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Experimental Demonstration of Radiation Effects on the Performance of a Stirling-Alternator Convertor and Candidate Mate...

Permanent Link: http://ufdc.ufl.edu/UFE0041517/00001

Material Information

Title: Experimental Demonstration of Radiation Effects on the Performance of a Stirling-Alternator Convertor and Candidate Materials Evaluation
Physical Description: 1 online resource (288 p.)
Language: english
Creator: Mireles, Omar
Publisher: University of Florida
Place of Publication: Gainesville, Fla.
Publication Date: 2010

Subjects

Subjects / Keywords: alternator, fission, materials, nuclear, performance, power, radiation, radioisotope, space, stirling
Nuclear and Radiological Engineering -- Dissertations, Academic -- UF
Genre: Nuclear Engineering Sciences thesis, Ph.D.
bibliography   ( marcgt )
theses   ( marcgt )
government publication (state, provincial, terriorial, dependent)   ( marcgt )
born-digital   ( sobekcm )
Electronic Thesis or Dissertation

Notes

Abstract: Free-piston Stirling power convertors are under consideration by NASA for service in the Advanced Stirling Radioisotope Generator (ASRG) and Fission Surface Power (FSP) systems to enable aggressive exploration missions by providing a reliable and constant power supply. The ASRG must withstand environmental radiation conditions, while the FSP system must tolerate a mixed neutron and ?-ray environment resulting from self-irradiation. Stirling-alternators utilize rare earth magnets and a variety of organic materials whose radiation limits dominate service life estimates and shielding requirements. The project objective was to demonstrate the performance of the alternator, identify materials that exhibit excessive radiation sensitivity, identify radiation tolerant substitutes, establish empirical dose limits, and demonstrate the feasibility of cost effective nuclear and radiation tests by selection of the appropriate personnel and test facilities as a function of hardware maturity. The Stirling Alternator Radiation Test Article (SARTA) was constructed from linear alternator components of a Stirling convertor and underwent significant pre-exposure characterization. The SARTA was operated at the Sandia National Laboratories Gamma Irradiation Facility to a dose of over 40 Mrad. Operating performance was within nominal variation, although modestly decreasing trends occurred in later runs as well as the detection of an electrical fault after the final exposure. Post-irradiation disassembly and internal inspection revealed minimal degradation of the majority of the organic components. Radiation testing of organic material coupons was conducted since the majority of the literature was inconsistent. These inconsistencies can be attributed to testing at environmental conditions vastly different than those Stirling-alternator organics will experience during operation. Samples were irradiated at the Texas A & M TRIGA reactor to above expected FSP neutron fluence. A thorough materials evaluation followed and results indicate that the majority of material properties experienced minimal statistically significant change.
General Note: In the series University of Florida Digital Collections.
General Note: Includes vita.
Bibliography: Includes bibliographical references.
Source of Description: Description based on online resource; title from PDF title page.
Source of Description: This bibliographic record is available under the Creative Commons CC0 public domain dedication. The University of Florida Libraries, as creator of this bibliographic record, has waived all rights to it worldwide under copyright law, including all related and neighboring rights, to the extent allowed by law.
Statement of Responsibility: by Omar Mireles.
Thesis: Thesis (Ph.D.)--University of Florida, 2010.
Local: Adviser: Dugan, Edward T.

Record Information

Source Institution: UFRGP
Rights Management: Applicable rights reserved.
Classification: lcc - LD1780 2010
System ID: UFE0041517:00001

Permanent Link: http://ufdc.ufl.edu/UFE0041517/00001

Material Information

Title: Experimental Demonstration of Radiation Effects on the Performance of a Stirling-Alternator Convertor and Candidate Materials Evaluation
Physical Description: 1 online resource (288 p.)
Language: english
Creator: Mireles, Omar
Publisher: University of Florida
Place of Publication: Gainesville, Fla.
Publication Date: 2010

Subjects

Subjects / Keywords: alternator, fission, materials, nuclear, performance, power, radiation, radioisotope, space, stirling
Nuclear and Radiological Engineering -- Dissertations, Academic -- UF
Genre: Nuclear Engineering Sciences thesis, Ph.D.
bibliography   ( marcgt )
theses   ( marcgt )
government publication (state, provincial, terriorial, dependent)   ( marcgt )
born-digital   ( sobekcm )
Electronic Thesis or Dissertation

Notes

Abstract: Free-piston Stirling power convertors are under consideration by NASA for service in the Advanced Stirling Radioisotope Generator (ASRG) and Fission Surface Power (FSP) systems to enable aggressive exploration missions by providing a reliable and constant power supply. The ASRG must withstand environmental radiation conditions, while the FSP system must tolerate a mixed neutron and ?-ray environment resulting from self-irradiation. Stirling-alternators utilize rare earth magnets and a variety of organic materials whose radiation limits dominate service life estimates and shielding requirements. The project objective was to demonstrate the performance of the alternator, identify materials that exhibit excessive radiation sensitivity, identify radiation tolerant substitutes, establish empirical dose limits, and demonstrate the feasibility of cost effective nuclear and radiation tests by selection of the appropriate personnel and test facilities as a function of hardware maturity. The Stirling Alternator Radiation Test Article (SARTA) was constructed from linear alternator components of a Stirling convertor and underwent significant pre-exposure characterization. The SARTA was operated at the Sandia National Laboratories Gamma Irradiation Facility to a dose of over 40 Mrad. Operating performance was within nominal variation, although modestly decreasing trends occurred in later runs as well as the detection of an electrical fault after the final exposure. Post-irradiation disassembly and internal inspection revealed minimal degradation of the majority of the organic components. Radiation testing of organic material coupons was conducted since the majority of the literature was inconsistent. These inconsistencies can be attributed to testing at environmental conditions vastly different than those Stirling-alternator organics will experience during operation. Samples were irradiated at the Texas A & M TRIGA reactor to above expected FSP neutron fluence. A thorough materials evaluation followed and results indicate that the majority of material properties experienced minimal statistically significant change.
General Note: In the series University of Florida Digital Collections.
General Note: Includes vita.
Bibliography: Includes bibliographical references.
Source of Description: Description based on online resource; title from PDF title page.
Source of Description: This bibliographic record is available under the Creative Commons CC0 public domain dedication. The University of Florida Libraries, as creator of this bibliographic record, has waived all rights to it worldwide under copyright law, including all related and neighboring rights, to the extent allowed by law.
Statement of Responsibility: by Omar Mireles.
Thesis: Thesis (Ph.D.)--University of Florida, 2010.
Local: Adviser: Dugan, Edward T.

Record Information

Source Institution: UFRGP
Rights Management: Applicable rights reserved.
Classification: lcc - LD1780 2010
System ID: UFE0041517:00001


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1 EXPERIMENTAL D EMONSTRATION OF RADIATION EFFECTS ON THE PERFORMANCE OF A STIRLING ALTERNATOR CONVERTOR AND CANDIDATE MATERIALS EVALUATION By OMAR R. MIRELES A DISSERTATION PRESENTED TO THE GRADUATE SCHOOL OF THE UNIVERSITY OF FLORIDA IN PARTIAL FULFILLMENT OF THE REQUIREMENTS FOR THE DEGREE OF DOCTOR OF PHILOSOPHY UNIVERSITY OF FLORIDA 2010

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2 2010 Omar R. Mireles

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3 To my mother Leticia, father Jose, and Step father Charlie for teaching me that with ambition, determination, effort and perseverance anything is possible

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4 ACKNOWLEDGMENTS I would like to thank Dr. Samim Anghaie, Dr. Edward Dugan, Dr. Valentin Craciun, Dr. William Vernetson Dr. James Baciak, Diana Dampier, Terri Sparks, Lynne Schreiber and Maxw ell Minch of the University of Florida. I would like to thank Dr. Cheryl Bowman, Dr. Eugene Shin, Lee Mason, Wayne Wong, Jeff Schreiber, Maxwell Briggs, Wayne Gerber Arianna Gunn, Salvitore Oriti Michael Brace, Peggy Cornell, Dan Scheiman, Robert Mattingly Christopher Blasio, Stacey Bagg and Terry McCue of NASA Glenn Research Center. I would like to thank Dr. Ross Rade l, Don Hason, James Andazola, and Jacqueline R. Tonigan of Sandia National Laboratories. I would like to thank Dr. Shannon Bragg Sitto n Dr. Latha Vasudevan, Jerry Newhouse, and Albert Tijerina of Texas A&M University I would also like to thank Dr. Dave Poston of Los Alamos National Laboratory and Dr. Lou Qualls of Oak Ridge National Laboratory. A special thanks goes to my family and friends for their encouragement and support throughout the years. A very special thanks goes to my long time advisor Dr. Michael Houts of NASA Marshall Space Flight Center for his guidance and incredible support of m y graduate research endeavors throughout the years, without whom it would not have been possible for me to achieve my masters degrees and a PhD. Funding for the author was provided by the NASA Graduate Student Research Program. Funding for the Stirling A lternator Radiation T est A rticle and for radiation testing at SNL was provided by the NASA Nuclear Power Radioisotope System Development program Funding for the materials test articles and radiation testing at Texas A&M University was provided by the NAS A Fiss ion Surface Power system project

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5 TABLE OF CONTENTS page ACKNOWLEDGMENTS .................................................................................................. 4 LIST OF FIGURES ........................................................................................................ 11 LIST OF ABBREVIA TIONS ........................................................................................... 21 ABSTRACT ................................................................................................................... 27 CHAPTER 1 BACKGROUND ...................................................................................................... 29 Introduction ............................................................................................................. 29 Radioisotope Power Systems ................................................................................. 29 Europa Radiation Environment ............................................................................... 31 Fission Surface Power Systems ............................................................................. 32 2 PROBLEM STATEMENT ........................................................................................ 41 Free Piston Stirling Power Convertor ...................................................................... 41 Objectives ............................................................................................................... 43 3 LITERATURE REVIEW AND HYPOTHESIS .......................................................... 46 Radiation Induced Damage .................................................................................... 46 Radiation Interaction with Organic Matter ............................................................... 47 Heavy Charged Particles .................................................................................. 48 Electrons and Positrons .................................................................................... 49 Particle Track Formation .................................................................................. 50 Photon .............................................................................................................. 50 Neutron ............................................................................................................. 51 Radiation Induced Chemical Effects in Organics .................................................... 53 Chemical Structure of Organics ........................................................................ 54 Ef fect of Crystallinity ......................................................................................... 54 Radiation Effects .............................................................................................. 55 Cross Linking ................................................................................................... 55 Scission ............................................................................................................ 56 Gas Production ................................................................................................. 56 Temperature Effect ........................................................................................... 57 Atmospheric Effect ........................................................................................... 58 Changes in Electrical Conductivity ................................................................... 58 Lubricant Degradation ...................................................................................... 59 Adhesives ......................................................................................................... 60 Additives ........................................................................................................... 61

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6 NdFeB and SmCo Rare Earth Permanent Magnets ............................................... 62 Time ................................................................................................................. 62 Thermal Effects ................................................................................................ 62 Shock, Stress, and Vibration ............................................................................ 63 Radiation Effects on Rare Earth Magnets ........................................................ 64 Photon .............................................................................................................. 64 Electron ............................................................................................................ 65 Neutron ............................................................................................................. 65 Proton ............................................................................................................... 66 Radiation Mitigation Techniques ...................................................................... 66 Design of Experiments: Accelerated Life Tests ...................................................... 67 Conclusions ............................................................................................................ 68 4 APPARATUS AND PROCEDURE: SARTA RADIATION TESTING ....................... 73 Stirling Alternator Component Radiation Testing .................................................... 73 Pre Exposure Characterization ............................................................................... 75 Thermal Response and Control .............................................................................. 76 Performance Mapping ............................................................................................. 76 Pressure cha nge sensitivity .................................................................................... 77 SARTA Radiation Testing ....................................................................................... 78 SARTA Dosimetry ............................................................................................ 79 SARTA Electrical Integrity ................................................................................ 80 SARTA Post Irradiation Performance Analysis ....................................................... 81 5 APPARATUS AND PROCEDURE: MIXED NEUTRON AND GAMMA RAY TESTING OF STIRLING ALTERNATOR CAND IDATE MATERIALS ..................... 87 Sample Irradiation Test Articles .............................................................................. 88 Mixed Neutron and Gammaray Testing Facility ..................................................... 88 Radiation Environment Characterization ................................................................. 89 Neutron Flux Spectrum Measurement .............................................................. 89 Dose Rate Measurement .................................................................................. 90 Pre Irradiation Test Article Preparation ................................................................... 90 Sample Irradiation ................................................................................................... 91 PostIrradiation Sample Processing ........................................................................ 92 6 APPARATUS AND PROCEDURE: POST IRRADIATION MATERIALS CHARACTERIZATION ......................................................................................... 104 Dimension and Weight Measurement ................................................................... 104 Optical Microscopy ............................................................................................... 104 Scanning Electron Microscopy & Energy Dispersive Spectroscopy ...................... 104 Differential Scanning Calorimetry ......................................................................... 105 Dynamic Mechanical Analysis .............................................................................. 106 Ther mo Gravimetric Analysis ................................................................................ 107 Fourier Transform Infrared Spectroscopy ............................................................. 108

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7 Surface Electrical Resistivity ................................................................................. 108 ORing Compression Set Testing ......................................................................... 109 Lap Shear Testing ................................................................................................ 109 7 RESULTS: SARTA RADIATION EXPOSURE OPERATION PERFORMANCE ... 114 SARTA Waveform Comparison ............................................................................ 118 SARTA Radiation Exposure Electrical Integrity .................................................... 119 SARTA Radiation Exposure RGA ......................................................................... 121 SARTA Radiation Exposure Leak Rate ................................................................ 122 Conclusions .......................................................................................................... 122 8 RESULTS: POST IRRADIATION EVALUATION OF THE SARTA ....................... 140 Electrical Integrity Measurements ......................................................................... 144 Changes in Pressure Leak Rate ........................................................................... 145 Conclusions .......................................................................................................... 1 45 9 RESULTS: SARTA POST IRRADIATION MATERIAL COUPON EVALUATION 153 Dimension and Weight Measurement ................................................................... 153 Optical Microscopy ............................................................................................... 153 Differential Scanning Calorimetry ......................................................................... 154 Thermo Gravimetric Analysis ................................................................................ 157 Conclusions .......................................................................................................... 159 10 RESULTS: MIXED NEUTRON & GAMMA RAY CANDIDATE MATERIAL EVALUATION ....................................................................................................... 169 Sample Transport, Control, and Preparation ........................................................ 169 Candidate Material Description ............................................................................. 169 Dimension Measurements .................................................................................... 170 Weight Measurement ............................................................................................ 171 Optical Microscopy ............................................................................................... 172 Scanning Electron Microscopy Energy Dispersive Spectroscopy ......................... 172 Fourier Transform Infrared Spectroscopy ............................................................. 173 Differential Scanning Calorimetry ......................................................................... 174 Thermo Gravimetric Analysis ................................................................................ 176 Dynamic Mechanic Analysis ................................................................................. 177 Electr ical Resistivity Measurement ....................................................................... 178 ORing Compression Set Tests ............................................................................ 178 Lap Shear Tensile Test ......................................................................................... 178 Conclusions .......................................................................................................... 180 11 RELEVANCE OF NUCL EAR AND RADIATION TESTS ....................................... 219 Nuclear and Radiation Testing Within A Research & Development Program ....... 219 Selection of an Appropriate Nuclear or Radiation Facility ..................................... 220

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8 Screening and Demonstration Tests .............................................................. 221 Flight Qualification and Acceptance Tests ...................................................... 222 Examples of Facility Selection and Hardware Maturity ................................... 223 Importance of Proper Facility Selection .......................................................... 223 Overcoming Political Constraints .......................................................................... 224 Utilization of Graduate Student Researchers ................................................. 224 Transition of Research Efforts ........................................................................ 225 Nuclear and/or Radiation Specific Test Factors to Consider ................................. 225 Activation of Experimental Equipment ............................................................ 226 PostIrradiation Disposal ................................................................................ 228 Conclusions .......................................................................................................... 229 12 CONCLUSIONS ................................................................................................... 231 SARTA Irradiation Testing and Evaluation ............................................................ 231 Mixed Neutron Ray Irradiation and Evaluation of Candidate Materials .............. 232 Relevance of Nuclear and Radiation Tests ........................................................... 234 Recommendations for Future Work ...................................................................... 235 Overall Findings .................................................................................................... 236 APPENDIX A SARTA TESTING ................................................................................................. 237 Supplemental Figures ........................................................................................... 237 Statistical Analysis of SARTA Confidence Intervals .............................................. 261 B MIXED NEUTRON GAMMA TESTING ................................................................. 262 C SARTA DISASSEMBLY AND INSPECTION ........................................................ 263 Analysis of Unknown Residue Deposit on SARTA Attachment Flange ................ 264 D MIXED NEUTRON & GAMMA RAY POST IRRADIATION MATERIALS EVALUATION ....................................................................................................... 265 Neutron Activation Analysis of Irradiated Samples ............................................... 265 Supplemental SEM EDS Data .............................................................................. 274 Electr ical Resistivity Measurement Data ............................................................... 278 ORing Compression Set Methodology ................................................................ 279 LIST OF REFERENCES ............................................................................................. 280 BIOGRAPHICAL SKETCH .......................................................................................... 287

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9 LIST OF TABLES Table page 1-1 Approximate isotopic composition of PuO2 GPHS fuel4 ...................................... 35 1-2 Radioisotope Power System Specifications1, 2, 5 ................................................. 35 3-1 Effects of radiation on lubricants24 ...................................................................... 72 3-2 NdFeB and SmCo Material Properties28, 37 ......................................................... 72 3-3 Neutron Irradiation Effects on NdFeB and SmCo Permanent Magnets .............. 72 4-1 SARTA Nominal Operation Limits ...................................................................... 84 4-2 Performance as a function of pressure ............................................................... 84 4-3 Pre exposure electrical values ............................................................................ 86 5-1 TAMU irradiation cell neutron spectrum measurements at 11 cm ...................... 99 5-2 TAMU irradiation cell neutron flux measurements as a function of distance. ..... 99 5-3 TAMU irradiation cell gamma ray dose rate measurements. .............................. 99 5-4 Stirling Alternator Candidate Organic Materials ................................................ 100 5-5 Test Conditions ................................................................................................. 100 7-2 SARTA leak rate vs. dose ................................................................................ 139 7-3 SARTA Operational Totals ............................................................................... 139 8-1 SARTA electrical integrity comparisons ............................................................ 152 10-1 Organic Materials Weight Measurement (Pre and Post Irradiation) ................. 186 11-1 Comparison of Select Radiation Test Facilities ................................................ 230 C-1 SARTA Organic Material Samples ................................................................... 263 D-1 Nuke 4 Silicone O ring Gamma Ray Spectroscopic Elemental Abundance Analysis ............................................................................................................ 265 D-2 Nuke 4 Kalrez O ring Gamma Ray Spectroscopic Elemental Abundance Analysis ............................................................................................................ 266 D-3 Nuke 4 Viton Heat Shrink GammaRay Spectroscopic Elemental Abundance Analysis ............................................................................................................ 267

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10 D-4 Nuke 4 Kynar Heat Shrink GammaRay Spectroscopic Elemental Abundance Analysis ............................................................................................................ 267 D-5 Nuke 4 Polyimide Insulated Cu Wire GammaRay Spectroscopic Elemental Abundance Analysis ........................................................................................ 268 D-6 Nuke 4 PTFE Insulated Cu Wire GammaRay Spectroscopic Elemental Abundance Analysis ......................................................................................... 269 D-7 Nuke 4 Xylan Coated Al Plate GammaRay Spectroscopic Elemental Abundance Analysis ......................................................................................... 270 D-8 TAMU Background GammaRay Spectroscopic Elemental Abundance Analysis. .......................................................................................................... 271 D-9 Comparative Sample and TAMU Background Spectroscopic Analysis. ........... 272 D10 Qualitative electrical resistivity measurements of Viton and Kynar heat shrink tubing, and PTFE and Polyimide wire insulation. .............................................. 278

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11 LIST OF FIGURES Figure page 1-1 Exploded View Schematic of the General Purpose Heat Source (GPHS) module3. ............................................................................................................. 35 1-2 Concept design schematic of the Advanced Stirling Radioisotope Generator (ASRG)6. ............................................................................................................. 36 1-3 90day Europa radiation environment based on Divine/GIRE model9. ............... 37 1-4 90day Europa orbiter expected dose based on Divine/GIRE model9. ............... 37 1-5 Fission Surface Power (FSP) general reactor system schematic11. ................... 38 1-6 Deployed emplaced Fission Surface Power (FSP) reactor concept configuration ...................................................................................................... 38 1-7 Estimated Fission Surface Power (FSP) gammaray dose along the axial centerline above the reactor core ...................................................................... 39 1-8 MCNP model of the Fission Surface Power (FSP) buried configuration neutron spectrum at the Stirling power convertors ............................................. 40 2-1 Free piston Stirling Convertor with linear alternator. Courtesy of Sunpower20. ... 45 3-1 Expected electron interaction with energy & atomic mass dependance25. ......... 69 3-2 Conceptual example of the molecular chain unit cell (crystallinity)28. ................. 69 3-3 Influence of crystallinity and molecular weight on physical properties28. ............ 70 3-4 Representation of A) linear branches crosslinked branches C) 3D highly cross linked branch network28. ........................................................................... 70 3-5 Relaxation modulus vs. Temperature for amorphous polystyrene28. .................. 71 3-6 Temperature effect on crystalline, lightly cross linked and amorphous polymers28. ......................................................................................................... 71 4-1 Stirling Alternator Radiation Test Article (SARTA) before radiat ion testing. ....... 82 4-2 SARTA support, control, instrumentation, and data acquisition equipment. ....... 82 4-3 Thermal time constant response runs (steady state, fan turned on, no heat, no insulation). ..................................................................................................... 83 4-4 Stoke variation performance mapping (90C, 500 psig, 6.5 8.5 mm). ............. 83

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12 4-5 Sandia National Laboratories Gamma Irradiation Facility cell #2 155.6 kCi 60Co source array. .............................................................................................. 84 4-6 Low dose rate configuration and h igh dose rate configuration. .......................... 85 4-7 RAD 1 dosimeter placeme nt and subsequent dosimeter placement. ................. 85 4-8 Sencore LC53 Z Meter Capacitor Inductor Analyzer. ...................................... 86 4-9 Fluke 1520 MegaOhm Meter a nd Fluke 189 Multi Meter. .................................. 86 5-1 Mixed neutron and gammaray material sample test article. .............................. 93 5-2 Material sample tray ........................................................................................... 94 5-3 Material sample tray position in the test article. .................................................. 94 5-4 Instrumentation and control rack used for data collection and test article temperature environment control. ....................................................................... 95 5-5 TAMU TRIGA Mark I reactor coupled to the irradiation cell during operation. .... 95 5-6 TAMU reactor coupled to irradiation cell schematic ............................................ 96 5-7 Gold foil, cadmium covered gold foil and iron wires for neutron spectrum characterization. ................................................................................................. 96 5-8 TLD 400 (CaF2) and radiachromic film dosimeters for dose rate measurement ...................................................................................................... 97 5-9 Dosimeter and foil/wire placement in the irradiation cell window ........................ 97 510 Dosimeter, gold foil, cadmium covered gold foil, and iron wire arrangement. ..... 98 511 Fully instrumented test articles and dosimeters in the TAMU irradiation cell window. .............................................................................................................. 98 512 Nuke 1 and Nuke 2 temperature profiles with respect to reactor power. .......... 101 513 Nuke 3 and Nuke 4 temperature profiles with respect to reactor power. .......... 101 514 Nuke 5 and Nuke 6 temperature profiles with respect to reactor power. .......... 102 515 Nuke 7 and Nuke 8 temperature profiles with respect to reactor power. .......... 102 516 Postirradiation sample removal from test articles in portable glove bag back filled with argon. ................................................................................................ 103 6-1 NASA GRC Keyence Digital Optical Microscope and image acquisition unit. .. 110

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13 6-2 NASA GRC Hitachi S 4700 Field Emission Scanning Electron Microscope. .... 110 6-3 NASA GRC TA Instruments Q 1000 Differential Scanning Calorimeter. .......... 111 6-4 NASA GRC TA Instruments 2980 Dynamic Mechanical Analyzer. ................... 111 6-5 NASA GRC TA Instruments Q 500 ThermoGravimetric Analyzer. .................. 112 6-6 NASA GRC Thermo Electron Nicolet 380 FTIR instrument and data acquisition system ........................................................................................... 112 6-7 Oring compression set apparatus ................................................................... 113 6-8 NASA GRC Instru Met Instron Uni Axial Load Frame used for lap shear testing ............................................................................................................... 113 7-1 Typical SARTA radiation test profile (power, current, voltage, and stroke). ...... 125 7-2 Typical SARTA radiation test profile (temperature and pressure). .................... 125 7-3 Lead sheet and blocks used to shield the SARTA pressure transducer. .......... 126 7-4 Influence of v arying operating pressure on SARTA performance. .................... 126 7-5 SARTA power factor comparison for 90C operating temperature runs. .......... 127 7-6 SARTA stroke vs. time for all tests conducted at 90C operating temperature. 128 7-7 SARTA power factor comparison for tests conducted at 125C operating temperature. ..................................................................................................... 129 7-8 SARTA stroke vs. time for runs at 125C operating temperature. .................... 130 7-9 End of SARTA RAD 18 exposure run voltage and current decrease. ............... 131 710 End of SARTA RAD 18 exposure run power and stroke decrease. .................. 131 711 SARTA pre exposure oscilloscope waveforms at 90C and 125C. ................. 132 712 RAD 12 waveform (90C, 80 minute exposure, 853.57 rad/s, 4.097 Mrad). ..... 132 713 RAD 16 waveform (125C, 40 min. exposure, 853.57 rad/s, 2.049 Mrad). ....... 133 714 SARTA pre exposure and post exposure waveform comparison at 90C. ....... 133 715 SARTA pre exposure and post exposure waveform comparison at 125C. ..... 134 716 SARTA FLDT electrical integrity as a function of increasing dose. ................... 135

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14 717 SARTA outer stator electrical integrity as a function of increasing dose. .......... 135 718 Close up view of the SARTA outer stator inductance as a function of dose. .... 136 719 Pre exposure RGA scan of the SARTA internal helium working fluid. .............. 136 720 RAD 1 RGA scan (90C, 30 minute exposure, 81.82 rad/s, 0.147 Mrad). ........ 137 721 RAD 13 RGA scan (90C, 195 minutes, 853.57 rad/s, 9.987 Mrad). ................ 137 722 RAD 16 RGA sc an (125C, 40 minutes, 853.57 rad/s, 2.049 Mrad). ................ 138 723 RAD 18 RGA scan (125C, 200 minutes, 853.57 rad/s, 10.243 Mrad). ............ 138 724 SARTA post RAD 1 exposure helium leak rate. ............................................... 139 8-1 NASA GRC SARTA disassembly glove box with argon backfill. ....................... 147 8-2 RGA scan of the glove box atmosphere prior to SARTA disassembly. ............ 147 8-3 SARTA internal atmosphere RGA scan prior to disassembly. .......................... 148 8-4 SARTA pressure vessel Silicone O ring with significant compression set. ....... 148 8-5 SARTA FLDT Viton h eat shrink tubing at the transition plate terminals. ........... 149 8-6 Crack along one of the outer stator PTFE insulated wires near the transition plate terminals. ................................................................................................. 149 8-7 Kynar heat shrink tubing embrittlement at the outer stator wire terminals. ....... 149 8-8 Outer stator section comparison of Control vs. SARTA. ................................... 149 8-9 Outer stator Hysol epoxy section comparison of Control vs. SARTA. .............. 150 810 Outer stator Nomex section comparison of Control vs. SARTA. ....................... 150 811 Front end inner stator section comparison of Control vs. SARTA. .................... 150 812 Back en d inner stator section comparison of Control vs. SARTA. .................... 151 813 Residue deposit on the inner transition plate surface. ...................................... 151 814 Rub marks on the SARTA piston outer running surface Xylan coating. ............ 151 815 SARTA Inner stator rub marks corresponding to piston rub marks. .................. 152 816 External PTFE wire insulation post irradiation aging embrittlement. ................. 152

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15 9-1 Degradation of PTFE alternator wire insulation near the transition plate terminals ........................................................................................................... 160 9-2 Micrograph of the coil PTFE hook up wire insulation fracture pattern (150x). .. 160 9-3 Micrograph of coil PTFE hook up wire insulation fracture pattern adjacent to the previous image (150x). ............................................................................... 161 9-4 3D composite micrograph of the PTFE hook up wire insulation spiral crack (100x). .............................................................................................................. 161 9-5 Angled 3D composite image of the PTFE hook up wire insulation spiral crack (100x). .............................................................................................................. 162 9-6 3D composite image of the PTFE hook up wire insulation spiral crack (100x). 162 9-7 DSC curve comparison of Kynar heat shrink tubing. ........................................ 163 9-8 DSC curve comparison of Viton heat shrink tubing. ......................................... 163 9-9 DSC curve comparison of Viton O ring ............................................................. 164 910 DSC curve comparison of Silicone O ring. ....................................................... 164 911 DSC curve comparison of PTFE hook up wire insulation. ................................ 165 912 DSC curve comparison of Xylan coating. ......................................................... 165 913 TGA weight curve comparison of Kynar heat shrink tubing. ............................. 166 914 TGA weight curve comparison of Viton heat shrink tubing. .............................. 166 915 TGA weight curve comparison of Viton O ring. ................................................ 1 67 916 TGA weight curve comparison of Silicone O -r ing. ............................................ 167 917 TGA weight curve comparison of PTFE hook up wire insulation. ..................... 168 918 TGA curve comparison of Xylan coating. ......................................................... 168 10-1 NASA GRC glove box for storage and processing of radioactive samples. ...... 182 10-2 Xylan coating dimension change as a function of condition. ............................ 182 10-3 Silicone O ring dimension change as a function of condition. .......................... 183 10-4 Kalrez Oring dimension change as a function of condition. ............................. 183 10-5 Kynar heat shrink dimension change as a function of condition. ...................... 184

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16 10-6 Viton heat shrink dimension change as a function of condition. ....................... 184 10-7 PTFE wire dimension change as a function of condition. ................................. 185 10-8 Polyimide wire dimension change as a function of condition. ........................... 185 10-9 Optical micrograph comparison at 30x of Silicone O ring ................................. 187 1010 Optical micrograph comparison at 30x of Kalrez O ring .................................. 187 1011 Optical micrograph comparison at 22.5x of Kynar heat shrink ......................... 188 1012 Optical micrograph comparison at 22.5x of Viton heat shrink .......................... 188 1013 Optical micrograph comparison at 37.5x of PTFE wire ..................................... 189 1014 Optical micrograph comparison at 22.5x of Polyimide wire .............................. 189 1015 Optical micrograph comparison at 37.5x of Xylan coated aluminum ................ 190 1016 Comparison of Xylan SEM SE images BOA 1 at 300x, NUKE 1 at 300x, BOA 1 at 1000x, and NUKE 1 at 1000x. ................................................................... 191 1017 Comparison of Xylan SEM SE images BOA 2 at 300x, NUKE 2 at 300x, BOA 2 at 1000x, and NUKE 2 at 1000x. ................................................................... 192 1018 Comparison of Xylan SEM SE images BOA 3 at 300x, NUKE 3 at 300x, BOA 3 at 1000x, and NUKE 3 at 1000x. ................................................................... 193 1019 Comparison of Xylan SEM SE images BOA 4 at 300x, NUKE 4 at 300x, BOA 4 at 1000x, and NUKE 4 at 1000x. ................................................................... 194 1020 FTIR spectra comparison for Silicone bakedout aged (BOA) vs. Nuke 1 and 2, Nuke 3 and 4. ............................................................................................... 195 1021 FTIR spectra comparison for Kalrez bakedout aged (BOA) controls vs. NUKE 1 and 2, NUKE 3 and 4 irradiated samples. ........................................... 196 1022 FTIR spectra comparison for Kynar bakedout aged (BOA) controls vs. NUKE 1 and 2, NUKE 3 and 4 irradiated samples. ........................................... 197 1023 FTIR spectra comparison for Viton bakedout aged (BOA) controls vs. NUKE 1 and 2, NUKE 3 and 4 irradiated samples. ........................................... 198 1024 FTIR spectra comparison for PTFE bakedout aged (BOA) controls vs. NUKE 1 and 2, NUKE 3 and 4 irradiated samples. ........................................... 199 1025 FTIR spectra comparison for Polyimide baked out aged (BOA) controls vs. NUKE 1 and 2, NUKE 3 and 4 irradiated samples. ........................................... 200

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17 1026 FT IR spectra comparison for Xylan bakedout aged (BOA) controls vs. NUKE 1 and 2, NUKE 3 and 4 irradiated samples. ........................................... 201 1027 Co mparison of DSC curves for Silicone irradiated at 125C (NUKE 1 & 3) and 50C (NUKE 2 & 4). ................................................................................... 202 1028 Comparison of DSC curves for Kalrez irradiated at 125C (NUKE 1 & 3) and 150C (NUKE 2 & 4). ........................................................................................ 203 1029 Comparison of DSC curves for Kynar irradiated at 125C (NUKE 1 & 3) and 150C (NUKE 2 & 4). ........................................................................................ 204 1030 Comparison of DSC curves for Viton irradiated at 125C (NUKE 1 & 3) and 150C (NUKE 2 & 4). ........................................................................................ 205 1031 Comparison of DSC curves for PTFE irradiated at 125C (NUKE 1 & 3) and 150C (NUKE 2 & 4). ........................................................................................ 206 1032 Comparison of DSC curves for Polyimide irradiated at 125C (NUKE 1 & 3) and 150C (NUKE 2 & 4). ................................................................................. 207 1033 Comparison of TGA curves for Silicone irradiated at 125C (NUKE 1 & 3) and 150C (NUKE 2 & 4). ................................................................................. 208 1034 Comparison of TGA curves for Kalrez irradiated at 125C (NUKE 1 & 3) and 150C (NUKE 2 & 4). ........................................................................................ 209 1035 Comparison of TGA curves for Kynar irradiated at 125C (NUKE 1 & 3) and 150C (NUKE 2 & 4). ........................................................................................ 210 1036 Comparison of TGA curves for Viton irradiated at 125C (NUKE 1 & 3) and 150C (NUKE 2 & 4). ........................................................................................ 211 1037 Comparison of TGA curves for PTFE irradiated at 125C (NUKE 1 & 3) and 150C (NUKE 2 & 4). ........................................................................................ 212 1038 Comparison of TGA curves for Polyimide irradiated at 125C (NUKE 1 & 3) a nd 150C (NUKE 2 & 4). ................................................................................. 213 1039 Comparison of TGA curves for Xylan irradiated at 125C (NUKE 1 & 3). ......... 214 1040 Comparison of PTFE DMA curves of bakedout aged (BAO) versus irradiated (N) for samples irradiated at 125C(NUKE 1 and 3), and 150C (NUKE 2 and 4). .............................................................................................................. 215 1042 Bond strength of Hysol and High Temperature Hysol epoxies lap shear samples tested at 120C. ................................................................................. 217

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18 1043 Toughness of Hysol and High Temperature Hysol epoxies lap shear samples tested at 120C. ................................................................................. 218 A-1 SARTA system instrumentation, power, and control schematic. ...................... 237 A-2 The Sandia National Laboratory Gamma Irradiation Facility (GIF) radial dose rate profile. ........................................................................................................ 238 A-3 GIF axial dose rate profile. Courtesy of Dr. Ross Radel, SNL49. ..................... 238 A-4 RAD 1 Performance (90C, 30 minutes, 81.82 rad/s, 0.147 Mrad). .................. 239 A-5 RAD 1 Temperatures (90C, 30 minutes, 81.82 rad/s, 0.147 Mrad). ................ 239 A-6 RAD 2 Performance (90C, 30 minutes, 75.94 rad/s, 0.137 Mrad). .................. 240 A-7 RAD 2 Temperatures (90C, 30 minutes, 75.94 rad/s, 0.137 Mrad). ................ 240 A-8 RAD 3 Performance (90C, 60 minutes, 98.74 rad/s, 0.335 Mrad). .................. 241 A-9 RAD 3 Temperatures (90C, 60 minutes, 98.74 rad/s, 0.335 Mrad). ................ 241 A10 RAD 4 Performance (90C, 60 minutes, 94.38 rad/s, 0.340 Mrad). .................. 242 A11 RAD 4 Temperatures (90C, 60 minutes, 94.38 rad/s, 0.340 Mr ad). ................ 242 A12 RAD 5 Performance (90C, 210 minutes, 78.88 rad/s, 0.994 Mrad). ................ 243 A13 RAD 5 Temperatures (90C, 210 minutes, 78.88 rad/s, 0.994 Mrad). .............. 243 A14 RAD 6 Performance (90C, 160 minutes, 78.88 rad/s, 0.757 Mrad). ................ 244 A15 RAD 6 Temperatures (90C, 160 minutes, 78.88 rad/s, 0.757 Mrad). .............. 244 A16 RAD 7 Performance (90C, 270 minutes, 78.88 rad/s, 1.278 Mrad). ................ 245 A17 RAD 7 Temperatures (90C, 270 minutes, 78.88 rad/s, 1.278 Mrad). .............. 245 A18 RAD 8 Performance (90C, 5 minutes, 859.39 rad/s, 0.258 Mrad). .................. 246 A19 RAD 8 Temperatures (90C, 5 minutes, 859.39 rad/s, 0.258 Mrad). ................ 246 A20 RAD 9 Performance (90C, 5 minutes, 847.75 rad/s, 0.254 Mrad). .................. 247 A21 RAD 9 Temperatures (90C, 5 minutes, 847.75 rad/s, 0.254 Mrad). ................ 247 A22 RAD 10 Performance (90C, 35 minutes, 853.57 rad/s, 1.793 Mrad). .............. 248 A23 RAD 10 Temperat ures (90C, 35 minutes, 853.57 rad/s, 1.793 Mrad). ............ 248

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19 A24 RAD 11 Performance (90C, 40 minutes, 853.57 rad/s, 2.049 Mrad). .............. 249 A25 RAD 11 Temperatures (90C, 40 minutes, 853.57 rad/s, 2.049 Mrad). ............ 249 A26 RAD 12 Performance (90C, 80 minutes, 853.57 rad/s, 4.097 Mrad). .............. 250 A27 RAD 12 Temperatures (90C, 80 minutes, 853.57 rad/s, 4.097 Mrad). ............ 250 A28 RAD 13 Performance (90C, 195 minutes, 853.57 rad/s, 9.987 Mrad). ............ 251 A29 RAD 13 Temperatures (90C, 195 minutes, 853.57 rad/s, 9.987 Mrad). .......... 251 A30 RAD 14 Performance (120C, 60 minutes, 78.88 rad/s, 0.284 Mrad). .............. 252 A31 RAD 14 Temperatures (125C, 60 minutes, 78.88 rad/s, 0.284 Mrad). ............ 252 A32 RAD 15 Performance (125C, 20 minutes, 853.57 rad/s, 1.024 Mrad). ............ 253 A33 RAD 15 Temperatures (125C, 20 minutes, 853.57 rad/s, 1.024 Mrad). .......... 253 A34 RAD 16 Performance (125C, 40 minutes, 853.57 rad/s, 2.049 Mrad). ............ 254 A35 RAD 16 Temperatures (125C, 40 minutes, 853.57 rad/s, 2.049 Mrad). .......... 254 A36 RAD 17 Performance (125C, 80 minutes, 853.57 rad/s, 4.097 Mrad). ............ 255 A37 RAD 17 Temp eratures (125C, 80 minutes, 853.57 rad/s, 4.097 Mrad). .......... 255 A38 RAD 18 Performance (125C, 200 minutes, 853.57 rad/s, 10.243 Mrad). ........ 256 A39 RAD 18 Temperatures (125C, 200 minutes, 853.57 rad/s, 10.243 Mrad). ...... 256 A40 SARTA voltage vs. time results for operation at 90C. ..................................... 257 A41 SARTA current vs. time for operation at 90C. ................................................. 257 A42 SARTA power vs. time results for operation at 90C tests. .............................. 258 A43 SARTA voltage vs. time results for operation at 125C. ................................... 258 A44 SARTA current vs. time results for operation at 125C. .................................... 259 A45 SARTA power vs. time results for operation at 125C. ..................................... 259 A46 SARTA average power factor as a function of dose for 90C tests. ................. 260 A47 SARTA average power factor as a function of dose for 125C tests. ............... 260 B-1 Mixed neutrongamma ray test system instrumentation, power, and control schematic. ........................................................................................................ 262

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20 C-1 FTIR spectra comparison of the unknown residue deposit and Hysol epoxy reference. ......................................................................................................... 264 D-1 Comparison of unirradiated Xylan SEM imag es secondary electron (SE) and BackScattered Electron (BSE) images at 500x. ....................................... 274 D-2 Comparison of unirradiated Xylan SEM images secondary electron (SE) and BackScattered Electron (BSE) images at 10,000x. .................................. 275 D-3 Comparison of irradiated (NUKE 4) Xylan SEM images secondary electron (SE) image at 5,000x. ....................................................................................... 275 D-4 Comparison of Xylan EDS spectra referencing Figure D 2B callout lo cations .. 276 D-5 Comparison of Xylan EDS spectra referencing Figure D 3 callout locations .... 277

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21 LIST OF ABBREVIATION S A Amps ACRR Annular Core Research Reactor ALT Alternator AR As Received ASC Advanced Stirling Convertor ASRG Advanced Stirling Radioisotope Generator ASTM American Society of Testing and Materials A(t) Activity ATR Advanced Test Reactor BHmax Energy Product BO BakedOut BOA BakedOut Aged BOM Beginning of Mission Br Remanence BNL Brookhaven National Laboratory BSE BackScattered Electron c Speed of light in a vacuum CB Compression Set Co Concentration of oxygen C Degrees Celsius cm2 Square Centimeter CSNR Center for Space Nuclear Research DEst Estimated Dose DMeas Measured Dose

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22 DTotal Total Dose D Gamma ray Dose Dose Rate Gamma ray Dose Rate D ia Diameter DSC Differential Scanning Calorimetry DMA Dynamic Mechanical Analysis E Energy Eb Binding energy e Electron e Magnitude of electron charge EDS Energy Dispersive Spectroscopy EPD Energetic Particle Detector eV Electron Volt FLDT Fast Linear Displacement Transducer FSP Fission Surface Power ft Feet GIF Gamma Irradiation Facility GIRE Galileo Interim Radiation Environment Model GPHS General Purpose Heat Source GRC NASA Glenn Research Center I M ean excitation energy of the medium INL Idaho National Laboratory h Hour Hc Coercivity

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23 HFEF Hot Fuel Examination Facility HFIR High Flux Isotope Reactor HIC Heavy Ion Counter HPGe High Purity Germanium hv Incident photon energy ko 8.99 x 109 N-M2/C2 kCi Kilo Curie krad Kilo Radiation Absorbed Dose kWe Kilo Watts Electric LANL Los Alamos National Laboratory LET Linear Energy Transfer LLW Low Level Waste m Neutron mass M Nuclide mass mo Electron rest mass energy MFC Materials and Fuels Complex MCNP Monte Carlo Neutral Particle code MeV Mega Electron Volt mg Milligram mH MilliHenry mm Millimeter MMRTG Multi Mission Radioisotope Thermoelectric Generator MPa Mega -Pascal Mrad Mega Radiation Absorbed Dose mR/hr MilliRoentgen per hour

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24 ms Millisecond MSFC NASA Marshall Space Flight Center MSL Mars Science Laboratory Msystem System Mass MWth Mega Watt thermal m/Z Mass To Charge Ratio n Neutron n Electron volumetric density NASA National Aeronautics and Space Administration NdFeB Neodymium Iron Boron NSC Nuclear Science Center NSUF National Scientific User Facility NSRL NASA Space Radiation Laboratory ORNL Oak Ridge National Laboratory OSU Ohio State University p Proton PCS Power Conversion Subsystem PID Proportional Integral -D erivative Ps Specific Power psia Poun ds per square inch (absolute) psig Pounds per square inch (gauge) PTFE P olytetrafluroethylene (Teflon trade name) PVC Polyvinyl chloride QMax Maximum neutron energy loss RAD Radiation Absorbed Dose

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25 RGA Residual Gas Analysis RDM Radiation Design Margin RPS Radioisotope Power System RTG Radioisotope Thermoelectric Generator SARTA Stirling Alternator Radiation Test Article SE Secondary Electron SEM Scanning Electron Microscopy SmCo Samarium Cobalt SNL Sandia National Laboratories TRL Test Readiness Level TAMU Texas A&M University Tc Curie Temperature Td Degradation Onset Temperature TDC Technology Demonstration Convertor TE Thermo Electric tf final specimen thickness Tg Glass Transition Temperature TGA Thermo Gravimetric Analysis TLD Thermo Luminescent Dosimeter Tmax Maximum Operating Temperature to O riginal specimen thickness tshim S him thickness UHP Ultra High Purity UHV Ultra high vacuum V Volts

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26 V Volume fraction crystallinity v Velocity of an incident particle V AC Volts AC W Watts We Watts Electric Wth Watts Thermal wt% Weight Percent XRD Xray Diffraction Z A tomic number S peed of an incident particle relative to c -de/dx Energy loss per unit path length (stopping power) P Change in pressure t Change in time ray Gamma ray Neutron flux ( n/cm2-s) t Neutron f luence ( n /cm2) Ohms Efficiency Density (g/cm3) c C rystal phase density (g/cm3) Photon scattering angle Micro Henry Micro Curie D egree of crystallinity

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27 Abstract of Dissertation Presented to the Graduate School of the University of Florida in Partial Fulfillment of the Requirements for the Degree of Doctor of Philosophy EXPERIMENTAL DEMONSTRATION OF RADIATION EFFECTS ON THE PERFORMANCE OF A ST IRLING ALTERNATOR CONVERTOR AND CANDIDATE MATERIALS EVALUATION By Omar R. Mireles May 2010 Chair: Edward Dugan Major: Nuclear Engineering Sciences Free piston Stirling power convertors are under consideration by NASA for service in the Advanced Stirling Radioisotope Generator (ASRG) and Fission Surface Power (FSP) systems to enable aggressive exploration missions by providing a reliable and constant power supply The ASRG must withstand environmental radiation conditions while the FSP system must tolerate a mixed neutron and ray environment resulting from self irradiation. Stirling alternators utilize rare earth magnets and a variety of organic materials whose radiation limits dominate service life estimates and shielding requirements The project objective was to demonstrate the performance of the alternator, identify materials that exhibit excessive radiation sensitivity, identify radiation tolerant substitutes, establish empirical dose limits, and demonstrate the feasibility of cost effective nuclear and radiation tests by selection of the appropr iate personnel and test facilities as a function of hardware maturity. The Stirling Alternator Radiation Test Article (SARTA) was constructed from linear alternator components of a Stirling convertor and underwent significant preexposure characterization. The SARTA was operated at the Sandia National Laboratories

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28 Gamma Irradiation Facility to a dose of over 40 Mrad. Operating performance was within nominal variation, although modestly decreasing trends occurred in later runs as well as the detection of a n electrical fault after the final exposure. Post irradiation disassembly and internal inspection revealed minimal degradation of the majority of the organic components. Radiation testing of organic material coupons was conducted since the majority of the literature was inconsistent. These inconsistencies can be attributed to testing at environmental conditions vastly different than those Stirling alternator organics will experience during operation. Samples were irradiated at the Texas A&M TRIGA reactor to above expected FSP neutron fluence. A thorough materials evaluation followed and r esults indicate that the majority of material properties experienced minimal statistically significant change.

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29 CHAPTER 1 BACKGROUND Introduction Novel technologies for power production and next generation space propulsion systems are critically needed in order to enable aggressive, long duration, space e xploration missions Long duration, deep space missions require compact energy generation systems capable of sustaining a reliable energy supply with little fuel and no maintenance. Radioactive decay and nuclear fission are excellent candidates for these missions b ecause they possess many of these favorable characteristics1. Examples include the r adioisotope power systems (RPS) and Fission Surface Power (FSP). Radioisotope Power Systems The traditional application of nuclear power in space applications (both flight and surface operations) has been accomplished through the utilization of the Radioisotope Thermoelectric Generator (RTG). Several models of RTGs have enjoyed an impressive success rate of 26 missions flying 45 RTGs over four decades with no failures2. T his success rate can be attributed to a simplified design that has no moving components. All modern RTG models, both in service and under development, utilize the G eneral Purpose Heat Source (GPHS) design as the basic fuel module. A GPHS module utilizes four PuO2 fuel pellets which are encased in iridium clad. The iridium is then coated with layers of graphite t o form an impact shell. The impact shell is encased within a graphite housing aeroshell which will allow for the GPHS to survive atmospheric re entry in the event of a launch accident or deorbit. A standard GPHS module has size dimensions of 9.32 x 9.72 x 5.30 cm, a maximum weight of 1.44 kg, and nominally produces 250 Wth power at beginning of mission (BOM)3 as shown in Figure 1-1.

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30 The long hal flife of the fuel 238Pu (t1/ 2 = 87.74 years) allows the GPHS module to provide a reliable supply of energy that will endure throughout the mission lifetime. The isotopic composition of the PuO2 at fabrication is shown in Table1-1. The radiation field outside the GPHS is relatively benign due to a low emission rate. The GPHS primarily emits neutrons and rays. Neutron emission comes from 17,18 spontaneous fission, and neutron induced fission. The decay of 238Pu is responsible for 99concern, the secondary radiation it produces does pose a problem. Naturally occurring oxygen is comprised of three isotopes : 16O, 17O, and 18particles emitted by the decay of 238Pu can induce 17O and 184. Therefore, to minimize the neutron emission rate from the GPHS module the fuel has been depleted of 17O and 18O. It was determined that the PuO2 in a GPHS has a specific neutron e mission rate of ~5 x 103 n/s g of 238Pu and that t he ray emission is a consequence of fission and radiative capture within the fuel4. Several GPHS based radioisotope power systems are currently under consideration for future space applications5. The primary concern over the utilization of RPS is that the U.S. domestic stockpile of 238Pu has become depleted and p lans to re start production will take several years. In addition, Russia will no longer sell the radionuclide to the U.S. due to the fact that Russias own stockpile has been depleted. The current RTG model known as the GPHS RTG requires large quantities of fuel and was originally developed to power larger robotic spacecraft than the missions envisioned for the next few decades. Therefore, a significant effort has been underway to develop an RPS that will require less fuel t o deliver mission power demands.

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31 Examples include the Multi Mission Radioisotope Thermoelectric Generator (MMRTG) and the Advanced Stirling Radioisotope Generator (ASRG). A comparison of the se three radioisotope systems is displayed in Table 1-2. The MMRTG is simply a GPHS RTG with fewer GPHS modules ; however, the A SRG uses significantly less GPHS modules and produces comparable power and therefore, this concept must be explored further The conceptual ASRG flight system is illustrated in Figure 1 -2. Although s elf irradiation from a GPHS module is not significant RPS technologies must be capable of operating in a high radiation environment A worst case radiation scenar io would be a Jovian mission, such as a Europa Orbiter and would be subjected to high energy electron and proton belts resulting from the Jovian magnetosphere. Such environments dictate high radiation tolerance requirements be established. Europa R adiation E nvironment Missions to Jupiter can be difficult due to the high radiation environment, especially within the orbit of Ganymede where radiation belt fluxes increase by two orders of magnitude. A spacecraft in a circular 100 km orbital trajectory above Europa (R = 9.4 RJ) w ould experience a flux dominated by electrons (hundreds of keV to tens of MeV), protons and heavy ions (tens to hundreds of MeV) as measured by the Galileo Energetic Particles Detector (EPD) and Heavy Ion Counter (HIC)7. These and other mission data were incorporated into the Divine model supplemented with the Galileo Interim Radiation Environment (GIRE) model7. Fortunately, models indicate that a spacecraft in Europa orbit will experience much lower fluxes than the surrounding space, despite Europa having no permanent magnetic field. Europas presence acts as 2 shield that will attenuate radiation levels. Therefore, an orbiting spacecraft will receive flux from only one field line direction at a time8. Missi on studies establ ished a

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32 mission duration of 90 days, which would be the minimum service life requirement for an RPS system From this timeline we estimate the radiation dose to the spacecraft. GIRE models suggest that a Europa orbiter spacecraft will receive a dose of 4. 3 Mrad (Si), assuming a 2.54 mm aluminum pressure vessel as shown in Figure 13 and Figure 1-4. Flight units will most likely be constructed of titanium with wall thickness on the order of 12 mm which should provide additional shielding 10. This dose estimate includes transit and a 90 day mission in Europa orbit. Fission Surface Power Systems NASA has been redirected towards a more focused goal technology development to include flight and surface exploration, such as the moon or Mars. The location of an outpost is a topic of debate between the planetary scientists who desire a location situated near areas of scientific interest (usually near equatorial) and the engineers who prefer to choose a more practical location based on engineering concerns such as lunar day/night thermal cycling, power, communication, orbit to surface site access and probable insitu station resources. For these reasons, lunar architecture studies call for lunar outpost s to be placed on the south pole. At this high latitude the sunlight is relatively constant allowing for solar power, negates severe thermal cycling issues and the 14 day lunar night, allows for continuous direct communication with earth without using relay satellites in lunar orbit, and access to detected stores of water ice located in permanently shaded craters. However, many of the power concerns on a proposed station are not necessarily electrical but thermal for temperature management. Therefore, photovoltaic power systems may not be capable of providing the majority of thermal power requirements at all proposed locations. As a result, Fission Surface Power (FSP) systems are under consideration for manned outpost scenarios.

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33 The proposed FSP reactor design concept will produce 40 kWe nominal output, N aK co oled primary loop, us es f our dual opposed freepiston Stirling power converters heat rejection via titanium water heat pipes coupled to radiators, weighs 5000 kg with an expected 510 year service cycle11. Each of the four Stirling converters produce 10 kWe (5 kWe per alternator) and operates nominally at 5 60C for the heater head and 150C for the cooler11. The general FSP system layout is illustrated in Figure 1-5. The current FSP concept calls for deliver y to the lunar surface with a cargo lander A fter off loading the core is surrounded with regolith, essentially burying the core in order to provide insitu shielding to supplement the onboard shield. Placing the reactor into the lunar regolith is known as the emplaced reactor configuration and is illustrated in Figure 16. The FSP system is to be located approximately 100 m from an outpost and reduces radiation levels to less than 5 rem/yr at 100 m in all directions (360) to meet maximum occupational dose requirements for radiation workers11. In addition, shielding requirements are largely driven by the dose limits to the power conversion subsystem (PCS). MCNP analysis performed by Dave Poston of Los Alamos National Laboratory (LANL) and Lou Qualls of Oak Ridge National Laboratory (ORNL) recomme nded Stirling alternator dose limits Although the shield will greatly attenuate neutrons and rays, coolant activation is the main source of radiation11. NaK (78% Na) is the primary loop coolant and is pumped into the reactor at a mass flow rate of 4 kg/s11. The mechanism responsible for NaK activation is neutron absorption initiating the 23Na(n, )24Na reaction. 24Na emits a 2.75 and 1.37 MeV ray per disintegration and has a t1/2 = 14.595 hours12. Since 23Na constitutes 100% of natural Na abundance the resulting induced radioactivity will be significant13. If

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34 activated NaK flow s directly to the PCS it was est imated that the resulting dose could be above proposed dose limits. Therefore, intermediate pumped NaK loops were included to minimize the S tirling Alternator dose. However, dose requirements were not experimentally verified and the intermediate loop may not be required if Stirling alternator radiation tolerance to expected radiation doses can be demonstrated11. Remov al of the intermediate coolant loops will eliminate numerous components such as heat exchangers, piping, valves, instrumentation, and support structure. The magnitude of the mass savings alone warrants the Stirling alternator radiation tolerance demonstration. The additional components that comprise the intermediate loop will simply increase the overall system complexity, which subsequently increase mass and significantly increase cost and decrease reliability. By removing the intermediate loop b alanceof power plan t issues that arise between the primary and secondary loop are no longer a concern. By only using one loop the design becomes much more simple, practical, and cost effective. Results obtained from M CNP analysis suggest that the Stirling alternators can ex pect to receive a neutr on fluence of 6.5 x1013 n/cm2 and a ray dose of 2 Mrad14. The combined NaK and reactor dose rate as a function of height above the reactor core is shown in Figure 17. The neutron spectrum at the Stirling convertors (1 to 2 m above the core) is illustrated in Figure 18. Completion of a Stirling alternator radiation test demonstration wil l show if the power conversion subsystem can tolerate the activated 24Na ray flux and neutron flux. This demonstration can increase confidence in the power conversion system, thus opening the possibility to simplify the design or relax several radiation safety factors that may potentially dec rease system efficiency

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35 Figure 11. Exploded View Schematic of the G eneral P urpose H eat S ource (GPHS) module3. Table 11. Approximate isotopic composition of PuO2 GPHS fue l4 Nuclide wt % 238 Pu 83.6 239 Pu 14.0 240 Pu 2.0 241 Pu 0.4 242 Pu 0.1 Table 1-2. Radioisotope Power System Specifications1 2 5 RPS Model P BOM (W th ) P BOM (W e ) (%) GPHS Blocks M System (kg) P s (W e /kg) Powe r Conversion Mission Example GPHS RTG 4400 290 6.6 16 54.4 5.2 SiGe TE Cassini MMRTG 2000 120 6 8 43 .0 2.8 SiGe TE MSL ASRG 500 140 28 2 27.1 4.2 Stirling Discovery

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36 Figure 12. Concept design schematic of the A dvanced S tirling R adioisotope G enerator (ASRG)6.

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37 Figure 13. 90day Europa radiation environment based on Divine/GIRE model9. Figure 14. 90day Europa orbiter expected dose based on Divine/GIRE model9.

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38 Figure 15. F ission S urface P ower (FSP) general reactor system schematic11. Figure 16. Deployed emplaced F ission S urface P ower (FSP) reactor concept configuration11.

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39 Figure 1-7 Estimated F ission S urface P ower (FSP) gamma ray d ose a long the a xial centerline above the reactor core Courtesy of Dr. Lou Qualls, ORNL15.

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40 Figure 1-8 MCNP model of the F ission S urface P ower (FSP) buried configuration neutron spectrum at the Stirling power convertors. Courtesy of Dr. Dave Poston, LANL16

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41 CHAPTER 2 PROBLEM STATEMENT A number of design, operational, and safety issues mus t be resolved before the application of Stirling power generation becomes flight mature These nuclear power systems will need to be extremely rugged, reliable and safe, especially when applied to long durationmanned missions. Although these designs see m feasible from a perspective of systems level engineering, the reality is that practical engineering problems are not always taken into consideration. Issues that at first seem trivial eventually require an immense amount of time, effort, and resources to re solve17. Such issues can be properly addressed by conducting subscale demonstrations that provide experimental evidence of design concept feasibility. Free Piston Stirling Power Convertor ASRG and FSP concepts b oth utilize a free piston Stirling convertor coupled to a linear alternator as shown in Figure 2-1 Radioisotope Stirling power concepts have a nominal power output o f 11 0 We using two dual opposed convertors14. FSP concepts produce approximately 40 kWe using four larger dual opposed convertors14. The basic operation of a freepiston Stirling convertor is accomplished by maintaining a fixed frequency with the piston amplitude varied to change the operation point. The piston amplitude is nearly proportional to output voltage of the linear alternator6. Several models of these convertors have been subjected to numerous tests in order to demonstrate capability under expected operational loads. Previous hardware demonstrations include service life trials of over 100,000 hours of continuous operation, thermal vacuum operational tes ts, and vibration tests at expected launch loads to demonstrate capability at operational conditions18.

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42 One concern was in regard to the radiation tolerance of the Stirling convertor alternator The most probable failure prone components in any ( nuclear ) system are those with moving parts operating at high temperature in a radiation environment with possibly radiation sensitive materials. The Stirling alternator, with many moving components constructed from a variety of materials, exemplifies such a conce rn The Stirling alternator employs potentially radiation sensitive materials whose radiation damage limits will dominate location, shielding requirements, and possibly service life14. Due to this perceived risk a literature study on the radiation sensitivity of the Stirling Technology Demonstrator Convertor (TDC) was conducted in 2000 by NASA Glenn Research Center (GRC)10. The TDC incorporated rare earth magnets and numerous organic materials that were chosen without consideration for operation in a radiation environment. It was determined that the magnets can meet the projected mission dose requirements H owever, it was not clear what testing conditions were used to qualify this conclusion. In addition, some of the organic materials would likely requir e replacement with radiation tolerant variants. Potentially radiation tolerant substitute materials of similar function as the original materials were identified but not tested or impl emented. A complete material and system radiation test on a Stirling alternator ha d not been performed; however, it was taken under consideration for radiation hardness verification and to establish a credible radiation design margin (RDM) for the altern ator10. Another item of concern is that the 2000 TDC materials radiation stability study conducted by GRC was compared to the 2007 ASC materials radiation stability study also conducted by GRC and it was found that the new ASC design did not incorporate the more radiation tolerant material recommendat ions.

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43 Reliable system operation requires substantial development to ensure that components meet required service live s at expected operating temperature and radiation conditions. Therefore, it is imperative that experimental hardware demonstrations be conducted under prototypic conditions in order to reduce uncertainty and development costs, while proving the reliability and viability of dynamic radioisotope and fission surface power systems19. Most importantly, these experimental demonstrations will identify unanticipated development issues, which have the potential to drastically increase system reliability while possibly alleviating shielding requirements that subsequently reduce system mass and overall cost. Objectives Th is research project serves as a first order analysis of potential radiation effects to a Stirling alternator. The overall project objectives of this study are to: Establish the Stirling alternator radiation tolerance by exposing candidate material samples and a full scale alternator operating at prototypic operating conditions to expected dose and fluence conditions Conduct post irradiation materials property evaluation. Identify Stirling alternator candidate materials that exhibit excessive radiation sensitivity to the expected radiation environments. Identify radiation sensitive materials and determine radiation tolerant substitutes. Demonstrate that relatively cost effective accelerated life tests in radiation of components an d materials is feasible if a proper understanding of the dominant degradation mechanisms are understood and if careful design of the experimental conditions are maintained. Of these objectives, demonstration of cost effective radiation testing is of critic al importance to all space flight programs. This project will show that similar damage response behavior can be obtained by testing in one type of radiation environment v ersus the exact radiation environment expected in flight. In addition, prototypic

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44 ra diation facilities are typically far more time constrained and expensive. Non prototypic radiation facilit ies can be used as long as the proper damage mechanisms are taken under consideration for each type of candidate material. From these tests we will develop a protocol for properly screening candidate materials and components for service in a space radiation environment before actual flight qualification and acceptance tests are conducted. These system tests expose unforeseen component interactions t hat can be addressed and mitigated before flight. Although exact flight designs are not yet fully established, these experiments allow for candidate materials to be selected based on radiation tolerance early in the design process Environmental t oleranc e is primarily a function of the expected application, not necessarily the specific design parameters For example materials that exhibit excessive property changes at doses of <10 Mrad and a neutron fluence < 1015 n/cm2 are considered radiation sensitive20. This research is unique and innovative in that this type of experimental approach has not been attempted on a full scale Stirling alternator system from a standpoint of operation at prototypic radiation environment s Several materials tradeoff studies have been conducted based on existing data; however, much of the data is inconsistent and not necessarily representative of the environment the Stirling alternator will be subjected to. This research also conducted materials characterization a nalysis on coupon samples and on samples obtained from the full scale system hardware after the testing was complete. Results provide insight into the ability of the system to meet service life requirements and in the process develop methods to improve performance of actual flight units from a radiation tolerance standpoint.

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45 Figure 21. Freepiston Stirling Convertor with linear alternator. Courtesy of Sunpower20.

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46 CHAPTER 3 LITERATURE REVIEW AND HYPOTHESIS The success of the project is heavily dependent on first performing a thorough literature search and conducting small scale experiments to validate the theoretical concepts before proceeding to full scale testing. In order to estimate Stirling alternator service life we must evaluate the performance stability of radiation sensitive component materials. A component has a higher probability of failure if even small property changes (on the order of a few percent ) occur ; however, changes do not necessarily r esult in failure. Depending on the function of the component, operability may still be preserved and the degree to which a system maintains operability is a function of a minimum damage threshold21. Service life can be improved by replacing radiation sen sitive covalent compounds with ionic or metallic compounds. However, this process is not always practical ; therefore, radiation tolerant replacements of similar function must be properly selected. Thus we evaluate the organic and magnetic materials in the Stirling alternator which have been deemed radiation sensitive. This study is done in order to better understand the physical processes that contribute to material degradation from basic principles in order to better select alternative candidate mater ials Radiation Induced Damage Radiation damage is defined as the transfer of energy from an incident radiation to matter resulting in an interaction that induces deleterious atomic structure change. Radiation effect is the subsequent property change (usually adverse) that results from interactions22. Radiation damage is proportional to the energy absorbed per unit mass ; however, this is an overly simplified assumption. Effects are depende nt on the type and

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47 energy of the incident radiation, material characteristics, temperature, and atmospheric conditions21, 23. Material radiation sensitivity is a function of the bond type Ionic materials are essentially ionized and metals are considered to be in an ionized state ; therefore, neither is particularly susceptible to excitation and ionization. Covalent materials are highly susceptible to electronic excitation and ionization21. Radiation interactions relevant to this project are now discussed, as well as the effects on Stirling alternator radiation sensitive materials. Radiation Interaction with Organic Matter The purpose of this section is to specify the numerous interactions that can occur within the alternator organics and how these affect testing. Specific organic material formulations are not listed due to intellectual property restrictions ; rather general name descriptors are used. Material functions required of the alternator include sealants, adhesives, running surface coatings, insulation, bonding, potting, and lubricants20. Covalent bond organic s are highly sensitive to radiation. Organic properties are a function of molecular structure, which ionizing radiation alters creating a nonhomogeneous distribution of ions, free radicals, and excited species22. The ionization density along a radiation path is a function of the type and energy of incident radiation. Linear Energy Transfer (LET) is the rate at which energy is lost by ionizing radiation per unit path length12. LET is characteristic of radiation type, energy, and the absorber. Generally, organics undergo property change proportional to the deposited energy. Photons, electrons, and heavy charged particles interact directly with bound orbital electrons of the atoms that make up the absorbing material, while neutrons interact with the nuclei and indirectly induce ionization. The specific mechanisms are now discussed in more detail.

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48 Heavy Charged Particles Incident charged particles do not cause drastic molecular alterations themselves ; rather, they lose energy by Coulomb force interaction (soft collisions) with bound orbital electrons of the absorbing material and this process excites the electrons24. H ard collisions or interactions with the nuclear field are not discussed for heavy charged particles If the excited electron energy is greater than the binding energy (ionization potential) of the atom or molecule, ionization occurs12. Incident radiation mass is not necessarily important ; however, incident particle energy is a major factor. For example, a n energetic proton will transfer less energy to a bound electron than a slow proton. During inelastic collisions the heavier particles undergo little to no deflection but the impinging electrons experience considerable deflection24. To estimate the energy loss per unit path length known as the stopping power of a heavy charged particle we use the Bethe expression as described by Equation 3-112. =4 224 22 ln2 22 ( 1 2) 2 (3 1) where: ko = 8.99 x 109 N-M2/C2 e = magnitude of the electron charge mo = electron rest mass z = particle charge atomic number n = volumetric electron density c = speed of light in vacuum = v /c = speed of the incident particle relative to c I = mean excitation energy of the medium The value I can be determined by the empirical relationship shown in Equation 32. 19 .0 =1 11 .2+ 11.7 ,2 13 52 .8+8. 71 > 13 (3 2)

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49 Electron s and Positrons Incident electrons interact differently than heavy charged particles. For nonrelativistic ( E < < moc2) electrons the energy loss per unit path length, or collisional stopping power is defined by E quation 3-312. 5656 56565656 =4 242 22 ln2 +2 2 + ( ) (3 -3) where: = T/moc2 T = kinetic energy of the electron or position F for electrons and positrons can be determined by the functions described by E quation 3-4 and Equation 35, respectively12. F( ) =1 22 1+28 ( 2 +1 ) ln 2 (3 -4) F+( ) = ln 2 224 23 +14 +2 +10 ( +2 )2 +4 ( +2 )3 (3 -5) From this we can see that many collisi ons are required to slow an energetic electron. Incident electrons may also lose energy by producing electromagnetic radiation ( xrays) when they are decelerated through an atoms electromagnetic field in a process known as B remsstrahlung radiation. Equation 3-6 approximates the ratio of radiative to collision energy loss12. 5E5E5E5E 5E5E5E5E5E5E 5656 56565656 58585858 80 0 (3 -6) These functions are generalizations to better understand the basic principles of interactions. Complex programs are used to model the systems and determine with excellent certainty the reactions expected to be produced in a system. Next, we discuss the resulting tracks that form in the medium from the passage of an incident particle.

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50 Particle Track Formation Io nizing radiation produces a series of randomly distributed primary events along a path with s econdary events clustered around primary interaction locations. Often, the secondary charged particle tracks look like a string of small randomly spaced spurs. Primary and secondary event density is a function of the radiation type and energy. E lectrons produce a relatively low ionization density, while heavy charged particles As the ionizing particles slow the ionization density increases. For proton and alpha radiation the end of track ionization density is very high due to the Bragg effect. At even lower energies the charged particles capture electrons, becoming neutralized. When reduced to thermal energies they undergo molecular collisions and induce local temperature increases. Photon Photons (e.g. xrays and rays) interact with orbital electrons of the atoms of the absorbing material primarily via the photoelectric effect, Compton effect, and pair production23. The type of interaction depends on the energy of the incident photon and the atomic n umber of the irradiated material as illustrated by Figure 3-1 and Figure 3-2 P hotoelectric absorption occurs when a photon of energy hv is completely absorbed by a bound electron, most likely a Kshell electron, and is ejected as a photoelectron with energy as shown in E quation 3-7. 5858 = 5858 (3-7) Here we define Eb as the binding energy of the photoelectron in a specific shell12. Compton scattering is the result of an incident photon interact ing with an electron within the absorbing material. The interaction produces both a recoil electron and a scattered photon with energy hv as described by E quation 3-8.

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51 = 1+ 2 ( 1 cos ) (3 -8) We define moc2 as the rest mass energy of the electron (0.511 MeV) and is the photon scatter angle25. Pair production occurs when an incident photon of E > 2moc2 (1.02 MeV) interacts with an electron, creating an electronpositron pair, with the positron subsequently undergoing annihilt ion25. The result of these interactions is the production of secondary electrons, which following eject ion by the incident photon will interact and cause damage by electron excitation and ionization. Such interactions can cause considerable damage to organic covalent compounds21. Compton scattering is primarily responsible for change in organic materials bec ause the energy range of the expected photons is a region where the Compton process dominates24. Neutron Neutrons do not interact with electrons ; rather they interact directly with nuclei. Neutrons indirectly cause secondary radiation effects due to the scatter of recoil nuclei scatter of recoil proton s, radiative capture, inelastic scatter, and the formation of transmutation impurity atoms within the absorbing material12. Higher energy neutrons lose energy by elastic scattering collisions with nuclei of absorbing material. Equation 3-9 is used to describe the maximum energy a neutron of mass m with kinetic energy E can transfer to a nucleus of mass M in a single headon elastic collision26. 5@5@5@5@ =4 5Z5Z5Z5Z ( + )2 (3 -9) From this function we see that more energy is transferred to lighter nuclei which makes up the bulk of organic materials ; therefore, these elastic collisions can cause considerable damage. This statement is a simplified generalization for there are

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52 regim es where inelastic collisions can dominate elastic collisions (such as neutrons with energies above approximately 1 MeV interacting with heavier mass targets) but does not apply to neutron energies expected at the Stirling alternators The nucleus is ejec ted as a recoil ion and interacts with other bound electrons producing excitation and ionization. The recoil ion neutralizes and the neutron scatters to interact with other nuclei, continually losing energy in the process Higher energy neutrons of inter est, from approximately 0.1 to 1 MeV, will be present outside the FSP shield but in small numbers. These neutrons lose energy by inelastic scattering with low mass nuclides is small when compared to elastic26. Conversely, the energy loss can be significant when energetic neutrons undergo inelastic scattering with high mass nuclides. A nucleus can capture a fast neutron and eject a charged particle, if the ejected particle has sufficient energy to overcome binding energy26. The threshold energy for proton ejection increases with the atomic number of the impacted nucleus. These fast recoil protons create ionization and excitation clusters along their tracks. A m ajor difference between neutron and proton reactions is that protons (the secondary particle fro m neutron interaction) leave a much denser ionization track than do electrons (secondary particle from photon interaction). Thus, for equivalent deposited energy the number of ionization tracks by neutrons is far fewer, but the number of events within the track is far greater24. Since most organics are hydrogenous in nature, neutrons are rapidly reduced to thermal energies and much of the deposited energy will result in elastic recoil of protons. These recoil protons are capable of inducing ionization, e xcitation and break chemical bonds throughout the absorbing material as mentioned previously24.

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53 Once thermalized, neutrons are absorbed by nuclei. In the case of organic materials, thermal neutrons undergo radiative capture by hydrogen. The 1H(n, )2H re action results in the emission of a 2.2 MeV ray that cause further ionization of the surrounding material. Thermal neutrons also cause the 14N (n,p) 14C reaction, resulting in a 0.626 MeV proton emission12. In summary, photons, electrons, and thermal neutrons via (n, ) reactions damage by the same mechanism ; thus, for a specific energy deposited an equivalent damage results. Fast neutrons, protons, alphas fission products, and thermal neutrons via ( n,p) and ( reactions damage by different mechanisms ; thus, a different (likely non linear) damage/energy relationship exists21. Radiation effects in organics are more a function of excitation and ionization rather than atomic displacement24. Although organic radiation effect s are dependent on total absor bed dose and not necessarily dose rate, there is sufficient evidence to show a dose rate dependence Thus, it is possible for a material to receive a dose at different dose rates and exhibit slightly different changes24. This behavior is explained later and is dependent on irradiation conditions. Radiation Induced Chemical Effects in Organics Radiation induced chemical effects are a function of the relative time required for various competing processes to occur. These processes include free radical production, radical scavenger (affinity for free radicals) reactions, and the production of evolved gases24. Addressing molecular radiation effects is difficult due to their complexity. Atoms only have electrically excited states ; however, molecules have elec trical, vibrational and rotational excited states. In addition, every electrically excited state has a corresponding set of vibrational and rotational states24.

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54 Chemical Structure of Organics The properties of polymers are a function of the material physical state (e.g. crystallin e glass, liquid, or rubber). The physical state is governed primarily by the chemical structure of the molecules. The two most important factors to chemical structure are forces between molecules and molecular shape. Intermolecular forces are the result of attraction between permanent dipoles, asymmetric charge distribution, and molecule polariz ation. Chemical structure also controls the stability of a polymer to chemical attack and atmospheric aging. Previous studies have show n that thermal resistance and radiation resistance do not necessarily correlate27. Effect of Crystallinity Polymer crystallinity is a term used to refer to the simpl e arrangement of molecular chains that pack in a repeatable manner and produce an ordered atomic array28. As with other crystalline materials they are characterized by unit cells and can be very complex as illustrated by Figure 3-3 Although the figure depicts a completely crystalline structure, in reality, a crystalline polymer is a mixture of completely crystalline regions and completely amorphous regions, whose combined presence can have an impact on the behavior of the polymeric material24. The degree of crystallinity depends on the rate of cooling during processing ; however, this ca n be modified in subsequent processes and during service life The physical properties of polymeric compounds are influenced by the degree of crystallinity because it affects intermolecular secondary bonding. It will later be shown that radiation can modify the molecular chain lengths, thereby, modify the average molecular weight of the polymer Changing of the molecular weight can have a direct impact on crystallinity and consequently the physical properties of the polymer.

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55 Radiation Effects Radiation e ffect s on polymers produce ionization and excitation, breaking the covalent bonds resulting in molecular fragments with unpaired electrons. These free radicals react to change the chemical structure of the polymer and alter the physical properties. Polym ers subjected to radiation may experience scission cross linking, and gas formation. These processes are described in more detail in the next several subsections. Various structural components, such as the elements that make up the main chain, side chains, and substituent groups determine the nature and rate of radiation induced effects. Polymers with high aromatic ring content are most resistant to radiolysis. Aromatic compounds have relatively stable ring structures, with a high resonance energy which allows the energy deposited to be shared by many electrons without bond breakage29. Highly aliphatic compounds (carbon atoms bonded in straight or branched chains or in rings) are least resistant24. Cross Linking Cross linking is the formation of new bonds between adjacent polymer molecules. A small side group, usually a hydrogen atom, is broken from the main chain and the free bond forms a link to a neighboring molecule as shown in Figure 3-4 b. This increases the average molecular weight, decrease s solubility, decreases elongation, increases softening temperature, and increases tensile strength. Further irradiation beyond the onset of cross linking stage results in the polymers forming one large threedimensional molecular network, which imped es v iscous flow as shown in Figure 3-4 c. Additional crosslinking increases network rigidity, thus, increasing Youngs modulus. Polymer properties are now governed not by chemical structure but by crosslink network density. As the cross link density increases, a soft amorphous polymer will change into a

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56 rubbery material, then become hard and glassy. During this process the molecules come closer together and decrease the specific volume. Highly cross linked polymers become hard and brittle, with poor impact strength. Scission Scission is the breaking of molecules into smaller fragments, which increases viscous flow and decreases molecular weight and softening point. Materials soften (reduction in modulus) due to the increased short chain population, which decreases tensile strength and allow s for increased elongation. Further scission can increase polymer crystallinity because of less restraint on shorter molecules. Crystallinity increase has a corresponding density increase and leads to hardness/embritt l ement as well as gas formation24. Gas Production Radiation generated gas is observed in polymer decomposition. The majority of the gas is hydrogen, CO, CO2 and low molecular weight hydrocarbons such as CH4. Gas evolution yields more hydrogen than carbon since C H bonds are more easily scissioned than C C bonds, which leads to an increase in C:H ratio30. The release of hydrogen from polymers under irradiation is assumed to be the result of radicals escaping the material. If in the presence of oxygen, the radicals will react with the oxygen greatly accelerating the oxidati on process31. The range of gas yield varies due to different irradiation conditions and/or sample materials24. The alternator organics are sealed in the pressure vessel and operate in u ltra high purity helium at ~ 3.5 MPa (500 psi), which should inhibit outgassing and radical oxidation induced aging Any g aseous products that form should deposit on the pressure vessel since it is the coldest part of the system and not pose a credible failure mechanism10.

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57 Temperature Effect Temperature is a dominant condition that directly affects physical and mechanical properties of organics. Pol ymer behavior is a strong function of operating temperature and of previous heat treatments. Temperature has a significant effect without radiation as shown in Figure 35 and Figure 3-6 Temperature becomes dominant at much lower temperatures when irradiating organics. Free radical production rate is determined by the intensity of the radiation. Free radical reaction rates are affected by mechanisms that change the mobility of species, such as temperature. Radical recombination competes with other reactions such as cross linking T herefore, given a radical production rate the cross linking rate can be influenced by changing the temperature. For example, at higher temperature more cross linking occurs due to less free radical recombination occurring wit hin a polymer24. When analyzing the combined effects of temperature and radiation on polymeric material physical properties the structure must first be addressed. As previously mentioned, irradiation will alter the molecular structure inducing cross linki ng and scission (which can increase crystallinity). The combined effect will simply appear as changes in the performance of the engineering properties at specific temperatures due to changes in molecular structure, which are well illustrated by Figure 3-5. A synergistic relationship exists between heat and radiation. Equation 310 is used to compare the rate constants of the C=O release by heat and radiation, kc(h + r), which was higher than the sum of heat or radiation alone32: ( + ) > ( ) + ( ) (3 10) Therefore, efforts to improve thermal stability of a polymer do not necessarily result in an improved radiation resistance.

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58 Atmospheric Effect Radiation induced free radicals and perhaps other reactive species have been experimentally shown to be highly sensitive to oxygen. In many cases, post irradiation aging effects have been observed. For example, irradiation under vacuum and subsequent exposure to air can induce rapid oxidation. It was found that the reaction rate is a function of oxygen diffusi on rate into a specimen and is greatly affected by sample thickness. Scission produced free radicals react with oxygen, aging the material. Free radical recombination involves one group of radicals reacting rapidly with their neighbors, while the other group of radicals take much longer to react due to limited mobility33. For this reason post irradiation material handling, characterization, and evaluation was conducted in an inert, low oxygen environment24. Since the alternator organics operate in an ult ra high purity helium environment, oxidation should be kept at a minimum. Materials characterization will provide data on the amount of irradiation oxidation that occurs after irradiation of polymers in ultrahigh purity helium Changes in Electrical Cond uctivity Polymer electrical resistance is known to decrease proportionally to radiation dose and radiation intensity34. Radiation induced conductivity is the result of the formation of mobile electron hole pairs that degrade the performance of organic dielectrics35. In high flux fields a decrease is up to three orders of magnitude is possible but is recovered when ionization terminates. This effect results from mobile charges available for co nduction yielding semiconductor like behavior. Assuming no severe deterioration of the electrical insul ation occurs, this decrease in electrical resistivity should not pose a major failure mechanism. A larger concern is more likely to be wire insulation embrittlement from the combined effects of temperature and radiation.

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59 Lubricant Degradation Lubricants typically consist of an organic anti wear additive (such as PTFE ) dispersed in an organic binder (such as a polyimide base matrix). Lubricants must resist changes in temperature and oxidation. Little damage occurs with increasing temperature until a thermal threshold is met, above which degradation rapidly accelerates. Experimental studies have demonstrated that increasing the oxygen content and/or temperature rapidly accelerates lubricant property change. Rad iation damage is a minor condition below the temperature threshold; however, radiation can exacerbate oxidation susceptibility24. Lubricant radiation stability is dominated by the binder stability. Lubricants harden due to binder cross linking of the base oil; however, anti wear additives degrade at doses well below binder doses. Experimental evidence indicates that it is preferential to choose radiation resistant bases rather than attempting to improve stability of inferior bases through the implementation of additives. Xylan s urface coatings (PTFE particles in a polyimide binder) are used in the Stirling alternator and provide a contact running surface only during brief start up or shut down periods. During normal operation gas bearings prevent contac t. If considerable degradation of Xylan running surface is observed, other options are available24. The use of inorganic anti wear particles such as graphite or MoS2 powder dispersed in an organic binder function as a lubricant at doses well above thos e that degrade the binder. T ypical lubricant behavior as a function of dose is detailed in Table 31. Solid lubricants are not greatly affected by radiation. Wear tests of graphite and MoS2 in a mixed neutron and ray field to 2x109 rad show no wear property deterioration24. Solid lubricants would be best used for the higher operating

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60 temperature and fluence/dose expected in the FSP system. In addition, machine elements do have some tolerance for function with degraded lubrication and frequently function at radiation doses above the point were predicted critical degradation occurs. The majority of the Stirling alternator components do not require running surface and conditions are not aggressive (relative to temperature and stresses); there fore, it is unlikely that a lubricant failure will constitute a system failure. However, it may be possible for the running surface coating to degrade into smaller particles that can distribute themselves throughout the interior of the alternator. Powder deposition can induce friction throughout moving components and decrease convertor performance. Adhesives Off the shelf a dhesives are available that exhibit excellent thermal and radiation stability. Improving adhesive radiation resistance is difficult because formulations are often proprietary and additives generally result in negligible improvement. Therefore, it is recommended that adhesives under consideration be tested at expected conditions24. Traditionally, adhesives undergo static exposures in air at ambient temperatures with no load applied to test specimens. Nondestructive test methods such as ultrasonic inspection have been used to detect radiation damage with limited success Destructive methods have been performed including tensile, shear, bend, and fatigue tests. It was found that electron and ray exposure deteriorated tensile shear strength of adhesives for a given dose and degraded with increasing temperature. Bend strength tests also indicated damage and increased with temperature f rom ambient to 673oC for a given dose of radiation 24. It must be noted that some adhesives exhibit property improvement with exposure, indicating that radiation completes the curing process36.

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61 Phenolic epoxy based adhesive (422) exhibits good thermal and radiation stability. 422 consistently retains up to 82% strength after 8.6x108 rad. 422 did not exhibit any negative effects when irradiated at ambient temperature with a mixed neutron field (6x1014 fast n/cm2 and 4x1013 thermal n/cm2) at a dose of 1.7x107 rad. Post irradiation mechanical testing was conducted from room temperature to 673C. 422 withstood greater fatigue stresses and slightly degraded with high doses. Based on the testing recommendations adhesives should be tested at expected operational conditions to include radiation type and energy, atmosphere, and temperature, followed by shear, bend, and peel stress tests. Having a better understanding of these potential effects will allow for improved estimation of an ac curate service life es timate for the components that utilize adhesives Adhesives under consideration in the Stirling alternator service include Hysol and high temperature Hysol epoxy resins. Additives Additives can increase the radiation tolerance of polymers and are either a ctive or inert in nature. Active additives act as either energy sink s or chemical reactants. Energy sinks absorb energy from the irradiated material and dissipate the energy as heat or light. Aromatic hydrocarbons are good protective agents. Chemical reactants are sacrificial agents that are preferentially destroyed such as antioxidants. The total damage to the system remains the same as would have been without the sacrificial agent. Iodine or other radical scavengers are also sacrificial. Inert additives known as filler will make up a significant portion of the polymer mass in which energy is deposited, which will subsequently reduce radiation effects24. Few if any Stirling alternator organics utilize additives. M aterials that do utilize fillers are used obtain desirable properties not necessarily for improving the radiation tolerance of the polymer

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62 NdFeB and SmCo Rare Earth Permanent Magnets Rare Earth (RE) magnets are utilized in applications where a compact, high magnetic property material is crucial. Permanent magnets are composed of small regions or domains each of which exhibit a magnetic moment28. In an unmagnetized magnet the domains are randomly oriented resulting in a net magnetic moment of zero. External magnetizing fields are used to align the domains to give a net magnetic field37. The magnets are magnetized to saturation although slight demagnetization will occur during the stabilization process described later. The two RE magnets under consideration for use in a Stirling alt ernator system are NdFeB and SmCo whose basic properties are illustrated in Table 3-2. The performance of the magnets will have a direct impact on the electrical power output of the convertor T herefore, it is important to address the mechanis ms that effect magnet stability in order to properly develop and radiation mitigation strategies if they are indeed required. Time Magnets experience some demagnetization with time, known as magnetic creep. H igh coercivity RE magnets are less susceptible to magnet ic creep For example, at over 100,000 hours SmCo experiences essentially no time demagnetization. Thermal Effects Increasing temperature induces lattice vibrations, which allow aligned magnetic moments to rotate and loose alignment. Demagnetization is a function of domain wall mobility, which is controlled by microstructure and is dependent on manufacturing. This phenomenon is dominated by the magnet Curie Temperature (Tc) where the m agnetic moments are randomized and the material is completely demagnet ized28. For this reason maximum operating temperatures are established for each magnet grade.

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63 Temperature induced demagnetization losses can be categorized into three c onditions Reversible demagnetization losses are recoverable when the magnet is returned to a lower temperature28. Irreversible but r ecoverable losses result from high temperature, which are not recovered when the temperature returns to the initial value with l osses recovered by remagnetization28. Irreversible and u nrecoverable lose s result from microstructural changes and losses are not recoverable28. Thermal stability is improved by carefully heating the magnet to elevated temperatures. Stabilization allows for weakly oriented domains to preferentially lose orientation. Slight demagnetization does occur ; however, the magnetic properties during operation are more predictable, exhibiting nearly constant behavior when heated to the expected operating conditions. A batch of stabilized magnets will exhibit lower variation in perform ance and is recommended for all high temperature applications37. Magnets with high er permeance coefficient s usually have improved high temperature performance. NdFeB magnetic properties deteriorate rapidly above 130C. NdFeB designed with a high Hci and operating at a high permeance coefficient can operate up to 180 C. SmCo exhibits superior thermal stability, especially Sm2Co17 and can operate up to 350C. However, SmCo is susceptible to thermal shock fracture when high temperature gradients exist within the material37. Shock, Stress, and Vibration Unfortunately, since SmCo is an inter metallic alloy these magnets are very brittle SmCo is known to be susceptible to chipping cracking and erosion of the surface layers through handling Electrolyt ic nickel surface plating is commonly used to improve the fracture toughness of SmCo. The nickel surface barrier exhibits good thermal stability and excellent surface adhesion throughout a wide range of temperatures38.

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64 Radiation Effects on Rare Earth Magn ets Rare Earth magnets are susceptible to ra diationinduced demagnetization; however, they are used in particle accelerators in radiation environments orders of magnitude higher than the Stirling alternator is expected to see. Numerous studies have reported on the radiation effects in RE magnets with the majority maintaining a low irradiation temperature in order to not have temperatureinduced demagnetization interfere with radiationinduced demagnetization. A 0.5% magnetization loss in synchrotron devic es is typically not tolerated; therefore, small variations (0.1 0 .2%) are considered significant39. Therefore, limits for demagnetization of the Stirling convertor must be established in order to determine the magnet radiation dose limit. It is difficult to compare investigations of NdFeB and SmCo radiation induced demagnetization due to the numerous variables involved. Examples include experimental procedure, radiation types (photon, electron, proton, and neutron), total absorbed dose, elemental composition, and variation in manufacturing processes39. The degree of demagnetization depends more on the radiation type, not necessarily the deposited energy. For example, electrons are more damaging than an equivalent dose of rays40. Next, we discuss and summarize many studies that have been conducted on the effects of radiation o n NdFeB and SmCo RE magnets. Photon NdFeB and SmCo magnets are not affected by photon radiation. NdFeB magnets were irradiated with X rays at <50C with an absorbed dose of 280 M rad and with a 60Co ray source at <60C with an absorbed dose between 50 and 700 Mrad40 41, 42 43. For the X ray and the ray exposures the average decrease in residual induction was

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65 less than 0.2% which was found to be statistically insignificant and most likely attributable to the experimental uncertainties39. These errors seem very small and may likely require additional verification to validate the statistical accuracy of the results. Electron NdFeB magnets w ere irradiated at 20C with 17 MeV electrons up to 170 Mrad and show ed a remanence loss ranging from 0.2 5.6%. SmCo magnets under the same conditions show negligible change, on the order of 0.1% gain, from the preexposure value. Remanence loss from electron irradiation is not apparent for temperature resistant grades of NdFeB, which suggests that losses are thermal effects resulting from local heating and is lik ely dependant on the incident radiation energy43 44. Neutron The e xpected FSP radiation environment primarily consists of fast neutrons and ray s. Although thermal neutron population will be minimal in the reactor core, the thermal population will be substantial at the location of the Stirling convertors after moderati on through shield. This poses a concern since NdFeB uses natural boron with 19.8% 10B abundance, which has a large thermal neutron absorption cross section. The 10B( n )7Li reaction can cause damage and swelling. NdFeB magnets can withstand some degree of irradiation which is dependent on microstructure. SmCo5 and especially Sm2Co17 are relatively insensitive to radiation damage by fast neutrons42. Evidence suggests that the mechanism for remanence loss is caused by the nucleation of reverse domains by collisions. These reverse domains have a higher probability of occurring when the temperature is closer to Tc effective45. Results from numerous sources have been compiled i n Table 3-3. Although there are

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66 differences in the data, the overlying result is that SmCo does indeed outperform NdFeB in a neutron environment Therefore, SmCo should undergo serious consideration for FSP Stirling convertor concepts Proton SmCo5, Sm2Co17, and NdFeB magnets were exposed to a 500 MeV proton beam, with a maximum temperature of 125C. SmCo magnets exhibit measureable demagnetization when irradiated with protons from 1091010 rad. Sm2Co17 is most resistant to proton radiation experiencing demagnetization <0.4 1.8%. Different vendor magnets have large differences in the irradiation demagnetization due to variations in manufacturing and composition. NdFeB proved extremely sensitive to proton irradiation experiencing up to 55.4% remanence loss at 70C and a fluence of ~1014 p/cm2 (4 Mrad) and essentially 100% at 70 Mrad45 4 6. SmCo demagnetization loss due only to thermal heating effects after 100 hours at 250C was 0.31 1.48% varying with vendor. NdFeB at the same conditions experienced less than 3% loss46. Radiation Mitigation Techniques Experiments have been conducted to determine the effect of radiation on RE magnets. In the majority of experiments the temperature was controlled in order to differentiate between thermal and radiation induced demagnetization. The experimental evidence suggests that fast neutron and proton irradiation damage is caused by a radiationinduced temperature spik es which can exceed the Curie temperature47. It has been determined that the radiation tolerance of a magnet is a function of thermal stabi lity, which is dependent on the Tc and Hc. Radiation induced localized temperature spiking is dependent on radiation intensity, thermal properties (e.g. specific heat capacity and thermal conductivities). Regardless of the radiation, source SmCo has

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67 superior radiation tolerance over NdFeB magnets48. Radiation mitigation methods include utilizing sufficient radiation shielding o perat ion at high permeance coefficients Pc, p re flight radiation stabilization to expected levels, designing in a high Hc and thermal stabilization37. Consequently, a h igh Hc equals higher radiation tolerance. NdFeB or SmCo grades can have Hc that may vary by more than a factor of three depending on the composition and the manufacturing process40. T hermal stabilization is done by baking magnets for several hours above 140C with some demagnetization occurring (0.1 0.4%). After 50x1013 e/cm2 non stabilized magnets had greater than 2% demagnetization while stabilized magnets had less than 0.5 1.5% demagnetization40. Desi gn of Experiments: Accelerated Life Tests A ppropriately designed experiments are essential to conduct accelerated li fe tests. From the design of experiments method we reduce the number of possible variables that may influence materials to four important f actors : Material Type Depend e nt on bond type (ionic, covalent, metallic) t hermal stability r adiation resistance, microstructure, and processing history. Temperature Depend e nt on r ate of change (dT/dt) and max imum operating condition (Tmax). Atmosphere Depend e nt on oxygen concentration (C0) Not necessarily dependent on pressure. Radiation Depend e nt on particle type, particle energy (E), cumulative dose (D), and cumulative fluence ( t) In certain circumstances the dose rate (dD/dt) and fluence rates (d t/dt) are important W hen considering organic materials there is little dependence on dose rate as long as irradiation occurs in an oxygen deficient environment. This relationship allows us to select the appropriate variables that can be modified to develop appr o priate accelerated life tests as described by E quation 3-11. 5555 5N5N5N5N5N5N5N5N5N5N = 5O5O5O5O5O5O 5Q5Q 5G5G 5@5@5@5@ 5]5]5]5]5]5]5]5]5]5]5]5]5]5] 5858 5a5a (3 -11)

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68 Conclusions In summary, radiation testing of candidate materials and flight components is absolutely necessary because much of the existing literature is insufficient or inconsistent to ensure mission success. As was demonstrated in this chapter much of the existing literature was insufficient because the m ajority of studies were conducted in a way that not all of the relevant environmental properties were controlled; therefore, yielding a different result than for operation of a Stirling Alternator (e.g. polymer irradiations in oxygen vs. inert gas, lack of sample temperature control, etc. ) As mentioned previously, tests must be conducted in controlled environment conditions in order to obtain relevant results. In addition, s ystem testing is of critical importance due to the p ossibility of synergistic effects that cannot be dis covered by coupon test s alone. Nearly all of the existing literature does not mention synergistic effects, which is a well known, and likely to occur phenomenon during the service life of any operating system. Test s must be considered to ensure that resul ts a re applicable to the appropriate environment. By understanding and controlling the radiation damage and subsequent radiation effects on materials we can select economic radiation test facilities that will yield equivalent response in materials and com ponents. Therefore, exposures to appropriate absorbed dose levels with equivalent damage mechanisms are a valuable screening tool for materials and components using cost effective radiation sources. For example, a 60Co source will provide a radiation environment with lower energy particles at a higher particle flux. Such an environment can encompass an equivalent portion of the same radiation effect that would be experienced in the Jovian or FSP radiation environment. Therefore, such ray testing can represent a worst case testing scenario.

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69 Figure 31. Expected electron interaction with energy & ato mic mass dependance25. Figure 32. Conceptual example of the molecular chain unit cell (crystallinity)28.

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70 Figure 33. Influence of crystallinity and molecular weight on physical properties28. C Figure 34. Representation of A ) linear bra n ches B ) cross linked branches C ) 3D highly cross linked branch network28. A B

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71 Figure 35. Relaxation modulus vs. Temperature for amorphous polystyrene28. Figure 36. Temperature effect on crystalline, lightly cross linked and amorphous polymers28.

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72 Table 31. Effects of radiation on lubricants24 Dose (rad) Observed Effect on Lubricant < 10 6 Little to no deterioration 10 6 10 7 Some lubricants usable, others exhibit marginal performance 10 8 10 9 Thermal stability & oxidation of most lubricants severely impaired 10 9 10 10 Polyphenyls, poly (phenyl ethers), alkylaromatics recommended > 10 10 Oil lubricants become hard/brittle solids, Dry lubricants recommended Table 32. NdFeB and SmCo Material Properties28 3 7 Material Br (kGauss) Hc (kOe) BHmax (MGOe) T c ( C) T max ( C) Radiation Resis tance Corrosion Resis tance Machine ability Nd2Fe14B 10.211.6 1017 1048 310 150180 Poor Poor Good Sm 2 Co 17 8.8 9.6 9 16 18 32 725 300 350 Good Good Poor Table 33. Neutron Irradiation Effects on NdFeB and SmCo Permanent Magnets Material Composition T max ( C) (n/cm 2 ) Energy (MeV) Irrad Time (h ou r s ) B r Loss (%) Source NdFeB Melt spun 4 7 80 1.4x10 16 >5 eV 1 1.5 Reactor NdFeB Melt spun 4 7 80 N/A >5 eV 5.3 3 Reactor NdFeB (sintered) 4 7 80 1.4x10 16 >5 eV 1 4.6 Reactor Nd 13 Dy 2 Fe 77 B 8 (48 ) 267* >10 16 0.02 10 N/A 100 Reactor NdFeB grade 3 9 25 10 12 Thermal N/A 0.03** 252 Cf NdFeB grade 39 25 10 13 Fast N/A 0.6 252 Cf NdFeB grade 39 25 10 14 Fast N/A 10 252 Cf SmCo grade 47 4 9 78 10 15 >0.1 41 <0.5 Reactor SmCo grade 48 300 10 18 0.02 10 N/A ~0** Reactor SmCo grade 48 500 10 18 0.02 10 N/A ~0 Reactor SmCo grade 48 77 6.1x10 16 Fast N/A 5 Reactor SmCo grade 48 153 5.0x10 16 Fast N/A 10 Reactor *Temperature is much higher than the recommended maximum operating temperature. **Within experimental uncertainties (not statistically significant)

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73 CHAPTER 4 APPARATUS AND PROCED URE: SARTA RADIATION TESTING The research project consisted of four phases. First, a Stirling alternator is radiation tested in a ray environment at accelerated rates and at expected operating conditions. Second, the alternator is disassembled and the internal components evaluated for changes in function. Organic materials from the alternator underwent a thorough characterization process The performance of the alternator was analyzed to determine if any degradation occurred during operation. Third, sample coupons of select cand idate organic materials are irradiated in a mixed neutron and ray flux at accelerated rates, under prototypic operating conditions. Fourth, the irradiated organics underwent characterization and are compared to control samples and samples removed from t he irradiated alternator. These combined results provide a much more clear understanding of the service life and credible failure mechanisms of a Stirling alternator and have direct impact to design considerations. Stirling Alternator Component Radiation Testing The Stirling Alternator Radiation Test Article (SARTA) design was based on the linear alternator of an Advanced Stirling Convertor ( A SC ) which was designed and built by SunPower Inc. The SARTA (F igure 4-1) differs from an AS C in that it lacks a heater head, heat exchangers, a displacer, and produces no electric power. The primary function of the SARTA is to operate a functionally equivalent Stirling alternator section in an accelerated radiation environment and investigate the durability of the polymeric and magnetic materials used in an AS C The SARTA operating conditions (temperature, pressure, power factor, and stroke) were chosen to match those of an AS C linear alternator. Two internal type K thermocouples were embedded

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74 into the inner laminations cylinder and coil to monitor temperature and used as an appropriate temperature control reference. Temperature is controlled with a resistance band heater mounted direct ly to the pressure vessel wall and a continuously operating muffin fan mounted above the SARTA. The band heater is used to sustain the SARTA at the appropriate operating temperature (90 C or 125C) through the use of a Eurotherm PID temperature controller linked to the inner lamination cylinder thermocouple as a reference source. Two additional type K thermocouples were externally mounted to the pressure vessel using Kapton tape. The two internal and two external thermocouple temperature measurements are used to develop a spatial temperature profile of the SARTA during steady state and transient responses to changing operating conditions. The pressure was measured using an Ashcroft K1 thin film high accuracy pressure transducer. Pressure was actively controlled to maintain 500 psig through the use of an ultra high purity helium gas k bottle a series of manually controlled valves, a pressure regulator, and automatic pressure relief safety valves. The SARTA was motored at a constant nominal input voltage to obtain a piston stroke of approximately 8.0 mm. The piston stroke was measured by the use of a Fast Linear Displacement Transducer (FLDT) located inside the SARTA. The stroke input voltage was provided by a manually controlled AC power supply operating at 60 Hz and a nominal input voltage of 8.2 VAC ; however, this varie d with operating temperature. SARTA power, voltage, current, stroke, pressure, and temperature were measured and recorded through the use of an Agilent 34970A data acquisition/switch unit used in conjunction with Agilent Data Logger software running on a laptop computer.

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75 The waveform produced during operation was monitored and recorded using a Tektronix TDS 3054 Oscilloscope. The period, frequency, amplitude, peak to peak, mean, and root mean squared values produced during radiation testing were manually recorded in the logbook and compared to those values obtained before radiation testing to look for changes as a function of increasing dose. The waveforms produced during radiation testing are also compared to each other as a function of increasing dose a nd to pre exposure waveforms A microphone coupled to a speaker system allowed the SARTA operator to monitor the audible sound produced by the SARTA dynamic balancer and the muffin fan during radiation testing. The microphone was required since the SARTA was not usually visible during radiation testing because the GIF cell glass was shielded during extended exposures by internal doors to preserve the glass lifetime. Gas content scans were obtained using a Dycor Dymaxion Residual Gas Analyzer (RGA) before t esting to determine background environment of the vacuum and the unirradiated system. Scans were also obtained after each exposure in order to detect evolved gases. The complete SARTA control and instrumentation system is shown in Figure 4-2 The compl ete system schematic is located in Appendix A Figure A 1. Pre Exposure Characterization The SARTA underwent extensive preirradiation testing at both Sunpower and NASA G lenn R esearch C enter (GRC) totaling over 64 hours of operation After assembly it was operated for approximately 15 hours at SunPower for system checkout testing. Once at GRC the SARTA was operated for approximately 49 hours in order to establish variable sensitivity, baseline operation criteria, develop operating procedures, and test the system hardware.

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76 Thermal Response and C ontrol Pre exposure test s conducted at NASA GRC characterized thermal time constants by observing system response as a function of temperature and pressure for a range of nominal and possible off nominal conditions as illustrated in Figure 4 3. In addition, several methods of temperature control were considered and it was decided to continuously provide cooling with a muffin fan mounted axially above the SARTA while simultaneous providing heat from the externally mounted band heater. This method allowed for operation at both 90C and 125C with sufficient temperature margin to account for possible radiation induced heating and in the event that the cooling fan would fail prematurely during the tests In order to prevent the PID controller from overshooting the set point temperature and overheating the SARTA, the maximum heater output power was limited to 40%. Performance M apping Performance mapping was also conducted at NASA GRC in order to gather sufficient data to perform a thorough statistical analysis to determine the acceptable operating limits of the SARTA. Several eight hour long steady state runs at nominal stroke (8.0 mm), pressure (500 psig), and temperature (90C and 125C) were conducted. Additional tests were carried out at off nominal stroke (6.50 mm 8.50 mm) in order to determine the behavior and magnitude of theorized steady state and transient deviations as shown in Figure 4-4. Selecting the acceptable operating limit range was accomplished by applying a 99% confidence interval to eight steady state runs. These established limits allow the operator to accurately identify the difference between nominal operation and off nominal performance due to degradation of the test article (both the SARTA and t he

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77 control/instrumentation support hardware). These parameters are used to provide acceptable bounds on SARTA performance that can be referenced to make the decision if the test can continue or must be terminated. During irradiation if operating parameters drift outside of the established limits testing is terminated to investigate for possible instrumentation failure or SARTA degradation. The SARTA operation limits are detailed in Table 41. Pressure change sensitivity The SARTA pr essure vessel leaked through the thermocouple feed throughs, requiring active control by the operator to maintain 500 psig. Therefore, it was important to quantify the expected change in SARTA performance as a function of changes in helium pressure. This was accomplished by establishing steady state conditions and varying the pressure 15 psig. Changes in performance from nominal were relatively small and indicated that the SARTA was not overly sensitive to changes in pressure at the scales under consideration. The results of the pressure change sensitivity tests are shown in Table 4-2 The accuracy ranges shown in Table 42 are based on the published experimental error for each instrumentation measurement Several hardware configuration changes took pl ace after performance characterization at NASA GRC and also before radiation testing. First, a safety review recommended that a fuse be added to the instrumentation control system for an additional safety margin. Second, the band heater originally used at NASA GRC was made of resistive heating elements encased in Silicone which would not have operated long in the radiation environment. Therefore, it was decided to use an all metal band heater that could survive to the end of the established test matrix. The metal band heater provided a similar power output when compared to the original Silicone band

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78 heater but is approximately 4 cm shorter in height (3.5 cm vs. 7.5 cm). This difference in heating area (i.e., heat flux applied to the pressure vessel wall) changed the spatial temperature distribution profile of the SARTA during all phases of operation (e.g. start up, steady state, and shut down). Third, at the radiation facility a different pressure regulator was mandated by a safety review board. The new regulator had very coarse adjustment and the operators were now forced to maintain pressure from 2 psig to 5 psig. In addition, the rate at which operator input was required to actively maintain operating pressure greatly increased. Finally, both the alternator coil and FLDT external leads, which were insulated with PTFE were replaced with more radiation resistant PVC insulated leads. These combined changes to the SARTA hardware resulted in changes in nominal operating parameters when compared to t hose parameters established during preexposure testing at NASA GRC. A discussion on how these changes affected operation and data processing is detailed later. SARTA Radiation Testing F rom May to June 2009 t he SARTA was subjected to radiation exposure te st s at the Sandia National Laboratories (SNL) Gamma Irradiation Facility (GIF) cell #2. The 155.6 kCi (as of May 20, 2009) 60Co source consists of a 20 pin array arranged in a 32.385 cm diameter circular geometry as shown in Figure 4-5 The source array rests on an elevator submerged in a pool of water that is remotely raised. Once raised the source surrounds a 19.685 cm diameter experiment basket. MCNP analysis of the GIF radiation environment suggested that that the expected dose rate at the array cent erline was on the order of 900 krad per hour with about one tenth the dose rate at one meter from the array centerline as shown in Appendix A, Figure A-2 and Figure A-350. Dose rates were then measured using CaF2 dosimeters.

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79 SARTA Dosimetry Radiation exposures were performed in steps with the alternator performance recorded during exposure by the data acquisition system. For the first exposure, 16 Thermo luminescent dosimeter (TLD) chips were assembled into arrays (four TLDs per array). TLD 400 (CaF2:Mn) was selected due to its high sensitivity to rays and moderate saturation threshold51. CaF2 dosimeters tend to approach saturation at approximately 450 krad; the refore, the first test was chosen to be 30 minute exposure at approximately one meter fr om the array centerline (low dose rate configuration) in order to accurately gauge the dose rate as shown in Figure 46a. When the SARTA is placed in a basket (high dose rate configuration) the dose rate was obtained by conducting two 5 minute long exposures as shown in Figure 46b, also to prevent saturation of the dosimeter arrays The TLD 400 arrays were positioned in various r adial and axial locations around the SARTA as shown in Figure 47a For the second through fourth tests, only two TLD 400 arrays were placed in front and two behind the pressure vessel as shown in Figure 47b The eighth and ninth tests were performed in the basket (high dose rate configuration) and had 16 and 8 dosimeters, respectively, radially surrounding the SARTA. Afte r irradiation the TLDs were removed and analyzed by the SNL Radiation Metrology Laboratory (RML) using a ThermoFisher model 5500 TLD reader. The dose and associated error was obtained for each TLD array by applying the appropriate statistical analysis to the TLD reader results By kn owing the dose and the length of exposure we determine the dose rate of each TLD52. The dose rates are used to determine the spatial dose distribution the SARTA experienced The results indicate

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80 that for the low dose configuration the dose on the front side of the SARTA pressure vessel was roughly twice that on the backside. To a first order approximation, the inter nal SARTA dose was taken as the average of several exterior TLD dose measurements with a simple calculation to account for pressure vessel shielding SARTA Electrical Integrity The electrical integrity of the SARTA wa s highly dependent on the proper function of the polymeric dielectric materials used to insulate certain internal components. Although the inert, hi gh pressure, highpurity helium environment helped to preserve the integrity of insulating materials, a nondestructive method of verifying this assumption was required. Between exposure steps the SARTA was shut down and allowed to cool to ambient temperature in order to verify the electrical insulation integrity. The insulation integrity was verified by measuring the alternator and FLDT inductance, resistance and resistance to ground. The inductance was obtained by connecting two probes across both the alternator leads located outside the pressure vessel, taking a measurement, then repeating the process across the FLDT leads. These probes were connected to a Sencore Capacitor Inductor analyzer as shown in Figure 48. Measuring resistance to ground would determine if a short of the internal components to structure was occurring. The resistanceto ground was measured using a Fluke 1520 MegaOhm Meter (Figure 49a) which provides a 250 V supply that corresponded to a particular resistance value of >1000 The coil and FLDT resistance were measured using a Fluke 189 Multi Meter as shown in Figure 49b. The preexposure values for inductance, resistance, and resistance to ground were measured at NASA GRC and verified prior to the first exposure test at SNL. The

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81 values corresponding to nominal SARTA electrical integrity are shown in Table 43. After each exposure these values were remeasured in order to supplement operational data The operational database is then used to aid in the decision making process of whether to continue with the test matrix or terminate testing operations SARTA Post Irradiation Performance Analysis Data collected from the Stirling alternator while operating in the aggressive radiation environment wa s thoroughly assessed to determine if changes in performance we re detected. These data are correlated as a function of operating time, radiation dose rate, cumulative dose and operating temperature. The performance of the Stirling alternator was also analyzed to determine if detectable degradation had occurred during operation with nondestructive evaluation methodologies Changes in component operating temperature, evolved gas, alternator efficiency, and overall power output w ere a lso compared to preirradiated val ues Performance data of a nonirradiated Stirling alternator is compared to the data from the photon irradiated Stirling alternator to determine if radiation will indeed compromise the power subsystem or if only minor changes are required to fully radiation harden the systems for spaceflight. In addition, these combined results are to be fed back into the original model in order to experimentally benchmark the code Improving design codes will provide more accurate lifetime predict ing capability of future Stirling alternator design concepts Destructive methodologies are also implemented. Of specific interest significant dimensional divergence is the most easily observed failure criteri on and in terms of practical engineering terms constitutes a mechanical component failure since components may exceed tolerance requirements53, 54.

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82 Figure 41. Stirling Alternator Radiation Test Article (SARTA) before radiation testing. Figure 42. SARTA support, control instrumentation and data acquisition equipment V V V a a a c c c u u u u u u m m m P P P u u u m m m p p p i i i n n n g g g S S S t t t a a a t t t i i i o o o n n n A A A C C C P P P o o o w w w e e e r r r S S S u u u p p p p p p l l l y y y O O O s s s c c c i i i l l l l l l o o o s s s c c c o o o p p p e e e D D D A A A Q Q Q S S S w w w i i i t t t c c c h h h Uit R R R G G G A A A Temperature Controller S S S p p p e e e a a a k k k e e e r r r Computer D D y y n n a a m m i i c c B B a a l l a a n n c c e e r r R R e e s s i i s s t t a a n n c c e e H H e e a a t t e e r r B B a a n n d d P P r r e e s s s s u u r r e e V V e e s s s s e e l l P P r r e e s s s s u u r r e e V V e e s s s s e e l l T T o o p p O O u u t t e e r r T T h h e e r r m m o o c c o o u u p p l l e e P P r r e e s s s s u u r r e e L L i i n n e e C C o o i i l l T T h h e e r r m m o o c c o o u u p p l l e e F F e e e e d d t t h h r r o o u u g g h h P P r r e e s s s s u u r r e e V V e e s s s s e e l l B B o o t t t t o o m m O O u u t t e e r r T T h h e e r r m m o o c c o o u u p p l l e e I I n n n n e e r r I I r r o o n n T T h h e e r r m m o o c c o o u u p p l l e e F F e e e e d d t t h h r r o o u u g g h h

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83 Figure 43. Thermal time constant response runs (steady state, fan turned on, no heat, no insulation) Figure 44. Stoke variation performance mapping (90C, 500 psig, 6.5 8.5 mm)

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84 Table 41. SARTA Nominal Operation Limits Temp ( C) Value Power (W) Current (A) Voltage (V) Stroke (mm) Power Factor 90 Nominal 27.03 0.74 4.652 0.08 8.207 0.10 8.002 0.03 0.708 0.02 Range 26.66 27.40 4.611 4.69 8.156 8.26 7.989 8.02 0.697 0.72 125 Nominal 27.10 0.8 4.66 0.08 8.01 0.1 8.02 0.02 0.73 0.02 Range 26.7 0 27.5 4.62 4.70 7.96 8.06 8.01 8.03 0.72 0.74 Table 42. Performance as a function of pressure Temp (C) Pressure (psig) Power (W) Current (A) Voltage (V) Stroke (mm) 485 0.01 27.027 1.35 4.806 0.12 8.317 0.208 8.020 0.876 90 2.2 500 0.01 27.190 1.36 4.828 0.12 8.316 0.208 8.006 0.874 514 0.01 27.333 1.37 4.846 0.12 8.314 0.208 7.992 0.872 485 0.01 27.220 1.36 4.847 0.12 8.143 0.204 8.012 0.875 125 2.2 500 0.01 27.373 1.37 4.848 0.12 8.143 0.204 8.010 0.875 513 0.01 27.478 1.37 4.865 0.12 8.139 0.203 7.994 0.873 Figure 45. S andia N ational L aboratories Gamma Irradiation Facility cell #2 155.6 kCi 60Co source array.

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85 A B Figure 46. A) Low dose rate configuration. B) High dose rate configuration. A B Figure 47. A) RAD 1 dosimeter placement B) Subsequent dosimeter placement S S S A A A R R R T T T A A A M M M i i i c c c r r r o o o p p p h h h o o o n n n e e e S S S A A A R R R T T T A A A i i i n n n B B B a a a s s s k k k e e e t t t S S S o o o u u u r r r c c c e e e A A A r r r r r r a a a y y y

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86 Figure 48. Sencore LC53 Z Meter Capacitor Inductor Analyzer A) B) Figure 49. A) Fluke 1520 MegaOhm Meter B) Fluke 189 Multi Meter Table 43. Pre exposure electrical values Component Resistance ( ) Inductance Resistance to Ground ( ) Alternator 0.2 0.0001 1.0 (relative nominal) > 1000 FLDT 5.1 0.0026 > 1000

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87 CHAPTER 5 APPARATUS AND PROCED URE: MIXED NEUTRON AND GAMMA RAY TESTING OF STIRLING ALTERNAT O R CANDIDATE MATERIALS Since the Stirling convertors are also under consideration for fission surface power (FSP) applications the linear alternator materials must be capable of tolerating a mixed neutron and ray radiation environment. The objective of the mixed neutron and ray exposure test phase is to evaluate the performance and characterize possibl e radiation induced changes of organic materials under consideration for use in current Stirling alternator design concepts These tests are performed by subjecting candidate material coupons to a similar FSP neutron fluence (at appropriate order of magnitude neutron energy levels) and combined neutron and ray doses while at prototypic atmospheric and temperature operating conditions. The Stirling alternator environment is a somewhat benign operating condition and will simplify data extrapolation as long as conditions are maintained. For example, in the absence of oxygen and water vapor there is evidence that cross linking is insensitive to dose rate and may be correlated to the total dose with high accuracy24. This allows the samples to be irradiated at accelerated rates in order to make practical utilization of irradiation facilities (irradiations on the order of hours instead of years). As mentioned previously, under certain conditions, equal energy absorbed yields an equal radiation effect. This rule applies only to similar irradiation conditions with a sin gle radiation type and irradiation conditions dictate the expected radiation effect. Therefore, it is incorrect to extrapolate to other conditions and more data is required to verify the theorized radiation effect. The radiation effect estimation must take into account the specific irradiation conditions such as temperature, atmosphere, pressure, radiation type and energy (e.g. dose rate), and sample thickness24.

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88 Sample Irradiation Test Articles Four ultrahigh vacuum (UHV) capable test articles fabricat ed from Al 6081 with stainless steel knifeedge seal grooves serve as sample chambers shown in Figure 5-1 O rganic material sample coupons are attached to an Al 6081 plate and arranged to receive uniform unobstructed neutron flux as shown in Figure 5-2 and Figure 5-3 The temperature of each test article is maintained by externally mounted resistance band heaters coupled to Eurotherm PID temperature controllers. The temperature of the test article was maintained at 125C or 150C 2C The temperature of the internal samples and the external test article wall wer e measured using type-k thermocouples and recorded through the use of an Agilent 34970A data acquisition/switch unit used in conjunction with Agilent Data Logger software running on a laptop computer. The instrumentation rack is shown in Figure 5-4 and the complete system schematic can be found in the A ppendix B Figure B1. T he PID temperature controller maximum output power was limited to 40% to prevent the possibility of a te mperature overshoot condition. In addition, the temperature ramp rate was manually controlled to allow for a gradual increase in the test article and sample tray temperature. Mixed Neutron and Gamma ray Testing Facility In December 2009 and January 2010 the test articles were irradiated at the Texas A&M University (TAMU) TRIGA Mark I reactor as illustrated in Figure 5 -5 The test articles were exposed in the confinement building which is a dry irradiation cell adjacent to the reactor pool Utilizing the irradiation cell as opposed to water immersion of the test articles allows for ease of access, simple test article manipulation and placement as well as the inclusion of support hardware such as instrument wires and power cables The reactor wa s coupled to the irradiation cell window as shown in Figure 5-6

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89 and the samples were irradiated to the FSP expected lifetime fluence. As discussed in Chapter 1, the FSP expected lifetime neutron fluence is approximately 6.5 x1013 n/cm2 (primarily soft spectrum); however more conservative fluence level s were established. We must consider the highly probable possibility that the reactor service life will be extended beyond the primary mission requirements as is commonly the case with many NASA missions Missions are often extended to include not only secondary but tertiary mission objectives Therefore, we increase the estimated maximum FSP lifetime fluence to 1x1014 and 5x1014 n/cm2 a nd test to these predetermined c onditions while at prototypic temperature and atmospheric conditions Radiation Environment Characterization In order to ensure that the radiation exposure tests of the candidate materials are of practical relevance to RPS and FSP projects we must accurately characterize the radiation envir onment. The approximate neutron flux, neutron energy, and dose rate was characterized before test article irradiation. The following sections detail the methodology utilized for the characterization process. Neutron Flux Spectrum Measurement The TAMU average thermal and fast neutron flux is experimentally characterized using activation threshold gold foils cadmium covered gold foils, and iron wires in an arrangement as shown in Figure 5-7 By measuring the activity of irradiated foils and knowing the foil elemental composition, sample mass, capture cross section at neutron irradiation temperature and the irradiation time, the approximate number of incident neutrons at a specific energy can be estimated12. The different foil s allow for the neutron spectrum to be estimated at discrete energies which are used to benchmark the existing MCNP models used at the TAMU NSC

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90 Dose Rate Measurement R adiachromic film dosimeters were used to characterize the spatial dose rates at the irra diation cell window as shown in Figure 5-8 The placement of the foils and dosimeters in the irradiation cell window is shown in Figure 5-9 The dosimeters were provided and processed by the ORNL RML. Initial estimates obtained from a n MCNP model were benchmarked with the experimental measurements and suggested an average total neutron flux of 1x1011 n/cm2s with a corresponding ray dose rate of 0.1 4 Mrad/min. The flux and dose data are illustrated by Tables 51, 5 2, and 5-3. Although the dose rate is rather high, the radiachromic film dosimeters have a dose range from 0.05 20 Mrad as well as dose rate independent behavior up to 1x1014 rad/s55. From the neutron spectrum and dose rate data we determine the optimal location in the irradiation cell window to place the test articles to achieve the required neutron fluence while staying within the dose limit of 1 0 Mrad. F ilm d osimeter s, foil s, and wires were arranged around the test articles as illustrated in Figure 510 The placement of two test ar ticles in the irradiation cell window is shown in Figure 511 Pre Irradiation Test Article Preparation One difficulty with neutron irradiation of organics is that samples exhibit self shielding and this may cause the spatial neutron spectrum distribution in a sample to vary24. To minimize this effect sample thicknesses are kept thin. In addition, the activation of components, particularly the metallic components may have a relatively high activity and long half life. Therefore, metallic material usage is deliberately limited. The test articles were evacuated to 1x10-3 torr using a roughing pump and vacuum baked at 1 10120 C for 1 015 minutes to remove volatiles such as water absorbed into the vessel walls Next, the test articles were allowed to cool while being back filled with

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91 ultra high purity helium and evacuated five times respectively. Finally the test articles were filled to 2 2 psig with helium in preparation for placement in the cell and subsequent irradiation. Although Stir ling alternator operation requires an operating pressure of approximately 500 psig such high pressures are not necessarily required for materials testing As long as the relative percentage of the reactive gas (e.g. oxygen and water vapor ) is the same in the cover gas, the testing can be conducted at near atmospheric pressure greatly simplifying the test article design requirements. Sample Irradiation The sample manifest is illustrated in Table 54 and test matrix with exposure conditions listed in Table 55. From ambient conditions the test articles are brought to steady state temperature for five minutes to establish preexposure base line operating conditions. Next, reactor start up wa s initiated and reactor power increased until a 1 MWth steady state power level wa s achieved. At this point the timer is activated and the test articles were irradiated to the proper predetermined irradiation time to achieve the desired neutron fluence. After the irradiation time had been met the reactor is shutdown and after several minutes ha d passed the reactor wa s de coupled from the irradiation cell and transported to the other side of the reactor coolant pool The reactor is suspended from a bridge that rides on rails on either side of the pool that allows it to be moved away from the irradiation cell to ensure a dose below 10 Mrad. The temperature profile s for irradiations NUKE 1 t o 8 are illustrated in Figure 512 through Figure 515 After irradiation the test articles decay before they are remov ed from the irradiation cell Irradiation tests NUKE 1, 2, 5, and 6 decayed anywhere from three to four days before removal from the irradiation cell Irradiation tests NUKE 3, 4, 7, and 8 required a decay period of approximately three weeks before removal from the irradiation cell

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92 Post Irradiation Sample Processing Radiation levels were measured using a handheld ionization chamber. The test article s radiation level was relatively low, with the stainless steel components showing the highest levels as expected. Surface radiation levels for NUKE 1 and 2 were 4.5 mR/h ou r 7.4 mR/h our for NUKE 3 and 4 12.7 mR/h our for NUKE 5 and 6, and 5.1 m R/h ou r for NUKE 7 and 8. After the test articles we re removed from the irradiation cell, they were pl aced in a portable glove bag back filled with argon as shown in Figure 516. The test articles were opened, the samples were removed and then the samples were bagged by group. At first glance the samples appear to be in pristine condition, showing no visible signs of degradation, other than darkening and shrinking of the Kynar heat shrink tubing. Whether the change in the Kynar is due to just thermal or thermal and radiation conditions will be determined by the materials characterization. S amples underwent ray spectroscopy using a High Purity Germanium (HPGe) detector Each sample type w as peak counted for 30 minutes in order to develop a spectr um for radioisotope species identification a nd to determine the overall activity of the samples25. A typical ray spectr oscopic abundance result for each type of irradiated material sample can be located in Appendix D, Table D1 through Table D-9. The dominating constraint for neutron testing is the time required for the activity of exposed materials to decay to acceptable levels before handl ing and shipping. In addition, highly detailed safety permits are required to evaluate samples and to prevent contamination of laboratory facilities. Fortunately, the organic samples cool relatively quickly, on the order of several weeks before samples can be transported for material characterization. The irradiated samples w ere subsequently shipped to NASA GRC in a White I ( Class VI I) radioactive material package with a silica gel desiccant.

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93 Figure 5-1 Mixed neutron and g amma ray material s ample test article Thermocouple Feed through Gas Feed Through Resistance Band Heater Outer Top Thermocouple Band Heater Power Leads Outer Bottom Thermocouple

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94 Figure 5-2. Material sample t ray The left side image A) represents the front face and the right image B) represents the back face. Figure 5-3. Material sample tray position in the test article. Arrow indicates the direction of neutron and gammaray flux emanating from the reactor. Xylan on Al Substrate Kalrez O Rings Kynar Heat Shrink Viton Heat Shrink PTFE Insulated Cu Wire Silicone O Rings Polyimide Insulated Cu Wire High Temperature Epoxy Lap Shear Hysol Epoxy Lap Shear Inner Thermocouple A A B B

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95 Figure 5-4. Instrumentation and control rack used for data collection and test article temperature environment control. Figure 5-5. TAMU TRIGA Mark I reactor coupled to the irradiation cell during operation. DAQ Switch Unit Temperature Controllers Computer Heater Power Control Switch Panels

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96 Figure 5-6. TAMU reactor coupled to irradiation cell schematic Courtesy of Dr. Latha Vasudevan, TAMU56. Figure 5-7 Gold foil, cadmium covered gold foil and iron wires for neutron spectrum characterization. Courtesy of Dr. Latha Vasudevan, TAMU56. Gold Foil Cadmium Covered Gold Foil Iron Wire Test article location Reactor

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97 Figure 5-8 TLD 400 (CaF2) and radiachromic film dosimeters for dose rate measurement. Courtesy of Dr. Latha Vasudevan, TAMU56. Figure 5-9 Dosimeter and foil/wire placement in the irradiation cell window Courtesy of Dr. Latha Vasudevan, TAMU56. Radiachormic Film Dosimeter TLD 400

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98 Figure 510. Dosimeter, gold foil, cadmium covered gold foil, and iron wire arrangement Figure 511. Fully instrumented test articles and dosimeters in the TAMU irradiation cell window

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99 Table 5-1 TAMU irradiation cell neutron spectrum measurements at 11 cm Courtesy of Dr. Latha Vasudevan, TAMU56. Data Source 5G5G (0.05 eV) (n/cm 2 s) (> 9keV) (n/cm 2 s) Model 1.03x10 10 3.12x10 11 Measured 1.04x10 10 2.90x10 11 An MCNP model of the TAMU reactor neutron spectra at the irradiation cell was utilized after the latest refuel. The measured data was obtained irradiating gold foils, cadmium covered gold foils, and iron wires. The foil or wire activity was then used to correlate the approximate neutron flux the samples were subjected to at a distance of 11 cm from the irradiation cell window, normal to the window surface. Table 5-2 TAMU irradiation cell neutron flux measurements as a function of distance. Courtesy o f Dr. Latha Vasudevan, TAMU56. Distance from window (cm) Total (n/cm 2 s) 13 2.9x10 11 55 2.94x10 10 34 1.6x10 11 34 with Factor of Safety 1.0x10 11 The measured data was obtained irradiating gold foils, cadmium covered gold foils, and iron wires. The foil or wire activity was then used to correlate the approximate neutron flux the samples were subjected to as a function of distance from the irradiation cell window (normal to the window surface). Table 5-3 TAMU irradiation cell gamma ray dose rate measurements Film Dosimeter Window Distance (cm) Film (Mrad/min) CaF2 (Mrad/min) 1A 13 0.122 0.00036 0.119 0.0115 2A 13 0.140 0.00042 3A 13 0.124 0.00037 3A 61 0.026 0.0018 4A 61 0.033 9.75E 5 5A 61 0.030 9.0E 5 Dose values were measured using radiachromic film and CaF2 thermoluminescent dosimeters. The irradiation period included reactor start up, 2 minutes at 1 MWth steady state power, reactor shutdown, and irradiation cell decay time to allow dosimeters to be remo ved for subsequent analysis. Dose values were measured as a function of distance from the irradiation cell window.

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100 Table 5-4 Stirling Alternator Candidate Organic Materials Test PTFE Polyimide Silicone Kalrez Viton Kynar Hysol HT Epoxy Xylan Number of samples per material type NUKE 1 3 3 2 2 3 3 4 4 1 NUKE 2 3 3 2 2 3 3 4 4 1 NUKE 3 3 3 2 2 3 3 4 4 1 NUKE 4 3 3 2 2 3 3 4 4 1 NUKE 5 3 3 2 2 3 3 4 4 1 NUKE 6 3 3 2 2 3 3 4 4 1 NUKE 7 3 3 2 2 3 3 4 4 1 NUKE 8 3 3 2 2 3 3 4 4 1 Table 5-5 Test Conditions Test T ( o C) t (n/cm 2 ) t Irrad (min) D Est (Mrad) D Meas (Mrad) NUKE 1 125 1x10 14 16.67 2.333 0.007 1.355 0.004 NUKE 2 150 1x10 14 16.67 2.333 0.007 1.355 0.004 NUKE 3 125 5x10 14 83.33 11.667 0.035 5.38 0 0.1076 NUKE 4 150 5x10 14 83.33 11.667 0.035 5.38 0 0.1076 NUKE 5 125 1x10 14 16.67 2.333 0.007 1.355 0.004 NUKE 6 150 1x10 14 16.67 2.333 0.007 1.355 0.004 NUKE 7 125 5x10 14 83.33 11.667 0.035 5.38 0 0.1076 NUKE 8 150 5x10 14 83.33 11.667 0.035 5.38 0 0.1076

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101 Figure 512. Nuke 1 and Nuke 2 t emperature p rofiles with respect to reactor power. Figure 513. Nuke 3 and Nuke 4 t emperature p rofiles with respect to reactor power.

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102 Figure 514. Nuke 5 and Nuke 6 t emperature p rofiles with respect to reactor power. Figure 515. Nuke 7 and Nuke 8 temperature profiles with respect to reactor power.

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103 Figure 516. Post irradiation sample removal from test articles in portable glove bag back filled with argon.

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104 CHAPTER 6 APPARATUS AND PROCED URE: POST IRRADIATION MATERIAL S CHARACTERIZATION The Stirling alternator wa s disassembled and inspected in a methodical manner in order to address hardware and material changes. The alternator post irradiation protocol first involved nondestructive evaluation of the system components and materials followed by material characterization and property evaluation. Dimension and Weight Measurement Sample averaged dimensions were measured before and after testing using a Mitutoyo Absolute Digimatic caliper with accuracy of 0.00 1 mm. The sample weight wa s also measured before and after testing using an Ohaus AS120 (at TAMU) or a Mettler AJ100 (at NASA GRC) analytic balance with an accuracy of 0.0001 g. Optical Microscopy Samples are viewed using an optical microscope coupled to a digital camera as shown in Figure 6-1 This system allows one to observe bulk topography, morphology, and discoloration while capturing the corresponding sample micrograph for subsequent comparative analysis. Optical micrographs provide a good starting point for higher magnification observation methods described below. Scanning Electron Microscopy & Energy Dispersive Spectroscopy Samples undergo characterization using a Scanning Electron Microscope (SEM) used in conjunction with an Energy Dispersive Spectrometer (EDS) that utilizes software to obtain and analyze xray spectra. SEM EDS allow for sample topography, morphology, and qualitative elemental abundance57. The complete SEM EDS system is shown in Figure 6-2. SEM use on organic samples is limited to the minimum time that is required to gather the required information due to sample altering effects29. These

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105 effects include heating, radiation induced damage, electrostatic charging, sputtering of low Z surface atom s (mass loss of hydr ogen, nitrogen, carbon, oxygen), and hydrocarbon contamination29. Of these issues, heating is of greatest concern. S ince the incident electrons transfer appreciable energy to bound electrons during inelastic scattering events, temperature spikes even at low current densities can result Localized heating is a concern for organic s since they have low thermal conductivities (0.2 2 W/m/oC ) and are susceptible to thermal degradation or even melting at moderately elevated temperatur es29. Hydrocarbon contamination within a specimen diffuses along the surface towards the electron probe. By scanning at lower magnification, surface hydrocarbons are fixed by polymerization and prevent ed diffusion towards the focused electron probe29. Rad iation induced surface roughness and imperfections often form blisters, voids, globules and cracks. Surface roughening is also indicative of no nuniform surface gas evolution30. EDS is used to detect oxidation effects by measuring the qualitative increase in oxygen content of a sample as a function of exposure. Unfortunately, EDS can not measure the expected increase in the C:H ratio because of H insensitivity30. Differential Scanning Calorimetry Differential Scanning Calorimetry ( DSC) is a technique where the difference in the quantity of heat required to change the temperature of a sample and reference sample maintained at the same temperature is measured. Discontinuities or slope changes in the heat flow curves are indicative of a physical change such as melting, crosslinking, species evaporation, phase transitions glass transitions, detection of thermodynamically stable crystals and curing state58. E ndothermic reactions indicate m elting and conversely e xothermic reactions indicate solidification.

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106 H eat flow difference curves are obtained for each of the organic samples using a TA Instruments Q1000 modulated DSC as illustrated in Figure 6-3 Samples weighing from 510 mg were heated in nitrogen gas from 90C to 370C at a ramp rate of 5C per minut e, modulated 0.5C every 40 seconds. The data files are processed using TA Universal Analysis 2000 software. Dynamic Mechanical Analysis Dynamic Mechanical Analysis (DMA) is a method where an applied sinusoidal oscillating stress (frequency and amplitude) is compared to the measured sinusoidal strain of a sample. The stress is applied from sub ambient temperatures (90C) through elevated temperature (400 C) as the stress and displacement is measured58. The DMA results provide insight on the viscoelastic nature of the polymeric sample. From the stress and deformation measurement s the modulus and Tg can be accurately derived. Tg measured with DMA is more sensitive than DSC. Initially, samples were going to undergo xray diffraction (XRD), which allows for the determination of crystal structure and qualitative phase analysis59. However, XRD could cause radiation induced chemical changes within the sample60. D ue to the limited availability of the XRD at N ASA GRC it was decided that DMA would be substituted as a n independent means for comparing DSC measured radiation induced changes. Density or specific volume measurements can also be correlated with crystallinity in polymers with DMA Crystallinity has been known to increase with dose. This phenomenon can be attributed to the scission of molecular chains in amorphous regions that yield short segments with higher mobility that can create entanglements leading to new crystalline regions or can be incorporated into preexisting crystalline regions31. The DMA assembly is shown in Figure 6-4 DMA results are used to provide

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107 independent compar i son to DSC results to help approximate the degree of crystallinity ( ) within a sample. The weight fraction crystallinity is given by E quation 6-154. = 5I5I (6 1) where: V c = crystalline phase volume fraction c = crystal phase density = sample density Crystallinity is not necessarily controlled in all the organic samples but those with known crystallinity control will be evaluated by this method. ThermoGravimetric Analysis Thermo Gravimetric Analysis (TGA) measures weight change with respect to temperature. A high precision analytical balance suspends a platinum pan containing the sample. The pan is surrounded by an electrically heated furnace with an integrated thermocouple to measure the sample temperature. The w eight change vs. temperature curves as obtained using a TA Instruments Q500 TGA are illustrated in Figure 6-5. The result is a weight change vs. temperature curve which is used in determining degradation temperatures, absorbed moisture content, and oxidation rates. TGA also provides insight to post irradiation reactions that occur as a result of uncombined free radicals in organic samples that react if the test is done in the presence of oxygen58. S amples weighing from 510 mg were heated from 25 C to 750C at a ramp rate of 10 C per minute using a nitrogen cover gas. The data files are processed using TA Universal Analysis 2000 software. This technique will help to determine degradation temperature (Td), absorbed moisture content, and oxidation rate. TGA will also provide insight to post irradiation reactions that occur as a result of uncombined free radicals in solid organic samples that will react if the test is done in the presence of oxygen.

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108 Fourier Transform Infrared Spectroscopy Fourier Transform Infrared Spectroscopy (FTIR) is a method commonly used for polymer degradation detection as well as identification of c hemical composition and compounds. The specific method used was attenuated total reflectance, which is a sampling technique for examining solid surfaces directly without significant sample preparation. A sample is pressed directly to the germanium based crystal to ensure close contact. A heated ceramic source is used to impinge infrared light through crystal and penetrates a few micrometers into the sample. The infrared beam then exits the crystal and is sampled by the detector that measures the total i nternal reflection or evanescent waves. The resulting absorption spectr um is then used to reveal qualitative changes in the polymeric samples. Samples are evaluated using a Thermo Electron Nicolet 380 FTIR as shown in Figure 66. Surface Electrical Resistivity For the electrically insulating organic materials (e.g. PTFE and Polyimide coated wires and Kynar and Viton heat shrink tubing) the electrical conductivity change is determined b y measuring the electrical resistivity Resistanceto ground measurements can determine if radiationinduced conductivity in the wire and heat shrink electrical insulation is occurring. Significant changes in conductivity can be correlated to resistanceto ground measurements taken using a Fluke 1520 MegaOhm Meter (the sam e used for the SARTA testing) and is illustrated in Figure 49a. The meter uses a 250 V supply that corresponded to a particular resistance value of either great er than 500, greater than 1000 or greater than 2000 Although coarse in nature this resistivity measurement method does allow for a qualitative analysis of polymeric electrical insulation behavior as a function of radiation exposure.

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109 ORing Compression Set Testing The Silicone and Kalrez O ring samples are subjected to compression set tests. The specific method used can be found in appendix D. The Orings undergo 25% diametric compression for a period of approximately 70 hours at room temperature. Appropriately thick stainless steel shim stock is used to ensure that the 25% deformation requirement is maintained throughout the compression process. After the compression period has passed the pressure applied to the Orings is relieved and the samples are allowed to recover for 30 minutes. The Oring diameter is then m easured using a set of calipers in order to investigate the changes in Oring compression set with radiation exposure. The irradiated sample measurements are compared to a variety of control sample measurements in order to observe statistically significant deviations from nominal compression set values The Oring arrangement on the stainless steel compression plate and the complete compression apparatus is illustrated in Figure 67. Lap Shear Testing Lap shear testing w as performed on the Hysol and H igh -T emperature Hysol lap shear samples. The samples are axially strained using an Instron uni axial load frame equipped with environmental chamber that maintains the sample temperature at 120C as illustrated in Fig ure 68. The sample is held inside the en vironmental chamber by two opposed gripper fixtures Sandpaper is used as a contact barrier separating the radioactive sample from the grip fixture, which prevents contamination exchange between the irradiated lap shear specimen and the grip fixtures. Th e purpose of these tests is to evaluate the Hysol epoxy and H igh T emperature Hysol epoxy adhesion strength and toughness as a function of radiation conditions. The samples are tested to failure so the amount of data collected is limited.

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110 Figure 61. NASA GRC Keyence Digital Optical Microscope and image acquisition unit Figure 6-2. NASA GRC Hitachi S 4700 Field Emission Scanning Electron Microscope. V V H H X X D D i i g g i i t t a a l l S S c c o o p p e e C C o o n n t t r r o o l l U U n n i i t t V V H H X X 5 5 1 1 5 5 P P r r o o f f i i l l e e M M e e a a s s u u r r e e m m e e n n t t U U n n i i t t V V H H X X 2 2 1 1 0 0 0 0 D D i i g g i i t t a a l l M M i i c c r r o o s s c c o o p p e e

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111 Figure 6-3. NASA GRC TA Instruments Q1000 D ifferential S canning C alorimeter. Figure 64. NASA GRC TA Instruments 2980 Dynamic Mechanical Analyzer

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112 Figure 6-5. NASA GRC TA Instruments Q500 T hermo Gravimetric Analyzer. Figure 66. NASA GRC Thermo Electron Nicolet 380 FTIR instrument and data acquisition system

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113 A B Figure 67. Oring compression set apparatus. A) O ring arrangement on compression plate and B) compression of the Orings using c -clamps. Figure 68. NASA GRC Instru Met Instron Uni -A xial Load Frame used for lap shear testing

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114 CHAPTER 7 RESULTS: SARTA RADIA TION EXPOSURE OPERAT ION PERFORMANCE The test matrix called for s everal irradiation steps to be used to gradually expose the SARTA to ever increased cum ulative radiation doses. Specific operating conditions for each radiation exposure conducted are summarized in Table 7-1. The first column lists the exposure step name and the last column lists the different RPS and FSP total dose milestones based on des ign requirements. RAD 1 through RAD 4 utilized dosimeters to measure the dose for the SARTA at 1 m from the source centerline, which was used to calculate the average dose rate. This average dose rate was used to estimate the dose for RAD 5 through RAD 7 and RAD 14 exposure steps. RAD 8 and RAD 9 also utilized dosimeters to measure the dose for the SARTA at the source centerline, which was used to calculate the average dose rate. This average dose rate was used to estimate the dose for RAD 10 through RAD 13 and RAD 15 through RAD 18 exposure steps. The total dose was estimated by summing the doses for each exposure step. In the low dose configuration the SARTA was rotated 180 between each exposures to minimize spatial variation. TLD measurements support ed model predictions that the dose rate was very uniform when the SARTA was placed in the basket centerline for the high dose rate configuration. After startup, the SARTA was brought to temperature and operated at steady state for five minutes before and after each radiation exposure. This was done to establish baseline operating parameters for each exposure step as shown in Figure 71 and Figure 7-2. Several changes in system hardware and operator procedures influenced differences in the nominal operating conditions. After a safety review, the board required

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115 the addition of a fuse in the instrumentation control system (for fail safe operation). In addition, modifications to the He gas supply system were required by SNL. These hardware changes resulted in operating conditions that shifted to slightly different values than those established during preexposure testing at NASA GRC. These new nominal operating conditions were found to be outside the established nominal operating limits to be used to identify failure during exposure. Therefore, new nominal conditions were required to be established at SNL prior to the first exposure. The new nominal operating parameters for operation at 90C and 125C were based on taking the average of the parameters while at temperature and pressure during brief steady state runs. The same operating limits range established at NASA GRC (by statistical analysis of the data) for steady state operation was applied to the new nominal values determined at SNL While at NASA GRC the SARTA was operated using piston stroke as the control variable. It was later determined to use input voltage as the control variable while operating at SNL due to concerns that the FLDT might fail leaving the operator without a primary control mechanism. Therefore, the data in which the stroke is the control variable is uncharacteristically uniform with voltage having larger variation. The inverse is true for the data where the input voltage was the primary means of SARTA control. Perhaps the largest influence affecting the SARTA performance was pressure fluctuation. Precise pressure control practiced at NASA GRC could not be achieved with the change to the SNL pressure regulator due to its rather coarse adjustment control. Initially the pressure transducer was located inside the test cell approximately five feet from the source and was shielded by numerous lead bricks as shown in Figure

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116 7-3. During the RAD 5 run several parameters were slowly approaching shut down limits and it was eventually discovered that the pressure transducer readings were far less ( approximately 35 psig) than what was being indicated by cross checking a mechanical pressure gauge. A pressure transducer calibrator was tied into the gas supply system and confirmed the pres sure transducer failure. After RAD 5 the failed transducer was replaced and a 25 ft flex hose was used to place the transducer outside the test cell. In addition, the SARTA leak rate coupled with the inaccurate pressure regulator caused considerable pres sure fluctuation and subsequent operation parameter fluctuation throughout the exposure tests. These slight changes in performance can be related directly to pressure changes as illustrated in Figure 7-4. Next, we analyze the SARTA performance at the higher dose rate. Figure 7-5 shows how the power has a gradually decreasing trend, which is to be expected in an accelerated life test and except for RAD 5, performance is well within the established operating limits. However, the more drastic changes in power factor such as RAD 5 can be attributed to the malfunction of the pressure transducer providing false readings. When this problem was isolated and a secondary means of pressure monitoring was utilized, and the power factor immediately returned to withi n normal acceptable limits. Downward trends in power factor (e.g. the ratio of applied to produced current and voltage from the alternator) and upward trends in stroke are evident for RAD 1 through RAD 5 exposures It is uncertain whether these changing trends are due to radiation effect s of the SARTA or due to a slow variance in operating pressure due to the gradually degrading pressure transducer. Replacing and relocating the pressure transducer eliminated pressure uncertainty but not pressure variabi lity.

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117 The loss of pressure over time required the operator to increase the pressure to the upper level of the pressure limit (505 psig), monitor it as it decreased, and again increase the pressure once it reached the lower bound (495 psig). Such coarse pr essure control resulted in the saw tooth appearance of the performance data as illustrated by the RAD 6 and RAD 7 data and cannot conclusively be attributed to the operation in radiation. When we compare the stroke of the 90C exposure tests we observe fai rly close agreement between data sets as shown in Figure 7-6 The exceptions include the end of RAD 4 and the majority of RAD 5, which show an increasing trend. As mentioned before this can be attributed to the gradual deterioration of the pressure trans ducer resulting in an under pressurization condition and subsequent characteristic stroke response. When the transducer was isolated as the source of the anomalous behavior a secondary pressure gauge was used to maintain pressure and the performance retu rned to nominal as seen at the end of RAD 5. The later tests at 90C (i.e. RAD 13), show a much more gradual pressure variability and can be attributed to the pressure regulator set point being optimized for steady state operation. Although small changes in operating parameters were observed during the 90C radiation tests (e.g. RAD 1 RAD 13), the values were within the predefined operating limits, with no distinct trends observed with the exception of those induced by pressure variation. The same power factor comparison is now examined for the SARTA operation at the 125C exposure steps as shown in Figure 7-7 By this late in the test matrix, SARTA operation was well understood and the performance is far more uniform, which is evident by the decrease in variance in the data.

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118 It must be noted that for approximately the first 15 minutes of RAD 14 the SARTA was inadvertently operated at the incorrect input voltage setting. The operator applied the voltage setting used for 90C operation as opposed to the voltage setting used for 125C operation. The error was discovered and the correct input voltage setting was applied. The change in voltage resulted in the SARTA operating parameters transitioning to within the expected nominal operating values, with variation within the established limits ( Figure 7-8 ). It must be noted that one performance anomaly was detected. The performance does show a noticeable decrease throughout the last 17 minutes of the RAD 18 exposure. Voltage and current ( Figure 7-9 ) as we ll as power and stroke ( Figure 710 ) display a coordinated and consistent decrease in performance starting at approximately 224 minutes into RAD 18. No changes in pressure or input voltage were applied by the operator, which indicate that the anomaly can be most likely attributed to changes within the SARTA or support hardware. The subtle but consistent decrease in SARTA performance is likely due to degradation of one or more of the SARTA internal components. Plots for each radiation exposure step, timeaveraged performance, and exposure step comparison can be found in Appendix A. SARTA Waveform Comparison The waveforms produced by the SARTA during operation at both 90 C and 125C were obtained in order to observe subtle changes in the alternator performance. Although the waveforms are identical when overlaid on one another the specific values are somewhat different at different operating conditions as shown in Figure 711. Waveforms were obtained before, during, and after most exposure steps and they con sistently displayed no change in shape or parameters as a function of radiation exposure, indicating normal operation was preserved as illustrated in Figure 712 and

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119 Figure 713. When comparing waveforms taken at SNL to those taken at GRC a change in phas e was observed. Waveforms obtained at SNL prior to the first exposure show a small displaced shift in phase. All subsequent waveforms also display this phase shift and can be attributed to a change in the scan trigger setpoint. Adjusting the waveforms with a trigger set point correction results in good agreement between waveforms. No changes in waveform are observable with increasing radiation dose at both 90C and 125C operating temperatures as shown in Figure s 714 and 715. SARTA Radiation Exposure Electrical Integrity Initially, no changes were detected in the post exposure electrical integrity measurements conducted between each exposure step, indicating that the insulating materials were continuing to maintain their function with increasing dose. Resistanceto ground measurements for both the alternator and the FLDT did not change throughout the entire test matrix ; however, both alternator coil resistance and inductance changed drastically when measured after final exposure (RAD 18). The alternat or measurements were repeated several times in order to rule out measurement error. Instrumentation error can also be ruled out due to the fact that changes in the alternator measurements should also have corresponding changes in FLDT measurements, which were not observed, as shown in Figure 716. The FLDT measurements show fairly uniform variation that is well within the experimental error bounds established by the statistical analysis The decrease in alternator resistance and inductance is illustrated in Figure 717. The resistance decreased from a mean of 0.182 to 0.13 ; more dramatic was the change in inductance from 1. 0 to 3.54x10-3 (normalized) as shown in Figure 718. Such

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120 a drastic change in inductance may be explained as either a failure of the insulation or a failure of the bonding materials that allowed for contact between adjacent electrical carriers. The likely culprit would be the breakdown of the alternator coil lead wire insulation or breakdown of insulation between windings becoming r esistive. The fault could be at a specific location or uniformly distributed degradation throughout the coil. Since resistance to ground measurements did not change for the SARTA internal components ( such as the alternator coil), there was no short to th e external st ructure. Measurements were taken after exposure so the fault could not be attributed to temperature effects or radiation induced conduction. The fact that the SARTA operated with no drastic changes in performance even with a significant elect rical fault and was not detected until after testing could be attributed to testing procedure. In between each exposure the basket was removed from the test stand then the SARTA was removed from the bask et in order to conduct the post exposure electrical integrity measurements. One possibility is that there was insufficient degradation of the insulating material during steady state exposure operation to initiate an electrical fault Removal of the SARTA from the basket inherently induced movement of the internal components. This movement after irradiation could have contributed to the cause of the short. Perhaps the insulation was showing embrittlement and when the SARTA was moved, some of the material could have flaked off allowing contract between adjacent wires. Therefore, we can conclude that sufficient degradation of the material occurred sometime during the RAD 18 run, or between approximately 30 Mrad and 40 Mrad. Subsequent disassembly and an account of the detailed observations of internal SARTA components are described in detail in Chapters 8 and 9.

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121 SARTA Radiation Exposure RGA RGA scans of the internal SARTA atmosphere were obtained at NASA GRC in order to provide a pre irradiation gaseous element distribution of the SARTA operating under normal conditions. The preirradiation RGA scan is used as the primary gas evolution reference as is illustrated in Figure 719. Subsequent RGA scans taken after irradiation are compared to preirradiation RGA scans to qualitatively determine the evoluti on of gases as a function of the combined temperature and dose effects as illustrated in Figure 720 and Figure 721 Postir radiation RGA s pectra d o not show increased trace amounts of CO and CO2. Based on the literature review, this type of gas evolution was expected. What is of particular interest is the increase of the CO and CO2 peak height relative to the other peaks as the dose increased with each respective exposure step. Again, this method is qualitative in nature and although the amount of evolved gas detected did decrease with additional exposures, the relative peak height of CO and CO2 did increase relative to the other detected gas peaks. The evolved gas peaks show considerable growth relative to the other peaks for exposure steps at 853.57 rad/s and operating at 125C ( Figure 722 and Figure 723). It is uncertain if there is sufficient data to say conclusively whether the outgassing is driven by dose and temperature or simply temperature driven via absorbed gas in the stainless steel pressure vessel walls volatilizing at higher temperature. As a result, the control samples should undergo similar thermal aging effects in order to compare RGA and FTIR data of irradiated and nonirradiated organic samples. Such a comparison is critical for th e confident separat ion of combined thermal and radiation effects from thermal only effect s.

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122 SARTA Radiation Exposure Leak Rate Possible changes in the O ring seal integrity were determined as a function of the pressure vessel helium leak rate. The leak rate was monitored and recorded overnight after the last exposure test of the day was completed. This was measured using both a mechanical pressure gauge and the pressure transducer. Accelerat ing helium leak rates could be indicat ive of temperature or radiation damage of the Silicone O-r ings Viton O rings, or thermocouple feedthrough failure of the thread sealant. A linear regression was applied to the leak rate data in order to estimate the isothermal leak rate after each radiation exposure step A d epiction of the regression analysis is illustrated in Figure 724. From the regression we observe that the leak rate decreased with increasing radiation dose indicating the possibility that the thermocouple feedthrough thread sealant (known to be the source of the leak prior to exposure) underwent some change such as radiation induced curing or expansion, which is detailed in Table 7-2. Subsequent tests are conducted to better understand possible O ring degradation with dose that can compromise the pressure sealing ability of Stirling alternator designs. Such Oring compression test results are thoroughly described in chapter 10. Conclusions The SARTA was successfully operated at 90 C to approximately 22 Mrad and at 125C to approximately 18 Mrad with no significant degradation of operating parameters Post irradiation electrical integrity measurements indicate that some discernable damage to the alternator did in fact occur somewhere between approximately 30 and 40 Mrad. A summary of the different conditions the SARTA was operated in is detailed in Table 7-3. The SARTA performed well throughout radiation exposure tests as confirmed by minimal change in operational parameters, no

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123 detectable changes in waveforms, and the minimal generation of radiati oninduced gases as confirmed by RGA scans. The SARTA operating parameters for all 18 exposure tests were within expected the nominal variation and show only minor change in steady state operation as a function of increasing dose. Although modestly decreasing trends in the performance were observed, it cannot be stated conclusively if this change was due to only radiation exposure or to the normal wear of the SARTA unit. To make the distinction between normal operation and radiation induced wear will require a series of subsequent trials each being disassembled at various periods throughout the expected service life. Although slight changes in performance were observed, it is unlikely that radiation exposure alone would induce such effects. The expected design life for the SARTA was on the order of 100 hours and as a result was not subject to the considerable quality control processes of a typical AS C system. With the exception of the end of RAD 18, the variation in performance data was found to be a function of support hardware and operator input. Decrease in performance observed throughout the last 17 minutes of RAD 18 also coincides well with the electrical fault detected in subsequent electrical integrity measurements. Disassembly and internal inspection of the components followed by a thorough materials analysis should not only provide insight into how the components and materials withstood the exposure tests but also help to isolate the electrical fault. It cannot be conclusively stated that the electrical fault is attributable to only radiation effects. It must also be mentioned that the SARTA did operate at cumulative doses well above FSP design requirements.

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124 T able 71. SARTA radiation test matrix Test T (C) (rad/s) t (min) DEst (Mrad) DMeas (Mrad) DTotal (Mrad) RAD 1 90 81.822 30 0.147 0.014 0.147 0.014 RAD 2 90 75.943 30 0.137 0.013 0.284 0.029 RAD 3 90 98.745 60 0.355 0.036 0.639 0.065 RAD 4 90 94.38 60 0.339 0.035 0.979 0.1 RAD 5 90 78.883 210 0.994 0.095 1.973 0.201 RAD 6 90 78.883 160 0.757 0.072 2.730 0.278 RAD 7 90 78.883 270 1.278 0.122 4.008 0.408 RAD 8 90 859.396 5 0.258 0.025 4.266 0.434 RAD 9 90 847.75 5 0.254 0.025 4.52 0.46 RAD10 90 853.573 35 1.793 0.175 6.313 0.643 RAD 11 90 853.573 40 2.049 0.2 8.362 0.851 RAD 12 90 853.573 80 4.097 0.4 12.459 1.268 RAD 13 90 853.573 195 9.987 0.973 22.446 2.285 RAD 14 125 78.883 60 0.284 0.028 22.729 2.314 RAD 15 125 853.573 20 1.024 0.1 23.754 2.418 RAD 16 125 853.573 40 2.049 0.2 25.802 2.627 RAD 17 125 853.573 80 4.097 0.4 29.899 3.044 RAD 18 125 853.573 200 10.243 0.998 40.142 4.086

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125 Figure 7-1. Typical SARTA radiation test profile (power, current, voltage, and stroke) Figure 7-2. Typical SARTA radiation test profile (temperature and pressure) 7.9 7.92 7.94 7.96 7.98 8 8.02 8.04 8.06 8.08 8.1 0 5 10 15 20 25 30 0 5 10 15 20 25 30 35 40 45 50 55 60 65 70 75 80 85 Stroke (mm) Power (W), Current (A), Voltage (VAC) Time (minutes) Power Current Voltage Stroke Source up t = 42.5 min Source down t = 72.5 min Start up t = 6.16 min Shut down t = 82.03 min 480 485 490 495 500 505 510 20 30 40 50 60 70 80 90 100 110 0 5 10 15 20 25 30 35 40 45 50 55 60 65 70 75 80 85 Pressure (psig) Temperature (C) Time (minutes) Pressure Vessel Bottom Temp Pressure Vessel Top Temp Inner Iron Temp Coil Temp Pressure SARTA Start up t = 6.16 min Source up t = 42.5 min Source down t = 72.5 min SARTA Shut down t = 82.03 min

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126 Figure 73. Lead sheet and blocks used to shield the SARTA pressure transducer. Figur e 74. Influence of varying operating pressure on SARTA performance Time (min) 0 50 100 150 200 Power (W) 26.4 26.6 26.8 27.0 27.2 27.4 27.6 27.8 Pressure (psig) 470 475 480 485 490 495 500 505 510 Nominal Power RAD 3 Power RAD 3 Pressure RAD 13 Power RAD 13 Pressure

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127 Figure 75. SARTA p ower factor comparison for 90C operating temperature runs Time (minutes) 0 50 100 150 200 Power Factor 0.702 0.704 0.706 0.708 0.710 0.712 Nom. PF RAD 1 RAD 2 RAD 3 RAD 4 RAD 5 RAD 6 RAD 7 RAD 8 RAD 9 RAD 10 RAD 11 RAD 12 RAD 13

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128 Figure 76. SARTA s troke vs. time for all tests conducted at 90C operating temperature. Time (min) 0 50 100 150 200 250 Stroke (mm) 7.90 7.95 8.00 8.05 8.10 8.15 Nom. Stroke 90 C RAD 2 RAD 3 RAD 1 RAD 4 RAD 5 RAD 6 RAD 7 RAD 8 RAD 9 RAD 10 RAD 11 RAD 12 RAD 13

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129 Figure 7-7 S ARTA p ower factor comparison for tests conducted at 125C operating temperature. Time (minutes) 0 50 100 150 200 Power Factor 0.724 0.726 0.728 0.730 RAD 14 RAD 15 RAD 16 RAD 17 RAD 18 Nom. PF

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130 Figure 78. S ARTA s troke vs. time for runs at 125C operating temperature. Time (minutes) 0 50 100 150 200 Stroke (mm) 7.98 8.00 8.02 8.04 8.06 Nom Stroke RAD 14 RAD 15 RAD 16 RAD 17 RAD 18

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131 Figure 79. End of SARTA RAD 18 exposure run voltage and current decrease Figure 710. End of SARTA RAD 18 exposure run power and stroke decrease.

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132 Figure 711. SARTA p re exposure oscilloscope waveforms at 90C and 125C. Figure 712. RAD 12 waveform (90C, 80 minute exposure, 853.57 rad/s, 4.097 Mrad) Time (ms) -0.02-0.010.000.010.020.03 Displacement (mm) -6 -4 -2 0 2 4 6 90 o C 125 o C T = 90 C Period = 16.61 16.72 ms Frequency = 59.83 60.30 Hz Amplitude = 7.32 7.36 V Peak to Peak = 7.64 7.72 V RMS = 2.62 V T = 125 C Period = 16.60 16.72 ms Frequency = 59.75 60.30 Hz Amplitude = 7.68 8.92 V Peak to Peak = 7.24 7.32 V RMS = 2.62 V Time (ms) 0.00 0.01 0.02 0.03 0.04 Displacement (mm) -4 -2 0 2 4 Pre-Exposure Exposure Post-Exposure

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133 Figure 713. RAD 16 wav eform (125C, 40 mi n. exposure, 853.57 rad/s, 2.049 Mrad) Figure 714. SARTA p re exposure and post exposure waveform comparison at 90C. Time (ms) 0.00 0.01 0.02 0.03 0.04 Displacement (mm) -4 -2 0 2 4 Pre-Exposure Exposure Post-Exposure Time (ms) 0.00 0.01 0.02 0.03 0.04 Displacement (mm) -4 -2 0 2 4 Pre-Exposure RAD 2 RAD 5 RAD 8 RAD 12

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134 Figure 715. SARTA p re exposure and post exposure waveform comparison at 125 C. Time (ms) 0.00 0.01 0.02 0.03 0.04 Displacement (mm) -4 -2 0 2 4 Pre-Exposure RAD 16 Post

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135 Figure 716. SARTA FLDT electrical integrity as a function of increasing dose Figure 717. SARTA outer stator electrical integrity as a function of increasing dose. 150 151 152 153 154 155 156 157 158 159 5 5.1 5.2 5.3 5.4 5.5 5.6 5.7 5.8 5.9 6 0 5 10 15 20 25 30 35 40 Inductance ( H) Resistance ( ) Dose (Mrad) Resistance Inductance 157 3.14 H = 0.352 5.177 0.0026 = 1.026 0.60 0.70 0.80 0.90 1.00 1.10 1.20 0.12 0.14 0.16 0.18 0.2 0.22 0.24 0.26 0 5 10 15 20 25 30 35 40 Inductance (Relative Nominal) Resistance ( ) Dose (Mrad) Resistance Inductance 0.182 0.000091 = 0.0946 3.54e3 Relative 0.13 1.00 0.02 Relative Nominal = 0.0147 Calc f(T)

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136 Figure 718. Close up view of the SARTA outer stator inductance as a function of dose. Figure 719. P re -e xposure RGA scan of the SARTA internal helium working fluid. 0.95 0.96 0.97 0.98 0.99 1.00 1.01 1.02 1.03 1.04 1.05 0 5 10 15 20 25 30 35 40 Inductance (Realative Nominal) Dose (Mrad) 1.00 0.02 Relative Nominal = 0.0147 3.54e3 Relative Nominal 0.00E+00 1.00E 06 2.00E 06 3.00E 06 4.00E 06 5.00E 06 6.00E 06 7.00E 06 0 5 10 15 20 25 30 35 40 45 50 55 60 65 70 75 80 85 90 Pressure (Torr) m/Z H 2 H He CO, N2O2CO2Ar F O N OH, NH3

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137 Figure 720. RAD 1 RGA scan (90 C, 30 minute exposure, 81.82 rad/s, 0.147 Mrad) Figure 721. RAD 13 RGA scan (90C, 195 minutes, 853.57 rad/s, 9.987 Mrad) 0.00E+00 5.00E 09 1.00E 08 1.50E 08 2.00E 08 2.50E 08 3.00E 08 0 5 10 15 20 25 30 35 40 45 50 55 60 65 70 75 80 85 90 Pressure (Torr) m/Z H He N H2O F CO, N2O2Ar CO2 0.00E+00 2.00E 09 4.00E 09 6.00E 09 8.00E 09 1.00E 08 0 5 10 15 20 25 30 35 40 45 50 55 60 65 70 75 80 85 90 Pressure (Torr) m/Z H H2H e CO2CO, N2F N O CH2Ar C

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138 Figure 722. RAD 16 RGA scan (125 C, 40 minutes, 853.57 rad/s, 2.049 Mrad) Figure 723. RAD 1 8 RGA scan (125 C, 200 minutes, 853.57 rad/s, 10.243 Mrad) 0.00E+00 5.00E 09 1.00E 08 1.50E 08 2.00E 08 0 5 10 15 20 25 30 35 40 45 50 55 60 65 70 75 80 85 90 Pressure (Torr) m/Z H H2He CO, N2CO2Ar H2O O F N CH3C BCl3Chlorinated Hydrocarbons 0.00E+00 5.00E 09 1.00E 08 1.50E 08 2.00E 08 2.50E 08 0 5 10 15 20 25 30 35 40 45 50 55 60 65 70 75 80 85 90 Pressure (Torr) m/Z H He H2CO, N2CO2BCl3C N CH3O F Ar Chlorinated Hydrocarbons

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139 Figure 724. SARTA post RAD 1 exposure helium leak rate. Linear fit used to estimate time to reach a prede termined pressure and compare to actual pressure. Table 72. SARTA leak rate vs. dose Test Name Dose (Mrad) P/ t (psig/hour) RAD 1 0.147 0.140 9.770 0.01 RAD 2 0.284 0.290 7.401 0.01 RAD 6 2.730 0.278 2.228 0.01 RAD 10 6.313 0.643 1.440 0.01 RAD 18 40.142 4.086 2.039 0.01 Table 73. SARTA Operational Totals Test Operation Condition Pre Exposure Operation 64.0 Hours Exposure Operation 26.37 Hours Total Operation 102.094 Hours Operating at 90C 54.654 Hours Operating at 125C 15.877 Hours Operation at 1 m from source centerline 4.301 0.418 Mrad (Si) Operating at source centerline 35.921 3.50 Mrad (Si) y = 9.7727x + 490.73 R = 0.9932 300 350 400 450 500 550 0 2 4 6 8 10 12 14 Pressure (psig) Time (hours)

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140 CHAPTER 8 RESULTS: POST IRRADIATION EVALUATI ON OF THE SARTA Following ray testing the SARTA was heat sealed in three bags, each back filled with argon. The SARTA was transported from SNL to GRC with an argon cover gas and stored in a glove box upon arrival at GRC as shown in Figure 81. The inert atmosphere minimizes potential reactions between oxygen and possible radiationinduced free radicals within the organics that can lead to post irradiation aging effects. The purpose of the disassembly wa s to thoroughly inspect the SARTAs internal components for changes that could explain the variation in operating parameters observed at SNL. These varying parameters include a considerable decrease in the helium leak rate measured after each exposure step as well as the significant decrease in the SARTA resistance and inductance. In addition, evidence of radiationinduced degradation of the internal components could be further investigated by collecting samples and performing a thorough materials characterization analysis. RGA scans of the glove box environment before and after opening the SARTA found only trace quantities of water vapor and oxygen at the resolution available. Trace amounts of CO2, CO and light molecular weight hydrocarbons were detected within the SARTA. No additional reactive species were identified as shown in Figure 82. The RGA scans also showed a slightly higher quantity of residual CO2 when compared to scans obtained immediately after the completion of irradiation RAD 18 as illustrated in Figure 83. The m inimal presence of water vapor and oxyg en indicate that the SARTA maintained an inert internal atmosphere during transport and storage. V erification of a contaminant free cover gas allowed for the disassembly and inspection of the SARTA internal components without concerns over post irradiatio n aging effects

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141 The SARTA was disassembled inside the glove box by Aaron Courtney of Sunpower, Inc. Mr. Courtneys considerable experience in assembly and disassembly of Stirling power convertors provided a wealth of knowledge to identify even minor changes that could possibly be attributed to the radiation exposure. Sunpower also provided a magnet can, inner stator assembly outer stator assembly and piston assembly as control references used for comparison with the SARTA components. The SARTA internal assembly was free of debris, deposits, condensation products, or major changes in appearance. The pressure vessel inner surface retained a lustrous finish, suggesting minimal deposition of radiationinduced gaseous products. The pressure vessel, transition plate, and displacer spring appeared unchanged. The pressure vessel Silicone Oring suffered considerable compression set, as evidenced by a flat appearance flush with the flange and permanently deformed as shown in Figure 8-4 The compression set in this Oring would not increase the sealing ability of the pressure vessel and cannot account for the reduced leak rate discussed previously. The compression set in the O ring could have increased the leak rate, meaning that the o ther possible leak sites improved much more than expected. The Viton heat shrink tubing at both the outer stator and FLDT lead wire terminal connection points appeared to be undamaged and in very good condition as shown in Figure 8-5 The Viton heat shrink tubing cut cleanly and r etained considerable ductility The thermocouples, PTFE insulated FLDT hook up wires, and PTFE insulated alternator hook up wires were cut in order to separate the outer stator, inner stator, magnet can, and piston assemblies from the transition plate. The hook up wires appeared to be in good shape except for a roughly 3 mm long crack along the length of

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142 one of the outer stator PTFE insulated wires as shown in Figure 8-6. Small cracks were also observed along the outer stator hookup wires length as well The hook up wires ran from both the outer stator and FLDT coils to their respective feed through pins. Viton heat shrink covered the connection ends of the FLDT hook up wires. Kynar heat shrink covered the connection ends of the outer stator hook up wires. After completion of the visual inspection, the shrink tube insulation was removed from the outer stator and FLDT leads at the point of connection with the transition plate feedthrough pins. The Viton heat shrink on the FLDT pins appeared to be in good conditions ; also it cut cleanly, and retained ductil ity Kynar heat shrink tubing was used to cover the connection terminals of the outer stator hook up wires. The Kynar was brittle when removed from outer stator leads as shown in Figure 8-7. Kynar is known to be b rittle if overheated during the shrink process or during operation. Whether the brittle behavior of Kynar was exacerbated by radiation requires additional investigation using analytical characterization methods The outer stator appeared to be in good shape; although the Hysol epoxy had taken on a green tint, which is typical of temperature aging effects. Fi gure 88 compares the control outer stator (left) and the SARTA outer stator (right). T he Hysol shown between windings also had a slightly green tint as mentioned previously and is shown in Figure 8-9 One noted design difference between the SARTA and control outer stator assemblies was that the hook up wire connections for the SARTA outer stator h ave greater separation than the control outer stator. The proximity between the adjacent hook up wire s could make the control outer stator less radiation tolerant especially if PTFE insulated wires were used and lost integrity during to operation.

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143 Discoloration and stains were observed comparing Nomex paper located on the outer stator as shown in Figure 810. The source of this discoloration is uncertain, but it is highly unlikely that the stain could have induced any significant degradation in performance. It was decided that this change did not pose a risk and did not warrant additional analysis to determine if this is a temperature and/or radiation effect When removed from the assembly the magnet can had a normal feel, with no more or less force required to remove it. Scour marks on the external face of the magnet were normal and can be attributed to a grind process of the external surface of the magnets i n order to allow for normal final assembly fit. The inner stator was removed from the transition flange and showed no noticeable changes in condition with the exception of a residue deposit on attachment flange and transition plate as shown in Figure 811 and Figure 812. The same residue was also found on the transition plate where the flange bolts into place as shown in Figure 813. It was suggested that the residue could be Hysol epoxy that leaked out from the threads o f the inner stator attachment flange. The residue underwent FTIR analy sis in order to determine its content and possible origin. The FTIR results can be found in Appendix C. Like the Hysol epoxy in the outer stator, the Hysol visible between the magnets had a slightly green tint. The inner Viton O ring retained its shape and was in good condition. The piston, w hen removed had a normal appearance, with no discernable changes in the color, dimensions, or feel of the Xylan coating. There w ere rub marks at two locations on the piston Xylan coating which extended on a section of the piston end with corresponding marks on the cylinder ( Figure 814 ) More rub marks w ere fou nd on the back end of the piston and corresponded to marks on the cylinder as shown in

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144 Figure 815. T he number and size of the rub marks observed was normal and not in any way excessive for an alternator with over 100 hours of operation. Significant degradation of external wire insulation (alternator, FLDT, band heater power) irradiated in air was observed as shown in Figure 816 The PTFE insulation suffered significant embrittlement and cracking. Insulation flaked off easily with even delicate manipulations. The insulation embrittlement was not immediately apparent after exposure and therefore can likely be attributed to post irradiation aging affects. Electrical Integrity Measurements Resistance and inductance measurements were performed on the feedthrough pins, outside and inside the transition plate, and on the outer stator terminals. E lectrical int egrity parameter s were measured throughout the disassembly process and are compared to those measured during radiation testing ( Table 8-1). The inductance values have been normalized to the nominal operation value for intellectual property re striction s, but the relative change shows the magnitude of variation. The Viton heat shrink was removed from the outer stator terminal connection points A second set of inductance and resistance measurements were taken during disassembly by connecting one probe to the end of a lead wire and the other to the outer stator assembly body which resulted in a measurement of zero resistance. These measurements were repeated to the outer stator lead wire terminals and also resulted in no measureable resistance. These m easurements suggest that the anomalous inductance and resistance behavior of the outer stator assembly are not due to PTFE lead wire insulation failure, but rather a short that is probably internal to the outer stator assembly A ttempts to determine the exact location by invasive exploration of the polyimide wire insulation w ould create additional damage that would mask the location of the short.

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145 Changes in Pressure Leak Rate Potential locations for the helium leak include the pressure vessel Silicone Ori ng, two thermocouple feed throughs, and a gas feedthrough line As previously discussed the compression set in the pressure vessel Oring would not improve sealing and can therefore be ruled out The thermocouple feed throughs were threaded and sealed with an organic thread sealant The thermocouple feedthroughs were known to leak prior to radiation exposure testing and are the most likely source of the pressure leak. The feedthroughs were inspected but no major changes in the feed through or th read sealant could be observed. T he most likely cause for the decreased leak rate is due to some sort of final cure that resulted from radiation exposure and/or heat treatments throughout operation. The cure likely allowed for the thread sealant to somehow expand into the threads, which improved helium gas retention. Unfortunately, there was not sufficient thread sealant material retrieved from the SARTA thermocouple feedthroughs to allow for a thorough material property analysis. Future radiation exposure testing that requires the use of a feedthrough should utilize a welded gasket design to ensure retention of the internal testing atmosphere. These weldable gasket designs are very common in ultrahigh vacuum and ultrahigh purity gas technologies. Although these feedthroughs are somewhat more ex pensive they are far superior to the threaded type feedthroughs combined with organic thread sealant s. Conclusions The evaluation of the SARTA internal components revealed minimal degradation even after operating in the radiation environment to over 40 Mrad. A c rack w as observed in one of the PTFE hook up wire s insulation but did not appear to contribute

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146 to changes in system response. Slight appearance changes such as discoloration or rub marks were observed; however, the magnitude of these changes was comparable to those found in normal operation of un irradiated Stirling alternators T he SARTA performed well throughout radiation exposure tests as confirmed by the internal inspection. The lack of debris, deposits, and major changes in component appearance when compared to control specimens indicate that few changes will be required to radiation harden Stirling alternator designs. The only recommendation to improve radiation tolerance is to replace PTFE insulated wires with PVC or Polyimide insulated wires. The inspection also confirmed the importance of minimizing exposure to oxygen during and after radiation tests to minimize the effects of radiationinduced aging of organic materials. These theorized aging effects were clearly confirmed by the d egradation of the external leads, which suff ered severe cracks and flaking. O f critical importance was isolating the general electrical fault location. Initially, it was thought that the short was the result of contact between frayed PTFE insulated lead wires. However, after resistance and inductance measurements were taken of the outer stator this assumption was no longer valid. Evaluations suggest that the short formed in the outer stator windings, which is unexpected since the windings utilize polyimide insulation with a high radiation tolerance. Therefore, it can be suggested that such a failure is most likely attributed to therm omechanical fatigue or a pre existing flaw in the insulation that was exacerbated throughout radiation exposure tests. It cannot be stated conclusively that the failure was due to radiation. It is recommended that further thermo mechanical testing and possibly additional radiation testing of the outer stator assembly be conducted in order to repeat the fault and isolate the failure mechanism.

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147 Figure 81. NASA GRC SARTA disassembly g love box with argon backfill Figure 82. RGA scan of the glove box atmosphere prior to SARTA disassembly. 0.00E+00 2.00E 07 4.00E 07 6.00E 07 8.00E 07 1.00E 06 1.20E 06 1.40E 06 0 5 10 15 20 25 30 35 40 45 50 55 60 65 70 75 80 85 90 Pressure (Torr) m/Z Ar H2O Ar+H OH, NH3CO, N2 P P u u m m p p i i n n g g S S t t a a t t i i o o n n R R G G A A

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148 Figure 83. SARTA internal atmosphere RGA scan prior to disassembly. Figure 84. SARTA p ressure vessel Silicone Oring wi th significant compression set. 0.00E+00 2.00E 07 4.00E 07 6.00E 07 8.00E 07 1.00E 06 1.20E 06 1.40E 06 0 5 10 15 20 25 30 35 40 45 50 55 60 65 70 75 80 85 90 Pressure (Torr) m/Z Ar H2O Ar+H CO, N2OH, NH3O2

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149 Figure 85. SARTA FLDT Viton heat shrink tubing at the transition plate terminals. Figur e 86. Crack along one of the outer stator PTFE insulated wire s near the transition plate terminals Figure 87. Kynar heat shrink tubing embrittlement at the outer stator wire terminals. Figure 88. O uter stator section comparison of A) Control vs. B) SARTA A A B B

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150 Figure 89. O uter stator Hysol epoxy section comparison of A) Control vs. B) SARTA. Figure 810. O uter stator Nomex section comparison of A) Control vs. B) SARTA. Figure 811. Front end inner stator section comparison of A) Control vs. B) SARTA A A B B A A B B A A B B

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151 Figure 812. Back end inner stator section comparison of A) Control vs. B) SARTA Figure 813. R esidue deposit on the inner t ransition plate surface. Figure 814. Rub marks on the SARTA piston outer running surface Xylan coating A A B B

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152 Figure 815. SARTA I nner stator rub marks corresponding to piston rub marks. Figure 816. External PTFE wire insulation post irradiation aging embrittlement. Table 81. SARTA electrical integrity comparisons Measurement (Location) Resistance ( ) Inductance ALT FLDT ALT (Relative Nominal) Pre Irrad. Nominal Operation 0.17 5.07 1.0 156 Post Rad 18 (SNL) 0.13 5.14 3.54 x10 3 157 Pr e Disassembly (GRC) 0.127 5.12 1.0 157 Inside transition plate (GRC) 0.146 5.028 1.98e 1 N/A Outer stator terminals (GRC) 0.139 N/A 1.98e 1 N/A

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153 CHAPTER 9 RESULTS: SARTA POST IRRADIATION MATERIAL COUPON EVALUATION After disassembly and inspection the SARTA had samples removed from it in order to first undergo non destructive evaluation followed by material and property characterization. Samples were prepared us ing industry standard methods60. Dimension and Weight Measurement Sample dimensions were obtained by taking the average of three measurements using calipers and weight was measured using an analytic balance as previously detailed (Appendix C, Table C1) As received measurements were not obtained before radiation testing by Sunpower ; t herefore, t he as re ceived dimension values published by the manufacturer were used as control reference values Optical Microscopy The cracked PTFE hook up wire w as examined as shown in Figure 9-1. The micrographs are used to observe shear or fracture patterns of the PTFE insulation as shown in Figure 92 and Figure 9-3 Without magnification, the crack appears to be short and linear. T he PTFE flakes that came off from the external PTFE insulated wires were elliptical in shape. In addition to the 2D images, the Keyence microscope has the capability of rendering 3D composite images. The control station software works in conjunction with the profile measurement unit that allows the system to obtain a series of images at various ch anges of focus as a function of height and compiles the focused pixel sections into a 3D image. Using the microscopes 3D composite image c apability allows us to observe a well defined crack at one end of the wire leading to a more subtle spiral crack tha t propagates along the longitudinal axis as illustrated in Figure 94.

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154 Re orienting the composite image shown above in 3D space allows us to observe how a crack originates at the highly visible crack where the underlying copper wire braid is visible and wraps around the wire surface spirally as illustrated by Figure 95. Another composite 3D image was obtained at the open end of the wire where the crack is cle arly visible. The image plainly shows the underlying copper braid as shown in Figure 96. Further examination reveals these spiral cracks propagate in opposite directions and intersect creating elliptical sections of insulation that flake off. Cracks seem to form along pre existing features that form during the manufacturing process. Differential Scanning Calorimetry The following comparisons were performed to look for evidence of chemical changes brought about by irradiation such as a shift in the polymer sample Tg, which refers to a shift from glassy behavior to rubbery behavior and is associated with relaxation or breaking sidechain bonds. In the following figures, green curves represent as received (AR) samples, blue represents bakedout (BO) samples, and red represents post irradiation samples taken from SARTA components. The baked-o ut thermal exposures represent the baking out process used on the entire SARTA assembly prior to radiation testing. This baking out process does not replicate the entire thermal history experienced by the SARTA components. The Kynar heat shrink material response is shown in Figure 97. The vertical off set in as received curves appears large only due to the choice of Yaxis scale and is probably indicative of the natural variation in sample density. The as received and bakedout curves show good agreement in peak identification with endothermic reactions occurring around 4050 C and 127 C. Temperature variation on the order of C is within the experimental error of the DSC hardware; however, we do see

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155 differences in excess of this value, indicating that thermally induced changes have taken place. The two SARTA curves show excellent agreement with each other, but peaks differ greatly from those identified in as received and bakedout samples. In addition to radiation exposure, the SARTA samples had a different heat treatment history due to the heat shrinking process itself. In practice, applying the heat shrink material over a joint was performed using a heat gun and the application temperatures were not well controlled. Therefore, it is uncertain to what extent the identified changes w ere due to heat shrink process thermal history versus radiation exposure. Variations in m anufactur ing could account for changes between control specimens obtained directly from the vendor and the material used on the SARTA. Viton heat shrink response curves are shown in Figure 9-8. Again the as received and bakedout samples were well behaved and identif ied a Tg of approximately 20C and another endothermic reaction at 155C The SARTA curves show a slight, but noticeable decrease of 5C Further investigation is required to confirm whether this change is due to irradiation or difference in thermal his tory M inimal observable change supports the higher thermal stability of the Viton heat shrink. V ariation in instrument calibration is a potentially problematic issue when comparing samples evaluated different times. For this reason samples are tested as close to each other as possible. Sample results from Viton Oring material are shown in Figure 9-9. In this case, there are marked changes between the as received and the bakeout sample s indicative of a thermal history response. T here is an exothermic reaction in the as received material, possibly advancement in cross linking that is completed sometime after heat treatment. The re was no difference in bakedout or SARTA sample curves

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156 T he Silicone Oring curves are illustrated in Figure 910. The endoth ermic peaks around 42C for as received and bakedout samples and 54 C for SARTA exposed samples are probably too low to be associated with melting or evaporation of additives but are at temperatures well above the anticipated glass transition. The exot hermic peaks around 300C for the as received and bakedout samples are not typical The PTFE wire insulation curve is depicted in Figure 911. Unlike the previous DSC plots, a bakedout sample was not available for comparison. Although the wire insulati on experienced embrittlement there wa s little difference between the as received and SARTA curves. We observe a slight increase in endothermic peaks occurring between 14 C and 17C. The sharp peak at 327C is the melting reaction. Finally, the comparison between as received and SARTA exposed Xylan samples are shown in Figure 912. A representative bakedout sample was not available. As discussed in Chapter 3, Xylan lubricant consists of a polyimide matrix dispersed with PTFE filler. The PTFE particles act as the primary lubricating agent and should display similar curve trends as for the PTFE wire insulation curve. The Xylan matrix is 100% polyimide after processing and will not show any transitions at these temperature ranges due to the thermal stability of polyimide. Comparing the as received sample curve with the irradiated SARTA sample curves yielded good agreement of an endothermic peak at 320C and is probably associated with the melting of the PTFE filler within the Xylan. C hanges in intermediate heat flow response are probably associated with thermal history but would require further investigation. Analysis of the DSC curves for the different polymeric sample conditions reveal changes that typical ly tracked with thermal history and that more samples are needed to improve results.

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157 ThermoGravimetric Analysis Weight change s as a function of temperature were obtained. Resulting weight loss curves provided insight to the thermal degradation behavior and post irradiation reactions that result from un combined free radicals reacting with oxygen. The first knee in the weight loss curve represents the onset of thermal degradation (Td) and the second knee represents completion of degradation. Once again, green curves represent as received samples, blue represents bakedout samples, and red represents post irradiation samples taken from SARTA components. TGA curves for the Kynar heat shrink materials a re shown in Figure 913. There is good repeatability between curves of similar condition and good agreement for the as received and bakedout curves. However, Td occurs at a much higher temperature for SARTA post irradiation samples suggesting advancement in cure state. Previous evaluations of Kynar heat shrink material have found that the heat shrink process temperature affects both the degradation temperature and the flexibility62. The Kynar did appear embrittled when pos tSARTA testing samples were obtained; even delicate manipulation of the heat shrink induced cracking. It is plausi ble that the high shrink temperature influenced an increased Td. Kynar is typically radiation crosslinked by the manufacturer in order to obtain desirable properties; therefore, differentiating between the temperature and additional radiation effects on an increasing Td is difficult. Viton heat shrink curves are shown in Figure 914. In this case Td was decreased by bake out and decreased further in SARTA testing. Kynar and Viton are fluoroelastomers ( fluorine, hydrogen, and carbon contained in the backbone) and we should not observe any major degradation at these conditions. Therefore, differences in thermal histories are most l ikely responsible for changes

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158 The Viton Oring curves are shown in Figure 915. As was also observed in the DSC results the bakedout and SARTA specimen demonstrated good agreement. A noticeable shift in behavior occurred between as received and baked out samples. Figure 916 shows Silicone Oring response and it can be seen that as received and bakedout curve s show minor variation. The SARTA curves show more scatter when compared to the as received and baked out sample curves. In particular, the onset Td increased slightly and the total weight loss was larger compared to control samples. The SARTA samples lost nearly 10% more weight than the control as received and bakedout samples, which is considerable when compared to the lack of scatter between the two as received and two bakedout curves. The TGA curve comparison for the PTFE wire insulation as received and taken from the SARTA is shown in Figure 917. Baked out control samples were not available for comparison. The TGA curves show excellent repeatability. The onset of weight loss begins earlier with the irradiated samples and may be due to the variation in thermal histories. This result is unexpected since the SARTA insulation showed considerable cracking and flaking when compared to the AR and control samples. F igure 918 shows that the onset of thermal degradation in the Xylan samples was comparable i n the as received and SARTA exposed conditions. A bakedout Xylan sample was not available for testing. The data plots closely match those of the PTFE plots ; h owever the total weight loss was noticeably different. T GA of the polymer samples revealed only minor changes in Td with the exception of the heat shrink materials. Due to the sensitivity of Td to thermal history, these results do not conclusively demonstrate radiation degradation or lack thereof.

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159 Conclusions Chemical changes characterized by DSC and TGA testing found some differences between polymeric bakedout samples and corresponding samples taken from the irradiated SARTA components However the SARTA did experience extended operation at elevated temperatures so these results can only serve as guidance for future testing. The lessons learned from this phase of testing are used to design the sample preparation and sample characterization methodology implemented in the subsequent mixed neutron and ray testing of candidate material coupons. PTFE wire insulation and Kynar heat shrink tubing were the two organic materials which experienced the most notable changes and are consequently identified as candidate materials which will warrant close r scrutiny in subsequent tests. Improved t hermal history control of the coupon samples and follow on characterization should reveal the nature of the observable deterioration of Kynar and PTFE. The operating performance changed insignificantly through 18 different irradiation runs amassing approximately 40 Mrad of exposure. Post operation measurements on the alternator suggested a short circuit occurred when the SARTA was removed from the test c ell. Further e valuation suggests that the short formed in the stator coil windings. Due to the nature of its construction, destructi ve evaluation of the coil was not practical or feasible. Since the outer stator windings utilize a polyimide based electrical insulation coating a high radiation tolerance would be expected from this component It is recommended that further testing of the outer stator be conducted in order to repeat the fault and isolate the failure mechanism. These tests should be conducted after a completed lifetime duty cycle to include unirradiated and irradiated outer stator specimens.

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160 Figure 91. Degradation of PTFE alternator wire insulation near the transition plate terminals Figure 92. Micrograph of the c oil PTFE hook up wire insulation fracture pattern (150x)

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161 Figure 93. Micrograph of coil PTFE hook up wire insulation fracture pattern adjacent to the previous image (150x) Figure 94. 3D composite micrograph of the PTFE hook up wire insulation spiral crack (100x) C C u u W W i i r r e e B B r r a a i i d d S S p p i i r r a a l l C C r r a a c c k k

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162 Figure 95. Angled 3D composite image of the PTFE hook up wire insulation spiral crack (100x) Figure 96. 3D composite image of the PTFE hook up wire insulation spiral crack (100x) S S p p i i r r a a l l C C r r a a c c k k

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163 Figure 97. DSC curve comparison of Kynar h eat sh rink tubing. Figure 98. DSC curve comparison of Viton heat shrink tubing.

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164 Figure 99. DSC curve comparison of Viton Oring. As received data courtesy of Dr. E. Shin NASA GRC62. Figure 910. DSC curve comparison of Silicone Oring.

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165 Figure 911. DSC curve comparison of PTFE hook up wire insulation. Figure 912. DSC curve comparison of Xylan coating. As received data courtesy of Dr. E. Shin, NASAS GRC62.

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166 Figure 913. TGA weight curve comparison of Kynar heat shrink tubing. Figure 914. TGA weight curve comparison of Viton heat shrink tubing.

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167 Figure 915. TGA weight curve comparison of Viton Oring. Figure 916. TGA weight curve comparison of Silicone Oring.

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168 Figure 917. TGA weight curve comparison of PTFE hook up wire insulation. Figure 918. TGA curve comparison of Xylan coating As received data courtesy of Dr. E. Shin NASA GRC62.

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169 CHAPTER 10 RESULTS: MIXED NEUTRON & GAMMA RAY CANDIDATE MATERIAL E VALUATION Sample Transport, Control, and Preparation Irradiated samples were shipped in radioactive shipping containers (White I Class VII label) from TAMU to NASA GRC. Upon arrival, the GRC radiation safety officer (RSO) measured radiation levels 1 m from the container, on the container surface, 1 m from the samples, and on the sample surface using a handheld scintillation detector. Samples were transferred to an inert glove box back filled with argon gas to prevent post irradiat ion aging and to control access the samples as shown in Figure 101 A detailed sample inventory (A ppendix D Table D-9 and D10) was used to track the samples during all stages of use to include preparation, transport, testing, and disposal. The majorit y of sample processing was conducted in the glove box. Samples were transported to and from the glove box in labeled glass vials sealed within small plastic bags. Four sample conditions were tested: As Received (AR), Baked -O ut (BO), Baked O ut -A ged with a thermal history matching that of the irradiated samples (BOA) and Irradiated in the mixed neutron and ray field (IR). Candidate Material Description As mentioned in Chapter 3, specific organic material formulations are not listed due to intellectual property restrictions ; instead general descriptors are used Material functions include Oring sealants, electrical insulation shrink tubes, electrical wire insulation, running surface coatings, and bonding adhesives. Candidate materials include Silicone Orings for sealing in moderate temperature and pressure applications. Kalrez Orings are f luorocarbon elastomer s with excellent thermal stability. Kynar heat shrink is a polyvinylidene fluoride with moderate thermal stability. Viton heat shrink is a

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170 f lu orocarbon elastomer; a copolymer of vinylidene fluoride and hexafluoropropylene with improved thermal and mechanical stability. P olytetrafluroethylene ( PTFE ) or Teflon is a standard for electrical wire i n sulation with good thermal and chemical resistance but known radiation susceptibility. Polyimide is a high temperature and high radiation tolerance electrical wire insulator. The Xylan running surface coatings consists of PTFE filler dispersed in a poly imide resin binder. Hysol is a poly functional bonding epoxy resin used widely in aerospace applications. High Temperature (HT) Hysol epoxy has improved thermal stability and strength at higher temperatures. These candidate mater ials were irradiated at the TAMU reactor and undergo post irradiation evaluation. Dimension Measurement s Sample dimensions were obtained by averaging several measurements along various locations of all the samples. The dimension al change s of the samples are illustrated in Figure 102 through Figure 108. Trends were observed, s uch as the swelling experienced by the irradiated Silicone and Kalrez Orings when compared to the bakedout aged control sample s. The magnitude of cha nge is small and well within the standard deviation for the measurements, particularly when considering the Xylan coated aluminum samples. The diameter s for Kynar and Viton heat shrink tubing also decreased when comparing the irradiated to the aged samples; however, these trends can be seen to be a function of temperature and not necessarily radiation expos ure PTFE and Polyimide wire insulat ion diameters varied the least of all the samples and showed no distinguishable correlation to temperature or radiation exposure. In summary, no statistically significant changes in sample diameter or thickness were observed when comparing irradiated and non irradiated samples even with the same thermal histories

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171 Weight Measurement A summary of t he weight change measurements is illustrated in Table 101. The average Silicone Oring weight decreased approximately 0.04%, while the average Kalrez weight increased approximately 0.03%. The Kynar and Viton heat shrink average weight s increased approximately 0.16% and 0.1% respectively The heat shrinks were held in place during irradiation with tread and some small strands were baked into the samples which can account for the weight increase. Control samples show a weight decrease which is normal for heat shrinks during the shrinking process. PTFE and Polyimide wire showed an average weight increase of 0.02% and 0.03% res pectively. These samples were fixed in place during irradiation using Kapton tape and some residue was evident on the wire ends during weighing. This residue was with a solvent removed before characterization of the samples began. The Hysol and high temperature Hysol epoxy lap shear samples experienced an average weight loss of 0.4% and 0.06% respectively. This observation is particularly interesting because these samples were also held in place with Kapton tape and had noticeable residue deposits. Thi s means that even though some weight was gained with the Kapton residue the overall weight still decreased. The Xylan coated aluminum plate experienced an average weight gain of 0.3% and like the lap shear samples this weight gain can be attributed to tape residue deposited during the placement process In summary, a few of the irradiated samples experienced an average weight change that was outside of experimental error; however, there was no direct dependence on fluence or dose observed. In order to resolve these measured changes it is critical that a much larger data sampling be obtained so that we can assess if these changes have significant statistical relevance.

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172 Optical Microscopy Moderate magnification optical m icrographs were obtained for as rec eived, bakedout bakedout aged, and irradiated samples. The corresponding micrograph comparisons are illustrated in Figure 109 to Figure 1015. Optical microscopy found minor anomalies within the sample distribution. When comparing AR, BO, BOA, and I R samples it is apparent that the majority of observable changes occur due to changes in temperature conditions. Particularly the Kynar heat shrink tubing undergoes considerable shrinkage and darkening of optical density as the various heat treatment proc esses are applied. When comparing these results to the dimensional measurements already discussed we can safely state that no significantly discernable changes in surface color, morphology, or topography were observed when comparing as received, bakedout bakedout aged, and irradiated samples. Scanning Electron Microscopy Energy Dispersive Spectroscopy A field emission SEM was used to analyze the Xylan coated aluminum samples. A l ow magnification was first utilized to fix possible surface hydrocarbon contaminants by polymerization induced by the electron beam. The polymerization prevents the contaminants from diffusing to the electron probe during scanning. The resulting Xylan secondary electron (SE) micrographs are illustrated in Figure 1016 to Figure 10 -19. Unlike the smooth, uniform surface observ ed with the optical microscope the SEM micrographs show a topography rich in striations, PTFE inclusions, and voids in the Polyimide mat r ix. When the bakedout aged control samples were compared to the irradiated samples we observe no discernable changes in surface color, topography, morphology or onset of degradation. No observable changes can be attributed to either the different thermal or radiation environments experienced by the samples

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173 Low magnification back scatter electron (BSE) images compared to SE images differentiate the higher atomic number (higher Z) elements present within the sample as shown in Figure D 1. From the low magnification SE/BSE comparisons we find a ubiquitous distribution of high Z particulates dispersed throughout the matrix. At higher magnifications we observe the same SE/BSE comparisons in both irradiated and nonirradiated samples as shown in Figures D 2 and D 3. EDS is applied to the large low Z inclusions, the small high Z inclusions, and the matrix to characterize the possible elemental change as shown in Figure D 4 and Figure D5. The large low Z inclusions have a strong carbon and fluorine component which matches well with the expected results for PTFE The smaller high Z inclusions have strong molybdenum and sulfur peaks which correspond to MoS2 particulates which are added by the manufacturer in order to obtain certain desired properties. The matrix shows only carbon and oxygen peaks consistent with the Polyimide matrix. From the SE, BSE, and EDS results show Xylan experienced little to no obs ervable change due to either the thermal or radiation operating environment. These findings do not rule out the possibility of thermal or radiation induced structure change leading to property change; however, confirming this assumption must be done using physical property methodologies such as FTIR, DSC, and TGA. Fourier Transform Infrared Spectroscopy The FTIR surface spectra comparisons of bakedout aged control samples to irradiated NUKE 1 through NUKE 4 samples are illustrated in Figure 1020 through Figure 1026. Spectra peak location and intensities for Silicone control and irradiated samples are very similar. No peak shifting is evident, although there is some change in absorptance intensity. The absorption peak height has a strong dependence on how

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174 well the sample is held in contact with the germanium crystal. Kalrez unirradiated controls and irradiated samples show excellent agreement in terms of both peak intensity and location. NUKE 3 and NUKE 4 Kalrez samples have larger peaks than the unirradiated control samples but no peak shifting is observed, which is indicative that no significant chemical or compound change has occurred. Kynar, Viton, PTFE, Polyimide, and Xylan spectra all show excellent agreement both in terms of peak location and peak intensity when comparing unirradiated and irradiated samples. From these results we can conclude with certainty that no significant change in surface chemistry occurred after mixed neutron and ray irradiation at the appropriate operating temperat ure and atmospheric conditions. Differential Scanning Calorimetry The DSC curve comparisons are illustrated in Figure 1027 through Figure 1032. Artifacts such as a shift in the Tg, exothermic peaks, and endothermic peaks would be indicati ve of radiationinduced effects that are unlikely to be caused by thermal history. Kynar heat shrink shows some vertical offset s in the curves and are likely indicative of the natural variation in sample density. All sample conditions show good agreement in peak identification, with endothermic reactions for the samples irradiated at 125C occurring at approximately 4849C and 130134C. For samples irradiated at 150C endothermic reactions occur between approximately 4749C and 125128C. Viton heat shrink tubing samples of all conditions are well behaved and show good agreement. A Tg can be identified at approximately 20 to 22C and another endothermic reaction occurs between approximately 154155C. No significant differences between bakedout aged and irradiat ed sample curves can be found. No significant differences are observed between the two different irradiation or temperature

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175 conditions as well. Analysis of the SARTA Kynar curves found a slight decrease of 5C for the endothermic reaction at 155C when comparing as received to irradiated samples. N o change was found in comparing bakedout aged and irradiated samples ; thus, it is evident SARTA data trend s were due to bake out or heat shrinking process es Kalrez O ring curves show good agreement in sample trends, with a wide scatter range in the data. A s ubambient Tg is identified between approximately 4 to 5C and appears to be consistent with all conditions. This lack of Tg shift is evidence that no significant thermal or radiationinduced changes have occurred. Silicone Oring curves show good trend agreement and endothermic reaction peaks are identified between 43 to 47C for all cases. These peaks are too low to be associated with melting or evaporation of additives but are at temperatures well above the anticipated Tg. SARTA sample exothermic peaks were identified at approximately 300C for the as received and bakedout samples. These peaks are not evident in these results are likely due to variation in instrument calibration. Analysis of PTFE results show little difference between the bakedout aged controls and irradiated sample curves. The sharp endothermic peak at approximately 327 to 329C is identified as the primary melting reaction. T he Polyimide graphs show appreciable scatter in the data, with no transitions or peaks identified. The thermal and radiation stability of Polyimide wire insulation is well known and this data merely complement s the existing literature. In summary, chemica l changes characterized by DSC found no major differences between bakedout aged control and irradiated samples. The degree of change was statistically insignificant and well within material design margins.

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176 ThermoGravimetric Analysis Silicone TGA curve comparisons are illustrated in Figure 10-33 through Figure 1039. The irradiated samples match closely with the respective bakedout aged control samples. There is some scatter between the different curves but there is no observable trend that either thermal aging or radiation significantly change Td. W e compar e the control samples and samples irradiated at 125C (NUKE 1 and 3) and 150C (NUKE 2 and 4) We observe that the irradiated Silicone curves match well with the aged control samples and have a Td offset of 4 -5C between NUKE 1 and NUKE 3 cases. We see the same result for NUKE 2 and NUKE 4 samples. This indicates that since the samples had the same thermal history any changes can likely be attributed to the difference in fluence and dose. The magnitude of change is slightly outside of experimental error but well within design limitations. Kalrez shows nearly identical weight loss curves for all cases. This apparent lack of curve trend changes indicates that neither thermal history nor radiat ion exposure history has a significant effect on the rate of weight loss or on the Td. The primary reason Kalrez is under consideration for service is because of its thermal stability. Irradiated Kynar sample curves closely match the control curves for the different cases. This trend illustrates the strong dependence between the thermal history and the expected performance of this particular heat shrink tubing. Irradiated Viton sample curves also match the thermal history control well for all cases. As w ith Kynar, the Viton heat shrink has a strong dependence on thermal history. The most change is actually observed when comparing as received to the first bake process. Subsequent baking and aging have minimal effect on properties.

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177 PTFE control and irradiated sample curves show excellent agreement, yet have some of the largest Td offsets of all the materials. There is no discernable difference between the different conditions so it is likely the offset is a function of the natural variation of the polymer itself resulting from the manufacturing process. Polyimide sample curves also show excellent agreement between irradiated and control cases and a moderate Td offset. In addition, these data curves have a significant amount of scatter in the data and this is attributed to the low sample mass. Xylan samples were low mass and this also result ed in significant scatter in the data. No trends are apparent with either thermal or irradiation conditions. In nearly all the materials evaluated the maximum changes i n Td were within experimental error and we assume that chemical changes characterized by TGA found no statistically significant differences between control and irradiated samples. Dynamic Mechanic Analysis PTFE DMA curve comparison is illustrated in Figure 10-40 The irradiated samples match nicely with their respective bakedout aged control samples. There is some scatter between the different curves but there is no observable trend that would indicat e of either thermal aging or radiation significantly c hang e Tg or Td. When comparing the samples irradiated at 125C (NUKE 1 and 3) we see a maximum difference of 5C in Tg and 4C in Td. When comparing samples irradiated at 150C (NUKE 2 and 4) we see a maximum difference of 1C in Tg and 2C in Td. In bo th cases, the maximum changes are within experimental error and variation in calibration period for the equipment. Therefore, we can say that chemical changes characterized by D MA found no statistically significant differences between control and irradiat ed samples. The results correlate well with those results obtained through DSC.

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178 Electrical Resistivity Measurement Electrical resistivity measurements can be found in Appendix D, Table D 10. The measurement method was coarse in that it simply provided a qualitative result of whether or not the electrical wire and heat shrink insulation maintained their dielectric properties at an applied voltage. Bakedout aged and irradiated samples were evaluated and all retained their insulating properties. However, these results are inconclusive in nature and do not indicate if a change occurred in the insulation resistance. When considering the minimal change in properties observed in all the previous evaluation methods it is likely that any change in electrical insulation performance is insignificant as long as the insulation mechanical integrity is maintained. ORing Compression Set Tests The results of the Oring compression set tests are shown in Figure 10-41. Silicone and Kalrez both show similar trends as a function of time; however, no t rends as a function of temperature or radiation exposure were observed. The Silicone Orings showed improved perform ance over the Kalrez O rings with regard to compression recovery This slight change in performance is more likely a function of thermal history and not necessarily due to irradiation. Alt h ough changes were measured the magnitude of performance degradation was well within design criteria. M any more irradiated samples are required in order to increase the statistical confidence of these initial findings for Kalrez and Silicone Orings. Lap Shear Tensile Test S tatic adhesion bond strengths of the Hysol and high temperature (HT) Hysol epoxies were determined as a function of mixed neutron and ray irradiation. Bond strength and toughness were determined by failure testing subscale sandwich lap shear

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179 specimens The measured results are used to calculate the resulting change in sample performance with respect to the performance of a control sample A set of c ontrol samples underwent similar thermal aging as the samples irradiated at TAMU. The calculated changes for the irradiated specimens NUKE 1 through 4 were based on properties of their corresponding thermal ly aged con trols. The calculated changes of the thermal ly aged controls w ere based on the properties of the bakedout control samples. The lap shear axial tension results for bond strength and toughness are illustrated in Figure 10-42 and Figure 10-4 3, respectively The influence of a mixed neutron and ray radiation field on the bond integrity was less significant than the influence of thermal history in both Hysol and H igh T emperature Hysol epoxies. The Hysol epoxy showed slightly different behavior, with the bo nd integrity positively affected by thermal aging but toughness degraded slightly after irradiation even though bond strengths mostly improved. In the case of the H igh T emperature Hysol epoxy thermal exposures actually lowered its strength and toughness; however, irradiation to both fluence levels actually improved properties, especially when irradiated at 125C. This improvement in strength and toughness is likely due to radiationinduced curing since the H igh T emperature Hysol epoxy samples were somewhat under cured before irradiation. In summary, both Hysol and H igh T emperature Hysol epoxies maintained the required bond integrity after irradiation in a mixed neutron and ray radiation environment The limited number of samples means there is signific ant uncertainty in the data and a larger set of irradiated samples is required in order to increase the statistical confidence of these findings for Hysol and H igh T emperature Hysol epoxies.

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180 Conclusions The Stirling alternator candidate organic material samples were irradiated in a mixed neutron and gamma ray field to estimated maximum FSP lifetime fluence s of 1x1014 and 5x1014 n/cm2 a nd a corresponding dose of 1.355 and 5.38 Mrad respectively, while under ultra high purity helium at 125C and 15 0C. No statistically significant changes in sample diameter or thickness were observed when comparing irradiated and non irradiated samples even with the same thermal histories. We conclude that these candidate materials are capable of maintaini ng their shape well within established dimensional tolerance requirements. A few of the material samples experienced an average weight change outside of experimental error; however, these changes can be attributed to artifacts not associated with the radiation exposure. A much larger of number samples is required in order to accurately address weight changes with any significant statistical relevance. FTIR spectra comparison between as received, bakedout, and baked out aged control samples with irradiated samples found little to no change in surface chemistry Since the surface composition was subjected to the largest number of incident neutron and rays, it is highly unlikely that any changes would be determined through FTIR within the sample by evaluating sectioned specimens. Bulk thermal and chemical changes characterized by DSC, TGA and DMA evaluation found only minor differences between bakedout, bakedout aged and irradiated sample conditions No significant shifts in Tg, Td, endothermic, and exothermic peak locations could be resolved. Those shifts that were found were within the calculated experimental error for each instrument. The degree of change was statistically insignificant and well within material design margins.

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181 Optical microscopy found no discernable changes in surface color or morphology when comparing as received, bakedout bakedout aged, and irradiated samples. SEM found no significant changes between control and irradiated Xylan coatings Oring compression set tests measured a slight change in recovery as a function of thermal history and radiation history. A larger number of irradiated samples is required to increase statistical confidence of these findings. Although these changes were measure able, the magnitude of performance degradation was well within design criteria and unlikely to pose a credible failure mechanism Lap shear axial tension tests results indicate some measureable changes in bond strength as a function of both thermal aging and irradiation. B oth Hysol and H igh T emperature Hysol epoxies maintained the required bond integrity after irradiation. A much larger sample data set of irradiated samples is required in order to increase the statistical confidence of these initial findings for the Hysol and HT Hysol epoxies. Electrical resistivity measurements although coarse in nature, show no appreciable change after irradiation. When considering the minimal change in properties observed in all the previous methods it is likely that change in electrical resistivity wa s insignificant. As with the results from the SARTA wire evaluation, insulation degradation will not pose a significant failure mechanism as long as cracking or flaking of the insulation (whether thermal and/or radiation induced) does not occur. In summary, o rganic material properties experienced n o significant degradation O rganic material candidates under consideration for Stirling alternator service sh ould be able to meet service life requirements It is recommended that PTFE wire insulation and Kynar heat shrink tubing be replaced with radiation tolerant substitutes.

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182 Figure 101. NASA GRC g love box for storage and processing of radioactive samples. Figure 102. Xylan coating on aluminum dimension change as a function of condition. 0.985 0.990 0.995 1.000 1.005 1 2 3 4 Thickness (normalized) Sample Condition N1 N2 N3 N4 AR BO BOA IR 125oC 1x1014n/cm21.35 Mrad 150oC 1x1014n/cm2 1.35 Mrad 125oC 5x1014n/cm25.38 Mrad 150oC 5x1014n/cm25.38 Mrad

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183 Figure 103. Silicone Oring dimension change as a function of condition. Figure 104. Kalrez Oring dimension change as a function of condition. 0.980 1.000 1.020 1.040 1.060 1.080 1.100 1 2 3 4 Diamter (normalized) Sample Condition N1 N2 N3 N4 BO BOA IR 125oC 1x1014n/cm21.35 Mrad 150oC 1x1014n/cm2 1.35 Mrad 125oC 5x1014n/cm25.38 Mrad 150oC 5x1014n/cm25.38 Mrad 0.980 0.990 1.000 1.010 1.020 1.030 1.040 1.050 1.060 1.070 1 2 3 4 Diameter (normalized) Sample Condition N1 N2 N3 N4 AR BO BOA IR 125oC 1x1014n/cm21.35 Mrad 150oC 1x1014n/cm2 1.35 Mrad 125oC 5x1014n/cm25.38 Mrad 150oC 5x1014n/cm25.38 Mrad

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184 Figure 105. Kynar heat shrink dimension change as a function of condition. Figure 106. Viton heat shrink dimension change as a function of condition. 0.540 0.740 0.940 1.140 1.340 1.540 1.740 1.940 1 2 3 4 Diameter (normalized) Sample Condition N1 N2 N3 N4 AR BO BOA IR 125 oC 1x1014n/cm21.35 Mrad 150 oC 1x1014n/cm2 1.35 Mrad 125 oC 5x1014n/cm25.38 Mrad 150 oC 5x1014n/cm25.38 Mrad 0.980 1.080 1.180 1.280 1.380 1.480 1.580 1 2 3 4 Diameter (normalized) Sample Condition N1 N2 N3 N4 AR BO BOA IR 125oC 1x1014n/cm21.35 Mrad 150oC 1x1014n/cm2 1.35 Mrad 125oC 5x1014n/cm25.38 Mrad 150oC 5x1014n/cm25.38 Mrad

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185 Figure 107. PTFE wire dimension change as a function of condition. Figure 108. Polyimide wire dimension change as a function of condition. 0.980 0.985 0.990 0.995 1.000 1.005 1.010 1.015 1.020 1 2 3 4 Diameter (normalized) Sample Condition N1 N2 N3 N4 AR BO BOA IR 125oC 1x1014n/cm21.35 Mrad 150oC 1x1014n/cm2 1.35 Mrad 125oC 5x1014n/cm25.38 Mrad 150oC 5x1014n/cm25.38 Mrad 0.985 0.990 0.995 1.000 1.005 1.010 1.015 1.020 1 2 3 4 Diameter (normalized) Sample Condition N1 N2 N3 N4 AR BO BOA IR 125oC 1x1014n/cm21.35 Mrad 150oC 1x1014n/cm2 1.35 Mrad 125oC 5x1014n/cm25.38 Mrad 150oC 5x1014n/cm25.38 Mrad

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186 Table 10 -1 Organic Materials W eight Measurement (Pre and Post Irradiation) Material Sample Wpre (g) Wpost (g) W (g) W (%) WAvg (%) Silicone O ring NUKE 5 0.9134 0.9127 0.0007 0.0766 0.0466 NUKE 6 0.9134 0.9127 0.0007 0.0766 NUKE 7 0.8926 0.8926 0.0000 0.0000 NUKE 8 0.9056 0.9053 0.0003 0.0331 Kalrez O ring NUKE 5 0.9973 0.9986 0.0013 0.1304 0.0375 NUKE 6 1.0255 1.0257 0.0002 0.0195 NUKE 7 1.0259 1.0259 0.0000 0.0000 NUKE 8 1.0172 1.0172 0.0000 0.0000 Kynar Heat Shrink Tubing NUKE 5 1.0591 1.0621 0.0030 0.2833 0.1660 NUKE 6 1.0768 1.0779 0.0011 0.1022 NUKE 7 1.0587 1.0599 0.0012 0.1133 NUKE 8 1.0290 1.0307 0.0017 0.1652 Viton Heat Shrink Tubing NUKE 5 1.9643 1.9667 0.0024 0.1222 0.0832 NUKE 6 1.9398 1.9411 0.0013 0.0670 NUKE 7 1.9348 1.9370 0.0022 0.1137 NUKE 8 1.9938 1.9944 0.0006 0.0301 PTFE Wire NUKE 5 1.9672 1.9682 0.0010 0.0508 0.0204 Insulation NUKE 6 1.9523 1.9529 0.0006 0.0307 NUKE 7 2.0062 2.0062 0.0000 0.0000 NUKE 8 1.9848 1.9848 0.0000 0.0000 Polyimide Wire NUKE 5 2.5457 2.5467 0.0010 0.0393 0.0322 Insulation NUKE 6 2.4527 2.4547 0.0020 0.0815 NUKE 7 2.4527 2.4528 0.0001 0.0041 NUKE 8 2.5011 2.5012 0.0001 0.0040 HT Hysol Epoxy NUKE 5 10.8708 10.8689 0.0019 0.0175 0.4020 Lap Shear NUKE 6 10.8997 10.8378 0.0619 0.5679 NUKE 7 11.1959 11.0813 0.1146 1.0236 NUKE 8 10.9946 10.9947 0.0001 0.0009 Hysol Epoxy NUKE 5 10.9005 10.8923 0.0082 0.0752 0.0617 Lap Shear NUKE 6 11.1045 11.1056 0.0011 0.0102 NUKE 7 10.9108 10.9448 0.0340 0.3116 NUKE 8 11.0868 11.0321 0.0547 0.4934 Xylan Coated NUKE 5 7.6333 7.6400 0.0067 0.0878 0.3455 Aluminum Plate NUKE 6 7.3256 7.3318 0.0062 0.0846 NUKE 7 7.6183 7.6616 0.0433 0.5684 NUKE 8 7.7660 7.8158 0.0498 0.6413

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187 A B C D Figure 109. Optical micrograph comparison at 30x of Silicone Oring (Left : bakedout aged, right : irradiated). A) NUKE 1, B) NUKE 2, C) NUKE 3, D) NUKE 4. A B C D Figure 1010. Optical micrograph comparison at 30x of Kalrez Oring (Left : bakedout aged, right: irradiated). A) NUKE 1, B) NUKE 2, C) NUKE 3, D) NUKE 4.

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188 A B C D Figure 1011. Optical micrograph comparison at 22.5x of Kynar heat shrink ( Left : asr eceived control, center : bakedout aged control right : irradiated) A) NUKE 1, B) N UKE 2, C) NUKE 3, D) NUKE 4. A B C D Figure 1012. Optical micrograph comparison at 22.5x of Viton heat shrink ( Left : asr eceived control, center : bakedout aged control right : irradiated) A) NUKE 1, B) NUKE 2, C) NUKE 3, D) NUKE 4.

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189 A B C D Figure 1013. Optical micrograph comparison at 37.5x of PTFE wire ( Left : as-r eceived control, center : bakedout aged control right : irradiated) A) NUKE 1, B) NUKE 2, C) NUKE 3, D) NUKE 4. A B C D Figure 1014. Optical micrograph comparison at 22.5x of Polyimide wire ( Left : asr eceived control, center : bakedout aged control right : irradiated) A) NUKE 1, B) NUKE 2, C) NUKE 3, D) NUKE 4.

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190 A B C D Figure 1015. Optical micrograph comparison at 37.5x of Xylan coated aluminum. Left hand side is the baked out aged control the right is irradiated. A) NUKE 1, B) NUKE 2, C) NUKE 3, D) NUKE 4.

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191 A B C D Figure 10-16 Comparison of Xylan SEM SE images A) BOA 1 at 300x, B) NUKE 1 at 300x, C) BOA 1 at 1000x, and D) NUKE 1 at 1000x.

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192 A B C D Figure 10-17 Comparison of Xylan SEM SE images A) BOA 2 at 300x, B) NUKE 2 at 300x, C) BOA 2 at 1000x, and D) NUKE 2 at 1000x.

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193 A B C D Figure 1018. Comparison of Xylan SEM SE images A) BOA 3 at 300x, B) NUKE 3 at 300x, C) BOA 3 at 1000x, and D) NUKE 3 at 1000x.

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194 A B C D Figure 1019. Comparison of Xylan SEM SE images A) BOA 4 at 300x, B) NUKE 4 at 300x, C) BOA 4 at 1000x, and D) NUKE 4 at 1000x.

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195 A B Figure 1020. FTIR spectra comparison for Silicone b akedout -a ged (BOA) vs. A) Nuke 1 and 2, B) Nuke 3 and 4.

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196 A B Figure 1021. FTIR spectra comparison for Kalrez bakedout aged (BOA) controls vs. A) NUKE 1 and 2, B) N UKE 3 and 4 irradiated samples

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197 A B Figure 1022. FTIR spectra comparison for Kynar bakedout aged (BOA) controls vs. A) NUKE 1 and 2, B) NUKE 3 and 4 irradiated samples.

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198 A B Figure 1023. FTIR spectra comparison for Viton bakedout aged (BOA) controls vs. A) NUKE 1 and 2, B) NUKE 3 and 4 irradiated samples.

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199 A B Figure 1024. FTIR spectra comparison for PTFE bakedout aged (BOA) controls vs. A) NUKE 1 and 2, B) NUKE 3 and 4 irradiated samples.

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200 A B Figure 1025. FTIR spectra comparison for Polyimide bakedout aged (BOA) controls vs. A) NUKE 1 and 2, B) NUKE 3 and 4 irradiated samples.

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201 A B Figure 1026. FTIR spectra comparison for Xylan bakedout aged (BOA) controls vs. A) NUKE 1 and 2, B) NUKE 3 and 4 irradiated samples.

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202 A B Figure 10-27 Comparison of DSC curves for Silicone irradiated at A) 125C (NUKE 1 & 3) and B) 150C (NUKE 2 & 4).

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203 A B Figure 10-28. Comparison of DSC curves for Kalrez irradiated at A) 125C (NUKE 1 & 3) and B) 150C (NUKE 2 & 4).

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204 A B Figure 10-29. Comparison of DSC curves for Kynar irradiated at A) 125C (NUKE 1 & 3) and B) 150 C (NUKE 2 & 4).

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205 A B Figure 10-30. Comparison of DSC curves for Viton irradiated at A) 125C (NUKE 1 & 3) and B) 150C (NUKE 2 & 4).

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206 A B Figure 1031. Comparison of DSC curves for PTFE irradiated at A) 125C (NUKE 1 & 3) and B) 150C (NUKE 2 & 4).

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207 A B Figure 1032. Comparison of DSC curves for Polyimide irradiated at A) 125C (NUKE 1 & 3) and B) 150C (NUKE 2 & 4).

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208 A B Figure 10-33. Comparison of TGA curves for Silicone irradiated at A) 125C (NUKE 1 & 3) and B) 150C (NUKE 2 & 4).

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209 A B Figure 10-34. Comparison of TGA curves for Kalrez irradiated at A) 125C (NUKE 1 & 3) and B) 150C (NUKE 2 & 4).

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210 A B Figure 10-35. Comparison of TGA curves for Kynar irradiated at A) 125C (NUKE 1 & 3) and B) 150C (NUKE 2 & 4).

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211 A B Figure 10-36. Comparison of TGA curves for Viton irradiated at A) 125C (NUKE 1 & 3) and B) 150C (NUKE 2 & 4).

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212 A B Figure 10-37. Comparison of TGA curves for PTFE irradiated at A) 125C (NUKE 1 & 3) and B) 150C (NUKE 2 & 4).

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213 A B Figure 10-38. Comparison of TGA curves for Polyimide irradiated at A) 125C (NUKE 1 & 3) and B) 150C (NUKE 2 & 4).

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214 Figure 1039. Comparison of TGA curves for Xylan irradiated at 125 C (NUKE 1 & 3).

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215 A B Figure 10-40. Comparison of PTFE DMA curves of bakedout aged (BAO) versus irradiated (N) for samples irradiated at A) 125C( NUKE 1 and 3) and B) 150C ( NUKE 2 and 4).

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216 A B Figure 10-41. Oring compression set test results for A) Silicone and B) Kalrez. 1.0 0.0 1.0 2.0 3.0 4.0 5.0 6.0 7.0 0.0 10.0 20.0 30.0 40.0 50.0 Average C B (%) Time After Unloading (Hours) AR BOA N1 N2 N3 N4 10.0 5.0 0.0 5.0 10.0 15.0 20.0 25.0 30.0 0.0 10.0 20.0 30.0 40.0 50.0 Average C B (%) Time After Unloading (Hours) AR BOA N1 N2 N3 N4

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217 A) B) Figure 10-42. B ond strength of A) Hysol and b) H igh T emperature Hysol epoxies lap shear samples tested at 120C. +6% +34% +6% +6% +27% 14% +36% +1% 0.0 200.0 400.0 600.0 800.0 1000.0 1200.0 1400.0 1600.0 1800.0 NUKE 1 NUKE 2 NUKE 3 NUKE 4 Lap Shear Bond Strength (psig) Baked Control Aged Control Irradiated 18% 7% 23% 13% +23% 5% +13% 21% 0.0 200.0 400.0 600.0 800.0 1000.0 1200.0 1400.0 1600.0 1800.0 2000.0 NUKE 1 NUKE 2 NUKE 3 NUKE 4 Lap Shear Bond Strength (psig) Baked Control Aged Control Irradiated

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218 A) B Figure 10-43. Toughness of A) Hysol and B ) High Temperature Hysol epoxies lap shear samples tested at 120 C. +5% +20% +5% +20% 29% 48% 22% 8% 0.0 2.0 4.0 6.0 8.0 10.0 12.0 14.0 16.0 18.0 NUKE 1 NUKE 2 NUKE 3 NUKE 4 Strain term to Failure (in 1) Baked Control Aged Control Irradiated 14% +1% 39% 35% +41% +22% +32% +16% 0.0 5.0 10.0 15.0 20.0 25.0 NUKE 1 NUKE 2 NUKE 3 NUKE 4 Strain term to Failure (in 1) Baked Control Aged Control Irradiated

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219 CHAPTER 11 RELEVANCE OF NUCLEAR AND RADIATION TESTS The s pace nuclear systems project must emphasize a hardware rich research and development strategy17. A primary objective of the space nuclear project is to establish technical credibility by specifically focusing efforts on practical and affordable performance demonstrations of materials, components and systems. Although nonnuclear issues constitute the vast majority of research and development efforts, this chapter will discuss considerations for identifying the need to conduct nuclear and/or radiation tests, selecting an appropriate test facility, factors unique to nuclear and/or radiation tests and navigating the political structure. Nuclear and Radiation Testing Within A Research & Development Program Testing is the only method that will provide a high degree of certainty with regard to material, component, or system performance63. However, tests are often expensive and the importance of the desired data must be compared against the overall project budget, utility, and timeline. Tes t effectiveness is used to define the degree to which the data generated in a test contributes to a successful mission application64. Test effectiveness factors include the test article maturity (e.g. representative of flight hardware) and the realism of the dominant test conditions, many of which may not require a representative nuclear environment64. A n important lesson learned from terrestrial power and research reactor development i s that the majority of research and development efforts are actually nonnuclear in nature. Dominating engineering issues that require mitigation include corrosion, thermomechanical fatigue, creep, coefficient of thermal expansion mismatch, conting ency operations, off nominal procedures, etc.

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220 Many in the space nuclear community consider nuclear testing too expensive for an incremental increase in technology validation63. A full power nuclear test is estimated to be on the order of one billion dollars63. Very similar data can be obtained by implementation of comprehensive nonnuclear methods, such as substitution of nuclear fuel with electric resistance heaters and well designed zeropower critical tests, which are estimated to cost on the order of a few million dollars63. Therefore, a primarily nonnuclear focused program will allow for both reasonable development periods and relatively affordable cost s. However, at some point nuclear and radiation test s must be carried out Radiation tests define subjecting equipment to radiation whereas nuclear tests describe react or specific tasks such as reactor control and fuel development. Traditionally, nuclear and radiation tests have been v ery expensive. Such tests can easily dominate a project budget if not carefully planned and executed. Cost effective nuclear and radiation test s can be conducted by utilizing methodologies such as collaborating with graduate students to carry out research, university facilities, and low cost national laboratory facilities. The type of facility used to test the equipment in question depends on the maturity level of the hardware. Selection of an Appropriate Nuclear or Radiation Facility A properly designed t est matrix can generally be divided into four phases which consequently also increase test effectiveness, fidelity, and cost63: Developmental (screening): Materials, components and subsystems are selected, designed, and tested at somewhat representative conditions. Demonstration: Functionality (nominal, off nominal, transient) established, limited component optimization, initial system integration and modification. Flight Qualification: Rigorous t ests to ensure syst ems perform as required. Acceptance: F light system tests conducted at prototypic operating conditions.

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221 Screening and Demonstration Tests I t has been suggested by some in the radiation effects community that components and materials must only be exposed to the exact radiation environment expected (e.g. particle type, flux, energy, dose rates, etc) during operation for all technology test maturity levels. An example would be to use particle accelerators if a mission is expected to operate in the Jovian radiation environment ( high energy electrons protons and heavy charged particles) or use of high power reactors for FSP testing ( with a representative neutron spectrum )65. However, r elevant results from radiation test s can be obtained in a cost effective and practical manner utilizing radiation sources that are not representative of the expected operating enviornment65. For example, r elatively low cost not necessarily prototypic radiation testing can be applied as long as the factors influencing radiation effects are properly taken into consideration using appropriate design of experiments methodologies. Utilizing low energy, moderate flux ray sources can be used for initial material and component level screening trials. Similar ray facilities with low energy but high fluxes are ideal higher maturity level component and system performance demonstrations. These low cost facilities enable long duration, low dose rate or short duration, high dose rate accelerated life test s. The selection of an appropriate nuclear or radiation test facility should be highly dependent on the test readiness level (TRL) of the technology to be tested. Low TRL technology meant to undergo screening trials should be irradiated at relatively inexpensive facilities such as universities or low cost national laboratory facilities. Several universities have research reactors particle accelerators, low dose rate photon sources, and high dose rate photon sources. Examples of photon sources include 60Co or 137Cs at institutions such as the University

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222 of Florida6 6,67, Texas A&M University, Ohio State University (OSU), University of Wisconsin, University of Michigan, Georgia Tech, Duke, and Univers ity of Chicago to name a few. Component or system level performance demonstrations should be conducted in an accelerated environment so that lessons learned from testing, such as synergistic effects, can be incorporated early in the design process. Appro priate facilities include low energy but high flux ray sources, again available at various university and national laboratories. Experienced and properly equipped radiological metrology laboratories are essential to ensure that the radiation environment is properly characterized so that results of the irradiations are scientifically accurate and will be of relevance to the supporting project. Flight Qualification and Acceptance Tests As a technology matures and approaches flight TRL the r adiation test objectives will shift from screening and demonstration trials to flight qualification and acceptance trials Subsequent flight qualification and acceptance trials should be conducted at facilities that offer more pr o to typic radiation environment s in order to build confidence in the systems being evaluated. These include high energy high flux particle accelerators and reactors with an appropriate neutron spectrum and ray environments. Such radiation facilities with corresponding radiation metrology experts can be found within the DOE national laboratories Although far more expensive, they do provide the required high fidelity environmental conditions necessary to establish a high level of confidence to approve a technology for space flight. However, it may be possible to use cost effective facilities that offer aggressive radiation environments to meet qualification criteria. The decision of whether to use prototypic environments versus cheaper nonprototypic radiation sources is likely to be a topic for debate.

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223 Examples of Facility Selection and Hardwar e Maturity Maturity based test methodology has seen some initial use in terrestrial nuclear applications. For example, OSU and TAMU were recently awarded 2010 DOE Nuclear Energy University Program (NEUP) awards to work with Luna innovations on fiber optic instrumentation development. OSU will conduct fiber fundamental materials evaluation to include measurement of optical fiber performance when subjected to highradiation and hightemperature environments using the 500 kWth research reactor, and modeling to p redict fiber reliability and lifetime accurately 68. A broader range of follow research will be conducted at TAMU with activities to include fabrication of the sensor hardware, test article design and fabrication, incore sensor demonstration using the 1 MWth TRIGA research reactor, and 3D modeling69. A similar, step wise approach is slowly being introduced into space nuclear power and propulsion as demonstrated by the completion of this and other similar research project s. Importance of Proper Facili ty Selection The consequences for selecting an inappropriate facility for screening or qualification radiation test s at the inappropriate time (e.g. low maturity technology that is not representative of a flight system tested at a high fidelity facility ) w ill very likely result in the costs o f such tests increasing by at least one or two order s of magnitude. Cost overruns will not only apply to each individual test but for all subsequent tests expected to be conducted throughout the entire development cycle of a project. In addition, the availability of high er fidelity facilities is extremely limited and may interfere or likely delay, the sch edule of a system attempting to meet a time critical launch window. Therefore, it behooves the designers of the test matrices to properly take facility selection into account and the political ramifications that may develop.

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224 Overcoming Political Constraints Several nuclear and radiation tests have been identified as critical milestones to advance space nuclear power and propulsion technology but t hese tests are often postponed. Delaying these tests towards the end of a development program is usually due to budgetary constraints, resulting in initial screening experiments often treated as flight qualification t rials Radiation and nuclear research efforts have occasionally face d institutional opposition and lack of cooperation, often attributed to competition over funds or credit to conduct such tests as a standalone entity. Institutional roadblocks create a condi tion where no nuclear or radiation testing will be conducted at all, seriously limiting the confidence and reliability of space nuclear systems. What is needed to prevent such scenarios is developing a project that emphasizes flight of a system as the succ ess criteri on. Using neutral team members to work between institutions is a methodology that can lead to successfu l complete of research tasks. Utilization of Graduate Student Researchers W hen projects are highly controlled by upper management the researc h efforts come under increased scrutiny excessive oversight, and intense debate over distribution of accolades. These behaviors, although social in nature, have a direct impact on the research and development process, increasing the probability of delay i ng a schedule and increasing costs. I t was the authors experience in completing this project that utilizing graduate student s allow s for institutions to bridge gaps since they are viewed as neutral researcher s. This dissertation project provided the auth or the flexibility to collaborate jointly with multiple NASA centers, DO E national laboratories, and universities. In addition, facility personnel were far more willing to make concessions, often showing

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225 great flexibility, cooperation, and a desire to see the students project succeed. Allowing graduate students to make their masters thesis or doctoral dissertation research topic an important aspect of a space nuclear project is mutually beneficial for both the student and the sponsoring institution (e.g. NASA or DOE). The student will have the opportunity to work on relatively well funded, cutting edge research and the supporting institutions will be able to supervise research efforts that otherwise may be politically difficult to conduct. There are numerous competitive fellowships available to graduate students that allow for independent or collaborative research with the students university. It is the authors opinion that the staff working at the engineering research and development level ( at all the participating facilities ) want such tests to succeed for the advancement of the program Many expressed enthusiasm that such a novel approach of having a graduate student conduct the research is likely the only way relevant results can be ob tained promptly and affordably. Transition of Research Efforts Much of the development efforts leading to qualification tests should be conducted using collaboration with inter disciplinary graduate student researchers. This approach of utilizing graduate students is the model for activities at various institutions, but has been championed by the Center for Space Nuclear Research (CSNR) at Idaho National Laboratory (INL). Once the technology is nearly at flight qualification status the inter agency negotiation process must involve full time personnel, possibly with support from the graduate student researchers that assisted in maturing the technology. Nuclear and/or Radiat ion Specific Test Factors to Consider Several variables must be considered before testing. The primary factors to consider when choosing an appropriate radiation test facility include the radiation type,

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226 energy, flux, dose rate, flexibility to tailor the spectrum, special design requirements (e.g. dry cell vs. water immersion), dimensional constraints, facility availability, safety analysis period, approval period, experiment set up fees, special design fees, irradiation hourly fees, technician support, radiation metrology laboratory support, and inter agency license requirements. Table 111 compares a few radiation test facilities that have been utilized or are under consideration by NASA in support of FSP and other programs. This list in no way constitu tes the only research facilities that may be able to offer the desired radiation environment. The majority of the modifiers can be controlled by the researcher, which has a direct influence on cost and schedule. The factor with perhaps the largest impact is the complexity of the experiment, which has a tremendous effect on testing operations and subsequently all associated costs. Other factors include handling of potentially radioactive equipment and post irradiation disposal. Activation of Experimental Equipment Non nuclear test methodology allow s for post test inspection of the experimental hardware and for modifications to be implemented relatively soon. However, in nuclear tests implementation of these same modifications on activated hardware is diff icult and cost prohibitive since specialized techniques are required. The problem with activated hardware is typically circumvented by building a completely new second generation apparatus integrating lessons learned, which obviously increases cost signif icantly. Additional difficulties with conducting nuclear tests include the possibility of long decay times for activated materials, transport and handling of activated materials, post irradiation disposal of activated hardware, and there is always the poss ibly of an unsupportive public opinion. Although all the facilities utilized in this project possessed NRC licenses for shipping, accepting, and handling activated materials this is not

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227 always the case T herefore, careful consideration must be taken to ensure that inter agency licenses are compatible with the experiment requirements. Processing of samples, particularly those that will destructively evaluate the samples and have the potential to contaminate evaluation hardware can be particularly limiting depending on isotope half life and activity. There are several national laboratory facilities that are specifically designed to transport, manipulate, and evaluate experiments that will generate extremely high radiation activity and/or long isotopic hal f lives. The INL Hot Fuel Examination Facility (HFEF) INL Materials & Fuels Complex (MFC), and ORNL HFIR Fuel Processing Facility provide an extensive inventory of nondestructive and destructive evaluation techniques available. Unfortunately, the cost of utilizing such hot cell facilities is extreme and should be reserved only for research efforts that cannot be conducted by any other means. These include nuclear fuel development, validation of decades old experimental data, transport codes, etc. A low cost alternative available to universities is participation in the National Scientific User Facility (NSUF) program The NSUF program allows universities to submit a proposal requesting access to irradiat e experiment s at one of several DOE laboratories that participate in the NSUF program (including HFIR, ATR, etc.)70. Approved proposals are granted access with the hosting laboratory incurring the majority of facility costs70. The visiting university researchers are still required to f und the safety analysis, approval costs and design fees but the irradiation time is free of charge. NASA and DOE consulting with universit ies to develop and submit NSUF research proposal s tied to flight critical experiments may be the only cost effective way to gain access to such high end research facilities in the near term.

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228 Post Irradiation Disposal Disposal of activated materials can dominate the cost of neutron and charged particle exposed hardware. Again, proper material selection early in the exper iment design process will limit the generat ion of high activity and/or long lived half life isotopes. Therefore, most properly designed experiment hardware could be treated as low level waste (LLW) within a relatively short period of time, greatly reducing the post irradiation disposal costs. Under certain circumstances activated hardware is unable to be transported due to NRC license restrictions, primarily due to generation of certain isotopes during testing. One very attractive alternative to disposing of the activated hardware is to donate it to the irradiation facility ; that is assuming that the equipment is fully functional after irradiation test s ha ve been completed. Often the irradiated hardware is safe to handle within a relatively short period of time and can be reused on subsequent irradiation tests in support of various research efforts Re using hardware is mutually beneficial to both the researcher and the irradiating facility in that post irradiation disposal costs are avoided and equipmen t donation can enable enhanced testing infrastructure at no cost to the irradiation facility. This model of hardware donation was exercised with the mixed neutronray tests conducted in this project between NASA GRC and TAMU. The four ultra high vacuum capable (and considerably expensive) aluminum test article assemblies were donated to the TAMU NSC after the first eight initial irradiation runs. These test articles are intended to be used in future research efforts internal to TAMU In addition, a se cond series of organic irradiations and NdFeB and SmCo magnetic material irradiations are already under serious consideration for follow on experiments with NASA GRC and possibly the DOE

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229 Conclusions It is no secret that different centers within any large organization do not always cooperate. Previous efforts to promote cooperation between different NASA centers h as proven difficult and the challenge of incorporating various DOE national laboratories and universities to effectively work together seems daunting. Yet, research and development efforts within space nuclear power, propulsion, and shielding communities require exactly this type of inter disciplinary and inter laboratory cooperation to succeed. If there is any lesson to be taken from the Apollo program it is that even Augean tasks (both technical and social) can be accomplished if all parties involved work together towards a common goal. Fortunately, such partnerships have been forged and continue to improve within the space nuclear community. The fact that the author was able to receive NASA GRC guidance while on a NASA MSFC funded GSRP fellow ship is unique in itself Center politic al rivalries could have easily derailed such a gradstudent exchange; however, t he relatively small and collaborative groups prevented center politics from becoming a major factor W hen considering the inter agency co operation between NASA ( GRC, MSFC ), DOE ( LANL, ORNL, SNL) and universities ( TAMU UF ) the success of this research pro ject becomes truly unique. This same model of collaboration should allow for s ubsequent space nuclear projects to share similar success It is the authors sincere hope that this level of cooperation within the space nuclear community will continue and accelerate on all levels; from initial investigations to flight qualification s and space flight operations. Ultimately, it is incumbent upon those managing the se experimental programs to determine the most cost effective methods to demonstrate performance, areas for improvement, and reliability63.

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230 Table 11 -1. Comparison of Select Radiation Test Facilities Particle Type Facility Test Location Requirements Safety Eval. (mo) Approval Period (mo) Safety & Approval Cost ($) Set Up, Design Cost ($) Irradiation Rate ($/t) Total Cost ($) Facility Usage (weeks) Photon University Sources Irradiation cell or beam port ~1 -2 ~1 < 1 k < 1 k ~0 100/h N/A N/A Sandia National Laboratories (SNL) Low Dose Facility71 Irradiation cell ~1 ~0.25 ~5 10 k ~5 k ~500/week N/A N/A *SNL Gamma Irradiation Facility (GIF)71 Irradiation cell ~1 ~0.25 ~10 k ~5 k ~200/h ~45 k ~2 *ORNL HFIR GIF72 Water immersion ~6 ~1 ~ 8 9 k ~2 k N/A ~40 k N/A Neutron SNL Annular Core Research Reactor (ACRR) 73 Beam port ~2 ~1 ~5 10 k ~5 k ~2.2 k/h N/A N/A *Ohio State Univ. Reactor74 Beam port, rabbit ~2 -3 ~1 ~ 70/h ~ 70/h ~ 260/h ~25 k ~4 *Texas A&M Univ. (TAMU) TRIGA Reactor75 Irradiation cell, water immersion, rabbit ~2 ~1 ~0.5 k ~0.5 k ~55/h ~8 k ~4 ORNL HFIR76 Water immersion or rabbit ~6 24 ~1 Design specific 100 k > 1 M 0 for open literature ~3 M N/A Idaho National Laboratory Advanced Test Reactor (ATR) 77 Water immersion or beam port ~12 ~1 Design specific >500 k ~100 500 k /cycle ~3 M N/A TAMU Cyclotron78 Beam port ~2 ~1 0 0 ~780/h N/A N/A Charged Particle Brookhaven National Laboratory NASA Space Radiation Laboratory (NSRL) 79 Beam port ~6 ~1 Design Specific Design Specific ~5 k/h N/A N/A = facilities already used in support of recent space nuclear power projects

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231 CHAPTER 12 CONCLUSIONS A variety of results were obtained from the series of experiments carried out with regard to the radiation tolerance of the Stirling alternator design. These include system level results from the SARTA radiation testing and from the coupon level mixed neutron and ray testing of Stirling alternator candidate organic materials. SARTA Irradiation Testing and Evaluation The SARTA was operated at 90C to approximately 22 Mrad and at 125C for approximately 18 Mrad with no significant degradation of operating parameters. Postirradiation electrical integrity measurements indicated some discernable damage to the alternator which occurred somewhere between approximately 30 and 40 Mrad. The SARTA performed well throughout radiation exposure tests as confirmed by minimal change in operational parameters, no detectable changes in waveforms, RGA scans, lack of debris, lack of deposits, and no major changes in internal component appearance. Internal components revealed minimal degradation after operating in the radiation environment Cracks were observed in PTFE wires insulation but did not appear to contribute to changes in system response. Slight appearance changes such as d iscoloration or rubbing were observed but are a consequence of n ormal operation. Chemical changes characterized by DSC and TGA testing found some differences between bakedout samples and samples taken from SARTA components However the SARTA experience extended operation at elevated temperature so these results ca n only serve as guidance for future testing. PTFE wire insulation and Kynar heat shrink tubing were the organic materials which experienced the most notable changes and may warrant close r scrutiny.

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232 Postoperation measurements on the alternator suggested a short occurred when the SARTA was removed from the test cell. Evaluation suggests that the short formed in the stator coil windings. Due to the nature of its construction, destructi ve evaluation of the coil did not seem feasible. Since the outer stator windings utilize a polyimide based coating a high radiation tolerance would be expected from this component Mixed Neutron Ray Irradiation and Evaluation of Candidate Materials The Stirling alternator candidate organic material samples were irradiated in a mixed neutron and gamma ray field to an estimated maximum FSP lifetime fluence of 1x1014 and 5x1014 n/cm2 a nd a corresponding dose of 1.355 and 5.38 Mrad while under ultra high pur ity helium at 125C and 15 0C respectively No statistically significant changes in sample diameter or thickness were observed when comparing irradiated and nonirradiated samples even with the same thermal histories. We can conclude that these candidate materials should be capable of maintaining their shape well within established dimensional tolerance requirements. A few of the material samples experienced an average weight change outside of experimental error; however, these changes can be attributed to artifacts not associated with the radiation exposure. A much larger of number samples is required in order to accurately address weight changes with any significant statistical relevance. FTIR spectra comparison between as received, bakedout, and baked out aged control samples with irradiated samples found little to no change in surface chemistry. Since the surface composition was subjected to the largest number of incident neutron and rays, it is highly unlikely that any changes would be determined through FTIR within the sample by evaluating sectioned specimens.

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233 Bulk thermal and chemical changes characterized by DSC, TGA and DMA evaluation found only minor differences between bakedout, bakedout aged and irradiated sample conditions. No significa nt shifts in Tg, Td, endothermic, and exothermic peak locations could be resolved. Those shifts that were found were within the calculated experimental error for each instrument. The degree of change was statistically insignificant and well within material design margins. Optical microscopy found no discernable changes in surface color or morphology when comparing as received, bakedout bakedout aged, and irradiated samples. SEM found no significant changes between control and irradiated Xylan coatings Oring compression set tests measured a slight change in recovery as a function of thermal history and to radiation history. A larger number of irradiated samples is required to increase statistical confidence of these findings. Although these changes were measureable, the magnitude of performance degradation was well within design criteria and unlikely to pose a credible failure mechanism. Lap shear axial tension tests results indicate some measureable changes in bond strength as a function of both thermal aging and irradiation. Both Hysol and H igh T emperature Hysol epoxies maintained the required bond integrity after irradiation. A much larger data set of irradiated samples is required in order to increase the statistical confidence of these initial findings for the Hysol and HT Hysol epoxies. Electrical resistivity measurements, although coarse in nature, show no appreciable change after irradiation. When considering the minimal change in properties observed in all the previous methods it is likely that change in electrical resistivity was insignificant. As with the results from the SARTA wire evaluation,

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234 insulation degradation will not pose a significant failure mechanism as long as cracking or flaking of the insulation (whether thermal and/or radiation induced) does not occur. In summary, organic material properties experienced no significant degradation Organic material candidates under consideration for Stirling alternator service should be able to meet service life requirements It is recommended that PTFE wire insulation and Kynar heat shrink tubing be replaced with radiation tolerant substitutes. Relevance of Nuclear an d Radiation Tests Research and development efforts within space nuclear power, propulsion, and shielding communities require inter disciplinary and inter laboratory cooperation to succeed. S pace nuclear systems must emphasize a hardware rich research and development strategy Testing is the only method that will provide a high degree of certainty with regard to material, component, or system performance C ost effective nuclear and radiation test s can be conducted by utilizing graduate student researcher s to bridge institutional gaps university facilities, and low cost national laboratory facilities. R elevant nuclear and radiation test results can be obtained in a cost effective and practical manner utilizing radiation sources that are not representative of the expected operating environment. The type of facility used to test the equipment in question depends on the maturity level of the hardware. Low maturity technology should undergo screening trials at relatively inexpensive facilities such as universities or low cost national laboratory facilities. Component or system level performance demonstrations should be conducted in an accelerated environment at low energy but high flux sources are av ailable at various university and national laboratories. Te st objectives will shift from screening and demonstration trials to flight qualification and acceptance trials as technology matures. F light qualification

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235 and acceptance trials should be conducted at facilities that offer more pr o to typic radiation environment s in order to build confidence in the systems being evaluated. The consequences for selecting an inappropriate facility at an inappropriate time will likely increas e costs by order s of magnitude. Ultimately, it is incumbent upon those managing these experimental programs to determine the most cost effective methods to demonstrate performance, areas for improvement, and reliability Recommendations for Future Work Several future stud ies to be conducted were identified throughout the literature review and testing process Recommendations for future investigation include : Continue testing of the Stirling alternator outer stator in order to repeat the electrical fault detected in the SARTA, isolate the failure mechanism, and implement design modifications. Test methods should include thermomechanical fatigue testing and possibly radiation testing under operating conditions. Determine if a dose or fluence threshold exists in order to achieve onset of degradation when irr adiating the candidate polymeric materials in an inert atmosphere at the appropriate operating temperature. Determine the dose and flux rate thresholds that will possibly induce radiation induced degradation in organic materials under appropriate atmospheric and temperature conditions Define the onset of irradiation induced degradation with increasing oxygen concentration at specific operating temperatures. Utilize XRD to independently verify and supplement organic crystallinity estimations obtained throug h DSC and DMA methods. Quantitative ly measure polymeric electrical resistance as a function of service life dose at maximum operating voltages. A factor of safety should be established to take into account the possibility of radiation induced conductivity. F ocus f uture component and system tests on transient analysis and aggressive nonnuclear environments as the most likely mode of failure80. Irradiate NdFeB and SmCo magnet grades under consideration for alternator use by both protons and neut ron s at expected operating temperatur es to quantify possible performance decrease during the proposed service life.

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236 Overall Findings The fluroelasomers (such as the Kalrez Oring s and Viton heat shrink tubing) demonstrated sufficient radiation tolerance to meet RPS and FSP requirements. P olyimide wire insulation, Silicone Oring, Xylan solid running surface, Hysol epoxy and HT Hysol epoxy also performed sufficiently well to state that they should meet service conditions without significant change in engineering properties. As mentioned previously, Kynar heat shrink tubing and PTFE wire insulation should be replaced with more radiation tolerant substitutes such as Polyimide, PVC, Viton, etc. A cost effective methodology to conduct accelerated radiation ser vice life tests was demonstrated. This method utilized two universities, two NASA centers, and three DOE national laboratories to develop these finding s. This inter agency and inter disciplinary cooperation serves as an excellent model for the level of c ooperation required by these same institutions to make space nuclear systems a reality. Th is project provided initial experimentally demonstrated evidence that the Stirling alternator s are capable of exceeding RPS and FSP radiation service life requirement s. These results should be used to advance hardware maturity, improve performance confidence in a radiation environment, and increase maximum fluence and dose limits Since it was demonstrated that the system can survive expected radiation conditions, this represents a significant savings of additional radiation qualification tests that must be conducted before flight units are utilized and improved confidence in the technology. The r esults presented in this document can be additionally utilized to provide experimentally derived benchmark s for models, identify radiation tolerant materials for next generation systems, and determined a first order lifetime prediction for an alternator in radiation conditions

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237 APPENDIX A SARTA TESTING Supplemental Figures Figure A -1. SARTA system instrumentation, power, and control schematic

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238 Figure A -2. The Sandia National Laboratory G amma I rradiation F acility (GIF) radial dose rate profile. Courtesy of Dr. Ross Rad el SNL49. Figure A -3. GIF axial dose rate profile. Courtesy of Dr. Ross Rad el SNL49.

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239 Figure A -4. RAD 1 Performance (90C, 30 minutes 81.82 rad/s, 0.147 Mrad) Figure A -5. RAD 1 Temperatures (90C, 30 minutes, 81.82 rad/s, 0.147 Mrad) 7.9 7.92 7.94 7.96 7.98 8 8.02 8.04 8.06 8.08 8.1 0 5 10 15 20 25 30 0 5 10 15 20 25 30 35 40 45 50 Stroke (mm) Power (W), Current (A), Voltage (VAC) Time (minutes) Power Current Voltage Stroke Source Up t = 12.4 mins Source Down t = 42.4 mins 480 485 490 495 500 505 510 80 82 84 86 88 90 92 94 96 0 5 10 15 20 25 30 35 40 45 50 Pressure (psig) Temperature (C) Time (minutes) PV B PV T Inner Iron Coil Pressure Source Up t = 12.4 mins Source Down t = 42.4 mins

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240 Figure A -6. RAD 2 Performance (90C, 30 minutes 75.94 rad/s, 0.137 Mrad) Figure A -7. RAD 2 Temperatures (90C, 30 minutes, 75.94 rad/s, 0.137 Mrad) 7.9 7.92 7.94 7.96 7.98 8 8.02 8.04 8.06 8.08 8.1 0 5 10 15 20 25 30 0 5 10 15 20 25 30 35 40 45 50 55 60 65 70 75 80 85 Stroke (mm) Power (W), Current (A), Voltage (VAC) Time (minutes) Power Current Voltage Stroke Source up t = 42.5 min Source down t = 72.5 min Start up t = 6.16 min Shut down t = 82.03 min 480 485 490 495 500 505 510 20 30 40 50 60 70 80 90 100 110 0 5 10 15 20 25 30 35 40 45 50 55 60 65 70 75 80 85 Pressure (psig) Temperature (C) Time (minutes) Pressure Vessel Bottom Temp Pressure Vessel Top Temp Inner Iron Temp Coil Temp Pressure SARTA Start up t = 6.16 min Source up t = 42.5 min Source down t = 72.5 min SARTA Shut down t = 82.03 min

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241 Figure A -8. RAD 3 Performance (90C, 60 minutes 98.74 rad/s, 0.335 Mrad) Figure A -9. RAD 3 Temperatures (90C, 60 minutes, 98.74 rad/s, 0.335 Mrad) 7.9 7.92 7.94 7.96 7.98 8 8.02 8.04 8.06 8.08 8.1 0 5 10 15 20 25 30 0 10 20 30 40 50 60 70 80 90 100 110 Stroke (mm) Power (W), Current (A), Voltage (VAC) Time (minutes) Power Current Voltage Stroke SARTA Shut down t = 106.5 min Source down t = 96.83 min Source up t = 26.99 min SARTA Start up t = 2.33 min 490 492 494 496 498 500 502 504 506 508 510 20 30 40 50 60 70 80 90 100 0 10 20 30 40 50 60 70 80 90 100 110 Pressure (psig) Temperature (C) Time (minutes) PV B PV T Inner Iron Coil Pressure

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242 Figure A 10. RAD 4 Performance (90C, 60 minutes, 94.38 rad/s, 0.340 Mrad) Figure A 11. RAD 4 Temperatur es (90 C, 60 minutes, 94.38 rad/s, 0.340 Mrad) 7.9 7.92 7.94 7.96 7.98 8 8.02 8.04 8.06 8.08 8.1 0 5 10 15 20 25 30 0 10 20 30 40 50 60 70 80 90 100 110 120 130 Stroke (mm) Power (W), Current (A), Voltage (VAC) Time (mintutes) Power Current Voltage Stroke SARTA Shut down t = 121.33 min Source down t = 114.75 min Source up t = 54.75 min SARTA Start up t = 15.16 min 490 492 494 496 498 500 502 504 506 508 510 20 30 40 50 60 70 80 90 100 0 10 20 30 40 50 60 70 80 90 100 110 120 130 Pressure (psig) Temperature (C) Time (minutes) PV B PV T Inner Iron Coil Pressure

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243 Figure A 12. RAD 5 Performance (90C, 210 minutes, 78.88 rad/s, 0.994 Mrad) Figure A 13. RAD 5 Temperatures (90C, 210 minute s, 78.88 rad/s, 0.994 Mrad)

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244 Figure A 14. RAD 6 Performance (90C, 160 minutes, 78.88 rad/s, 0.757 Mrad) Figure A 15. RAD 6 Temperatures (90C, 160 minute s, 78.88 rad/s, 0.757 Mrad) 7.9 7.92 7.94 7.96 7.98 8 8.02 8.04 8.06 8.08 8.1 0 5 10 15 20 25 30 0 20 40 60 80 100 120 140 160 180 200 Stroke (mm) Power (W), Current (A), Voltage (VAC) Time (minutes) Power Current Voltage Stroke Source up t = 37.75 min SARTA Start up t = 7.41 min Source down t = 198.0 min SARTA Shut down t = 202.33 min 480 485 490 495 500 505 510 515 520 525 530 20 30 40 50 60 70 80 90 100 0 20 40 60 80 100 120 140 160 180 200 Pressure (psig) Temperature (C) Time (minutes) PV B PV T Inner Iron Coil Pressure

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245 Figure A 16. RAD 7 Performance (90C, 270 minutes, 78.88 rad/s, 1.278 Mrad) Figure A 17. RAD 7 Temperatures (90C, 270 minute s, 78.88 rad/s, 1.278 Mrad)

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246 Figure A 18. RAD 8 Performance (90C, 5 minutes 859.39 rad/s, 0.258 Mrad) Figure A 19. RAD 8 Temperatures (90C, 5 minute s, 859.39 rad/s, 0.258 Mrad) 7.9 7.92 7.94 7.96 7.98 8 8.02 8.04 8.06 8.08 8.1 0 5 10 15 20 25 30 0 5 10 15 20 25 30 35 40 45 Stroke (mm) Power (W), Current (A), Voltage (VAC) Time (minutes) Power Current Voltage Stroke SARTA Start up t = 0.833 min Source up t = 32.667 min Source down t = 37.667 min SARTA Shut down t = 43.833 min 470 475 480 485 490 495 500 505 510 20 30 40 50 60 70 80 90 100 0 5 10 15 20 25 30 35 40 45 Pressure (psig) Temperature (C) Time (minutes) PV B Inner Iron Coil Pressure

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247 Figure A 20. RAD 9 Performance (90C, 5 minutes 847.75 rad/s, 0.254 Mrad) Figure A 21. RAD 9 Temperatures (90C, 5 minute s, 847.75 rad/s, 0.254 Mrad) 7.9 7.92 7.94 7.96 7.98 8 8.02 8.04 8.06 8.08 8.1 0 5 10 15 20 25 30 0 5 10 15 20 25 30 35 40 45 Stroke (mm) Power (W), Current (A), Voltage (VAC) Time (minutes) Power Current Voltage Stroke SARTA Start up t = 1.25 min Source up t = 30.08 min Source down t = 35.08 min SARTA Shut down t = 41.75 min 480 485 490 495 500 505 510 20 30 40 50 60 70 80 90 100 0 5 10 15 20 25 30 35 40 45 Pressure (psig) Temperature (C) Time (minutes) PV B PV T Inner Iron Coil Pressure

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248 Figure A 22. RAD 10 Performance (90C, 35 minutes, 853.57 rad/s, 1.793 Mrad) Figure A 23. RAD 10 Temperatures (90C, 35 minutes, 853.57 rad/s, 1.793 Mrad) 7.9 7.92 7.94 7.96 7.98 8 8.02 8.04 8.06 8.08 8.1 0 5 10 15 20 25 30 0 5 10 15 20 25 30 35 40 45 50 55 60 65 70 75 Stroke (mm) Power (W), Current (A), Voltage (VAC) Time (minutes) Power Current Voltage Stroke SARTA Start up t = 1.25 min Source up t = 31.08 min Source down t = 66.08 min SARTA Shut down t = 73.16 min 490 492 494 496 498 500 502 504 506 508 510 20 30 40 50 60 70 80 90 100 0 5 10 15 20 25 30 35 40 45 50 55 60 65 70 75 Pressure (psig) Temperature (C) Time (mintues) PV B PV T Inner Iron Coil Pressure

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249 Figure A 24. RAD 11 Performance (90C, 40 minutes, 853.57 rad/s, 2.049 Mrad) Figure A 25. RAD 11 Temperatures (90C, 40 minute s, 853.57 rad/s, 2.049 Mrad) 7.9 7.92 7.94 7.96 7.98 8 8.02 8.04 8.06 8.08 8.1 0 5 10 15 20 25 30 0 5 10 15 20 25 30 35 40 45 50 55 60 65 70 75 80 Stroke (mm) Power (W), Current (A), Voltage (VAC) Time (minutes) Power Current Voltage Stroke SARTA Start up t = 3.16 min Source up t = 30.16 min Source down t = 70.33 min SARTA Shut down t = 75.83 min 480 485 490 495 500 505 510 20 30 40 50 60 70 80 90 100 0 5 10 15 20 25 30 35 40 45 50 55 60 65 70 75 80 Pressure (psig) Temperature (C) Time (minutes) PV B Inner Iron Coil Pressure

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250 Figure A 26. RAD 12 Performance (90C, 80 minutes, 853.57 rad/s, 4.097 Mrad) Figure A 27. RAD 12 Temperatures (90C, 80 minute s, 853.57 rad/s, 4.097 Mrad) 7.9 7.92 7.94 7.96 7.98 8 8.02 8.04 8.06 8.08 8.1 0 5 10 15 20 25 30 0 10 20 30 40 50 60 70 80 90 100 110 Stroke (mm) Power (W), Current (A), Voltage (VAC) Time (minutes) Power Current Voltage Stroke SARTA Start up t = 0.41 min Source up t = 28.00 min Source down t = 108.25 min SARTA Shut down t = 113.50 min 480 485 490 495 500 505 510 20 30 40 50 60 70 80 90 100 0 10 20 30 40 50 60 70 80 90 100 110 Pressure (psig) Temperature (C) Time (minutes) PV T PV B Inner Iron Coil Pressure

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251 Figure A 28. RAD 13 Performance (90C, 195 minute s, 853.57 rad/s, 9.987 Mrad) Figure A 29. RAD 13 Temperatures (90C, 195 minutes 853.57 rad/s, 9.987 Mrad)

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252 Figure A 30. RAD 14 Performance (120C, 60 minute s, 78.88 rad/s, 0.284 Mrad) Figure A 31. RAD 14 Temperatures (125C, 60 minute s 78.88 rad/s, 0.284 Mrad) 7.9 7.95 8 8.05 8.1 8.15 8.2 0 5 10 15 20 25 30 35 0 10 20 30 40 50 60 70 80 90 100 110 120 Stroke (mm) Power (W), Current (A), Voltage (VAC) Time (minutes) Power Current Voltage Stroke SARTA Start up t = 1.41 min Source up t = 54.16 min Source down t = 114.41 min SARTA Shut down t = 119.83 min 480 485 490 495 500 505 510 20 40 60 80 100 120 140 0 10 20 30 40 50 60 70 80 90 100 110 120 Pressure (psig) Temperature (C) Time (minutes) Inner Iron Coil Pressure

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253 Figure A 32. RAD 15 Performance (125C, 20 minute s, 853.57 rad/s, 1.024 Mrad) Figure A 33. RAD 15 Temperatures (125C, 20 minute s 853.57 rad/s, 1.024 Mrad) 7.9 7.95 8 8.05 8.1 8.15 0 5 10 15 20 25 30 35 0 10 20 30 40 50 60 70 Stroke (mm) Power (W), Current (A), Voltage (VAC) Time (minutes) Power Current Voltage Stroke SARTA Start up t = 1.41 min Source up t = 44.91 min Source down t = 62.25 min SARTA Shut down t = 70.41 min 480 485 490 495 500 505 510 20 40 60 80 100 120 140 0 10 20 30 40 50 60 70 Pressure (psig) Temperature (C) Time (minutes) PV B PV T Inner Iron Coil Pressure

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254 Figure A 34. RAD 16 Performance (125C, 40 minute s, 853.57 rad/s, 2.049 Mrad) Figure A 35. RAD 16 Temperatures (125C, 40 minute s 853.57 rad/s, 2.049 Mrad) 7.9 7.95 8 8.05 8.1 8.15 0 5 10 15 20 25 30 0 10 20 30 40 50 60 70 80 Stroke (mm) Power (W), Current (A), Voltage (VAC) Time (minutes) Power Current Voltage Stroke SARTA Start up t = 0.33 min Source up t = 36.67 min Source down t = 76.83 min SARTA Shut down t = 82.08 min 480 485 490 495 500 505 510 20 40 60 80 100 120 140 0 10 20 30 40 50 60 70 80 Pressure (psig) Temperature (C) Time (minutes) PV B PV T Inner Iron Coil Pressure

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255 Figure A 36. RAD 17 Performance (125C, 80 minute s, 853.57 rad/s, 4.097 Mrad) Figure A 37. RAD 17 Temperatures (125C, 80 minute s 853.57 rad/s, 4.097 Mrad) 7.9 7.92 7.94 7.96 7.98 8 8.02 8.04 8.06 8.08 8.1 0 5 10 15 20 25 30 0 10 20 30 40 50 60 70 80 90 100 110 120 Stroke (mm) Power (W), Current (A), Voltage (VAC) Time (minutes) Power Current Voltage Stroke SARTA Start up t = 0.25 min Source up t = 34.58 min Source down t = 114.75 min SARTA Shut down t = 119.83 min 480 485 490 495 500 505 510 20 40 60 80 100 120 140 0 10 20 30 40 50 60 70 80 90 100 110 120 Pressure (psig) Tempreature (C) Time (minutes) PV B Inner Iron Coil Pressure

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256 Figure A 38. RAD 18 Performance (125C, 200 minutes 853.57 rad/s, 10.243 Mrad) Figure A 39. RAD 18 Temperatures (125C, 200 minutes, 853.57 rad/s, 10.243 Mrad)

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257 Figure A 40. SARTA v oltage vs. time results for operation at 90C. Figure A 41. SARTA c urrent vs. time for operation at 90C. Time (min) 0 50 100 150 200 250 Voltage (VAC) 8.10 8.15 8.20 8.25 8.30 Nom. Voltage 90 C RAD 2 RAD 3 RAD 1 RAD 4 RAD 5 RAD 6 RAD 7 RAD 8 RAD 9 RAD 10 RAD 11 RAD 12 RAD 13 Time (min) 0 50 100 150 200 250 Current (A) 4.50 4.55 4.60 4.65 4.70 Nom. Current 90 C RAD 2 RAD 3 RAD 1 RAD 4 RAD 5 RAD 6 RAD 7 RAD 8 RAD 9 RAD 10 RAD 11 RAD 12 RAD 13

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258 Figure A 42. SARTA p ower vs. time results for oper ation at 90 C tests Figure A 43. SARTA v oltage vs. time results for operation at 125C. Time (minutes) 0 50 100 150 200 250 Power (W) 26.0 26.5 27.0 27.5 Nom Power RAD 1 RAD 2 RAD 3 RAD 4 RAD 5 RAD 6 RAD 7 RAD 8 RAD 9 RAD 10 RAD 11 RAD 12 RAD 13 Time (minutes) 0 50 100 150 200 Voltage (VAC) 8.00 8.05 8.10 8.15 8.20 Nom Voltage RAD 14 RAD 15 RAD 16 RAD 17 RAD 18

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259 Figure A 44. SARTA c urrent vs. time results for operation at 125C. Figure A 45. SARTA p ower vs. time results for operation at 125 C. Time (minutes) 0 50 100 150 200 Current (A) 4.60 4.65 4.70 4.75 4.80 Nom Current RAD 14 RAD 15 RAD 16 RAD 17 RAD 18 Time (minutes) 0 50 100 150 200 Power (W) 27.0 27.5 28.0 28.5 Nom Power RAD 14 RAD 15 RAD 16 RAD 17 RAD 18

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260 Figure A 46. SARTA a verage power factor as a function of dose for 90C tests Figure A 47. SARTA a verage power factor as a function of dose for 125C tests 0.703 0.704 0.705 0.706 0.707 0.708 0.709 0.71 0 2 4 6 8 10 12 14 16 18 20 22 Average Power Factor Dose (Mrad) RAD 1 RAD 2 RAD 3 RAD 4 RAD 5 RAD 6 RAD 7 RAD 8 RAD 9 RAD 10 RAD 11 RAD 12 RAD 13 0.727 0.7272 0.7274 0.7276 0.7278 0.728 0.7282 0.7284 22 24 26 28 30 32 34 36 38 40 Average Power Factor Dose (Mrad) RAD 14 RAD 15 RAD 16 RAD 17 RAD 18

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261 S tatistical Analysis of SARTA C onfidence I ntervals The nominal operating limits are established by first determini ng the mean and the standard deviation for the steady state preexposure data runs. We utilize the student Tdistribution, where the population standard deviation is unknown but the sample standard deviation is known, to determine the multiplier necessary to produce a 99 % total confidence interval. We assume the distribution about the population mean was normally distributed such that the confidence interval could be expressed by the function shown in equation B 1. 5a5a /2 (B 1) = population mean t = student t value at a specified value of for t distribution with ( n1) degrees of freedom = significance level = standard deviation of the sample measurements n = sample size In order to accurately utilize such a function to determine measureabl e off nominal operation requires a substantial amount of nominal operation data. Having sufficient data will improve the counting statistics required to develop confidence in the statistical analysis being applied. We apply a multiplier of 3.499, which yields a 99.5% confidence interval Using this confidence interval we can essentially determin e how likely a measured value is deviating beyond the normal operational range. In the case of the SARTA operation measured values 3.499 above or below the no minal value is indicative of actually observable off nominal operation.

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262 APPENDIX B MIXED NEUTRON GAMMA TESTING Figure B-1. Mixed neutrongamma ray test system instrumentation, power, and control schematic

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263 APPENDIX C SARTA DISASSEMBLY AN D INSPECTION Table C-1. SARTA Organic Material Samples SARTA Sample Location Material W (g) L(mm) Dia (%) R coil heat shrink Trans. Plate Kynar 0.0173 0.013 L coil heat shrink Trans. Plate Kynar 0.0703 0.013 R coil lead wire Trans. Plate PTFE 0.3989 31 L coil lead wire Trans. Plate PTFE 0.3454 26 R FLDT heat shrink Trans. Plate Viton 0.0751 14 0.128 L FLDT heat shrink Trans. Plate Viton 0.0067 8 0.05 R FLDT wire Trans. Plate PTFE 0.0261 45 0.016 L FLDT wire Trans. Plate PTFE 0.0336 29 Inner piston O ring Trans. Plate Viton 0.1116 PTFE wire flakes Outer Stator PTFE 0.0319 23 R coil heat shrink Outer Stator Viton 0.1857 19 L coil heat shrink Outer Stator Viton 0.2163 20 R coil lead wire Outer Stator PTFE 0.2243 22 L coil lead wire Outer Stator PTFE 0.2588 21 Xylan shavings Piston Coat Xylan 0.0039 PV O ring Cylinder Silicone 0.7794 0.105

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264 A nalysis of Unknown Residue Deposit on SARTA Attachment Flange Figure C 1 illustrates the FTIR comparison of the unknown residue deposit on the SARTA flange to the FTIR spectra of a Hysol epoxy standard reference sample. The spectra shows close agreement, however, the trends do not perfectly match (in terms of wavelength and intensity). This slight mismatch indicates that either there are multiple chemical components in the residue not referenced in the standard or different curing conditions caused variation in the polymer reactants observed. The different c uring conditions is the most plausible explanation since the SARTA and the Hysol reference sample had far different thermal histories, therefore, different curing states. Figure C -1. FTIR spectra comparison of the unknown residue deposit and Hysol epoxy reference. C ourtesy of Dr. Kenneth Street, NASA G RC81. From spectra comparison there is a reasonably good match to assume that H ysol (or some very closely related polymer) constitute s a m ajor component of the d eposit.

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265 APPENDIX D MIXED NEUTRON & GAMMA RAY POST IRRADIATION MATERIAL S EVALUATION N eutron Activation Analysis of Irradiated Samples The ray spectroscopy results are shown in Table D1 to D -9. When background spectra (Table D 8) are compared to the sample spectra we see low isotopic activities, with some i sotopes having low confidence factors indicating thei r unlikely presence. Table D-1 Nuke 4 Silicone Oring Gamma Ray Spectroscopic Elemental Abundance Analysis C ourtesy of Dr. Latha Vasudevan, TAMU56. Nuclide Confidence E (keV) Yield (%) A (t) Uncertainty NA 22 1 1274.54* 99.94 2.88E 04 2.40E 05 K 40 1 1460.81* 10.67 1.53E 03 1.77E 04 CO 60 0.994 1173.22* 100 4.68E 05 9.99E 06 1332.49* 100 6.49E 05 1.21E 05 BI 211 0.368 72.87* 1.2 3.09E 03 4.83E 04 351.10* 12.2 1.83E 04 7.66E 05 404.8 4.1 426.9 1.9 831.8 3.3 BI 214 0.337 609.31* 46.3 9.95E 05 1.69E 05 768.36 5.04 806.17 1.23 934.06 3.21 1120.29 15.1 1155.19 1.69 1238.11 5.94 1280.96 1.47 1377.67 4.11 1385.31 0.78 1401.5 1.39 1407.98 2.48 1509.19 2.19 1661.28 1.15 1729.6 3.05 1764.49* 15.8 3.45E 04 7.67E 05 1847.44 2.12 2118.54 1.21 = Energy line found in the spectrum

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266 Table D-2 Nuke 4 Kalrez O ring Gamma Ray Spectroscopic Elemental Abundance Analysis Courtesy of Dr. Latha Vasudevan, TAMU56. Nuclide Confidence E (keV) Yield (%) A (t) Uncertainty NA 22 1 1274.54* 99.94 2.88E 04 2.40E 05 K 40 1 1460.81* 10.67 1.53E 03 1.77E 04 CO 60 0.994 1173.22* 100 4.68E 05 9.99E 06 1332.49* 100 6.49E 05 1.21E 05 BI 211 0.368 72.87* 1.2 3.09E 03 4.83E 04 351.10* 12.2 1.83E 04 7.66E 05 404.8 4.1 426.9 1.9 831.8 3.3 BI 214 0.337 609.31* 46.3 9.95E 05 1.69E 05 768.36 5.04 806.17 1.23 934.06 3.21 1120.29 15.1 1155.19 1.69 1238.11 5.94 1280.96 1.47 1377.67 4.11 1385.31 0.78 1401.5 1.39 1407.98 2.48 1509.19 2.19 1661.28 1.15 1729.6 3.05 1764.49* 15.8 3.45E 04 7.67E 05 1847.44 2.12 2118.54 1.21 = Energy line found in the spectrum

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267 Table D-3. Nuke 4 Viton Heat Shrink Gamma Ray Spectroscopic Elemental Abundance Analysis Courtesy of Dr. Latha Vasudevan, TAMU56. Nuclide Confidence E (keV) Yield (%) A (t) ( Uncertainty NA 22 1 1274.54* 99.94 2.72E 04 2.55E 05 K 40 0.984 1460.81* 10.67 1.32E 03 1.75E 04 CO 60 0.985 1173.22* 100.00 6.04E 05 1.27E 05 1332.49* 100.00 5.68E 05 1.15E 05 ZN 65 0.998 1115.52* 50.75 5.48E 05 2.17E 05 RU 103 0.903 497.08* 89.00 4.17E 04 4.40E 05 610.33* 5.60 7.71E 04 1.34E 04 SB 124 0.887 602.71* 97.87 2.10E 04 1.92E 05 645.85* 7.26 2.37E 04 1.21E 04 709.31 1.42 713.82 2.38 722.78* 11.10 1.95E 03 1.92E 04 968.20 1.92 1045.16 1.86 1325.50 1.50 1355.24 1.00 1368.21 2.51 1436.60 1.14 1691.02* 49.00 2.07E 04 3.21E 05 = Energy line found in the spectrum Table D-4. Nuke 4 Kynar Heat Shrink Gamma Ray Spectroscopic Elemental Abundance Analysis Courtesy of Dr. Latha Vasudevan, TAMU56. Nuclide Confidence E (keV) Yield (%) A (t) Uncertainty NA 22 0.997 1274.54* 99.94 2.59E 04 2.43E 05 K 40 1 1460.81* 10.67 1.61E 03 1.81E 04 CO 60 0.991 1173.22* 100 6.63E 05 1.35E 05 1332.49* 100 5.44E 05 9.90E 06 BI 211 0.367 72.87* 1.2 2.66E 03 4.43E 04 351.10* 12.2 1.56E 04 6.09E 05 404.8 4.1 426.9 1.9 831.8 3.3 = Energy line found in the spectrum

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268 Table D-5. Nuke 4 Polyimide Insulated Cu Wire Gamma Ray Spectroscopic Elemental Abundance Analysis Courtesy of Dr. Latha Vasudevan, TAMU56. Nuclide Confidence E (keV) Yield (%) A (t) Uncertainty NA 22 1 1274.54* 99.94 2.53E 04 2.28E 05 K 40 0.997 1460.81* 10.67 1.54E 03 1.73E 04 CO 60 1 1173.22* 100 5.40E 04 3.76E 05 1332.49* 100 5.77E 04 3.83E 05 AG 110M 0.546 446.8 3.64 620.35 2.77 657.75* 94.4 7.57E 05 1.62E 05 677.61 10.68 686.99 6.47 706.67 16.68 744.26 4.64 763.93* 22.28 6.76E 05 4.05E 05 818.02 7.3 884.67* 72.6 1.26E 04 1.99E 05 937.48* 34.2 5.32E 05 3.20E 05 1384.27 24.26 1475.76 3.97 1505 13.06 1562.27 1.18 = Energy line found in the spectrum

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269 Table D-6. Nuke 4 PTFE Insulated Cu Wire Gamma Ray Spectroscopic Elemental Abundance Analysis Courtesy of Dr. Latha Vasudevan, TAMU56. Nuclide Confidence E (keV) Yield (%) A (t) Uncertainty NA 22 0.999 1274.54* 99.94 1.98E 04 5.55E 05 K 40 0.994 1460.81* 10.67 1.22E 03 3.41E 04 K 42 0.987 1524.67* 17.90 2.35E 04 1.86E 04 CO 60 0.995 1173.22* 100.00 2.42E 04 6.05E 05 1332.49* 100.00 1.45E 04 5.27E 05 SR 92 0.717 241.52 3.00 430.56 3.30 953.32 3.60 1142.30 2.90 1383.94* 90.00 2.09E 02 2.48E 03 AG 110M 1 446.80* 3.64 5.78E 02 5.13E 03 620.35* 2.77 6.25E 02 4.42E 03 657.75* 94.40 6.65E 02 4.02E 03 677.61* 10.68 6.24E 02 3.92E 03 686.99* 6.47 6.83E 02 4.46E 03 706.67* 16.68 6.52E 02 3.99E 03 744.26* 4.64 6.62E 02 4.44E 03 763.93* 22.28 6.57E 02 3.99E 03 818.02* 7.30 6.08E 02 3.85E 03 884.67* 72.60 6.65E 02 4.01E 03 937.48* 34.20 6.64E 02 3.83E 03 1384.27* 24.26 7.28E 02 3.00E 03 1475.76* 3.97 7.78E 02 3.70E 03 1505.00* 13.06 7.17E 02 3.03E 03 1562.27* 1.18 8.60E 02 5.70E 03 Au 198 0.999 411.80* 95.00 1.07E 05 6.00E 05 = Energy line found in the spectrum

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270 Table D-7. Nuke 4 Xylan Coated Al Plate Gamma Ray Spectroscopic Elemental Abundance Analysis Courtesy of Dr. Latha Vasudevan, TAMU56. Nuclide Confidence E (keV) Yield (%) A (t) Uncertainty NA 22 0.996 1274.54* 99.94 2.79E 04 2.46E 05 K 40 0.997 1460.81* 10.67 1.55E 03 1.74E 04 SC 46 0.999 889.25* 99.98 8.75E 05 1.69E 05 1120.51* 99.99 1.01E 04 1.23E 05 CR 51 1 320.08* 9.83 2.49E 02 2.05E 03 MN 54 1 834.83* 99.97 4.49E 04 3.61E 05 CO 58 1 810.76* 99.40 3.17E 04 2.82E 05 FE 59 0.869 142.65 1.03 192.34 3.11 1099.22* 56.50 2.62E 04 3.41E 05 1291.56* 43.20 2.33E 04 3.60E 05 CO 60 0.999 1173.22* 100.00 1.12E 04 1.69E 05 1332.49* 100.00 1.69E 04 1.83E 05 ZN 65 1 1115.52* 50.75 3.99E 04 3.59E 05 MO 99 0.759 140.51* 88.70 8.92E 05 1.49E 05 181.06 6.20 366.43 1.37 739.58 12.80 778.00 4.50 TC 99M 1 140.51* 89.07 9.13E 05 1.53E 05 = Energy line found in the spectrum

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271 Table D-8. TAMU Background Gamma Ray Spectroscopic Elemental Abundance Analysis Courtesy of Dr. Latha Vasudevan, TAMU56. Nuclide Confidence E (keV) Yield (%) A (t) Uncertainty NA 22 0.999 1274.54* 99.94 2.60E 04 1.85E 05 K40 1 1460.81* 10.67 1.66E 03 1.36E 04 CO 60 0.995 1173.22* 100 5.30E 05 7.52E 06 1332.49* 100 5.58E 05 8.12E 06 BI 211 0.362 72.87* 1.2 2.05E 03 3.10E 04 351.10* 12.2 3.02E 04 5.80E 05 404.8 4.1 426.9 1.9 831.8 3.3 BI 214 0.384 609.31* 46.3 8.77E 05 1.22E 05 768.36 5.04 806.17 1.23 934.06 3.21 1120.29 15.1 1155.19 1.69 1238.11 5.94 1280.96 1.47 1377.67* 4.11 1.79E 04 9.42E 05 1385.31 0.78 1401.5 1.39 1407.98 2.48 1509.19 2.19 1661.28 1.15 1729.6 3.05 1764.49* 15.8 2.66E 04 5.01E 05 1847.44 2.12 2118.54 1.21 PB214 0.484 74.81* 6.33 6.43E 04 8.32E 05 77.11 10.7 87.2 3.7 89.8 1.03 241.98 7.49 295.21* 19.2 3.74E 06 2.13E 05 351.92* 37.2 9.92E 05 1.90E 05 785.91 1.1 = Energy line found in the spectrum

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272 Table D -9. Comparative Sample and TAMU Background Spectroscopic Analysis. Material Irradiated Sample Counts Background Counts Nuclide Confidence A (t) Nuclide Confidence A (t) Silicone NA 22 1 2.88E 04 NA 22 0.999 2.60E 04 O ring K 40 1 1.53E 03 K 40 1 1.66E 03 CO 60 0.994 4.68E 05 CO 60 0.995 5.30E 05 BI 211 0.368 3.09E 03 BI 211 0.362 2.05E 03 BI 214 0.337 3.45E 04 BI 214 0.384 8.77E 05 PB 214 0.484 6.43E 04 Kalrez NA 22 1 2.73E 04 NA 22 0.999 2.60E 04 O ring K 40 1 1.53E 03 K 40 1 1.66E 03 CO 60 0.994 6.46E 05 CO 60 0.995 5.30E 05 BI 211 0.368 1.82E 03 BI 211 0.362 2.05E 03 BI 214 0.337 1.37E 04 BI 214 0.384 8.77E 05 PB 214 0.484 6.43E 04 Viton Heat NA 22 1 2.72E 04 NA 22 0.999 2.60E 04 Shrink K 40 0.984 1.32E 03 K 40 1 1.66E 03 Tube CO 60 0.985 6.04E 05 CO 60 0.995 5.30E 05 BI 211 0.362 2.05E 03 BI 214 0.384 8.77E 05 PB 214 0.484 6.43E 04 ZN 65 0.998 5.48E 05 RU 103 0.903 7.71E 04 SB 124 0.887 1.95E 03 Kynar NA 22 0.997 2.59E 04 NA 22 0.999 2.60E 04 Heat K 40 1 1.61E 03 K 40 1 1.66E 03 Shrink CO 60 0.991 6.63E 05 CO 60 0.995 5.30E 05 Tube BI 211 0.367 2.66E 03 BI 211 0.362 2.05E 03 BI 214 0.384 8.77E 05 PB 214 0.484 6.43E 04 Polyimide NA 22 1 2.53E 04 NA 22 0.999 2.60E 04 Coated K 40 0.997 1.54E 03 K 40 1 1.66E 03 Copper CO 60 1 5.77E 04 CO 60 0.995 5.30E 05 Wire BI 211 0.362 2.05E 03 BI 214 0.384 8.77E 05 PB 214 0.484 6.43E 04 AG 110M 0.546 1.26E 04 PTFE NA 22 0.999 1.98E 04 NA 22 0.999 2.60E 04 Coated K 40 0.994 1.22E 03 K 40 1 1.66E 03 Copper K 42 0.987 2.35E 04 Wire CO 60 0.995 2.42E 04 CO 60 0.995 5.30E 05

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273 Table D 9. Continued. Material Irradiated Sample Counts Background Counts Nuclide Confidence A (t) Nuclide Confidence A (t) BI 211 0.362 2.05E 03 BI 214 0.384 8.77E 05 PB 214 0.484 6.43E 04 SR 92 0.717 2.09E 02 AG 110M 1 8.60E 02 Au 198 0.999 1.07E 05 Xylan NA 22 0.996 2.79E 04 NA 22 0.999 2.60E 04 Coated K 40 0.997 1.55E 03 K 40 1 1.66E 03 Aluminum CO 60 0.999 1.69E 04 CO 60 0.995 5.30E 05 Plate BI 211 0.362 2.05E 03 BI 214 0.384 8.77E 05 PB 214 0.484 6.43E 04 CR 51 1 2.49E 02 MN 54 1 4.49E 04 CO 58 1 3.17E 04 FE 59 0.869 2.62E 04 SC 46 0.999 8.75E 05 ZN 65 1 3.99E 04 MO 99 0.759 8.92E 05 TC 99M 1 9.13E 05 Hysol NA 22 0.999 2.60E 04 Epoxy Lap K40 0.992 1.60E 03 K40 1 1.66E 03 Shear CO 60 1 1.58E 02 CO 60 0.995 5.30E 05 BI 211 0.362 2.05E 03 BI 214 0.384 8.77E 05 PB 214 0.484 6.43E 04 CO 58 0.999 1.16E 03 FE 59 0.867 2.79E 03 ZN 65 0.971 2.48E 02 KR 88 0.345 1.50E 01 NB 95 0.954 1.56E 02 RH 105 0.601 6.46E 03 SN 113 0.848 1.56E 02 SC 46 1 1.14E 01 CR 51 0.962 1.56E 02 SB 124 0.712 1.65E 03

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274 Table D9. Continued. Material Irradiated Sample Counts Background Counts Nuclide Confidence A (t) Nuclide Confidence A (t) H T Epoxy NA 22 0.999 2.60E 04 Lap Shear K 40 0.993 1.35E 03 K 40 1 1.66E 03 CO 60 1 2.27E 02 CO 60 0.995 5.30E 05 BI 211 0.362 2.05E 03 BI 214 0.384 8.77E 05 PB 214 0.484 6.43E 04 SC 46 1 1.31E 01 CR 51 0.969 1.88E 02 CO 58 1 1.26E 03 ZN 65 0.974 3.68E 02 KR 88 0.345 2.67E 01 NB 95 0.955 2.35E 02 RH 105 0.593 9.79E 03 SN 113 0.847 2.35E 02 SB 124 0.726 4.74E 03 S upplemental SEM EDS D ata A B Figure D1. Comparison of unirradiated Xylan SEM images A) secondary electron (SE) and B) Back Scattered Electron (BSE) images at 500x.

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275 A B Figure D2. Comparison of unirradiated Xylan SEM images A) secondary electron (SE) and B) Back Scattered Electron (BSE) images at 10,000x. Figure D3. Comparison of irradiated (NUKE 4) Xylan SEM images secondary electron (SE) image at 5,000x.

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276 A B C Figure D4. Comparison of Xylan EDS spectra referencing Figure D 2B callout locations for A) EDS A, B) EDSB, C)EDS C.

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277 A B C Figure D5. Comparison of Xylan EDS spectra referencing Figure D 3 callout locations for A) EDS A, B) EDSB, C)EDS C.

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278 E lectrical R esistivity M easurement Data Table D 10. Qualitative electrical resistivity measurements of Viton and Kynar heat shrink tubing, and PTFE and Polyimide wire insulation. Resistance at Voltage (M ) Sample Condition 250 V 500 V 1000 V Viton BOA 1 > 1000 > 2000 > 4000 BOA 2 > 1000 > 2000 > 4000 BOA 3 > 1000 > 2000 > 4000 BOA 4 > 1000 > 2000 > 4000 N1 > 1000 > 2000 > 4000 N2 > 1000 > 2000 > 4000 N3 > 1000 > 2000 > 4000 N4 > 1000 > 2000 > 4000 Kynar BOA 1 > 1000 > 2000 > 4000 BOA 2 > 1000 > 2000 > 4000 BOA 3 > 1000 > 2000 > 4000 BOA 4 > 1000 > 2000 > 4000 N1 > 1000 > 2000 > 4000 N2 > 1000 > 2000 > 4000 N3 > 1000 > 2000 > 4000 N4 > 1000 > 2000 > 4000 PTFE BOA 1 > 1000 > 2000 > 4000 BOA 2 > 1000 > 2000 > 4000 BOA 3 > 1000 > 2000 > 4000 BOA 4 > 1000 > 2000 > 4000 N1 > 1000 > 2000 > 4000 N2 > 1000 > 2000 > 4000 N3 > 1000 > 2000 > 4000 N4 > 1000 > 2000 > 4000 Polyimide BOA 1 > 1000 > 2000 > 4000 BOA 2 > 1000 > 2000 > 4000 BOA 3 > 1000 > 2000 > 4000 BOA 4 > 1000 > 2000 > 4000 N1 > 1000 > 2000 > 4000 N2 > 1000 > 2000 > 4000 N3 > 1000 > 2000 > 4000 N4 > 1000 > 2000 > 4000

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279 ORing C ompression Set M ethodology The Oring compression set testing procedure was conducted in accordance to the American Society of Testing and Materials ( ASTM ) D395, test method B standard82. The method calls for Orings to be compressed at 25% of the original Oring thickness for 70 hours at ambient temperature conditions (23oC). The compression set can be defined according to Equation D 1. ( % ) = 5a5a 5a5a 5a5a 0 5a5a5`5` 5V5V5V5V 100 (Eq D 1) where: CB = Compression set to = original specimen thickness tf = final specimen thickness tshim = 0.75 to = shim thickness After the Orings are compressed for 70 hours the pressure is unloaded. The samples experience dimensional recovery for a period of 30 minutes at which time the Oring thickness is measured using high accuracy calipers. The thickness is measured at several locations throughout the sample and measured several times throughout a period of between 24 and 48 hours after the pressure is unloaded. The lower the value for CB the better the performance of the Oring. Several different types of Orings are tested to include as received, bakedout aged, irradiated in a ray radiation environment at 125C and 150C, and irradiated in a mixed neutron and ray radiation environment in an ultrahigh purity helium environment at 125C and 150 C.

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280 LIST O F REFERENCES 1. J.A. ANGELO and D. BUDEN, Space Nuclear Power Ch 9, Krieger Publishing, Malabar, FL (1985). 2. U.S. Department of Energy. Space Radioisotope Power Systems: Multi Mission Radioisotope Thermoelectric Generator. Cited: August, 30 2007, http://www.ne.doe.gov/pdfFiles/MMRTG.pdf. 3. U.S. Department of Energy. Radioisotope Power Systems: General Purpose Heat Source. Cited: August, 30 2007, http://www.ne.doe.gov/space/neSpace2g.html 4. M. E. ANERSON, Neutron Flux,Spectrum, and Dose Equivalent Measurements for a 4500W(th) 238PuO2 General Purpose Heat Source. Monsanto Research Corporation. Springfield, VA, National Technical Information Service (1985). 5. L. G. THIEME AND J. G. SCHREIBER, Advanced Technology Development for Stirling Convertors, NASA Glenn Research Center, NASA/TM 2004 213186 (2004). 6. J.G. SCHRIEBER AND L.G. THIEME, Overview of NASA GRC Stirling Technology Development, NASA Glenn Research Center, NASA/TM 2004213186 (2004). 7. JUN, H. B. GARRETT, R EVANS, High Energy Trapped Particle Environments at Jupiter: An Update, IEEE Transactions on Nuclear Science, 52, 6 2282 (2005). 8. C. PARANICAS, et al., Europas Near Surface Radiation Environment American Geophysical Union Research Letters 34, 1, L15102 (2007). 9. I. JUN AND H. B. GARRET, Comparison of High Energy Trapped Particle Environments at the Earth and Jupiter Radiation Protection Dosimetry 116, 14, 50 (2005). 10. E. L. GOLLIHER and S. V. PEPPER, Organic materials ionizing radiation susceptibility for the Outer Planet/Solar Probe radioisotope power source, 35th Energy Conversion Engineering Conference &Exhibit, Las Vegas, NV, July 2428, 2000, p. 1491, American Institute of Aeronautics and Astronautics (2000). 11. L. MASON, Progress Made in Power Conversion Technologies For Fission Surface Power NASA Glenn Research Center, NASA/TM 2009022083 (200 8 ). 12. J. TURNER, Atoms, Radiation, and Radiation Protection, Ch 5, John Wiley & Sons Inc, New York, NY (1995).

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287 BIOGRAPHICAL SKETCH Omar Roberto Mireles was born in El Paso, TX Mr. Mireles was raised in southern New Mexico and graduated from Santa Teresa High School in May 1997. He obtained B achelor of S cience degrees in mechanical engineering and applied mathematics from New Mexico State University (NMSU) in May 2002, a Master of S cience in mechanical engineering from the Georgia Institute of Technology in May 2004, a M aster of S cience in nuclear and radiological engineering from the University of Florida in August 2007, and a D octor of Philosophy in nuclear and radiological engineering with a minor in materials science in May 2010 also from the University of Florida From May 1998 to July 2000, M r. Mireles worked as an undergraduate research associate with the NMSU astronomy department on e clipsing binary star system s using optical telescopes and modeling. His efforts r esulted in an undergraduate honors thesis Modeling Algol Binaries Through B lue V isual R ed and I nfrared Light Curve Analysis. Mr. Mireles completed two c oop tours fr om July to December 2000 and from May to August 2001 at the NASA Jet Propulsion Laboratory. His duties included performance evaluation of an ultrasonic rock abrasion tool, contamination studies for sample return, Mars dri lling mission concepts, and concur rent design methodologies Mr. Mireles e stablished NMSUs first team selected to fly experiments in microgravity aboard a KC 135 as part of the NASA Reduced Gravity Student Flight Opportunities Program in August 2001 and March 2002. Mr. Mireles also participated in the 2002 NASA Academy at Goddard Space Flight Center where he d esign ed Wide Field Camera 3 hardware for the Hubble Space Telescopes fourth servicing mission. Mr. Mireles completed three graduate research tours at the NASA Marshall Space Fl ight Center (MSFC) from May 2003 to August 2005 investigating high temperature

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288 materials compatibility, alkali metal heat pipe testing, and design /testing of inert gas/vacuum systems. These efforts r esulted in his first masters thesis Non Nuclear Materi als Compatibility Testing of Niobium 1% Zirconium & 316 Stainless Steel for Space Fission Reactor Applications. During this time he also served as a mentor to a student participating in the NASA Undergraduate Student Research Program and was a volunteer part time staffer for the 2005 NASA Academy at MSFC. Once at the University of Florida, from August 2005 to October 2007, Mr. Mireles conducted collaborative experimental materials research with NASA MSFC resulting in his second masters project Material Performance Evaluation of TaC, WC, and ZrC Under Prototypic Nuclear Thermal Propulsion Hot Hydrogen Environment F rom June 2007 to February 2010 Mr. Mireles served as a member of the New Mexico Air National Guard, 150th Fighter Wing, 188th Fighter Squad ron. After commission ing as a second lieutenant he attended the United States Air Force EuroNATO Joint Jet Pilot Training Program, receiving flight instruction in the both the T37B Tweet and T 38C Talon From March 2009 to May 2010 he resumed his doctorate research with NASA Glenn Research Center ultimately resulting in the completion of this dissertation. M r. Mireles is also an accomplished skydiver s kydiving instructor, pilot, SCUBA div er, hiker, and amateur astronomer