ML22342B278

Report - White Paper on the Enduring Legacy of ACRS: Contributing to Safety-Licensing Review of Reactor Facilities

ML22342B278
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Issue date:	12/31/2022
From:	Hossein Nourbakhsh Advisory Committee on Reactor Safeguards
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The Enduring Legacy of ACRS:

Contributing to Safety-Licensing Review of Reactor Facilities Prepared by:

Hossein P. Nourbakhsh Senior Technical Advisor for Reactor Safety December 2022 Advisory Committee on Reactor Safeguards U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

ABSTRACT Since 1957, the Advisory Committee on Reactor Safeguards (ACRS) has had a continuing statutory responsibility for providing independent reviews of, and advising on, the safety of proposed or existing reactor facilities and the adequacy of proposed reactor safety standards in the United States. This white paper begins with a history of the ACRS, noting some of its significant contributions to reactor safety. The paper then presents a historical perspective on ACRS reactor licensing reviews. The essential role of the ACRS on reviewing the currently proposed new reactor designs, which are radically different from the current fleet of light water reactors, is also discussed.

The views expressed in this report are solely those of the author and do not necessarily represent the views of the ACRS.

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TABLE OF CONTENTS ABSTRACT ..................................................................................................................... ii ABBREVIATIONS ... iv

1. INTRODUCTION ..... 1

2. A BRIEF HISTORY OF ACRS ........3 2.1 Creation of ACRS ................................................................................................ 3 2.2 ACRS as a Statutory Committee .... 3 2.3 Role of ACRS Over Its History ........... 5

3. REACTOR LICENSING REVIEWS, A HISTORICAL PERSPECTIVE .............. 8 3.1 Early Years of Reactor Licensing Reviews ... 8 3.2 Licensing Reviews of Early High-Power LWRs ........... 10 3.3 Reviews of Latter Non-LWR Designs . 12 3.4 Pre-application Reviews of Earlier Advanced Non-LWR Designs .. 16 3.5 Licensing Reviews of Advanced Light Water Reactors .... 19

4. LOOKING AHEAD TO ACRS LICENSING REVIEWS OF NEW ADVANCED REACTOR DESIGNS . 23 4.1 Essential Role of ACRS ... 23 4.2 Enhancing the Efficiency of the Review Process . 23

5.

SUMMARY

AND CONCLUSIONS .27

6. REFERENCES .... 28 iii

TABLES

1. Early U.S. Nuclear Power Reactors .. 9

2. The ACRS Review of Latter Non-LWR Designs . 14

3. The ACRS Pre-application Reviews of Non-LWR Designs . 17

4. The Design Certification Applications Reviewed by the ACRS . 20 iv

ABBREVIATIONS ABB-CE ASEA Brown Boveri - Combustion Engineering ABWR Advanced Boiling Water Reactor ACNW Advisory Committee on Nuclear Waste ACNW&M Advisory Committee on Nuclear Waste and Materials ACRS Advisory Committee on Reactor Safeguards AEC Atomic Energy Commission AP600 Advanced Passive 600 AP1000 Advanced Passive 1000 ATWS Anticipated Transients Without Scram BWR boiling water reactor CP construction permit CRBRP Clinch River Breeder Reactor Plant DCA design certification application DI&C digital instrumentation and control DOE Department of Energy ECCS emergency core cooling system ESBWR Economic Simplified Boiling-Water Reactor FACA Federal Advisory Committee Act FFTF Fast Flux Test Facility GCFBR gas cooled fast breeder reactor GDC General Design Criteria GE General Electric GENE General Electric Nuclear Energy GSI Generic Safety Issue HTGR high temperature gas-cooled reactor LMR liquid metal reactor LOCA loss-of-coolant accident LWR light water reactor MHTGR modular high temperature gas cooled reactor NPM NuScale Power Module NRC Nuclear Regulatory Commission NSRRC Nuclear Safety Research Review Committee v

PCCS passive containment cooling system PCRV prestressed concrete reactor vessel PHWR pressurized heavy-water reactor PRA probabilistic risk assessment PRDC Power Reactor Development Company PRISM Power Reactor Innovative Small Module PSER preapplication safety evaluation report PWR pressurized water reactor RG regulatory guide SAFR Sodium Advanced Fast Reactor SE safety evaluation US-APWR U.S. Advanced Pressurized-Water Reactor vi

1. INTRODUCTION The 1957 amendment to the Atomic

• Upon request from the Department of Energy Act of 1954 established the Energy (DOE) and with the consent Advisory Committee on Reactor of the Commission, ACRS provides Safeguards (ACRS) as a statutory advice on U.S. naval reactor designs, committee with its independent advisory and hazards associated with DOE role, and the responsibility to review nuclear activities and facilities.

safety studies and facility license applications... and to advise the U.S.

• Upon request and with the consent of Atomic Energy Commission (AEC) with the Commission, ACRS provides regard to the hazards of proposed or technical advice to the Defense existing reactor facilities and the Nuclear Facilities Safety Board (per adequacy of reactor safety standards. the National Defense Authorization Act of 1989).

With the enactment of the Energy Reorganization Act of 1974, the ACRS ACRS operations are governed by was assigned to the newly established the Federal Advisory Committee Act NRC with its statutory requirements (FACA), which is implemented through intact. NRC regulations (10 CFR Part 7). ACRS operational practices encourage the The ACRS consists of up to 15 public, industry, State and local members who are well recognized governments, and other stakeholders to experts in technical areas that are key to express their views on matters related to nuclear safety and with a breadth of safety of reactor facilities. All ACRS experience in all aspects of the nuclear records, reports, transcripts, or other enterprise: industry, universities. national documents, which are made available to laboratories, and government. or prepared for or by the Committee, are publicly available subject to the The present-day matters referred to provisions of the Freedom of Information the Committee can be grouped into five Act (5 U.S.C. 552) and NRCs Freedom categories: licensing reviews, regulatory of Information Act regulations at 10 CFR policies and practices, operating reactors part 9, sub-part A.

safety oversight, safety research reviews, and nuclear materials and Throughout its history, the ACRS waste. The scope of the present day review has been an important element of ACRS review activities include: the reactor licensing process. The Committees licensing reviews led to

• Specific statutory review functions evolution of many new safety established in NRC regulations requirements and design changes dealing with a wide range of technical

• Reviews of any generic issues or issues.

other matters referred to it by the Commission for advice As NRC is preparing for review of new reactor designs that are radically

• Reviews of specific generic matters different from the current fleet of LWRs, or nuclear facility safety-related the role of ACRS, with its diverse items, on ACRS own initiative (per technical expertise, will continue to be 10 CFR 1.13) 7

essential for integrated/multi-disciplinary on ACRS reactor licensing reviews. The independent review and advice. essential role of the ACRS on reviewing the new advanced non-light-water This white paper begins with a brief reactor designs is also discussed.

history of the ACRS, noting some of its significant contributions to reactor safety.

It then presents a historical perspective 8

2. A BRIEF HISTORY OF ACRS 2.1 Creation of ACRS Committee and the Industrial Committee on Reactor Location Problems were The history of ACRS goes back to combined by the AEC and the ACRS was 1947 when AEC soon after its formally born.

establishment recognized the need for an independent technical group to review and provide advice on reactor safety matters and thus a Reactor Safeguard Committee, chaired by Dr. Edward Teller, was established. Dr. Teller has been quoted saying that the Reactor Safeguard Committee was about as popular - and as necessary - as a traffic cop [Mazuzan & Walker, 1985]. As stated by Richard Meserve, former NRC Chairman, the Reactor Safeguard Committee clearly established an enduring characteristic of the ACRS - a willingness to provide candid views on Dr. C. Rogers McCullough First ACRS Chairman, 1953-1960 reactor safety issues, even at the risk of taking unpopular positions [NRC, 2003].

2.2 ACRS as a Statutory Committee In 1957, an amendment to the Atomic Energy Act of 1954 established the ACRS as a statutory committee advising the AEC. According to Section 29 of the Act the Committee shall review safety studies and facility license applications referred to it and shall make reports thereon, shall advise the Commission with regard to the hazards of proposed or existing reactor facilities and the adequacy of proposed reactor safety standards and shall perform other such Dr. Edward Teller duties as the Commission may request.

Chairman of Reactor Safeguard Committee, Subsection 182b of the Act requires that 1947-1953 the Committees report be made part of the record of the application and In 1950, the AEC established a available to the public, except to the second advisory committee, the extent that security classification Industrial Committee on Reactor prevents disclosure.

Location Problems, charged with the responsibility of advising on what we Establishing the ACRS as a statutory would today consider siting issues, committee resulted, in part, from a including seismic and hydrological controversy involving licensing of the characteristics of proposed sites. In Fermi-1 liquid metal fast breeder reactor.

1953, the Reactor Safeguards 3

An account of this, as reported by J. The AEC was unwilling to provide a Samuel Walker and Thomas R. Wellock copy of the ACRS report to the Joint

[Walker & Wellock, 2010], is summarized Committee without the condition that it here. would be kept administratively confidential. The AEC also refused to In 1956, the Power Reactor provide a copy of the ACRS report to the Development Company (PRDC), a State of Michigan on the grounds that it consortium of utilities led by Detroit would be inappropriate to disclose the Edison Company, applied for a contents of internal documents.

construction permit (CP) to build a fast Meanwhile, the AEC was completing its breeder reactor, located on Lake Erie review of PRDCs application. The AEC within 30 miles of both Detroit and took a more optimistic view of the safety Toledo. The fast breeder reactor that of the proposed reactor than ACRS had.

PRDC planned was far more advanced Since the company had agreed to in its technological complexity than the perform tests to answer the questions light-water designs proposed in earlier raised by ACRS, the AEC decided to applications. After review of PRDCs issue the construction permit. However, application and discussions with it acknowledged the ACRS concerns by company representatives, the ACRS inserting the word conditional in the concluded in an internal report to the construction permit to emphasize that the Commission that there was insufficient company would have to resolve the information available at the time to give uncertainties about safety before it could assurance that the PRDC reactor could receive an operating license.

be operated without public hazard.

ACRS also expressed doubt that its safety concerns could be resolved within PRDCs proposed schedule for obtaining an operating license. The ACRS urged that the AEC expand its experimental programs with fast breeder reactors to seek more complete data on the issues raised in the PRDC application.

During congressional hearings, members of the Joint Committee on Atomic Energy (the AEC congressional oversight committee) were troubled by Fermi Unit 1: the worlds first commercial liquid-revelations of safety concerns and the metal fast breeder reactor. Establishing the ACRS AEC Chairmans intention to attend the as a statutory committee resulted, in part, from a controversy involving licensing. Fermi 1 started groundbreaking ceremony for a reactor operation in 1963.In October 1966, it experienced whose construction permit was still being fuel melting from partial core flow blockage. Three evaluated by the AEC. They were years and nine months later, with cleanup particularly disturbed by the AECs failure completed and fuel replaced, Fermi 1 was restarted. In November 1972, the PRDC made the to inform them about the ACRS decision to decommission Fermi 1 for economic reservations. The AEC was obligated by reasons.

the Atomic Energy Act of 1954 to keep the Joint Committee fully and currently To prevent a recurrence of the AECs informed about its activities, and Joint conduct in the PRDC case, the Joint Committee members believed that, in the Committee soon introduced legislation to case of the ACRS report, the agency had establish ACRS as a statutory body, failed to carry out its responsibility. direct that its reports on licensing cases 4

be made public and require public report offered assurances about the hearings on all reactor applications. The reliability of ECCS designs and AEC opposed all three measures but improbability of a core meltdown, but it muted its objections because they were also acknowledged that a loss-of-coolant presented as amendments to a bill to accident (LOCA) could cause a breach of provide insurance coverage for reactor containment if the ECCS failed to owners, which the agency strongly perform. In an ACRS letter on the task favored. Establishment of the ACRS as force report, dated February 26, 1968, a statutory committee was accompanied the Committee strongly recommended by a significant expansion of public that a positive approach be adopted access into the regulatory and licensing toward studying the workability of activities of the AEC. protective measures to cope with core meltdown [ACRS, 1968a]. The 2.3 Role of ACRS Over Its History Committee also recommended, as it did in its 1966 report on safety research, that The role of ACRS has evolved over a vigorous program be aimed at gaining its history. The early licensing reviews better understanding of the phenomena were generally based on the engineering and mechanisms important to the course experience and judgment of regulatory of large-scale core meltdown. The task staff working closely with the ACRS, force report and ACRS without the availability of the regulatory recommendations formed the basis of guidance and structure established later some of the most important research during commercial development of initiatives and regulatory decisions by the LWRs. Most of todays U.S. nuclear AEC and the NRC, including the AECs power plants were licensed during the decision to undertake a study to estimate 1960s and 1970s, when both the the probability of a severe accident, technology and its governing regulations which resulted in the publication of the were in the formative stages. The landmark Reactor Safety Study (WASH-Committees licensing reviews led to 1400) [NRC,1975] and the beginning of evolution of many new safety the science of probabilistic risk requirements and design changes assessment as applied to nuclear power dealing with a wide range of technical plant safety.

issues [Nourbakhsh, 2018a]. The ACRS reactor licensing reviews are further As the ACRS moved into the 1980s, discussed in Section 3 of this white and continuing to the present day, the paper. Committee shifted much of its attention from plant design and construction to Early in the development of improvements in both the operation and commercial nuclear power, the ACRS regulation of nuclear power plants became concerned with core meltdown [Nourbakhsh, 2018b]. The ACRS has accidents, particularly one in which the made valuable contributions over a wide plants emergency core cooling system range of issues at operating plants, (ECCS) might fail to operate as including fire safety, operator training and designed, could lead to a breach of human performance, digital containment. In 1966, at the prodding instrumentation and control (DI&C) of ACRS, the AEC established a special upgrades, extended power uprates, plant task force to look into the problem of core aging, and license renewal.

meltdown [Walker & Wellock, 2010]. The task force, chaired by William K. Ergen, a A list of generic items related to former ACRS member, issued its report construction or operation of light-water in October 1967 [Ergen, 1967]. The reactors, which was developed initially by 5

the ACRS, and the work of the NRC staff to resolve those items became steadily more formal, stemming from the Some of the most significant requirement of Section 210 of the Energy successes of the Nuclear Reorganization Act of 1974 which Regulatory Commission were required the NRC to develop a plan achieved in large part with the providing for the specification and benefit of the wise counsel -

analysis of unresolved safety issues perhaps even the prodding - of relating to nuclear reactors and take such action as may be necessary to the ACRS implement corrective measures with Excerpt from a speech by Richard Meserve, respect to such issues. The ACRS has former NRC Chairman, before a symposium made significant contributions toward honoring the 50th anniversary of the ACRS, resolution of many generic safety issues March 4, 2003 (GSIs). One recent example is the Committees role in the resolution of GSI-191, Assessment of Debris Accumulation on PWR Sump Performance. ACRS was first to express concerns about the effects of In the early 1990s, the ACRS chemical reaction products and became concerned about the particle/fiber mats that could form on inconsistent use of PRA in the NRC. In a screens. The Committee was also the July 19, 1991, letter on the consistent use first to recognize that increasing screen of PRA [ACRS, 1991], the ACRS area, though it could reduce head loss, acknowledged, PRA can be a valuable might result in more fiber debris passing tool for judging the quality of regulation, through the screens and increase and for helping to ensure the optimal use downstream effects [Nourbakhsh & of regulatory and industry resources.

Banerjee, 2011]. The Committee also stated that it would have liked to see a deeper and more The ACRS was at the forefront of the deliberate integration of the methodology development of quantitative safety goals. into the NRC activities. The ACRS also In its May 16, 1979, letter on quantitative pointed to issues such as the safety goals [ACRS, 1979], the ACRS inconsistent use of conservatism and the recognized the difficulties and lack of the treatment of uncertainties. In uncertainties in the quantification of risk response to the ACRS, NRC chartered a and acknowledged that in many PRA Working Group and a Regulatory situations engineering judgment would Review Group to review processes, be the only or the primary basis for a programs, and practices to identify the decision. Nevertheless, the Committee feasibility of substituting performance-believed that the existence of quantitative based requirements and guidance safety goals and criteria could provide founded on risk insights in place of important yardsticks for such judgment. prescriptive requirements [NRC, 2006].

The first set of trial goals (NUREG-0739) These efforts led the Commission to

[ACRS, 1980] was developed by the issue a policy statement on the use of ACRS in 1980. These safety goals were PRA so that the many potential the basis for later NRC work on the applications of PRA can be implemented development of an NRC Safety Goal in a consistent and predictable manner Policy [NRC, 1986]. that would promote regulatory stability and efficiency [NRC, 1995].

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on NRC Safety Research Program to The ACRS has been very supportive Congress from 1977 until 1997. In 1998, of the evolution toward a risk-informed Public Law 105-362 struck those two and performance-based regulatory sentences in Section 29.

system and has taken a leading role in considering some of the challenging In 1997, the Commission transferred issues that have arisen in this effort. the research advisory function of the Nuclear Safety Research Review Since early 1990s, the ACRS has Committee (NSRRC) to the ACRS. In played an important role in design this role, the ACRS was directed to certification processes and has identified examine the need, scope, and balance many technical safety issues during its of the reactor safety research program reviews of advanced LWR designs, [NRC, 1997]. The Committee was also which were resolved before the directed to consider how well the Office Committee provided its final of Research anticipates research needs recommendations for approvals. More and how it is positioned for the changing discussions on the ACRSs role in design environment [NRC, 1997]. Since 1998, certification review process is provided in ACRS has been submitting reports to the Section 3 of this report. Commission on review and evaluation of the NRC Safety Research Program, Throughout its history, an essential annually at first, and after 2004 activity of the ACRS has also been biennially.

reviewing the research sponsored by the agency. This includes evaluation of Before the establishment of the technical and programmatic aspects of Advisory Committee on Nuclear Waste the overall reactor safety research (ACNW) in 1988, ACRS reviewed program as well as episodic review of matters related to the long-term particularly important ongoing research. management of radioactive wastes produced within the nuclear industry. In In 1977, Section 29 of the Atomic 2007, ACNW was renamed to Advisory Energy Act was amended to add the Committee on Nuclear Waste and following two sentences: In addition to Materials (ACNW&M), and in 2008 its other duties under this section, the ACNW&M was merged into the ACRS.

Committee, making use of all available Since then, the Committee has reviewed sources, shall undertake a study of many aspects of nuclear waste reactor safety research and prepare and management such as handling, submit annually to the Congress a report processing, transportation, and storage containing the results of such study. The of nuclear wastes including spent fuel first such report shall be submitted to the and nuclear wastes mixed with other Congress no later than December 31, hazardous substances.

1977." ACRS submitted an annual report 7

3. ACRS REACTOR LICENSING REVIEWS, A HISTORICAL PERSPECTIVE 3.1 Early Years of Reactor Licensing However, at the time, the operating Reviews experience with power reactors and the state of knowledge of safety analysis had The passage of the 1954 Atomic not progressed to the point where it was Energy Act made it possible for private possible to use quantitative techniques to companies to build and operate nuclear estimate the probabilities and reactors under license. This Act also consequences of accidents. Instead, assigned to the AEC the responsibility of conservative assumptions were used to protecting the health and safety of the provide upper bounds of the potential public through a licensing process. public consequences resulting from certain hypothetical accidents (the so-In the early years of development of called deterministic approach). The nuclear power plants, both the fundamental concept of defense-in-depth technology and its governing regulations was invoked at the time to ensure that the were in the formative stages. The early unquantified probabilities of accidents licensing reviews were highly customized were small [Nourbakhsh, et.al., 2018].

and were generally based on the engineering experience and judgment of The Atomic Energy Act of 1954 regulatory staff working closely with the prescribed a two-step licensing process.

ACRS, without the availability of the The AEC would issue a construction regulatory guidance and structure permit based on the safety of the established later during commercial preliminary plant design and the development of LWRs. The AEC safety suitability of the prospective site. Only philosophy, as summarized in a when AEC determined that the plant fully March 14, 1956, AEC letter to the met safety requirements, the applicant Congress of the United States would receive a license to load fuel and (responding to a February 16, 1956, begin operation. This two-step licensing letter to the ACRS) [AEC, 1956] was process allowed enough time to based on the proposition that the ultimate investigate outstanding safety questions safety of the public depends on three and to prescribe modifications to initial factors: plans. It was recognized that the wisdom of permitting construction to proceed

1. Recognizing all possible without first resolving all potential safety accidents that could release unsafe issues was disputable, but there were no amounts of radioactive materials alternatives in light of the existing state of the technology and the commitment to

2. Designing and operating the the rapid development of atomic power reactor in such a way that the [Walker & Wellock, 2010].

probability of such accidents is reduced to an acceptable minimum Table 1 depicts the power reactors approved for construction up through

3. By the appropriate combination of 1960. These early power reactors containment and isolation, protecting included small prototypes developed by the public from the consequences of the AEC in cooperation with electric such an accident, should it occur. utilities and reactor manufacturers.

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Table 1 Early U.S. Nuclear Power Reactors Name Type Power Construction Operation (MWt) Approved Shippingport PWR 231 1954 1957-1982 Indian Point 1 PWR 585 1955 1962-1974 Dresden 1 BWR 630 1955 1969-1978 Fermi 1 Fast Reactor 300 1955 1963-1972 Yankee PWR 485 1957 1960-1992 Elk River BWR 58 1958 1964-1968 Piqua Organic 48 1959 1963-1966 Carolina-VA PHWR 63 1959 1963-1967 Hallam Sodium Graphite 240 1959 1963-1964 Saxton PWR 20 1959 1967-1972 Pathfinder BWR 203 1959 1966-1967 Big Rock BWR 240 1960 1962-1997 Humboldt Bay BWR 202 1960 1963-1976 Bonus BWR 50 1960 1967-1974 Peach Bottom 1 HTGR 115 1960 1967-1974 The ACRS concerns during the early The licensing review of the proposed years of reactor licensing included construction of the 48 MWt Piqua reactor siting, need for a dependable organically cooled and moderated containment system as a means of reactor is particularly noteworthy. The protecting the public against the reactor designer, Atomic International, consequences of reactor accidents, and did not initially propose a containment lack of sufficient information regarding based on the argument that no accident certain features to arrive at a conclusion had been found which could release concerning the construction of the plants. significant quantities of radioactive materials. The AEC concluded that containment was needed in a moderately populated region. Atomic International then proposed a new site and an unconventional form of containment.

The ACRS wrote a report stating that the Committee does not consider the installation at this site of a nuclear power plant of this capacity of a relatively untried type to be without undue public hazard until the proposed unconventional type of containment is replaced by a more substantial and The Piqua Nuclear Power Facility, an organic dependable system [Okrent, 1981].

cooled and moderated nuclear reactor started operation in 1963 as a demonstration project by Finally, Atomic International proposed a the AEC. In 1966, problems with control rods and more conventional containment for the fouling on cooling surfaces led to ceased Piqua reactor, and the Committee, in its operations. The neutron flux within the reactor May 18, 1959, report [ACRS,1959], was core induced polymerization of Terphenyl, leading favorable to the new site.

to increased viscosity of the coolant and fouling.

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3.2 Licensing Reviews of Early High- the engineered safeguards (containment Power LWRs plus sprays and/or filters) needed to meet the dose guidelines of Part 100. The In the early 1960s, the AEC began ACRS licensing reviews also led to defining a standard regulatory evolution of many new safety prescription to licensing of nuclear requirements dealing with a wide range reactors. Reactor siting was the first of technical issues. The following are issue addressed with the new approach. some examples of the issues raised by Regulations for site selection were the ACRS during its licensing reviews of developed as 10 CFR Part 100, Reactor early high-power reactors.

Site Criteria. Part 100 was developed, in part, based on the assumptions that an

• Control Rod Ejection Accidents:

upper limit of fission product release The ACRS report on licensing review could be estimated and the containment of the Connecticut Yankee plant building, as a final element of defense [ACRS, 1964a] was the first to call out against the release of radiation, would the requirements for study of the hold even if a severe accident were to control rod ejection accident. This occur. In conjunction with Part 100, the led to design changes in large LWRs, concept of a maximum credible accident either to limit the reactivity worth of was developed to evaluate the control rods or to add an additional acceptability of a potential site (siting mechanical restraint to control rod limits) and containment design ejection (an approach taken in requirements. BWRs) [Okrent, 1981].

After publication of the proposed

• Design Considerations for a Part 100 Reactor Site Criteria in 1961, Tsunami Following a Major several high-power reactors were Earthquake: The ACRS report on proposed for construction. These large the proposed 1473 MWt Malibu LWRs were proposed at sites that did not Nuclear Plant Unit 1 for construction meet Part 100 without taking credit in at Corral Canyon (twenty-nine miles calculating off-site doses, either for a west of Los Angles) was the first to reduction in leak rate (e.g., by use of raise the issue of the adequate double containment), or by reducing the protection against a tsunami postulated fission product source following a major seismic event. The available to leak out of containment (e.g., following paragraph from the July 15, using containment atmosphere cleanup 1964, ACRS report on the proposed systems, such as containment sprays or Malibu Plant is particularly closed-loop filter systems) [Okrent, noteworthy:

1979]. The early large LWRs approved for construction using the Part 100 The ability of the plant to Reactor Site Criteria included San withstand the effects of a tsunami Onofre (1347 MWt PWR), Connecticut following a major earthquake has Yankee (1473 MWt PWR), and Oyster been discussed with the applicant.

Creek (1600 MWt BWR). It should be There has not been agreement noted that not all reactor proposals among consultants about the height received approval and not all approved of water to be expected should a reactors were ultimately constructed. tsunami occur in this area. The Committee is not prepared to resolve The principal focus of the ACRS the conflicting opinions and suggests licensing reviews of early large LWRs that intensive efforts be made to appears to have been on the efficacy of establish rational and consistent 10

parameters for this phenomenon. core spray or core-flooding system.

The applicant has stated that the Later that year Westinghouse containment structure will not be introduced accumulators.

impaired by inundation to a height of fifty feet above mean sea level. The

• Emergency Planning: As the size of integrity of emergency in-house proposed nuclear power plants power supplies should also be increased and containment could no assured by location at a suitable longer be regarded as an height and by using water-proof unchallengeable barrier to the techniques for the vital power escape of radioactivity, ACRS paid system. The emergency power more attention to emergency system should be sized to allow planning. In 1966, the Committee simultaneous operation of the noted that many applicants and containment building spray system licensees would rely heavily on local and the recirculation and cooling authorities to carry out evacuation if it system. Ability to remove shutdown should become necessary. There core heat under conditions of total were also no guidelines for judging loss of normal electrical supply when an evacuation would be should be assured. If these advisable. The ACRS decided that it provisions are made, the Committee should alert the AEC to a problem believes that the plant will be area where little efforts being adequately protected [ACRS, exerted [Walker, 1992]. Pressed by 1964b]. the ACRS, the AEC undertook a study of emergency plans and The proposed Malibu reactor was procedures that eventually led to never built. An intervener group adding a new Appendix E to 10 CFR successfully contested the Part 50, Emergency Planning and construction of the proposed Malibu Preparedness for Production and plant. The adequacy of seismic Utilization Facilities.

design was one of the main points of contention [Okrent, 1981].

• Reactor Pressure Vessel Integrity:

In 1965, Dresden 2 (a 2250 MWt

• Effectiveness of ECCS Design: By BWR) was proposed for construction the mid-1960s, as proposed plants at the site of Dresden Unit 1.

increased significantly in power level, Dresden 2 represented a large jump the ACRS became concerned that a in power level as compared to Oyster core meltdown accident, particularly Creek (1600 MWt), the largest one in which the plants emergency reactor previously approved for core cooling system (ECCS) might construction. Many ACRS members fail to operate as designed, could considered Dresden 2 as a likely lead to a breach of containment. The prototype for other reactors which ACRS emphasized the need for would be proposed for metropolitan improved ECCS. By August 1966, areas. For this and other reasons, General Electric responded in Dresden 2 received much attention support of the Dresden 3 plant by by the ACRS, and the potential proposing a redundant core-flooding resolution of certain generic matters system and an automatic such as reactor pressure vessel depressurization system, which integrity became tied to its licensing would reduce the primary system review [Okrent, 1981].

pressure sufficiently to maximize the effectiveness of the low-pressure 11

The concern of ACRS for reactor

• Anticipated Transients Without pressure vessel integrity went back to Scram (ATWS): The issue of ATWS 1961 when the Committee, in its was first raised by E. P. Epler, an May 20, 1961, report, raised the ACRS consultant, in a January 21, issue of the potential damage to 1969, letter to the ACRS executive reactor pressure vessel by virtue of secretary. Soon after, the ACRS the neutron flux to which they are decided to identify the issue in its subjected during their life [ACRS, letter reports on Hatch unit 1 [ACRS, 1961]. However, prior to Dresden 2 1969a] and on the application for the review, the vessel failure was not construction authorization for the considered credible. Brunswick Units 1 and 2 {ACRS, 1969b]. In each report the Committee recommended a study be made by the applicant of further means of preventing common failure modes from negating scram action and of design features to make tolerable the consequences of failure to scram during anticipated transients.

3.3 Reviews of Latter Non-LWR Designs Dresden Nuclear Power Station, Unit 2 (a 2250 MWt BWR 4 with Mark 1 Containment}, proposed The ACRS has a long history of for construction in 1965, represented a large jump review and evaluation of non-LWRs.

in power level as compared to the largest reactor Early safety reviews of non-water-cooled previously approved for construction. For this and power reactors (refer to Section 3.1), like other reasons, Dresden 2 received much attention those performed for the early LWRs, by the ACRS, and the potential resolution of certain generic matters tied to its licensing review. were highly customized and were generally based on the engineering In 1965, the concern about experience and judgment of the pressure vessels had been regulatory staff working closely with the growing, in part, due to the 1964 ACRS, without the availability of the failure of a very large heat exchanger regulatory guidance and structure at a temperature near the nil ductility established later during LWR commercial temperature while being tested by the development. In latter reviews, explicit Foster Wheeler Corporation [Okrent, use was made of LWR regulatory 1981]. During the Dresden 2 guidance where applicable.

licensing review, the ACRS discussed this matter extensively and Prior to the 1986 policy statement on debated whether this issue should be regulation of advanced reactors [NRC, handled in a generic way or by 1986], the principal statement on non-specifically addressing this applicant. LWR review policy was given in the The ACRS finally decided to issue a introduction to Appendix A to 10 CFR 50:

report favorable to the construction of General Design Criteria (GDC) for Dresden 2 and at the same time Nuclear Power Plants. Specifically, this wrote a general report regarding introduction states: These General reactor pressure vessels. Design Criteria establish minimum requirements for the principal design criteria for water-cooled nuclear power 12

plants similar in design and location to plants for which construction permits For those portions of non-LWR have been issued by the Commission. designs that were uniquely different from The General Design Criteria are also those of LWR designs (e.g.,

considered to be generally applicable requirements for handling a sodium to other types of nuclear power units coolant or the use of a concrete reactor and are intended to provide guidance vessel for HTGRs), adoption or in establishing the principal design adaptation of existing regulations or criteria for such other units. standards was not possible or desirable.

In such cases, design-specific licensing The regulatory staff worked closely criteria were developed by engineering with the ACRS in developing the GDC. judgment and analysis.

Versions of general design criteria were publicly available as early as 1965 before initially being incorporated into Part 50 in February 1971.

Development of the GDC led to the comparable level of a safety philosophy under which proposed high temperature gas-cooled reactors (HTGRs) and liquid metal reactors (LMRs) were reviewed in later years [NRC,1988]. This philosophy was based on the notion that a Fort St. Vrain Generating Station: A 842 MWt comparable level of safety would be HTGR cooled with helium. The prestressed established for all reactor types, concrete reactor vessel (PCRV) of this plant was recognizing that the licensing criteria for the first in the United States. The ACRS issued a favorable, but cautious report on its construction non-LWRs could be developed, to the permit. It operated sporadically for ten years extent feasible, using those for LWRs. (1979-1989). .

This philosophy was implemented, with respect to the existing criteria, by direct Table 2 depicts ACRS review adoption, suitable adaptation, or activities associated with latter non-LWR development of design-specific criteria (if designs. Short summaries of those needed). Direct adoption of the existing reviews are provided in the following:

criteria in many instances could provide a ready means of ensuring a comparable Fort St. Vrain: The ACRS licensing level of safety [NRC, 1988]. review of the proposed construction of Fort St. Vrain is particularly noteworthy.

For those existing criteria that could Fort St. Vrain was a HTGR cooled with not be viewed as clearly applicable, helium and designed to produce 842 MWt suitable adaptations were developed to (330 MWe). The prestressed concrete permit the use of the phrase meets the reactor vessel (PCRV) proposed for this objectives of or words to this effect. plant was the first in the United States.

Development of such adaptations was This PCRV was to contain not only the usually a straightforward practice of the core, but the entire primary coolant applicant identifying and justifying system. The plant utilized a confinement discrepancies from the criteria building equipped with ventilation filters

[NRC,1988]. An early example of the for removing particulates and iodine from adaptive approach was the means for the building exhaust.

conformance of the Fort St. Vrain design to the GDC for LWRs.

13

Table 2 The ACRS Review of Latter Non-LWR Designs Design ACRS Reviews Remarks ACRS Report on CP ACRS issued a favorable, but cautious May 15, 1968 report on CP.

Fort St. Vrain -

Final ACRS Report, HTGR, 842 MWt May 12, 1971 Sporadic operation (1979-1989), mainly caused by water ingress from helium circulator bearings.

1100 MWe HTGR ACRS reviewed the A conventional low-leakage containment (A 1969 study to conceptual design building similar to those used for PWRs upgrade HTGR power ACRS Report: determined to be necessary for an HTGR level) November 12, 1969 of this size.

Concept was reviewed ACRS indicated certain safety by ACRS, 1971-1974 disadvantages unique to the GCFBR, as Gas Cooled Fast ACRS Report, well as some safety problems common to Breeder Reactor, November 8, 1974 all fast reactors. General Atomic pursued GCFBR a program to resolve the outstanding safety/licensing issues.

Licensing activities Sited in Delaware and Pennsylvania, but Summit and Fulton 1973 to 1975 plants cancelled for economic reasons Plants ACRS Reports issued prior to public hearings and CP issuance.

HTGR, 700-1000 MWe. March 12, 1975 April 8, 1975 FFTF (Fast Flux Test Safety review by the Operated (1982-1992) as a national Facility) ACRS, but a license research facility to test various aspects of sodium-cooled fast was not required by law commercial reactor design and operation.

reactor, 400 MWt ACRS Final Report November 8, 1978 Clinch River Breeder ACRS reviewed the CP NRC completed its review and public Reactor, CRBR application hearing for CP in 1983. Plant never built liquid-sodium-cooled fast- Final Report: due to termination of the project by breeder reactor. April 19, 1983 Congress.

975 MWt On May 15, 1968, the ACRS issued plant shut down to repair a stuck control a favorable, but cautious report on its rod pair, numerous cracks were review of the Fort St. Vrain CP discovered in several steam generator application, with supplemental added main steam ring headers. The required comments by two members [ACRS, repairs were determined to be too 1968b]. In its report, the ACRS stated extensive to justify continued operation.

that since this was a first-of-a-kind reactor, particular attention must be paid 1100 MWe HTGR: In the late 1960s, to final design, construction, and quality the ACRS reviewed a conceptual design control. The ACRS report also identified and proposed design bases for an 1100 several items of safety-oriented research MWe HTGR. The reactor concept and development as critical to the safety utilized a helium-cooled, graphite-of this system. Fort St. Vrain operated moderated ceramic fuel core similar to sporadically for a little more than 10 that of the Fort St. Vrain reactor. In its years (1979-1989). In 1989, during a November 12, 1969, report, the ACRS 14

stated that although insufficient and to all large HTGRs. The ACRS information has been provided for the comments in its April 8,1975, report on Committee to make a recommendation Fulton [ACRS, 1975b] were essentially as to the adequacy of the specific design the same as those made on the Summit bases proposed, it believes that an 1100 Power Station.

MWe HTGR plant along the general lines of the preliminary concept described in GCFBR: In the mid-1970s, the ACRS the referenced documents may reviewed a conceptual design of a gas constitute a system which can be so cooled fast breeder reactor (GCFBR) by engineered and operated [ACRS, General Atomic Company. The purpose 1969c]. of this review was to acquaint the Committee with the conceptual design Summit and Fulton Plants: In the and to identify those areas which the 1970s, the ACRS also reviewed the Committee believed required further Summit and Fulton applications for large technological development, or which the HTGRs that were not subsequently built. Committee deemed unacceptable at the time. In its November 8,1974, report The proposed Summit Power Station [ACRS, 1974], the ACRS indicated consisted of two nuclear units, each certain safety disadvantages unique to using a General Atomic HTGR having a the GCFBR, as well as some safety rated power level of 1532 MWe. The problems common to all fast reactors.

Summit Power Station would have been Based on the licensing concerns of the located in Delaware, approximately 15 ACRS and NRC, General Atomic miles from Wilmington. In its March 12, pursued a program to resolve the 1975, report, the ACRS recognized that outstanding safety/licensing issues. One the Summit Power Station represented a example was studies on how other fast new design so that many of the proposed and thermal reactor systems could systems and components were relatively provide a combination of built-in thermal untested at the time. This aspect was capacity and/or natural circulation apparent in the NRC Staff Safety cooling, avoiding the need for rapid Evaluation Report (SER) where several restoration of forced circulation of items were unresolved, or resolution was primary coolant to prevent fuel to be deferred until the post-construction overheating. Studies were also permit period. The Committee urged the performed on other GCFR design resolution of those outstanding items well arrangements which could provide before equipment was installed [ACRS, sufficient natural circulation cooling if 1975a]. coolant pressure were maintained

[Lipinski et. al, 1978].

The proposed Fulton Generating Station consisted of two nuclear units, FFTF: In 1970s, the ACRS reviewed each using a General Atomic HTGR the proposed construction and operation having a rated power level of 1160 of the DOE Fast Flux Test Facility (FFTF)

HW(e). The Fulton Generating Station [ACRS,1978]. The FFTF was a 400 MWt would have been located approximately sodium-cooled fast reactor located at 17 miles south of Lancaster, DOE's Hanford Reservation. Since the Pennsylvania. The similarities between FFTF was a DOE facility, the scope of the the Fulton and Summit Stations were review was defined by the DOE request considered in the ACRS review of Fulton that NRC provide only advice regarding

[1975b]. With no significant exceptions, the adequacy of its design and technical the Committees concerns were generic specifications to ensure safe operation.

to both the Fulton and Summit Stations 15

its April 19, 1983, report, the ACRS recommended that timely completion of the PRA by the applicants, to permit its review and evaluation by the NRC Staff and the ACRS, be a condition of the construction permit. ACRS also raised a number of safety issues for which more work must be done prior to their resolution. The ACRS believed that, if the matters raised by the Committee and the open items described in the SER were resolved in a satisfactory manner, Fast Flux Test Facility (FFTF): A 400 MWt sodium-cooled fast reactor located at the DOEs the CRBRP could be constructed with Hanford Reservation. The ACRS provided advice reasonable assurance that it could be regarding the adequacy of its design and technical operated without undue risk to the health specifications. The FFTF operated (1982-1992) as and safety of the public [ACRS, 1983].

a national research facility to test various aspects of commercial reactor design and operation.

The NRC completed its review and public hearing for CP in 1983. The plant The FFTF did not have a Class 1E was never built due to termination of the power supply to provide decay heat project by Congress.

removal. Instead, the project depended upon natural convection cooling in the 3.4 Pre-application Reviews of Earlier event of loss of offsite power and failure Advanced Non-LWR Designs of the onsite diesel generators. The project's calculations indicated that In 1986, the Commission issued its natural circulation would provide decay policy statement on advanced reactors heat removal. It was proposed that the

[NRC, 1986]. Advanced reactors were to natural circulation decay heat removal be include evolutionary LWRs, non-LWRs, measured during the startup testing. In and small modular light water reactors.

its November 8, 1978, report, the ACRS In its policy statement, the Commission concurred that the adequacy of the decay expected that advanced reactors would heat removal by natural circulation provide enhanced margins of safety should be experimentally verified [ACRS, and/or utilize simplified, inherent, 1978]. The FFTF operated (1982-1992) passive, or other innovative means to as a national research facility to test accomplish their safety functions. The various aspects of commercial reactor Commissions policy statement design and operation.

regarding regulation of advanced nuclear power plants has encouraged early CRBRP: In the early 1980s, the interaction (prior to a license application)

ACRS reviewed the CP application for between vendors and the NRC to the Clinch River Breeder Reactor Plant provide for early identification of (CRBRP). The CRBRP was to be a regulatory requirements for advanced liquid-sodium-cooled, mixed-oxide-reactors, and to provide all interested fueled, fast-breeder reactor parties, including the public, with a timely, demonstration power plant. Design independent assessment of the safety power was 975 MWt (350 MWe).

characteristics of advanced reactor designs. The NRC has been particularly The applicants were conducting a interested in any regulatory issues which full-scope PRA on the CRBRP design could lead to the need for Commission during its licensing review by the NRC. In policy decisions, or technical issues 16

unique to the design that could require MHTGR: The modular high extensive effort and a long lead time to temperature gas cooled reactor resolve. (MHTGR) concept was a product of a joint DOE/industry program to develop a Consistent with the policy statement, design for a nuclear power plant using the ACRS has also been holding pre- HTGR technology and having important application meetings and discussions to inherently safe characteristics. In its familiarize the Committee with the design October 13, 1988, report [ACRS, 1988]

and to identify topics for more detailed on preapplication safety evaluation of discussions before the application is MHTGRs, the ACRS raised a number of submitted. Table 3 depicts the ACRS safety issues to be adequately pre-application reviews of non-LWR addressed to assure the key safety designs in the late 1980s and early characteristics claimed for the design 1990s. Short summaries of those pre- would be realized in an actual plant.

application reviews are provided in the following.

Table 3 The ACRS Pre-application Reviews of Non-LWR Designs Design ACRS Reviews Remarks MHTGR (Modular Preapplication The ACRS raised a number of safety HTGR) design review by issues that had to be adequately 350 MWt, annular ACRS addressed to assure the key safety prismatic block core ACRS report: characteristics claimed for the design arrangement October 13,1988 would be realized in an actual plant.

SAFR (Sodium Preapplication The conceptual design was under Advanced Fast design review, ACRS review when the DOE terminated the Reactor) Report: work in September 1988.

3600 MWt (each January 19, 1989 "power pack" comprising four reactor modules)

PRISM (Power Preapplication The staffs preliminary findings were Reactor Innovative Design Review, reviewed by the ACRS. After the Small Module) 1986-1994 review, several revisions to the Pool-type, LMR. 475 Final ACRS Report: conceptual design were made.

MWt November 10, 1993 Toshiba 4S (Super- Preapplication review The NRC ceased its review of the Safe, Small, and meetings in 2007- Toshiba 4S in 2013 without issuing Simple) 2008 any review documents.

30 MWt pool type LMR ACRS was briefed and held discussions regarding the design 17

SAFR: The Sodium Advanced Fast draft PSER for the PRISM liquid-metal Reactor (SAFR) conceptual design was reactor [ACRS,1993], the Committee another product of a DOE program to stated that although our own review of develop designs for possible future the PSER was less detailed than would power reactor systems that would have have been appropriate for a safety enhanced safety characteristics. Other evaluation report on an actual design projects in the program were the application, we believe that the staff has MHTGR and the PRISM. The SAFR satisfactorily fulfilled its role in the conceptual design was under review preapplication process.

when the DOE decided to discontinue its development and concentrate LMR efforts in the PRISM design organization.

In its January 19, 1989, report on safety evaluation of SAFR design, the ACRS stated that a continuing program of research and development, including plans for extensive prototype testing, would be necessary to support further design [ACRS, 1989]. The Committee also commented on a number of specific safety issues that DOE should consider if it continued design and development of the SAFR concept.

Power Reactor Innovative Small Module PRISM: The PRISM was the only DOE (PRISM): The NRC conducted a thorough review sponsored design that was developed to of the 475 MWt design between 1986 and 1994.

Consistent with the Commission's advanced the point that a safety review was reactor policy, the NRC staff, to the extent conducted. The NRC conducted a feasible, used existing regulations at the time to thorough review of the 475 MWt design formulate criteria and procedures for review of this between 1986 and 1994. Consistent with design. The ACRS reviewed the staffs preliminary findings. After the review, several revisions to the the Commission's advanced reactor conceptual design were made.

policy, the staff, to the extent feasible, used existing regulations at the time to Toshiba 4S: The Toshiba 4S (Super-formulate criteria and procedures for Safe, Small, and Simple) was a small review of this design. The ACRS sized (30 MWt) sodium-cooled fast reviewed the staffs preliminary findings. reactor designed for remote locations After the review, several revisions to the with small grids. The NRC held a series conceptual design were made. Because of preapplication review meetings in the staff review was based on a 2007-2008. From 2008 to 2013, Toshiba conceptual design, the preapplication continued to submit a series of technical safety evaluation report (PSER) did not, reports that responded to the policy nor was it intended to, result in an statement on the licensing of advanced approval of the design. Instead, it reactors. In 2010, the ACRS was briefed identified certain key safety issues, by and held discussions with provided some guidance on applicable representatives of the NRC staff licensing criteria, assessed the adequacy regarding the Toshiba 4S design.

of the research and development Toshiba did not proceed with an programs, and concluded that no obvious application for certification of the design.

impediments to licensing the PRISM NRC ceased its review of the Toshiba 4S design had been identified. In an ACRS in 2013 without issuing any review report, dated November 10, 1993, on the documents [NRC, 2019].

18

3.5 Licensing Reviews of Advanced 15 years from the date of issuance but Light Water Reactors can be renewed for an additional 10 to 15 years. According to NRC regulation (10 In the early 1980s, in cooperation CFR 52.53), the design certification with DOE, the U.S. nuclear utility application (DCA) is referred to the industry, with support from the Electric ACRS for a review and report.

Power Research Institute (EPRI),

initiated the Advanced Light Water Reactor Program to ensure a viable nuclear power generation option for the 1990s and beyond. A major objective of the program was to develop designs for future LWRs that were safer, more reliable, easier to operate, and more certain of being licensed without delays.

A means to assure this outcome was the development of the ALWR Utility Requirements Document by senior, experienced utility personnel in the U.S.

and overseas that incorporated the Advanced Passive 1000 (AP1000): A two-loop lessons learned from decades of pressurized water reactor (PWR) with passive worldwide operating experience with safety features. The ACRS played an important LWRs [DOE, 2001]. The ACRS followed role in the AP1000 design certification process by the development of the EPRI ALWR providing an independent review of the NRC program from its inception and offered staffs determination of compliance with the applicable standards and requirements of the suggestions regarding safety Atomic Energy Act and the Commissions improvements on several occasions regulations. The ACRS identified many technical

[ACRS, 1992]. issues during its review process which were resolved before the Committee provided its final recommendations on the design certification.

In 1989, the NRC established alternative licensing processes to The ACRS has played an important improve regulatory efficiency and add role in design certification processes by greater predictability to the licensing providing an independent review of the process. The alternative licensing determination of compliance with the processes include a combined license applicable standards and requirements that essentially combines a construction of the Atomic Energy Act and the permit and an operating license, with Commissions regulations.

certain conditions, into a single license.

Other licensing alternatives established The ACRS has identified many in 1989 are early site permits, which technical issues during its design allow an applicant to obtain approval for certification reviews, which were a reactor site for future use, and certified resolved before the Committee provided standard plant designs, which can be its final recommendations for approvals.

used as pre-approved designs. The Table 4 lists the design certification NRC may approve and certify a standard applications that the ACRS has reviewed nuclear plant design through a to date. Short summaries of those design rulemaking, independent of a specific site certification reviews are provided in the and an application to construct or operate following.

a plant. A design certification is valid for 19

Table 4 The Design Certification Applications Reviewed by the ACRS Design Applicant Final ACRS Report Advanced Boiling Water General Electric (GE) Nuclear April 14, 1994 Reactor (ABWR) Energy System 80+ ASEA Brown Boveri - Combustion May 11, 1994 Engineering (ABB-CE)

Advanced Passive 600 Westinghouse Electric Company July 23, 1998 (AP600)

Advanced Passive 1000 Westinghouse Electric Company July 20, 2004 (AP1000)

Economic Simplified Boiling- GE Hitachi Nuclear Energy October 20, 2010 Water Reactor (ESBWR) April 17, 2014 U.S. EPR AREVA NP, Inc. Suspended U.S. Advanced Pressurized- Mitsubishi Heavy Industries, Ltd Suspended Water Reactor (US-APWR)

Advanced Power Reactor Korea Electric Power Corporation July 26, 2018 1400 (APR1400) and Korea Hydro & Nuclear Power Co., Ltd.

NuScale Small Modular NuScale Power, LLC July 29, 2020 Reactor ABWR: The Advanced Boiling Water Reactor Water Cleanup (CUW) supply Reactor (ABWR) is a forced circulation line located inside of primary boiling water reactor with a rated power containment.

of 3926 MWt. The U.S. version of the ABWR standard design utilized a In its April 14,1994, report, the ACRS significant portion of the detailed design stated that acceptable bases and information developed jointly by General requirements had been established in the Electric Nuclear Energy (GENE), Hitachi, application to assure that the U.S.

and Toshiba for the international version version of the ABWR standard design which was being built in Japan. could be used to engineer and construct plants that with reasonable assurance The ACRS reviewed the ABWR could be operated without undue risk to design and issued its final report on the the health and safety of the public safety aspects of the GENE application {ACRS, 1994a].

for certification of the ABWR design in April 1994 [ACRS,1994a]. The ACRS System 80+: The ASEA Brown Boveri -

played an important role in GENE ABWR Combustion Engineering (ABB-CE) design certification process [Nourbakhsh System 80+ standard plant design

& Banerjee, 2015]. The ACRS review of evolved from the CE System 80 plant ABWR design led to significant design design. The ABB-CE System 80+ design changes, such as a GENE proposal that included several features that would safety-related equipment inside of the have enhanced safety relative to past ABWR secondary containment be PWR designs. Some of those features environmentally qualified for steam at 15 resulted from the use of PRA psig and about 248°F and adding a third methodology by ABB-CE during the break isolation valve in the 8-inch System 80+ design process.

20

Based on the results of ACRS review for AP600). The ACRS concluded that of those portions of the ABB-CE System most of the AP600 review findings were 80+ application which concerned safety, applicable to the AP1000 design. This the Committee concluded that the conclusion greatly enhanced the System 80+ standard plant design can efficiency of the reviews of the AP1000 be used to engineer and construct plants safety assessments [ACRS, 2004].

that, with reasonable assurance, can be operated without undue risk to the health During ACRS review process, the and safety of the public [ACRS, 1994b]. Committee identified many technical safety issues of concern and areas for AP600: The 1933 MWt (600 MWe) which it needed additional discussions Advanced Passive 600 (AP600) design [Nourbakhsh, et al., 2013]. The ACRS represented a significant departure from agreed with the staffs proposed previous commercial nuclear reactor resolution of all but two of those issues.

technology in that it placed more The Committee developed its own dependence on passive systems for arguments for the resolution of the accident response. Unique features of remaining two issues [ACRS, 2004].

the AP600 design included an improved reactor core design, a large reactor ESBWR: The Economic Simplified vessel, a large pressurizer, an in- Boiling-Water Reactor (ESBWR) is a containment refueling water storage tank direct-cycle, natural circulation BWR and (IRWST), an automatic depressurization has passive safety features to cope with system, a digital microprocessor-based a range of design basis accidents I&C system, hermetically sealed canned (DBAs).

rotor coolant pumps mounted to the steam generator, and increased battery During its review, the ACRS identified capacity. many issues of concern and areas for which the Committee needed additional Westinghouse conducted an discussion. Those issues included extensive test and analysis program, combustion control of flammable non-utilizing separate-effects and integral- condensable gases in the passive system facilities both to investigate the containment cooling system (PCCS),

behavior of the AP600 passive safety which led the General Electric-Hitachi systems and to develop a database for Nuclear Energy (GEH) to revise the validation of the computer codes used to design of the isolation condensers (IC) perform accident and transient analyses. and PCCS to address the potential for During the extensive reviews of the hydrogen detonations within the Westinghouse test and analysis condenser tubes or the lower plenum program, the ACRS raised numerous [ACRS, 2010]. GEH also provided more safety issues, which were adequately detailed explanations and tabular resolved before the Committee information in the ESBWR Design concluded that the AP600 design can be Control Document revisions to give the used to engineer and construct plants Committee confidence that the four that with reasonable assurance can be fundamental principles are inherent in the operated without undue risk to the health hardware and software DI&C and safety of the public [ACRS, 1998]. architectures, i.e., redundancy, independence, determinate behavior, AP1000: The AP1000 design is similar in and diversity and defense in depth concept to the AP600 design but [ACRS, 2010].

provides much higher power levels (1000 MWe for AP1000 compared to 600 MWe 21

In its report, dated October 20, 2010, strength, steel containment vessel that the ACRS concluded that the ESBWR closely surrounds the reactor vessel.

design is robust and there is reasonable Each NPM is rated at 160 MWt, with an assurance that it can be built and output of approximately 50 MWe.

operated without undue risk to the health and safety of the public [ACRS, 2010].

APR1400: The APR1400 is a pressurized-water reactor, which evolved from the System 80+ design. APR1400 includes several features that are designed to further improve safety and operability over that of the System 80+.

Enhancements include the combination of four trains of safety injection with direct vessel injection, and a unique safety injection tank fluidic device, which optimizes the safety injection flow rate during the initial blowdown and subsequent, long-term, core reflood phase. The performance of the fluidic device was verified via full scale testing.

The plant is designed to be fiber free, which along with experimental verification using conservative conditions, benefits long-term core cooling following a LOCA.

NuScale Small Modular Reactor: An integral During its reviews, the ACRS pressurized-water reactor designed by NuScale identified many issues of concern and Power LLC. The ACRS identified many technical areas needing additional discussion that safety issues during the NuScale design were addressed in nine ACRS letters. In certification review. The ACRS review led to design and setpoint changes to the NuScale its report, dated July 26, 2018, the ACRS Power Module concluded that the APR1400 design is mature and robust. There is reasonable The ACRS identified many technical assurance that it can be constructed and safety issues during the NuScale design operated without undue risk to the health certification review. As a result of the and safety of the public [ACRS, 2018]. ACRS review, NuScale incorporated design and setpoint changes to the NuScale: The certified NuScale design NuScale Power Module to mitigate the consists of up to 12 NuScale Power effects of boron dilution in the Modules (NPMs) in a single reactor downcomer for uncontrolled passive building. The NPMs are largely cooling events [ACRS, 2020a].

immersed in a large pool of borated water, also serving as the ultimate heat In its report, dated July 29, 2020, the sink. Each NPM is a small, integrated, ACRS concluded that there is reasonable natural-circulation PWR composed of a assurance that the NuScale small shrouded reactor core and riser, a modular reactor can be constructed and pressurizer, and two helical-tube steam operated without undue risk to the health generators within a reactor pressure and safety of the public [ACRS, 2020b].

vessel, and housed integral to a high-22

4. LOOKING AHEAD TO ACRS LICENSING REVIEWS OF NEW ADVANCED REACTOR DESIGNS 4.1 Essential Role of ACRS in Reactor As the NRC staff strategizes to Licensing Reviews assure that the Agency is ready to review potential licensing applications for new Throughout ACRSs history, the advanced reactor designs, the role of Committees independent review has ACRS, with its diverse technical been an essential element of the reactor expertise, will continue to be essential for licensing process. The Committees an independent integrated/multi-licensing reviews led to evolution of many disciplinary review. The Committees new safety requirements and design role is particularly important because the changes dealing with a wide range of new advanced reactor designs, currently technical issues. under development, pose new challenges for safety-licensing reviews.

A March 14, 1956, AEC letter to the These challenges include:

Congress [AEC, 1956], in response to a letter to ACRS on the question of public

• First-of-a-kind designs with a variety safety of nuclear reactors, states the of coolants, fuel forms, and following: innovative configurations The financial incentive of the owners

• Designs that do not have the same of the reactor to take all steps levels of operating and regulatory necessary to protect their investment, experience as that of LWRs as well as to decrease their potential public liability, and the legal and

• Limited experimental database and moral responsibilities of the validation Commission to protect the public from overexposure to radioactivity,

• Implementation of a new licensing are resulting in a system which is approach characterized by an attitude of caution and thoroughness of 4.2 Enhancing the Efficiency of the evaluation unique in industrial Review Process history. Every phase of the reactor design and operating procedure is The ACRS continues to perform reviewed separately and as a part of introspective evaluations to identify ways the whole. The inherent nuclear, to improve its own effectiveness and chemical, metallurgical, physical, and efficiency as the NRC accomplishes its mechanical characteristics of the mission within a changing regulatory fuel, moderator, coolant, neutron framework and culture, and in an absorbers, and structural materials expanding industry environment. Over are carefully considered to the past several years, the ACRS has assure that the probability of an also collaborated with NRC program operating mishap has by adequate offices to achieve greater efficiencies design and operating precautions while maintaining its independence been brought to an acceptably low [ACRS, 2019].

level.

23

As part of its continuing effort to detailed component and system become more effective and assist the analysis, it may be difficult for NRC to NRC in its transformation initiatives, the make a technically sound finding on any ACRS recently conducted a self- requested deviation (exemption) from assessment. This self-assessment was historical regulatory requirements (e.g.,

based on the Committees observations General Design Criteria).

and lessons-learned from its recent NuScale design certification and Design changes during the review standard design approval application process, as observed in the past, may reviews, informed also by its prior design also adversely impact the efficiency of certification and early site permit reviews, the review process.

as well as interactions with the NRC staff.

Observations and lessons-learned from Comprehensiveness of Knowledge the ACRS self-assessment led to several Base recommendations by the Committee that could improve future NRC reviews of All safety decisions are based, either advanced reactor designs [ACRS, explicitly or implicitly, on identifying 2020c]. For a more effective and radiological hazards and addressing the expeditious review, the ACRS has risk triplet questions: "What can go adopted a cross-cutting approach, wrong?; "How likely is it?"; and "What focusing on key safety-significant design are the consequences?" The NRC issues. It is expected that this will addresses these three questions through streamline reviews, resulting in more the body of its regulations and guidance.

efficiency and shorter schedules. The comprehensiveness of the knowledge base (experimental data Ultimately it is the completeness and base, operational experience, relevant quality of a license application and analyses, etc.), to support the safety associated supporting documents that decisions, has significant impacts on the significantly impacts the efficiency of the review process efficiency.

review process (both for the NRC staff and the ACRS). The desired attributes Both traditional deterministic and that would improve the quality and probabilistic approaches to safety completeness of future applications of analyses are based on identification of advanced reactor designs include the hazards, initiating events that disturb following: normal operation, scenarios (event sequences) that could evolve from the Completeness of the Design initiating events, and their associated consequences. Theoretical and Design completeness has a profound experimental bases are needed for impact on the efficiency of the review understanding the associated process. Proposed new reactor designs phenomenology of possible scenarios.

should be sufficiently complete to demonstrate that all Structures, For the new non-LWRs, their design Systems, and Components (SSCs) maturity and knowledge base are not important-to-safety are appropriately likely to be as comprehensive as for identified, designed, and tested, to be evolutionary LWR-based designs. The commensurate with their functions and to limited knowledge base may impact the provide adequate defense in depth. regulatory review.

Without having an essentially When there is a lack of operating complete design and performing experience or an inability to perform 24

experiments with sufficient similitude to are addressed in PRA models and not all the planned full-scale design, one phenomena and processes are approach for dealing with limited addressed in deterministic safety knowledge base, as suggested by the evaluation models. The safety ACRS, is limitations on power ascension philosophy of defense in depth and and focused surveillance tests during safety margins has been the traditional initial operation [ACRS, 2020d]. means of dealing with uncertainties.

Proper Consideration of Uncertainties The novel aspects of new technologies and first-of-a-kind reactor Safety-licensing decisions are made concepts make the identification of in the face of uncertainties and within the hazards, initiating events, and scenarios boundaries of the state of knowledge of more challenging. To address how the proposed reactor design would uncertainties caused by limited behave under both normal and accident information, the ACRS has conditions. Both deterministic and recommended critical examination of the probabilistic safety evaluations must deal design, its safety behavior, and all with uncertainties. Proper consideration aspects of operations, starting from a of uncertainties significantly helps the blank sheet of paper to avoid bias. The review process. Addressing Committee has also suggested use of uncertainties affects the reviewer several analysis tools that have been confidence regarding the results of safety developed to improve the search evaluations and the resulting safety process. These tools apply equally to margins. traditional and probabilistic safety analyses [ACRS, 2020c]. Such analysis Two major groups of uncertainty that tools can help formalize and add have been recognized are aleatory (or structure to the safety assessment and stochastic) and epistemic (or state-of- reduce completeness uncertainty.

knowledge) uncertainty. The key distinction between these two types of Appropriate Timing of Supporting uncertainty is that aleatory uncertainty is Documents Submittals irreducible. Epistemic uncertainty, in contrast, can be reduced by further The proper timing of the supporting study. documents (e.g., licensing topical reports) submittals may also have a There are two classes of epistemic significant impact on the efficiency of the uncertainty: parameter uncertainty and review process.

model uncertainty. Parameter uncertainties are those associated with Submittal of critical licensing topical the values of the fundamental reports late in the review process or parameters of a model, such as parallel with related chapters of the DCA, equipment failure rates that are used in as observed in the past, could reduce quantifying the accident sequence efficiency. The proper timing would be frequencies in PRAs. Model the sequential hierarchical order of uncertainties reflect the limited ability to submittals wherein licensing topical model accurately the specific events and reports on methodology description, phenomena. Completeness, including demonstration, and verification and possible unknown unknowns, can also validation precede their applications.

be considered as one aspect of model uncertainty. Completeness uncertainty The proper timing of critical topical arises because not all contributors to risk reports submittals is vital for review of 25

non-LWR concepts, which are likely to especially non-LWRs, will generally be have more uncertainty associated with more dependent on analytical methods analytical methods and their application, for understanding the safety response of underlying experimental bases, and the system [ACRS, 2020c].

validation of models. The licensing topical reports that support the design basis and safety analyses should be reviewed as early in the process as possible because new reactor designs, 26

5.

SUMMARY

AND CONCLUSIONS For almost 65 years the ACRS has advanced reactor designs include design had a continuing statutory responsibility completeness, knowledge base for providing independent reviews of, and (experimental data base, operational advising on, the safety of proposed or experience, relevant analyses, etc.)

existing reactor facilities and the comprehensiveness, proper adequacy of proposed reactor safety consideration of uncertainties, and standards in the United States. appropriate timing of supporting document submittals.

Throughout ACRSs history, the Committees independent review has been an essential element of the reactor licensing process. The Committees licensing reviews led to evolution of many new safety requirements and design changes dealing with a wide range of technical issues.

As the NRC staff strategizes to .

assure that the Agency is ready to review potential licensing applications for new advanced reactor designs that are radically different from the current fleet of LWRs, the role of ACRS, with its diverse technical expertise, will continue to be essential for an independent integrated/multi-disciplinary review.

The ACRS continues to perform introspective evaluations to identify ways to improve its effectiveness and efficiency as the NRC accomplishes its mission within a changing regulatory framework and culture, and in an expanding industry environment. Over the past several years, the ACRS has also collaborated with NRC program offices to achieve greater efficiencies while maintaining its independence.

Ultimately it is the completeness and quality of a license application and associated supporting documents that significantly impacts the efficiency of the review process (both for the NRC staff and the ACRS). The desired attributes that would improve the quality and completeness of future applications of 27

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32