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Quantitative Assessment of the Impact of the Previous Cycle's Core Exposure on the Transient Response of Boiling Water R...

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Title: Quantitative Assessment of the Impact of the Previous Cycle's Core Exposure on the Transient Response of Boiling Water Reactor Systems
Physical Description: 1 online resource (81 p.)
Language: english
Creator: Nalepa, Lauren
Publisher: University of Florida
Place of Publication: Gainesville, Fla.
Publication Date: 2008

Subjects

Subjects / Keywords: bwr, exposure, lhgr, licensing, mcpr, reloads, transients
Nuclear and Radiological Engineering -- Dissertations, Academic -- UF
Genre: Nuclear Engineering Sciences thesis, M.S.
bibliography   ( marcgt )
theses   ( marcgt )
government publication (state, provincial, terriorial, dependent)   ( marcgt )
born-digital   ( sobekcm )
Electronic Thesis or Dissertation

Notes

Abstract: We conducted a more intensive study of the effect of previous cycle exposure than is performed in a normal reload licensing analysis. For this research, analyses were performed to determine the validity of the Cycle N licensing limits for MCPR (minimum critical power ratio), LHGR (linear heat generation rate), and peak vessel pressure based on nominal EOC N-1 shutdown compared to the results obtained from early or late EOC N-1 shutdown. Early/late conditions as large as 1000 MWd/ST (as much as plus/minus 30 Effective Full Power Days (EFPD)), as well as intermediate early/late shutdown conditions of plus/minus 500 MWd/ST were examined. The analyses were performed in the same manner as the nominal exposure EOC N-1 reload licensing analysis. Design variations were minimized except for the EOC N-1 exposure effects in order to establish a valid metric for the dependence of results on previous cycle exposure variation. It was found that the plus/minus 500 MWd/ST cases did not challenge the nominal exposure reload licensing limits. The plus/minus 1000 MWd/ST cases were closer to the nominal limits and challenged them in some domains.
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 Lauren Nalepa.
Thesis: Thesis (M.S.)--University of Florida, 2008.
Local: Adviser: Anghaie, Samim.

Record Information

Source Institution: UFRGP
Rights Management: Applicable rights reserved.
Classification: lcc - LD1780 2008
System ID: UFE0022859:00001

Permanent Link: http://ufdc.ufl.edu/UFE0022859/00001

Material Information

Title: Quantitative Assessment of the Impact of the Previous Cycle's Core Exposure on the Transient Response of Boiling Water Reactor Systems
Physical Description: 1 online resource (81 p.)
Language: english
Creator: Nalepa, Lauren
Publisher: University of Florida
Place of Publication: Gainesville, Fla.
Publication Date: 2008

Subjects

Subjects / Keywords: bwr, exposure, lhgr, licensing, mcpr, reloads, transients
Nuclear and Radiological Engineering -- Dissertations, Academic -- UF
Genre: Nuclear Engineering Sciences thesis, M.S.
bibliography   ( marcgt )
theses   ( marcgt )
government publication (state, provincial, terriorial, dependent)   ( marcgt )
born-digital   ( sobekcm )
Electronic Thesis or Dissertation

Notes

Abstract: We conducted a more intensive study of the effect of previous cycle exposure than is performed in a normal reload licensing analysis. For this research, analyses were performed to determine the validity of the Cycle N licensing limits for MCPR (minimum critical power ratio), LHGR (linear heat generation rate), and peak vessel pressure based on nominal EOC N-1 shutdown compared to the results obtained from early or late EOC N-1 shutdown. Early/late conditions as large as 1000 MWd/ST (as much as plus/minus 30 Effective Full Power Days (EFPD)), as well as intermediate early/late shutdown conditions of plus/minus 500 MWd/ST were examined. The analyses were performed in the same manner as the nominal exposure EOC N-1 reload licensing analysis. Design variations were minimized except for the EOC N-1 exposure effects in order to establish a valid metric for the dependence of results on previous cycle exposure variation. It was found that the plus/minus 500 MWd/ST cases did not challenge the nominal exposure reload licensing limits. The plus/minus 1000 MWd/ST cases were closer to the nominal limits and challenged them in some domains.
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 Lauren Nalepa.
Thesis: Thesis (M.S.)--University of Florida, 2008.
Local: Adviser: Anghaie, Samim.

Record Information

Source Institution: UFRGP
Rights Management: Applicable rights reserved.
Classification: lcc - LD1780 2008
System ID: UFE0022859:00001


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1 QUANTITATIVE ASSESSM ENT OF THE IMPACT OF THE PREVIOUS CYCLE'S CORE EXPOSURE ON THE TRAN SIENT RESPONSE OF BO ILING WATER REACTOR SYSTEMS By LAUREN ELIZABETH NALEPA A THESIS PRESENTED TO THE GRADUATE SCHOOL OF THE UNIVERSITY OF FLORIDA IN PARTIAL FULFILLMENT OF THE REQUIREMENTS FOR THE DEGREE OF MASTER OF SCIENCE UNIVERSITY OF FLORIDA 2008

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2 2008 Lauren Elizabeth Nalepa

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3 To my family

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4 ACKNOWLEDGMENTS I would like to thank e veryone who assisted in the research and writing of this thesis. At GE Hitachi, the assistance of Randy Jacobs and Pauline Stier was invaluable. At GE I would also like to thank Jennifer Bowie, Brian Triplett, Cindy Fung, and Keith Bentley for their help in navigating through GE and the University of Florida graduate school. At the University of also like to thank Mireille Rowe for being my homework partner throu ghout our undergraduate and graduate studies Finally, I would like to thank my family: my parents, Ryan, and Emma, and especially my fianc, Tyler Schweitzer.

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5 TABLE OF CONTENTS page ACKNOWLEDGMENTS ................................ ................................ ................................ ........... 4 LIST OF TABLES ................................ ................................ ................................ ...................... 7 LIST OF FIGURES ................................ ................................ ................................ .................... 8 LIST OF ABBREVIATIONS ................................ ................................ ................................ ...... 9 ABSTRACT ................................ ................................ ................................ ............................. 11 CHAPTER 1 INTRODUCTION ................................ ................................ ................................ ............. 12 Boiling Water Reactors ................................ ................................ ................................ ...... 13 Boiling Water Reactor Evolution ................................ ................................ ........................ 15 Boiling Water Reactor Fuel Types ................................ ................................ ...................... 16 Safety Parameters of Interest ................................ ................................ .............................. 17 Minimum Critical Power Ratio ................................ ................................ .................... 17 Linear Heat Generation Rate ................................ ................................ ....................... 18 Peak Vessel Pressure ................................ ................................ ................................ ... 18 2 APPROACH AND METHOD ................................ ................................ ........................... 26 Exposure Cases ................................ ................................ ................................ .................. 26 Operating Domains ................................ ................................ ................................ ............ 26 Reactor and Fuel Parameters ................................ ................................ .............................. 27 Transients ................................ ................................ ................................ ........................... 28 Method ................................ ................................ ................................ ............................... 33 3 RESULTS ................................ ................................ ................................ .......................... 57 Safety Results of Interest ................................ ................................ ................................ .... 57 Delta Critical Power Ratio ................................ ................................ ........................... 57 Thermal and Mechanical Over Power ................................ ................................ ......... 58 Vessel Overpressure ................................ ................................ ................................ .... 58 The BWR/4 ................................ ................................ ................................ ........................ 58 Delta Critical Power Ratio ................................ ................................ ........................... 58 Load Rejection, No Bypass ................................ ................................ .................. 58 Turbine Trip, No Bypass ................................ ................................ ...................... 59 Feedwater Control Failure ................................ ................................ .................... 59 Thermal and Mechanical Over Power ................................ ................................ ......... 59 Load Rejection, No Bypass ................................ ................................ .................. 60 Turbine Trip, No Bypass ................................ ................................ ...................... 60

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6 Feedwater Control Failure ................................ ................................ .................... 60 Vessel Over Pressure ................................ ................................ ................................ .. 60 The BWR/6 ................................ ................................ ................................ ........................ 61 Delta Critical Power Ratio ................................ ................................ ........................... 61 Load Rejectio n, No Bypass ................................ ................................ .................. 61 Turbine Trip, No Bypass ................................ ................................ ...................... 61 Feedwater Control Failure ................................ ................................ .................... 62 P ressure Regulator Failure, Downstream ................................ .............................. 62 Thermal and Mechanical Over Power ................................ ................................ ......... 63 Load Rejection, No Bypass ................................ ................................ .................. 63 Turbine Trip, No Bypass ................................ ................................ ...................... 63 Feedwater Control Failure ................................ ................................ .................... 63 Pressure Regulator Failure, Downstream ................................ .............................. 63 Vessel Over Pressure ................................ ................................ ................................ .. 64 4 CONCLUSIONS ................................ ................................ ................................ ................ 79 LIST OF REFERENCES ................................ ................................ ................................ .......... 80 BIOGRAPHICAL SKETCH ................................ ................................ ................................ ..... 81

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7 LIST OF TABLES Table page 2 1 Case matrix performed for the BWR/4 analyses. ................................ ............................ 37 2 2 Transient matrix performed for the BWR/4 500MWd/ST case ................................ ..... 38 2 3 Case matrix performed for the BWR/6 analyses. ................................ ............................ 39 2 4 Transient matrix performed for the BWR/6 500MWd/ST case ................................ ..... 40 3 1 ................................ ........................... 67 3 2 ................................ .......................... 68 3 3 The BWR/4 GE14 thermal overpower results: margin to screening criteria. ................... 71 3 4 The BWR/4 GE14 mechanical overpower results: margin to screening criteria. ............. 72 3 5 The BWR/4 peak vessel pressure results. ................................ ................................ ....... 73 3 6 ................................ ................................ ... 74 3 7 The BWR/6 thermal overpower results: margin to screening criteria. ............................. 76 3 8 The BWR/6 mech anical overpower results: margin to screening criteria. ....................... 77 3 9 The BWR/6 peak vessel pressure results. ................................ ................................ ....... 78

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8 LIST OF FIGURES Figure page 1 2 BWR/6 steam and recirculation systems (BWR/6, 1980) ................................ ............... 20 1 3 BWR/6 pressure vessel (BWR/6, 1980) ................................ ................................ ......... 21 2 2 Response to LRNBP of system flux and core flow ................................ ......................... 41 2 3 Response to LRNBP of system pressure rise and valve flow ................................ .......... 42 2 4 Response to LRNBP of typical system flow response ................................ .................... 43 2 5 Response to TTNBP of system flux and core flow ................................ ........................ 44 2 6 Response to TTNBP of system pressure rise and valve flow ................................ ........ 45 2 7 Response to TTNBP of system flow ................................ ................................ .............. 46 2 8 Response to MSIVF of system flux and core flow ................................ ........................ 47 2 9 Response to MSIVF of system pressure rise and valve flow ................................ ........... 48 2 10 Response to MSIVF of system flow ................................ ................................ .............. 49 2 11 Response to FWCF with bypass of system flux, core fl ow, and core inlet subcooling .... 50 2 12 Response to FWCF with bypass of system pressure rise and valve fl ow ......................... 51 2 13 Response to FWCF with bypass of system flow ................................ ............................ 52 2 14 Response to PRFDS of system flux and core flow ................................ ......................... 53 2 15 Response to PRFDS of system pressure rise and valve flow ................................ .......... 54 2 16 Response to PRFDS of system flow ................................ ................................ ............... 55 2 17 Transient methodology ................................ ................................ ................................ .. 56 3 1 Critical power ratio versus time for BWR/6 LRNBP. ................................ ..................... 65 3 2 ................................ ...... 66 3 3 ................................ ................. 69 3 4 by domain and case for the BWR/4 ................................ ................ 70 3 6 ................................ ................. 75

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9 LIST OF ABBREVIATION S 2SRVOS Two safety relief valves out of service 6 SVOS Six safety valves out of service ASME American Society of Mechanical Engineers BWR Boiling water reactor BWR/N Boiling water reactor generation N CLTP Current licensed thermal power CPR Critical power ratio Cycle N Current cycle Delta critical po wer ratio EFPD Effective full power days EOC N 1 End of previous cycle EOC End of cycle FWCF Feedwater control failure GE General Electric GNF Global Nuclear Fuel HBB Hard bottom burn ICF Increased core flow ICPR Initial critical power ratio LCF Low core f low LHGR Linear heat generation rate LRNBP Load rejection, no bypass MCPR Minimum critical power MELLLA Maximum extended load line limit analysis MOC Middle of cycle

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10 MOP Mechanical overpower MSIV Main steam isolation valve MSIVF Main steam isolation valve failure MWd/ST Megawatt day per short ton NBP No turbine bypass NFW Normal feedwater temperature NRC Nuclear Regulatory Commission OLMCPR Operating limit critical power ratio PEOC Projected end of cycle PRFDS Pressure regulator failure, downscale Psig Pres sure per square inch, gauge RFW Reduced feedwater temperature SLMCPR Safety limit critical power ratio TCV Turbine control valve TOP Thermal overpower TSV Turbine safety valve TTNBP Turbine trip, no bypass

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11 Abstract of Thesis Presented to the Graduate S chool of the University of Florida in Partial Fulfillment of the Requirements for the Degree of Master of Science QUANTITATIVE ASSESSM ENT OF THE IMPACT OF THE PREVIOUS CYCLE'S CORE EXPOSURE ON THE TRAN SIENT RESPONSE OF BO ILING WATER REACTOR SYSTEMS By Lau ren Elizabeth Nalepa December 2008 Chair: Samim Anghaie Major: Nuclear Engineering Sciences We conducted a more intensive study of the effect of previous cycle exposure than is performed in a normal reload licensing analysis. For this research, analyse s were performed to determine the validity of the Cycle N licensing limits for MCPR (minimum critical power ratio), LHGR (linear heat generation rate), and peak vessel pressure based on nominal EOC N 1 shutdown compared to the results obtained from early o r late EOC N 1 shutdown. Early/late conditions as large as 1000 MWd/ST (as much as plus/minus 30 Effective Full Power Days (EFPD)), as well as intermediate early/late shutdown conditions of plus/minus 500 MWd/ST were examined. The analyses were performe d in the same manner as the nominal exposure EOC N 1 reload licensing analysis. Design variations were minimized except for the EOC N 1 exposure effects in order to establish a valid metric for the dependence of results on previous cycle exposure variatio n. It was found that the plus/minus 500 MWd/ST cases did not challenge the nominal exposure reload licensing limits. The plus/minus 1000 MWd/ST cases were closer to the nominal limits and challenged them in some domains.

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12 CHAPTER 1 INTRODUCTION The wor k undertaken in this thesis research was performed at GE Hitachi Nuclear Energy in Wilmington, North Carolina, in the Transient Analysis group. As part of the Nuclear Analysis Center of Excellence, the Transient Analysis group at GE Hitachi performs trans ient analyses for each new fuel cycle. The potentially limiting transients for each plant are analyzed each cycle. These transients are then analyzed at rated and sometimes off rated power levels to demonstrate compliance with the fuel thermal margin req uirements and American Society of Mechanical Engineers (ASME) reactor vessel overpressure protection criteria. This study on expanded exposure windows was performed because utilities often desire or request additional flexibility for operation of Cycle N 1 such that it does not affect the licensing of Cycle N. The intent was to show that the current method of performing licensing calculations is valid even if the previous cycle ends thirty days earlier or later than expected. This study focused on a BWR /4 and a BWR/6 plant. The BWR/4 wa s based on a 560 bundle core at 2536 MWt and the BWR/6 wa s based on 748 bundle core at 3758 MWt. The objective of the research was to perform transient analyses based on early/late shutdown conditions at the end of the pr evious cycle (EOC N 1). Early/late conditions as large as 1000 MWd/ST ( 30 Effective Full Power Days (EFPD)), as well as intermediate early/late conditions of 500 MWd/ST were examined. The analyses were performed in the same manner as the nominal expo sure EOC N 1 reload licensing analysis. Design variations were minimized except for the EOC N 1 exposure effects in order to establish a valid metric for the dependence of results on previous cycle exposure variation. The early/late shutdown results were compared to nominal licensing results.

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13 The research was a more intensive study of the effect of previous cycle exposure than is performed in a normal reload licensing analysis. Normal reload licensing considers an early shutdown window to assure that col d reactivity and shutdown margin concerns are acceptable (one rod out and standby liquid control system) at these potential early EOC N 1 conditions. Previous transient analysis studies have also examined smaller exposure windows, and have found an accept ably small impact on transient performance. For this research, analyses were performed to determine the validity of the Cycle N licensing limits for MCPR (minimum critical power ratio), LHGR (linear heat generation rate), and peak vessel pressure based on nominal EOC N 1 shutdown compared to the results obtained from early or late EOC N 1 shutdown. The expected outcome was to show that the current licensing process is acceptably bounding even for these very large variations in Cycle N 1 shutdown and corre sponding variation in exposure distribution. Boiling Water Reactors Boiling water reactor technology was developed at Argonne National Laboratory and the Nuclear Energy Division of General Electric (GE) in the 1950s (Lahey, 1993) The first reactor lic ense issued by the Atomic Energy Commission was to Vallecitos 1, which came online near San Jose, California in 1957. This reactor produced 5 MWe and provided power to the Pacific Electric and Gas Company grid. The first commercial plant was the 180 MWe Dresden 1 in Morris, Illinois, in 1961 (BWR/6, 1980). Figure 1 1 shows the layout of a direct cycle reactor system, such as a BWR. Steam is produced in the vessel itself, where it travels through steam separators and dryers before exiting and traveling t o the turbine. Most BWRs have three stage turbines, of which one is high pressure and the other two are low pressure. The steam goes through moisture separators and reheaters between each turbine stage. This is crucial because moisture is very damaging to

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14 turbine blades. After exiting the turbine the steam is passed through a condenser to turn it back into water. The water is demineralized and heated, and then again is pumped to the reactor. This water that travels through the reactor and turbine is r adioactive due primarily to N 16 which has a half life of seven seconds. Figure 1 2 shows the steam and recirculation systems for the BWR. Shown on either side are the jet pumps. These pumps recirculate water through the reactor vessel; they generate about two thirds of the recirculation flow within the reactor vessel. They also allow for load following by the plant through increases or decreases in the coolant flow. The main water/steam path through the vessel is also shown in the figure. Cold wate r flows down the sides of the vessel in the downcomer region. The recirculation and jet pumps provide the recirculation head to drive the flow through the fuel bundles, where it turns to steam. The steam is then separated from the water in the steam sepa rators and additional moisture is removed when the steam passes through the steam dryer assembly near the top of the vessel. The steam then flows to the turbine. Figure 1 3 shows a detailed view of the BWR pressure vessel. The steam dryers and separators are shown in more detail. These two systems are static and utilize baffles and turning vanes to remove moisture from the steam. One notable item from the figure is the relatively small size of the core itself the pressure vessel has many components, o f which the fuel itself is a small part. The control rods enter the vessel from the bottom by hydraulic control rod drives. The control rods themselves are cruciform in shape (when viewed from the top) and fit between sets of four fuel bundles. They ar e filled with B 4 C powder. In addition to the control rods, the BWR power can be controlled by the amount of recirculating water. An increase in the recirculating

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15 water flow rate decreases the void fraction and thus causes power to increase. A decrease i n the flow rate will cause power to decrease. B oiling W ater R eactor Evolution In the 18 years between the first BWR/1 and BWR/6 startup dates, the GE BWR design underwent many changes. The design variations can be classified into two groups the reactor systems and the containment design. The BWR has six distinct reactor system design generations and three containment designs. The Dresden 1 BWR/1 plant, with its ex vessel steam generators and dry containment is very different from the Perry BWR/6 plant, with its Mark III containment and lack of steam generators. The steam generators disappeared in the BWR/2, the first of which was Oyster Creek in 1959. These plants were purchased solely for economic reasons, and were thus much larger than the BWR/1 pl ants. Five recirculation loops were used to remove heat from the core. This number was reduced to two by the introduction of internal jet pumps in the BWR/3 design, such as in plants like Dresden 2. The BWR/3 also improved the Emergency Core Cooling Sys tem (ECCS) in the plants by adding spray and reflood capability. Figure 1 4 shows the evolution of the BWR. Changes in the next generations of the BWR family were smaller in scale, and focused on improving power density in the BWR/4 (first online in 1972) and improving the ECCS and introducing recirculation valve flow control in the BWR/5 (first online in 1972). The BWR/6 (first online in 1978) improved the nuclear system protection system and reduced the size of the control room. These changes in the r eactor systems generally improved the transient and accident response of BWRs. One major difference between BWR/4 and BWR/6 plants, as seen in this

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16 research, is their neutron flux response during transients. The BWR/4 response is much more severe than th e BWR/6 increase due to the faster scram speed introduced with the BWR/6. The containments for BWR plants changed greatly, as well. The first BWRs used dry, spherical containments. The elimination of the steam generators and reduction of the outside re circulation loops in turn allowed for the reduction of the containment volumes. The Mark I containment, shaped like an inverted light bulb, holds water in its torus at the bottom of containment. The Mark II containment has a conical shape and a large con tainment drywell. The Mark III containment is universally used in BWR/6 plants. It is an easily constructed right circular cylinder within a free standing steel structure. Figure 1 5 shows the e volution of the BWR containment (ABWR, 2006). B oiling W ater R eactor Fuel Types The two most advanced fuel bundles offered by GNF were considered in this study, GE14 and GNF2 Advantage Both fuel product lines contain tie plates, spacers, channel boxes, fuel rods, and water rods made up of Zircalloy 2, an alloy of zirconium, tin, iron, chromium, and nickel. The fuel bundles are 10x10 arrays and contain 92 fuel rods, 14 of which are part length rods. GNF2 Advantage bundles contain eight part length rods that are approximately two thirds the full length a nd six shor ter part length rods (GNF2 Advantage, 2007). The part length rods of the GE14 bundle are all the same length (DiFillipis, 2005). The bundle also contains two large water rods which take the place of eight fuel rods. The rods are all surrounded by a chan nel box that prevents cross flow between channels. Tie plates at the top and bottom, as well as spacers spaced axially, hold the rods in place (GNF2 Advantage, 2007). Figure 1 6 shows the GE14 fuel bundle. Figure 1 7 shows the GNF2 Advantage fuel bundle The fuel rods are filled with stacked high density, ceramic UO 2 or (U, Gd)O 2 pellets. The zircalloy cladding is lined with a thin barrier layer of zirconium and the rod is pressurized with

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17 helium gas. Global Nuclear Fuel (GNF) uses gadolinia as its bu rnable poison of choice, in the (U, Gd)O 2 pellets. In reload fuel, GNF2 Advantage and GE14 fuel rods typically have a 6 12 inch blanket at the bottom and typically a six inch blanket at the top. In addition, in the gadolinia rods the enrichment can also vary axially. Safety Parameters of Interest The transient analysis process is undertaken to determine the values for several safety factors: MCPR, fuel centerline melting, excessive cladding strain, and peak vessel pressure. Minimum C ritical P ower R atio The minimum critical power ratio is in essence a measure of how far a fuel rod is from transition boiling; that is, how far the fuel rod is from leaving the highly efficient nucleate boiling behind and moving into transition boiling and film boiling. The value used to determine this margin is the critical power ratio (CPR) (Todrias, 1990). To determine the CPR, a critical steam quality is defined where the onset of transition boiling would occur. This quality is defined in terms of boiling length, mass f low rate, power level, pressure, local steam quality, bundle geometry, and local peaking power. A critical power is found that would produce the critical quality, and then the critical power ratio is the ratio of the critical power to the operating bundle power. The minimum critical power ratio is the ratio of the critical power to the maximum operating bundle power (Lahey, 1993). No reactor operator wishes to operate at the CPR, so several layers of conservatism are applied for actual reactor operation. Each plant has a safety limit minimum critical power ratio (SLMCPR), which usually ranges from 1.07 to 1.10, and then operating limit minimum critical power ratios (OLMCPR) are defined for each cycle and operating condition, in an attempt to be certain th e MCPR never approaches the value of one for an anticipated operational occurrence

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18 for a Feedwater Controller Failure (FWCF) or 0.19 for a Load Rejection with No Bypass (LRNBP), is added to the SLMCPR (Ishigai, 1999) The highest, or most limiting, OLMCPR is set as the limit for the cycle. Linear Heat Generation Rate The LHGR is a measure of the power produced by a fuel pin divided by the length of active fuel. In the core, a maximum linear heat generation rate is found by finding the fuel rod with the highest surface heat flux. This value is monitored to be sure that fuel thermal and mechanical limits are not exceeded. For operational transients, the fuel ther mal and mechanical overpowers are evaluated for margin to thermal mechanical limits; these values are margins to fuel melt and margin to 1% cladding plastic strain. Peak Vessel Pressure The ASME limit for peak vessel pressure is 1375 psig for upset cond itions, which is 10% greater than the design pressure. This assures the integrity of the reactor vessel during postulated overpressure events. To prevent the vessel pressure from exceeding this limit, safety / relief valves are set to open if the vessel pressure reaches the opening setpoint for the safety / relief valves; these valves will vent steam to the containment to relieve the system pressure. The most limiting transient for peak vessel pressure is Main Steam Isolation Valve Closure with a Flux Sc ram (MSIVF).

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19 Figure 1 1. Direct cycle reactor system ( BWR/6, 1980 )

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20 Figure 1 2. BWR/6 steam and recirculation systems (BWR/6, 1980)

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21 Figure 1 3. The BWR/6 pressure vessel (BWR/6, 1980)

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22 Figure 1 4. Evolution of the BWR reactor syste m design (A BWR, 2006 )

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23 Figure 1 5. Containment evolution of the BWR (ABWR, 2006 )

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24 Figure 1 6. The GE14 fuel bundle. ( DeFilippis, 2005 )

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25 Figure 1 7. The GNF2 Advantage fuel bundle (GNF2 Advantage 2007)

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26 CHAPTER 2 APPROACH AND METHOD Exposure Cases Five e xposure cases were analyzed for this research. In addition to the nominal case, state point nuclear cross sections (wrap ups) were created for both plus and minus 500 and 1000 MWd/ST. These wrap ups contain the fuel bundle exposure and control rod patter ns at multiple exposure points throughout the cycle. For both the BWR/4 and the BWR/6, end of cycle (EOC) points were investigated, as well as middle of cycle (MOC) points for the BWR/4. Additionally, normal and reduced feedwater temperatures were analyz ed, as well as increased core flow and low core flow conditions. Operating Domains B oiling water reactor s can operate at varied flow and power configurations. To operate safely, the reactor must operate within a certain area of the power versus flow map, shown in Figure 2 1. The reactor must operate between the minimum and maximum core flow at the various power levels shown on Figure 2 1. This area is shown in Figure 2 1 by the quadrangle formed by the minimum pump speed line, minimum rod line, 100% cur rent licensed thermal power (CLTP) rod line that pass through point E, and line EH. In addition, the plant may opt to operate in low core flow or increased core flow conditions. The low core flow region, called the maximum extended load line limit analys is (MELLLA) region, is bound between the line CD and the 100% CLTP rod line. It permits operation at lower core flows. The increased core flow region is bound in the figure by 100% and 105% core flow, and lines EF and HG. Increased core flow provides fl exibility at rated power and can be used to extend the operating cycle by the additional reactivity associated with the increased core flow.

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27 In addition, the feedwater temperature may be altered during operation. The plant usually operates at its normal f eedwater temperature (NFW), but a reduced feedwater temperature will result in higher reactivity because the reduced temperature feedwater (RFW) is more dense, resulting in more moderator in the core. This may be used by the plant when a feedwater heater is out of service or for use in extending the cycle with the increased moderation and reactivity. Combinations of ICF, MELLLA, RFW, and NFW are used to produce multiple operating domains for a variety of situations. These combinations form the various ini tial conditions studied in the transient analysis. Reactor and Fuel Parameters The BWR/4 ha d a licensed thermal power of 2536 MW and a rated core flow of 77 Mlb/hr. The increased core flow domains were analyzed at 105% of rated flow, or 80.85 Mlb/hr. The low flow domains were analyzed at 79.8% of rated flow, or 61.45 Mlb/hr. The increased core flow and low flow values we re analyzed to envelope the allowed operating range at the license power. The license (rated) power was used for all MCPR analyses. Th e normal feedwater temperature wa s 424 o F, while the reduced feedwater temperature wa s analyzed at 344 o F. The cycle analyzed for BWR/4 contained a mixed core of GNF2 Advantage and GE14 fuel. The BWR/6 analyzed ha d a licensed thermal power of 3758 MW and a rated core flow of 104 Mlb/hr. The increased core flow domains were analyzed at 105% of rated flow, or 109.2 Mlb/hr. The low flow domains were analyzed at 81% of rated flow, or 84.24 Mlb/hr. The license power was used for all MCPR analyses. The n orma l feedwater temperature wa s 425.5 o F, while the reduced feedwater temperature wa s 255.5 o F. The BWR/6 core contain ed GE14 fuel bundles.

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28 Transients Five transients were analyzed to determine required operating limits: Main Steam Isolation Valve Failure Flux Scram (MSIVF), Load Rejection without Bypass (LRNBP), Turbine Trip without Bypass (TTNBP), Feedwater Control Failure (FWCF), and Pressure Regulator Failure, Downscale (PRFDS). The MSIVF transient was analyzed at a thermal power of 102% of rated to d etermine the margin to the ASME upset code limit for the peak vessel pressure and the technical specification safety limit for the dome pressure. The LRNBP, TTNBP, FWCF, and PRFDS were analyzed at rated power to determine the MCPR limit and to show compli ance with the LHGR overpower limits. Additionally, the FWCF was also analyzed with the turbine bypass valve out of service (NBP). For the BWR/4, LRNBP, TTNBP, FWCF, and MSIVF were all analyzed; additionally, the analysis considered two safety relief val ves out of service (2SRVOS) because that is a Nuclear Regulatory Commission (NRC) approved operating strategy for this plant and this flexibility causes more limiting overpressure conditions than with the safety relief valves in service. Table 2 1 shows the analysis matrix for the BWR/4, while Table 2 2 shows the transient matrix performed for the BWR/4 500 MWd/ST case. Similar matrices were performed for all the BWR/4 cases. For the BWR/6, LRNBP, TTNBP, FWCF, and PRFDS were all run with two relief valv es out of service (2RVOS) because it is an NRC approved operating strategy and more limiting than with the valves in service. The MSIVF transient was run with six safety valves out of service (6SVOS). Table 2 3 shows the case matrix for the BWR/6 analyse s, while Table 2 4 shows the transient matrix performed for the BWR/6 500MWd/ST case. Similar matrices were performed for all the BWR/6 cases.

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29 During LRNBP, the generator experience d a loss of electrical load. To prevent a turbine/generator over speed c ondition, the turbine control valves must close as quickly as possible. This cut off the flow of steam through the turbine, which remove d the working fluid and causes the turbine to slow to a stop. This create d an increase in pressure as the steam headed to the turbine is stopped, resulting in a pressure wave that transmit ted through the system back to the reacto r vessel. Positive reactivity wa s introduced into the core as voids collapse d under the pressure wave, resulting in an increase in neutron flux. However, this also increase d the heat flux, which, combined with control rod scram, insert ed negative reactivity to bring the system under control. If needed, safety relief valves could also open, depressurizing the system by piping steam into the suppr ession pool. Figure 2 2 shows the typical system flux and core flow response after LRNBP. A fter the turbine control valves we re completely shut at 0.152 sec, the neutron flux spike d to several times its rated value. This increase in neutron flux wa s q uickly turned around by the start of control bla de motion at 0.281 sec. There wa s also a moderate increase in the average surface heat flux, which follow s the neutron flux, but this wa s turned around, as well, by the insertion of the control rods. The da ta for the figure was from the BWR/4, which has a much more severe neutron flux response than BWR/6 plants. Figure 2 3 shows the increase in vessel pressure and the flow of the relief valves, as well as the turbine bypass flow during LRNBP. The vessel p ressure rose to above 175 psi above the initial condition before the relief valves open ed at 1.975 sec. The relief valves br ought the pressure down to the closing setpoints associated with the safety relief valves. It should be no ted that the bypass valv e flow wa s constant at zero because this transient d id not credit the turbine bypass valves.

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30 Figure 2 4 shows vessel steam flow, turbine steam flow, and feedwater flow in the core during LRNBP. The turbine steam flow cut off at 0.15 sec and stay ed at zero for the rest of the transient. The vessel steam flow show ed oscillatory behavior; this is a result of the pressure wave propagating between the vessel and the turbine control valves. Once the safety relief valves open ed the magnitude of the vessel stea m flow oscillations dampen ed to the safety relief valve flow rate. During TTNBP, the turbine trip ped off line and the turbine stop valves close d This trip c ould have be en caused by several events, including large vibrations, high water level in the vess el, and low condenser vacuum. A similar course of events occur ed as with LRNBP. Figure 2 5 shows the neutron and average surface heat flux, as well as the core inlet flow during the TTNBP. The figure is very similar to Figure 2 2 because after the initi ation, the TTNBP ha d a very similar course of events as the LRNBP. Figures 2 6 and 2 7 are also similar to their counterparts for LRNBP. During the MSIVF, the reactor wa s assumed to be operating at 102% of rated power. A main steam isolation valve (MSIV) close d in the minimum time and the first available scram on MSIV position wa s not credited. The pressure increase d in the core, collapsing voids, causing an increase in power. The neutron flux increase d and reache d the scram setpoint, which trigger ed th e control rod scram and decrease d the power to bring the system back under control. The safety relief valves could also open to reduce the vessel pressure. Figure 2 8 shows the neutron and average surface heat flux, as well as the core inlet flow, durin g MSIVF. The initial pressure of the vessel was 1094 psig. At time zero the MSIVs beg a n closing and were fully closed at 3.00 sec. When the MSIVs beg a n to have a significant effect on the steam flow (at approximately 0.6 seconds) there wa s a sp ike in ne utron flux that resulted in a

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31 scram. The control rods bega n motion at 1.81 sec and turn ed the neutron flux increase around. The heat flux again follow ed the neutron flux, but at much smaller values. Figure 2 9 shows the vessel pressure rise, relief va lve flow, and bypass valve flow for the MSIVF. Just like in the LRNBP and TTNBP, the vessel pressure increase d until the relief valves open ed where they vent ed the steam and allow ed the pressure to decrease. Figure 2 10 shows the vessel steam flow and tu rbine steam flow during MSIVF. The turbine steam flow drop ped off throughout the transient to zero. The ve ssel steam flow once again showed oscillatory behavior, with a spike in steam flow at about the same time as the second spike in neutron flux. Also noteworthy is that the oscillations were more frequent with the MSIV closure because the main steam isolation valves were much closer to the reactor vessel. During a FWCF, a postulated failure occurred in the feedwater control system, which demand ed the m aximum feedwater flow. This caus ed the water level to rise until it reache d the high water set point, at which time the turbine and feedwater pumps tripped, the turbine stop valve and turbine control valves were demanded to close, and the control rods scr am med Normally, the turbine bypass system would be demanded to open, directing steam to the condenser. However, if the turbine bypass valves were inoperable the pressure at the turbine inlet and the vessel would have a larger pressure increase due to th e buildup of steam. The safety relief valves could open if needed to relieve the pressure. Figure 2 11 shows the neutron flux, average surface heat flux, core inlet flow, and core inlet subcooling for FWCF with bypass. One factor that is immediately ap parent is how long it took for the transient to take effect FWCF is a relatively slow transient and the effects are not seen until several seconds after the increase in feedwater flow. At zero seconds the feedwater

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32 flow wa s increased to the maximum dema nd, about 145 % of rated feedwater flow. Feedwater flood ed the vessel until the level reache d the high water level trip setpoint at 11.5 sec. To protect the turbine, the turbine stop valves and turbine control valves beg a n closi ng at 11.5 sec and the turb ine was tripped. Once the turbine was tripped the FWCF behavior wa s much like a turbine trip except the turbine bypass system is normally credited in this event. The control blade motion beg an at 11.82 sec. The characteristic spikes in neutron flux and average surface heat flux are shown in the figure. Also, the core inlet subcooling rose because of the influx of cold water in the core. The large change in inlet subcool ing at approximately 12 second wa s due to the vessel pressurization. Figure 2 12 sh ows the vessel pressure rise, relief valve flow, and bypass valve flow for the FWCF. Note that because this transient takes credit for the turbine bypas s system, the bypass valves beg a n to open at 1 1.6 sec. The relief valves beg a n to open at 13.8 sec and turn the pressure wa s turned around. Figure 2 13 shows the vessel steam flow, turbine steam flow, and feedwater flow for FWCF. The feedwater flow wa s a constant 145% until the vessel reache d the vessel high level setpoint, and the increase in vessel water can be seen in the vessel level plot in the figure. Once the turbine trip ped and the turbine steam flow drop ped to zero, the pressure wave propagate d from the turbine to the vessel and produce d the oscillatory behavior o f the vessel steam flow, which wa s very similar to a TTNBP event. A transient unique to the BWR/6 licensing is PRFDS. During this transient, the pressure regulators fail ed which demand ed the turbine control valves to close. The turbine control valves fully close d and the bypass flow wa s rendered i noperable due to the pressure regulator failure, causing an increase in reactor power and pressure. This event result e d in a slow closure of the

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33 turbine control valves and thus no direct reactor scram occur red The reactor wa s scrammed once the high neutron flux setpoint is reached. Figure 2 14 shows the typical neutron flux, average surface heat flux, core inlet flow, and core inlet subcooling responses during PRFDS. The neutron flux spike wa s much smaller than that seen du ring the other transients, and wa s mostly due to the slower closure of the turbine control valves. The core inlet subcooling rose to over 200% of its rated value because of the sudden pressure increase. Figure 2 15 shows the typical system pressure rise and valve flow response after PRFD S. As with the other transients, the vessel pressure rose until the relief valves open and vent the steam. Figure 2 16 shows the typical vessel and turbine steam flow after PRFDS. The oscillatory behavior of the steam flow seen in other transients wa s no t seen during PRFDS. This is because during PRFDS the ~2.5 second TCV closure time was much slower than the ~0.1 second TCV/TSV closure time of transients like LRNBP. Because there was a gradual reduction in steam flow, a large pressure wave is n ot created at the turbine inlet (Watford, 2000). Method Several codes were utilized to perform the transient analysis. PANACEA was a three dimensional, nodal diffusion BWR simulator. It couple d the neutronics and thermal hydraulics and determine d both the steady st ate conditions in the core and the nuclear parameters such as k effective, cross sections for the statepoint of interest and the delayed neutron fraction. CRNC collapse d the three dimensional parameters to a one dimensional form of averaged cross sectio ns so that they could be used by OYDN, the one dimensional transient simulator. ISCOR and TASC calculate d the thermal hydraulic response and hot channel analysis. TACLE was an automation code that is used to run the codes.

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34 ODYN was the transient simulator an d was the code that produce d the transient results described in Figures 2 1 through 2 16. Coupling the neutron kinetics with the thermal hydraulics and heat transfer in the core, ODYN was a best estimate one dimensional core model. It model ed a multi node c ore consistent with the axial noding used in PANACEA to determine the core average response. The neutron kinetics in ODYN were assumed to be one dimensional and var ied axially with time. Parameters from PANACEA, collapsed by CRNC, were used for the one en ergy group diffusion and six delayed neutron groups. To collapse the cross sections, radial spatial weighting factors were applied to the nuclear parameters. These preserve d the dynamic response of the core during core wide abnormal operating events. ODYN model ed the average core response and therefore the parameters such as the flow are a and gap conductance are represented by averaged values of the fuel types modeled. The model was initialized to the core pressure drop calculated by the BWR steady state si mulator. The core model was tuned to be sure that calculated parameters accurately reflect physical parameters in the core, recirculation system and steam line. The steady state axial power distribution was calculated in ODYN with its one dimensional kinet ics model. The cross section collapsing process assure d that it is a close match to the PANACEA results. Once the power distribution was calculated, the steady state fuel temperature distribution was calculated. These distributions provide d a starting poin t for the transient calculations. The transient analysis model simulate d the reactor core model and the balance of plant. The reactor core model calculate d the pressure, flow, neutron flux, heat flux, fuel temperatures,

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35 reactor core exit quality, and co re pressure drop. The plant model calculate d steam line dynamic response along with the recirculation loop, feedwater and pressure control systems. During transients, pressure, flow, neutron flux, and heat flux were dependent on time and are determined a t each time step. These factors provide d boundary conditions for the next time step, and a new neutron flux, fuel temperatures, pressure, heat flux, and other parameters are determined again for the new time. To calculate the thermal hydraulic behavior o f the core, a combination of mass and energy conservation for liquid and vapor plus a momentum conservation equation for the mixture was used in two channels a heated one representing the average core conditions and one simulating th e bypass region (Supp lemental Safety Evaluation, 1981). ISCOR and TASC took the core pressure, core flow, inlet subcooling, power generation, and axial power shape, all time dependent, as inputs from ODYN. TASC calculate d the change in CPR during the transient, using the core pressure, axial power shape and inlet subcooling from ODYN. To determine the hot channel power and flow, ISCOR calculate d the hot channel inlet flow from the core pressure, core inlet flow, core inlet enthalpy, and core power. The hot channel flow and p ower were input to TASC and TASC determine d the MCPR. This process iterate d until the calculated MCPR was equal to the SLMCPR. Figure 2 17 shows the relationship between ODYN, ISCOR, and TASC, and the inputs and outputs associated with the codes. The one dimensional nuclear data and core geometry were fed into ODYN, which in turn provide d the core inlet flow to ISCOR. The core power and power shape, core pressure, and core inlet enthalpy from ODYN, plus the hot channel power and hot channel inlet flow fro m ISCOR, were fed into TASC, which calculate d the critical power ratio and maximum rod temperature versus time.

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36 Figure 2 1: Power/Flow map for the BWR/4.

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37 Table 2 1. Case matrix performed for the BWR/4 analyses. Core Flow Feedwater Temperature Nominal +1 000 MWd/ST +500 MWd/ST 500 MWd/ST 1000 MWd/ST Increased Normal ICF___TN IF_P1ETN IF_P5ETN IF_ 5ETN IF_ 1ETN Increased Normal ICF___TN (MOC) IF_P1MTN IF_P5MTN IF_ 5MTN IF_ 1MTN Low Normal MEL___TN LF_P1ETN LF_P5ETN LF_ 5ETN LF_ 1ETN Low Normal MEL___T N (MOC) LF_P1MTN LF_P5MTN LF_ 5MTN LF_ 1MTN Increased Reduced ICF___TR IF_P1ETR IF_P5ETR IF_ 5ETR IF_ 1ETR Low Reduced MEL___TR LF_P1ETR LF_P5ETR LF_ 5ETR LF_ 1ETR Increased Normal ICF__BTN* Low Normal MEL__BTN* Increased Reduced ICF__BTR* Low Reduced MEL__BTR* Note: The eight digit names in the table are the domain identifiers for the analysis. The FWCF/NBP event was included in the domains for the varied exposure cases instead of being in its own domain. *Domain only analyzes FW CF/NBP event.

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38 Table 2 2. Transient matrix performed for the BWR/4 500MWd/ST case Domain Flow (% of rated) Exposure Feedwater t emperature ( o F) Transient Equipment o ut of s ervice IF_E5LTN 105 EOC 424 MSIVF 2SROVS IF_E5LTN 105 EOC 424 LRNBP 2SROVS IF_E5 LTN 105 EOC 424 TTNBP 2SROVS IF_E5LTN 105 EOC 424 FWCF 2SRVOS IF_E5LTN 105 EOC 424 FWCF NBP/2SRVOS LF_E5LTN 79.8 EOC 424 MSIVF 2SROVS LF_E5LTN 79.8 EOC 424 LRNBP 2SROVS LF_E5LTN 79.8 EOC 424 TTNBP 2SROVS LF_E5LTN 79.8 EOC 424 FWCF 2SRVOS LF_E5LTN 79 .8 EOC 424 FWCF NBP/2SRVOS IF_E5LTR 105 EOC 344 FWCF 2SRVOS IF_E5LTR 105 EOC 344 FWCF NBP/2SRVOS LF_E5LTR 79.8 EOC 344 FWCF 2SRVOS LF_E5LTR 79.8 EOC 344 FWCF NBP/2SRVOS IF_M5LTN 105 MOC 424 MSIVF 2SROVS IF_M5LTN 105 MOC 424 LRNBP 2SROVS IF_M5LTN 105 MOC 424 TTNBP 2SROVS IF_M5LTN 105 MOC 424 FWCF 2SRVOS LF_M5LTN 79.8 MOC 424 MSIVF 2SROVS LF_M5LTN 79.8 MOC 424 LRNBP 2SROVS LF_M5LTN 79.8 MOC 424 TTNBP 2SROVS LF_M5LTN 79.8 MOC 424 FWCF 2SRVOS

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39 Table 2 3. Case matrix performed for the BWR/6 analy ses. Core f low Feedwater t emperature Nominal +1000 MWd/ST +500 MWd/ST 500 MWd/ST 1000 MWd/ST Increased Normal ICF___TN IF_P1ETN IF_P5ETN IF_ 5ETN IF_ 1ETN Low Normal MEO___TN LF_P1ETN LF_P5ETN LF_ 5ETN LF_ 1ETN Increased Reduced ICF___TR IF_P1ETR IF_P 5ETR IF_ 5ETR IF_ 1ETR Increased Normal ICF__PTN* Increased Reduced ICF__PTR* Note: The eight digit names in the table are the domain identifiers for the analysis. The PRFDS event was included in the domains for the varied exposure cases ins tead of being in its own domain. *Domain only analyzed the PRFDS event.

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40 Table 2 4. Transient matrix performed for the BWR/6 500MWd/ST case Domain Flow (% of rated) Exposure Feedwater t emperature ( o F) Transient Equipment o ut of s ervice IF_ 5ETN 105 EO C 425.5 LRNBP 2RVOS IF_ 5ETN 105 EOC 425.5 TTNBP 2RVOS IF_ 5ETN 105 EOC 425.5 FWCF 2RVOS IF_ 5ETN 105 EOC 425.5 MSIFV 6SVOS IF_ 5ETN 105 EOC 425.5 PRFDS 2RVOS LF_ 5ETN 84.24 EOC 425.5 LRNBP 2RVOS LF_ 5ETN 84.24 EOC 425.5 TTNBP 2RVOS LF_ 5ETN 84.24 E OC 425.5 FWCF 2RVOS LF_ 5ETN 84.24 EOC 425.5 MSIFV 6SVOS IF_ 5ETR 105 EOC 255.5 FWCF 2RVOS IF_ 5ETR 105 EOC 255.5 PRFDS 2RVOS

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41 Figure 2 2. Response to LRNBP of sys tem flux and core flow. Data are from the BWR/4 500 MWd/ST, ICF, NFW case.

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42 Figure 2 3. Response to LRNBP of system pressure rise and valve flow. Data are from the BWR/4 500 MWd/ST, ICF, NFW case.

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43 Figure 2 4. Response to LRNBP of t ypical system flow response. Data are from the BWR/4 500 MWd/ST, ICF, NFW case.

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44 Figure 2 5. Res ponse to TTNBP of system flux and core flow. Data are from the BWR/4 500 MWd/ST, ICF, NFW case.

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45 Figure 2 6. Response to TTNBP of system pressure rise and valve flow. Data are from the BWR/4 500 MWd/ST, ICF, NFW case.

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46 Figure 2 7. Response to TT NBP of system flow. Data are from the BWR/4 500 MWd/ST, ICF, NFW case.

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47 Figure 2 8. Response to MSIVF of system flux and core flow. Data are from the BWR/4 500 MWd/ST, ICF, NFW case.

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48 Figure 2 9. Response to MSIVF of system pressure rise and valv e flow. Data are from the BWR/4 500 MWd/ST, ICF, NFW case.

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49 Figure 2 10. Response to MSIVF of system flow. Data are from the BWR/4 500 MWd/ST, ICF, NFW case.

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50 Figure 2 11. Response to FWCF with bypass of system flux, core flow, and core inlet sub cooling. Data are from the BWR/4 500 MWd/ST, ICF, NFW case.

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51 Figure 2 12. Response to FWCF with bypass of system pressure rise and valve flow Data are from the BWR/4 500 MWd/ST, ICF, NFW case.

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52 Figure 2 13. Response to FWCF with bypass of syste m flow Data are from the BWR/4 500 MWd/ST, ICF, NFW case.

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53 Figure 2 14. Response to PRFDS of system flux and core flow. Data are from the BWR/6 500 MWd/ST, ICF, NFW case.

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54 Figure 2 15. Response to PRFDS of system pressure rise and valve flow. Data are from the BWR/6 500 MWd/ST, ICF, NFW case.

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55 Figure 2 16. Response to PRFDS of system flow Data are from the BWR/6 500 MWd/ST, ICF, NFW case.

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56 Figure 2 17. Transient methodology. Inputs and outputs associated with ODYN, ISCOR, and TASC ( Lamb, 2007) ODYN ISCOR TASC 1D nuclear data Core geometry Core inlet flow CPR Rod temperature Hot channel power Hot channel inlet flow Core power and shape Core pressure Core inlet enthalpy

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57 CHAPTER 3 RESULTS The domains analyzed could not be compared directly; it wa s not an apples to apples comparison because the projected end of cycle (PEOC) or nominal shutdown exposure, cases were bounding hard bottom burns (HBB), while the vari ed exposure cases were nominal burns. The intent was to show that the HBB analyzed in the original/licensing transient analysis, using the projected end of cycle, would bound any nominal burns that arise from exposure shortfalls or extensions. Safety Resul ts of Interest Delta Critical Power Ratio was the value measured during the transients; it was a measure of the amount the critical power ratio drops between the initial CPR and the Minimum CPR (MCPR) of the transient. Because each transient in a domain generally b egins at the same was an easy way to measure which transient was most severe in terms of MCPR. The transients analyzed to determine CPR were LRNBP, TTNBP, and FWCF (both with and without bypass). Figure 3 1 shows a comparison of cri tical power ratio versus time for the BWR/6 LRNBP. All of the domains had similar trend for CPR versus time. Figure 3 2 is a increased core flow region had CPR values than the low core flow region, and that PRFDS and FWCF had

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58 Thermal and Mechanical Over Power T hermal over power and MOP were values that express ed the percentage above the steady state fuel thermal and mech anical limits. These values used were screening criteria for determining margins to the overpower limits. Vessel Overpressure To prevent vessel overpressure, the peak vessel pressure was found and checked to see if it exceed ed the ASME limit for vessel press ure. This was important because an overpressure event could cause failure of the reactor vessel. The BWR/4 Though the BWR/4 core contained both GE14 and GNF2 Advantage fuel bundles, only the GE14 results are reported below. The GNF2 Advantage results requ ire d further sensitivity studies to determine the impact of shutdown exposure variation on the safety factors. Delta Critical Power Ratio increased core flow, normal feedwater temperature domain. The results are br oken down by transient below. The results were split into EOC and MOC results. The EOC results were further split into with and without bypass. Table 3 2 shows the corrected Load Rejection, No Bypass 1 cycles were larger than the nominal cycle; however, these differences were less than 0.01 and thus not significant. The sensitivity was very small, on the order of 0.003 and is conside red an equivalent result. Figure 3 flow domain was the most limiting domain for all cases.

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59 Turbine Trip, No Bypass The same trends were seen in the TTNBP results as were seen in the LRNBP results the increased core flow exposure cases that fell short of the nominal exposure. The GE14 values of the varied exposure cases were not significantly larger than those of the nominal case, and were smaller than the was very small, on the order of 0.003 and is considered an equivalent result. Figure 3 for TTNBP. Feedwater Control Failure The FWCF transients in the nominal case had several challenges by the varied exposure ain had the most 1000 MWd/ST, +500 MWd/ST, and +1000 MWd/ST cases were all within 0.01of that of the nominal, a trend that was also apparent in the ICF, RFW domain. The sensitivity was very small, on the order of 0 .003 and is considered an equivalent operable. In both the ICF, NFW and ICF, RFW domains, the three exposure cases mentioned previously were very close to the limiting va lues set by the nominal case. Thermal and Mechanical Over Power In terms of overpower s the ICF, RFW domain was more limiting for GE14 thermal overpower (TOP), while the ICF, NFW was more limiting for mechanical overpower (MOP). Table 3 3 shows the marg in to the screening criterion for thermal overpower for GE14 and Table 3 4 shows the margin to the screening criterion for mechanical overpower for GE14 The results are broken down by transient below.

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60 Load Rejection, No Bypass For the LRNBP transient, t he nominal, increased core flow domain had the smallest margins to the TOP and MOP limits. For all transients, the nominal exposure case was the most limiting for both TOP and MOP. For TOP, the EOC nominal exposure case had a margin of 13.5% to the scree ning criterion. The closest TOP value was 14.5% in the EOC +1000 MWd/ST exposure case. For MOP, the EOC nominal shutdown exposure domain had the smallest margin of 13.5% and was the most limiting for all transients The next smallest margin was 1 4 5 % i n the EOC +1000 MWd/ST case. Turbine Trip, No Bypass The TTNBP transient TOP and MOP results were similar to those of LRNBP. Once again, the EOC nominal, increased core flow domain had the smallest margins to the TOP and MOP screening criteria. For TOP, the nominal exposure case had the smallest margin of 14.1%; the next smallest was in the +1000 MWd/ST case, which had a margin of 15.0 % to the screening criterion. For MOP, the nominal exposure domain had a margin of 14.1% to the screening criterion. The next closest case was the EOC +1000 MWd/ST case, which had a margin of 15.0%. Feedwater Control Failure For the FWCF transient, much smaller margins were observed for TOP and MOP. The EOC nominal exposure case had the smallest margins to the screen criter ion for both TOP and MOP, with values of 11.7% for both for FWCF and 6.4% for both for FWCF/NBP. Vessel Over Pressure Analyzed at end of cycle, t he nominal case had the most limiting values in terms of vessel over pressure. The MSIVF transient tradition ally has the smallest margin to the ASME vessel pressure limit. For increased core flow, the peak dome pressure was 1301 psig, less than the

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61 limit of 1325 psig. The peak vessel pressure at increased core flow was 1337 psig, less than the limit of 1375 p sig. For low core flow operation, the peak dome pressure was 1298 and the peak vessel pressure was 1327. Therefore, the varied exposure cases did not challenge the ASME pressure limit. Table 3 5 shows the peak vessel pressure results for the BWR/4. The BWR/6 Delta Critical Power Ratio temperature domain. The results are broken down by transient below. Table 3 6 shows the Load Rejection, No Bypas s 1000 MWd/ST case was only 0.003 greater than that of the nominal, which in nor mal transient analysis was not considered a significant difference. The other varied exposure cases showed case, which was 0.046 less than that of the nominal. Figure 3 values. In the low core flow domain, the +1000 MWd/ST case showed a 0.101 improvement over the nominal case. All of the varied exposure cases showed improvement over the nominal in the low core flow domain. Turbine Trip, No Bypass Because the TTNBP transient is very similar to the LRNBP, it was expected that those results would follow the same trends as the LRNBP ones. This proved true; only the 1000

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62 it. The increased core flow domain was the more limiting of the two domains, and as the N 1 F eedwater Control Failure The FWCF transient in the nominal case was the other burn strategies. Once again the increased core flow, normal feedwater temperature case was the most limiting, followed by the increased core flow, reduced feedwater temperature case. The 1000 MWd/ST case came close to challenging the nominal case in both the ICF, NFW and rtable margin in the varied exposure cases to the nominal Pressure Regulator Failure, Downstream The PRFDS transient only occurred in two domains: increased core flow, normal feedwater temperature, and increased core flow, reduced feedwater tem perature. Both of these domains caused the varied exposure cases to challenge the nominal case during other transients and this trend continued during the PRFDS transients. For the increased core flow, reduced feedwater temperature, the 1000 MWd/ST case was 0.005 greater than the nominal case. It was not a significant difference. In the increased core flow, reduced feedwater temperature domain, both of the smaller exposure cases challenged the nominal case. While the differences were not significant (0 .005, 0.002), the difference was shown. In both domains, the cases not noted had that the PRFDS is not close to the limiting TTNBP event.

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63 Thermal and Mechanical Over Power The increased core flow, reduced feedwater temperature domain was the most limiting for both thermal and mechanical overpower. Table 3 7 shows the margin to the screening criterion for thermal overpower for GE14 and Table 3 8 shows the margin to the screening criterion for mechanical overpower The results are broken down by transient below. Load Rejection, No Bypass For LRNBP, the increased core flow domain for the nominal case was the most limiting. For TOP, there was a margin of 34.5%, and for MOP, a margin of 34.2%. The case with the closes t margins to the nominal case was the increased core flow, normal feedwater temperature, 1000 MWd/ST case, which had a TOP margin of 34.7% and a MOP margin of 34.7% Turbine Trip, No Bypass The TTNBP transient showed the same trends as the LRNBP. The TO P margin for the increased core flow, nominal domain was 33.6% and the MOP margin was 33.3 % The case with the closet margins was the increased core flow, 1000 MWd/ST case, which had margins about 1% greater. Feedwater Control Failure The case with the s mallest margin during a FWCF transient was the nominal case, increased core flow and reduced feedwater temperature. The TOP margin was 31.8% and the MOP margin was 27.8%. The closest margins were of the +500 MWd/ST case, increased core flow and reduced feedwater temperature domain. The TOP margin for that domain was 32.1% and the MOP margin was 29.5%. Pressure Regulator Failure, Downstream The most limiting case in terms of TOP and MOP for the PRFDS transient was the 1000 MWd/ST, increased core flow and reduced feedwater temperature, however there are still large

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64 margins for this event. The TOP margin was 17.2 % and the MOP margin was 38.7 %. The clos est margins were those of the 5 00 MWd/ST, increased core flow and reduced feedwater temperature; the TOP margin was 20 0 % and the MOP margin was 4 1.2 %. The nominal case, increased core flow and reduced feedwater temperature domain had a TOP marg in of 22.9% and a MOP margin of 44.3 %. Vessel Over Pressure For the increased core flow, normal feedwater te mperature and the low core flow, normal feedwater temperature domains, the nominal case had the most limiting peak dome and vessel pressures. For the ICF, NFW domain, the nominal case had a peak dome pressure of 1271 psig and a peak vessel pressure of 129 9. For the LCF, NFW domain, the nominal case had a peak dome pressure of 1269 psig and a peak vessel pressure of 1290 psig. All of those values are comfortably below the limits of 1325 psig for peak dome pressure and 1375 psig for peak vessel pressure. The case that came closest to the nominal case for both domains was the 500 MWd/ST case, which was still had peak pressures several psi below the nominal values reported above. Table 3 9 shows the peak vessel pressures for the BWR/6.

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65 Figure 3 1. Cr itical power ratio versus time for BWR/6 LRNBP.

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66 Figure 3 Note: The following describes the domain labels for the chart above. Domain Description ICF___TN Increased Core Flow, Hard Bottom Burn Licensing power shape IF_P1ETN Increased Core Flo w, N 1 Exposure +1000, EOC, normal FW temp. IF_P5ETN Increased Core Flow, N 1 Exposure +500, EOC, normal FW temp. IF_ 5ETN Increased Core Flow, N 1 Exposure 500, EOC, normal FW temp. IF_ 1ETN Increased Core Flow, N 1 Exposure 1000, EOC, normal FW temp MEO___TN MELLLA (Low) Core Flow, Hard Bottom Burn Licensing power shape LF_P1ETN Low Core Flow, N 1 Exposure +1000, EOC, normal FW temp. LF_P5ETN Low Core Flow, N 1 Exposure +500, EOC, normal FW temp. LF_ 5ETN Low Core Flow, N 1 Exposure 500, EOC, n ormal FW temp. LF_ 1ETN Low Core Flow, N 1 Exposure 1000, EOC, normal FW temp.

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67 Table 3 1. The BWR/4 EOC GE14 c r esults GE14 c EOC18 1000 MWd/ST C19 n ominal EOC18 500 MWd/ST C19 n ominal EOC18 PEOC C19 HBB EOC18 +500 MWd/ ST C19 n ominal EOC18 +1000 MWd/ST C19 n ominal LRNBP ICF 0.359 0.327 0.358 0.360 0.361 TTNBP ICF 0.355 0.323 0.353 0.354 0.356 FWCF ICF 0.356 0.327 0.353 0.354 0.354 FWCF/NBP ICF 0.395 0.372 0.392 0.394 0.393 LRNBP LCF 0.306 0.238 0.317 0.308 0.306 TT NBP LCF 0.302 0.237 0.313 0.300 0.303 FWCF LCF 0.298 0.239 0.309 0.296 0.298 FWCF/NBP LCF 0.343 0.281 0.357 0.340 0.343 FWCF ICF, RFW 0.343 0.336 0.340 0.344 0.342 FWCF/NBP ICF, RFW 0.382 0.375 0.381 0.384 0.383 FWCF LCF, RFW 0.298 0.262 0.303 0.298 0 .299 FWCF/NBP LCF, RFW 0.339 0.318 0.345 0.339 0.339

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68 Table 3 2. The BWR/4 MOC GE14 esults GE14 c EOC18 1000 MWd/ST C19 n ominal EOC18 500 MWd/ST C19 n ominal EOC18 PEOC C19 HBB EOC18 +500 MWd/ST C19 n ominal EOC18 +1000 M Wd/ST C19 n ominal LRNBP ICF, MOC 0.309 0.253 0.321 0.310 0.308 TTNBP ICF, MOC 0.306 0.250 0.315 0.306 0.304 FWCF ICF, MOC 0.308 0.253 0.319 0.308 0.306 LRNBP LCF, MOC 0.251 0.187 0.270 0.250 0.251 TTNBP LCF, MOC 0.248 0.186 0.266 0.249 0.249 FWCF LCF MOC 0.247 0.187 0.262 0.241 0.243

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69 Figure 3 3. The LRNBP the BWR/4. The following describes the domain labels for the chart above. Domain Description ICF___TN Increased Core Flow, Hard Bottom Burn Licensing power shape ICF___BN Increased Core Flow, Under Burn Licensing power shape IF_P1ETN Increased Core Flow, N 1 Exposure +1000, EOC, normal FW temp. IF_P5ETN Increased Core Flow, N 1 Exposure +500, EOC, normal FW temp. IF_ 5ETN Increased Core Flow, N 1 Exposure 500, EOC, normal FW temp. IF_ 1ETN Increased Core Flow, N 1 Exposur e 1000, EOC, normal FW temp. IF_P1MTN Increased Core Flow, N 1 Exposure +1000, MOC, normal FW temp. IF_P5MTN Increased Core Flow, N 1 Exposure +500, MOC, normal FW temp. IF_ 5MTN Increased Core Flow, N 1 Exposure 500, MOC, normal FW temp. IF_ 1MTN In creased Core Flow, N 1 Exposure 1000, MOC, normal FW temp. MEL___TN MELLLA (Low) Core Flow, Hard Bottom Burn Licensing power shape MEL___BN MELLLA (Low) Core Flow, Under Burn Licensing power shape LF_P1ETN Low Core Flow, N 1 Exposure +1000, EOC, normal FW temp. LF_P5ETN Low Core Flow, N 1 Exposure +500, EOC, normal FW temp. LF_ 5ETN Low Core Flow, N 1 Exposure 500, EOC, normal FW temp. LF_ 1ETN Low Core Flow, N 1 Exposure 1000, EOC, normal FW temp. LF_P1MTN Low Core Flow, N 1 Exposure +1000, MOC, normal FW temp. LF_P5MTN Low Core Flow, N 1 Exposure +500, MOC, normal FW temp. LF_ 5MTN Low Core Flow, N 1 Exposure 500, MOC, normal FW temp. LF_ 1MTN Low Core Flow, N 1 Exposure 1000, MOC, normal FW temp.

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70 Figure 3 4. The TTNBB d case for the BWR/ 4. The following describes the domain labels for the chart above. Domain Description ICF___TN Increased Core Flow, Hard Bottom Burn Licensing power shape ICF___BN Increased Core Flow, Under Burn Licensing power shape IF_P1ETN Increa sed Core Flow, N 1 Exposure +1000, EOC, normal FW temp. IF_P5ETN Increased Core Flow, N 1 Exposure +500, EOC, normal FW temp. IF_ 5ETN Increased Core Flow, N 1 Exposure 500, EOC, normal FW temp. IF_ 1ETN Increased Core Flow, N 1 Exposure 1000, EOC, no rmal FW temp. IF_P1MTN Increased Core Flow, N 1 Exposure +1000, MOC, normal FW temp. IF_P5MTN Increased Core Flow, N 1 Exposure +500, MOC, normal FW temp. IF_ 5MTN Increased Core Flow, N 1 Exposure 500, MOC, normal FW temp. IF_ 1MTN Increased Core Flo w, N 1 Exposure 1000, MOC, normal FW temp. MEL___TN MELLLA (Low) Core Flow, Hard Bottom Burn Licensing power shape MEL___BN MELLLA (Low) Core Flow, Under Burn Licensing power shape LF_P1ETN Low Core Flow, N 1 Exposure +1000, EOC, normal FW temp. LF_P5 ETN Low Core Flow, N 1 Exposure +500, EOC, normal FW temp. LF_ 5ETN Low Core Flow, N 1 Exposure 500, EOC, normal FW temp. LF_ 1ETN Low Core Flow, N 1 Exposure 1000, EOC, normal FW temp. LF_P1MTN Low Core Flow, N 1 Exposure +1000, MOC, normal FW temp. LF_P5MTN Low Core Flow, N 1 Exposure +500, MOC, normal FW temp. LF_ 5MTN Low Core Flow, N 1 Exposure 500, MOC, normal FW temp. LF_ 1MTN Low Core Flow, N 1 Exposure 1000, MOC, normal FW temp.

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71 Table 3 3. The BWR/4 GE14 t hermal o verpower r esults : m arg in to s creening c riteria GE14 TOP m argin (%) EOC18 1000 MWd/ST C19 n ominal EOC18 500 MWd/ST C19 n ominal EOC18 PEOC C19 HBB EOC18 +500 MWd/ST C19 n ominal EOC18 +1000 MWd/ST C19 n ominal LRNBP ICF 14.8 19.1 13.5 14.6 14.5 TTNBP ICF 15.3 19.7 14.1 15.5 1 5 FWCF ICF 13.4 16.6 11.7 13.4 13.4 FWCF/NBP ICF 8.1 11.1 6.4 8.1 8.1 LRNBP LCF 23.4 29.5 21.5 23.7 24 TTNBP LCF 24.2 29.8 21.9 24.9 24.9 FWCF LCF 22.2 27.5 19.7 23.1 22.9 FWCF/NBP LCF 17.2 22.5 13.9 18.1 17.9 FWCF ICF, RFW 13.7 14.7 12 13.5 13.4 F WCF/NBP ICF, RFW 8.6 9.6 6.8 8.5 8.3 FWCF LCF, RFW 19.8 22.4 17.4 20.1 19.7 FWCF/NBP LCF, RFW 14.7 15.5 12.3 15.1 14.9 LRNBP ICF, MOC 15.9 23.6 15.1 16.7 16.9 TTNBP ICF, MOC 16 24.3 15.5 16.9 17.4 FWCF ICF, MOC 13.2 21.6 13.3 14.3 14.3 LRNBP LCF, MOC 25.4 31.8 23.3 26.7 25.9 TTNBP LCF, MOC 26 32.1 24 27.1 26.1 FWCF LCF, MOC 23.1 30.5 22.1 24.7 23.7

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72 Table 3 4. The BWR/4 GE14 m echanical o verpower r esult s: margin to screening c riteria GE14 MOP margin (%) EOC18 1000 MWd/ST C19 n ominal EOC18 500 M Wd/ST C19 n ominal EOC18 PEOC C19 HBB EOC18 +500 MWd/ST C19 n ominal EOC18 +1000 MWd/ST C19 n ominal LRNBP ICF 14.8 19.1 13.5 14.6 14.5 TTNBP ICF 15.3 19.7 14.1 15.5 15 FWCF ICF 13.4 16.6 11.7 13.4 13.4 FWCF/NBP ICF 8.1 11.1 6.4 8.1 8.1 LRNBP LCF 23.2 2 9.5 21.5 23.4 23.6 TTNBP LCF 23.8 29.8 21.9 24.4 24.2 FWCF LCF 20.6 27.5 19.7 21.4 20.9 FWCF/NBP LCF 15.4 22.5 13.9 16.2 15.7 FWCF ICF, RFW 13.5 14.7 12 13 12.8 FWCF/NBP ICF, RFW 8.6 9.6 6.8 8.1 7.8 FWCF LCF, RFW 17.6 22.4 17.4 18.1 17.6 FWCF/NBP LC F, RFW 12.3 15.5 12.3 12.9 12.5 LRNBP ICF, MOC 15.9 23.6 15.1 16.3 16.7 TTNBP ICF, MOC 16 24.3 15.5 16.5 17.2 FWCF ICF, MOC 13 21.6 11.9 13.8 14.3 LRNBP LCF, MOC 24.2 31.7 22.9 25.2 25 TTNBP LCF, MOC 24.8 31.8 23.3 25.6 25.3 FWCF LCF, MOC 21.8 30.4 2 0.2 23.2 22.8

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73 Table 3 5 The BWR/4 peak vessel pressure results Peak v essel p ressure (psig) EOC18 1000 MWd/ST C19 n ominal EOC18 500 MWd/ST C19 n ominal EOC18 PEOC C19 HBB EOC18 +500 MWd/ST C19 n ominal EOC18 +1000 MWd/ST C19 n ominal MSIVF ICF 1333. 3 1325.7 1336.8 1333.7 1334.2 MSIVF LCF 1322.2 1312.7 1327.3 1322.2 1322.3

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74 Table 3 6 The BWR/6 GE14 c r esults GE14 c EOC12 1000 MWd/ST C13 n ominal EOC12 500 MWd/ST C13 n ominal EOC12 PEOC C13 HBB EOC12 +500 MWd/ST C13 n ominal EOC12 +1000 MWd/ST C13 n ominal LRNBP ICF 0.246 0.242 0.243 0.209 0.197 TTNBP ICF 0.247 0.244 0.248 0 .212 0.200 FWCF ICF 0.211 0.212 0.220 0.176 0.163 PRFDS ICF 0.140 0.136 0.140 0.122 0.120 LRNBP LCF 0.174 0.169 0.194 0.117 0.093 TTNBP LCF 0.176 0.173 0.201 0.125 0.098 FWCF LCF 0.149 0.147 0.169 0.113 0.093 FWCF ICF/RFW 0.241 0.238 0.243 0.236 0.23 8 PRFDS ICF/RFW 0.140 0.137 0.135 0.124 0.129

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75 Figure 3 6. The LRNBP the BWR/6. The following describes the domain labels for the chart above. Domain Description ICF___TN Increased Core Flow, Hard Bottom Burn Licensing power shape IF_P1ETN Increased Core Flow, N 1 Exposure +1000, EOC, normal F W temp. IF_P5ETN Increased Core Flow, N 1 Exposure +500, EOC, normal FW temp. IF_ 5ETN Increased Core Flow, N 1 Exposure 500, EOC, normal FW temp. IF_ 1ETN Increased Core Flow, N 1 Exposure 1000, EOC, normal FW temp. MEO___TN MELLLA (Low) Core Flow, Hard Bottom Burn Licensing power shape LF_P1ETN Low Core Flow, N 1 Exposure +1000, EOC, normal FW temp. LF_P5ETN Low Core Flow, N 1 Exposure +500, EOC, normal FW temp. LF_ 5ETN Low Core Flow, N 1 Exposure 500, EOC, normal FW temp. LF_ 1ETN Low Core Fl ow, N 1 Exposure 1000, EOC, normal FW temp.

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76 Table 3 7 The BWR/6 thermal o verpower r esults : m argin to s creening c riteria GE14 TOP m argin (%) EOC12 1000 MWd/ST C13 n ominal EOC12 500 MWd/ST C13 n ominal EOC12 PEOC C13 HBB EOC12 +500 MWd/ST C13 n omi nal EOC12 +1000 MWd/ST C13 n ominal LRNBP ICF 34.7 35.5 34.5 38 37.3 TTNBP ICF 34.6 35.4 33.6 37.7 37 FWCF ICF 38.2 38.5 35 41 41.2 PRFDS ICF 20.9 23.7 22.6 24.4 24.8 LRNBP LCF 42.4 43.2 41.7 50 52 TTNBP LCF 41.5 42.4 40.8 48.9 52 FWCF LCF 44.4 44.8 43.1 42.6 42.5 FWCF ICF/RFW 36.2 34.4 31.8 32.1 37 PRFDS ICF/RFW 17.2 20 22.9 22.1 21.5

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77 Table 3 8 The BWR/6 m echanical o verpower r esults : m argin to s creening c riteria GE14 MOP Margin (%) EOC12 1000 MWd/ST C13 n ominal EOC12 500 MWd/ST C13 n omina l EOC12 PEOC C13 HBB EOC12 +500 MWd/ST C13 n ominal EOC12 +1000 MWd/ST C13 n ominal LRNBP ICF 34.7 35.5 34.2 37.8 37 TTNBP ICF 34.6 35.4 33.3 37.5 36.7 FWCF ICF 38.2 38.5 34.7 40.9 41 PRFDS ICF 42.1 44.8 43.9 45.3 45.7 LRNBP MEL 42.2 43 41.7 47.4 49.1 TTNBP MEL 41.3 42.2 40.8 46.4 48.8 FWCF MEL 43.9 43.9 43.1 42.4 42.5 FWCF ICF/RFW 32 30.1 27.8 29.5 32.2 PRFDS ICF/RFW 38.7 41.2 44.3 42.4 42.1

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78 Table 3 9 The BWR/6 peak vessel pressure results Peak v essel p ressure (psig) EOC12 1000 MWd/ST C13 n ominal EOC12 500 MWd/ST C13 n ominal EOC12 PEOC C13 HBB EOC12 +500 MWd/ST C13 n ominal EOC12 +1000 MWd/ST C13 n ominal MSIVF ICF 1295.3 1295.7 1298.6 1291.4 1290.4 MSIVF LCF 1285.4 1285.9 1289.5 1279.9 1276.4

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79 CHAPTER 4 CONCLUSIONS This research a more intensive study of the effect o f previous cycle exposure than is performed in a normal reload licensing analysis found that the current licensing process at GE Hitachi Nuclear for transient analysis is acceptably bounding. This result agreed with the i nitial hypothesis and the results from previous studies of smaller exposure windows. For the BWR/4, the nominal exposure case had the most limiting over pressure results. In the differences were not significant. Likewise, the overpower results we re not significantly affected and all cases showed acceptable margins to the screening criteria. In addition, the screening criteria are known to be conservative and there is additional margin to fuel melt and cladding strain criteria. For the BWR/6, the variation in shutdown exposure did not seem to have a large impact on the operating limits determined using the projected HBB. Interestingly, for the BWR/6 the exposure cases that fell short of the projected end of cycle exposure were more limiting than t hose from the lengthened cycle exposure cases. Previous studies have supported the 500 MWd/ST exposure window. Overall, this study confirms that the 500 MWd/ST exposure steps did not significantly challenge the limits, and therefore a change of fiftee n days in the shutdown will be acceptably bounded by the transient analyses done for the reload core using the projected end of cycle data for plants using GE14 fuel. In addition, these results also support that the 1000 MWd/ST sensitivities are suffici ently small that the licensing results are acceptable. The licensee would continue exposure monitoring through Cycle N to assure compliance with the licensing basis power shape assumptions.

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80 LIST OF REFERENCES Advanced Boiling Water Plant General Descript ion. 2006. General Electric, Wilmington, North Carolina. BWR/6 General Description of a Boiling Water Reactor. 1980. Proposal Engineering, Nuclear Power Systems Engineeri ng Department, General Electric, San Jose, California. DeFilippis, M., Higgins, R. 2005. GE14 Fuel Assembly Mechanical Desi gn Report. Global Nuclear Fuel, Wilmington, North Carolina. GNF2 Advantage Generic Compliance with NEDE 24011 P A (GESTAR I I). 2007. Global Nuclear Fuel, Wilmington, North Carolina. Ishigai, Seikan (Ed.) 1999. Steam Power Engineering: Thermal and Hydraulic Design Principles. Cambridge University Press, New York City Lahey, F., Moody, F. 1993. The Thermal hydraulics of a Boiling Water Nuclear Reactor. American Nuclear Society. La Grange Park, Illinois. Lamb, Charles F. 2007. Transient Train ing. GE Hitachi Nuclear Energy, Wilmington, North Carolina. Supplemental Safety Evaluation for the General Electric Topical Report Qualification of the One Dimensional Core Transient Model for Boiling Water Reactors. 19 81. Reactor Systems Branch, DSI, Nuclear Regulatory Commission, Washington, D.C. Todreas, N., Kazimi, M. 1990. Nuclear Systems: Elements of Thermal Hydraulic Design. Taylor & Francis Group Hemisphere, N ew Y ork Watford, G. A. (Ed.) 2000. General El ectric Standard Application for Reactor Fuel (Supplement for Unite d States). Global Nuclear Fuel, Wilmington, North Carolina.

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81 BIOGRAPHICAL SKETCH Lauren Elizabeth Nalepa was born in 1985 in Biloxi, Mississippi, to Robert and Lisa Nalepa. She has two yo unger siblings, Ryan and Emma. Lauren attended elementary school in graduate of Kadena High School on Kadena Air Base, Okinawa, Japan. A second generation Gat o r, Lauren received her B.S. in n uclea r e ngineering in May 2007 and began graduate school the next fall. In school, Lauren was active in Phi Sigma Rho, Tau Chapter. She served as Vice President of Standards her junior and senior years. She was also secre tary of the UF student chapter of the American Nuclear Society, and presented a paper at the 2006 ANS Student Conference at Rensselaer Polytechnic Institute and a poster at the 2007 ANS Winter Conference in Washington, D.C. Lauren interned for six semester Crystal River nuclear plant in Crystal River, Florida, and GE Hitachi Nuclear Energy in Engineering Developme nt Program as a nuclear engineer in Wilmington, North Carolina. Lauren enjoys traveling, reading, baking, tennis, and Disney World.