NUREG/CP-0175

Proceedings of the 

Advisory Committee 

on Reactor Safeguards 

Workshop on Future 

Reactors 

June 4 - 5, 2001 

U.S. Nuclear Regulatory Commission 

Advisory Committee on Reactor Safeguards 

Washington, DC 20555-0001
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I
NUREG/CP-0175

Proceedings of the Advisory Committee 

on Reactor Safeguards 

Workshop on Future Reactors 

June 4 - 5,2001

Manuscript Completed: December 2001 

Date Published: December 2001 

U.S. Nuclear Regulatory Commission 

Advisory Committee on Reactor Safeguards 

Washington, DC 20555-0001
NUREG/CP-0175 has been reproduced 

from the best available copy.
ABSTRACT

This report contains the information presented at the Advisory Committee on Reactor 

Safeguards Workshop on Future Reactors held at the Nuclear Regulatory Commission 

headquarters in Rockville, Maryland, on June 4-5, 2001. Included are the subject matter 

summaries, followed by the presentation material and selected participants discussions. 

The primary purpose of the workshop was to identify the regulatory challenges associated with 

future reactor designs. A list of such challenges was developed from the workshop notes, the 

various presentations, the panel discussions and the question and answer sessions. This list 

is included in the Introduction section of this document. 

The titles of the papers and the names of the authors have been updated and may differ from 

those that appeared in the final workshop agenda.

iii

In addition to the summaries and presentation materials, these 

Proceedings contain selected discussions which were extracted from the 

workshop transcripts. Where practical, the participants were given an 

opportunity to review and edit their individual contribution. The 

discussions can be found immediately following each presentation.
CONTENTS 

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

EXECUTIVE SUMMARY .............................................................................................................................. vii 

A B B R E V IA T IO N S ....................................................................................................................................... xvii 

I. IN T R O D U C T IO N ....................................................................................................................................... 1 

II. KEYNOTEADDRESS ............................................................................................................................... 5 

1. Disciplined - Meaningful - Scrutable 

Commissioner Nils J. Diaz 

Ill. UNITED STATES DEPARTMENT OF ENERGY PRESENTATIONS 

1. Overview and Introduction to Generation IV Initiative ........................................................... 21 

W. Magwood 

2. Generation IV Goals and Roadmap Effort ............................................................................. 27 

R. Versluis 

3. Near-Term Deployment Efforts .............................................................................................. 35 

T. Miller 

4. Generation IV Concepts .............................................................................................................. 39 

R. Versluis 

5. NextStepslll and d IV ................................................................................................................... 49 

IV. GENERATION IV DESIGN CONCEPTS 

1. Pebble Bed Modular Reactor ................................................................................................ 59 

W. Sproat - Exelon 

J. Slabber - PBMR Pty. 

2. International Reactor Innovative and Secure ........................................................................ 77 

M. Carelli - Westinghouse 

3. General Atomic - Gas Turbine Modular Helium Reactor .......................................................... 101 

L. Parme - General Atomics 

4. European Simplified Boiling Water Reactor Program ............................................................ 119 

and Regulatory Challenges 

A. Rao - General Electric Nuclear Energy 

5. Advanced Liquid Metal Reactor ............................................................................................... 135 

C. Boardman - General Electric Nuclear Energy 

V. NUCLEAR REGULATORY COMMISSION PRESENTATIONS 

1. NRR Response to SRM on Evaluation of NRC Licensing Infrastructure ................................ 165 

M. Gamberoni, T. Kenyon, E. Benner, A. Rae, N. Giles - NRR 

2. Planed RES Activities ............................................................................................................... 171 

J. Flack, S. Rubin - RES 

VI. NUCLEAR ENERGY INSTITUTE 

1. Advanced Reactors Initiatives ................................................................................................. 183 

R. Simard, - NEI 

VII. TECHNICAL PRESENTATIONS 

1. Safety Goals for Future Nuclear Power Plants ........................................................................ 193 

N. Todreas - MIT

V
2. License by Test ........................................................................................................................ 207 

A. Kadak - MIT 

3. NERI Project on Risk-Inform ed Regulation .............................................................................. 237 

G. Davis - Westinghouse 

M. Golay, MIT 

4. Advanced Safety Concepts ....................................................................................................... 273 

C. Forsberg, ORNL 

5. Requlatory Fram ework for Nuclear Power Plants ..................................................................... 293 

A. Heymer - NEI 

VII. PANEL DISCUSSION 

1. Regulatory Challenges for the Licensing of 

Future Nuclear Power Plants ................................................................................................. 303 

N. Todreas - MIT 

E. Lyman - NCI 

ENDNOTES ................................................................................................................................................ 315

vi
EXECUTIVE SUMMARY

Introductory Remarks 

T. Kress, Chairman of the Advisory Committee on Reactor Safeguards (ACRS) 

Subcommittee on Future Reactors convened the meeting and introduced 

Subcommittee members in attendance, key participants, and presenters. He presented 

the planned agenda for the first day of the Subcommittee meeting/workshop and 

offered members of the public opportunities to ask questions and to provide comments 

on the matters discussed. G. Apostolakis, ACRS Chairman, introduced the keynote 

speaker, NRC Commissioner Nils J. Diaz, and provided a brief summary of his 

extensive experience in matters related to nuclear power and research and 

development of nuclear technology. 

Subcommittee Presentations 

Commissioner Diaz provided an overview of his paper entitled, "Disciplined - Meaningful 

- Scrutable." He stated nuclear power has entered the national energy debate on the 

future of America's energy supply and emphasized that nuclear safety is a priority on 

everyone's agenda. He stated that the priority should be on what should be done better 

rather than what was done wrong in the past. Commissioner Diaz stated that the 

Commission relies on the ACRS for expert advice and the recommendations of the 

Committee will be valuable to the Commission as regulatory changes are made. He 

noted that an important change to the regulatory structure has been risk-informed 

regulation which has enabled both the licensee and NRC to focus on safety issues and 

reduce unnecessary regulatory burden. He stated that the future of nuclear power is 

dependent on economic trends and events, the safety and reliability of plants, and the 

political environment. He expressed the view that it is possible to resolve safety and 

environmental issues before nuclear plants are built. An important element will be the 

readiness of the NRC for potential new plant applications but also that the NRC should 

not become an impediment to meeting the energy demands of the country. He 

reiterated that every step will need to be disciplined, meaningful, and scrutable and 

suggested that the industry and NRC will need to proceed in a disciplined and patient 

manner to ensure that errors are avoided. Commissioner Diaz qualified these 

statements as being his individual views and noted that they do not represent the views 

of his fellow Commissioners or the NRC. 

W. Magwood of the U.S. Department of Energy (DOE) led the discussions for the DOE 

staff. W. Magwood provided an overview of the Generation IV Initiative to evaluate 

candidate technology concepts for a new generation of nuclear power plants. R 

Verslius, DOE, presented the Generation IV goals, roadmap effort, and concept 

evaluation. T. Miller discussed the Near-Term Deployment Working Group (NTDG) 

formed to identify actions and evaluate options necessary for DOE to support new 

plants. DOE has established a Nuclear Energy Research Advisory Committee

vii
(NERAC) to provide independent evaluation and feedback on the establishment of 

goals and objectives and to examine progress in evaluating candidate nuclear energy 

concepts. DOE has also established a Generation IV Roadmap NERAC Subcommittee 

(GRNS) to serve as an advisory group in establishing a proposed roadmap along with a 

Roadmap Integration Team (RIT) for its implementation. Candidate technologies must 

be deployable by 2030. Nuclear systems are expected to meet sustainability goals 

(resource inputs, waste outputs, and nonproliferation), safety and reliability goals 

(operating maintainability excellence, limiting core damage risk, and reduced need for 

emergency response), and economic goals (reduced life-cycle costs and risk to capital). 

Criteria and metrics for each goal are being developed by an Evaluation Methodology 

Group (EMG), RIT, and the GRNS. DOE plans to evaluate all candidate concepts 

equally without prejudice toward existing technologies (e.g., light-water reactors) but 

recognizes that most energy generation units are likely to be fission based. DOE is 

presently considering 94 concepts. The output of the Generation IV Program is 

expected to be a research and development plan to support future commercialization of 

the best concepts. 

W. Sprout of Exelon Generation and J. Slabber of the Pebble Bed Modular Reactor 

(PBMR) Demonstration Project in the Republic of South Africa (RSA) provided a 

presentation on the safety design aspects and licensing challenges for the PBMR. The 

PBMR is a modular high-temperature gas-cooled reactor (HTGR). It is helium cooled 

and uses a graphite moderator (approximately 110 MWe). The PBMR is nearing 

completion of the preliminary design phase. The feasibility study for application in the 

United States is in preparation for investor decisions by the end of 2001. RSA 

demonstration plant construction is expected to begin in late 2002. The PBMR design 

approach is intended to employ both passive and active design features, provide 

prevention and mitigation capability, and reduce dependence on operator actions. 

Central to this approach is the spherical fuel design involving carbon-coated uranium 

oxide fuel manufactured into a fuel particle or sphere. Key technical licensing 

challenges include: lack of a gas reactor technical licensing framework; fuel qualification 

and fabrication; source term; containment performance requirements; probabilistic risk 

assessment (PRA); regulatory treatment of non-safety systems; classification of 

structures, systems, and components (SSCs); and lack of technical expertise on gas 

reactors for both the NRC and the industry. Key licensing challenges include: Price

Anderson Act indemnity, NRC operational fees, decommissioning trust funding, 

untested provisions of 10 CFR Part 52, and the potential number of exemptions that 

may be required by the NRC. 

M. Carelli of Westinghouse Science and Technology provided a presentation on the 

International Reactor Innovative and Secure (IRIS) nuclear reactor design. IRIS is a 

small to medium sized pressurized water reactor (100-300 MWe) that utilizes a 5- to 8

year option fuel cycle. The IRIS safety philosophy is "safety by design." Like current 

generation PWRs, IRIS is designed to have a reactor containment structure. However, 

Westinghouse proposes to perform scaling tests rather than loss-of-coolant accident

viii
(LOCA) analysis. IRIS is scheduled for initial deployment in 2010-2015. 

L. Parme of General Atomics (GA) provided a presentation on the GA Gas Turbine 

Modular Helium Reactor. He discussed the history of GA as a pioneer of gas reactor 

technology and noted that the proposed GA design is similar to the PBMR in its use of 

ceramic carbon-coated spherical fuel. The fuel is passive by design in that the fission 

products are retained in the coated particles or spheres. Worst-case fuel temperature 

is limited by low-power density, low thermal rating per module, use of an annular core 

design, and passive heat removal. GA proposes to apply a risk-informed approach to 

licensing using performance assessment methods. 

A. Rao of GE Nuclear Energy provided a presentation on the Evolutionary Simplified 

Boiling Water Reactor (ESBWR). The ESBWR is a 1380 MWe boiling water reactor 

with improved operating safety margins and passive safety systems. He stated that the 

ESBWR derived from earlier GE plant design certification efforts and is the result of 

eight years of international cooperative work. He stated that the biggest challenge is to 

cross the regulatory hurdles associated with the inspections, tests, analyses, and 

acceptance criteria (ITAAC) and combined license (COL) programs. He further stated 

that he did not know how long it might take to license the ESBWR, in part, because the 

last GE design certification took about 8 to 10 years. Dr. Rao also provided a brief 

overview of the GE Nuclear Advance Liquid Metal S-PRISM design. 

M. Gamberoni, NRR, led the discussion for the NRC staff. N. Gilles, NRR discussed 

the future licensing organization and inspection readiness assessment (FLIRA). T. 

Kenyon, NRR, discussed early site permits (ESPs), ITAAC and COL programs. A. Rae 

discussed the Westinghouse AP1000 review and E. Benner, NRR, discussed issues 

related to the regulatory infrastructure. J. Wilson, NRR, also participated. J. Flack and 

S. Rubin, RES, provided a brief discussion on research activities in support of possible 

future plants. The staff stated that an assessment of licensing and inspection 

readiness is ongoing and is scheduled to be completed by September 28, 2001. The 

staff is working to develop lessons-learned from past design certifications, preparing 

guidance on ESPs, and responding to the Nuclear Energy Institute (NEI) petition for 

rulemaking to 10 CFR Part 52. The staff is reevaluating its ITAAC/COL programs. 

Short-term plans are to address existing regulations, license conditions, and 

exemptions. Long-term actions are expected to addressed via rulemaking. The staff 

stated that there is a limit on how far they can pursue these initiatives and/or allocate 

resources without formal submittals by licensees and industry organizations. 

Subcommittee Questions/Comments on Presentations 

Significant points raised by members of the Subcommittee during the presentations 

include: 

G. Apostolakis questioned what DOE representatives considered to be the two most

ix
important regulatory challenges facing the NRC in licensing new reactors. DOE 

representatives stated that the key challenges will be related to making the regulatory 

environment as risk-informed and performance-based as practicable. DOE 

representatives stated that the NRC process must be predictable in both its review time 

and its decisions. D. Powers questioned the extent to which performance indicators 

(PIs) might further performance-based considerations. G. Apostolakis suggested that 

reliability goals be numerical. DOE representatives stated that the it is difficult to place 

goals on PIs or reliability without knowing more about the detailed designs. 

T. Kress and D. Powers questioned the nature of fuel performance for the PBMR. T. 

Kress questioned how fuel manufacturing quality and integrity will be ensured. D. 

Powers questioned how friction, ramp rates, and other operating characteristics would 

be addressed considering the fact that there was limited operating experience for this 

type of fuel. Exelon and RSA representatives stated that fuel would be subjected to 

extensive quality assurance and quality control requirements during fabrication and that 

operating performance would be monitored using gamma spectroscopy for each of the 

212,000 fuel spheres cycled through the core. 

G. Apostolakis and J. Garrick questioned how the Commission's Safety Goal Policy 

Statement would be considered for the PBMR. They noted that Safety Goal's use of 

core damage frequency (CDF) might be challenged if applied to the collective 

population of modular units at reactor sites across the country. Exelon and RSA 

acknowledged that this is an issue to be addressed in characterizing the risk metrics. 

They noted that the modular approach to siting will have substantial licensing expense 

ramifications as well (i.e., licensing fees per reactor). 

T. Kress questioned the PBMR and GA Gas Turbine - Modular Helium Reactor 

proposals to limit or eliminate the use of primary containment structures and reducing 

emergency planning zones. He questioned the prudency of this given that the 

uncertainties that have not been quantified. He also noted that Chernobyl had a 

graphite core and it burned. D. Powers noted that there is a substantial difference 

between point-ignition and diffuse-ignition of core materials and that one of the largest 

catalysts in fuel performance is cesium. The GA representative stated that the fuel will 

not burn in the normal sense of a chain reaction and that most analyzed failures have 

been associated with fuel oxidation. He also stated that the MHTGR has circulators 

designed to reduce temperature. 

Panel Discussion 

The Subcommittee and participants extensively discussed the use of risk information in 

considering future nuclear plants. G. Apostolakis stated that there seems to be a gap 

between the staff and industry thinking concerning the importance of risk assessment. 

He stated that he is not sure that there is a full appreciation how important risk 

assessment is in the design, licensing, and operation of nuclear power plants. M.

x
Bonaca stated that there seems to be a perception that risk is a regulatory constraint 

rather than a safety benefit. The staff stated that the Commission has been very clear 

in directing the staff to use risk analysis in deciding what information and analysis is 

needed. The staff also stated that more confidence is needed than just demonstrating 

that the Commission's Safety Goals are met. 

J. Garrick expressed concern that an important opportunity was being missed in the 

rush to license new reactors. He stated that there could not be a better time to consider 

risk. D. Powers stated that there is not much risk information available concerning the 

proposed plants designs and suggested that the NRC will need to perform confirmatory 

analyses to ensure that vulnerabilities have not been missed. He also stated that the 

staff will need to perform tests (e.g., to ensure that particle-type fuel does not burn) and 

testing programs to ensure that actual operating performance reflects design 

characteristics and to validate thermal-hydraulic modeling and component performance. 

The staff stated that 10 CFR Part 52 requires licensees to conduct PRAs. Exelon 

representatives stated that existing bodies of data must be utilized and that they must 

pursue a COL first, rather than design certification, based on the RSA Demonstration 

Project. 

P. Ford noted that the presentations involved little discussion of material degradation, 

embrittlement, or cracking. Industry representatives stated that materials were not a 

top priority at this early stage. They stated that their focus was on design first with 

consideration of materials later. The staff stated that the Commission expects these 

designs to be safer than the current generation of plants and that issues such as 

pressurized thermal shock (PTS) will certainly be addressed. 

T. Kress questioned how defense in depth will be considered in new plant designs. 

Commissioner Diaz offered his views on the importance of considering defense in 

depth in the design stage of reactors. G. Apostolakis stated that he was encouraged br 

recent government-wide initiatives to consider both risk information and defense in 

depth. He expressed concern, however, over the argument that PRA might be viewed 

as a major challenge if it makes plants uneconomical. He stated that risk analysis is 

necessary to reduce the uncertainty in new and untested designs. 

June 5. 2001 

Introductory Remarks 

T. Kress, Chairman of the ACRS Subcommittee on Future Reactors convened the 

meeting and introduced Subcommittee members in attendance, key participants, and 

presenters. He presented the planned agenda for the second day of the Subcommittee 

meeting/workshop and offered members of the public opportunities to ask questions 

and to provide comments on the matters discussed.

Xi
Subcommittee Presentations

R. Simard of the Nuclear Energy Institute (NEI) provided a brief presentation on the 

state of energy demand in the United States and discussed the improving economics 

for new nuclear power plants. He discussed the consolidation of companies under 

deregulation and suggested that these larger companies will be better able to undertake 

large capital projects such as nuclear power plant construction. He discussed efforts 

under way to support a new generation of plants but noted that there needs to be 

greater certainty in the licensing process. He discussed infrastructure challenges in 

terms of people, hardware, and services to support new and current plants. He stated 

that there needs to be fair and equitable licensing fees and decommissioning funding 

assurance for innovative modular designs such as the PBMR. He concluded that NRC 

challenges will include resolving 10 CFR Part 52 implementation issues; establishing an 

efficient and predictable process for siting, COL permits and inspection; and an 

increasing regulatory workload. 

N. Todreas of the Massachusetts Institute of Technology (MIT) provided a discussion 

on safety goals for future nuclear power plants. He stated that this effort is focused 

solely on future power plants and not the current NRC Safety Goals and associated 

quantitative health objectives that use core damage frequency (CDF) and large early 

release frequency (LERF) as surrogate measures. This work is being sponsored by 

DOE for Generation IV Initiative technology goals. These goals are being developed for 

systems to be deployed from 2011 to 2030. They are intended to guide in making 

trade-offs in the evaluation of candidate technologies. The goals will partition the 

systems according to categories of sustainability, safety and reliability, and economics. 

The outcome is expected to a framework that encourages fundamental design 

directions that promote safety. 

A. Kadak of MIT presented an approach to licensing Generation IV technologies 

entitled "License by Test." He stated that the major challenges for new reactors are 

driven by a regulatory framework that generally supports light-water-reactor technology. 

He stated that both licensees and the NRC staff lack sufficient knowledge in non-light

water reactor technologies and that the regulatory system is overly rigid in adjusting to 

change. He suggested that the NRC adopt a risk-informed approach to licensing 

whereby a safety basis would be established using risk-based techniques to identify 

dominant accident sequences and systems and components, establishing confidence 

levels to bridge deterministic and probabilistic approaches, and implementing a license 

by test approach using a full-size demonstration plant. Successful demonstration would 

provide the basis for reducing uncertainty and for certifying the design. Traditional 

performance tests would still be required to demonstrate reliability. However, license by 

test would serve to validate analyses, shorten time for paper reviews, and demonstrate 

safety. He suggested that the PBMR be used as the prototype for this licensing 

approach.

xii
M.Golay of MIT and G. Davis of Westinghouse provided a presentation on a Nuclear 

Energy Research Initiative (NERI) Project sponsored by DOE. The focus of the NERI 

Project is to take future plant designs and use risk information to evaluate what new 

design and regulatory processes must be developed to support new plant license 

applications for Generation IV concepts. M. Golay stated that there is a need to 

improve the regulatory process and suggested that the overall national effort in support 

for reactors suggests that there is a need for change. These activities are being 

coordinated with NEI which will be initiating the industry-sponsored development of new 

regulations. NERI will address the overall risk-informed design and regulatory process. 

Sandia National Laboratories (SNL) is also providing technical support. 

C. Forsberg of Oak Ridge National Laboratory provided a presentation on the economy 

of nuclear-generated hydrogen production. He stated that there is enormous need for 

increased hydrogen production to support the U.S. chemical industry (oil refineries) 

which uses 5% of all the natural gas consumed in this country. He stated that the major 

reason for the need is increased use of more abundant heavy-sour crude oils which 

require more energy to process than the more scarce light-sweet crude oil. He noted 

that non-light-water reactors (e.g., molten salts) are better suited for this type of 

application and suggested that an advanced high-temperature reactor (AHTR) could 

provide dual-purpose electric generation and hydrogen production. This is a joint DOE 

effort with Sandia National Laboratories (SNL). 

A. Heymer of NEI provided a brief discussion on the benefits of establishing a new 

regulatory framework. He suggested that a new paradigm in regulatory thinking is 

needed and stated that the reactor oversight process (ROP) serves as the appropriate 

basis for starting these discussions. He suggested that the ROP cornerstones of safety 

be used as the starting point for developing a new set of General Design Criteria (10 

CFR Part 50, Appendix A). He suggested that new operating criteria, generic risk

informed and performance-based regulations be developed with associated design

specific and regulation-specific regulatory guides. 

Subcommittee Questions/Comments on Presentations 

Significant points raised by members of the Subcommittee during the presentations 

include: 

D. Powers questioned the NEI contention that DOE energy demand estimates are 

consistently low. He stated that the critics have argued that efficiency and conservation 

can do the job. R. Simard agreed that efficiency and conservation play an important 

role but concluded that it is unrealistic to suggest that new electricity generation is not 

needed. 

D. Powers expressed appreciation for the systems-approach and use of trade-off 

studies in evaluating new plant designs and safety goals. N. Todreas stated that the

xiii
goal is to stimulate innovation and not to go back to existing reactors as the standard 

for the future. He stated that they are looking at a balance of utilization in terms of 

whole fuel cycle, e.g., economics, waste, diversion, etc. 

D. Powers questioned why the safety goals could not be expressed in terms of release 

of radioactivity. G. Wallis expressed concern that this approach might overly constrain 

the evaluation of certain designs and lock the evaluation into certain design directions. 

J. Garrick stated that the evaluation should not focus too heavily on fission products as 

the actinides drive much of the risk in high-level waste. G. Apostolakis suggested that 

safety and reliability can also be expressed in terms of investment protection. He noted 

that serious plant damage can occur without having releases and suggested that it may 

be worthwhile to distinguish between technology goals and safety goals. G. Wallis 

suggested that life-cycle costs also be expressed in terms of external costs in 

comparing candidate nuclear technologies with alternate fuels, e.g., adverse effects of 

fossil fuels killing fish in New England via acid rain. 

G. Wallis questioned how human performance would be evaluated using the "license by 

test" approach. G. Leitch stated that the major advantage of license by test appears to 

be a reduction in the time and costs for paper reviews associated with the licensing 

process and questioned what technical merits would be derived. J. Sieber questioned 

who should finance the costs of such a facility. A. Kadak stated that a containment 

should be constructed on the PBMR Demonstration Project only for the purpose of 

demonstrating safety and suggested that operators be allowed to take non-conservative 

actions to test the robustness of the design. A Kadak stated that the PBMR 

Demonstration Project should be a legitimate government expense (i.e., DOE) as it is 

still a concept, and the plant has not yet been designed. He stated that much work 

needs to be done to develop the models and codes necessary to validate the design. 

G. Apostolakis questioned whether the licensing process can be made performance

based. A. Heymer of NEI stated that the inspection process can be made performance

based as evidenced by the reactor oversight process (ROP). He also noted that certain 

regulations can be made more performance-based (e.g. 10 CFR Part 20). A. Heymer 

suggested that risk-informing 10 CFR Part 52 will be very important for new reactors. 

G. Apostolakis stated that the ROP is an evolution of the existing regulatory system and 

suggested that the risk for new reactors may be different thereby requiring a different 

approach. He noted that NEI does not normally want to depart too substantially from 

the existing regulatory structure. 

Panel Discussion 

R. Barrett, NRR, offered a four-pillar approach to licensing new nuclear power plants. 

He stated that success will be based on assuring safety, streamlining the organization 

to be efficient and effective, not imposing unnecessary regulatory burden, and 

maintaining public confidence. G. Wallis stated that it is not good enough to provide

xiv
public access to NRC decisionmaking. R. Barrett agreed and stated that they need to 

identify public concerns and act on them. 

N. Todreas of MIT provided a brief presentation on regulatory challenges mostly related 

to fuel and clad materials. He stated that longer operating cycles and higher operating 

temperatures will result in challenges related to waste toxicity and volume, corrosion 

control of coolant impurities, qualification of fuel particles or spheres, and new 

maintenance practices to support longer operating cycles. T. Kress suggested that new 

reactor licensing may be somewhat like digital instrumentation and control in that the 

NRC controls the process and not the product. J. Garrick stated that the regulatory 

process, like people, are slow to change. 

E. Lyman of the Nuclear Control Institute (NCI) provided a presentation that focused on 

the role of government in energy matters. He stated that public money should not be 

spent as a taxpayer subsidy for utilities. He stated that the performance data on PBMR 

fuel is "spotty" and that the German graphs illustrating the 10% release fraction of Cs

137 were flawed. He also stated that British Nuclear Fuels falsified fuel performance 

data sent to Japan on this matter. E. Lyman suggested that the NRC establish an 

ITAAC for PBMR fuel manufacture and acceptance. He questioned how the Chernobyl 

event could not happen at a PBMR and suggested that ignition fuel temperatures could 

be achieved through sabotage. He stated that the Commission's Safety Goals are not 

conservative enough and concluded that there is no technical basis for relaxing 

containment and emergency preparedness requirements. He noted that about half of 

the U.S. nuclear plants failed the NRC Operational Safeguards Response Evaluation 

(OSRE) safeguards inspection. 

W. Hauter of Public Citizen provided a brief presentation concerning the state of energy 

deregulation and the need for new nuclear power plants. She stated that the demand 

for and acceptance of nuclear power is being painted as a "rosy picture" based on a 

recent poll in California. She stated that 58% of the public disapprove of President 

Bush's energy plan and the public always supports renewable energy as the first option. 

She suggested that the apparent energy crisis is being misrepresented in order to 

justify using taxpayer money to subsidize a resurgence of nuclear power and the 

associated research and development costs for new reactors. She questioned the 

safety of "merchant" nuclear plants and expressed concern that the recent work on 

health effects is being conducted with the improper intent of reducing the waste 

classification of certain radiological materials. W. Hauter suggested that licensing is 

being used as a new code word for deregulation. She stated that the biggest challenge 

is the issue of subsidies to the utilities and questioned the theme of the Subcommittee 

meeting/workshop as being biased toward further deregulation that favors getting new 

plants licensed. T. Kress and G. Wallis expressed concern over the lack of public 

interest in ACRS meetings and questioned how to get the public more involved in 

providing broader perspective. W. Hauter suggested that meetings be held around the 

country outside normal business hours (i.e., in the evening) so that interested parties

XV
could more conveniently attend after work.

Expected Subcommittee Action 

At the conclusion of the meeting, T. Kress stated that the purpose of this meeting was 

to explore the regulatory challenges associated with future nuclear power plants and for 

the Subcommittee to examine technical issues for the ACRS to consider in evaluating 

the safety of candidate reactor designs and applications. The Subcommittee plans to 

continue its discussion of these matters during future meetings. 

Summary prepared by Michael T. Markley, ACRS Senior Staff Engineer.

xvi
ABBREVIATIONS

ACRS 

ABWR 

ACRS 

AHTR 

ALARA 

ALWR 

ATWS 

BWR 

CBD 

CFR 

COMNJD 

COL 

DBA 

DBE 

DOE 

EMG 

EPZ 

ESP 

ESBWR 

FLIRA 

FP

GIF 

GCFR 

GDC 

GDCS 

GRNS 

IC

IRIS 

IAEA 

LBT 

LMFBR 

LOCA 

LEU 

LWRs 

MIT 

MLD 

MOX 

MHTGR 

NERAC 

NRC 

NRR 

NTDG 

NEI -

xvii

Advisory Committee on Reactor Safeguards 

Advanced Boiling Water Reactor 

Advisory Committee on Reactor Safeguards 

Advanced High-Temperature Reactor 

As Low As Reasonably Achievable 

Advanced Light Water Reactor 

Anticipated Transients Without Scram 

Boiling Water Reactor 

Commerce Business Daily 

Code of Federal Regulations 

Commissioner Niles J. Diaz 

Combined Operating License 

Design Basis Accident 

Design Basis Event 

Department of Energy 

Evaluation Methodology Group 

Emergency Planning Zone 

Early Site Permit 

European Simplified Boiling Water Reactor 

Future Licensing and Inspection Readiness Assessment 

Fuel Particles 

Generation IV International Forum 

Gas-Cooled Fast Reactor 

General Design Criteria 

Gravity Driven Cooling System 

Generation IV Roadmap NERAC Subcommittee 

Isolation Condenser 

International Reactor Innovative and Secure 

International Atomic Energy Agency 

License By Test 

Liquid Metal Fast Breeder Reactor 

Loss of Coolant Accident 

Low Enriched Uranium 

Light Water Reactors 

Massachusetts Institute of Technology 

Master Logic Diagram 

Mixed-Oxide Fuel 

Modular High-Temperature Gas Reactor 

Nuclear Energy Research Advisory Committee 

Nuclear Regulatory Commission 

Office of Nuclear Regulatory Regulation (NRC) 

Near-Term Deployment Group 

Nuclear Energy Institute
OECD - Organization For Economic Cooperation Development 

ORNL - Oak Ridge National Laboratory 

PBMR - Pebble Bed Modular Reactor 

PCCS - Passive Containment Cooling System 

PRA - Probabilistc Risk Assessment 

RAI - Request for Additional Information 

R&D - Research and Development 

RES - Office of Nuclear Regulatory Research (NRC) 

RFI - Request for Information 

RIT - Roadmap Integration Team 

RCCA - Rod Cluster Control Assembly 

RDAS - Risk Dominant Accident Sequence 

RSA - Republic of South Africa 

SAR - Safety Analysis Report 

SBWR - Simplified Boiling Water Reactor 

SG - Steam Generator 

SSC - Systems, Structure and Component 

T/H - Thermal Hydraulic 

TWG - Technical Working Group 

WANO - World Association of Nuclear Power Operators

xviii
INTRODUCTION

Because of the large amount of regulatory activity that is anticipated for licensing future 

reactor concepts, the ACRS decided to hold this workshop on "Regulatory Challenges 

for Future Reactor Designs." The workshop was held primarily for the benefit of the 

Committee - to acquaint the members with the various design concepts and to identify 

potential regulatory and policy issues for which ACRS may be called upon to give 

advice to the Commission. It was also believed that the workshop would be of benefit 

to the NRC staff as well as to the industry in getting an early dialogue started on the 

possible regulatory approaches to licensing future reactor designs. These designs are 

expected to be significantly different from the LWRs which are the primary focus of the 

current regulations and regulatory system. 

The primary purpose of the workshop, as indicated by its title, was to identify the 

regulatory challenges. A list of such challenges identified by the workshop was 

developed from the workshop notes, the various presentations, the panel discussions, 

and the question and answer sessions. 

The concept of defense in depth and its application to future reactor designs is of great 

interest to the Committee. The traditional interpretation of this concept asserts that 

defense in depth is embodied in the structure of the regulations and in the design of the 

facilities built to comply with those regulations. The requirements for defense in depth 

are derived by repeated application of the question, "What if this barrier or safety 

feature fails?" This is the structuralist interpretation of defense in depth1 . It is the 

cornerstone of traditional regulations. In recent years, the maturity of Probabilistic Risk 

Assessment (PRA) methodology has led to the rationalist interpretation. This 

interpretation asserts that defense in depth is the aggregate of provisions made to 

compensate for uncertainty and incompleteness in our knowledge of accident initiation 

and progression. The successful implementation of this interpretation of defense in 

depth requires robust PRAs and risk acceptance criteria. In this context, the Committee 

was interested in the following questions: 

1. Should we "force-fit" non-LWR designs into the current regulatory structure that 

is heavily focused on LWRs and the structuralist approach to defense in depth or 

should licensing take a "clean-sheet" risk-based approach? 

2. Do we need additional risk acceptance criteria (e.g., frequency-consequence 

curves) for designs for which core damage frequency and large, early release 

frequency are ill-posed concepts? 

3. How do we quantify and deal with PRA uncertainty for the new concepts? 

'J. N. Sorensen, G. E. Apostolakis, T. S. Kress, and D. A. Powers, "On the Role of Defense in Depth in 

Risk-Informed Regulation," Proceedings of PSA '99, International Topical Meeting on Probabilistic Safety 

Assessment, Washington, DC, August 22 - 26, 1999, American Nuclear Society, La Grange Park, Illinois.

1
4. How robust must the containment (if any) be in view of the lack of experience with new designs and potentially large uncertainty in risk assessments? 

5. What is the acceptability of significantly reduced emergency response for 

designs with low source terms and long warning times before release? 

6. What is the acceptability of no additional ECCS for designs that have a great 

deal of water in the primary vessel and no pipes to break at locations that would 

drain the water (e.g., IRIS)? 

7. What role will "licensing by test" play in the regulatory process? 

8. Can the frequency of air ingress and a graphite/fuel fire be demonstrated to be 

acceptably low for those designs that use coated particle fuel? 

9. How de we establish LOCA frequencies for new plants for which there is not an 

extensive data base? The reliability of simplified passive safety systems? 

10. For the new designs, will we have appropriately qualified/validated PRAs and 

T/H, neutronic, and safety assessment codes? 

11. Are there any new human performance and I&C issues? 

In addition to these questions, the Committee raised a number of issues that will have 

to be addressed in the licensing process of future reactor designs: 

12. What process will be used to develop design basis accidents for the new 

designs? How will risk-significant SSCs be identified? What will be the 

regulatory treatment of non-safety systems? 

13 What use can be made of previous NRC reviews of derivative designs? 

14. For the new designs, what will be the regulatory approach with respect to: a) 

many new plants; b) multi-unit/module sites? 

15. How can we assure the required fuel quality for designs for which the safety case 

relies heavily on fuel integrity (focus on process vs. product)? Validation of fuel 

performance? Accident source terms for design basis events? High burnup? 

16. How will once-through cores that last 5 - 10 years be monitored and inspected? 

Acceptability of new "smart" on-line health monitoring systems? 

17. How will financial related requirements apply to the new concepts and business 

environment? Extension of Price-Anderson? NRC fee structure for multi

modules? Decommissioning?

2
18. What is the acceptability of natural circulation cooling as an accident recovery 

strategy? What is the availability and validation status of appropriate analytical 

models? 

19. What database exists for the lifetime temperature and irradiation behavior of 

graphite? How do the "new" graphites compare to the "old" ones? 

20. Can the NRC develop the manpower, resources, and technical expertise 

required for assessing the future concepts? 

21. Is there a sufficient database on high temperature material behavior for the gas

cooled concepts with high exit temperatures? What are the gas turbine safety 

issues for high temperature helium cycles? 

22. What new NRC research is needed? Participation with Industry? Use of 

international research and data bases? 

23. Are new approaches/criteria needed for licensing multi-purpose plants (e.g., 

power plus desalination, industrial/residential heating, hydrogen production, coal 

conversion)? 

24. What will be the spent fuel storage requirements for new fuel types/geometries? 

Potentially damaged fuel spheres/particles? 

The workshop provided a forum for raising and discussing these questions. We hope 

that both the industry and the NRC will find these proceedings useful in their search for 

answers. 

We thank all the speakers and participants for their contributions in making this 

workshop a success. We thank Richard P. Savio, Michael Markley, Medhat EI-Zeftawy, 

and the staff of the ACRS Operations Support Branch for their efforts in the conduct of 

the workshop. Special thanks are given to Jenny Gallo of the ACRS Office for her 

outstanding efforts in overseeing the publication of this document. 

Thomas S. Kress, Chairman 

Subcommittee on Future Reactors 

George E. Apostolakis 

ACRS Chairman

3
Disciplined - Meaningful - Scrutable

Remarks of Commissioner Nils J. Diaz 

United States Nuclear Regulatory Commission 

ACRS Workshop on Advanced Reactors 

June 4, 2001 

It is a real pleasure to participate in this workshop to discuss regulatory challenges for 

advanced nuclear power plants. It is particularly appropriate that the Advisory 

Committee on Reactor Safeguards is hosting this meeting, at this time. The discussion 

on nuclear power has now fully entered the national debate on the future of America's 

energy supply, and nuclear safety is going to be a priority on everybody's agenda. The 

Commission relies on the ACRS for expert advice on the safety of reactors, existing or 

submitted for licensing. The recommendations of the Committee will be of particular 

value for the Commission deliberations on the licensing of new reactors. I will be 

presenting my individual views today. They do not necessarily represent the views of 

the U.S. Nuclear Regulatory Commission (NRC), except when indicated. 

I want to premise my remarks with a few selected quotes from a "couple" of speeches 

during my tenure as a Commissioner. 

* 'There is no credible regulator without a credible industry. There is no 

credible industry without a credible regulator." 

* "It is essential for the regulator to be cognizant of the technology. It is 

essential for the industry and technologists to be cognizant of the 

regulations." 

0 "Regulations need to result in a benefit or they will result in a loss." 

* "My goal is to ensure the paths are clearly marked. A path that is clear of 

obstacles and unnecessary impediments, with well defined processes, will 

provide regulatory predictability, equity and fairness." 

* "We are learning how to define adequate protection in more precise 

terms, and to define it in terms that make sense to the American people." 

* "We have learned from our mistakes and we are bound not to repeat 

them." 

At the 2001 US NRC Regulatory Information Conference, I said: "We might be asked, 

as would other government agencies and the private sector, to sharpen our skills, and 

improve our efficiency to meet the needs of the country". We have been asked. It is 

worthwhile to try to understand why the President and the Vice-President of the United 

States have brought nuclear power generation to center-stage in the debate on the

5
energy policy for our country. Shown in Table 1 is a compilation of important aspects of 

the debate, summarizing what has changed in 20 plus years. 

The NRC has been changing to meet the challenge of what must be changed and to 

strengthen what must be conserved. I submit to you that we have changed for the 

better, especially the last three years, and that improvements in regulatory 

effectiveness and efficiency are changing from goals into reality. It has not been easy, and there are still lessons to be learned. I must say that there is one change that I 

believe speaks louder than words for the NRC staff and the agency as a whole: priority 

is now placed on what should be done better rather than on what was done wrong. 

This is a cultural change that is needed to enable the consideration of newer, better and 

enduring ways to exercise the mandate entrusted to the NRC by the people of this 

country: to license and regulate the peaceful uses of nuclear energy, with adequate 

assurance of public health and safety. I believe that we are now capable of meeting the 

regulatory challenges that we face today regarding advanced nuclear power plants. 

The improved industry performance over the past decade has enabled the NRC to 

initiate and implement reforms that are progressively more safety-focused. Furthermore, it allowed the industry to concentrate resources on the issues important to 

safety which provided a sharper focus to regulatory improvements. Safety and overall 

performance, including productivity, became supporters of each other, with the clear 

and unmistakable proviso that safety is first. 

For existing nuclear power plants, the list of profound regulatory changes and 

accomplishments, many done under the mantle of the so-called risk-informed 

regulation, would occupy the rest of this meeting. Five of them stand out: the revised 

rule on changes, tests, and experiments for nuclear power facilities (10 CFR § 50.59); 

the new risk-informed maintenance rule (10 CFR § 50.65 (a)(4)); the revised reactor 

oversight process; the new guidance on the use of PRA in risk-informed 

decision-making (Regulatory Guide 1.174); and the revised license renewal process (10 CFR Part 54). The list is growing. About two weeks ago, the Commission approved COMNJD-01 -0001 instructing the staff to give high priority to power uprates and 

allocate appropriate resources to streamline the NRC power uprate review process to 

ensure that it is conducted in the most effective and efficient manner. All of these and 

most of the other regulatory improvements conform to the Commission's decision to 

focus attention on real safety. The resulting improvements in rules, regulations and 

processes, including changes to the hearing process and enhanced stakeholders 

participation, are assuring the nation that a fair, equitable, and safety-driven process is 

being used. 

I mentioned risk-informed regulation as an important component of the changed NRC 

regulatory structure. I want to be sure you know what I mean when I use the term risk

informed regulation, so I am going to present you with my own, personal definition of it: 

Risk-informed regulation is an integral, increasingly quantitative approach to 

regulatory decision-making that incorporates deterministic, experiential and

6
probabilistic components to focus on issues important to safety, which avoids 

unnecessary burden to society. 

The definition can also be used for risk-informed operations, risk-informed 

maintenance, risk-informed engineering, risk-informed design.... 

For new license applications, much groundwork has been done, and a lot of it is useful 

to address today's issues. In the statements of consideration for 10 CFR Part 52, the 

Commission stated that the intent of the regulation was to achieve the early resolution 

of licensing issues and enhance the safety and reliability of nuclear power plants. The 

Commission sought nuclear power plant standardization and the enhanced safety and 

licensing reform which standardization could make possible. In addition, the 10 CFR 

Part 52 process provides for the early resolution of safety and environmental issues in 

licensing proceedings. The statement of considerations for 10 CFR Part 52 goes on to 

say "...the Commission is not out to secure, single-handedly, the viability of the [nuclear] 

industry or to shut the general public out. The future of nuclear power depends not only 

on the licensing process but also on economic trends and events, the safety and 

reliability of the plants, political fortunes, and much else. The Commission's intent with 

this rulemaking is to have a sensible and stable procedural framework in place for the 

consideration of future designs, and to make it possible to resolve safety and 

environmental issues before plants are built, rather than after." 

In February of this year, the Commission directed the staff in COMJSM-00-0003 to 

assess its technical, licensing, and inspection capabilities and identify enhancements, if 

any, that would be necessary to ensure that the agency can effectively carry out its 

responsibilities associated with an early site permit application, a license application 

and the construction of a new power plant. In addition, the Commission directed the 

staff to critically assess the regulatory infrastructure supporting both 10 CFR Parts 50 

and 52 with particular emphasis on early identification of regulatory issues and potential 

process improvements. The focus of these efforts is to ensure that the NRC is ready 

for potential applications for early site permits and new nuclear power plants, certified 

designs or designs to be certified, and the NRC does not become an impediment 

should society decide that additional nuclear plants are needed to meet the energy 

demands of the country. Necessary safety-focused regulations, yes; unnecessary, not 

safety-focused regulations, no. The staff is working hard to carry out this direction and I 

am sure you will hear about some of our efforts over the next two days. 

Risking being repetitive, I am going to re-start at the beginning. The US Nuclear 

Regulatory Commission has a three-pronged mandate: 

* To protect the common defense and security 

• To protect public health and safety, and 

* To protect the environment 

by the licensing and regulation of peaceful uses of atomic energy. I have long 

advocated that an adequate and reliable energy supply is an important component of 

our national security. I firmly believe that our three-pronged mandate is going to endure 

the test of time because it is good, and it is balanced.

7
Within that mandate, I am an advocate of change, functioning under the rule of law. As 

we face the regulatory challenges that are sure to be posed by the certification and 

licensing of new designs, a series of familiar requirements will have to be met, 

regardless of the licensing path chosen: 

* Public Involvement 

* Safety Reviews 

* Independent ACRS Review 

* Environmental Review 

• Public Hearing 

* NRC Oversight 

I am convinced, by practical experience, that the present pathway for potential licensing 

success of certified or certifiable new reactor applications is Part 52. First, it exists - not 

a minor issue; second, it contains the requirements for assurance of safety and the 

processes for their implementation. Lastly, it can be upgraded to meet technological advances that require new licensing paths, without compromising safety. Windows of 

opportunity can be opened, yet the price is always the same: reasonable assurance of public health and safety. A new technology, with different design basis 

phenomenology, e.g., single phase coolant, could present the need for a different 

pathway. Yet, it would have to face the same requirements listed above. What could 

be different is the manner in which some of these requirements are addressed. There is definitely room for innovation and improvement, within the safety envelope that has to 

be provided for assurance of public health and safety. 

I am also convinced that the NRC and all stakeholders need to apply common criteria 

to the tasks at hand. Every success path, however success is defined, should follow 

these simple criteria: Every path, every step has to be disciplined, meaningful and 

scrutable. 

Allow me to consider widely different roles. The NRC has the statutory responsibility for conducting licensing and regulation in a predictable, fair, equitable and efficient 

manner to ensure safety. Every step of the licensing and oversight has to be 

disciplined, meaningful and 

scrutable. 

Applicants need to satisfy the technical, financial, and marketplace requirements, and 

meet the NRC and other regulatory requirements. Every step has to be disciplined, meaningful and scrutable. 

I have no doubt that there will be objections and opposition and the law of the land will 

respect them and give them full consideration. The objections will have to be 

disciplined, meaningful and scrutable. 

These common criteria are necessary but they are not sufficient. It is indispensable that what we have learned - and it is much - be incorporated into the science, engineering and technology supporting any new reactors; they have to be as good as

8
the state-of-the-art permits. So it should be for the regulatory processes. I happen to 

believe that risk information can be a contributor to disciplined, meaningful and 

scrutable processes, and to the underlying science and technology. 

Someone once wrote a phrase framing how to achieve high performance expectations, 

and it may be appropriate for this occasion: 

Promise... to think only the best, 

to work only for the best 

and 

to expect only the best

9
Nuclear Power Generation 

- Perception and Reality -

1973-1982 2001 

Interest Rates High & Unstable Low & Stable 

Inflation High & Unstable Low & Stable 

Electrical Demand Decreasing Increasing 

Socio-political Climate Negative Improving 

Technical Maturity Low High 

Regulatory Framework Low Predictability High Predictability 

Economical Performance Poor & Unstable Good & Improving 

Environmental Image Poor Improving 

Safety Image Poor Good & Improving 

Expectations Too High Realistic 

Competition/Deregulation None High 

Standard (certified) Designs None Three + 

Combined License No Yes 

Important to National Security Yes Yes 

Financial Risk High Improving 

Public Credibility Low Good & Improving 

Bottom Line ( Low Predictability Good Predictability

8
11

United States 

Nuclear Regulatory Commission 

Disciplined - Meaningful - Scrutable 

Commissioner Nils J. Diaz 

Remarks Before the ACRS Workshop 

Advanced Reactors 

June 4, 2001

• "There is no credible regulator without a 

credible industry. There is no credible industry 

without a credible regulator." 

• "It is essential for the regulator to be cognizant 

of the technology. It is essential for the industry 

and technologists to be cognizant of the 

regulations." 

• "Regulations need to result in a benefit or they 

will result is a loss." 

Figure 1
12

"• "My goal is to ensure the paths are clearly 

marked. A path that is clear of obstacles 

and unnecessary impediments, with well 

defined processes, will provide regulatory 

predictability, equity and fairness." 

* "We are learning how to define adequate 

protection in more precise terms, and to 

define it in terms that make sense to the 

American people." 

" "We have learned from our mistakes and we 

are bound not to repeat them." 

Figure 2

Nuclear Power Generation 

- Perception and Reality 

1973- 1982 1 2001 

Interest Rates High & Unstable Low & Stable 

Inflation High & Unstable Low & Stable 

Electrical Demand Decreasing Increasing 

Socio-political Clmate Negative [ Improving 

Technical Maturity Low F High 

Regulatory Framework Low Predictability L High Predictability 

Economical Perlormance Poor & Unstable Good & Improving 

Environmental Image Poor I Improving 

Safety Image Poor I Good & Improving 

Expectations I Too High Realistic 

Competition/Deregulation None H Nigh 

Standard (certified) Designs None Three 

Combined License No F Yes 

Important to National Security Yes Y Yes 

Financial Risk High Improving 

Pubic Credibility Low Good & Improving 

Bottom Line Low Predlctablllty Good Predictability 

Figure 3

I
13

Priority is now placed on 

what should be done better 

rather than on 

what was done wrong 

Figure 4

Key NRC Regulatory Improvements 

"* revised rule on changes, tests, and experiments 

(10 CFR 50.59) 

"* new risk-informed maintenance rule 

( 10 CFR 50.65 A.4 ) 

"* revised reactor oversight process 

"• new guidance on the use of PRA in risk-informed 

decision-making ( Regulatory Guide 1.174) 

"• revised license renewal process ( 10 CFR 54) 

Figure S
14

Improvements in rules, 

regulations, and processes 

are assuring the nation that a 

fair, equitable, and safety

driven process is being used. 

Figure 6

Risk-informed regulation is an 

integral, increasingly quantitative 

approach to regulatory decision

making that incorporates 

deterministic, experiential and 

probabilistic components to focus 

on issues important to safety, 

which avoids unnecessary 

burdens to society. 

Figure 7
15

Statement of Consideration Part 52 

"The future of nuclear power 

depends not only on the licensing 

process but also on economic 

trends and events, the safety and 

reliability of the plants, political 

fortunes, and much else." 

Figure 8

Necessary 

safety focused regulation - YES 

Unnecessary, 

not safety focused regulation - NO 

Figure 9
Regulatory Requirements

16

° Public Involvement 

* Safety Reviews 

0 Independent ACRS Review 

* Environmental Review 

* Public Hearing 

- NRC Oversight 

Figure 10

Criteria for Success 

Every path, every step has to be: 

Disciplined 

Meaningful 

Scrutable 

Figure 11
17
E. Quinn, Consultant, General Atomics: The combined operating license part of Part 

52 is unproven. We haven't run through that yet, as well as early plant siting. Can you define how the Commission can help the staff to provide, to make this a more stable 

process as we go through it so that the financial community will help us to get these 

through? 

Commissioner Diaz: It's a very good point. We have it, it's there. We've been looking at it for some time, but it's not been tested. The issue is how do we make sure that it 

works the way it should be, effectively and efficiently. 

I think we learned a lot at the license renewal process. I believe that what I have learned the last few years is that Commission involvement is very necessary in this 

step. 

I will use one of the first phrases I used in a meeting down there that the enemy of the good is the better and the enemy of the better is the best. Therefore, we are going to have to be in very close contact with the staff. I believe the Commission will actually take an important role in making sure that the processes are timely. 

In this respect what we have done among many other things the last 31/2 years, is we 

have maintained our doors open. We have allowed stakeholders from all different 

areas to come and visit and let us sometimes close this little gap that exists, it is vital information to us how stakeholders, whether they're industry or there are other, you know, groups that have an interest in the proceedings, let us know how things are 

going. That has worked very well. It keeps the Commission informed early. Sometimes, the staff protects the Commission and shields us from knowing the little problems that are happening. And sometimes that is fine. There are times in which we 

need to know ahead of time. 

I think this combined operating licensing process should be very similar with the 

Commission, really on top of it all the time. 

D. Powers, ACRS Member: I'd like to phase in the issue of nuclear waste, which 

comes up repeatedly in connection with all the discussions of nuclear power, especially 

as we go to looking at maybe an increased use of nuclear power. 

Are we making any progress on this nuclear waste issue? Is there something that the 

NRC can do or is this totally in the hands of the Department of Energy? 

Commissioner Diaz: I think the NRC has done as much as it can do. We have 

engaged in the process all the way. We have tried to make sure that everybody understands that we believe there is the science and technology that offers a better 

pathway that ensures public health and safety. 

I think the decisions right now are practically at final stages. I cannot comment on them. 

I think that we are going to do what we do best; we're going to take whatever the

18
country decides in the Congress of the United States and the President, and EPA and 

we're going to work with them. We're going to try to make it a scrutable process. That 

is what we do best. 

Whatever is coming down, we're going to use it. If an application is submitted, we're 

going to try to license it working through a process. The licensing, is not assured and 

we're going to have to look at the process every step of the way. Hopefully the 

Department of Energy will do a good job. We would like to ensure that the process is 

open to the public. We need to make sure that this is disciplined, meaningful and 

scrutable. 

T. Kress, Chairman, Subcommittee on Future Reactors: With some of the new 

reactor concepts, I see one of the hard places in the regulatory challenges to be in the 

area of defense- in-depth, which is a general guiding principle for regulation. 

Do you think the concept of defense-in-depth is sufficiently rigorously defined for some 

of the newer reactor concepts or will we have to rethink what we think defense-in-depth 

is? 

Commissioner Diaz: I think, those of us who worked in reactor science know what 

defense-in-depth really is and what are its limitations. I think we have actually reached 

the limitations of defense-in-depth, and that it is time to move forward and use it in the 

best possible manner, but complimented with everything else that we can to make sure 

that we don't make cumbersome design requirements or cumbersome regulatory 

requirements. I go back to that definition, the end of the definition and risk-informed 

regulation, which avoids unreasonable burden. That is what we have to do, because 

eventually the burden is on the people of the United States. 

So, I believe that we need to relook and resharpen our focus. I know the ACRS has 

been working on this, and I share a lot of your views. 

G. Apostolakis, Chairman, ACRS: This is related to the use of risk-information in 

licensing and regulations. We hear that the agency may, in fact, receive license 

application in the very near future. Do you believe, Commissioner, that the regulatory 

system is ready to review such a license application or does it require some 

fundamental changes, which will take time, of course? 

Commissioner Diaz: We will work hard at it. The ACRS is going to need to come and 

pitch in. I think everybody is getting their attention focused on how can we move in this 

area, what is it that we know sufficiently that will provide answers within that envelope 

that I keep referring to provide the protection of all the processes. I think there are hard 

decisions to be made. 

H. Feinroth, Gamma Engineering: As I listened to some questions from the ACRS 

and also the DOE presentation I see it different -- there's a gap between what the DOE 

is focusing on, which is the entire fuel cycle not just the reactor and their interest is in

19
the goals that they've described to achieve safety and public health for the entire fuel 

cycle. Whereas the ACRS is focused, I believe, in the past and I think still on reactors 

only. It seems to me that this is more of an observation than a question. I don't believe 

that the question has an answer. The regulators need to look at the whole fuel cycle as 

well and not just the reactor as they provide advice or input to the DOE in their section 

process. 

The gentleman asked about the source term. Well, the source term of importance to 

public health is not just what's in the reactor, but what gets transported, but gets recycled, what gets sent to a repository. So I think the context that DOE is looking at 

this is correct. And I think the regulatory agency needs to figure out how to address the 

imbalance and the public health from the different parts of the fuel cycle. My concern is 

the ACRS just looks at the reactor. 

I don't know if anybody has a response to that, but I think that's an issue that needs to 

be addressed by the regulatory agency. 

D. Powers, ACRS Member: Well, we'll comment quickly that we do have the 

Chairman of the Advisory Committee on Nuclear Waste look at the waste portion of it. 

And that ACRS does also look at the fuel fabrication part of the problem as well, though 

we probably haven't focused on it very much in the discussion today because the fuel 

cycle has only been mentioned briefly here as being changed. 

T. Kress, Chairman, Subcommittee on Future Reactors: I think the questioner had a 

good point. I did want to point out that the ACNW also focuses on regulations related 

to sensitive materials and materials applications. 

Perhaps ACRS could do a little more on the fuel cycle parts, but our conception, at 

least our feeling is, the real risk part of the thing is in the reactor or perhaps in the fuel 

fabrication. 

G. Apostolakis, Chairman, ACRS: We also have a joint Committee with the ACNW 

when the issues warrant it. It's certainly a good thought.

20
Department Of Energy (DOE) Summary

Prepared by ACRS Staff for DOE 

William D. Magwood IV of the U.S. Department of Energy (DOE) led the discussions for 

the DOE staff. Dr. Magwood provided an overview of the Generation IV Initiative to 

evaluate candidate technology concepts for a new generation of nuclear power plants. 

Robert Verslius, DOE, presented the Generation IV goals, road map effort, and concept 

evaluation. Mr. Thomas P. Miller discussed the formation of a Near-Term Deployment 

Working Group (NTDG) formed to identify actions and evaluate options necessary for 

DOE to support new plants. DOE has established a Nuclear Energy Research Advisory 

Committee (NERAC) to provide independent evaluation and feedback on the 

establishment of goals and objectives and progress in evaluating candidate nuclear 

energy concepts. DOE has also established a Generation IV Road map NERAC 

Subcommittee (GRNS) to serve as an advisory group in establishing the road map 

along with a Road map Integration Team (RIT). Candidate technologies must be 

deployable by 2030. Nuclear systems are expected to meet sustainability goals 

(resource inputs, waste outputs, and nonproliferation), safety and reliability goals 

(operating maintainability excellence, limiting core damage risk, and reduced need for 

emergency response), and economic goals (reduced life-cycle costs and risk to capital). 

Criteria and metrics for each goal are being developed by an Evaluation Methodology 

Group (EMG), RIT, and the GRNS. DOE plans to evaluate all candidate concepts 

equally without prejudice toward existing technologies (e.g., light-water reactors) but 

recognizes that most primary energy generators are likely to be fission based. DOE is 

presently considering 94 concepts. The output of the Generation IV Program is 

expected to be a research and development plan to support future commercialization of 

the best concepts.

21
od IV Initiative 

Presentation at ACRS Workshop 

"Regulatory Challenges for Future Nuclear 

Power Plants" S 

June 4, 2001 

William D. Magwood IV, Director 

Office of Nuclear Energy, Science and Technology 

Office of Nuclear Energy, Science and Technology 

O'Nuclear energy systems deployable by 2030 

-'Systems offering significant advances in 

r-9sustainability 

r3safety and reliability 

c9economics 

,O'Systems include fuel cycle and power conversion 

,O'Diversity of applications (electricity, H2, water, heat) 

oDeployable in a wide range of markets

22
Office of Nuclear Energy, Science and Technology

Generation I 

Early Protoype Reactors 

- Shippingpor, 

- Dresden. Fermi I 

- Magnox

Generation 1I 

Commercial Power 

Reactors

- LWR-PWR. BWR 

- CANDU 

- VVER/RBMK

Generation III Near-Tens 

Deployment Advanced MIKE _ LWRs 

Generation IlIl 

Evolutionary 

Designs Offenng 

Improved 

Economics 

- ABWR 

- System 80, 

- AP600 

- EPR

Generation rV 

- Highly 

Economical 

- Enhanced 

Salety 

- Minimal 

Waste 

- Proliferation 

Resistart

1950 1960 1970 1980 1990 2000 2010 2020 2030 

II I I I

Office of Nuclear Energy, Science and Technology f

Committee (NERAC)

* Subcommittee on Generation IV Technology Planning 

-Established in October 2000 

to provide guidance on development 

of the Generation IV Technology 

Roadmap 

:Membership from U. S. 

Industry, laboratories, 

and academia 

•Co-chaired by 

Neil Todreas, MIT and 

Sal Levy, GE (retired)

23
Office of Nuclear Energy. Science and Technology 

sor .Committee (NERAC) 

-Subcommittee Charter: Gen IV Technology Roadmap 

•Establish goals that define the requirements for 

Generation IV nuclear energy plants 

•Suggest paths forward to resolve technical and 

institutional issues for Near-Term Deployment 

(by 2010) 

E_•Recommend Gen IV R&D Plan

id Sequencing of R&D task and initial 

cost estimates 

9 National and international collaboration 

U Systems must be deployable by 2030

Office of Nuclear Energy, Science and Technology 

.0"Facilitate research planning 

and international cooperation 

between countries interested in the 

future of Nuclear Energy 

#APLed by Policy Committee, composed ~ited SKn Kingdom of senior nuclear technology official representing member governments 

' 'Observers from: 

South Korea -:lntem ational Atomic I= , A

-SOECD/Nuclear Energy Agency 

ý-European Commission 

ý-U.S. Nuclear Regulatory 

Commission 

-•U.S. Department of State

Japan

C-----'

24

I

I

R

t
Office of Nuclear Energy, Science and Technology

!Endorsed Gen-IV 

technology goals 

!Internationalized 

the Gen-IV 

Technology 

Roadmap effort 

/Finalized charter 

governing 

memberships and 

objectives

1 = Meeting ?3d Meeting 

Washington, D.C. Paris, France 

January 2001,Api 2000 Mach 2001 

4t Meeting ofGen-IV 

International FFLowm 

Miami, RFtida 

October 2001

Office of Nuclear Energy, Science and Technology 

Near-term Objectives 

,-"Establish Near-term Deployment Working Group 

*Identify institutional and regulatory barriers to new plant deployment in 

the U.S. 

-Provide recommendations on appropriate government actions to assist 

in addressing barriers (complete by September 2001) 

Long-term Objectives 

,Establish Gen-IV Technology Project 

Identify and evaluate most promising nuclear energy system concepts 

!Provide comprehensive R&D plan to support future commercialization 

of the best concepts (complete by September 2002)

25

Seoul, Korea 

August 2000
Ž Geeraio IV Goals 

odmap 0 Effort 

Presentation at ACRS Workshop 

"Regulatory Challenges for Future Nuclear 

Power Plants"

June 4, 2001

Dr. Rob M. Versluis 

Office of Technology and International Cooperation

27

Office of Nuclear Energy, Science and Technology 

2.hch ol-og Roadmap 

* Identify and evaluate most promising nuclear energy 

system concepts (Oct '00 - Sep '02) 

"* Advisory group: Generation IV Roadmap NERAC 

Subcommittee (GRNS) 

"* Working Groups: 

- -50 U.S. experts from industry, labs, academia 

- -40 experts from Generation IV International Forum (GIF) 

member countries & organizations 

"* R&D Plan to support future commercialization of the best 

concepts
Office of Nucdear Energy, Science and Technology 

ec ology Roadmap: Goals 

Goals 

"* Reflect mid-century vision of energy needs (2030) 

"* Provide basis for evaluating nuclear energy systems 

and identify the most promising concepts 

Sustainability Goals Safety & Reliability Goals Economics Goals 

-Resource inputs -Excellence -Life cycle cost 

-Waste outputs -Core damage *Risk to capital 

*Nonproliferation -Emergency response 

Office of Nuclear Energy, Science and Technology f 

7Generation IV Systeemn: 

* An entire energy production system, including 

* nuclear fuel cycle front and back end 

* nuclear reactor 

* power conversion equipment and its connection to the 

distribution system 

* electricity, hydrogen, fresh water, process heat, district 

heat, propulsion 

* infrastructure for manufacture and deployment of the plant 

"* Limited to systems that are likely to be commercially viable by 2030 

"* Primary energy generators based on critical fission 

reactors

28
Office of Nuclear Energy. Science and Technology

Office of Nuclear Energy, Science and Technology

rWrducing te Roadmap 

Four Phases over Two Years:

Phase 1: Initial work 

Oct '00 - Jan '01 

Phase [I: Needs assessment 

Jan '01 - Jan '02 

Phase III: Response developm 

Oct '01 - May '02

- Completed 

- Jan '02 Draft Roadmap 

tent 

- May '02 Interim Roadmap

Phase IV: Implementation planning 

May '02 - Sep '02 - Sep '02 Final Roadmap

29

~~o~Roadeo o amap: Organization 

Generation IV 

International 

NERAC ---- ------ Forum (G.... 

Near'Term Deployment S Group (NTDG) 

GEN IV Roadmap • = 

NERACSubcommittee Roadmap Integration Team (R' 

(GRNS) 

Technical Community Evaluation Methodology (EMG) - .... 

-industry Technical Working Groups G): 

-Academia - - Water Coolant - 2 ..... 

"* National1 

Laboratores -- Gas Coolant - ... 

,rInternational~ros --- Liquid-Metal Coolant 

GNnl aConcepts L-- . 

SNon-Classical
Office of Nuclear Energy, Science and Technology 

ss ndPPlansnI 

Derive technology goals based on industry needs 

"* Goals have been drafted by GRNS and GIF 

"* Captured in Technology Goals Document 

Plan the activity 

"* Roadmap Development Guide drafted by RIT 

"* Working groups have been convened including international participation 

Determine how to measure concepts against goals 

* Develop criteria and metrics for each goal 

* Continue on to develop evaluation methodology 

0 Conducted by EMG, with the RIT and GRNS

30

Office of Nuclear Energy, Science and Technology 0 

Identify concepts for evaluation 

"* Drawn from a broad international base 

"* Concepts adopted or synthesized by TWGs 

"* Concepts grouped into "concept sets" 

Detail the most promising concepts 

* Interactions between TWGs & concept teams/advocates 

* Active study and comparison of underlying technology 

"* eScreening for Potential" guided by EMG criteria 

* Evaluations guided by EMG metrics

I"
Office of Nuclear Energy. Science and Technology 0 

I .4~- -ea- •.-Evafu -te-• As7,'";semble -

Evaluate the most viable concepts 

* Compare concept performance to goals 

* identify technology gaps 

* TWGs lead - RIT!EMG reviews - DOE approves - GIF 

endorses 

Assemble Roadmap to support the most promising 

concepts 

* Identify R&D needed to close gaps in areas of 

crosscutting technology 

* Assemble a program plan with recommended phases 

* Groups report - RIT integrates - DOE approves - GIF 

endorses

31

Office of Nuclear Energy, Science and Technology0 

Concept: 

A technical approach for a Gen IV system with enough 

detail to allow evaluation against the goals, but broad 

enough to allow for optional features and trades. 

Concept Set: 

A logical grouping of concepts that are similar enough to 

allow their common evaluation.
Office of Nuclear Energy, Science and Technology 

"• Request for information March 2001 

Concept elicitation, sorting, and characterization "* Screening for Potential July 2001 

Concept studies 

(assessment of technical needs by concept) "* Final screeninq April2002 

R&D plan development "• Roadmap completion September2002

Viability R&D 

• First down-selection 

Performance R&D (industry participation) 

* Second down-selection 

Demonstration wfindustry, design, regulatory reviews

Office of Nuclear Energy, Science and Technology 

torkin IGroups;1-4: 

Charter 

Identify Gen IV concepts for evaluation, evaluate their 

potential against the goals, their technology gaps and 

needs, and recommended R&D priority. 

Special Features 

"• Groups will author major sections of the roadmap on 

concepts, technology gaps and R&D needs 

"• Group members will staff the crosscut groups in the 

second year

32

I

I
Office of Nuclear Energy, Science and Technoiogy 

••ology.~Group ,i/. . 

Charter 

- Develop a process for the systematic evaluation of the 

comparative performance of proposed Gen IV concepts 

against the established Gen IV goals. 

Special Features 

"* Early delivery of products in Feb/Mar and May 2001 

"* Continued refinement of methodology 

"* Review of the TWG analyses to assure a consistent 

approach 

Office of Nuclear Energy, Science and Technology 

Charter 

Examine fuel resource input and waste output from a 

survey of Generation IV fuel cycles, consistent with 

projected energy demand scenarios. The survey of fuel 

cycles will include currently deployed and proposed fuel 

cycles. 

Special Features 

- Members mostly drawn from the TWGs and EMG 

* 8-10 month time frame for delivery of products

33
34
Office of Nuclear Energy, Science and TechnoIogy 

rtD loym,.ent Gru :_ 

Mission - Identify the technical, institutional and regulatory gaps to 

the near term deployment of new nuclear plants and recommend 

actions that should be taken by DOE. 

G Orders by 2005 

G Multiple plants in commercial operation by 2010 

1 Participants - multi-disciplined nuclear industry group 

" Nuclear Utilities - Duke, Southern Nuclear, Exelon 

" Reactor Vendors - Westinghouse, General Electric, General 

Atomics 

®National Laboratories - ANL, INEEL 

(;Academia - Penn State 

G Industry - EPRI 

( Government - DOE-NE 

eNERAC

35
Office of Nuclear Eergy, Science and Technology 

SNear-Term Deployment Report 

GTo be Issued by September 30, 2001 

G Based on evaluation of industry response to RFI 

• Request for Information (RFI) 

G Issued April 4, 2001 to reactor designers, AEs, nuclear plant 

owners/operators, Gen IV participants, and other stakeholders 

G Issued to NEI New Plant Task Force members 

G Public notice through Commerce Business Daily (CBD) 

G Solicits identification of design-specific, site-related and generic 

barriers to deployment of new nuclear plants by 2010 

G Responses due May 4, 2001- received responses from 

12 organizations 

G RFI response under review

36

Office of Nuclear E-rFgy, Science and Tech•nolog• 

•eployrnent Group 

I Deliverables 

( Near-Term Actions for New Plant Deployment 

- Near-Term Deployment Report (Roadmap) 

SNear-Term Actions For New Plant Deployment 

* Overview of recommended DOE activities and FY 02/03 funding needs 

* Intended for use during DOE budget hearing process and DOE-NE 

input to VP Energy Task Force 

D Presented to NEI and New Plant Task Force 

¢ Significant Activities include: 

- Early Site Permit Demonstration (1 OCFR52) 

- Combined Construction/Operating License (COL)Demonstration (1 OCFR52) 

- Design Certification of 1000+ MWe ALWR 

- Confirmatory Testing and Code Validation of Advanced Reactor Utilizing 

New Technology
Office of Nuclear E=ergy. Science and Technology 

rDesign Specific Responses 

qSW 1000 Framatome 

®PBMR Exelon/PBMR 

®AP600/AP1000 Westinghouse 

MIRIS Westinghouse 

®GT-MHR General Atomics 

®ABWR General Electric

37

Office of Nuclear E•ergy. Science and Technology 

•••eployment'iroup 

RFI requested information in two areas: 

. Specific Deployment Candidate Designs that meet six criteria 

®Credible plan for gaining regulatory acceptance 

G Existence of industrial infrastructure 

G Credible plan for commercialization 

q0 Cost-sharing between industry and government 

¢ Demonstration of economic competitiveness 

SReliance on existing fuel cycle structure 

I Generic & Design Specific Gaps 

SKnown gaps provided requiring ranking and possible solutions 

G, Other gaps to be identified by respondent
Offnce of Nuclear Enerfy. Science and Tectxoolog',

SDtgloyment Group.

SGeneric Gaps Responses 

q ESP Demonstration 

G COL Demonstration 

G Construction Inspection & ITACC 

e Risk-Informed Regulation for Future Design Certifications 

Emergency Planning and Plant Security 

3 Advanced Fabrication, Modularization and Construction Technologies, 

D Standardized Life-Cycle Information & Configuration Control Systems 

G High Level Waste Disposal Resolution 

G, Risk Management Tool 

" Public Influence and Acceptance 

" Appropriate Resource and Financial Arrangements

38

-1
Mofce of Nuclear Energy. Scence and Technology 

* Request for concept information (RFI) 

"* RFI response 

"* Concept statistics & key features 

"• Grouping of concepts 

"° Current activities on concept evaluation

39
Office of Nuclear Energy, Science and Technology

Total: 94 

By reactor coolant type

By country 

- France 3

32 

17

By organization type 

- University 27

"• UK 

"* US

4 

45

S7 Others-13 

.Argentina, Brazil, Canada, 

Germany, Italy, Netherlands, 

Russian Federation

* Industry 22

* Laboratory 45

40

Office of Nuclear Energy. Sceence and Technology 

fifomg concepts 

Concept: 

A technical approach for a Gen IV system with enough detail to allow evaluation against the goals, but broad 

enough to allow for optional features and trades. 

Concept Set: 

A logical grouping of concepts that are similar enough to 

allow their common evaluation.

"* Water 

"* Gas 

"* Liquid Metal 

"• Non-classical

28 - Japan 19 

17 - Korea 10

I
Office of Nuclear Energy. Science and Technology 

Variables 

; Coolant (H20, D20) 

> Coolant phase & conditions 

> Spectrum (thermal, epi-thermal, fast) 

ý- Primary system layout (conventional, integral) 

> Fuel cycle (U vs.Th, once-through vs. recycle) 

> Thermal output 

> Maturity 

Office of Nuclear Energy. Science and Technology 

Crosscutting R&D Issues 

> High temperature materials 

> Modular manufacturing technologies 

> Internal control rods 

>- I&C

41
Office of Nuclear Energy, Soence and Technology 0 

Variables 

"* Reactor concepts 

> GT-MHR 

> PBMR 

> Fluidized Bed Reactor 

> GCFR 

"• Applications of fission heat 

> Electricity generation: direct vs. indirect cycle 

> Process heat applications (industrial smelting, petroleum refining, hydrocarbon reforming, coal 

conversion, etc.) 

> Desalination

42

Office of Nuclear Energy, Science and Technology 

Gas'Colant(conIfd) 

Fuel forms and fuel cycles 

SLEU 

SThorium 

> U-Pu 

Generic R&D issues 

> Fuel fabrication quality assurance 

> Fuel performance -- integrity and FP retention 

> Lifetime temperature and irradiation behavior of graphite 

structure 

> High temperature materials and equipment 

e Passive decay heat removal for fast-spectrum concepts
Office of Nuclear Energy. Science and Technology - .) 

Variables 

> Size (large/monolithic, modular, transportable) and targeted 

clients 

> Coolant (Na, Pb-alloy, Pb, ... ) 

> Fuel type (oxide, metal, nitride, composites) 

; Primary system layout (loop, pool) 

> BOP options and energy products 

> Energy conversion options 

> Fuel recycle technology (aqueous, dry) 

Office of Nudear Energy. Scence and Technology 

* Focus: adequately defined concepts with significant potential 

* Variables 

> Cooling approach (convection, conduction, radiation) 

> Coolant (molten salt, organic coolant) 

> Fuel phase (solid, liquid, gastvapor) 

> Electricity generation technology conversion (turbine, gas 

MHD, direct conversion of fission-fragment energy) 

> Alternative energy products or services 

> Fuel cycle

43
Office of Nuclear Energy. Science and Tochnology 

boupin 

TWG's have grouped concepts into "concept sets" 

Concept sets share 

> Technology base 

1• Design approach 

Rationale for grouping 

> Efficient division of TWG analysis effort 

> Streamline evaluation process 

> Avoid premature down-selection

44

Office of Nuclear Ene•gy, Scence and Tecoirroogy 

Crosscut issues 

> Modular deployable 

> Hydrogen production and very high temperature systems 

> Advanced fuels and fuel management techniques 

; Energy conversion systems (esp. non-Rankine)
Office of Nuclear Energy, Science and Tecnmology • : 

"* PWR loop reactors (3) 

"* Integral primary system PWR's (6) 

"* Integral BWRs (6) 

"* Pressure tube reactors (3) 

"* High conversion cores (11) 

"• Supercritical water reactors (3) 

"* Advanced fuel cycle concepts (14) 

> MOX 

> Thorium 

> DUPIC 

> Marble Fuel 

> Neptunium 

Office of Nucdear Energy, Soence and Tedhnology 

* Pebble bed modular reactor concepts (5) 

"* Prismatic modular reactor concepts (5) 

"* Very high temperature (-1 5000C) reactor (1) 

"• Fast-spectrum reactor concepts (5) 

"* Others (4) 

> Fluidized bed 

> Moving ignition zone concepts

45
Office of Nuclear Energy. Scence and Technology 

-g ,on-Classical Systems TWG 

"* Eutectic metallic fuel (2) 

"• Molten salt fuel (4) 

"* Gas core reactor (1) 

"* Molten salt cooled/solid fuel (1) 

"* Organic cooled reactor (1) 

"* Solid conduction/heat pipe (1) 

"* Fission product direct energy 

conversion (2)

46

Office of Nuclear Energy. Scence and Technology 

i•..-..q•uid Metal '"G 

Four major categories of concepts: 

> Medium-to-large oxide-fueled systems (6) 

SMedium-sized metal-fueled systems (8) 

SMedium-sized Pb/Pb-Bi systems (8) 

SSmall-sized Pb/Pb-Bi systems (6) 

Liquid Metal TWG is also examining three supporting 

technology areas 

> Fuels (oxide, metal, nitride) 

; Coolants (Na, Pb/Pb-Bi) 

SFuel Cycle (advanced aqueous, pyroprocess)
47

Oflrce of Nuclear Energy. Science an- Tecr'.nology 

TWG's are analyzing the candidate concepts for 

> Performance potential relative to the technology goals 

> Technology gaps 

A report will be prepared this fiscal year describing 

> Concepts 

> R&D needs 

> Results of the initial "screening for potential" evaluations
ration III+/IV

Presentation at ACRS Workshop 

"Regulatory Challenges for Future Nuclear 

Power Plants"

June 4, 2001 

R. Shane Johnson, Associate Director 

Office of Technology 

and International Cooperation

Office of Nuclear Energy, Science and Technology

anced Reactors

o•Near-Term Actions

"* Complete report on recommended DOE activities 

- Report will reflect generic and design specific issues 

- Report to be issued by September 30, 2001 

"* Significant activities expected to include: 

- Development of Regulatory Framework for Gas Reactor 

Technologies 

- Early Site Permit Demonstration 

- Combined Construction/Operating License Demonstration 

- Design Certification of Advanced Reactors

49
Office of Nuclear Energy, Science and Technology 

S....map 

Near-Term Actions 

"* Evaluate the most viable concepts 

"* Compare concept performance to technology goals 

"* Identify technology gaps 

"* Identify R&D needed to close technology gaps 

"• Prepare comprehensive report on most promising concepts 

including detailed R&D plan

50
T. Clements, Nuclear Control Institute: I was a little confused during the DOE 

presentation about the relationship between the roadmap and the review you're doing 

and what's happening with the Exelon pebble bed reactor. From what I hear, 

depending on what happens in South Africa, they plan to start construction in 2004 and 

have a reactor operating in this country in 2006. It sounds to me like you're behind the 

curve on what's happening with that reactor. Are you going to ask them to slow down 

their decision process in pursuing this with NRC? You're behind the curve on what 

they're doing here on the ground with the NRC or do you assume that you're going to 

include this reactor in your roadmrap? I'm just confused about the relationship between 

what you're doing and the pebble bed. 

W. Magwood IV, DOE: The pebble bed reactor that R. Versluis spoke to, is a class of 

PBMRs, those are not necessarily, in fact may not really be the reactor that Exelon is 

interested in and is now being discussed in South Africa. That specific design is being 

discussed as part of the near-term deployment activities. And, as I've mentioned, those 

activities are largely complete and will be final -- scheduled to be final through the 

NERAC process in September, and include largely institutional issues that are being 

raised by NERAC that are fully in concert with the schedule that PBMR corporation is 

on. 

And, in fact, there are representatives of Exelon on some of the working groups that are 

providing information about the schedule and trying to keep everything in concert. 

So that PBMR is slated for near-term deployment as opposed to being in the longer 

term Generation IV activities. And that's simply because of the fact that it's of near-term 

interest to a utility and, therefore, it's appropriate that we look at it as something to be 

deployed by 2010. And whether it actually gets deployed by 2010 or not is up to Exelon 

and others. 

E. Quinn, Consultant, General Atomics: We've read the Vice President's report -- or 

the President's report and it addresses investment in new technologies for renewables, 

for coal for example, and some of the 105 recommendations address advance nuclear 

designs. Can you advise in FY '02 and beyond how those recommendations will come 

into DOE planning? 

W. Magwood IV, DOE: No. To expand on. No. Let me just say that, obviously, our 

international partners are all very pleased with the outcomes that were in the Vice 

President's review and have every hope that eventually there'll be more resources 

devoted to nuclear research and development by the government. Certainly there 

would have to be to do any of the things that we've talked about today. 

What will happen in specific fiscal years, 2002 in particular, I simply don't have an 

answer for you. I think that as the government continues to digest results of the review, 

we'll begin to talk more in terms of what do we have to do, to actually implement those 

things, and those discussions have already started moving.

51
But I wouldn't expect to hear any specific implementation announcements other than 

what you may have already heard from the Secretary. I think he made some 

announcements recently about specific things in non-nuclear aspects. But on the 

nuclear aspects it's going to take a while to adjust it, move on it and to formulate those 

implementation activities. 

So I would expect that over the course of the next few months, those would start to 

come out. 

G. Apostolakis, Chairman, ACRS: If you had to give us the two most important 

regulatory challenges for meeting all these wonderful initiatives, what would they be? 

W. Magwood IV, DOE: That's a good question. I think I'll answer the question a little 

more generic. 

I think that it's extremely important the NRC move as close to performance based risk

informed regulation as possible. Because these technologies are dissimilar in so many 

ways, and you're already starting to see it. There's already a large discussion going 

forward about the pebble bed reactor versus light water reactor technology and how 

you license those. 

The only way to do that successfully with these different concepts floating around out 

there is to move to a technology- independent regulatory approach. And unless you do 

that, you're going to inhibit the development of these new technologies because people 

will not have the confidence that NRC can respond quickly enough to regulate these 

technologies. 

I know there's a lot of concern about how long it's going to take to get regulations for the 

pebble bed reactor. And we're working with General Atomics at DOE with the 

development of their system, and that presents similar challenges. So I think that the 

larger issue is the one you have to deal with. 

In the nearer term, I think it's really more a job of demonstrating the pieces are already 

out there. But even as we look at these newer technologies coming in before now, they 

present issues, many that you are already very familiar with. 

So I would say that pushing as fast as possible towards a new regulatory regime that 

will support new technologies in the next century is really going to be -- should be a high 

priority. 

D. Powers, ACRS Member: Well, it seems to me that if you're going to encourage 

people to move to a performance- based regulatory system, that must mean surely 

you're looking at performance indicators for these new generation? Is that the case? 

W. Magwood IV, DOE: I think the answer to that is yes. If you look at our technology 

goals, and I think you're going to get a rundown of that.

52
You'll see a very high level version of what those performance goals are. On a 

regulatory space, you're talking about safety. You'll see some indications where we 

think things should go, but not to the level of detail because these technology goals are 

very, very high level. You're not going to see a low level of detail, but you will see an 

overall vision. 

E. Lyman, Nuclear Control Institute: I think there are public issues that really have to 

be thought about before large expansion in DOE's research budget has to be 

contemplated. Because these days you have to really worry about whether what looks 

like government subsidization of one energy technology over another, how that will be 

perceived, especially by small scale generators using other competitive fossil fuel 

technology. And in a deregulated environment that's going to be a greater concern. 

So, I was encouraged when these reports of a task force on near-term deployment that 

recently reported to NERAC discussed a cost sharing program with industry for near

term deployment. I was wondering if industry had actually made any firm commitments 

in that regard, since this would be a positive step since I don't think they've put any 

money down so far in these initiatives? 

W. Magwood IV, DOE: First, it's important to clarify, and I think you raised a good 

point. There's two things really important to clarify. 

First, in general, you know our office is not in the business of corporate welfare. We're 

not here to make technologies marketable that wouldn't otherwise be marketable, you 

wouldn't otherwise compete on it. In fact, our goals, and you'll hear about it, for our 

Generation IV have a lot of built into them about the need to be economically 

competitive. That's a hallmark of what we're trying to do. 

And let me say for the record that there should not be a new nuclear power plant that's 

not economically competitive in this country. It shouldn't be built because we're not 

going to subsidize it and if industry is not willing to go off and do it because they can 

make money, it shouldn't happen. It shouldn't be done. 

Now, regarding the specific point you raised, I think that where we are right now -- well, 

first it's important to recognize that this is a NERAC advisory group, so we're not at the 

point where we're making commitments on a policy basis on behalf of the industry. We 

have asked certain experts in industry along with academia and working with our 

national laboratories to come together and make recommendations. These 

recommendations will flow up through the NERAC process and if it comes out the other 

side, NERAC will make a recommendation to DOE that we should go pursue a program 

in that vein. 

But at that stage, if that were to happen, we would be in a position to approach the 

industry and say "Okay, your people were on this panel, here's the recommendation 

that they made, Mr. CEO do you want to buy into this?" And if they don't want to buy

53
into it, we don't have to do it. But, you know, it's a recommendation. It's not a 

commitment on anyone's part, especially ours. 

You know, with my budget I couldn't commit to anything they recommended at this 

point. So, it's really a recommendation for the future. 

The question we asked was if we were going to solve these problems, how would we go 

about it? And that's what these recommendations gives us. It gives us a way of solving 

the problems. 

It doesn't mean that we have to do it. It doesn't mean the industry has to do it, but it 

gives us a methodology. 

So the answer to your question is no, no one's made any commitments, nor would it be 

appropriate to at this point in time. 

D. Powers, ACRS Member: I have a question that comes to mind when I see these 

plans for Generation IV reactors. My good friends at the Nuclear Energy Institute 

regularly provide me metrics on the performance of the current generation of plants in a 

variety of areas, including resources, safety and economics. And they show excellent 

performance, just outstanding performance in the last ten years. 

In all this roadmapping exercise, do you carry along some representative of the current 

generation plants as a comparison so you can see if you're really going to accomplish 

anything with these new plants. 

R. Versluis, DOE: Well, it's a good question because the initial screenings are really 

not much more than comparing in a number of different areas with the Generation III 

technology. So, they are qualitative comparisons, and that's how we approach it, is 

comparing it with the Generation III technology. 

D. Powers, ACRS Member: We don't have a whole lot of performance and data on 

those Generation Ill plants the way we do with the existing plants? 

R. Versluis, DOE: We think at this point with the amount of data that we have on the 

various concepts, there is no need to be very, very precise about these things. What 

the schedule, the last slide really showed is that we need to do a certain amount of 

viability research where we get a better handle on how to measure, how we can 

measure the various indicators before we can do a more sophisticated screening. 

J. Garrick, Chairman, ACNW: It might be important to point out, too, that GRNS has 

put a lot of emphasis on the total energy system concept, and that has kind of evolved. 

When we first got together, that wasn't so much an emphasis. And when you think 

about performance indicators, you've also got to think about the scope that we're 

addressing this time, namely the total energy system.

54
So, it would seem that if we're going to go in the direction of performance indicators that 

are compatible with risk-informed performance based regulatory practice, we'll be 

talking about probably a different structure and at least a more range of indicators that 

we've perhaps ever seen before. Is that not correct? 

R. Versluis, DOE: Yes. For example, the base case we're comparing with, of course, 

has a once through fuel cycle. We have various criteria that have to do with the waste 

and use of fuel, but particularly the waste forms that can be achieved by other fuel 

cycles. 

So, you're very right that we are not just looking at the reactor, but the entire system 

from soup to nuts, so to speak. 

G. Apostolakis, Chairman, ACRS: Now when you say reliability goals, I mean are 

they goals the way we understand them, numerical goals for reliability? 

R. Versluis, DOE: That's where we would like to end up, but reliability you can't really 

put a metric of reliability together until you know the design pretty well. And so early on 

we are really looking at very general indicators that might lead to reliability, but it's not -

as I remember well, it's actually not a screen for potential criteria. It doesn't come into 

play until later. 

W. Shack, ACRS Member: One of the things I noticed this morning in the whole 

discussion of the Generation IV thing was that the word "severe accident" never 

appeared anywhere. Do you envision that as being a technology need that will have to 

be addressed in the R&D program? 

R. Versluis, DOE: Yes. One of the goals, the second safety and reliability goal has to 

do with core damage. And then the third goal has to do with the emergency response. 

So in both of these goals severe accidents are an issue. 

And the second goal will assume the performance of a PRA. And the third goal will 

have to involve all the severe accident, that could lead to a release off-site. 

P. Ford, ACRS Member: We've been told earlier on that risk-informed regulation is 

going to be a part of your strategy, and yet we're looking at a whole lot of new systems 

here for which we have no experience at all in terms of time dependent degradation. 

As you're going through your screening process, does the time needed for R&D to 

resolve those questions, enter into your timing and your decision making? 

R. Versluis, DOE: Yes, it does. And certainly we hope or we intend but in early on in 

particular to focus on those issues where there's a large amount of uncertainty and try 

to reduce that uncertainty. That's how we will focus what we call the viability R&D, so 

that we have a better idea of what the potential is.

55
P. Ford, ACRS Member: And have you also taken into account the question of the 

manpower capability of doing that research? 

R. Versluis, DOE: Well, there will of course be as part of the roadmap an estimate of 

required manpower, resources and infrastructure. But we are certainly aware that there 

is a lot of work needed there and a lot of investment needs to be made. 

E. Lyman, Nuclear Control Institute: I just have to follow up from my earlier question, 

because I think what we've just heard is a list of activities which I don't think it's 

appropriate for the government to be funding. These are activities which are associated 

with providing a regulatory climate or easing licensing advanced reactors. And I think in 

today's context, that's a cost that really should be born by the applicants. 

Licensing is expensive, but that is part of the package for trying to develop a new 

nuclear reactor and market it. And so I think it raises real questions whether DOE 

should be involved in trying to facilitate or come up with ways of easing the site permits 

and other regulatory activities. 

I'm also concerned about DOE proposing a licensing framework for reactors and then a 

way of meeting those licensing criteria. I think there really has to be a separation 

maintained between the licensing standards and the actual applicant. Because 

otherwise these criteria could be gerry-rigged to justify or to facilitate the particular 

reactor you're pushing. 

W. Magwood IV, DOE: What we're doing, Ed, and for everyone else who had concern 

about this, is we're focusing on generic issues, and this is something that DOE has 

done basically throughout history. 

For example, in the case of gas reactors there are some very generic issues related to 

the implementation of gas reactor technology in the United States whether it's a pebble 

bed or GT-MHR or something else, you have to deal with, for example -- and this is 

something that we've had a lot of very important discussions about. If in the case of a 

gas reactor you're relying very heavily on the quality of the fuel, how does one go about 

thinking about fuel manufacturing in concert with the design of a power plant? You 

can't separate it as easily as you can in the light water reactor. That's a very, very 

broad generic technology issue. And I think it's entirely appropriate for DOE to be 

involved in that. 

What we will not be involved in are the specific -- and NRC, by the way I'll point this out, 

NRC's Office of General Counsel has been very, very diligent about keeping both NRC 

and DOE straight about this issue. 

We will not contribute to the specific design related regulatory activities NRC will be 

participating in with the vendors. There will be a separate activity that will probably be 

coming on in the next year or so. We expect that Exelon, or whoever, will come to the

56
NRC and will be obligated to pay for those activities. We don't anticipate being involved 

in that. 

But the generic activities are things that we think the government ought to be involved 

in and should be involved in. And I'll be happy to talk with you more about that later, 

but I think it's entirely appropriate what we're doing as long as you stay on this generic 

level. I think there has to be a distinction. 

R. Uhrig, ACRS Member: There's a number of rather exotic materials involved in the 

various concepts that have been talked about this morning. Is there any consideration 

or any time being spent looking at the availability of these? Even something as 

common as helium, there's a limited amount of that unless you want to produce it 

artificially. And I just wondered if this is an issue that's going to be brought into the 

consideration? 

W. Magwood IV, DOE: That's a really good question, and something that I've actually 

started to worry about myself. The answer to the question is no, we haven't done this 

stage. And the reason we haven't is because we haven't reached this 2002 target of 

narrowing down the number of options. When we know what concepts we're really 

going to spend our energies on, we're going to really have to deal with those materials 

issues. 

And I can't talk too much about this, but we are expecting in the next few weeks to 

really strengthen our materials activities within the DOE infrastructure and start to have 

more focus on these issues. Because I think they're too disperse right now. We need 

to really focus our energies there, and we're going to be doing that very soon. We'll 

make some announcements about that. 

But your question is really good one, and we're worried about it but it's too early for us 

to really go a whole lot further.

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58
Presentation Summary 

Safety Design Aspects 

And 

U.S. Licensing Challenges 

Of the 

Pebble Bed Modular Reactor 

By 

Ward Sproat - Exelon Generation 

Dr. Johan Slabber - PBMR Pty. 

This presentation consists of three sections: An overview of the status of the 

PBMR project in South Africa, a review of the design features and philosophy 

being utilized to design the PBMR, and a summary of the key licensing issues 

that Exelon has identified in assessing the licensability of the PBMR for 

application in this country. 

Project Status 

The PBMR project is currently completing the Preliminary Design Phase. Based 

on the information developed, a Detailed Feasibility Study is currently being 

performed and the report will be issued in mid-summer. The investor companies 

will make their decisions regarding investing in the next phase of the project 

based on the report's findings and their own investigations. If the decision is 

made to proceed, and if the South African government agrees, a PBMR 

demonstration plant will be constructed near Cape Town, starting in late 2002. 

Safety Design Aspects 

The PBMR design philosophy is to use both passive and active engineered 

features to provide both prevention and mitigation capability as well as to reduce 

dependence on operator actions. The PBMR reactor is being designed to a set 

of principles that is intended to assure fuel integrity, provide multiple fission 

product barriers to the environment, and to provide safeguards against nuclear 

material proliferation. 

Fuel integrity will be assured by quality controls on the fuel manufacturing 

process, minimizing excess reactivity, assuring excess heat removal from the 

fuel, prevention of chemical attack on the fuel, and preventing excess burnup. 

Multiple layers of fuel particle coatings, the primary pressure boundary system, 

and the containment structure will provide barriers to fission product release. 

Nuclear material proliferation is addressed in the inherent design of the fuel and 

politically through agreements by the Republic of South Africa with international 

agencies.

59
Licensing Issues

Exelon has determined that a number of technical and non-technical issues will 

need to be addressed during the licensing of the PBMR. These issues may 

affect the ability to license the design in the US at a cost-competitive price. Some 

of the issues will require exemptions, rulemaking or legislation to address the 

unique aspects of small modular reactors.

60
61

Safety Design Aspects and U.S. 

Licensing Challenges of the 

PBMR 

Ward Sproat - Exelon Generation 

Dr. Johan Slabber - PBMR Pty.

Agenda 

"* Project Overview 

"• PBMR Safety Design Features 

"* U.S. Licensing Challenges
62

PBMR Project Overview 

* Ending Preliminary Design Phase 

"* Feasibility Study in preparation 

"* Investors' decisions by end of year 

"* RSA demonstration plant construction start 

in late 2002 pending approvals 

"* Exelon decisions hinge on economics and 

technical risks
63
64

Reactor Safety Design 

Principles 

"• Assure fuel integrity 

"* Multiple fission product barriers to the 

environment 

"* Nuclear material proliferation safeguards

FUEL ELEMENT DESIGN FOR PBMR 

S.rGraphite layer 

Coated prtid les; ibedde 

i: Graphite Matrix 

Fuel Sphere S--":,: c Bc, CoooF.ue 

S235/:000mm 

Half Setion \\Po=o Cxbo.

HafSeto 95/1C00 mm 

Dia. O. 9•n

Coated farticle 

U,-niu, " :io de 

Fuel 

Jan 3i 200i
65

Reactor Design Principles 

Assure Fuel Integrity 

- Assure Fuel Quality 

- Control Excess Reactivity 

- Assure Heat Removal from Fuel 

- Prevention of Chemical Attack 

- Prevent Excess Bumup

Assure Fuel Integrity 

Assure Fuel Quality 

- Fuel Design has been proven internationally 

- Fuel Qualification Program 

"• Fuel Performance Testing Program 

"* Fuel Fabrication Quality Assurance Program 

- Operational fuel integrity assurance by 

monitoring primary coolant activity online
66

Assure Fuel Integrity (cont'd) 

- Control of Excess Reactivity 

- Low Excess Reactivity = 1.3% delta k effective 

- Core geometry maintained by design for all credible 

events 

- PBMR core design precludes Xenon oscillations 

- Demonstrable large Negative Temperature Coefficient 

of Reactivity 

- Criticality safety assured for spent and used fuel

Assure Fuel Integrity (cont'd) 

- Assure Heat Removal From Fuel 

- Materials properties and design features 

assure heat transfer from fuel to RPV 

- Passive heat sink provided by the Reactor 

Cavity Cooling System for extended period 

- The reactor cavity including its structures 

will maintain geometry during all credible 

events.
67

Fuel Performance at Elevated 

Temperatures 

1E+0O0 ' 

1 E-01

i1 E-02 

1E-03 

1E-04 T 

S1 E-05 4

1E-06 

1000 1200 1400 1600 1800 2000 2200 2400 2600 

Fuel Temperatures ['C]

Nominal Fuel Performance
68

Assure Fuel Integrity (cont'd) 

Prevention of Chemical Attack 

- Water systems at a lower pressure than that of the 

primary coolant system during operation 

- Water ingress to reactor when depressurized 

prevented by physical design 

- Primary coolant system monitored to detect, and 

cleaned to remove moisture and air 

- Graphite oxidation due to air ingress prevented by 

physical design of reactor, gas manifold and citadel

Assure Fuel Integrity (cont'd) 

Prevention of Excess Burn-up 

- Physical core design 

- On-Line gamma spectrometric system 

to measure fuel bum-up
69

Fission Product Barriers to 

Environment 

"* Individual fuel kernels with 3 layers 

• High integrity primary pressure boundary 

"• Containment (Confinement) 

- Reinforced concrete structure 

- Filtered vent path 

- Hold up of fission products 

- Plate out 

- Auto-close blowout panels 

- Late release

Nuclear Material Proliferation 

Safeguards 

• International Atomic Energy Agency 

(IAEA) / Government of the Republic of 

South Africa Safeguards Agreement 

"• Non-Proliferation attributes inherent in fuel 

design
70

Key Technical Licensing Challenges 

"o Lack of eas reactor technical licensing 

framework 

"* Fuel qualification and fabrication process 

licensing (South African Fuel) 

" Source Term: Mechanistic or Deterministic 

"° Containment performance requirements 

"* Computer code V&V 

"* PRA - Uncertainties, Initiators and End States 

" Regulatory treatment of non-safety svstenm 

"o Classification of SSC's 

"* Lack of technical expertise on gas reactors

Key Legal Licensing 

Challenges 

"* Price Anderson indemnity 

"* NRC operational fees 

"* Decommissioning trust funding 

"* Untested Part 52 process 

"* Potential number of exemptions
T. Kress, Chairman, Subcommittee on Future Reactors: Do you have a goal for 

how many particles can be failed within the core before you violate 10 CFR 100? 

W. Sproat, Exelon: This is clearly an issue that we're going to have to wrestle with the 

staff, once we decide ourselves how we think the appropriate way of addressing it, is 

what's the source term? Is it mechanistically determined source term or 

deterministically determined source term. 

T. Kress, Chairman, Subcommittee on Future Reactors: Isn't the answer obvious 

there? 

W. Sproat, Exelon: No, the answer's not obvious. I know what we would like to do, 

but the issue of how good are your goods analyzing your diffusion coefficients and 

being able to provide an analytic framework for migration of fission products from the 

core to the environment is going to be a challenge. It's going to be a challenge. 

Obviously, containment performance requirements, Johan talked about the containment 

design and whether or not a zero leakage or a LWR type containment would be 

required versus moderate to high leakage filtered containment would be required is 

obviously an issue that's going to be discussed at some length. 

T. Kress, Chairman, Subcommittee on Future Reactors: And that would be linked to 

the fuel quality? 

W. Sproat, Exelon: Absolutely, and to the source term. The issue of the various 

computer codes that are being used in South Africa to design this plant, how they're 

verified and validated and how they're benchmarked against the other existing codes 

will be an extensive effort associated with that. 

The PRA itself that's being developed in South Africa that we're advising them on, it's 

kind of interesting. If you have -- what's your endstate if core melt isn't a valid endstate 

for your reactor? Than what is your endstate? What are your initiators and how do you 

determine your uncertainties of the various accident sequences? 

T. Kress, Chairman, Subcommittee on Future Reactors: Your endstate is quantity 

of fission products. Frequency of fission products. 

W. Sproat, Exelon: It might be. The point is that we're exploring some new ground 

here and, obviously, there'll be some discussions with staff about how we go and do 

that. 

The regulatory treatment of nonsafety systems and how we classify the SSCs, the 

safety system components, will really be a key issue. 

Finally, an issue that I lumped in the technical area, but it's a real practical issue is there 

aren't a lot of people left in the U.S., in the NRC, in the national labs or in DOE that

71
have gas reactor experience and understanding. Obviously, I think you've gotten a 

sense as we go forward with this, if we submit an application having people who 

understand the technology, understand the science and can provide good independent review of the submittal is going to be a real challenge. 

On the last slide I have is the nontechnical, what I'll call the legal licensing challenges. I personally believe we have a very good chance at satisfactorily resolving a number of the technical issues that I showed on the previous slide. I'm not as confident about 

some of these, because some of these are potential deal breakers for moving forward 

with merchant nuclear power plants in this country. That's what we're talking about 

here; this is not a power plant or nuclear plant that's going to go into a rate base 

somewhere. This is a merchant plant where the shareholders are going to take the risk 

of building and operating this plant and whether or not it makes money in the 

deregulated marketplace is solely dependent on the technology and the company that 

runs it. 

So, the first issue up here is Price Anderson. The current law and the way it's currently interpreted by the NRC is that each reactor in the country is assessed a retrospective premium of $90 million per reactor in the case of an accident anywhere in the U.S. 

associated with any reactor. If I've got a 2200 megawatt light water reactor plant, like 

our Limerick plant, that means my retrospective premium at risk due to a reactor 

accident somewhere in the U.S. is $180 million retrospective premium associated with 

that plant. 

If I have the same capacity of pebble bed modular reactors under today's law, my retrospective premium would be $1.8 billion for that same amount of capacity. Even I 

would have difficulties selling our board of directors to take that kind of a risk associated with that kind of retrospective premium associated with an accident from a reactor that 

we don't own or operate. So that's got to be addressed somehow. 

The second issue up there is the NRC operational fees. Right now the operational fees 

are approximately $3 million per reactor. Again, say at the Limrick plant, that means about $6 million a year for the two reactors. The same size for 2200 megawatts, you're talking about $60 million a year in NRC licensing fees for a 2200 megawatt set of string of PBMRs. 

The decommissioning trust fund is another issue that's clearly going to have to be 

addressed. The law gives a number of different alternatives, but those alternatives have presupposed that generally the plant is going to be operated by a regulated utility and that in the rate base in which the plant is based rate, you have a set aside income 

stream that goes and funds the decommissioning trust fund. In our case that won't be 

the case. These plants won't be in a rate base. How we fund the decommissioning trust fund, how much we have to put up front and what we can put into a sinking fund needs to be resolved. The law is not clear on that at this point in time.

72
Clearly, Part 52 licensing process which is, we think, the right way to go is untested at 

this point in time. Nobody's actually done it. So the staff will be learning, the applicants 

will be learning, and how we actually work our way through that and how long it takes is 

going to be a key challenge for us. 

Finally, I have up there the potential number of exemptions. As I talked about earlier, 

there is no gas reactor licensing framework. If there's not when we go with an 

application, the staff might decide that a number of the things we're asking for are very 

appropriate to license this plant, but will require exemptions from the existing regulatory 

framework. Obviously, it would be undesirable to all of us to have the first advanced 

reactor in place with a significant number of exemptions. It just doesn't work. 

So, those are the key issues and challenges we see on the licensing side, both from the 

technical side and the legal side. As I said, we are considering all that and now we'll go 

into our decision making process as to whether or not to proceed with both the venture 

in South Africa and the licensing process here in the U.S. by sometime around the end 

of the year. 

T. Kress, Chairman, Subcommittee on Future Reactors: These appear to me like 

mostly policy issues rather than technical ones related to the reactor design? 

W. Sproat, Exelon: A number of these will require some policy statements and 

decisions by the Commission itself, yes. 

T. Kress, Chairman, Subcommittee on Future Reactors: Very good. Is there any 

discussion or questions for either of our two speakers? 

G. Apostolakis, Chairman, ACRS: Yes, I have a question. As I recall in one of your 

communications to the staff in addressing these issues, the key legal licensing issues, 

you proposed that a site with ten units be considered as one reactor? 

W. Sproat, Exelon: One facility. 

G. Apostolakis, Chairman, ACRS: One facility. 

Now, if this is accepted by the staff, then should we also be applying the same idea to 

various safety goals and say, assuming that the concept of core damage makes sense 

here, that if the goal is 10 to the minus 4 and that would apply to the facility, so each 

unit then would have to be ten to the minus 5? Given the fact that you have ten of 

them, you have some synergistic effects, maybe it'll have to be even lower than ten to 

the minus 5. 

W. Sproat, Exelon: Well, synergistic effects is not intuitively obvious to me. There 

are synergistic effects when in fact the risk is from one reactor to the other. I'm not 

ready to concede that point at this point.

73
G. Apostolakis, Chairman, ACRS: Okay. Fine.

T. Kress, Chairman, Subcommittee on Future Reactors: Some common mode. 

G. Apostolakis, Chairman, ACRS: Some common mode, perhaps. Anyway, but how 

about the thought process here that you would apply stricter criteria -

T. Kress, Chairman, Subcommittee on Future Reactors: Yes, instead of calling it 

core melt, call it fission product release -

G. Apostolakis, Chairman, ACRS: Call it something else. Yes, fission product 

release. 

If we treat 10 PBMRs as one facility with respect to these five bullets that you showed 

us, shouldn't we be doing the same when it came to risk and treat it as one facility and apply the goals to the facility, in which case of course we will have much lower goals for 

each individual unit? 

W. Sproat, Exelon: Well, we certainly haven't done that for two and three unit light water reactors. So, I hesitate to do that for a smaller, supposedly safer reactor. 

G. Apostolakis, Chairman, ACRS: For a two unit reactor there are some PRAs 

where they look at these things. A factor of two in the goals really doesn't mean 

anything. When you talk about ten, a factor of ten, then you're beginning to see some 

difference. 

So it seems to me that if we are to apply this idea to the five legal licensing challenges you mentioned, maybe we ought to think about doing the same thing to the goals. Now, 

you don't have to answer right now, but -

W. Sproat, Exelon: I would probably disagree with that, but that's okay. 

D. Powers, ACRS Member: Explain why you would disagree other than the fact that 

you wouldn't like the numbers when they came out. 

W. Sproat, Exelon: No. What would the basis be for doing that? For example, in 

airline travel there's a certain risk associated with flying on an airplane. Now, the fact that them are increasing numbers of airplanes in the air doesn't necessarily mean that 

your risk of being killed on an airplane has proportionally increased. 

G. Apostolakis, Chairman, ACRS: The societal risk has. 

D. Powers, ACRS Member: Right. 

G. Apostolakis, Chairman, ACRS: The individual risk has not.

74
T. Kress, Chairman, Subcommittee on Future Reactors: You don't fly the same 

number of people on the airplanes. What you have is a site with a given fixed 

population around it, for example. That population is exposed to either one module or 

ten modules who could fail independently of each other, and in fact that's probably the 

assumption. But the risk of being on that site and associated with those reactors is, in 

my mind, ten times when you have ten modules over one module. 

D. Powers, ACRS Member: Isn't it even higher than that because you've got a mode 

failure with theT. Kress, Chairman, Subcommittee on Future Reactors: Yes. And then if there's 

common mode failures, it's even higher. 

D. Powers, ACRS Member: Especially if you go up -

T. Kress, Chairman, Subcommittee on Future Reactors: And that would be the 

reasoning behind -

D. Powers, ACRS Member: to a centralized control room? 

T. Kress, Chairman, Subcommittee on Future Reactors: Yes. So you treat it as 

one reactor, but in order to accommodate the ten of them you have to do something to 

one end; you either up the frequency by ten or the lower safety goal by -

W. Sproat, Exelon: Well, then clearly you have to take into account in that kind of an 

analysis the concept of coincident events happening in multiple units at the same time. 

T. Kress, Chairman, Subcommittee on Future Reactors: No, no, that's not -

D. Powers, ACRS Member: It's just common mode failure is what we are talking about 

here. 

T. Kress, Chairman, Subcommittee on Future Reactors: But that's not what I had in 

mind. 

W. Sproat, Exelon: Assuming there is a common mode failure that -

D. Powers, ACRS Member: But that's not what we're saying. 

T. Kress, Chairman, Subcommittee on Future Reactors: Yes, but that's not what 

we're saying. I mean, that's another issue, coincidence events and common mode 

failures. No, I'm not just talking about an independent frequency of something 

happening to one or something happen to the other independently. 

W. Shack, ACRS Member: Of course, now he does get something back because he 

probably has a smaller source term.

75
T. Kress, Chairman, Subcommittee on Future Reactors: Oh, I think that's a -- for 

this concept, that's -

G. Apostolakis, Chairman, ACRS: I didn't say anything about the assessment. 

G. Apostolakis, Chairman, ACRS: I'm just talking about the goals. 

T. Kress, Chairman, Subcommittee on Future Reactors: I'm sure they could meet 

the ten times or the ten percent -

G. Apostolakis, Chairman, ACRS: You don't use a facility of ten PBMRs only on 

these things. 

L. Parme, General Atomics: George, I might add in the mid-80s submittal on the 

MHTGR where there were multiple reactors coupled to a common steam plant, it was 

viewed as a plant and we took the safety goals and the release limits that we were 

analyzing it and considered multiple reactors. In fact, if you look back in the mid-80s submittal you'll see there is at least one event that has all four MHTGR models leaking simultaneously without cooling, and it was handled that way. 

It's not quite the case where these reactors are truly independent, but we did consider 

the four modules to be a plant consistent with your thinking. What would you do with 

truly independent modules, I guess, is something that one might want to think of. 

G. Apostolakis, Chairman, ACRS: If we decide, for example, that the appropriate way to formulate the goals here would be through frequency/consequence curves, then it 

seems to me that you would have one such curve or a family of curves for the facility. 

J. Garrick, ACNW Chairman: Yes. Well, why wouldn't you have a CCDF for the 

facility? 

G. Apostolakis, Chairman, ACRS: For the facility, that's what I'm saying. 

J. Garrick, ACNW Chairman: Every time you add a module, you get a new CCDF. 

T. Kress, Chairman, Subcommittee on Future Reactors: Yes, absolutely. 

G. Apostolakis, Chairman, ACRS: Anyway, that's just a point. 

T. Kress, Chairman, Subcommittee on Future Reactors: But it's a thought. 

W. Sproat, Exelon: Understood. 

P. Gunter, Nuclear Information Resource Service: Obviously fuel integrity is a big question here. What I would like to get a little better idea of, is have you looked at the

76
IRIS SUMMARY

Mario D. Carelli 

Westinghouse Science & Technology 

The IRIS (International Reactor Innovative & Secure) reactor is described in the 

first part of the presentation. IRIS is a light water cooled reactor with an integral 

configuration, where steam generators, pumps and pressurizer are inside the 

reactor vessel. Partially funded by the DOE NERI program, IRIS is being 

developed by an international consortium of 16 organizations from seven 

countries. A key IRIS characteristic is its "safety by design" approach which 

strives to eliminate, by design, as many accidents as possible rather than coping 

with their consequences. Initial returns are very positive; out of the eight Class 

IV accidents considered in the AP600 only one remains as a Class IV in IRIS, 

and at much reduced probability. Small-to-medium LOCAs have minimal 

consequences as the core remains safely under water for days, without the need 

for safety injection or water makeup. In spite of its novelty IRIS is firmly 

grounded on proven LWR technology and therefore a prototype is not needed to 

assure design certification. Rather, very extensive scaled tests will be performed 

to investigate the performance of in-vessel components such as steam 

generators and pumps, both individually and as interactive systems. Accident 

sequences will also be simulated and tested to prove IRIS safety by design 

claims. The first core fuel is less than 5% enriched and the fuel assembly is very 

similar to existing PWR assemblies, so there is no licensing challenge regarding 

the fuel. Because of the safety by design approach, yielding simplifications in 

design and accident management (e.g., IRIS does not have an emergency core 

cooling system), some accident scenarios are eliminated and others have lesser 

consequences. Thus, simplification and streamlining of the regulatory process 

might be possible. Risk informed regulation will be coupled with safety by design 

to show lower accident and damage probabilities. This could lead to a relaxation 

of siting regulatory requirements. It is believed that the IRIS licensing process 

will not present major challenges, actually it might be simplified. An aggressive 

schedule could lead to deployment early in the next decade. It is imperative, 

however, that tests, planning, and pre-licensing activities start as early as FY02.

77
78

IRIS 

International Reactor Innovative 

and Secure 

M. D. Carelli 

Westinghouse Science & Technology 

ACRS Subcommittee Workshop on 

Advanced Reactors 

June 4, 2001 

W101 ( Westinghouse Science v-7.ýh• & Technology

OUTLINE 

- Overview 

- Team Partnership 

- Funding 

- Schedular Objectives 

- Fuel Designs 

- Configuration (Integral vessel, internal shield, 

steam generators) 

* Enhanced Safety Approach (Safety by Design) 

* Maintenance Optimization 

* Issues 

* Conclusions 

v, (• Westinghouse Science 

v-.7N* 2 t &Technology
79

OVERVIEW 

SWestinghouse Science 

6,3. t & Technology

IRIS is a Modular LWR, with Emphasis on Proliferation 

Resistance and Enhanced Safety 

"* Small-to-medium (100-300 MWe) 

power module 

"* Integral primary system 

" 5- and 8-year straight burn core 

"• Utilizes LWR technology, newly 

engineered for improved 

performance 

"* Most accident initiators are 

prevented by design 

"* Potential to be cost competitive 

with other options 

"* Development, construction and N 

deployment by international team 

"* First module projected 

deployment in 2010-2015

SWestinghouse Science w*-7. =, & Technology
80

IRIS AND GENERATION IV GOALS 

GOAL 

Design feature Sustainable devlopent : Safe con mic and Economics ' 

F developmentReablt • ~Reliabilit 

Modular design " / 

Long core life (single bum, no shuffling) i / i , 

Extended fuel bumup ", 

Integral primary circuit , / , 

High degree of natural circulation 

High pressure containment wih inside

the-vessel heat removal 

Optimized maintenance V z , 

Attractive Commercial Market Entry 

() Westinghouse Science 

& Technology
IRIS Consortium Members

Team Member Function Scope 

Engineering Supplier Development 

Westinghouse Electric LLC, USA * * Overall coordination, leadership 

and interfacing, licensing 

Polytechnic Institute of Milan, Italy (POLIMI) Core design, in-vessel thermal 

hydraulics, steam generators, 

containment 

Massachusetts Institute of Technology, USA (MIT) Core thermal hydraulics, novel fuel 

rod geometries, safety, 

maintenance 

University of California at Berkeley, USA (UCB) * Core neutronics design 

Japan Atomic Power Company, Japan (JAPC) * * Maintenance, utility feedback 

Mitsubishi Heavy Industries, Japan (MHI) * * * Steam generators, modularization 

British Nuclear Fuels plc, UK (BNFL) * * * Fuel and fuel cycle, economic 

evaluation 

Tokyo Institute of Technology, Japan (TIT) Novel fuel rod geometries, detailed 

3D T&H subchannel 

characterization, PSA 

Bechtel Power Corp., USA (Bechtel) * * Balance of plant, cost evaluation, 

construction 

University of Pisa, Italy (UNIPI) * Containment analyses, transient 

analyses 

Ansaldo, Italy * * Steam generators, reactor systems 

National Institute Nuclear Studies, Mexico (ININ) * Core neutronics 

NUCLEP, Brazil * * Containment, vessel, pressurizer 

ENSA, Spain * * Reactor internals, steam 

generators, vessel 

Nuclear Energy Commission, Brazil (CNEN) * * Transient, structural analyses, 

(Pending) testing 

Oak Ridge National Laboratory, USA (ORNL) * * Core analyses, safety, cost 

evaluation, diagnostic 

Associates 

University of Tennessee, USA * Modularization, transportability 

Ohio State University, USA * Novel In-Core Power Monitor 

Iowa State University (Ames Lab), USA * NDE
FUNDING

DOE NERI - $1.6M over 3 years 

(9/99 - 8/02)

Consortium Members - $4M 

- $8M 

$10-12M

VeW:ol B

in 2000 

in 2001 

anticipated in 2002

(\ Westinghouse Science 

& Technology

IRIS SCHEDULAR OBJECTIVES

"* Assess key technical & economic 

feasibilities (completed) 

"* Perform conceptual design, 

preliminary cost estimate 

"* Perform preliminary design 

"* Pre-application submitted

End 2000 

End 2001 

End 2002 

?,

"* Decision to proceed to commercialization End 2002 

"* Complete SAR 2005

"* Obtain design certification 

"* First-of-a-kind deployment

6/401 

V• 9 4~ 9

2007 

2010-2015

Westinghouse Science 

& Technology

82
83

IRIS FUEL DESIGN OPTIONS 

IRIS 5-YEAR DESIGN FIRST CORE 

CURRENT FUEL TECHNOLOGY 

PROVIDES MINIMUM-RISK PATH FORWARD 

(DETAILED CORE DESIGN IN PROGRESS) 

IRIS 8-YEAR DESIGN RELOADS 

BOTH U0 2 and MOX MAY BE USED 

EMPHASIZES PROLIFERATION RESISTANCE 

(SCOPED INTERCHANGEABLE CORE DESIGN) 

(• Westinghouse Science 

v..v .o "0 & Technology

CONFIGURATION 

® Westinghouse Science 

& Technology
84

335 MWe LAYOUT 

SWestinghouse Science 

,*-' & Technology

INTERNAL SHIELDS 

0 A "gift" of integral configuration 

* Dose rate outside vessel surface as low as 

10-6 mSv/h 

* No restrictions to workers in containment 

* Simplified decommissioning 

* Vessel (minus fuel) acts as sarcophagus 

( ) Westinghouse Science V,9Vph 13 & Technology
HELICAL STEAM GENERATOR 

* LWR and LMFBR experience 

* Fabricated and tested 

* Test confirmed performance (thermal, 

pressure losses, vibration, stability) 

* 8 SGs practically identical to Ansaldo 

modules will be installed in IRIS 

SWestinghouse Science 

&T echnology

85
86

ENHANCED SAFETY APPROACH 

(Safety by Design) 

,,) Westinghouse Science v •- -h 16 & Technology

SAFETY PHILOSOPHY 

" Generation II reactors cope with accidents 

via active means 

" Generation III reactors cope with accidents 

via passive means 

" Generation IV reactors (IRIS) emphasize 

prevention of accidents through "safety by 

design" 

(• Westinghouse Science Vw=g,,K 17 & Technology
87

IRIS SAFETY BY DESIGN APPROACH 

Exploit to the fullest what is offered by IRIS 

design characteristics (chiefly, integral 

configuration and long life core) to: 

"• Physically eliminate possibility for 

accident(s) to occur 

"• Lessen consequences 

"* Decrease probability of occurrence 

V("' Westinghouse Science 

v,-,s ' a 9 & Technology
IMPLEMENTATION OF IRIS SAFETY BY DESIGN 

Design Characteristic Safety Implication Related Accident Disposition 

Integral reactor No external loop piping Large LOCAs Eliminated 

configuration 

Tall vessel with elevated Can accommodate internal Reactivity insertion due to 

steam generators control rod drives control rod ejection Can be eliminated 

High degree of natural Either eliminated (full natural 

circulation circulation) or mitigated 

LOFAs (e.g., pump seizure consequences (high partial 

or shaft break) natural circulation) 

Low pressure drop flow N-1 pumps keep core flow 

path and multiple RCPs above DNB limit, no core 

damage occurs 

Primary system cannot SGTR Automatic isolation, accident 

High pressure steam over-pressure secondary terminates quickly 

generator system system 

No SG safety valves Reduced probability 

required Steam and feed line breaks Reduced consequences 

Once through SG design Low water inventory 

Long life core No partial refueling Refueling accidents Reduced probability 

Large water inventory Slows transient evolution Core remains covered with no 

inside vessel Helps to keep core covered Small-medium LOCAs safety injection 

Reduced size, higher Reduced driving force 

pressure containment through primary opening 

Inside the vessel heat 

removal

00 

co
89

AP600 CLASS IV ACCIDENTS AND IRIS RESOLUTION 

Accident I IRIS Safety by Design IRIS Resolution 

1 Steam system piping failure Reduced probability Can be 

(major) Reduced cobabieity reclassified as 

2. Feedwater system pipe break Reduced consequences ,Class III 

SReactor coolant pump shaft Can be 

seizure or locked rotor Reduced consequences I reclassified as 

4. Reactor coolant pump shaft Class Ill 

break 

S Not applicable 

5Spectrum of RCCA ejecaon Can be eliminated (with internal 

5. accidents I _CRDMs) 

Can be 

6. Steam generator tube rupture Reduced consequences reclassified as 

__Class III 

7. Large LOCAs Eliminated I Not applicable 

SDesign basis fuel handling Stil Class IV 

8.accidents Reduced probability 11/3-1/5 lower 

_accidents_ !probability 

c• Westinghouse Science 

v& Technology

IRIS CONTAINMENT 

"* It performs containment function 

plus 

"* In concert with integral vessel, it practically 

eliminates LOCAs as a safety concern 

On first principles 

Pressure differential (driving force through rupture) 

is lower in IRIS because 

" Containment pressure higher (lower volume, 

higher allowable pressure) 

"• Vessel pressure lower (internal heat removal) 

(Q Westinghouse Science 

6u ,Z/ & Technology
AP600/IRIS Containment Size Comparison

MWe IRiS CONTAINMENT (25 meter diameler)

100MWe IRIS CONTAINMENT (20meter dameter)

VV~ph22 SWestinghouse Science & 

Technology Department

90

ANALYSES PERFORMED 

"* Break size: 1, 2, 4" 

"* Elevation: Bottom of vessel, above core 

(inside and outside cavity), 12.5 m above 

bottom 

* No water makeup or safety injection 

* Three codes provided consistent results 

- Proprietary (POLIMI) 

- GOTHIC (Westinghouse) 

- FUMO (Univ. Pisa) 

(,Y( Westinghouse Science 

\.;' & Technology
91

REACTOR VESSEL/CONTAINMENT PRESSURE 

DIFFERENTIAL EQUALIZES QUICKLY 

900 1900 2900 3900 4900 

2500 3 

2000 2 

a 1500 

O.S 

"0.5 

0 5000 10000 15000 20000 25000 30000 35000 40000 

Trme [s] 

Westinghouse Science 

V,•.V, 2. 0 &Technology

CORE STILL UNDER 2 METERS OF WATER AFTER 2 DAYS 

10 

4" Break, 12.5m high 

No Gravity Make-Up I 

-Z 7 - -- - - - - - - - - - - -

0 Liquid Level in the Reactor T 

- Tbp---6-ofte core ------- i 

4-,i 

0 0.5 1 1.5 2 2.5 

T401rime (days) Westinghouse Science 

V-V,, 2S (\- & Technology
A LICENSING CHALLENGE 

".".... simultaneous loss-of-coolant accident, loss of residual heat removal 

system, and loss of emergency core cooling ..... PMBR can meet that 

challenge ..... but "you can't assume that sequence for any LWR" even 

advanced units ....... 

Nucleonics Week 5/10101 Pg. 10 

IRIS CAN MEET THAT CHALLENGE

"• Loss of coolant accident 

"* Loss of residual heat removal system 

"* Loss of emergency core cooling 

V61.1. 26

Safety by design 

Three independent 

diverse systems 

Not needed 

(gravity makeup 

available anyway) 

SWestinghouse Science 

& Technology

92

MAINTENANCE OPTIMIZATION 

V(• Westinghouse Science 

& Technology
SURVEILLANCE STRATEGY 1j

"defer if practical, perform on-line when possible, and 

eliminate by design where necessary"

Design where necessary. 

"* Utilize existing components 

"* Utilize existing technologies 

"• Request rule changes 

"• Develop new components/systems 

"* Develop new technologies

V-/Ql

Dir, .ction of 

increasing cost, 

desiln effort, 

a d3r sk

(• Westinghouse Science 

(t & Technology

93

GOAL 

• Perform maintenance shutdowns no sooner 

than 48 months 

t• Westinghouse Science 

n (t & Technology

3
EXTENDED FUEL CYCLE

o Study completed in 

1996 investigated 

extending PWR to 

48 month cycle 

Recategorized all off

line maintenance as 

either: 

- Defer to 48 months 

- Perform on-line 

- Unresolved

0 0 

C.,M 

0 e 0 

2 

C. 

a) 

C.

PROJECT 

PWR Surveillance Program 

Comparison

0 1000 2000 3000 4000

E3 Unresolved N On-line 0 Off-line 

() Westinghouse Science 

( & Technology v.•3a

94

Q THE BOTTOM LINE 

• IRIS must utilize components and systems 

which are either accessible on-line for 

maintenance or do not require any off-line 

maintenance for the duration of the 

operating cycle 

° IRIS must utilize high reliability components 

and systems to minimize the probability of 

failure leading to unplanned down-time 

during the operating cycle 

(0) Westinghouse Science 

&Technology

3
95

ISSUES 

SWestinghouse Science V-.•l 32 & Technology

DEVELOPMENT APPROACH 

"* No need for prototype since no major 

technology development is required 

"* First-of-a-kind IRIS module can be deployed 

in 2010 or soon after 

"* Future improvements can be implemented 

in later modules (Nth-of-a-kind) 

Westinghou se Science 

r..7ý 33& Technology
96

LICENSING CHALLENGES AND OPPORTUNITIES 

VS. GEN II REACTORS 

"* First core fuel well within current state of the art 

"* Reload, higher enrichment fuel (post 2015) handled through 

licensing extension 

"* IRIS does have containment which in addition to its classic 

function is thermal-hydraulically coupled with integral vessel to 

choke small/medium LOCAs 

"* Safety by design approach eliminates some accident scenarios 

and significantly diminishes consequences of others. 

Simplification and streamlining possible. 

"* Risk informed regulation will be coupled with safety by design to 

show lower accidents and damage probabilities 

"* How can we translate IRIS improved safety into licensing 

opportunity, e.g., site requirements relaxation? 

"• Are regulatory changes necessary to accommodate extended 

maintenance? 

"° Multiple modules plants with common functions, e.g., control 

room Westinghouse Science 

a't & Technology

IRIS APPROACH TO LICENSING, CONSTRUCTION 

AND OPERATION VS. GEN II REACTORS 

" Licensing 

- No unique major changes identified at this time 

- Testing to confirm IRIS unique traits (safety by design, integral 

components, maintenance optimizations, inspections) 

"* Construction 

- Modular fabrication and assembly 

- Use of advanced EPC tool sets (Bechtel) 

- Multiple, parallel suppliers 

- Staggered modules construction 

"* Operation 

- Extended cycle length straight bum 

- Maintenance shutdown intervals no shorter than 48 months 

- Refueling shutdowns every 5 to 10 years 

- Reduced number of plant personnel 

- Multiple modules operation 

(-w-) Westinghouse Science 

V asn- 3 & Technology
97

DO SCHEDULES SUPPORT PLANNED 

LICENSE APPLICATIONS/DEPLOYMENT? 

Achieving 2007 design certification requires: 

"• Lead testing (safety by design) be initiated in 2002 

"* IRIS Consortium members decision by end 2002 to 

pursue commercial effort 

"• Continuous NRC interaction beginning late 

2001/early 2002 

Achieving early deployment (2010 or soon after) 

requires US generator interested by 2005 

) Westinghouse Science 

V-7V, M "& Technology

SUMMARY AND CONCLUSIONS 

- IRIS specifically designed to address Gen IV 

requirements 

* Modularity and flexibility address utility needs 

- Enhanced safety through safety by design and 

simplicity 

* IRIS is based on proven LWR technology, newly 

engineered for improved performance 

- Testing program needs to start in 2002 on selected 

high priority tests. Early interaction with NRC and 

ACRS will be extremely beneficial. 

Westinghouse Science 

yes-W 37 & Technology
T. Kress, Chairman, Subcommittee on Future Reactors: Would your SAR follow the 

SAR process that we use now for light water reactors? 

M. Carelli, Westinghouse: Yes. When the issue is safety, I think it should be 

simplified. Should be a simplified SAR. We'll see. 

T. Kress, Chairman, Subcommittee on Future Reactors: When you change out the 

core, do you also change out the steam generator? 

M. Carelli, Westinghouse: No. I'm coming to the steam generators. The IRIS steam 

generators are based on the helical steam generators which Ansaldo designed for 

Super Phenix. They tested the steam generators and the picture in viewgraph number 

14 shows a 20 megawatt -- steam generator mockup. They tested it. Now, I want to 

make a point. The perception is we have so much trouble with steam generators. This 

crazy guy wants to put the steam generator inside the reactor and this makes even 

worse. But there are things you have to think about. 

First of all, if you put a steam generator inside, the primary fluid is now outside the 

tubes so the tubes are in compression instead of tension. Now, you don't have any 

more of the tensile stress cracking. Also, our IRIS steam generator doesn't have a 

bottom in terms of deposits. The bottom of the steam generator is the bottom of the vessel, and the chemistry is much better. So there are a bunch of things that the steam 

generator has a different environment in an integral reactor versus a loop reactor. 

What they did in Ansaldo, is that they tested the steam generators. First of all, there is 

experience with Super Phenix and the LMFBR experience. Then they designed an 

intergal LWR reactor with similar helical steam generators. So they fabricated, tested, and confirmed the performance and by some stroke of luck, The IRIS design is such 

that it has eight steam generators practically identical to the model Ansaldo had 

fabricated. This brings up another important item. What we have now, is eight steam 

generators for a total of 300 megawatts. So we're talking a high level of redundancy. That's exactly what we want to do because the steam generators have a very critical 

safety function and you are going to see in a second what it is. 

G. Leitch, ACRS Member: The reactor vessel in the drawing looks as though it's large enough to facilitate internal control rod drives. 

M. Carelli, Westinghouse: Absolutely. When I look at that geometry, it is a waste of a prime real estate to have all that room above the core full of control drivelines. The 

internal CRDMs are ideally set for integral reactor. Absolutely. 

G. Leitch, ACRS Member: The CRDMs are going to be internal? Has that decision 

been made?

98
M. Carelli, Westinghouse: I would like to have internal CRDMs. The present design 

shows the CRDMs as regular CRDMs because the technology has not yet been 

developed to the point out that we are comfortable in incorporating the internal CRDMs 

as the reference design. There are essentially two designs of internal CRDMs. One is 

electromagnetic driven internal CRDMs done by the Japanese. MHI is the one that's 

been testing for 10 years and again, MHI is one of our team members. The second 

design is hydraulically controlled rods. That is a solution chosen by the Argentinean, in 

the CAREM reactor, and also by the Chinese. The Chinese have a reactor in Beijing 

that is running right now, operating with internal CRDMs. 

So both technologies are not a far fetch. There are reactors already operating or being 

designed. What, right now, I do not know is which one is better. There are two. So I 

have to decide which one. 

G. Leitch, ACRS Member: If they are external, you haven't eliminated the rod ejection 

"•problem. If they're internal, you have introduced some new technology. 

M. Carelli, Westinghouse: Yes. You're absolutely right. The issue is whether a 

deployment by 2012 is compatible with incorporating internal CRDMs. However, we're 

not starting from scratch. It has been done. There has been 10 years work on that. 

What I need is about one or two years to select a technology. At that point, we'll see 

how long it takes to implement. Can we make it or 2012 or not? That will be the 

decision.

99
PAGE LEFT BLANK INTENTIONALLY 0

100
Presentation Summary 

GT-MHR 

U.S. and European technology provide the bases for the Gas Turbine - Modular 

Helium Reactor (GT-MHR). For more than 4 decades, High Temperature Gas

cooled Reactors (HTGRs) have been under development in multiple countries. 

Numerous prototypes and demonstration plants have been constructed and 

operated beginning with the Dragon plant in the early 1960s. At the time of 

these initial plants, the vision was one of scaling up the technology to large, 

steam cycle plants comparable to modern LWRs, thus benefitting from economy 

of scale. However, in the early 1980s, both in the U.S. and the Federal Republic 

of Germany, a shift in paradigm occurred. Smaller, modular plants offered 

simplification in safety design, shortened construction schedules, and 

incremental capacity addition. The MHTGR was the U.S. developed modular 

plant and underwent preapplication review by NRC. The GT-MHR represents a 

further refinement on this concept with the steam cycle being replaced by a 

closed loop gas turbine (Brayton) cycle. 

The reactor system is contained in a 3 vessel, side-by-side arrangement. The 

reactor and a shutdown cooling system are in one vessel, and the gas turbine 

based power conversion system, including the generator, in a second parallel 

vessel. A small horizontal vessel provides coaxial ducting of gas between the 

reactor and power conversion system. The entire nuclear unit is located in a 

below grade silo with service areas above. The silo provides containment and 

protection of the reactor but is not designed to hold pressure. Naturally 

circulating water or air in panels around the reactor vessel carry off heat radiated 

from the uninsulated vessel and provide reactor cavity cooling. 

A more detailed look at the system shows the compact arrangement of gas 

turbine, compressors, recuperator, heat exchanges, and generator. All rotating 

machinery is on a common shaft. A central feature of the U.S. modular reactors 

is the annular core. Notice that the annular arrangement provides a high surface 

to volume ratio and a relatively short conduction path between any fueled block 

and the vessel wall. Fueled blocks are stacked in three concentric rings with 

inert graphite blocks making up the inner and outer reflectors. Operating control 

rods are located outside the active core while startup control rods and channels 

for reserve shutdown pellets are located near the core center. 

Ceramic coated fuel is the key to the GT-MHR's safety and economics. A kernel 

of Uranium oxycarbide (or U0 2) is placed in a porus carbon buffer and then 

encapsulated in multiple layers of pyrolytic carbon and silicon carbide. These 

micro pressure vessels withstand internal pressures of up to 2,000 psi and 

temperatures of nearly 2,000 'C providing extremely resilient containment of 

fission products under both normal operating and accident conditions. The fuel 

particles are blended in carbon pitch, forming fuel rods, and then loaded into 

holes within large graphite fuel elements. Fuel elements are stacked to form the 

core.

+ GENERAL ATOMICS

101
As alluded to earlier, modular gas reactors and the GT-MHR represent a 

fundamental shift in reactor design and safety philosophy. Up through approximately 1980, HTGR development proceeded on a path of scaling up core size in the interests of economics. In the process of this scaleup, core power density was kept nearly constant while the LID ration was kept as close to unity 

as possible. As a consequence of this thinking, maximum accident temperatures increased well above the temperature capabilities of the fuel particles. This 

placed ever increasing reliance on engineered safety features to assure 

continued core cooling and to contain released fission products should this 

cooling be lost. The modular reactor represents a 180 degree turn around in 

design philosophy. From its inception, the modular design first addresses safety, sacrificing size and optimized nuclear geometry to ensure that regardless of 

cooling system operation or coolant boundary integrity, fuel temperatures will 

never exceed the point at which fission products would be released. Having first 

addressed safety with the inherent features available in the gas-cooled reactor, good economics are sought in the efficient Brayton cycle and plant simplification. 

Fuel particle testing in Japan, Germany, and U.S. has repeatedly demonstrated 

the high temperature resilience of coated particle fuel to temperature approaching 2,0000C. As an conservative design goal, GT-MHR has been sized 

to keep maximum fuel temperatures below 1,6000C during the limiting accident 

condition of lost coolant circulation, pressure, and all AC power. Like other 

reactor types, the GT-MHR has a negative temperature coefficient. But unique to reactors with an all refractory, high temperature core there a several hundred 

degrees of temperature margin in the core design to make full use of this 

feedback mechanism. 

The GT-MHR licensing builds on the mid-80s submittal to NRC for the steam 

cycle MHTGR. Because of the unique design approach, especially with regards to safety, employed, the a licensing approach returns to the basics goals and 

rebuilds a licensing frame work that includes conventional deterministic analyses and safety classification. However, this framework is derived recognizing that safety functions and the means to achieve them in this concept can be expected 

to differ from those of LWRs. Furthermore, they are derived using the 

systematic plant evaluation offered by PRA techniques. 

The GT-MHR is now being developed in Russia under a joint U.S./Russian 

Federation agreement aimed at the destruction of surplus weapons plutonium. In addition to the U.S. and Russia, the program is sponsored by Japan and 

France. Conceptual design is completed and preliminary design is on schedule 

for completion in early CY 2002. Startup of the first module is currently scheduled for 2009 with a 4 module plant scheduled for completion in 2015. 

The path to commercialization involves the importation of the Russia design with 

a U.S. designed LEU core replacing the Plutonium core. A licensing submittal 

would be prepared in the U.S. and submitted to NRC. The first U.S. module 

could be online approximately 1 year after the first Russian module.

+ GENERAL ATOMICS

102
In summary, the GT-MHR is rooted in decades of international HTGR technology 

development and builds on the mid-1 980s MHTGR experience. The design 

features optimization of characteristics inherent to high temperature gas reactors 

to achieve high thermal efficiency, and easily understood, assured safety. The 

international program facilitates the near term deployment of this concept. 

Presenter: 

Laurence L Parme 

Manager - Safety & Licensing 

General Atomics Company

+ GENERAL ATOMICS

103
104

ACRS WORKSHOP 

Regulatory Challenges for Future 

Nuclear Power Plants 

Gas Turbine - Modular Helium Reactor 

4 - 5 June 2001 

Laurence L Parme 

Manager: Safety & Licensing 

Power Reactor Division 

+ GENERAL ATOMICS

Presentation Outline 

"* Background and design description 

"* Key safety features 

"* Licensing approach 

"• Design status and deployment schedule 

"* Conclusions 

+ GENERAL ATOMICS
105

U.S. AND EUROPEAN TECHNOLOGY BASES FOR 

MODULAR HIGH TEMPERATURE REACTORS

BROAD FOUNDATION OF HELIUM REACTOR TECHNOLOGY

DEMONSTRATION OF EXPERIMENTAL. REATORS BASIC OTGR TECHNOLOGY 

)RAGON AVR PEACH BOTTOM 1 FORT ST. VRAIN THTR 

(U.K.) (FRG) (U.S A.) (U.S.A) (FRG) 963-76 1967 - 1W 1067 -1974 1976-1989 1986-1909

3D Arrangement of Plant 

Reactor equipment Pos•Iioner Refueling Reactor 

maintenance and machine 2uxiliaiy 

repoa biditag building 

n c 600 MW(t) - 285 MW(e) bu•ulc•ing * Power conversion 

baddoag system integrated in 

single vessel 

" Vented, below grade 

reactor building 

•L]• 1. • Continuously !I operating, natural 

Reactor •circulating, air cooled nrPoaector reactor cavity cooling 

Reoactor 

lecto bafu`&-g--- +W GENERAL ATOMICS
106

GT-MHR 

COMBINES 

MEL TDOWN-PROOF 

ADVANCED 

REACTOR 

AND 

GAS TURBINE 

BASED POWER 

CONVERS ION 

SYSTEM 

+ GENERAL ATOMICS

ANNULAR REACTOR CORE LIMITS FUEL 

TEMPERATURE DURING ACCIDENTS

36 X OPERATING SIDE REFLECTOR CONTROL RODS 

CORE BARREL BORATED PINS (TYP) 

REFUELING 

A R -PENETRATIONS 

ACTIVE CORE 102 COLUMNS 10 BLOCKS HIG'•• 

CONTROL RODS 

SIDE 18 X RESERVE RELECTO SHUTDOWN 

CHANNELS 

... ANNULAR CORE USES EXISTING TECHNOLOGY 

+ GENERAL ATOMICS L9I•DO)
107

CERAMIC COATED FUEL IS KEY TO 

GT-MHR SAFETY AND ECONOMICS 

Pyrolytic Carbon 

Silicon Carbide 

Porous Carbon Buffer 

Uranium Oxycarbide 

TRISO Coated fuel particles (left) are formed into fuel 

rods (center) and inserted into graphite fuel elements S ~(right). 

PARTICLES COMPACTS FUEL ELEMENTS 

+ VEHERAL ATOMICS 

L-029(5) 

4-14-94
108

MODULAR HELIUM REACTOR REPRESENTS A FUNDAMENTAL 

CHANGE IN REACTOR DESIGN AND SAFETY PHILOSOPHY 

4000 W 

3000 ---------- RADIONUCLIDE 3000 

LARGE HTGRs RETENTION IN [30o0MW(t)]/ FUEL PARTICLES i ~FSV/ U- .([842 MWMT)] 

PEACH BOTTOM U (115 Mw•] 

1000 1000 

MHR 

II• 

1967 1973 1900 1985 

CHRONOLOGY 

...SIZED AND CONFIGURED TO TOLERATE EVEN A SEVERE ACCIDENT 

GENERAL ATOMICS 

1-12-96

COATED PARTICLES STABLE TO BEYOND 

MAXIMUM ACCIDENT TEMPERA TURES 

1.0 

0.8 0 

z 

0 C P- 0 = 0.6 0 U. 

- S0.4 NORMAL PEAK 

TEMPERATURE 0 

U. MAXIMUM 

0.2 - DESIGN BASIS EVENT 

TEMPERATURE O 

00 

1000 1200 1400 1600 1800 2000 2200 2400 2600 

FUEL TEMPERATURE (

0C) 

L-25601) + GECENA2L ATOMICS 

7-28-94 

W-9
109

FUEL TEMPERATURES REMAIN BELOW DESIGN 

LIMITS DURING LOSS OF COOLING EVENTS 

1800 

Design Goal = 16001C 

- 1400 - Depressurized 

. 1200 . 

Pressurized - 1 

1000 

0 2 4 68 

Time After Initiation (Days) 

... PASSIVE DESIGN FEATURES ENSURE FUEL REMAINS BELOW 1600'C 

S+ GEAiERAL ATOMICS 

L-34193) 

11-16-94

PASSIVE SAFETY BY DESIGN 

" Fission Products Retained in Coated Particles 

- High temperature stability materials 

- Refractory coated fuel 

- Graphite moderator 

" Worst case fuel temperature limited by design 

features 

- Low power density 

- Low thermal rating per module 

- Annular Core 

- Passive heat removal .... CORE CAN'T MELT 

"° Core Shuts Down Without Rod Motion 

+ CENERAL ATOMICS
110

Licensing Approach Builds on 

Mid-80s Submittal to NRC 

The DOE MHTGR program in the mid-80's utilized a "clean 

sheet of paper" integrated approach to the conceptual 

design 

- utilized participant experience in PRA's of HTGRs 

- approach underwent a preapplication review by the NRC/ACRS 

Provided risk-informed MHTGR Licensing Bases 

- Top Level Regulatory Criteria 

- Licensing Bases Events 

- Equipment Safety Classification 

- Safety Related Design Conditions 

- Basis design criteria 

+ GENERAL A7OMICS

Bases for 

Top Level Regulatory Criteria 

• Direct statements of acceptable consequences or risks 

to the public or the environment 

"* Quantifiable statements 

"* Independent of plant design 

" Top Level criteria include 

- 51 FR1 30 individual acute and latent fatality risks 

5xl0 7

/yr and 2xl10 1yr, respectively 

- 10CFR50 Appendix I annualized offsite dose guidelines 

5 mrernlyr whole body 

- 10CFR100 accident offsite doses 

25 rem whole body and 300 rem thyroid 

- EPA-520/1-75-001 protective action guideline doses 

1 rem whole body and 5 rem thyroid 

+ GENERAL ATOMICS
111

Licensing Basis Events 

"* Off-normal or accident events used for demonstrating 

design compliance with the Top Level Regulatory Criteria 

"* Collectively, analyzed in PRAs for demonstrating 

compliance with the 51 FR1 30 safety goals 

"• Encompass following event categories 

- Anticipated Operational Occurrences 

- Design Basis Events 

- Emergency Planning Basis Events 

+ GENERAL ATOMICS

Ranges of Top Level Regulatory Criteria 

and MHTGR Licensing Basis Events

t- -| 10-4 Nt-3 111-2 sv wll- iI tal t 10 l i0) so, 

UK"0A W'tIt SWtI GAMMA DOSE AT FAI IRIt)tU 

+ GENERAL ATOMICS
112

Equipment Safety Classification 

* Safety related systems, structures, and 

components (SSC) are those performing required 

functions to meet 1 OCFR1 00 doses for DBEs

Retain Radionuclides in Fuel

ReGemove ore Heat Control 

MHTGR functions for 10CFR 100 focus 

on retention within fuel particles 

+ GENERAL ATOMICS

Licensing Bases Application 

to GT-MHR 

" The above process is generic and should be directly 

applicable to the GT-MHR 

" Prior application to the MHTGR did not reveal a large 

sensitivity to the power conversion system 

GT-MHR would be expected to have some different LBEs 

and therefore some differences in safety related SSC 

- potential for new initiating events with rotating 

equipment in primary system 

- potential for different consequences with higher core 

rating 

- LBEs involving water ingress very unlikely-no SGs 

+ GENERAL ATOMICS

Control Heat Generation
113

GT-MHR NOW BEING DEVELOPED 

IN INTERNATIONAL PROGRAM 

"* In Russia under joint US/RF agreement for 

destruction of surplus weapons Plutonium 

"* Sponsored jointly by US (DOE) and RF (Minatom); 

supported by Japan and EU 

"* Conceptual design completed; preliminary design 

complete early 2002 

+ GENERAL ATOM•CS
COMMERCIALIZATION PROGRAM

COMMERCIAL 

PROGRAM

INTERNATIONAL 

PROGRAM + 

TECHNOLOGY

URANIUM FUEL 

RATHER THAN 

PuFUEL

Plant construction can start in 5 years 

+ GENERAL ATOMICS

114

LIMITED ENGINEERING WORK REQUIRED

+ GENERAL ATOMICS
115

COMMERCIAL PROGRAM FOLLOWS 

INTERNATIONAL PROGRAM

+ GENERAL ATOMICS

SUMMARY 

"- GT-MHR 

- Rooted in decades of international HTGR technology 

- Builds on 1980's (MHTGR) experience 

"* Optimization of inherent gas-reactor features 

provides 

- High thermal efficiency 

- Easily understood, assured safety 

"* International program facilitates near term 

deployment 

+ GENERAL ATOMICS
T. Kress, Chairman, Subcommittee on Future Reactors: For light water reactors, the 

safety goal that you have is 5 X 10-7 for early fatalities. You hear statements like, well, that's for light water reactors because we can live with that number. We have some 

idea of what the uncertainty is in the determination of it. But because those 

uncertainties are pretty big, we hear statements like well, we're going to not let you do 

that all with preventing the core damage. We're going to make you have a containment 

because of uncertainties. There's no quantification in my mind of what that uncertainty level is where you no longer have to have a containment. How are you going to deal 

with that concept in the regulatory arena? 

L. Parme, General Atomics: I've heard those kind of questions multiple times. In the 

'80s, what we submitted first of all is we argued that the goal of the NRC should be to 

assure the safety of the public, environment if that be also the case, but the criteria for 

the top level regulatory criteria and going and giving me a criteria on core melt or core 

damage is not really telling me anything about how safe you want the public. I will 

admit they didn't full accept that response, but in the case of the high temperature gas cooled reactor, I'd come back in a second. Perhaps it's not such a concern if 

something like that were imposed on me. In all of the accidents -- and some of the 

accidents I plotted up there. You'll notice all of those things are less than a rem and 

typically they're on the order of tens of millirems. Some of those things include 

assuming that in the steam cycle plant we had lost all electric power on one module, took a break in a steam generator, lost our forced cooling, started pumping steam from 

one module back to the others for hours on end with nobody taking action. Those are 

still the kind of doses we got. There's no damage to the core. 

However, I will add, we mistakenly in the mid '80s said, what do you mean by core 

damage? There's no damage. The graphite will stand up to 5,000 degrees Fahrenheit 

or more before it starts to sublime. It won't be damaged. Well then they started 

redefining it as a dose over 100 millirem or something like that. I think the argument is 

tell me how safe you want me to be. If Generation IV or if these newer reactors are 

supposed to be quantitatively safer -

T. Kress, Chairman, Subcommittee on Future Reactors: I'm quite pleased to see 

your frequency consequence curves because some of us on the ACRS think that's a 

good way to go, particularly when you don't have core melts. 

The other question I wanted to ask you that may come up, I don't know. Chernobyl had 

a lot of graphite and it apparently burned. You have an air cooled cavity where you're encouraging natural convection. Is there an issue there? 

L. Parme, General Atomics: Let me say a couple of words. In the NRC interactions we 

had in the '80s, we did do some analysis of broken vessels, failed vessels, and air 

ingress. First of all, reactor grade graphite in the U.S., H451 for pebble bed modular 

reactor. I'm not sure what the grade is but typically the German graphites. They will not 

burn in the sense of a self-sustaining chain reaction.

116
D. Powers, ACRS Member: Why do you say that?

L. Parme, General Atomics: I will say that exactly as follows. Coal will burn, charcoal 

will burn because of its impurities. Reactor grade graphite -- and there's been tests 

done at Oak Ridge where an oxyacetylene torch was placed on the graphite. 

D. Powers, ACRS Member: You're talking of the difference between a point ignition 

and a homogeneous ignition. 

L. Parme, General Atomics: In the case where we analyzed air going into the core, 

and here I'll speak only of the blocks, the reaction rate is driven by temperature that is 

held up by decay heat. The heat generated from oxidation of the graphite was aboutand it's been 10 years -- but on the order of 10 to 20 percent of the total heat generated 

was -- in fact, 10 percent or less was due to oxidation. Also the reaction then becomes 

oxygen-limited as the air passes up the channels. We did an analysis assuming a 

vessel failure in that cross vessel that connects the two vessels and then assumed that 

the silo was open and you could get air in that. What you would get was air coming in 

the hot duct, going up through the core, down through the vessel and out the return 

duct. 

We did the analysis for about 24 hours and I think we did it beyond that but, once 

again, I'd have to go back and look at the calculations, though it is in Appendix G to the 

preliminary safety information document that was submitted. I think you see there's no 

increase in particle failures, but what you do is you are getting releases. They're pretty 

substantial because they're a driving force and the releases you're seeing and the 

doses that come with it are due to picking up the contaminants that are within the 

graphite. As you oxidize the graphite, there are contaminants there. They were -- I 

want to be careful about quoting the doses. I rather doubt that they stayed within the 

protection action guides for that accident. However, they were well within the limits of 

10 CFR 100. 

My comment on combustion was implying just primarily that the reaction is driven by 

decay heat. It's not as if you had a charcoal pile there. But you will oxidize. There's no 

question you will oxidize graphite. 

Incidentally, in the large HTGR, the approach to that, if you got a break and the primary 

cooling system got air in the system, it's a coolant. What you do is if you've got a 

circulator, you turn the circulator on and you cool the core with air. Once the core 

temperature is down, it will not oxidize so you just run the circulator. That was the 

design approach for the large HTGRs. If you had a circulator running, that's how you 

do it. You just turn the circulator on, blow the air around and cool it off.

117
General Electric Nuclear Energy

Prepared by ACRS Staff for A. Rao 

A. Rao of GE Nuclear Energy provided a presentation on the Evolutionary Simplified 

Boiling Water Reactor (ESBWR). The ESBWR is a 1380 MWe boiling water reactor 

with improved operating safety margins and passive safety systems. He stated that the 

ESBWR derived from earlier"GE plant design certification efforts and is the result of 

eight years of international cooperative work. He stated that the biggest challenge is to 

cross the regulatory hurdles associated with the inspections, tests, analyses, and 

acceptance criteria (ITAAC) and combined license (COL) programs. He further stated 

that he did not know how long it might take to license the ESBWR, in part, because the 

last GE design certification took about 8 to 10 years. Dr. Rao also provided a brief 

overview of the GE Nuclear Advance Liquid Metal S-PRISM design.

119
g GE Nuclear Energy

ESBWR Program and 

Regulatory Challenges 

Atam Rao 

GE Nuclear Energy, USA 

ACRS Workshop - Regulatory Chall g r q a 

June 4/5, 2001, Rockville, Maryland

120

Overview 

"* Design is based on SBWR and ABWR components 

Natural Circulation, ABWR Fuel, Vessel, CRD - just less 

Passive safety systems - based on NRC reviewed SBWR 

Optimized buildings/structures - economics/construction 

8 year international design and technology program 

Goal was to improve performance/safety and economics 

"* Regulatory Issues 

How much use can be made of SBWR review by NRC? 

Extensive new testing completed - is it enough? 

Is the regulatory hurdle too high for new plants?
Evolution of BWR Containments

Mark II Mark III 

ABWR SBWR

ESBWR 

Simpler Structures 

EY' Higher Margins 

Easier Construction 

Improved Economics 

Reference 

ESBWR ARW1034

121

Evolution of the BWR Reactor Design

ABWR ESBWR 

Evolution Towards Simplicity

Mark I
Comparison of Key Parameters

Paraee ABWRI BW

* Power (MWt) 

* Power (MWe)

ARD1O-.

122
123

ESBWR Program Plan

Requirements

Design

Technology

Licensing

FO PHASE 3 P2AE 4 I• :• i~ I!• 1• 1[2000-2002 Mt- ?I

ESBWR Design/Technology based on SBWR and ABWR 

too3
• Rais 

. Hig 

All1 

". Dec

Safety Systems Inside Containment Envelope 

sed Suppression Pool 

h Elevation Gravity Drain Pool 

Pipes/Valves Inside Containment 

ay Heat Condensers Above Drywell

124

Main steam

Feedwater 

Annulus 

D Saturated Water 

U Subcooled Water 

] Saturated Steam

Natural Circulation is y 

Standard BWR Technology
SBWR (670 MWe) ESBWR (1380 MWe)

Significant Reduction in Systems & BuildingsI

Bundle Power vs. Flow for various BWRs

20 

o 22 a 

Or

oo 2. 4.o .oo- &. 1o.. 12-o 1-00 -6o -00 2,20 

Avera&e Flow per Bundle (kgqs) 

POWFIO-2ý.2 011 

ESBWR has 100% flow margin to stability data boundary 

AR010! 12

125
Comparison of Plant Performance

Parameter 

Natural Circulation flow/bundle, kg/s 

Power/Flow Ratio, M WI(kg/s) 

Transient pressure rate, MPa/s 

Margin to SRV setpoint during isolation 

transient, MPa 

Minimum water level after accident, m 

above top of fuel 

Post accident containment pressure 

margin, KPa below design pressure

Typical 

BWR 

3.5-5 

0.25 

0.8 

valve 

opens 

0.0 

40

Passive BWR 

SBWR ESBWR 

8.5 10.6 

0.31 0.26 

0.4 0.4 

0.52 0.32 

1.5 2.8 

100 200

ESBWR Performance is Better Than or Equal to Most Plants I 

AJUI103 14

126

Natural Circulation Technology Program I I I I 

SSVWR ESBWR Phwe•I ESBVWRPase2 ESBWR PShase 3 

ATLAS Tests - 529 

smooth rs•t geometry I reduced pitch d.,..F ci 

(305 mm - 292 mm) •r ./CEA Chsney Tests 

OntaChm HydroTes Vo l scaled ESBWR conditions 

Ontario~~ Teter Plyo n chirnhey partition Stransiiert test (pump induced) -supported by EdF 

- round pipe (0.518 m ID) s tb Starto Floiw Oscilltilon - relatvey flat void distribubon PSI1 IRI Testig 

Oscilatiotn2 -• full range parameters -ESSWR condions CRIEPI Tests C=ge•imei -rbgenmrler scaling and other effecs• 

-single chimnney 

- SSW conditdons 

large margin to osclation regr-nee 

IRI / ETH Projects S -code development 

4•• 7 and analyses 

-studs pedorm'ed by PSI cor -he efect -su-pported by 

SWUtltiessW
Reactor pressure response to isolation events

127

0 10 20 30 40 50 

TIME (..-) 

ESBWR has slowerpressurization I 

No SRV opening I

Passive Safety Systems - Simplify the Plant 

* Reactivity Control 

Electro-hydraulic control rod drive system 

Accumulator driven backup boron injection system 

* Inventory Control 

Large vessel with additional inventory 

High pressure isolation condensers (IC) 

Depressurization and gravity driven cooling system (GDCS) 

* Decay Heat Removal 

Isolation condensers for transients 

Passive Containment Cooling System (PCCS) condensers for pipe 

breaks 

Fission Product Control and Plant Accident Release 

Passive condensers 

Retention and holdup with multiple barriers 

Simplified Systems Extending Operating Plant Technology AROlOO IC
Water Level in Shroud Following a Pipe Break 

10 

9 

7 

~-6 

S5- ABWVR ESBWVR 

> 4

- 1 JP PLANT 

........ ..... .... .... -1 

IPWUMP IrIJEC.TION 

-2. (ASWR) 

100 200 300 400 500 600 700 800 900 1000 1100 1200 PUMP INJECTION TIME AFTER PIPE BREAK (SEC) (JP PLANT)

128
Safety System (GIST) Test Facility 

and Depressurization Valve

Reactor Depressurization Valve in the Test Facility

129

ESBWR Decay Heat Removal 

"Remove Decay Heat From Vessel 

- Main Condenser 

- Normal shutdown cooling system 

- Isolation condensers 

- Remove vessel heat through valve opening 

"If Needed, Remove Heat From Containment 

- Suppression pool cooling 

- Containment sprays 

- Passive containment cooling (PCCS) condensers 

Several Diverse Means of Decay Heat Removal 

1 RIS2
JLT 

0- 

-

- ---~ 

o 0o 

4-, 

ku- S 0p
131

Extensive Technology Program to 

Qualify Features New to SBWR 

"- Component and Integral tests as part of the SBWR 

program 

- Full scale components tests - condensers, valves 

- Integral tests at different scales, with the largest test at 

PANDA 

"* Testing extended to incorporate European requirements 

- Large hydrogen releases and severe accidents 

- Improvements in the plant design 

"* Ongoing programs will further quantify margins 

- Natural circulation in the vessel 

- Severe accident performance/features for passive 

systems 

" Testing used to qualify computer codes 

"* Extensive international cooperation 

A Complete and Thorough Technology Programi 

Supports the Design

TEPSS Program 

Suppression Pool stratification and mixing 

- 9+ tests with flow visualization in LINX 

- CFD analysis using CFX 

"a Passive Decay Heat Removal 

- 8 Integrated system tests run in PANDA 

- Pre- and post-test predictions using TRACG, TRAC-BF1, 

RELAP5 and MELCOR 

"* Passive Aerosol Removal 

- PCCS testing in AIDA 

- Analysis with MELCOR 

- Demonstrate PCCS as fission product aerosol filter 

- Demonstrate ability of PCC to remove decay heat with 

aerosol build-up 

3 partprogram extended the SBWR database to ESBWR 

AROIO2- 24
132

Ongoing Simplification Studies 

" Reduce Fuel Bundles, CRD, Vessel - COMPLETE 

Increase Fuel Length 

" Improve Plant Availability - 5% 

Refueling and Outage Plan and System Improvements 

" Reduce Buildings and Structures - 30% 

Reduce Basemat Thickness 

Reduce Containment Design Pressure 

Move Spent Fuel Pool to Grade Elevation/Separate Building 

Separate Reactor Building From Containment 

Normal performance margins maintained while reducingi 

excessive conservatisms in other areas

Ongoing Technology Programs 

* Quantify Natural Circulation Performance Margins 

NACUSP Programs at IRI, NRG, CEA and PSI 

Additional Testing at IRI and CRIEPI 

Independent Stability Assessment at ETH, IRI 

* Reduce Uncertainty in Natural Circulation Parameters 

Chimney Tests at CEA 

* Develop Confidence in Safety System Performance 

TEMPEST Programs at PSI, VTT, NRG, CEA 

a Develop Back-up Systems to Provide Additional Margin 

TEMPEST Programs at PSI 

Technology programs to confirm that design is robust and 

provide additional data for code qualification I AROIQO- 26
133

Program Summary and Conclusion 

* 8 year ESBWR program 

Reduced Components and Systems -simplify 

Reduced the Structures and Buildings - simplify 

* 8 year Technology Studies 

Large margins confirmed - increased over SBWR 

Qualified codes for incremental changes for ESBWR 

* Challenges for the Coming Years 

Crossing the regulatory minefield? hurdles? resources? 

Improved Safety/Performance and Economics 

Completed Extensive Technology Program 

SBWR and ABWR Programs ease Regulatory Challenges 

AR010 V7