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New Reactor Design and Regulation

Permanent Link: http://ufdc.ufl.edu/UFE0022266/00001

Material Information

Title: New Reactor Design and Regulation
Physical Description: 1 online resource (61 p.)
Language: english
Creator: Chomat, Alejandro
Publisher: University of Florida
Place of Publication: Gainesville, Fla.
Publication Date: 2008

Subjects

Subjects / Keywords: nuclear, reactors
Nuclear and Radiological Engineering -- Dissertations, Academic -- UF
Genre: Nuclear Engineering Sciences thesis, M.E.
bibliography   ( marcgt )
theses   ( marcgt )
government publication (state, provincial, terriorial, dependent)   ( marcgt )
born-digital   ( sobekcm )
Electronic Thesis or Dissertation

Notes

Abstract: This study will address the improved safety features of advanced light water reactor (ALWR) systems that are under consideration for near term deployment in the Unites States and abroad. The safety features of ALWR systems can be divided into two major categories based on their reliance on passive or active design for the emergency core cooling system. The ALWR systems evaluated in this study are actively safe ABWR, EPR and APWR, and passively safe AP 1000 and ESBWR. Each of the passive or active ALWR systems includes significantly improved safety features and larger operational margins. In each of the BWR and PWR categories, actively cooled ABWR and EPR feature the lowest average LHGR. The AP 1000 and the ESBWR feature passive safety systems that use gravity-driven flow for emergency core cooling with no need for secondary power sources such as emergency diesel generators (EDGs) in case of loss of off-site power. The ABWR and the EPR both use the proven technology and use safety grade EDGs. They must be maintained and have surveillances performed like existing diesels at current nuclear reactor sites. There is also a finite probability of EDGs failing to start or to run for a variety of causes including problems associated with cooling, engine, fuel oil, generator, instrumentation and control, breaker, and starting air. Though the lack of safety grade EDGs in passive systems is a clear advantage, the ABWR and EPR by employing multiple and redundant EDGs and the APWR by utilizing the GTG provide for an order of magnitude safety improvement over the current generation of operating LWR reactors in US.
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 Alejandro Chomat.
Thesis: Thesis (M.E.)--University of Florida, 2008.
Local: Adviser: Anghaie, Samim.

Record Information

Source Institution: UFRGP
Rights Management: Applicable rights reserved.
Classification: lcc - LD1780 2008
System ID: UFE0022266:00001

Permanent Link: http://ufdc.ufl.edu/UFE0022266/00001

Material Information

Title: New Reactor Design and Regulation
Physical Description: 1 online resource (61 p.)
Language: english
Creator: Chomat, Alejandro
Publisher: University of Florida
Place of Publication: Gainesville, Fla.
Publication Date: 2008

Subjects

Subjects / Keywords: nuclear, reactors
Nuclear and Radiological Engineering -- Dissertations, Academic -- UF
Genre: Nuclear Engineering Sciences thesis, M.E.
bibliography   ( marcgt )
theses   ( marcgt )
government publication (state, provincial, terriorial, dependent)   ( marcgt )
born-digital   ( sobekcm )
Electronic Thesis or Dissertation

Notes

Abstract: This study will address the improved safety features of advanced light water reactor (ALWR) systems that are under consideration for near term deployment in the Unites States and abroad. The safety features of ALWR systems can be divided into two major categories based on their reliance on passive or active design for the emergency core cooling system. The ALWR systems evaluated in this study are actively safe ABWR, EPR and APWR, and passively safe AP 1000 and ESBWR. Each of the passive or active ALWR systems includes significantly improved safety features and larger operational margins. In each of the BWR and PWR categories, actively cooled ABWR and EPR feature the lowest average LHGR. The AP 1000 and the ESBWR feature passive safety systems that use gravity-driven flow for emergency core cooling with no need for secondary power sources such as emergency diesel generators (EDGs) in case of loss of off-site power. The ABWR and the EPR both use the proven technology and use safety grade EDGs. They must be maintained and have surveillances performed like existing diesels at current nuclear reactor sites. There is also a finite probability of EDGs failing to start or to run for a variety of causes including problems associated with cooling, engine, fuel oil, generator, instrumentation and control, breaker, and starting air. Though the lack of safety grade EDGs in passive systems is a clear advantage, the ABWR and EPR by employing multiple and redundant EDGs and the APWR by utilizing the GTG provide for an order of magnitude safety improvement over the current generation of operating LWR reactors in US.
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 Alejandro Chomat.
Thesis: Thesis (M.E.)--University of Florida, 2008.
Local: Adviser: Anghaie, Samim.

Record Information

Source Institution: UFRGP
Rights Management: Applicable rights reserved.
Classification: lcc - LD1780 2008
System ID: UFE0022266:00001


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NEW REACTOR DESIGN AND REGULATION


By

ALEJANDRO CHOMAT















A THESIS PRESENTED TO THE GRADUATE SCHOOL
OF THE UNIVERSITY OF FLORIDA IN PARTIAL FULFILLMENT
OF THE REQUIREMENTS FOR THE DEGREE OF
MASTER OF ENGINEERING

UNIVERSITY OF FLORIDA

2008































2008 Alejandro Chomat































To my family









ACKNOWLEDGMENTS

I would like to take this opportunity to thank all my advisors, family, and friends for all of

their support.









TABLE OF CONTENTS

Page

A CK N O W LED G M EN T S ................................................................. ........... ............. .....

LIST OF TABLES ......... .........................................................7

LIST OF FIGURES .................................. .. ..... ..... ................. .8

LIST O F A B B R EV IA TIO N S ......... ................. .................................................................. 9

A B S T R A C T ......... ....................... ............................................................ 12

CHAPTER

1 BACKGROUND ....................................... ..... .............................. 14

2 SCOPE OF STUDY ............................... .................................. 21

3 NEW REACTOR LICENSING PROCESS ........................................ ....................... 22

E early S ite P erm it ....................................................................... 2 2
C om binded O operating L license ...............................................................................................23
D design C certification ......... ............................................. .. ... ...... ......24

4 R E A C T O R D E SIG N S ........................................................................... ........................... 27

A P 1 0 0 0 ............................................................................2 7
R e liab ility ..............................................................2 9
T technology M maturity ....................................... .... .... ........ .............. 30
E c o n o m ic s ....................................................................................................................... 3 0
S u p p ly C h a in ........................................................................................................ 3 1
A B W R ................... ................... ...................1..........
R e liab ility ..............................................................3 4
T ech n ology M atu rity ................................................................ ...............................34
E c o n o m ic s ...............................................................................3 5
S u p p ly C h a in ............................................................................................................. 3 5
E S B W R ................... ................... ...................5..........
R e liab ility ..............................................................3 7
T ech n ology M atu rity ................................................................ ...............................3 8
E c o n o m ic s ...............................................................................3 8
S u p p ly C h a in ............................................................................................................. 3 8
A P W R ................... ...................3...................9..........
R e liab ility ..............................................................4 0
T ech n ology M atu rity ................................................................ ...............................4 0
E c o n o m ic s ..............................................................................4 0
S u p p ly C h ain .............................................................................4 1









E P R .................................................................................................................................... 4 1
R e liab ility ..............................................................4 2
T technology M maturity ....................................... .... ...... ...... .............. 42
E c o n o m ic s ..............................................................................4 2
S u p p ly C h ain .............................................................................4 3

5 DISCU SSION ........... .... ..... ... ........... ....................... ........ 44

Robustness ........................................ 44
Safety .........................................45
R liability ................ .. ...............................................................4 8
Technology M aturity .............................................................48
E conom ics............... .........................................................49
S u p p ly C h ain ................................................................5 1
L icen seab ility ............................................................................... 5 1
S e cu rity ................... ...................5...................2..........
R isk and U uncertainty .............................................................53
S ite S p e c ific Issu e s ........................................................................................................... 5 4

6 CON CLU SION ... .................................................56

LIST OF REFERENCES ..............................................................................58

B IO G R A PH IC A L SK E T C H ................................................................. ................................6 1





























6









LIST OF TABLES

Table page

3-1 C chapter L ist ................ .......... ............................. .............................25

4-1 D ose to W workers A P 1000 .................................................................. ... ......................29

4-2 D ose to W workers A B W R ......................................................................... ....................34

4-3 D ose to W workers E SB W R ......................................................... ...................................37

5-1 L near H eat G generation R ate ..................................................................... ..................44

5-2 C D F and L R F ......... ...... ............ .................................... ..........................47

5-3 C capacity F actor ............. ..... ............ ............ .......................................48

5-4 R actor O overnight C ost .............. .................................................................................50

5-5 Operations and M maintenance Cost .............................................................................. 50

5-6 Regulatory Status ............ .................................. .................. ............. 52

5-7 Electric Power ............. ............................ ................... .........54









LIST OF FIGURES

Figure page

1-1 Simplified PW R ............... ............... ......... ........................... .... 15

1-2 Simplified BW R ............... ................. ........... .................... ......... 16

1-3 Sustained Reliability and Productivity..................................................... ..... ......... 17

1-4 T total C ost of G generation ................................................................................. .......... 18

1-5 Public Support ................................................. 18

1-6 Power Uprates ...................................................................... ......... 19

1 -7 S tate P o lic ie s ................................................................................................................ 2 0

3 -1 C O L R ev iew P ro cess ................................................................................................... 2 4

4-1 Passive Containment Cooling ......................................... ................. ....... 28

4-2 ABW R Reactor Pressure Vessel ................................................................. 32

4-3 C om prison of P SA R esults......................................................................................... 33

4 -4 N atu ral C ircu latio n ....................................................................................................... 3 7

4-5 EPR Component Reduction ........................ ........ ..........43

5 1 E D G F a ilu re s ............................................................................................................... 4 6






















8









LIST OF ABBREVIATIONS

ABWR Advanced Boiling Water Reactor

ADS Automatic Depressurization System

ALWR Advanced Light Water Reactor

APWR Advanced Pressurized Water Reactor

B Boron

BWR Boiling Water Reactor

CDF Core Damage Frequency

CFR Code of Federal Regulation

CMT Core Makeup Tank

COL Combined Operating License

COLAs Combined Operating License Application

CST Condensate Storage Tank

DCD Design Control Document

ECCS Emergency Core Cooling System

EDG Emergency Diesel Generator

EPR Evolutionary Power Reactor

ESBWR Economically Simplified Boiling Water Reactor

FOAK First of a Kind

FPL Florida Power and Light

GE General Electric

GTG Gas Turbine Generator

HPCF High Pressure Core Flooder

IC Isolation Condenser

IRWST In-Containment Refueling Water Storage Tank









ITAAC Inspections, Test, Analyses, and Acceptance Criteria

LHGR Linear Heat Generation Rate

LOCA Loss of Coolant Accident

LRF Large Release Frequency

LWR Light Water Reactor

m meter

MOX Mix Oxide Fuel

MTU Metric Ton Uranium

MWD Mega Watt Day

MWe Mega Watts electric

MWh Mega Watt Hour

MWt Mega Watts thermal

NRC Nuclear Regulatory Commission

NUREG US Nuclear Regulatory Commission Regulation

PCC Passive Containment Cooling

PRA Probabilistic Risk Assessment

PSI Pounds per Square Inch

PSIA Pounds per Square Inch Absolute

PSIG Pounds per Square Inch Gauge

PWR Pressurized Water Reactor

RCIC Reactor Core Isolation System

RCP Reactor Coolant Pump

RCS Reactor Coolant System

RHR Residual Heat Removal

RPV Reactor Pressure Vessel









SER Safety Evaluation Report

SLCS Standby Liquid Control System

SP Suppression Pool









Abstract of Thesis Presented to the Graduate School
of the University of Florida in Partial Fulfillment of the
Requirements for the Degree of Master of Engineering

NEW REACTOR DESIGN AND REGULATION

By

Alejandro Chomat

August 2008

Chair: Samim Anghaie
Major: Nuclear Engineering Sciences

This study will address the improved safety features of advanced light water reactor

(ALWR) systems that are under consideration for near term deployment in the Unites States and

abroad. The safety features of ALWR systems can be divided into two major categories based on

their reliance on passive or active design for the emergency core cooling system. The ALWR

systems evaluated in this study are actively safe ABWR, EPR and APWR, and passively safe AP

1000 and ESBWR. Each of the passive or active ALWR systems includes significantly improved

safety features and larger operational margins. In each of the BWR and PWR categories, actively

cooled ABWR and EPR feature the lowest average LHGR. The AP 1000 and the ESBWR

feature passive safety systems that use gravity-driven flow for emergency core cooling with no

need for secondary power sources such as emergency diesel generators (EDGs) in case of loss of

off-site power. The ABWR and the EPR both use the proven technology and use safety grade

EDGs. They must be maintained and have surveillance performed like existing diesels at current

nuclear reactor sites. There is also a finite probability of EDGs failing to start or to run for a

variety of causes including problems associated with cooling, engine, fuel oil, generator,

instrumentation and control, breaker, and starting air. Though the lack of safety grade EDGs in

passive systems is a clear advantage, the ABWR and EPR by employing multiple and redundant









EDGs and the APWR by utilizing the GTG provide for an order of magnitude safety

improvement over the current generation of operating LWR reactors in US.









CHAPTER 1
BACKGROUND

Nuclear power reactors play a huge role in our nation's energy diversity plan. Currently,

20% of the electric needs of the U.S. are provided by nuclear reactors. There are two forms of

nuclear reactors used for commercial power production in the U.S., the pressurized water reactor

(PWR) and the boiling water reactor (BWR). In total there are 104 operating commercial

reactors of which 69 are PWRs and 35 are BWRs.1 In order for these reactors to be operating

they must be approved by the Nuclear Regulatory Commission (NRC) which continues to

monitor their safety during all phases of plant life from construction to decommissioning. Both

PWRs and BWRs have a unique history that has led to the development and success of the

reactor designs.

The first PWR built was the Shippingport Atomic Power Station, in Pennsylvania. The

reactor began commercial operation in 1958 and ran until 1982. The Shippingport Atomic Power

Station produced 60 MWe. But to this day, concepts used in this reactor still apply. These

concepts were originally developed from the navy nuclear programs.2

A pressurized water reactor consist of 3 or 4 loops in the primary side which is

radioactive, a secondary side which creates power by turning a turbine and a tertiary side which

provides an ultimate head sink for the system. Fuel in the form of uranium fuel assemblies is

controlled by both control rods and boron. As the fuel heats up it is cooled by the primary water.

The water is at 2250 PSIA which does not allow it to boil. The water then goes out the hot leg

into the steam generator. There are multiple hot legs out of the reactor pressure vessel one has a

pressurizer which maintains the system pressure. In the steam generator the water from the

reactor is cooled by water coming from the secondary system. The steam generator is either a U-

tube steam generator like most designs or a once through steam generator that was produced by










Babcock and Wilcox. The water from the secondary system is flashed to steam which will be

used to turn the turbine. The primary water then leaves the steam generator goes to the reactor

coolant pumps and back into the reactor. The primary side is mostly kept inside the containment

building to avoided radioactive release. The steam produced in the steam generator is then used

to turn the turbine which turns a shaft to create electricity in the generator. The steam is then

cooled in the condenser and pumped back into the steam generator to create steam again.3 In the

condenser the steam is cooled by the tertiary system. A basic PWR is seen in Figure 1-1 Basic

PWR.




Stem Line
Reactor Vessel P r P's e
Control Rods t T









BR ws s VcCl oding Wate

~Condenser

Figure 1-1. Simplified PWR

In the figure the system is cooled by a cooling tower but power plants also are cooled by

canal systems, lakes, or oceans. Note the above system is simplified; real systems are much more

complex and have many more components.

Thirty five of the nuclear reactors in the U.S. do not operate as described above but use the

concept of a BWR in which we boil the primary water to turn the turbine. The first commercial

BWR was GE's Valcitos BWR which was licensed by the United States Atomic Energy









Commission the predecessor to the NRC. Designs for both dual cycle and single cycle BWRs

exist. A dual cycle is not preferred because it has a higher capital cost associated with it.4 A

single cycle design takes the steam directly from the core and turns a turbine as seen in Figure 1-

2 Basic BWR.

ConlairiernlI 5Iructure

Reagt Reoels














Figure 1-2. Simplified BWR

As in the PWR the fuel heats up and the heat is removed by the primary water. In the BWR

design the primary water is allowed to boil. In order to remove the moisture left in the steam it is
passed through steam dryers and separators. If water were to pass through the turbine, the blades


are damaged, so steam is approximately at 100 percent quality. The system here is pressurized to

1000 PSIG.5 The primary steam then spins the turbine which spins a shaft in the generator to

create electricity. The steam is condensed in the condenser by the secondary system and pumped

back into the core. By using primary steam to spin the turbine and through the subsequent
systems there is more contamination tie e oninment building in a B n a .

Another major difference between the two is that the control rods in a BWR come in from the










bottom of the core and in a PWR they come in from the top of the core. A BWR does not use


boron for control during normal operation like a PWR it only uses boron during emergencies.3










Currently, the industry has undertaken tasks to start the licensing projects within the U.S.

Both vendors and utilities have announced and or submitted applications for various tasks. While

no utility has actually ordered a reactor, which has not been done since the 1970's, they have

initiated the process to seek early site permits from the NRC. Exelon, Systems Energy Resources

Inc., Dominion and Southern Nuclear Operating Company have all applied for early site permits

with many more utilities claiming they too will follow.6 In the state of Florida, Progress Energy

has selected a site in Levy County and submitted an application for building a new nuclear plant.

The other major Florida electric power utility, FPL, recently has received the approval from the

Public Service Commission to build a pair of new nuclear units at their existing site, Turkey

Point. Many other utilities have expressed intent to apply for COLAs.

Besides the ongoing work by the utilities and the vendors, the current work and

dependability of the current fleet is remarkable. As can be seen in Figure 1-3 Sustained

Reliability and Productivity the US nuclear power reactors have become more efficient. Gains in

productivity are based on our increased experience operating the existing 104 power plants.


Sustained Reliability and Productivity
U.S. Nuclear Capacity Factor
95-





-------- ----- -- --- -- -
- -. - -- --... - - - - - -


Hi 2 ED 84 'B5 86 807 8 *l O 91 Q9 -)3 94 'A !J6 ']1 -S *>0 1) 0. W. t* I (0 W Vi 06
8 E I _** .;....

Figure 1-3. Sustained Reliability and Productivity (Source:
http://nei.org/filefolder/thechangingclimatefornuclearenergy-
wallstreetbriefing2007.pdf. Last accessed July 21, 2007).










In Figure 1-4 Cost of Generation shows how nuclear energy is becoming more cost

efficient. In 2006 the fuel, operation, maintenance and Nuclear Waste Fund fee cost was 1.7

cents per kilowatt-hr.8



Cents

10.00
8.00 --- Nuclear
6.00 E Coal
4.00 a Gas
2.00 E Petroleum

Y -N

17 D C
Year

Figure 1-4. Total Cost of Generation

Another important factor we must consider is the public perception of nuclear energy. The

former head and co-founder of Green Peace Patrick Moore openly supports nuclear power even

though he used to oppose it. The American public also realizes the need for new sources of

energy and is supporting the growth of nuclear power as can be seen in Figure 1-5 Public

Support.



Strong Public Support Continues















Figure 1-5. Public Support
(Source:http://nei.org/filefolder/thechangingclimatefornuclearenergy-
wallstreetbriefing2007.pdf. Last accessed July 21, 2007).










Many say that since we are not building any new nuclear reactors we have not increased

nuclear power generation since the last new reactors came online in the early 1990's. But as can

be seen in Figure 1-6 Power Uprates that is not true and while there are no new plants we are

adding nuclear power generation by increasing power in existing plants.8



U.S. Nuclear Plant Uprates
Cumulative Capacity Additions at Existing Plants
2000-2011
6,000
1,383 MWe Expected
5,000



10057WeUndeRevie





2000 7001 7002 10.1 A004 J005 )01.6 S00? ?O-lI 2009 20 JC1 Poil


Figure 1-6. Power Uprates (Source:
http://nei.org/filefolder/thechangingclimatefornuclearenergy-
wallstreetbriefing2007.pdf. Last accessed July 21, 2007).

The government is also providing incentives for new nuclear power generation helping to

ease the worries of the utilities and the financial institutions that will fund the new construction.

The first 6000MWe built will receive $18/MWh produced tax break but the tax break will only

help the utility after construction.8 The federal government is also providing insurance for delays

caused by licensing or litigation, the first two plants will have $500 million dollar policies to

cover 100% delay cost and no waiting period for claims, the second four plants will have a $250

million dollar policy with only 50% delay cost after 6 months delay.8 The federal government

also proposed in the 2005 Energy Policy Act to provide loan guarantees for up to 80% of project










cost but the regulations will not be enacted until late 2007.8 Some state governments are adopting

a pronuclear policy as can be seen in Figure 1-7 State Policies.8



State Policies Supporting
Nuclear Construction
Legislation In place that helps secure Legislation under consideration that
financing elpsecurelinncing
Regulatory ploce)edings ongoing that S Legislation and regulation In place that
could help secure financing help secure financing












Figure 1-7. State Policies (Source: http://nei.org/filefolder/thechangingclimatefornuclearenergy-
wallstreetbriefing2007.pdf Last accessed July 21, 2007).

These states tend to have nuclear power plants already and utilities are more prone to apply for

licenses in the states. This also allows for a utility and investor to worry less about negative

reception from the citizens.









CHAPTER 2
SCOPE OF STUDY

This study examines the new technologies available for nuclear power plant selection and

the new regulatory procedure available. It will look at 10 Code of Federal Regulations Part 52

and describe the processes the CFR covers. The study looks at the safety, reliability, technology

maturity, economics, supply chains, licenseability, security, risk and uncertainties, and site

specific issues associated with the new reactor designs. Safety concerns with the performance of

engineered and passive systems that respond to an accident. Reliability is inclusive of issues

such as the capacity factor and ability to keep the plant online. Technology maturity deals with

the status of the engineering design and the system readiness to go to the construction phase.

Economics is divided into 2 sections; the initial capital cost and operation and maintenance cost.

Licenseability looks into the reactor design and its place in the current regulatory space. The risk

and uncertainties and site specific issues sections provide information on the problems that are

encountered with certain reactor designs.









CHAPTER 3
NEW REACTOR LICENSING PROCESS

To simplify nuclear reactor licensing the NRC has developed 10 Code of Federal

Regulations Part 52. Before utilities or investors would first apply for the construction license

and then later reapply to the commission for an operating license. This posed a problem because

most of the cost in a nuclear plant is construction cost; the cost of the fuel is much lower. Some

plants received there construction licenses but could not get approval for the operating license.

This also caused many utilities to cancel orders for other new power plants fearing that the same

fate would occur to their investment and the money would be lost due to the political landscape.

Regulators realizing the need for new nuclear power plants created the new licensing process

allowing for the hopeful revival of nuclear power plant construction.

Early Site Permit

The new license process is a 3 prong approach which will allow the utility greater ease

with the licensing process. The first part is the early site permit. The early site permit allows for

the utility to apply for the possibility to build a reactor or reactors at a site specifically identified

in the permit. Before a permit is granted hearings must take place which are called for in 10 CFR

Part 52. Once the permit is granted then it is valid for no less than 10 years but no more than 20

years. Also in the event that the permit expires but a proceeding on a combined license

application is in effect the permit remains valid. An early site permit may is to in a combined

licensed application at the discretion of the utility and not a requirement for a combined license

application. However it is expected that many utilities will first get the early site permit.9

The early site permit must contain all information required by 10 CFR Part 50.33 a through

d, the information required by 10 CFR Part 50.34 a 12 and b 10, and to the extent approval of

emergency plans is sought under paragraph b 2 ii of this section, the information required by 10









CFR Part 50.33 g and j, and 10 CFR Part 50.34 b 6 v. It must also contain the description and

safety assessment of the site. An evaluation of the major structures, systems, and components of

the facility that affect the acceptability of the site under the radiological consequence evaluation

factors described in 10 CFR Part 50.34 a 1. It also describes the number, type, and thermal power

rating of the plants, boundary of the site, general location of each facility, maximum effluents

produced, the cooling systems to be used, the geological site characteristics, the location and

purpose of nearby surroundings, and a future population profile for the site. Lastly a complete

environmental report is included with the application.9

Combinded Operating License

The next prong is the Combined Operating License COL, which allows the utility to both

construct and operate the plant. The purpose of this part was not to block the public from

objecting to a nuclear power plant but to streamline the construction process and to stop cost

overruns. This process allows for public hearings at the beginning of the licensing process and

during the acceptance of the Inspections, Test, Analyses, and Acceptance Criteria (ITAAC). Also

note that all the information required in 10 CFR Part 50 is required in 10 CFR Part 52 and that if

the utility chooses 10 CFR Part 50 is available for use. It also allows for the licensee to refer to

either the early site permit, the design certifications sought by the reactor vendors, or both. If the

utility chooses they submit their own design and not seek an early site permit but then the

information required in both the design certification and the early site permit must be included in

the COL. The ITAAC is required in the COL application. The ITAAC contains all the material,

locations, test, and criteria that must be met and certified. These are met at the end of

construction before the facility is used for commercial operations. If there is any modification to

the design of the plant during construction the new design must be submitted to the NRC for

review because it changes the license that the plant is authorized to operate under. 10 CFR Part










52 also allows the Commission to make effective an amendment to the license that involves no

significant hazardous conditions; the Commission also makes effective an amendment in

advance of holding a hearing on the proposed amendment if it deems appropriate.9

The NRC is confident that the new COL process provides a timely response so that new

nuclear reactors are built on schedule and without cost overruns due to regulatory problems.

David Matthews the director of the Division of New Reactor Licensing in a presentation given

on September 13, 2006 used Figure 3-1 COL Review Process to describe how the new proposed

COL applications would proceed.









I ,linh r' h -h -i..FP

ScL il .. 1=a .... .

.'12 Ii.n














Figure 3-1. COL Review Process (Source:
http://www.engr.utexas.edu/trtr/agenda/documents/Matthews-
StatusofNewReactorLicensingActivities.pdf Last accessed June 21, 2007).

Design Certification

The last major part of the new licensing process is the design certification. In the past

utilities ask a vendor to design a specific system that meets some utility requirements at a site of
Si- ~






















utilities ask a vendor to design a specific system that meets some utility requirements at a site of









their choosing. This causes all the sites in the US to be different in one form or another and no

standard design is kept. The new way of thinking is that all the new plants for a vendor have a

similar design. In order to ease the regulatory review during the COL application a utility

chooses a certified design. In order to obtain a certified design a vendor pays the U.S. NRC to

review a design. Once the NRC reviews the design if it is certified then a utility can refer to it

during the COL application; a utility can also refer to a design under review. A design

certification is valid for 15 years plus the duration of any ongoing activity for which the design is

used as long as the activity is docketed before the date of expiration.9

The design certification must contain all technical information required from applicants on

COL that are located in 10 CFR Part 20, 50, 73, and 100, all the technical information required

after Three Mile Island found in 10 CFR 50.34 f except paragraphs f 1 xii, f2 ix and f3 v. The

design certification is not site specific but includes generic site parameters called for in the

design, proposed resolutions to safety issues identified in NUREG-0933, and a design specific

probabilistic risk assessment (PRA). Proposed ITACC requirements are included and the

technical reasons that these requirements will allow for the proper testing of their system.9

This information is separated into 2 Tiers. Tier 1 includes definitions, general provisions,

design descriptions, and the ITAAC. Tier 2 includes information required by 10 CFR 52.47, with

the exception of technical specifications and conceptual design information, FSAR, supporting

information on ITAAC test, and items needed in the COL. A sample Table of Contents for Tier 2

from the AP 1000 is seen in Table 3-1 Tier 2 Chapter List.1

Table 3-1. Chapter List
Chapter Number Chapter Title
Chapter 1 Introduction and General Plant Description of Plant
Chapter 2 Site Characteristics
Chapter 3 Design of Structures, Components, Equipment and
Systems









Table 3-1. Continued
Chapter Number
Chapter 4
Chapter 5
Chapter 6
Chapter 7
Chapter 8
Chapter 9
Chapter 10
Chapter 11
Chapter 12
Chapter 13
Chapter 14
Chapter 15
Chapter 16
Chapter 17
Chapter 18
Chapter 19


Chapter Title
Reactor
Reactor Coolant System and Connected Systems
Engineered Safety Features
Instrumentation and Control Systems
Electric Power
Auxiliary Systems
Steam and Power Conversion System
Radioactive Waste Management
Radiation Protection
Conduct of Operations
Initial Test Program
Accident and Analysis
Technical Specifications
Quality Assurance
Human Factors Engineering
Probabilistic Risk Assessment


Tier 1 information comes from Tier 2 information and all Tier 1 information is fulfilled

with unless a plant specific exemption is granted by the NRC. Differences in design descriptions

found in Tier 1 and the Tier 2 Information, the design descriptions take precedence but the

information in Tier 2 is not the only way to comply with the information in Tier 1.11


If there is a certain part of the design that the applicant does not seek certification for a

conceptual design is needed. The applicant should be sure they provided enough information for

the Commission to be able to review and accept the proposed design and if there are any

questions the applicant must submit the information that the Commission request. The design

must also include information on the performance of the safety features of the design. The utility

applying for the COL provides all the missing information related to site specific information.9









CHAPTER 4
REACTOR DESIGNS

AP 1000

The AP 1000 is Westinghouse's version of a generation 3+ design. The design is for a two-

loop reactor with an 1117 MWe output. The total reactor power is 3,400 MWt. It contains 157

fuel assemblies in a 17 X 17 array that is 14 ft tall. The total fuel weight is 211,588 lb and has a

5.72 kW/ft average linear power density. The average core burnup is approximately 60,000

MWD/MTU.11

The AP 1000 uses passive safety systems so that in an accident scenario no on-site or off-

site power is needed. There are 5 systems that are considered nuclear safety systems in the DCD

tier 1 for the AP 1000. The steam generator system transports the main steam produced to the

turbine and secondary components. It is considered a safety system because it must isolate the

primary system from the secondary system in a design basis accident. The valves associated with

containment isolation in the steam generator fail closed so that primary water does not leave

containment." The second safety system is the passive core cooling system. The passive core

cooling performs safety injection and core makeup and passive residual heat removal. There are

2 nozzles on the reactor vessel dedicated to safety injection. The makeup water can be found in

core makeup tanks (CMTs), accumulators, or in-containment refueling water storage tanks

(IRWST). High pressure safety injection would be handled by CMTs which are located above

the reactor coolant system (RCS) loop piping. If a signal is given that the water level or pressure

in the pressurizer is too low, the reactor coolant pumps trip and the CMT are emptied by gravity

into the reactor vessel. The CMTs contain borated water and are in parallel trains on each leg of

the RCS. The accumulators is actuated after pressure deferential between the accumulators and

the RCS drops, the check valves isforced open and water refills the downcomer and the lower











plenum to help the CMTs in re-flooding the core. After the pressure is low enough, water from

the IRWST which is located above the RCS loop piping flows into the core. The IRWST is kept


at atmospheric pressure.12

The passive heat removal system is located in containment and the flow for the passive

system is generated by natural circulation. The heat exchanger is located in the IRWST and has

100 percent capacity. Water volume is sufficient to delay boiling in the IRWST for 2 hours.


Once boiling starts the steam produced condenses on the steel containment vessel wall, and with

the help of special gutters, flows back into the IRWST. Another passive system is the passive

containment cooling system which again is driven by gravity. As seen in Figure 4-1 Passive

Containment Cooling air enters the containment vessel through the top of the sides and flows

down an air baffle. Air then rises up the steal walls of the inner containment and leaves through

the roof of containment.12


Natural convection
air discharge \
i-CCS aa-.-T. dr,r, .
eli ank
Water film evaporation

Outside cooling air intake -

? t
Intemrd condensation
and
Steel containment vessel -- ne ,rel iia u lalloll

Air baffle 1'







Figure 4-1. Passive Containment Cooling (Source:
http://www.westinghousenuclear.com/docs/AP 1000_brochure.pdf. Last accessed
June 21, 2007).

The tank on top of the containment vessel releases water to help cool the steal containment

vessel if more cooling is needed because of high containment pressure. There is enough water in

these tanks for 3 days. After three days the tanks must be refilled, but calculations show that if









they are not refilled, peak pressure will only reach 90% of design pressure. The main control

room emergency habitability system is the last of the 5 systems mentioned in tier one. The

system is made of compressed air tanks to provide a livable environment for 11 personnel. The

compressed air tanks are connected to the control room by a main and alternate delivery line so

that no single failure can occur.12

Since all passive systems are being used there is no need for safety related diesel

generators. The AP 1000 like most new designs also reduces the number of pumps and feet of

cable needed to construct and operate the plant.13 The spent fuel pool allows for enough storage

for 10 years of operation plus one full core. This is important because we currently do not have a

national spent fuel storage plan.8

Many of the active safety systems used in other plants though are retained and are no

longer considered safety related.

Reliability

The AP 1000 is expected to achieve over 93% capacity factor.12 With a high capacity

factor and defense in depth for the AP 1000 system the core damage frequency (CDF) is 4.2 x

10-7 per year and a large release frequency is 3.7 x 10-8 per year.13 While the large release

frequency may be low, workers do receive radiation. Table 4-1 Doses to workers shows the

doses the workers receive at the site."

Table 4-1. Dose to Workers AP 1000
Category Percent of Total Estimated Annual (man rem)
Reactor operations and surveillance 20.6% 13.8
Routine inspection and maintenance 18% 12.1
Inservice inspection 24.7% 16.6
Special maintenance 22.4% 15
Waste processing 7.7% 5.2
Refueling 6.6% 4.4
Total 100% 67.1









Technology Maturity

The AP 1000 is the only generation 3+ reactor to receive a design certification from the

NRC, though there has been an amendment submitted for review. The fuel used in the AP 1000

is currently being used in 4 other U.S. Power plants. The reactor vessel and internals are used in

Doel 4 and Tihange 3. The control rod drive motors are the same in use throughout the world.

The large model F steam generators are currently used in 4 other sites like Waterford. The

reactor coolant pump (RCP) is based on a canned motor pump which has been used by the U.S.

Navy and fossil boilers. The pressurizer is used in 70 other plants worldwide.12

Economics

In order to lower the construction costs, the buildings needing seismic stability are reduced

but there is still working space due to the reduction in active equipment. Modular construction

will also be used to decrease construction time.14 The AP 1000 plant uses over 270 modules to

help reduce construction time, reduce man power needed, reduce site congestion, and increase

reliability in construction. It is expected that the first two reactors will share the cost for the first

of a kind (FOAK) engineering. With the first of a kind cost included, the first 2 units are

estimated by the vendor to cost around $1488/KWe. The nth plant which is the plant that comes

after the first 2 which includes the lessons learned from construction is expected to cost about

$1134/kWe. Westinghouse also estimates that it cost between 3 to 3.5 cents/kWh to operate a

duel unit plant. These cost estimates are based on calculations made for the AP 600 plant which

is a smaller version of the AP 1000.15 The major factor in cost though is the time delay from

order to operation. The construction schedule calls for a 5-year delay.

The delay is factored into 3 sections. Based on the vendor estimation, the construction site

preparation time is 18 months. The construction time is 36 months. The last 6 months includes

the start up and testing of the reactor.12









Supply Chain

The supply chain of the AP 1000 is indicative of the industry as a whole. There are long

lead times for the large items such as the reactor pressure vessels and other large forgings.

Unlike some of the other plant designs the AP 1000 and its predecessor the AP 600 have not

been built. The supply chain is based on smaller firms providing some of the module

construction. Many of the firms that used to be certified to produce the nuclear grade equipment

have not kept the certification. Another industry-wide problem is the personnel supply chain.

Nuclear power plant companies have a hard time finding employees to fill the required positions.

ABWR

The ABWR is a generation 3 BWR design by GE, Hitachi, and Toshiba. The design is a

1325 MWe output reactor with 4 main steam lines with flow restrictors and isolation valves on

each line. The total reactor power is 3,926 MWt. It contains 872 fuel assemblies in a 10 X 10

array that are 12 ft inches tall. The fuel has a 4.28 kW/ft average linear power density. Control

for the reactor is provided by 205 control blades driven by fine motor control rod drives that

drive the blades in from the bottom of the core. The average core burnup is about 60,000

MWD/MTU.16

The ABWR is the GE-Hitachi-Toshiba first step into creating a new design for generation

3 reactors. The ABWR has safety grade Emergency Diesel Generators (EDGs) which provide

power to the systems safety features in an accident scenario. The system is equipped with 3

diesels that need to be serviced and maintained along with many other safety equipment

currently found in other BWR plants. The ABWR includes design features to help improving the

safety and simplicity of the system. For instance, the external recirculation pump in ABWR is

replaced by sealed internal pumps that improves the operational flexibility and safety









performance of the system.7 The reactor has 10 of these internal pumps located inside the

reactor pressure vessel (RPV) as it is depicted in Figure 4-2. ABWR Reactor Pressure Vessel.


















Figure 4-2. ABWR Reactor Pressure Vessel (Source:
http://www.gepower.com/prod_serv/products/nuclearenergy/en/downloads/gea 457
6e_abwr.pdf Last accessed June 5, 2007).

One of the major features in the ABWR is that there are no large pipe nozzles or external

recirculation loops bellow the top of the core. This feature helps keep the core covered in the

case of a loss of coolant accident (LOCA). If a LOCA were to occur two motorized high pressure

core flooders (HPCF) in each division of the ECCS network start filling the reactor core. The

suction for the HPCF comes from the condensate storage tank (CST) or the suppression pool

(SP) as an alternate. The reactor core isolation Cooling (RCIC) primary purpose is to provide

reactor cooling when the core is isolated from the turbine as a main heat sink. The RCIC operates

using steam from the reactor to turn turbine pumps. The RCIC has the capacity to provide

cooling when the vessel is in hot standby, during a plant shutdown with loss of feedwater before

alternate systems come online, and during a loss of AC power. The suction for the system comes

from the CST or the SP with the drain from the steam turbine going to the main condenser. After

a short delay the automatic depressurization system (ADS) is initiated and the reactor goes from










1040 psia to pressures where the residual heat removal system can be used. The residual heat

removal system has 6 major functions: low pressure flooder, suppression pool cooling, reactor

shutdown cooling, primary containment vessel spray cooling, supplemental fuel pool cooling,

and AC independent water addition with water from the fire protection system.17

During an accident or refueling operations a standby gas treatment system is used to

remove radioactivity from the air and release through the plant stacks. During an accident there

is an atmospheric control system that will keep an inert atmosphere in the containment building.

Due to the Three Mile Island accident and lessons learned a flammable control system which

will not let the hydrogen ignite in containment is installed. In the event that the control rods

cannot shut down the reactor a standby liquid control system (SLCS) pumps boron into the

reactor. Many of these systems need AC power that is why the EDGs are safety related.17 An

order of magnitude improvement in safety including the core-recovery during LOCA is achieved

by the ABWR. The low frequency of the ABWR core damage as compared with other US BWR

and PWR plants as well as with the standardized Japanese BWR (BWR-5) is shown in Figure 4-

3 Comparison of PSA Results. 19,20

1E-3


-IAEA Safety Goal
1E-5




1E-8
Zion' Sequoyah* Surry* Peach Grand Gulf* BWRRS ABWR**
(PWR) (PWR) (PWR) Bottom* (BWR6)
(BWR4)
Figure 4-3. Comparison of PSA Results (Source: CSNI Workshop on PSA Applications and
Limitations, NEA/CSNI/R (91) 2, Santa Fe, New Mexico, September 4-6, 1990.
Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plant, NUREG
1150, Vol. 3, 1989 (released 1991)).









Besides safety system the ABWR employs many technological advances since the last

BWRs. It uses a fully digital control system for accurate plant monitoring and uses better fuel to

reduce radioactive waste.18 The ABWR keeps the radioactive waste volume to around 735 ft3/yr,

which is a significant reduction from the average waste volume for the previous generation of

BWR plants. It has enough storage in the spent fuel pool for 10 years and one core.17

Reliability

The ABWR is expected to achieve 95% capacity factor.17 With a high capacity factor and

active defense in depth for the ABWR system the core damage frequency (CDF) is 1.6 x 10-7 per

year and a large release frequency is 1 x 10-8 per year.17 Along with better monitoring of the

reactor system and a low LRF the annual man-rem to workers in ABWR system is to 98.9 man-

rem/ year as can be seen in Table 4-2. Dose to Workers ABWR.16

Table 4-2. Dose to Workers ABWR
Task Location Percent of Estimated Annual
Total (man rem)
Drywell 46.7% 46.2
Reactor 14.7% 14.5
Building
Radwaste 10.6% 10.5
Building
Turbine 11.8% 11.7
Building
Work at Power 16.2% 16
Total 100% 98.9

Technology Maturity

The ABWR is the only design evaluated in this study that has operating experience. There

are 4 such units in Japan operating since 1996, 1997, 2004, and March 2006, an additional 3

under construction, 2 in Taiwan and 1 in Japan. Eleven more are planned for construction, 9 in

Japan and 2 in US. The ABWR technology is based on 50 years of experience using BWR power









plants. The ABWR is also a certified design which would allow for construction to begin
18
sooner.1

Economics

The ABWR has a proven record of reducing capital, operational, and maintenance cost. Its

construction time is approximately 39 months which is proven in Japan. It has a smaller reactor

building to decrease construction time. The ABWR also uses modular design which allows for

high quality and reliability in the construction. The ABWR also has no FOAK cost associated

with the plant design since the ABWR is fully engineered and fully constructed in other

countries. Based on the experience in building ABWR systems in Japan, a more realistic cost

estimate for this system is possible. In a recent DOE document the estimated overnight cost of

the ABWR is cited between $1400-1600/kWe.15

Supply Chain

GE with the help of Hitachi and Toshiba is currently building two ABWR plants in

Taiwan. Hitachi is building another ABWR plant in Japan at predictable construction schedule

and cost. GE is building the Taiwan plants using ASME codes and meeting all US regulatory

requirements.15

ESBWR

The ESBR is the newest reactor design from GE. It is based on BWR technology but has

not been certified by the NRC. GE, with this design, is simplifying the plant to remove

complexity. The design is a 1535 MWe output reactor with 4 main steam lines with flow

restrictors and isolation valves on each line. The total reactor power is 4500 MWt. It contains

1132 fuel assemblies in a 10 X 10 array that is 10 ft tall. The fuel has a 4.6 kW/ft average linear

power density. Control is provided by fine motor control drives that drive 260 control blades into

the core to control power. The average core burnup is around 60,000 MWD/MTU.21'22









The safety systems in the ESBWR are passive and therefore do not need AC power in an

accident. The isolation condenser (IC) system is located above the reactor removing the need for

pumps and allowing the system to be gravity driven and is attached to the passive containment

cooling (PCC) pool. Water travels from the isolation condenser to the reactor core and steam

goes back to the condenser. A vent is provided in the system to allow non-condensable gases to

be vented and the water in the system to stay in the IS and PCC. In order to depressurize the

system the ADS actuates and opens the relief valves during a LOCA. Due to the ability of the

ADS system to depressurize the core quickly, the ECCS is now driven by gravity. In order to

ensure there is no containment failure a passive containment cooling system is employed. There

are a total of 6 loops to condense steam and cool the containment. Each loop is rated for 11 MWt

and without makeup to the PCC pool; they can cool containment for 72 hours. The SLCS system

in the ESBWR has two methods of delivery. Pressurized accumulators which require no AC

power inject borated water into the core via squib valves, or a non safety related nitrogen

pressure charging system is used. In the event of an accident which places the area outside of the

control room uninhabitable the emergency breathing air system provides enough air for 72 hours

after the accident. The air is in compressed air tanks that also pressurize the control room to

minimize in-leakage from the outside air.21

The ESBWR however is mainly known for the fact that it uses natural circulation. GE

achieves this by partitioning a chimney directing the steam flow above the core. The downcomer

is also enlarged to reduce flow resistance and the entrance of the feedwater is placed high on the

reactor vessel to provide more driving head. A basic drawing is depicted in Figure 4-4. Natural

Circulation.23












MAIN STU,,'1
STEAM FEEowATER
SEPARATO R
ANNULUS

CHIMNEY
*: rt.;-


CORE


U


r SATURATEDWATER
SUBCOLED WATER
SSATURATED STEAM


Figure 4-4. Natural Circulation (Source:
http://www.gepower.com/prod_serv/products/nuclear energy/en/downloads/natural_c
irculation_esbwr.pdf. Last accessed May 10, 2007).

There are 11 fewer systems than in previous power plants and a 25% reduction in pumps,

valves, and motors.21 Along with the reduction in systems and components the ESBWR keeps

the radioactive waste to around 735 ft3/yr. It has enough storage in the spent fuel pool for 10

years and one core. These numbers are estimated based on the experience from the ABWRs

already built.21

Reliability

The ESBWR is expected to achieve above 95% capacity factor.21 With a high capacity

factor and active defense in depth the core damage frequency (CDF) is 3 x 10-7 per year and a

large release frequency is 1 x 10-9 per year.2'24 The estimated dose to radiation workers in the

ESBWR is 60.4 man rem/yr Table 4-3 Dose to Workers ESBWR.22

Table 4-3. Dose to Workers ESBWR
Task Percent of Estimated Annual
Total (man rem)
Drywell 34.3% 20.7
Reactor Building 28.0% 16.9
Radwaste Building 4.1% 2.5
Turbine Building 18.5% 11.2
Work at Power 13.2% 8
Fuel Building 1.8% 1.1
Total 100.0% 60.4









Technology Maturity

ESBWR is considered a generation 3+ reactor design. It is currently going through the

NRC review for the Design Certification and the Safety Evaluation Report (SER) is expected to

be released was filed in 2008.25 The ESBWR is basing much of the technology it uses on

experience from the ABWR. Reactors in the past have run on natural circulation even though this

reactor has much more power the technology is proven effective by Dodewaard a 60 MWe

reactor that ran for 25 years. Also many of our current BWRs can run at about 50% power on

natural circulation. Also many of the passive safety features envisioned for the SBWR are

incorporated in the ESBWR design. Effectively no new systems are designed for the ESBWR

they are adjusted, uprated, or simplified for the ESBWR.24

Economics

The ESBWR is designed to use many of the construction techniques that are used in the

ABWR. Construction time provides a substantial cost increase, but with the use of modular

construction technology the GE estimated time of construction is about 36 months.26 The first

few plants must account for the FOAK engineering cost. Based on the GE's estimates, due to the

simplification of the design even with the FOAK the cost of building ESBWR could be less than

ABWR.15

Supply Chain

The ESBWR uses the supply chain already in use for the ABWR for most of its

components. The modules are built offsite and brought onsite for final construction. GE plans to

construct the turbine island, associated equipment and the instrumentation and controls in the

U.S. The large forgings are made overseas, depending on Japan Steel Works for the reactor

pressure vessel forgings. Due to the ties between the ABWR and ESBWR supply chains a bottle

neck may occur in the future.1









APWR

The APWR is a 1,700 MWe plant. The design calls for 4 loop PWR system with a core

power of 4,451 MWt. It uses a 17 x 17 assembly design and has 257 assemblies in the core. The

average linear power density is 5.36 kW/ ft. The core burn up is estimated to be 62,000

MWD/MTU.27 The attachment of the reactor internals is made without the use of welds to

facilitate the quality control during installation and for later inspection. The reactor vessel itself

is made out of partially low alloy steel to minimize welds that would require regular inspection.

Also forged rings are used to reduce welding by completely eliminating welds along the belt line.

Control is provided by both boric acid and 69 control rods. The reactor coolant pump is a vertical

shaft, single-stage suction diffuser with a non-contact controlled leakage system.28

The containment vessel is a cylindrical prestressed concrete containment vessel with a

hemispherical dome and a carbon steel liner. It has an inner diameter of 43 m and is 1.3 m thick.

The design pressure for the containment building is 83 psia. In case of a core damage accident

there are hydrogen igniters inside containment to prevent hydrogen explosions. Surrounding

containment is an annulus that is kept at a slight negative pressure to keep the radioactive

material from being released in an accident. In the case of a LOCA the APWR uses accumulators

to provide large-flow, low-pressure injection and safety injection pumps to provide high head

injection. The safety injection pumps take suction from the refueling water storage tanks located

in containment. Containment cooling is achieved by the containment spray system. The

containment sprays also take suction from the refueling water storage tank. Residual heat

removal (RHR) is responsible for removing the decay heat from the reactor, the system is tied to

the containment sprays. In accident scenarios power to the safety systems is provided through the

safety-related gas turbine generators.2









Along with improving the system associated with plant safety, the APWR is more

environmentally friendly and produces about 60 drums/yr of radioactive waste vs. the 100

drums/yr currently produced which is approximately 441 ft3/yr of radioactive waste. This will be

possible by using enriched B-10, a better drain recovery system, a boric acid recovery system,

and a high performance cement solidification system.28

Like the other designs the APWR uses a digital control system. The system allows for

daily load following, automatic frequency control and house load operation. Mitsubishi uses

advanced methods, including the prefabricated unit method and the steel-plated concrete method

to deliver the plants on time with the highest quality.28

Reliability

The thermal efficiency is projected to be 39% giving it the highest efficiency estimate for

the new reactors.28 The PRA demonstrates a reduction of 1/10 the prior value of the CDF. The

target CDF and LRF is lower than 1x10 5.27

Technology Maturity

The APWR was submitted for the NRC Design Certification review on January 4, 2008. It

is a generation 3 design like most of the reactors in this study.25 Although Mitsubishi has never

built a reactor in the U.S., it has built reactors in Japan. With this continued experience the

APWR has become more of a work in progress than a new reactor design.29

Economics

APWR like all other reactors has many factors controlling the cost per kWe installed.

Construction time is a major cost factor so the highest levels of control is exercised, including

utilization of the Integrated Project Schedule Control System. Construction is therefore estimated

to be 46 months. Since this is the first time an APWR will be built in the U.S. there will be

FOAK costs associated with the APWR, which is targeted to be $1500/kWe.28









Supply Chain

Since 1970, Mitsubishi has been constructing nuclear reactors, it currently plans to

continue with its rich history. Throughout its operating history not only has Mitsubishi

constructed reactors but it has also exported major components to the world. Mitsubishi currently

plans to continue making the parts in Japan and exporting them to its worldwide clients.

Mitsubishi has fabricated a total of over 323 major components and is confident that they can

meet the supply demand for new reactor construction.28

EPR

The EPR is the new reactor design by AREVA NP. The EPR is a 1600 MWe reactor that

has 4,500 MW of thermal power. The power comes from 241 fuel assemblies made from UO2;

the reactor however is capable of using MOX fuel. Each fuel assembly is 14 ft tall and arranged

in a 17 x 17 array. The core has an average linear power density of 4.75 kW/ft.30 Each assembly

has a maximum discharge burnup of greater than 70,000 MWD/MTU. Control is provided by 89

control rods and boric acid. The EPR uses a four-train safety system in which each train is

independent of the other so that a common mode failure cannot occur. Each train is capable of

performing all safety functions.30

Each train has a safety injection that provides water from the accumulators located in

containment. Medium head safety injection is provided via safety injection pumps which pump

water through the cold legs and take suction from the IRWST. Low-head water is provided by

the RHR system. Two of the trains have extra boron systems that allow for 7,000 ppm of boric

acid to ensure the unit is shut down. In order to ensure power to these systems in an accident

there are 4 Emergency Diesel Generators in 2 separate buildings. In the event of a station black

out 2 additional diesel generators provide power to the safety busbars of 2 trains. In addition to

many safety features to protect the core, if there is a core melt the EPR is designed for









mitigation. The containment building is a double wall design with each wall being 4.3 ft thick to

prevent releases. There are dedicated relief valves to mitigate a high pressure core melt even if

the pressurizer relief valves fail. The high strength pressure vessel prevents damage from

reactions between corium and water. Hydrogen explosions are mitigated by hydrogen

recombiners that will keep the hydrogen levels inside containment below 10 percent. Corium

flows from the reactor vessel in the case of a reactor vessel failure into a core catcher area where

a long term containment heat removal system cools the corium and allows it to solidify. The

control room is fully digital and is located in one of the safeguard buildings allowing for it to be

manned during most accidents.31

Reliability

The EPR operates within large margins and flexibility in order to allow for future

regulations and standards. It is expected to have a 92% capacity factor and operate for 60 years.32

The dose to workers is estimated at 40 man rem/yr due to its safety, simplicity, and health

physics programs. Major accidents should not occur with a core damage frequency of 4 x 10-7

and a large release frequency of < 10-7.30

Technology Maturity

EPR technology is based on the French N4 reactor and the German Konvoi reactor design.

There is one EPR being built in Finland and another one in France. Currently in the US the EPR

is under review for design certification which was filed at the end of 2007.25

Economics

As with most new plants long construction times cause significant cost increases for the

buyers. The preconstruction period of 15 months is needed for components with long lead times.

Once construction begins it is expected to take 48 months for commercial operation. Like other

plants modular construction is used, and experiences from Finland and France is used to stream










line the construction.30 The Unistar group of nuclear plant operators has the overnight cost of the

EPR at $2400/kWe if the FOAK cost are spread over 4 units.32

Another cost-saving measure is reducing the components in the system. Figure 4-5 shows

the percentages reduced.


COMPARISON OF EPR EQUIPMENT WITH A TYPICAL 4-LOOP UNIT






120%

z

VALVES PUMPS TANKS HXS
COMPONENT TYPES
Figure 4-5. EPR Component Reduction (Source: http://unistarnuclear.com/RightTeam.pdf Last
accessed June 29, 2007).

Supply Chain

Areva has over 4,200 employees in the U.S. working on nuclear projects. Eighty percent of

the work for an EPR order is done by U.S. employees with most of the engineering work done in

Areva's offices in North Carolina and Virginia. Areva is also the world's largest nuclear supplier

with manufacturing capability. The large components are fabricated by FANP and the other

components such as the electrical systems are fabricated in the U.S.31









CHAPTER 5
DISCUSSION

Robustness

Robustness in a design is the ability to keep a plant within a safe operating condition. New

reactor designs have incorporated features that allow operation of the plant more safely and with

more margin. All of the control rooms are fully automated and computers disseminate

information much more efficiently to the operators allowing for a less stressful environment. But

no matter how easy a plant is to operate; there is no substitute for design margin. In the future,

new regulations, unforeseen problems, design modifications, and improvements may arise and

having extra margin allows for accommodation. Many factors can be categorized as margin but

this study will look at thermodynamic margins in the power plants, specifically linear heat

generation. The IAEA published a document which lists items that should be limited and

controlled to keep fuel failures to an acceptably low level. These items included linear heat

generation rate, critical power ratios, minimum departure from nucleate boiling, and peak fuel

and temperature.34 The study uses the linear heat generation rate because the amount of heat

produced in a specific area will determine the temperatures in the area that will affect all the

other parameters. Table 5-1 Linear Heat Generation Rate shows the linear heat generation rates

for the studied reactors.

Table 5-1. Linear Heat Generation Rate
Reactor Design Linear Heat Leneration(kW/ft)
AP 1000 5.72
ABWR 4.28
ESBWR 4.6
APWR 5.36
EPR 4.75









The ABWR has the lowest LHGR. Among the Generation 3+ reactors, the ESBWR has the

lowest LHGR and the AP 1000 has the highest. The lower the LHGR the more margins available

during an accident.

Safety

Each design has features that make it robust and safe. The safety systems for all the

reactors are carefully designed and perform their needed functions in accident scenarios. The

main choice in the safety category is whether to go with an all-passive safety system design or an

active safety system design that will include another form of electricity in an emergency. Both

the AP 1000 and the ESBWR are passive safety systems that use gravity and pressure

differentials in emergencies. They have no need for secondary power sources in emergencies

other than for instrumentation and can cool the core while off-site power is being restored. They

both also use compressed air for the control room during postulated accidents. Emergency diesel

generators (EDGs) are therefore no longer safety-grade. Systems that were in previous PWRs

and BWRs are no longer safety-grade but like the diesels may still be present. The ABWR and

the EPR both use the proven technology and use safety-grade EDGs. They must be maintained

and have surveillance performed like existing diesels at current nuclear reactor sites. EDGs can

have many problems, in 1999 a study was prepared for the NRC looking for a common cause of

failure in EDGs while the study found no firm conclusions Figure 5-1 EDG Failures shows that

from 1980 to 1995 there were 131 events that caused the diesel to fail the testing.35










25

20
00
( 15

U
C 10



- M
a) M


ti


* Series
* Series


L... -- L C uo o
o 30 0 O 00 A
LU C w 3 i _W o u C-
Ln

Figure 5-1. EDG Failures

In the Figure 5-1 series means failed to run and series means failed to start. As depicted

in the chart emergency diesel generators fail for many different reasons and must be maintained

properly. Both the ABWR and EPR have multiple diesels therefore maintenance activities are

factored by the number of diesels and so are chances of human performance errors.35 In the U.S.-

APWR, Mitsubishi has decided against using emergency diesel generators and instead electing to

go with gas turbine generators. The GTG though have a slower start time of about 40 seconds but

Mitsubishi feels the 10-3 failures/demand is better than the 10-2 failures/demand of the EDG.28

Defense in depth is a key word in nuclear power however the EPR has 4 independent trains

that each has 100% capacity to cool the core in an accident. Though each reactor uses the

defense in depth principle, the EPR has the most systems due to the 4 redundant trains and the

preparations the EPR has in case of a core melt such as the core catcher. The EPR has the most

defense in depth mechanisms than any other reactor design. The 4 redundant systems not only

have separate wiring and piping to prevent common cause failures between trains, but the trains

are physically separated by buildings. Many incidents occur where a component on one train can

be confused with the same component on another, building separation of the trains allow for

human performance techniques to be more apparent since the trains are not located in the same









work space. This also allows for a catastrophe to occur on one side of the plant and have minimal

or no impact on the trains on the opposite side. The AP 1000 and the ESBWR are passive

systems during accident scenarios. Passive systems allow for operators to respond to the accident

rather than worrying if their equipment works due to lack of power. When running on the diesel

engines, operators must check the logic and satisfactory loading of plant equipment to supply

water to the core. If not loaded properly the diesel may fail or a component might not provide

proper core cooling. The APWR has an advantage over the ABWR since it has no EDG and the

failure/ demand rate of a steam turbine generator is lower than an EDG. While there is no U.S.

experience with safety-grade gas turbine generators Mitsubishi has successfully used them in

Japan.

Another safety consideration that should be taken into account is core damage frequency

(CDF) and large release frequency (LRF). These new reactor designs are a large step forward in

safety and therefore have lower CDF's and LRF's. As can be seen in Table 5-2 CDF and LRF

the CDF and LRF values are lower than those of today's reactors.

Table 5-2. CDF and LRF
Reactor CDF LRF
Design
ABWR 1.60E-07 1E-08
ESBWR 3.2E-08 1E-09
APWR Target <1.OE-05 Target <1.OE-05
EPR 4.0E-07 <1E-07
AP 1000 4.2E-07 3.7E-08

The ESBWR has the lowest CDF and LRF followed by ABWR. No value is given for the

core damage frequency and large release frequency for the APWR design. The values in Table 5-

2 indicate the target goals for the CDF and LRF.









Reliability

All new reactors featured are expected to have good reliability. The capacity factors for all

the reactors are above 90%, the national average for the current reactor fleets best reactors. As

expected the new technology has a lower CDF and LRF than current reactors due to enhanced

safety features in the power plants. The NRC requires the CDF to be at 10-4 or lower and all

these reactors meet that criterion.

With all the systems, new generation reactors are more reliable and less likely to be

damaged. Capacity factors are a good measure on the reliability because they show how plant

equipment failures are not the cause of unit downpowers or shutdowns. Table 5-3 Capacity

Factor shows the capacity factor data for the reactors in the study.

Table 5-3. Capacity Factor
Reactor Capacity
Design Factor
AP 1000 93%
ABWR 95%
ESBWR 95%
APWR 92%
EPR 92%

The highest capacity factors are for the ABWR and ESBWR. For only the ABWR system

the estimated capacity factor is based on operational data. The EPR and APWR have the lowest

estimated capacity factors. The capacity factors have a direct correlation to money because the

longer a plant is down the more cost is incurred by the utility to buy other forms of fuel to make

up the MWs needed for their customers.

Technology Maturity

Each reactor is in a different stage of development. Some of the reactors are already built

overseas and have a rich history of operation and others have never been built before.









The ABWR is the most mature design since it has been built and new units are on order.

The ABWR design is also approved by the NRC. The EPR is the next most mature reactor since

it is currently under construction in Finland. Since the design is completed and is under

construction the reactor must undergo changes only to meet U.S. regulatory requirements and

constraints. The AP 1000 is the next most mature design because it is licensed by the NRC;

however, a revision is submitted to the NRC. In order for the reactor design to be approved, most

of the engineering design is completed and therefore the AP 1000 is more mature than the other

reactors which have not yet constructed. The ESBWR is based on ABWR and SBWR

technology and many of the systems are tested and accepted. The U.S.-APWR is now in the

preliminary stages of the application process and none are being constructed currently. The

APWR however is currently under review and is selected for construction by 2015-2016 in

Japan.

Economics

Costs associated with nuclear power plants are mainly initial capital costs and interest on

that capital cost. However Operations and maintenance cost can and will affect the plant and its

revenue. The NRC also estimates it costs $1000/ man rem and each reactor tries to achieve lower

man rem/yr values to save employees from doses and save money.36 The lowest is the EPR

followed by the ESBWR, AP 1000, and finally the ABWR. There is no information on doses to

workers for the APWR. One surprising find is that the ESBWR has a lower dose to the worker

than the AP 1000. As the trend in industry shows BWRs have more dose to workers.

Depreciation rates also play a large role in the final cost of nuclear power. Construction

times for the new reactors are concerns since most states do not allow for companies to bill

consumers until the plants have been built. Each company predicts that it will take 4 to 5 years to

build one of these reactors which are in line with NRC predictions for how long it takes for









reactor construction. The estimated overnight costs for the reactors as reported earlier are

summarized in Table 5-4.

Table 5-4. Reactor Overnight Cost
Reactor Design Overnight Cost ($/kWe)
AP 1000 1488
ABWR 1400-1600
ESBWR <1400-1600
APWR 1500
EPR 2400

As is seen in Table 5-4 Reactor Overnight Cost many of the reactor vendors feel the cost is

around $1400/kWe so the cost is about equal for the 4 reactors. However, based on more recent

estimates these numbers are low by a factor of two or more. The EPR estimate is based on the

fixed price contract for the unit that is under construction in Finland.

After the reactors have been built they must make profit in order to pay off the initial

investment. Operations and maintenance costs will play a role in reducing the company's profit.

Table 5-5 Operations and Maintenance Cost shows the operations and maintenance costs for the

reactors.15,37,38

Table 5-5. Operations and Maintenance Cost
Reactor Design Cost (mils/kWe)
AP 1000 8.17
ESBWR 6.83
ABWR 6.71
APWR 9.10
EPR 7.00

Allowing for reactor construction at locations where reactors already exist lowers cost.

Existing reactors have large support staffing that may be shared with the new reactors reducing

the cost. An interesting fact is how the operations and maintenance costs for the generation 3+

reactors are higher than that for the ABWR.









Supply Chain

Each reactor vendor claims to be able to deliver the components and parts needed for the

timely completion of construction. Each though is plagued by large lead times for the large

forgings and component manufacturing. Another problem is that the U.S. no longer has the

manufacturing capabilities for large components so many have to be imported from overseas.

Another problem is lack of sufficient number of U.S. skilled craftsmen to construct these

reactors. The nuclear industry has an ageing workforce and recent grads are not filling the posts

being vacated by retirees.38

Hitachi has proven that it can construct the ABWR in less than 4 years and that its vendors

have the capability of delivering the components built to U.S. standards. ABWRs are currently

being ordered around the world and there is confidence in the supply chain. Since the same

vendors are used for the ESBWR as the ABWR; there is confidence in the supply chain for the

ESBWR. The EPR design from AREVA is currently being constructed in Finland and the supply

chain is proven. Next is the U.S.-APWR which is developed by Mitsubishi Heavy Industries.

Mitsubishi has built reactors in Japan as recently as 1997 and is capable of producing the

components overseas. Mitsubishi has stated that it will import components from overseas and

therefore the supply chain is verified. There is a lower confidence in the supply chain for AP

1000 since there is no history of building any system that utilizes similar components.

Licenseability

Regulatory issues are the cause of many nuclear units to be canceled or left without being

completed. Many financial institutions are not providing the capital for the investment knowing

that the investment may fail because the design is not at par with regulatory standards. The NRC

has been diligently working on the design certifications and reviews of many of the systems and

on fixing the open questions with those already certified.25









Table 5-6. Regulatory Status
Reactor Design Regulatory Status Projected Date
AP 1000 Certified/amended under review 3/2010
ABWR Certified Certified
ESBWR in certification process No Date
APWR in certification process No Date
EPR in certification process 5/2011

The ABWR has the greatest advantage in the licenseability since it is certified and

constructed. The AP 1000 is the next reactor. Even though the AP 1000 design is certified, it is

now being amended and re-reviewed as seen in table 9. Next is the EPR with a projected year of

2011. The ESBWR and APWR do not have a projected date. GE is currently resubmitting an

updated schedule for submission of revision 5 of the DCD. The APWR is currently waiting for

an acceptance of docketing and schedule for review from the NRC. The order of design status

depends on the NRC projected dates.

Security

Security at a reactor site involves the reduction of threats from outside sources. There are

many threats that can have a negative impact on control and safety at a nuclear site. September

11 really opened the eyes of the public to the threat of airplanes crashing into buildings. The

threat may come from terrorists or even a malfunction in the system, never the less it is still a

danger however improbable. The NRC has recently ruled that the new reactor design has to

withstand a large plane crash and be able to protect the health and safety of the public in such an

event. Another concern is a major fire that may engulf the spent fuel pit. Like many industrial

sites there are many hazards that can affect the security of the equipment.

The EPR has a distinct advantage over the other reactors because it has two containment

buildings that are 1.3 meters thick each. Each one of the four trains is located in different

buildings or partitions and on different sides of the plant, so if one side is hit the other trains on









the opposite sides are available. The ESBWR and ABWR are equal because they both have a

primary and secondary containment. The first containment building is the immediate area

surrounding the reactor including some of the pools needed for safety cooling. The secondary

containment is the reactor building itself which envelops the primary containment. The next

reactor is the APWR. The APWR has a containment vessel very similar to current reactor

designs. The vessel uses buttresses to tie the horizontal tension wires and vertical tension wires

fixed in the tendon gallery at the base. The building is made of pre-stressed concrete. Lastly, the

AP 1000's primary containment building is made of steel not a concrete structure. The shield

building surrounding containment has holes along the sides to allow for passive containment

cooling. These holes are a point of concern, because even though improbable there exists the

possibility of fuel getting into these holes and catching fire.

Risk and Uncertainty

There are many risks and uncertainties associated with building a nuclear reactor in the

U.S. In the past regulatory issues caused the cancelation of many nuclear plant orders and the

stagnation of the nuclear industry. When building new reactors many investors feel the

investment risks are large and are hesitant to provide the capital without assurances. All first of a

kind designs have uncertainties associated with them. All FOAK units have unforeseen issues

that may arise during operation and or construction that may cause the reactors not to operate or

operate with substantial increases in cost. The large amounts of capital investment requires that

the plants be built within the allotted time, if not interest rates will cause another downfall of the

industry. Regulatory bodies have also not seen these FOAK reactors operate and heavy scrutiny

when examining the designs is used. Regulatory delays also cause cost overruns and lack of

confidence in the NRC's ability to provide timely review of the reactors. The risks and

uncertainties are caused by different reasons but outside factors causing catastrophic events for









the industry and financial institutions are possible. One uncertainty worthy of mention has to do

with the security rule the NRC approved requiring the containment building to be able to

withstand an airplane crash. The AP 1000 which is an approved design will have to relook at the

containment structure and determine what the corrective actions are so that the reactor can be

built. Issues like these will cause problems during COL applications and construction.

Site Specific Issues

Reactor type selection has many factors as described above that can go into the decision

making process. Another topic that must be considered is site location, and logistics. Utility

expertise is also a factor. A utility which operates only PWRs may not want to select a BWR

because all the engineering analysis and training techniques are based on PWR information.

Another issue along the same lines is if you have a PWR located at the selected site selecting a

PWR again for the same reasons is preferable. Another choice a utility must look at is MWe

needed. Each reactor has a different power rating as shown in Table 5-7 Electric Power.

Table 5-7. Electric Power
Reactor Design Power (MWe)
AP 1000 1154
ABWR 1325
ESBWR 1535
APWR 1700
EPR 1600

If a utility needs less power in an area due to lower demand it would not necessarily want to buy

an APWR or EPR.

Another factor that must be considered is the actual site location. Nuclear power plants

require large components and the components are transported to the site. One particular

component is the reactor pressure vessel. The largest vessel is that of the ESBWR. The fuel

height was shortened creating a need for more assemblies. The chimney is designed to help the









natural circulation, creating a need for a larger vessel. On the other hand the AP 1000 has the

smallest reactor pressure vessel making it easier to transport. Fuel for the emergency equipment

is another transportation requirement. Diesel fuel is needed for the emergency diesels and natural

gas for the APWR's gas turbines. A pipeline or trucking capabilities are a requirement so that the

fuel source can be used. Is locating the facility near an ocean or major water source a viability

allowing for the use of barge transportation?

Nuclear plants also require large amounts of water for cooling. A question that must be

analyzed is where will this water be supplied from a lake or the ocean?









CHAPTER 6
CONCLUSION

This study reviews and compares the new reactor designs in reliability, technology

maturity, cost, supply chains, licenseability, O&M cost, safety, and security. The study also

brings out risks, uncertainties, and site specific issues. The new reactors studied are the AP 1000,

ABWR, ESBWR, APWR, and the EPR. The study also describes new regulations in place for

licensing reactors. These regulations are in place to stream line the reactor operation while

providing adequate review processes.

The AP 1000 is the Westinghouse version of a generation 3+ reactor. It currently has

resubmitted the design review for an amendment. It is a passive system and therefore needs no

power to mitigate accidents. This is accomplished by the use of accumulators and gravity.

The ABWR is a generation 3 reactor which uses active safety features during accident

scenarios. The ABWR seems to be the most deployable reactor design mainly due to technology

maturity, cost, and supply chains. All of these factors are demonstrated by the construction of

ABWRs in Japan and Taiwan. The ABWR is a proven technology and has been certified by the

NRC.

The ESBWR uses natural circulation and has no pumps to remove heat from the core. The

ESBWR also uses passive safety features eliminating the need for safety grade diesels to cool the

core during an accident. The main drawback in ESBWR is the lack of technology maturity.

APWR is an active design system that is currently under the NRC design review. It has

annulus surrounding containment that is kept at a slight negative pressure to keep the radioactive

material from being released in an accident. In accident scenarios power to the safety systems

can be provided through the safety-related gas turbine generators.









The EPR is also under rge design review by the NRC. It is a 4 loop PWR. The main

characterizing features of this system are the containment building design and the 4 trains of

safety features.

In conclusion it is quite evident that all new reactor systems feature improved safety and

reliability over the current generation of US nuclear power plants. A utility has many good

technology choices. Deciding which reactor is the best is dependent upon many factors; one

factor that is missing is the plant operational data. For a truly fair comparison operational data

must be used and the data is not available. The only reactor with operating experience is the

ABWR. All other reactor systems are in the early stage of engineering design with a long way

before the operational data will become available. Based on what we know at this point, it is not

possible to make a definitive conclusion on the overall relative merits of these reactor systems.









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BIOGRAPHICAL SKETCH

Alejandro Chomat was born on April 24, 1985, in Miami, Fl. He went to high school at

Christopher Columbus High School which prepared him for a college career at the University of

Florida. At UF, he majored in nuclear engineering for his undergraduate degree. Alex is a

member of the American Nuclear Society and ANX the American Nuclear Society's honor society.

In college, he worked as a lab assistant in the neutron activation analysis lab. During two summers,

he interned for FPL at Turkey Point Nuclear Station. Throughout his internships at FPL, he was able

to work in operations, maintenance, and engineering.

Alejandro Chomat is currently employed by FPL at Turkey Point Nuclear Station as a nuclear

systems operator. He plans to continue working at FPL after the completion of his master's.





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NEW REACTOR DESIGN AND REGULATION By ALEJANDRO CHOMAT A THESIS PRESENTED TO THE GRADUATE SCHOOL OF THE UNIVERSITY OF FLOR IDA IN PARTIAL FULFILLMENT OF THE REQUIREMENTS FOR THE DEGREE OF MASTER OF ENGINEERING UNIVERSITY OF FLORIDA 2008 1

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2008 Alejandro Chomat 2

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To my family 3

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ACKNOWLEDGMENTS I would like to take this opportuni ty to thank all my advisors, family, and friends for all of their support. 4

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TABLE OF CONTENTS Page ACKNOWLEDGMENTS ............................................................................................................... 4LIST OF TABLES ...........................................................................................................................7LIST OF FIGURES .........................................................................................................................8LIST OF ABBREVIATIONS ......................................................................................................... .9ABSTRACT ...................................................................................................................... .............12 CHAPTER 1 BACKGROUND ....................................................................................................................142 SCOPE OF STUDY .............................................................................................................. .213 NEW REACTOR LICENSING PROCESS ...........................................................................22Early Site Permit ............................................................................................................. ........22Combinded Operating License ...............................................................................................23Design Certification .......................................................................................................... ......244 REACTOR DESIGNS ............................................................................................................2 7AP 1000 ..................................................................................................................................27Reliability ................................................................................................................... .....29Technology Maturity .......................................................................................................30Economics ..................................................................................................................... ..30Supply Chain ...................................................................................................................31ABWR ....................................................................................................................................31Reliability ................................................................................................................... .....34Technology Maturity .......................................................................................................34Economics ..................................................................................................................... ..35Supply Chain ...................................................................................................................35ESBWR ...................................................................................................................................35Reliability ................................................................................................................... .....37Technology Maturity .......................................................................................................38Economics ..................................................................................................................... ..38Supply Chain ...................................................................................................................38APWR .......................................................................................................................... ...........39Reliability ................................................................................................................... .....40Technology Maturity .......................................................................................................40Economics ..................................................................................................................... ..40Supply Chain ...................................................................................................................41 5

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EPR ........................................................................................................................... ..............41Reliability ................................................................................................................... .....42Technology Maturity .......................................................................................................42Economics ..................................................................................................................... ..42Supply Chain ...................................................................................................................435 DISCUSSION .................................................................................................................. .......44Robustness .................................................................................................................... ..........44Safety ......................................................................................................................................45Reliability ................................................................................................................... ............48Technology Maturity ........................................................................................................... ...48Economics ..................................................................................................................... ..........49Supply Chain ..........................................................................................................................51Licenseability ................................................................................................................ ..........51Security ...................................................................................................................... .............52Risk and Uncertainty ..............................................................................................................53Site Specific Issues .................................................................................................................546 CONCLUSION .................................................................................................................. .....56LIST OF REFERENCES ...............................................................................................................58BIOGRAPHICAL SKETCH .........................................................................................................61 6

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LIST OF TABLES Table page 3-1 Chapter List ........................................................................................................................254-1 Dose to Workers AP 1000 .................................................................................................294-2 Dose to Workers ABWR ...................................................................................................344-3 Dose to Workers ESBWR ..................................................................................................375-1 Linear Heat Generation Rate .............................................................................................445-2 CDF and LRF ............................................................................................................... ......475-3 Capacity Factor ........................................................................................................... .......485-4 Reactor Overnight Cost ......................................................................................................505-5 Operations and Maintenance Cost .....................................................................................505-6 Regulatory Status ......................................................................................................... ......525-7 Electric Power ............................................................................................................ ........54 7

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LIST OF FIGURES Figure page 1-1 Simplified PWR ............................................................................................................ .....151-2 Simplified BWR............................................................................................................. ....161-3 Sustained Reliability and Productivity...............................................................................171-4 Total Cost of Generation ....................................................................................................181-5 Public Support ............................................................................................................ ........181-6 Power Uprates ............................................................................................................. .......191-7 State Policies ......................................................................................................................203-1 COL Review Process ........................................................................................................ .244-1 Passive Containment Cooling ............................................................................................284-2 ABWR Reactor Pressure Vessel ........................................................................................324-3 Comparison of PSA Results...............................................................................................334-4 Natural Circulation ....................................................................................................... ......374-5 EPR Component Reduction ...............................................................................................435-1 EDG Failures .............................................................................................................. .......46 8

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LIST OF ABBREVIATIONS ABWR Advanced Boiling Water Reactor ADS Automatic Depressurization System ALWR Advanced Light Water Reactor APWR Advanced Pressurized Water Reactor B Boron BWR Boiling Water Reactor CDF Core Damage Frequency CFR Code of Federal Regulation CMT Core Makeup Tank COL Combined Operating License COLAs Combined Operating License Application CST Condensate Storage Tank DCD Design Control Document ECCS Emergency Core Cooling System EDG Emergency Diesel Generator EPR Evolutionary Power Reactor ESBWR Economically Simplified Boiling Water Reactor FOAK First of a Kind FPL Florida Power and Light GE General Electric GTG Gas Turbine Generator HPCF High Pressure Core Flooder IC Isolation Condenser IRWST In-Containment Refue ling Water Storage Tank 9

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ITAAC Inspections, Test, Anal yses, and Acceptance Criteria LHGR Linear Heat Generation Rate LOCA Loss of Coolant Accident LRF Large Release Frequency LWR Light Water Reactor m meter MOX Mix Oxide Fuel MTU Metric Ton Uranium MWD Mega Watt Day MWe Mega Watts electric MWh Mega Watt Hour MWt Mega Watts thermal NRC Nuclear Regulatory Commission NUREG US Nuclear Regulatory Commission Regulation PCC Passive Containment Cooling PRA Probabilistic Risk Assessment PSI Pounds per Square Inch PSIA Pounds per Square Inch Absolute PSIG Pounds per Square Inch Gauge PWR Pressurized Water Reactor RCIC Reactor Core Isolation System RCP Reactor Coolant Pump RCS Reactor Coolant System RHR Residual Heat Removal RPV Reactor Pressure Vessel 10

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SER Safety Evaluation Report SLCS Standby Liquid Control System SP Suppression Pool 11

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Abstract of Thesis Presen ted to the Graduate School of the University of Florida in Partial Fulfillment of the Requirements for the Degree of Master of Engineering NEW REACTOR DESIGN AND REGULATION By Alejandro Chomat August 2008 Chair: Samim Anghaie Major: Nuclear Engineering Sciences This study will address the improved safety f eatures of advanced light water reactor (ALWR) systems that are under consideration for near term deployment in the Unites States and abroad. The safety features of ALWR systems can be divided into two major categories based on their reliance on passive or act ive design for the emergency co re cooling system. The ALWR systems evaluated in this study are actively safe ABWR, EPR and APWR, and passively safe AP 1000 and ESBWR. Each of the passive or active ALWR systems includes significantly improved safety features and larger opera tional margins. In each of the BWR and PWR categories, actively cooled ABWR and EPR feature the lowest average LHGR. The AP 1000 and the ESBWR feature passive safety systems that use gravitydriven flow for emergency core cooling with no need for secondary power sources such as emerge ncy diesel generators (E DGs) in case of loss of off-site power. The ABWR and the EPR both use the proven technology and use safety grade EDGs. They must be maintained and have survei llances performed like existing diesels at current nuclear reactor sites. There is al so a finite probability of EDGs failing to start or to run for a variety of causes including problems associated with cooling, engine, fuel oil, generator, instrumentation and control, breaker, and starti ng air. Though the lack of safety grade EDGs in passive systems is a clear advantage, the ABWR and EPR by employing multiple and redundant 12

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13 EDGs and the APWR by utilizing the GTG pr ovide for an order of magnitude safety improvement over the current genera tion of operating LWR reactors in US

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CHAPTER 1 BACKGROUND Nuclear power reactors play a huge role in our nations en ergy diversity plan. Currently, 20% of the electric needs of the U.S. are provid ed by nuclear reactors. There are two forms of nuclear reactors used for commercial power produc tion in the U.S., the pressurized water reactor (PWR) and the boiling water reactor (BWR). In total there are 104 operating commercial reactors of which 69 ar e PWRs and 35 are BWRs.1 In order for these reactors to be operating they must be approved by the Nuclear Regulatory Commission (NRC) which continues to monitor their safety during all phases of plant life from construction to decommissioning. Both PWRs and BWRs have a unique history that ha s led to the development and success of the reactor designs. The first PWR built was the Shippingport Atom ic Power Station, in Pennsylvania. The reactor began commercial operation in 1958 and ran until 1982. The Shippingport Atomic Power Station produced 60 MWe. But to this day, con cepts used in this r eactor still apply. These concepts were originally devel oped from the navy nuclear programs.2 A pressurized water reactor consist of 3 or 4 loops in the primary side which is radioactive, a secondary side wh ich creates power by turning a turb ine and a tertiary side which provides an ultimate head sink for the system. Fuel in the form of uranium fuel assemblies is controlled by both control rods and boron. As the fu el heats up it is cooled by the primary water. The water is at 2250 PSIA which does not allow it to boil. The water then goes out the hot leg into the steam generator. There are multiple hot legs out of the reactor pressure vessel one has a pressurizer which maintains the system pressure In the steam generator the water from the reactor is cooled by water coming from the second ary system. The steam generator is either a Utube steam generator like most designs or a once through steam generator that was produced by 14

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Babcock and Wilcox. The water from the secondar y system is flashed to steam which will be used to turn the turbine. The primary water then leaves the steam generator goes to the reactor coolant pumps and back into the reactor. The primar y side is mostly kept inside the containment building to avoided radioactive re lease. The steam produced in the steam generator is then used to turn the turbine which turns a shaft to create electricity in the generator. The steam is then cooled in the condenser and pumped back into the steam generator to create steam again.3 In the condenser the steam is cooled by the tertiary system. A basic PWR is seen in Figure 1-1 Basic PWR. Figure 1-1. Simplified PWR In the figure the system is cooled by a cooli ng tower but power plants also are cooled by canal systems, lakes, or oceans. Note the above sy stem is simplified; real systems are much more complex and have many more components. Thirty five of the nuclear reactors in the U. S. do not operate as described above but use the concept of a BWR in which we boil the primary wa ter to turn the turbine. The first commercial BWR was GEs Valcitos BWR which was li censed by the United States Atomic Energy 15

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Commission the predecessor to the NRC. Designs for both dual cycle and single cycle BWRs exist. A dual cycle is not preferred because it has a higher capital cost associated with it.4 A single cycle design takes the steam directly from the core and turns a turbine as seen in Figure 12 Basic BWR. Figure 1-2. Simplified BWR As in the PWR the fuel heats up and the heat is removed by the primary water. In the BWR design the primary water is allowed to boil. In order to remove the moisture left in the steam it is passed through steam dryers and separators. If wate r were to pass through th e turbine, the blades are damaged, so steam is approximately at 100 percent quality. The system here is pressurized to 1000 PSIG.5 The primary steam then spins the turbine which spins a shaft in the generator to create electricity. The steam is condensed in the condenser by th e secondary system and pumped back into the core. By using primary steam to spin the turbine and through the subsequent systems there is more contamination outside th e containment building in a BWR than a PWR. Another major difference between th e two is that the control rods in a BWR come in from the bottom of the core and in a PWR they come in from the top of the core. A BWR does not use boron for control during normal operation like a PWR it only uses boron during emergencies.3 16

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Currently, the industry has undertaken tasks to start the licensing projec ts within the U.S. Both vendors and utilities have announced and or submitted applications for various tasks. While no utility has actually ordered a reactor, which has not been d one since the 1970s, they have initiated the process to seek ear ly site permits from the NRC. Exelon, Systems Energy Resources Inc., Dominion and Southern Nuclear Operating Co mpany have all applied for early site permits with many more utilities claiming they too will follow.6 In the state of Florida, Progress Energy has selected a site in Levy County and submitted an application for building a new nuclear plant. The other major Florida electric power utility, FPL, recently has received the approval from the Public Service Commission to build a pair of ne w nuclear units at their existing site, Turkey Point. Many other utilities have ex pressed intent to apply for COLAs.7 Besides the ongoing work by th e utilities and the vendors, the current work and dependability of the current fleet is remarkab le. As can be seen in Figure 1-3 Sustained Reliability and Productivity the US nuclear power reactors have become more efficient. Gains in productivity are based on our increased expe rience operating the existing 104 power plants. Figure 1-3. Sustained Reliabi lity and Productivity (Source: http://nei.org/filefolder/thech angingclimatefornuclearenergywallstreetbriefing2007.pdf Last accessed July 21, 2007). 17

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In Figure 1-4 Cost of Generation shows how nuclear energy is becoming more cost efficient. In 2006 the fuel, operation, mainte nance and Nuclear Waste Fund fee cost was 1.7 cents per kilowatt-hr.8 Cents Figure 1-4. Total Cost of Generation Year Another important factor we must consider is the public perception of nuclear energy. The former head and co-founder of Green Peace Patrick Moore openly supports nuclear power even though he used to oppose it. The American public also realizes the need for new sources of energy and is supporting the grow th of nuclear power as can be seen in Figure 1-5 Public Support. Figure 1-5. Public Support (Source: http://nei.org/filefolder/thechangingclimatefornuclearenergywallstreetbriefing2007.pdf Last accessed July 21, 2007). 18

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Many say that since we are not building any new nuclear react ors we have not increased nuclear power generation since the last new reac tors came online in the early 1990s. But as can be seen in Figure 1-6 Power Uprates that is not true and while there are no new plants we are adding nuclear power generation by in creasing power in existing plants.8 Figure 1-6. Power Uprates (Source: http://nei.org/filefolder/thech angingclimatefornuclearenergywallstreetbriefing2007.pdf Last accessed July 21, 2007). The government is also providi ng incentives for new nuclear power generation helping to ease the worries of the utilities and the financial institutions that will fund the new construction. The first 6000MWe built will receive $18/MWh pr oduced tax break but the tax break will only help the utility after construction.8 The federal government is also providing insurance for delays caused by licensing or litigation, the first two plants will have $500 million dollar policies to cover 100% delay cost and no waiting period for cl aims, the second four plants will have a $250 million dollar policy with only 50% delay cost after 6 months delay.8 The federal government also proposed in the 2005 Energy Policy Act to pr ovide loan guarantees fo r up to 80% of project 19

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20 cost but the regulations will not be enacted until late 2007.8 Some state governments are adopting a pronuclear policy as can be s een in Figure 1-7 State Policies.8 Figure 1-7. State Policies (Source: http://nei.org/filefolder/thechangingclimatefornuclearenergywallstreetbriefing2007.pdf Last accessed July 21, 2007). These states tend to have nuclear power plants already and utilitie s are more prone to apply for licenses in the states. This also allows for a utility and invest or to worry less about negative reception from the citizens.

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CHAPTER 2 SCOPE OF STUDY This study examines the new technologies available for nuclear power plant selection and the new regulatory procedure available. It will look at 10 Code of Federal Regulations Part 52 and describe the processes the CFR covers. The study looks at the safety reliability, technology maturity, economics, supply chains licenseability, security, risk and uncertainties, and site specific issues associated with the new reactor de signs. Safety concerns with the performance of engineered and passive systems that respond to an accident. Reliability is inclusive of issues such as the capacity factor and ability to keep the plant online. Technol ogy maturity deals with the status of the engineering design and the system readiness to go to the construction phase. Economics is divided into 2 sections; the initial capital cost and operation and maintenance cost. Licenseability looks into the reactor design and its place in the current regulatory space. The risk and uncertainties and site specific issues secti ons provide information on the problems that are encountered with certa in reactor designs. 21

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CHAPTER 3 NEW REACTOR LICENSING PROCESS To simplify nuclear reactor licensing th e NRC has developed 10 Code of Federal Regulations Part 52. Before utili ties or investors would first a pply for the construction license and then later reapply to the commission for an operating license. This posed a problem because most of the cost in a nuclear plan t is construction cost; the cost of the fuel is much lower. Some plants received there construc tion licenses but could not get a pproval for the operating license. This also caused many utilities to cancel orders for other new power plants fearing that the same fate would occur to their investment and the mone y would be lost due to the political landscape. Regulators realizing the need fo r new nuclear power plants crea ted the new licensing process allowing for the hopeful revival of nuclear power plant construction. Early Site Permit The new license process is a 3 prong approach which will allow the utility greater ease with the licensing process. The first part is the ear ly site permit. The early site permit allows for the utility to apply for th e possibility to build a re actor or reactors at a site specifically identified in the permit. Before a permit is granted hearings must take place which are called for in 10 CFR Part 52. Once the permit is grante d then it is valid for no less than 10 years but no more than 20 years. Also in the event that the permit ex pires but a proceeding on a combined license application is in effect the permit remains valid. An early site permit may is to in a combined licensed application at the discre tion of the utility and not a requi rement for a combined license application. However it is expected that many utilities will first get the early site permit.9 The early site permit must contain all in formation required by 10 CFR Part 50.33 a through d, the information required by 10 CFR Part 50.34 a 12 and b 10, and to the extent approval of emergency plans is sought under pa ragraph b 2 ii of this secti on, the information required by 10 22

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CFR Part 50.33 g and j, and 10 CFR Part 50.34 b 6 v. It must also contain the description and safety assessment of the site. An evaluation of the major structures, systems, and components of the facility that affect the accep tability of the site under the radiological consequence evaluation factors described in 10 CFR Part 50.34 a 1. It also describes the number, type, and thermal power rating of the plants, boundary of the site, gene ral location of each faci lity, maximum effluents produced, the cooling systems to be used, the ge ological site characteristics, the location and purpose of nearby surroundings, and a future populat ion profile for the site. Lastly a complete environmental report is included with the application.9 Combinded Operating License The next prong is the Combined Operating Licen se COL, which allows the utility to both construct and operate the plant. The purpose of this part was not to block the public from objecting to a nuclear power plant but to streamline the construc tion process and to stop cost overruns. This process allows for public hearings at the beginning of the licensing process and during the acceptance of the Inspections, Test, Analyses, and Acceptance Criteria (ITAAC). Also note that all the information required in 10 CFR Pa rt 50 is required in 10 CFR Part 52 and that if the utility chooses 10 CFR Part 50 is available for us e. It also allows for the licensee to refer to either the early site pe rmit, the design certifications sought by the reactor vendors, or both. If the utility chooses they submit their own design and not seek an early site permit but then the information required in both the design certification and the early site perm it must be included in the COL. The ITAAC is required in the COL app lication. The ITAAC contains all the material, locations, test, and criteria that must be me t and certified. These are met at the end of construction before the facility is used for comm ercial operations. If there is any modification to the design of the plant during construction the new design must be submitted to the NRC for review because it changes the licen se that the plant is authorized to operate under. 10 CFR Part 23

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52 also allows the Commission to make effectiv e an amendment to the license that involves no significant hazardous conditions; the Commission also makes effective an amendment in advance of holding a hearing on the propos ed amendment if it deems appropriate.9 The NRC is confident that the new COL pro cess provides a timely response so that new nuclear reactors are built on schedule and without cost overruns due to regulatory problems. David Matthews the director of the Division of New Reactor Licensing in a presentation given on September 13, 2006 used Figure 3-1 COL Review Process to describe how the new proposed COL applications would proceed. Figure 3-1. COL Revi ew Process (Source: http://www.engr.utexas.edu/trtr /agenda/documents/MatthewsStatusofNewReactorLicensingActivities.pdf Last accessed June 21, 2007). Design Certification The last major part of the new licensing pro cess is the design certification. In the past utilities ask a vendor to design a sp ecific system that meets some utility requirements at a site of 24

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their choosing. This causes all the sites in the US to be different in one form or another and no standard design is kept. The new way of thinking is that all the new plants for a vendor have a similar design. In order to ease the regulatory review during the COL application a utility chooses a certified design. In orde r to obtain a certified design a vendor pays the U.S. NRC to review a design. Once the NRC reviews the design if it is certified then a utility can refer to it during the COL application; a utility can also refer to a design under review. A design certification is valid for 15 year s plus the duration of any ongoing activity for which the design is used as long as the activity is doc keted before the date of expiration.9 The design certification must c ontain all technical information required from applicants on COL that are located in 10 CFR Part 20, 50, 73, and 100, all the technical information required after Three Mile Island found in 10 CFR 50.34 f except paragraphs f 1 xii, f 2 ix and f 3 v. The design certification is not site specific but includes generic site parameters called for in the design, proposed resolutions to safety issues identified in NUREG-0933, and a design specific probabilistic risk assessment (PRA). Proposed ITACC requi rements are included and the technical reasons that these re quirements will allow for the proper testing of their system.9 This information is separated into 2 Tiers. Tier 1 includes definiti ons, general provisions, design descriptions, and the IT AAC. Tier 2 includes informa tion required by 10 CFR 52.47, with the exception of technical spec ifications and conceptual design information, FSAR, supporting information on ITAAC test, and items needed in th e COL. A sample Table of Contents for Tier 2 from the AP 1000 is seen in Table 3-1 Tier 2 Chapter List.11 Table 3-1. Chapter List Chapter Number Chapter Title Chapter 1 Introduction and General Plant Description of Plant Chapter 2 Site Characteristics Chapter 3 Design of Structur es, Components, Equipment and Systems 25

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26 Table 3-1. Continued Chapter Number Chapter Title Chapter 4 Reactor Chapter 5 Reactor Coolant Sy stem and Connected Systems Chapter 6 Engineered Safety Features Chapter 7 Instrumentation and Control Systems Chapter 8 Electric Power Chapter 9 Auxiliary Systems Chapter 10 Steam and Power Conversion System Chapter 11 Radioactive Waste Management Chapter 12 Radiation Protection Chapter 13 Conduct of Operations Chapter 14 Initial Test Program Chapter 15 Accident and Analysis Chapter 16 Technical Specifications Chapter 17 Quality Assurance Chapter 18 Human Factors Engineering Chapter 19 Probabilistic Risk Assessment Tier 1 information comes from Tier 2 informa tion and all Tier 1 information is fulfilled with unless a plant specific exemption is granted by the NRC. Differences in design descriptions found in Tier 1 and the Tier 2 Information, the design descriptions take precedence but the information in Tier 2 is not the only way to comply with the information in Tier 1.11 If there is a certain part of the design that the applicant doe s not seek certification for a conceptual design is needed. The applicant shou ld be sure they provided enough information for the Commission to be able to review and accept the proposed design and if there are any questions the applicant must submit the inform ation that the Commission request. The design must also include information on the performance of the safety features of the design. The utility applying for the COL provides all the missing info rmation related to site specific information.9

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CHAPTER 4 REACTOR DESIGNS AP 1000 The AP 1000 is Westinghouses version of a ge neration 3+ design. The design is for a twoloop reactor with an 1117 MWe output. The total reactor power is 3,400 MWt. It contains 157 fuel assemblies in a 17 X 17 array that is 14 ft tall. The total fuel wei ght is 211,588 lb and has a 5.72 kW/ft average linear power density. The av erage core burnup is approximately 60,000 MWD/MTU.11 The AP 1000 uses passive safety systems so that in an accident scenar io no on-site or offsite power is needed. There are 5 systems that are considered nuclear safety systems in the DCD tier 1 for the AP 1000. The steam generator system transports the main steam produced to the turbine and secondary components. It is considered a safety system because it must isolate the primary system from the secondary system in a de sign basis accident. The valves associated with containment isolation in the st eam generator fail closed so th at primary water does not leave containment.11 The second safety system is the passive core cooling system. The passive core cooling performs safety injecti on and core makeup and passive residual heat removal. There are 2 nozzles on the reactor vessel de dicated to safety injection. Th e makeup water can be found in core makeup tanks (CMTs), accumulators, or in-containment refueling water storage tanks (IRWST). High pressure safety in jection would be handled by CMTs which are located above the reactor coolant system (RCS) loop piping. If a signal is given th at the water level or pressure in the pressurizer is too low, the reactor cool ant pumps trip and the CMT are emptied by gravity into the reactor vessel. The CMTs contain borated water and are in parallel trains on each leg of the RCS. The accumulators is actuated after pres sure deferential between the accumulators and the RCS drops, the check valves isforced open and water refills the downcomer and the lower 27

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plenum to help the CMTs in re-flooding the co re. After the pressure is low enough, water from the IRWST which is located above the RCS loop pi ping flows into the core. The IRWST is kept at atmospheric pressure.12 The passive heat removal system is located in containment and the flow for the passive system is generated by natural circulation. The h eat exchanger is locate d in the IRWST and has 100 percent capacity. Water volume is sufficient to delay boil ing in the IRWST for 2 hours. Once boiling starts the steam produced condenses on the steel containment vessel wall, and with the help of special gutters, flows back into th e IRWST. Another passive system is the passive containment cooling system which again is driv en by gravity. As seen in Figure 4-1 Passive Containment Cooling air enters the containment vessel through the top of the sides and flows down an air baffle. Air then rises up the steal wa lls of the inner containment and leaves through the roof of containment.12 Figure 4-1. Passive Containment Cooling (Source: http://www.westinghousenuclear .com/docs/AP1000_brochure.pdf Last accessed June 21, 2007). The tank on top of the containmen t vessel releases water to help cool the steal containment vessel if more cooling is needed because of hi gh containment pressure. There is enough water in these tanks for 3 days. After three days the tanks mu st be refilled, but calculations show that if 28

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they are not refilled, peak pressure will only reach 90% of design pressure. The main control room emergency habitability system is the last of the 5 systems mentioned in tier one. The system is made of compressed air tanks to provide a livable environment for 11 personnel. The compressed air tanks are connected to the control room by a main and alternate delivery line so that no single failure can occur.12 Since all passive systems are being used there is no need for safety related diesel generators. The AP 1000 like most new designs al so reduces the number of pumps and feet of cable needed to construct and operate the plant.13 The spent fuel pool allows for enough storage for 10 years of operation plus one full core. This is important because we currently do not have a national spent fuel storage plan.8 Many of the active safety systems used in other plants though ar e retained and are no longer considered safety related. Reliability The AP 1000 is expected to ach ieve over 93% capacity factor.12 With a high capacity factor and defense in depth for the AP 1000 syst em the core damage frequency (CDF) is 4.2 x 10-7 per year and a large re lease frequency is 3.7 x 10-8 per year.13 While the large release frequency may be low, workers do receive radi ation. Table 4-1 Doses to workers shows the doses the workers receive at the site.11 Table 4-1. Dose to Workers AP 1000 Category Percent of Total Estimated Annual (man rem) Reactor operations and surveillance 20.6% 13.8 Routine inspection and maintenance 18% 12.1 Inservice inspection 24.7% 16.6 Special maintenance 22.4% 15 Waste processing 7.7% 5.2 Refueling 6.6% 4.4 Total 100% 67.1 29

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Technology Maturity The AP 1000 is the only gene ration 3+ reactor to receive a design certification from the NRC, though there has been an amendment submitt ed for review. The fuel used in the AP 1000 is currently being used in 4 othe r U.S. Power plants. The reactor ve ssel and internals are used in Doel 4 and Tihange 3. The control rod drive mo tors are the same in use throughout the world. The large model F steam generators are curren tly used in 4 other sites like Waterford. The reactor coolant pump (RCP) is based on a canned motor pump which has been used by the U.S. Navy and fossil boilers. The pressurizer is used in 70 other plants worldwide.12 Economics In order to lower the construction costs, the buildings needing seismic stability are reduced but there is still working space due to the reduction in active equipment. Modular construction will also be used to decrease construction time.14 The AP 1000 plant uses over 270 modules to help reduce construction time, reduce man power needed, reduce site co ngestion, and increase reliability in construction. It is expected that the first two reactors will share the cost for the first of a kind (FOAK) engineering. With the first of a kind cost included, the first 2 units are estimated by the vendor to cost around $1488/KWe. The nth plant which is the plant that comes after the first 2 which in cludes the lessons learned from constr uction is expected to cost about $1134/kWe. Westinghouse also estm ates that it cost between 3 to 3.5 cents/kWh to operate a duel unit plant. These cost estimates are based on calculations made fo r the AP 600 plant which is a smaller version of the AP 1000.15 The major factor in cost though is the time delay from order to operation. The construction schedule calls for a 5-year delay. The delay is factored into 3 sections. Base d on the vendor estimation, the construction site preparation time is 18 months. The construction time is 36 mont hs. The last 6 months includes the start up and testing of the reactor.12 30

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Supply Chain The supply chain of the AP 1000 is indicative of the industry as a whole. There are long lead times for the large items such as the reactor pressure vessels and other large forgings. Unlike some of the other plant designs the AP 1000 and its predecessor the AP 600 have not been built. The supply chain is based on smaller firms providing some of the module construction. Many of the firms th at used to be certified to pr oduce the nuclear grade equipment have not kept the certification. Another industry-wide problem is the personnel supply chain. Nuclear power plant companies have a hard time finding employees to fill the required positions. ABWR The ABWR is a generation 3 BWR design by GE Hitachi, and Toshiba. The design is a 1325 MWe output reactor with 4 main steam lines w ith flow restrictors and isolation valves on each line. The total reactor power is 3,926 MWt. It contains 872 fuel assemblies in a 10 X 10 array that are 12 ft 6inches tall. The fuel has a 4.28 kW/ft average linear power density. Control for the reactor is provided by 205 control blades driven by fine motor control rod drives that drive the blades in from the bottom of th e core. The average core burnup is about 60,000 MWD/MTU.16 The ABWR is the GE-Hitachi-Toshiba first step into creating a new design for generation 3 reactors. The ABWR has safety grade Emergency Diesel Generators (EDGs) which provide power to the systems safety features in an acc ident scenario. The system is equipped with 3 diesels that need to be serviced and main tained along with many other safety equipment currently found in other BWR plants. The ABWR includes desi gn features to help improving the safety and simplicity of the system. For instance, the external recirculation pump in ABWR is replaced by sealed internal pumps that improves the operational flexibility and safety 31

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performance of the system.17 The reactor has 10 of these internal pumps located inside the reactor pressure vessel (RPV) as it is depicted in Figure 4-2. ABWR Reactor Pressure Vessel. Figure 4-2. ABWR Reactor Pressure Vessel (Source: http://www.gepower.com/prod_serv/products /nuclear_energy/en/downloads/gea1457 6e_abwr.pdf Last accessed June 5, 2007). One of the major features in the ABWR is that there are no large pipe nozzles or external recirculation loops bellow the top of the core. This feature helps keep th e core covered in the case of a loss of coolant accident (LOCA). If a LOCA were to occur two motorized high pressure core flooders (HPCF) in each division of the E CCS network start filling the reactor core. The suction for the HPCF comes from the condensat e storage tank (CST) or the suppression pool (SP) as an alternate. The reactor core isola tion Cooling (RCIC) primary purpose is to provide reactor cooling when the core is isolated from the turbine as a main heat sink. The RCIC operates using steam from the reactor to turn turbin e pumps. The RCIC has the capacity to provide cooling when the vessel is in hot standby, during a plant shutdown with loss of feedwater before alternate systems come online, a nd during a loss of AC power. Th e suction for the system comes from the CST or the SP with the drain from the steam turbine going to the main condenser. After a short delay the automatic depressurization system (ADS) is initiated and the reactor goes from 32

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1040 psia to pressures where the residual heat re moval system can be used. The residual heat removal system has 6 major functions: low pre ssure flooder, suppression pool cooling, reactor shutdown cooling, primary containment vessel sp ray cooling, supplemental fuel pool cooling, and AC independent water addition with water from the fire protection system.17 During an accident or refueling operations a standby gas treatment system is used to remove radioactivity from the ai r and release through the plant st acks. During an accident there is an atmospheric control system that will keep an inert atmosphere in the containment building. Due to the Three Mile Island accident and less ons learned a flammable control system which will not let the hydrogen ignite in containment is installed. In the event that the control rods cannot shut down the reactor a standby liquid contro l system (SLCS) pumps boron into the reactor. Many of these systems need AC power that is why the EDGs are safety related.17 An order of magnitude improvement in safety incl uding the core-recovery during LOCA is achieved by the ABWR. The low frequency of the ABWR core damage as compared with other US BWR and PWR plants as well as with the standardi zed Japanese BWR (BWR-5) is shown in Figure 43 Comparison of PSA Results. 19,20 Figure 4-3. Comparison of PSA Results (Sourc e: CSNI Workshop on PSA Applications and Limitations, NEA/CSNI/R (91) 2, Sant a Fe, New Mexico, September 4-6, 1990. Severe Accident Risks: An Assessment fo r Five U.S. Nuclear Power Plant, NUREG 1150, Vol. 3, 1989 (released 1991)). 33

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Besides safety system the ABWR employs ma ny technological advan ces since the last BWRs. It uses a fully digital cont rol system for accurate plant mon itoring and uses better fuel to reduce radioactive waste.18 The ABWR keeps the radioactive waste volume to around 735 ft3/yr, which is a significant reduction from the averag e waste volume for the previous generation of BWR plants. It has enough storage in the spent fuel pool for 10 years and one core.17 Reliability The ABWR is expected to achieve 95% capacity factor.17 With a high capacity factor and active defense in depth for the ABWR system the core damage frequency (CDF) is 1.6 x 10-7 per year and a large releas e frequency is 1 x 10-8 per year.17 Along with better monitoring of the reactor system and a low LRF the annual man-re m to workers in ABWR system is to 98.9 manrem/ year as can be seen in Table 4-2. Dose to Workers ABWR.16 Table 4-2. Dose to Workers ABWR Task Location Percent of Total Estimated Annual (man rem) Drywell 46.7% 46.2 Reactor Building 14.7% 14.5 Radwaste Building 10.6% 10.5 Turbine Building 11.8% 11.7 Work at Power 16.2% 16 Total 100% 98.9 Technology Maturity The ABWR is the only design evaluated in th is study that has opera ting experience. There are 4 such units in Japan operating sin ce 1996, 1997, 2004, and March 2006, an additional 3 under construction, 2 in Taiwan a nd 1 in Japan. Eleven more are planned for construction, 9 in Japan and 2 in US. The ABWR technology is ba sed on 50 years of experience using BWR power 34

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plants. The ABWR is also a certified design which would allow for construction to begin sooner.18 Economics The ABWR has a proven record of reducing cap ital, operational, and maintenance cost. Its construction time is approximately 39 months which is proven in Japan. It has a smaller reactor building to decrease c onstruction time. The ABWR also uses modular design which allows for high quality and reliability in the constructi on. The ABWR also has no FOAK cost associated with the plant design since the ABWR is fully engineered and fully constructed in other countries. Based on the experience in building AB WR systems in Japan, a more realistic cost estimate for this system is possible. In a recent DOE document the estimated overnight cost of the ABWR is cited between $1400-1600/kWe.15 Supply Chain GE with the help of Hitachi and Toshiba is currently building two ABWR plants in Taiwan. Hitachi is building anot her ABWR plant in Japan at predictable construction schedule and cost. GE is building the Taiwan plants us ing ASME codes and meeting all US regulatory requirements.15 ESBWR The ESBR is the newest reactor design from GE. It is based on BWR technology but has not been certified by th e NRC. GE, with this design, is simplifing the plant to remove complexity. The design is a 1535 MWe output react or with 4 main steam lines with flow restrictors and isolation valves on each line. The total reactor power is 45 00 MWt. It contains 1132 fuel assemblies in a 10 X 10 array that is 10 ft tall. The fuel has a 4.6 kW/ft average linear power density. Control is provided by fine motor c ontrol drives that drive 260 control blades into the core to control power. The aver age core burnup is around 60,000 MWD/MTU.21,22 35

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The safety systems in the ESBWR are passive and therefore do not need AC power in an accident. The isolation condenser (IC) system is located above the reactor removing the need for pumps and allowing the system to be gravity driv en and is attached to the passive containment cooling (PCC) pool. Water travels from the isolation condenser to the reactor core and steam goes back to the condenser. A vent is provided in the system to allow non-condensable gases to be vented and the water in the system to stay in the IS and PCC. In order to depressurize the system the ADS actuates and opens the relief valv es during a LOCA. Due to the ability of the ADS system to depressurize the core quickly, the ECCS is now driven by gravity. In order to ensure there is no containment failure a passive containment cooling system is employed. There are a total of 6 loops to condense steam and cool the containment. Each loop is rated for 11 MWt and without makeup to the PCC pool; they can co ol containment for 72 hours. The SLCS system in the ESBWR has two methods of delivery. Pr essurized accumulators which require no AC power inject borated water into the core via squib valves, or a non safety related nitrogen pressure charging system is used. In the event of an accident which places the area outside of the control room uninhabitable the emergency breat hing air system provides enough air for 72 hours after the accident. The air is in compressed air tanks that also pressurize the control room to minimize in-leakage from the outside air.21 The ESBWR however is mainly known for the f act that it uses natural circulation. GE achieves this by partitioning a ch imney directing the steam flow above the core. The downcomer is also enlarged to reduce flow resistance and the entrance of the feedwater is placed high on the reactor vessel to provide more driving head. A basi c drawing is depicted in Figure 4-4. Natural Circulation.23 36

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Figure 4-4. Natural Ci rculation (Source: http://www.gepower.com/prod_serv/products/nuclear_energy/en/ downloads/natural_c irculation_esbwr.pdf Last accessed May 10, 2007). There are 11 fewer systems than in previous power plants and a 25% reduction in pumps, valves, and motors.21 Along with the reduction in systems and components the ESBWR keeps the radioactive waste to around 735 ft3/yr. It has enough storage in the spent fuel pool for 10 years and one core. These numbers are estima ted based on the experience from the ABWRs already built.21 Reliability The ESBWR is expected to achie ve above 95% capacity factor.21 With a high capacity factor and active defense in depth the core damage frequency (CDF) is 3 x 10-7 per year and a large release frequency is 1 x 10-9 per year.21,24 The estimated dose to radiation workers in the ESBWR is 60.4 man rem/yr Table 4-3 Dose to Workers ESBWR.22 Table 4-3. Dose to Workers ESBWR Task Percent of Total Estimated Annual (man rem) Drywell 34.3% 20.7 Reactor Building 28.0% 16.9 Radwaste Building 4.1% 2.5 Turbine Building 18.5% 11.2 Work at Power 13.2% 8 Fuel Building 1.8% 1.1 Total 100.0% 60.4 37

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Technology Maturity ESBWR is considered a generation 3+ reactor design. It is currently going through the NRC review for the Design Certification and the Sa fety Evaluation Report (SER) is expected to be released was filed in 2008.25 The ESBWR is basing much of the technology it uses on experience from the ABWR. Reactors in the past ha ve run on natural circulation even though this reactor has much more power the technology is proven effective by Dodewaard a 60 MWe reactor that ran for 25 years. Also many of our current BWRs can run at about 50% power on natural circulation. Also many of the passive safety features envisioned for the SBWR are incorporated in the ESBWR design. Effectivel y no new systems are designed for the ESBWR they are adjusted, uprated, or simplified for the ESBWR.24 Economics The ESBWR is designed to use many of the c onstruction techniques that are used in the ABWR. Construction time provides a substantial cost increase, but with the use of modular construction technology the GE estimated tim e of construction is about 36 months.26 The first few plants must account for the FOAK engineering cost. Based on the GEs estimates, due to the simplification of the design even with the FOAK the cost of building ESBWR could be less than ABWR.15 Supply Chain The ESBWR uses the supply chain already in use for the ABWR for most of its components. The modules are built offsite and brought onsite for final construction. GE plans to construct the turbine island, asso ciated equipment and the instru mentation and controls in the U.S. The large forgings are made overseas, depending on Japan Steel Works for the reactor pressure vessel forgings. Due to the ties be tween the ABWR and ESBW R supply chains a bottle neck may occur in the future.15 38

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APWR The APWR is a 1,700 MWe plant. The design ca lls for 4 loop PWR system with a core power of 4,451 MWt. It uses a 17 x 17 assembly design and has 257 assemblies in the core. The average linear power density is 5.36 kW/ ft. Th e core burn up is estimated to be 62,000 MWD/MTU.27 The attachment of the reactor internal s is made without the use of welds to facilitate the quality control dur ing installation and for later inspection. The r eactor vessel itself is made out of partially low alloy steel to minimize welds that would require regular inspection. Also forged rings are used to reduce welding by completely eliminating welds along the belt line. Control is provided by both boric acid and 69 contro l rods. The reactor coolant pump is a vertical shaft, single-stage suction diffuser with a non-contact controlled leakage system.28 The containment vessel is a cylindrical pres tressed concrete containment vessel with a hemispherical dome and a carbon steel liner. It has an inner diameter of 43 m and is 1.3 m thick. The design pressure for the containment building is 83 psia. In case of a core damage accident there are hydrogen igniters inside containm ent to prevent hydrogen explosions. Surrounding containment is an annulus that is kept at a slight negative pressure to keep the radioactive material from being released in an accident. In the case of a LOCA the APWR uses accumulators to provide large-flow, low-pressure injection an d safety injection pumps to provide high head injection. The safety injection pum ps take suction from the refueling water storage tanks located in containment. Containment cooling is achieved by the containment spray system. The containment sprays also take suction from th e refueling water storage tank. Residual heat removal (RHR) is responsible for removing the decay heat from the reactor, the system is tied to the containment sprays. In acciden t scenarios power to the safety systems is provided through the safety-related gas turbine generators.27 39

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Along with improving the system associated with plant safety, the APWR is more environmentally friendly and produces about 60 drums/yr of radioac tive waste vs. the 100 drums/yr currently produced which is approximately 441 ft3/yr of radioactive waste. This will be possible by using enriched B-10, a better drain rec overy system, a boric acid recovery system, and a high performance cement solidification system.28 Like the other designs the APWR uses a dig ital control system. The system allows for daily load following, automatic frequency cont rol and house load operation. Mitsubishi uses advanced methods, including the prefabricated unit method and the steel-p lated concrete method to deliver the plants on time with the highest quality.28 Reliability The thermal efficiency is projected to be 39% giving it the highest efficiency estimate for the new reactors.28 The PRA demonstrates a reduction of 1/10 the prior value of the CDF. The target CDF and LRF is lower than 1x10-5.27 Technology Maturity The APWR was submitted for the NRC Design Ce rtification review on January 4, 2008. It is a generation 3 design like most of the reactors in this study.25 Although Mitsubishi has never built a reactor in the U.S., it has built reactors in Japan. With this continued experience the APWR has become more of a work in progress than a new reactor design.29 Economics APWR like all other reactors has many factor s controlling the cost per kWe installed. Construction time is a major cost factor so the highest levels of contro l is exercised, including utilization of the Integrated Project Schedule Cont rol System. Construction is therefore estimated to be 46 months. Since this is the first time an APWR will be built in the U.S. there will be FOAK costs associated with the APWR which is targeted to be $1500/kWe.28 40

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Supply Chain Since 1970, Mitsubishi has been constructing nuclear reactors, it currently plans to continue with its rich history. Throughout its operating history not only has Mitsubishi constructed reactors but it has also exported major components to the world. Mitsubishi currently plans to continue making the parts in Japan and exporting them to its worldwide clients. Mitsubishi has fabricated a total of over 323 major components and is confident that they can meet the supply demand for new reactor construction.28 EPR The EPR is the new reactor design by AREV A NP. The EPR is a 1600 MWe reactor that has 4,500 MW of thermal power. The power come s from 241 fuel assemblies made from UO2; the reactor however is capable of using MOX fuel. Each fuel assembly is 14 ft tall and arranged in a 17 x 17 array. The core has an av erage linear power de nsity of 4.75 kW/ft.30 Each assembly has a maximum discharge burnup of greater th an 70,000 MWD/MTU. Control is provided by 89 control rods and boric acid. The EPR uses a four-train safety system in which each train is independent of the other so that a common mode failure cannot o ccur. Each train is capable of performing all safety functions.30 Each train has a safety injection that provi des water from the accumulators located in containment. Medium head safety injection is provided via safety inj ection pumps which pump water through the cold legs and take suction from the IRWST. Low-head water is provided by the RHR system. Two of the trains have extra boron systems that allow for 7,000 ppm of boric acid to ensure the unit is shut down. In order to ensure power to these systems in an accident there are 4 Emergency Diesel Generators in 2 sepa rate buildings. In the event of a station black out 2 additional diesel generators provide power to the safety busbars of 2 trains. In addition to many safety features to protec t the core, if there is a core melt the EPR is designed for 41

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mitigation. The containment building is a double wa ll design with each wall being 4.3 ft thick to prevent releases. There are dedicated relief valves to mitigate a high pressure core melt even if the pressurizer relief valves fa il. The high strength pressure vessel prevents damage from reactions between corium and water. H ydrogen explosions are mitigated by hydrogen recombiners that will keep the hydrogen levels inside containment belo w 10 percent. Corium flows from the reactor vessel in th e case of a reactor vessel failure into a core ca tcher area where a long term containment heat removal system co ols the corium and allows it to solidify. The control room is fully digital and is located in on e of the safeguard buildings allowing for it to be manned during most accidents.31 Reliability The EPR operates within large margins and flexibility in order to allow for future regulations and standards. It is expected to have a 92% capacit y factor and operate for 60 years.32 The dose to workers is estimated at 40 man re m/yr due to its safety, simplicity, and health physics programs. Major accidents should not occur with a core damage frequency of 4 x 10-7 and a large release frequency of < 10-7.30 Technology Maturity EPR technology is based on the French N4 reactor and the German Konvoi reactor design. There is one EPR being built in Finland and anothe r one in France. Currently in the US the EPR is under review for design certificatio n which was filed at the end of 2007.25 Economics As with most new plants long construction times cause signif icant cost increases for the buyers. The preconstruction period of 15 months is needed for com ponents with long lead times. Once construction begins it is expected to take 48 months for commercial operation. Like other plants modular construction is used, and experiences from Finland and France is used to stream 42

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43 line the construction.30 The Unistar group of nuclear plant oper ators has the overnight cost of the EPR at $2400/kWe if the FOAK cost are spread over 4 units.32 Another cost-saving measure is reducing the components in the syst em. Figure 4-5 shows the percentages reduced. Figure 4-5. EPR Component Reduction (Source: http://unistarnuclear.com/Right_Team.pdf Last accessed June 29, 2007). Supply Chain Areva has over 4,200 employees in the U.S. wo rking on nuclear projects. Eighty percent of the work for an EPR order is done by U.S. employ ees with most of the engineering work done in Arevas offices in North Carolina and Virginia. Arev a is also the worlds largest nuclear supplier with manufacturing capability. The large components are fabricated by FANP and the other components such as the electrical sy stems are fabricated in the U.S.31

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CHAPTER 5 DISCUSSION Robustness Robustness in a design is the ability to keep a plant within a safe operating condition. New reactor designs have incorporated features that al low operation of the plant more safely and with more margin. All of the control rooms are fully automated and computers disseminate information much more efficiently to the operators allowing for a less stressful environment. But no matter how easy a plant is to operate; there is no substitute for design margin. In the future, new regulations, unforeseen problems, design mo difications, and improvements may arise and having extra margin allows for accommodation. Many factors can be categor ized as margin but this study will look at thermodynamic margins in the power plants, spec ifically linear heat generation. The IAEA published a document wh ich lists items that should be limited and controlled to keep fuel failures to an acceptabl y low level. These items included linear heat generation rate, critical power ratios, minimum depa rture from nucleate bo iling, and peak fuel and temperature.34 The study uses the linear heat generation rate because the amount of heat produced in a specific area will determine the temper atures in the area that will affect all the other parameters. Table 5-1 Line ar Heat Generation Rate shows the linear heat generation rates for the studied reactors. Table 5-1. Linear He at Generation Rate Reactor Design Linear Heat Leneration(kW/ft) AP 1000 5.72 ABWR 4.28 ESBWR 4.6 APWR 5.36 EPR 4.75 44

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The ABWR has the lowest LHGR. Among the Ge neration 3+ reactors the ESBWR has the lowest LHGR and the AP 1000 has the highest. Th e lower the LHGR the more margins available during an accident. Safety Each design has features that make it robus t and safe. The safety systems for all the reactors are carefully designed and perform their needed functi ons in accident scenarios. The main choice in the safety category is whether to go with an all-passive safety system design or an active safety system design that will include anot her form of electricity in an emergency. Both the AP 1000 and the ESBWR are passive safety systems that use gravity and pressure differentials in emergencies. They have no need for secondary power sources in emergencies other than for instrumentation and can cool the co re while off-site power is being restored. They both also use compressed air for the control room duri ng postulated accidents. Emergency diesel generators (EDGs) are therefore no longer safety-grade. Systems that were in previous PWRs and BWRs are no longer safety-grade but like the diesels may still be present. The ABWR and the EPR both use the proven technology and use sa fety-grade EDGs. They must be maintained and have surveillances performed like existing dies els at current nuclear r eactor sites. EDGs can have many problems, in 1999 a study was prepar ed for the NRC looking for a common cause of failure in EDGs while the study found no firm c onclusions Figure 5-1 EDG Failures shows that from 1980 to 1995 there were 131 events that caused the diesel to fail the testing.35 45

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Figure 5-1. EDG Failures In the Figure 5-1 series1 means failed to run a nd series2 means failed to start. As depicted in the chart emergency diesel generators fail fo r many different reasons and must be maintained properly. Both the ABWR and EPR have multiple diesels therefore maintenance activities are factored by the number of diesels and so are chances of human performance errors.35 In the U.S.APWR, Mitsubishi has decided agai nst using emergency diesel gene rators and inst ead electing to go with gas turbine generators. The GTG though have a slower st art time of about 40 seconds but Mitsubishi feels the 10-3 failures/demand is better than the 10-2 failures/demand of the EDG.28 Defense in depth is a key word in nuclear pow er however the EPR has 4 independent trains that each has 100% capacity to cool the core in an accident. Though each reactor uses the defense in depth principle, the EPR has the most systems due to the 4 redundant trains and the preparations the EPR has in case of a core melt su ch as the core catcher. The EPR has the most defense in depth mechanisms than any other reactor design. The 4 re dundant systems not only have separate wiring and piping to prevent common cause failures between trains, but the trains are physically separated by buildings. Many incide nts occur where a component on one train can be confused with the same component on anothe r, building separation of the trains allow for human performance techniques to be more apparent since the trains are not located in the same 46

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work space. This also allows for a catastrophe to occur on one side of the plant and have minimal or no impact on the trains on the opposite side. The AP 1000 and the ESBWR are passive systems during accident scenarios. Passive system s allow for operators to respond to the accident rather than worrying if their equipment works due to lack of power. When running on the diesel engines, operators must check the logic and sa tisfactory loading of pl ant equipment to supply water to the core. If not loaded properly the diesel may fail or a component might not provide proper core cooling. The APWR has an advantage over the ABWR since it has no EDG and the failure/ demand rate of a steam turbine generator is lower than an EDG. While there is no U.S. experience with safety-grade gas turbine generators Mitsubishi has successfully used them in Japan. Another safety consideration th at should be taken into accoun t is core damage frequency (CDF) and large release frequency (LRF). These ne w reactor designs are a large step forward in safety and therefore have lower CDFs and LRF s. As can be seen in Table 5-2 CDF and LRF the CDF and LRF values are lower than those of t odays reactors. Table 5-2. CDF and LRF Reactor Design CDF LRF ABWR 1.60E-07 1E-08 ESBWR 3.2E-08 1E-09 APWR Target <1.0E-05 Target <1.0E-05 EPR 4.0E-07 <1E-07 AP 1000 4.2E-07 3.7E-08 The ESBWR has the lowest CDF and LRF follo wed by ABWR. No value is given for the core damage frequency and large release freque ncy for the APWR design. The values in Table 52 indicate the target goa ls for the CDF and LRF. 47

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Reliability All new reactors featured are expected to ha ve good reliability. The capacity factors for all the reactors are above 90%, the na tional average for the current reac tor fleets best reactors. As expected the new technology has a lower CDF and LRF than current reactors due to enhanced safety features in the power plants. The NRC requires the CDF to be at 10-4 or lower and all these reactors meet that criterion. With all the systems, new generation reacto rs are more reliable and less likely to be damaged. Capacity factors are a good measure on the reliability because they show how plant equipment failures are not the cause of unit downpowers or shutdowns. Table 5-3 Capacity Factor shows the capacity factor da ta for the reactors in the study. Table 5-3. Capacity Factor Reactor Design Capacity Factor AP 1000 93% ABWR 95% ESBWR 95% APWR 92% EPR 92% The highest capacity factors are for the AB WR and ESBWR. For only the ABWR system the estimated capacity factor is based on operati onal data. The EPR and APWR have the lowest estimated capacity factors. The capacity factors have a direct correlation to money because the longer a plant is down the more cost is incurred by the utility to buy other fo rms of fuel to make up the MWs needed for their customers. Technology Maturity Each reactor is in a different stage of devel opment. Some of the reactors are already built overseas and have a rich history of operation and others have never been built before. 48

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The ABWR is the most mature design since it has been built and new units are on order. The ABWR design is also approved by the NRC. The EPR is the next most mature reactor since it is currently under construction in Finland. Since the design is completed and is under construction the reactor must unde rgo changes only to meet U.S. regulatory requirements and constraints. The AP 1000 is th e next most mature design beca use it is licensed by the NRC; however, a revision is submitted to the NRC. In or der for the reactor design to be approved, most of the engineering design is completed and ther efore the AP 1000 is more mature than the other reactors which have not yet constructe d. The ESBWR is based on ABWR and SBWR technology and many of the systems are tested and accepted. The U.S.-APWR is now in the preliminary stages of the application proce ss and none are being constructed currently. The APWR however is currently under review and is selected for construction by 2015-2016 in Japan. Economics Costs associated with nuclear power plants are mainly initia l capital costs and interest on that capital cost. However Operations and mainte nance cost can and will affect the plant and its revenue. The NRC also estimates it costs $1000/ ma n rem and each reactor tr ies to achieve lower man rem/yr values to save employees from doses and save money.36 The lowest is the EPR followed by the ESBWR, AP 1000, and finally the ABWR. There is no information on doses to workers for the APWR. One surprising find is th at the ESBWR has a lower dose to the worker than the AP 1000. As the trend in industry shows BWRs have more dose to workers. Depreciation rates also play a large role in the final cost of nuclear power. Construction times for the new reactors are concerns since most states do not allow for companies to bill consumers until the plants have been built. Each co mpany predicts that it will take 4 to 5 years to build one of these reactors which are in line with NRC predictions for how long it takes for 49

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reactor construction. The estimated overnight co sts for the reactors as reported earlier are summarized in Table 5-4. Table 5-4. Reactor Overnight Cost Reactor Design Overnight Cost ($/kWe) AP 1000 1488 ABWR 1400-1600 ESBWR <1400-1600 APWR 1500 EPR 2400 As is seen in Table 5-4 Reactor Overnight Cost many of the reactor vend ors feel the cost is around $1400/kWe so the cost is about equal for th e 4 reactors. However, based on more recent estimates these numbers are low by a factor of two or more. The EPR estimate is based on the fixed price contract for the unit that is under construction in Finland. After the reactors have been built they must make profit in order to pay off the initial investment. Operations and maintenance costs will play a role in reduci ng the companys profit. Table 5-5 Operations and Maintenance Cost shows the operations and maintenance costs for the reactors.15,37,38 Table 5-5. Operations and Maintenance Cost Reactor Design Cost (mils/kWe) AP 1000 8.17 ESBWR 6.83 ABWR 6.71 APWR 9.10 EPR 7.00 Allowing for reactor construction at locations where reactors already exist lowers cost. Existing reactors have large support staffing that may be shared with the new reactors reducing the cost. An interesting fact is how the operations and maintenance costs for the generation 3+ reactors are higher than that for the ABWR. 50

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Supply Chain Each reactor vendor claims to be able to de liver the components and parts needed for the timely completion of construction. Each though is plagued by large lead times for the large forgings and component manuf acturing. Another problem is that the U.S. no longer has the manufacturing capabilities for la rge components so many have to be imported from overseas. Another problem is lack of sufficient number of U.S. skilled craftsmen to construct these reactors. The nuclear industry has an ageing work force and recent grads are not filling the posts being vacated by retirees.38 Hitachi has proven that it can construct the AB WR in less than 4 years and that its vendors have the capability of delivering the component s built to U.S. standards. ABWRs are currently being ordered around the world and there is confidence in the supply chain. Since the same vendors are used for the ESBWR as the ABWR; there is confiden ce in the supply chain for the ESBWR. The EPR design from AREVA is currently being constructed in Finland and the supply chain is proven. Next is the U.S.-APWR which is developed by Mitsubishi Heavy Industries. Mitsubishi has built reactors in Japan as re cently as 1997 and is capable of producing the components overseas. Mitsubishi has stated that it will import components from overseas and therefore the supply chain is verified. There is a lower confidence in the supply chain for AP 1000 since there is no history of building any sy stem that utilizes similar components. Licenseability Regulatory issues are the cause of many nuclear units to be canceled or left without being completed. Many financial institutions are not pr oviding the capital for the investment knowing that the investment may fail because the design is not at par with regulatory standards. The NRC has been diligently working on the design certifications and reviews of many of the systems and on fixing the open questions with those already certified.25 51

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Table 5-6. Regulatory Status Reactor Design Regulatory Status Projected Date AP 1000 Certified/amended under review 3/2010 ABWR Certified Certified ESBWR in certification process No Date APWR in certification process No Date EPR in certification process 5/2011 The ABWR has the greatest advantage in th e licenseability since it is certified and constructed. The AP 1000 is the next reactor. Even though the AP 1000 design is certified, it is now being amended and re-reviewed as seen in tabl e 9. Next is the EPR with a projected year of 2011. The ESBWR and APWR do not have a project ed date. GE is currently resubmitting an updated schedule for submission of revision 5 of the DCD. The APWR is currently waiting for an acceptance of docketing and schedule for review from the NRC. The order of design status depends on the NRC projected dates. Security Security at a reactor site i nvolves the reduction of threats fr om outside sources. There are many threats that can have a negative impact on control and safety at a nuclear site. September 11 really opened the eyes of th e public to the threat of airplanes crashing into buildings. The threat may come from terrorists or even a malfun ction in the system, neve r the less it is still a danger however improbable. The NRC has recently ruled that the new reactor design has to withstand a large plane crash and be able to protect the health and safety of the public in such an event. Another concern is a major fire that ma y engulf the spent fuel pit. Like many industrial sites there are many hazards that can a ffect the security of the equipment. The EPR has a distinct advantage over the ot her reactors because it has two containment buildings that are 1.3 meters thick each. Each on e of the four trains is located in different buildings or partitions and on different sides of the plant, so if one side is hit the other trains on 52

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the opposite sides are available. The ESBWR and ABWR are equa l because they both have a primary and secondary containment. The firs t containment building is the immediate area surrounding the reactor including some of the pool s needed for safety cooling. The secondary containment is the reactor building itself whic h envelops the primary containment. The next reactor is the APWR. The APWR has a containm ent vessel very similar to current reactor designs. The vessel uses buttresses to tie the horizontal tension wires and vertical tension wires fixed in the tendon gallery at the base. The buildi ng is made of pre-stressed concrete. Lastly, the AP 1000s primary containment building is made of steel not a concrete structure. The shield building surrounding containment has holes along the sides to allow for passive containment cooling. These holes are a point of concern, be cause even though improbable there exists the possibility of fuel getting into these holes and catching fire. Risk and Uncertainty There are many risks and uncertainties associat ed with building a nuclear reactor in the U.S. In the past regulatory issues caused the cancelation of many nuclear plant orders and the stagnation of the nuclear industry. When bu ilding new reactors many investors feel the investment risks are large and are hesitant to prov ide the capital without a ssurances. All first of a kind designs have uncertainties associated with them. All FOAK units have unforeseen issues that may arise during operation a nd or construction that may cause the reactors not to operate or operate with substantial increases in cost. The large amounts of capital inve stment requires that the plants be built within the allotted time, if not interest rates will cause another downfall of the industry. Regulatory bodies have also not seen these FOAK react ors operate and heavy scrutiny when examining the designs is used. Regulatory delays also cause cost overruns and lack of confidence in the NRCs ability to provide timely review of the r eactors. The risks and uncertainties are caused by differe nt reasons but outside factors causing catastrophic events for 53

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the industry and financial institutions are possible. One uncerta inty worthy of mention has to do with the security rule the NRC approved requi ring the containment building to be able to withstand an airplane crash. The AP 1000 which is an approved design will have to relook at the containment structure and determine what the corr ective actions are so th at the reactor can be built. Issues like these will cause problems during COL applications and construction. Site Specific Issues Reactor type selection has many factors as described above that can go into the decision making process. Another topic that must be cons idered is site location, and logistics. Utility expertise is also a factor. A utility which operates only PWRs may not want to select a BWR because all the engineering analysis and training techniques are ba sed on PWR information. Another issue along the same lines is if you have a PWR located at the selected s ite selecting a PWR again for the same reasons is preferable. Another choice a utility must look at is MWe needed. Each reactor has a different power rati ng as shown in Table 5-7 Electric Power. Table 5-7. Electric Power Reactor Design Power (MWe) AP 1000 1154 ABWR 1325 ESBWR 1535 APWR 1700 EPR 1600 If a utility needs less power in an area due to lower demand it would not necessarily want to buy an APWR or EPR. Another factor that must be considered is the actual site location. Nuclear power plants require large components and the components ar e transported to the site. One particular component is the reactor pressure vessel. The la rgest vessel is that of the ESBWR. The fuel height was shortened creating a n eed for more assemblies. The chimney is designed to help the 54

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55 natural circulation, creating a need for a larger vessel. On the other hand the AP 1000 has the smallest reactor pressure vessel making it easier to transport. Fuel for the emergency equipment is another transportation requireme nt. Diesel fuel is needed for the emergency diesels and natural gas for the APWRs gas turbines. A pipeline or trucking ca pabilities are a requi rement so that the fuel source can be used. Is locating the facility near an ocean or major water source a viability allowing for the use of barge transportation? Nuclear plants also require large amounts of water for cool ing. A question that must be analyzed is where will this water be supplied from a lake or the ocean?

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CHAPTER 6 CONCLUSION This study reviews and compares the new reactor designs in reliability, technology maturity, cost, supply chains, licenseability, O&M cost, safety, and security. The study also brings out risks, uncertainties, and site specifi c issues. The new reactors studied are the AP 1000, ABWR, ESBWR, APWR, and the EPR. The study al so describes new regulations in place for licensing reactors. These regulations are in pl ace to stream line the reactor operation while providing adequate review processes. The AP 1000 is the Westinghouse version of a generation 3+ reactor. It currently has resubmitted the design review for an amendment. It is a passive system and therefore needs no power to mitigate accidents. This is accomplished by the use of accumulators and gravity. The ABWR is a generation 3 reactor which us es active safety features during accident scenarios. The ABWR seems to be the most de ployable reactor design mainly due to technology maturity, cost, and supply chains. All of these f actors are demonstrated by the construction of ABWRs in Japan and Taiwan. The ABWR is a proven technology and has been certified by the NRC. The ESBWR uses natural circulation and has no pumps to remove heat from the core. The ESBWR also uses passive safety features eliminati ng the need for safety grade diesels to cool the core during an accident. The main drawback in ESBWR is the lack of technology maturity. APWR is an active design system that is currently under the NRC design review. It has annulus surrounding containment that is kept at a slight negative pr essure to keep the radioactive material from being released in an accident. In accident scenarios power to the safety systems can be provided through the safety -related gas turbine generators. 56

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57 The EPR is also under rge design review by the NRC. It is a 4 loop PWR. The main characterizing features of this system are the containment building design and the 4 trains of safety features. In conclusion it is quite evident that all ne w reactor systems feature improved safety and reliability over the current generation of US nuclear power plants. A utility has many good technology choices. Deciding which reactor is th e best is dependent upon many factors; one factor that is missing is the plant operational data. For a truly fair comparison operational data must be used and the data is not available. The only reactor with ope rating experience is the ABWR. All other reactor systems are in the earl y stage of engineering design with a long way before the operational data will become available. Based on what we know at this point, it is not possible to make a definitive conclusion on the overall relative merits of these reactor systems.

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LIST OF REFERENCES 1. US NRC. Power Reactors. [Online] Febrary 12, 2007. [Cited: April 11, 2008.] http://www.nrc.gov/reactors/power.html 2. Glastone, Samuel and Jordan, Walter H. Nuclear Power and its Environmental effects. La Grange Park, Illinois : Am erican Nuclear Society, 1980. 3. Stacey, Weston M. Nuclear Reactor Physics. s.l. : John Wiley and Sons, Inc., 2001. 4. Fretwell, Leah. Reator Basics. The Nuclear History Site. [Online] 2006. [Cited: July 21, 2007.] http://www.nuclear-history.org/reactor.html 5. Cochran, Robert G. and Tsoulfanidis, Nichol as. The Nuclear Fuel Cycle: Analysis and Management. La Grange Park, Illinois : American Nuclear Society, 1990. 6. US NRC. Early Site Permits Licensing Re views. [Online] Febrary 11, 2008. [Cited: April 11, 2008.] http://www.nrc.gov/reactors /new-licensing/esp.html. 7. US NRC. Expected New Nuclear Plant Applications. [Online] April 8, 2008. [Cited: April 11, 2008.] http://www.nrc.gov/reactors/new-licensing/new-lic ensing-files/expectednew-rx-applications.pdf 8. NEI. The Changing Climate for Nuclear En ergy. Annual Briefing for the Financial Community. February 2007, New York, New York [Online] 2007. [Cited: July 21, 2007.] http://nei.org/filefolder/t hechangingclimatefornuclearen ergy-wallstreetbriefing2007.pdf 9. U.S. NRC. 10 Code of Federal Regulations Part 52Early Site Permits; Standard Design Certifications; and Combined Licenses for Nuclear Power Plants. NRC. [Online] April 23, 2007. [Cited: April 24, 2007.] http://www.nrc.gov/reading-rm/doccollections/cfr/part052/full-text.html 10. Matthews, David. Status of New Reactor Li censing Activities. s.l. : NRC Division of New Reactor Licensing, 2006. 11. Westinghouse Electric Company, LLC. Desi gn Control Document AP 1000. Pittsburgh, PA : s.n., 2005. 15. [Online] April 14, 2008. [Cited: April 15, 2008.] http://www.nrc.gov/reactors/new -licensing/design-cert/ap1000.html 12. Westinghouse Electric Company, LLC. AP 1000 Safe Simple Innovative. Pittsburgh, PA : s.n., 2006. http://www.westinghousenuclear.com/docs/AP1000_brochure.pdf 13. Cummins W. E., Corletti, M.M., Schulz T.L. Westinghouse AP1000 Advanced Passive Plant. Proceedings of ICAPP Cordoba, Spain : s.n., 2003. Paper 3235. 14. Westinghouse Electric Comp any, LLC. AP1000. [Online] 2006. [Cited: Apri 22, 2007.] http://ap1000.westinghouse nuclear.com/ap1000_ec.html 58

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15. A Roadmap to Deploy New Nucl ear Power Palnts in the United States by 2010 Volume II. s.l. : United States Department of Energy, 2001. 16. GE, Hitachi, Toshiba. Design Control Document ABWR. [Online] March 20, 2008. [Cited: March 30, 2008.] http://www.nrc.gov/reactors/new-licensing/designcert/abwr.html. 17. GE, Hitachi, Toshiba. ABWR General Plan t Description. [Online] June 2000. [Cited: September 21, 2007.] http://www.ftj.agh.edu.pl/~cetnar/ABWR/Advanced %20BWR%20General%20Description.pdf 18. GE. New Reactors. [Online] 2007. [Cited: April 23, 2007.] http://www.gepower.com /prod_serv/products/nuclear_ener gy/en/downloads/gea14576e_abwr.pdf 19. CSNI Workshop on PSA Applications and Li mitations, NEA/CSNI/R (91) 2, Santa Fe, New Mexico, September 4-6, 1990. 20. Severe Accident Risks: An Assessment fo r Five U.S. Nuclear Power Plant, NUREG 1150, Vol. 3, 1989 (released 1991). 21. GE Energy. ESBWR General Description. 2006. [Online] June 2006. [Cited: December 30, 2007.] http://www.ne.doe.gov/np2010/pdfs/esbwrGenera%20DescriptionR4.pdf 22. GE Energy. ESBWR Design Control Docume nt. 2007. [Online] March 20, 2008. [Cited: March 30, 2008.] http://adamswebsearch2.nrc.gov/id mws/ViewDocByAccession.asp? AccessionNumber=ML052450245 23. GE Energy. Natural Circulation in ESBW R. [Online] 2007. [Cited: April 25, 2007.]. http://www.gepower.com/prod_serv/products/nuc lear_energy/en/downl oads/natural_circ ulation_esbwr.pdf 24. Hinds, David and Maslak, Chris. Next-generation nuclear energy: The ESBWR. Nuclear News. January, 2006. 25. U.S. NRC. Design Certifications Licen sing Reviews. [Online] February 11, 2008. [Cited: February 11, 2008.] http://www.nrc.gov/reactors/new -licensing/design-cert.html 26. GE Energy. ESBWR Fact Sheet. [O nline] 2006. [Cited: April 25, 2007.]. http://www.gepower.com/prod_serv/products /nuclear_energy/en/downloads/gea14429g_ esbwr.pdf 27. Mitsubushi Heay Industries, LTD. US-APWR General Description. 2006. 28. Mitsubishi Heavy Industries, LTD. Mitsubishi Advanced PWR Plant. [Online] 2001. [Cited: April 22, 2007.] http://www.mhi.co.jp/atom/hq/atome_e/apwr/index.html 59

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60 29. Miyake, Yoshio and Mukai, Hiroshi. For the Long-Term Stable Supply of Electric Energy Technical Review. s.l. : Mitsubishi Heavy Industry, Ltd, 2003. Vol. 40, 1. 30. AREVA. EPR. 2007. [Online] Feburary 2007. [Cited: April 23, 2007.] http://www.arevanp.com/us/liblocal/docs/EPR/U.S.EPRbrochure_1.07_FINAL.pdf 31. AREVA. EPR: the first-built generation III+ reactor. [Online] April 2007. [Cited: April 23, 2007.] http://www.areva-np.com/scripts/info/pub ligen/content/ templates/show.asp? P=1659&L=US 32. Unistar Nuclear Energy. Unistar Nuclear Energy. [Online] Unistar Nuclear Energy, LLC, 2007. [Cited: June 25, 2007.] http://unistarnuclear.com/ 33. Holm, Jerald S. Pre-Application Review of the EPR. s.l. : Areva, 2005. 34. International Atomic Energy Agency. Design of Reactor Core for Nuclear Power Plants. Safety Guide No. NS-G-1.12. 2005. 35. Marshall, Frances and Wierman, Thomas and Rasmuson, Dale and Mosleh Ali. Insights about Emergency Diesel Generator Failures for the US NRCs Common Cause Failure Database. 1999. 36. US NRC. 10 CFR Appendix I to Part 50 [Online] Feburary 12, 2007. [Cited: April 23, 2007.] http://www.nrc.gov/reading-rm/doc-colle ctions/cfr/part050/part050-appi.html 37. Dominion Energy Inc. Bechtel Power Cor poration, TLG, Inc., MPR Associates. Study of Construction Technologies and Schedules O&M Staffing and Cost, Decommissioning Costs and Funding Requirements for Advanced Reactor Designs. 2004. 38. Cohen, Bernard L. The Nuclear Energy Option. Plenum Press 1990

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BIOGRAPHICAL SKETCH Alejandro Chomat was born on April 24, 1985, in Miami, Fl. He went to high school at Christopher Columbus High School wh ich prepared him for a college career at the University of Florida. At UF, he majored in nuclear engineering for his undergraduat e degree. Alex is a member of the American Nuclear Society and the American Nuclear Societys honor society. In college, he worked as a lab assistant in the neutron activation analysis lab. During two summers, he interned for FPL at Turkey Point Nuclear Station. Throughout his internships at FPL, he was able to work in operations, maintenance, and engineering. Alejandro Chomat is currently employed by FPL at Turkey Point Nuclear Station as a nuclear systems operator. He plans to continue working at FPL after the completion of his masters. 61


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