ML18214A292

Acrs' Enduring Legacy Contributing to Reactor Safety - Proceedings of the 2018 26th International Conference on Nuclear Engineering

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1 Copyright © 2018 by ASME Proceedings of the 2018 26th International Conference on Nuclear Engineering ICONE26 July 22-26, 2018, London, England ICONE26-82275 ACRS' ENDURING LEGACY CONTRIBUTING TO REACTOR SAFETY Hossein Nourbakhsh Senior Technical Advisor for Reactor Safety Office of Advisory Committee on Reactor Safeguards (ACRS) U.S. Nuclear Regulatory Commission Washington, DC 20555

-0001 Hossein.Nourbakhsh@nrc.gov ABSTRACT 1 For over 60 years 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 paper discusses the role of the Committee as it has evolved during its more than 60 years of history, noting some of its significant contributions to reactor safety.

1. INTRODUCTION The history of ACRS goes back to 1947 when the U.S. Atomic Energy Commission (AEC) recognized the need for an independent technical group to review and provide advice on reactor safety matters and thus a Reactor Safeguard s Committee, chaired by Dr. Edward Teller was established. Dr. Teller has been quoted to recall that Reactor Safeguard s Committee "was about as popular

- and as necessary

- as a traffic cop" [1]. As stated by former NR C Chairman, Richard Meserve, the Reactor Safeguard s Committee "clearly established an enduring characteristic of the ACRS - a willingness to provide candid views on reactor safety issues, even at the risk of taking unpopular positions" [2].

1 The views expressed in this paper are solely those of the author and do not necessarily represent those of either the ACRS or NRC

. In 1950, the AEC established a second advisory committee, Industrial Committee on Reactor Location Problems, charged with the responsibility of advising on what we would today consider siting issues, including seismic and hydrological characteristics of proposed sites. In 1953, the Reactor Safeguard s Committee and the Industrial Committee on Reactor Location Problems were combined by the AEC and the ACRS was formally born.

The 1957 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 duties as the Commission may request." With the enactment of the Energy Reorganization Act of 1974, the AEC was reorganized into two separate entities, the Nuclear Regulatory Commission (NRC) and the Energy Research and Development Administration (now the Department of Energy). The ACRS was assigned to the newly established NRC with its statutory requirements intact.

This paper discusses the role of ACRS as it has evolved during its more than 60 years of history, noting some of its significant contributions to reactor safety.

2 Copyright © 2018 by ASME 2. ROLE OF ACRS OVER ITS HISTORY The role of ACRS has evolved over its 60 years of history.

Figure 1 depicts the number of reports/letters issue d by the Committee on various topics over its history.

The passage of the 1954 Atomic Energy A ct made it possible for private companies to build and operate nuclear reactors under license. This Act also assigned to the AEC the responsibility of protecting the health and safety of the public through licensing process. Most of today's U.S. nuclear power plants were licensed during the 1960s and 1970s, when both the technology and its governing regulations were in the formative stages. The ACRS review has always been an important element of the reactor licensing process.

After 1975, not only the orders for new reactor s plummeted, but there were also many cancellations of existing orders. Following the March 28, 1979 accident at Three Mile Island Unit 2 (TMI

-2), power reactor licensing was suspended for a year. As the ACRS moved into the 1980s, the Committee shifted much of its attention from plant design and construction to improvements in both the operation and regulation of nuclear power plants. The ACRS has been very supportive of the evolution toward a risk-informed and performance

-based regulatory system and has taken a leading role in considering some of the challenging issues that have arisen in this effort. Throughout its history, an essential activity of the ACRS has also been reviewing the research sponsored by the agency. This include s evaluation of technical and programmatic aspects of the overall reactor safety research program as well as episodic review of particularly important ongoing research

. In 1988, the Commission established the Advisory Committee on Nuclear Waste (ACNW) to advise the Commission on high-level waste and low

-level waste issues. Previously, the ACRS and its Waste Management Subcommittee performed this function. In 2007, ACNW was renamed to Advisory Committee on Nuclear Waste and Materials (ACNW&M). In 2008 the ACNW&M merged into the ACRS. The decision to merge ACNW&M into ACRS was based on the changing workload and technical challenges facing the agency and the anticipated increased need for expertise in health physics, waste management, and earth sciences in the agency's licensing reviews.

Figure 1. The number of reports/letters issued by the ACRS on various topics (1957-2017) 3 Copyright © 2018 by ASME 3. LICENSING REVIEWS The ACRS has played an important role in licensing reviews of reactors

[3]. The ACRS early reviews also led to evolution of many new safety requirements dealing with a wide range of technical issues. Following are some examples of the issues raised by the ACRS during its early licensing reviews. Control Rod Ejection Accidents:

The ACRS report on licensing review of the Connecticut Yankee plant [4] was the first to call out the requirements for study of the control rod ejection accident. This led to design changes in large LWRs, either t o limit the reactivity worth of control rods or to add an additional mechanical restraint to control rod ejection (an approach taken in BWRs)

[5]. Design Considerations for a Tsunami Following a Major Earthquake

The ACRS report on the proposed 1473 MW t Malibu Nuclear Plant Unit 1 for construction at Corral Canyon (twenty

-nine miles west of Los Angles) was the first to raise the issue of the adequate protection against a tsunami following a major seismic event. The following paragraph from the July 15, 19 64 ACRS report on the proposed Malibu Plant is particularly noteworthy.

"The ability of the plant to withstand the effects of a tsunami following a major earthquake has been discussed with the applicant. There has not been agreement among consultants about the height of water to be expected should a tsunami occur in this area. The Committee is not prepared to resolve the conflicting opinions, and suggests that intensive efforts be made to establish rational and consistent parameters for this phenomenon. The applicant has stated that the containment structure will not be impaired by inundation to a height of fifty feet above mean sea level. The integrity of emergency in

-house power supplies should also be assured by location at a suitable height and by using water-proof techniques for the vital power system. The emergency power system should be sized to allow simultaneous operation of the containment building spray system and the recirculation and cooling system. Ability to remove shutdown core heat under conditions of total loss of normal electrical supply should be assured. If these provisions are made, the Committee believes that the plant will be adequately protected

" [6]. The proposed Malibu reactor was never built. An intervener group successfully contested the construction of the proposed Malibu plant. The adequacy of seismic design was one o f the main points of contention [5].

Effectiveness of ECCS Design

By the mid

-1960s, as proposed plants increased significantly in power level, the ACRS became concerned that a core meltdown accident, particularly one in which the plant's emergency core cooling system (ECCS) might fail to operate as designed, could lead to a breach of containment.

The ACRS emphasized the need for improved emergency core-cooling systems (ECCS). By August 1966, General Electric responded in support of the Dresden 3 plant by proposing a redundant core

-flooding system and an automatic depressurization system, which would reduce the primary system pressure sufficiently to maximize the effectiveness of the low

-pressure core spray or core-flooding system. Later that year Westinghouse introduced accumulators.

Anticipated Transients without Scram (ATWS): The issue of ATWS was first raised by E. P. Epler , an ACRS consultant , in a January 21, 1969 letter to the ACRS executive secretary [5]. Few months later, the ACRS decided to identify the issue in its letter reports on Hatch unit 1

[6] and on the application for the construction authorization for the Brunswick Units 1 and 2

[7]. 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.

" In the early 1980s, the U.S. nuclear utility industry in cooperation with DOE, and with support from the Electric Power Research Institute (EPRI), initiated the Advanced Light Water Reactor (ALWR) program to ensure a viable nuclear power generation option for the 1990s and beyond. The ACRS followed the development of the ALWR program from its inception and offered suggestions regarding safety improvements on several occasions

[9]. In 1989, the NRC established alternative licensing processes to improve regulatory efficiency and add greater predictability to the licensing process.

The ACRS has played an important role in new reactor licensing reviews including the design 4 Copyright © 2018 by ASME certification process. According to NRC regulation (10 CFR 52.53), the design certification application is referred to the ACRS for a review and report.

The ACRS has identified many technical issues during its design certification reviews which were resolved before the Committee provided its final recommendations for approvals

[10-11]. The License Renewal Rule (10 CFR Part 54), first issued in 1991, establishes the technical and procedural requirements for renewing power reactor operating licenses.

According to 10 CFR 54.25 each license renewal application shall be referred to the ACRS for a review and report. To date, the ACRS has completed the review of 56 license renewal applications and the associated NRC staff safety evaluation reports involving 90 nuclear power units. The ACRS has contributed significantly to the success of the license renewal program by establishing expectations on the quality of the submittals and of the license renewal programs committed to by licensees.

Utilities have been using power uprates since the 1970s as a way to increase the power output of their nuclear plants. Power uprates are submitted to NRC as license amendment requests. ACRS reviews the power uprates that are amounting to power increase greater than 5 percent above originally licensed value. Since 1998, The Committee has reviewed 26 applications for power uprates.

ACRS has contributed significantly to the success of the power uprate program by establishing expectations on the quality of the power uprate license amendment requests and supporting documentations. The Committee was instrumental in the staff development of a review standard for extended power uprates [12]. 4. REGULATORY POLICIES AND PRACTICES The ACRS has played a significant role in the review and resolution of key technical issues associated with regulation of nuclear power plants.

Th e Committee has a history of recommending innovative approaches to regulatory problems. Following are some examples of the roles ACRS has played in shaping the regulatory policies and practices of the agency:

Emergency Planning

As the size of proposed nuclear power plants increased and containment could no longer be regarded as an unchallengeable barrier to the escape of radioactivity, ACRS paid more attention to emergency planning. In 1966, the Committee noted that many applicants and licensees would rely heavily on local authorities to carry out evacuation, if it should become necessary. There were also no guidelines for judging when an evacuation would be advisable.

The ACRS decided that it should alert the AEC "to a problem area where little efforts being exerted

" [13]. Pressed by the ACRS, the AEC undertook a study of emergency plans and procedures that eventually led to adding a new Appendix E to 10 CFR Part 50

, "Emergency Planning and Preparedness for Production and Utilization Facilities

." Quantitative Safety Goals

In 1979, The ACRS recommended that consideration be given to the establishment of quantitative safety goals for nuclear power reactors. In its May 16, 1979 letter on quantitative safety goals

[14], the ACRS recognized the difficulties and uncertainties in the quantification of risk and acknowledged that in many situations engineering judgment would be the only or the primary basis for a decision. Nevertheless, the Committee believed that the existence of quantitative safety goals and criteria could provide important yardsticks for such judgment [14]. The ACRS was at the forefront of the development of quantitative safety goals. The first set of trial goals (NUREG

-0739) [15] was developed by the ACRS in 1980. These safety goals were the basis for the later NRC work on the development of an NRC Safety Goal Policy in 1983 [16]. PRA Policy Statement

In the early 1990s, the ACRS became concerned about the inconsisten t use of PRA in NRC. In a July 19, 1991 lett er on the consistent use of PRA [17], the ACRS acknowledged

, "PRA can be a valuable tool for judging the quality of regulation, and for helping to ensure the optimal use of regulatory and industry resources." The Committee also stated that it "would have liked to see a deeper and more deliberate integration of the methodology into the NRC activities." The ACRS also pointed to issues such as the inconsistent use of conservatism and the lack of the treatment of uncertainties.

In response to the ACRS, NRC chartered a PRA Working Group an d a Regulatory Review Group to review processes, programs, and practices to identify the feasibility of substituting performance

-based requirements 5 Copyright © 2018 by ASME and guidance founded on risk insights in place of prescriptive requirements [

18]. These efforts led the Commission to issue a policy statement on the use of PRA so that the many potential applications of PRA can be implemented in a consistent and predictable manner that would promote regulatory stability and efficiency

[19]. Risk-Informed Regulations and Practices: ACRS has been very supportive of the evolution toward a risk-informed and performance

-based regulatory system

[20] and had technical oversight of this transformation

. The Committee has followed closely the development of regulatory guidance for the implementation of risk-informed programs and processes, and of PRAs standards that have been used to support these programs and processes.

The ACRS has also taken a leading role in considering some of the challenging issues that have arisen in this effort , such as the application of defense

-in-depth in a risk-informed context.

In its May 19, 1999 letter on the Role of Defense in Depth in a Risk-Informed Regulatory System

[21], ACRS forwarded a paper, prepared by several of its members and an ACRS Senior Fellow, in which two ("Structuralist" and "Rationalist") views of defense-in-depth were discussed along with a preliminary proposal regarding its role in a risk

-informed regulatory system. The ACRS motivation for this had arisen because of instances in which seemingly arbitrary appeals to defense in depth had been used to avoid making changes in regulations or regulatory practices that seemed appropriate in the light of results of quantitative risk analyses.

5. OPERATING REACTORS SAFETY OVERSIGHT The ACRS has always been attentive to safety improvements in operation of nuclear power plants.

In fact , ACRS was the first to recommend periodic comprehensive (ten year) review of operating power reactors. In a June 14, 1966 letter on the subject, the Committee recommended that the AEC "institute a program of periodic comprehensive review of operating licensed power reactors

" [22]. The Committee also recommended that the reports, to be submitted by reactor operators for these comprehensive reviews, "contain summaries of operating history with special emphasis on significant problems" [22]. The ACRS believed that "each reactor operator should be responsible for the maintenance of appropriate records of the design, fabrication, installation testing, and operating history of important plant components

" and this information should be adequate for the comprehensive reviews by the regulatory staff. [22].

ACRS has made valuable contributions over a wide range of issues at operating plants including the following:

Generic Safety Issues (GSIs)

Starting in 1972, the ACRS developed a list of generic items related to construction or operation of light

-water reactors [23]. This grew into a list which was last reported in an ACRS letter dated March 21, 1979

[24]. The work of the NRC staff to resolve these items, as well as the generic items identified by them, became steadily more formal, stemming from the requirement of Section 210 of the Energy Reorganization Act of 1974 which required the NRC to "develop a plan providing for the specification and analysis of unresolved safety issues relating to nuclear reactors

" and "take such action as may be necessary to implement corrective measures with respect to such issues

." The ACRS has made significant contribution toward resolution of many generic safety issues (GSIs). One recent example is the Committee's role in the resolution of GSI

-191 , "Assessment of Debris Accumulation on PWR Sump Performanc e." ACRS was first to express concerns about the effects of chemical reaction products and particle/fiber mats that could form on screens. The Committee was also the first to alarm that increasing screen area, though it could reduce head loss, might result in more fiber debris passing through the screens and increase downstream effects

[25]. Safety Culture:

The concept of safety culture received much attention in the aftermath of the 1979 accident at TMI

-2 which underscored the importance of management and organizational factors to the safe operation of nuclear power plants. The 2002 incident at the Davis

-Besse nuclear power plant renewed the interest in safety culture and provided a n impetus to the deliberations of the ACRS regarding the role and effectiveness of the NRC Reactor Oversight Process (ROP) in monitoring organizational performance.

ACRS organized a workshop on safety culture on June 12, 2003

[26] and issued a report to the Commission on July 16, 2003

[27].

6 Copyright © 2018 by ASME 6. SAFETY RESEARCH REVIEWS Throughout its history

, ACRS has made significant contribution to safety research conducted by the agency. From the very beginning of its establishment, the Committee recognized the need for a safety research program. In a November 16, 1959 letter on proposed study of the reactor hazard and criteria problem, the Committee acknowledged the increasing difficulties which it foresaw in the adequate evaluation of the hazards of reactor facilities due to the absence of a "critical evaluation of the existing data relating to reactor safety" and "the absence of written and agreed upon criteria for judging the adequacy of the proposed design, construction and operation of the various parts of a reactor" [28]. The ACRS believed that the problem required "a study of the available information on reactor safety, arranging it so it is readily available and deriving from it logical conclusions pertinent to answering the questions

A) Is the available knowledge sufficient to set criteria? B) Is more research needed and of what kind?

C) Is this the sort of problem that is not susceptible to solution by planned research and therefore, must primary reliance be placed upon judgment and experience

?"[28]. During its early licensing reviews, the ACRS identified ma n y technical safety issues for further research. Following are some examples of technical issues raised by ACRS for further study:

Radiation Damage to Reactor Pressure Vessel:

In a May 20, 1961 letter on the subject of radiation damage to reactor pressure vessel, ACRS raised its concern about "the potential damage to reactor pressure vessel by virtue of the neutron flux to which they are subjected during their life." [29]. However at that time the vessel failure was not considered "credible". In the mid-1960s, as the size of proposed reactors increased substantially, the ACRS discussed this matter extensively and in its November 24, 1965 letter on the subject of reactor pressure vessels stated the following:

"To reduce further the already small probability of pressure vessel failure, the Committee suggests that the industry and the AEC give still further attention to methods and details of stress analysis, to the development and implementation of improved methods of inspection during fabrication and vessel service life, and to the improvement of means for evaluating the factors that may affect the nil ductility transition temperature and the propagation of flaws during the vessel life

" [30]. Early Concerns about Core Meltdown Accidents:

In 1966, at the "prodding" of ACRS, the AEC established a special task force to look in to the problem of core meltdown [31

]. The task force, chaired by William K. Ergen, a former ACRS member, issued its report in October 1967 [

32]. The report offered assurances about the reliability of ECCS designs and improbability of a core meltdown, but it also acknowledged that a loss-of-coolant accident (LOCA) could cause a breach of containment if the ECCS failed to perform. In an ACRS letter on the task force report, dated February 26, 1968, the Committee strongly recommended that a "positive approach be adopted toward studying the workability of protective measures to cope with core meltdown

" [33]. The Committee also recommended

, as it did in its 1966 report on safety research

, that a "vigorous program be aimed at gaining better understanding of the phenomena and mechanisms important to the course of large

-scale core meltdown

." The task force report and ACRS recommendations formed the basis of some of the most important research initiatives and regulatory decisions by the AEC and the NRC , including the AEC's decision to undertake a study to estimate the probability of a severe accident which resulted in the publication of the landmark Reactor Safety Study (WASH

-1400) [34] and the beginning of the science of probabilistic risk assessment as applied to nuclear power plant safety [2].

In 1977, Section 29 of the Atomic Energy Act was amended to add the following two sentences: "In addition to its other duties under this section, the Committee, making use of all available sources, shall undertake a study of reactor safety research and prepare and submit annually to the Congress a report containing the results of such study. The first such report shall be submitted to the Congress no later than December 31, 1977." ACRS had been submitting an annual report on NRC Safety Research Program to Congress from 1977 until 1997. In 1998, Public Law 105

-362 struck those two sentences in Section 29. In 1997, the Commission transferred the research advisory function of the Nuclear Safety Research Review Committee (NSRRC) to the ACRS. In this 7 Copyright © 2018 by ASME role, the ACRS was directed to "examine the need, scope, and balance of the reactor safety research program" [35]. The Committee was also directed to "consider how well the Office of Research anticipates research needs and how it is positioned for the changing environment

" [35]. Since 1998, ACRS has been submitting reports to the Commission on review and evaluation of the NRC Safety Research Program, initially annually and after 2004 biennially.

Since 2004, the ACRS has also been assisting the NRC Office of Nuclear Regulatory Research in an independent evaluation of the quality of its research programs. An analytical/deliberative decision

-making framework has been adopted for evaluating the quality of NRC research projects. The definition of quality research adopted by the Committee includes general attributes such as soundness of technical approach and results, justification of major assumptions, and treatment of uncertainties/ sensitivities.

7. NUCLEAR MATERIALS AND WASTE Before the establishment of ACNW in 1988 , ACRS reviewed matters related to the long

-term management of radioactive wastes produced within the nuclear industry. Since the merging of ACNW&M into the ACRS in 2008, the Committee has been reviewing many aspects of nuclear waste management such as handling, processing, transportation, and storage of nuclear wastes including spent fuel and nuclear wastes mixed with other hazardous substances.

8.

SUMMARY

AND CONCLUSION Through more than 60 years of its history, the ACRS has made significant contributions to nuclear safety. . The Committee's early licensing reviews led to evolution of many new safety requirements dealing with a wide range of technical issues.

As the ACRS moved into the 1980s, the Committee shifted much of its attention from plant design and construction to improvements in both the operation and regulation of nuclear power plants. Throughout its history, an essential activity of the ACRS has also been reviewing the research sponsored by the agency.

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2. U.S. Nuclear Regulatory Commission, "The Role of the ACRS in Nuclear Regulation and Safety:

Looking Back, Looking Forward," Remarks by Chairman Richard A. Meserve before the ACRS Symposium on Role of Advisory Committees in March 4, 2003.

3. Nourbakhsh, H. P. "60 Years of Contributing to Reactor Safety: Insights on ACRS Power Reactor Licensing Reviews," To be presented at the 2018 International Congress on Advances in Nuclear Power Plants (ICAPP 18), Charlotte, NC, April 8-11, 2018. 4. Advisory Committee on Reactor Safeguards, Report from Herbert Kouts, Chairman of ACRS, to the Honorable Glenn T. Seaborg, Chairman, U.S. Atomic Energy Commission,

Subject:

Report on Connecticut

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Subject:

Report on the City of Los Angeles

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Subject:

Report on Edwin I. Hatch Nuclear Plant, May 15, 1969.

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Subject:

Report on Brunswick Steam Electric Plant Units 1 and 2, May 15, 1969.

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Powers, "The Evolution of the U.S. Nuclear Regulatory Process, "Progress in Nuclear Energy, Vol. 102 , pp. 79

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21. Advisory Committee on Reactor Safeguards, Letter from Dana A. Powers, ACRS Chairman, to Shirley Ann Jackson, NRC Chairman,

Subject:

The Role of Defense in Depth in a Risk

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22. Advisory Committee on Reactor Safeguards, Letter from David Okrent, ACRS Chairman, to Glenn T. Seaborg, NRC Chairman,

Subject:

Periodic Comprehensive (Ten Year) Review of Operating Power reactors., June 14, 1966.

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Status of Generic Items Relating to Light

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Subject:

Status of Generic Items Relating to Light

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Subject:

Safety Culture, July 16, 2003.

28. Advisory Committee on Reactor Safeguards, Letter from C. Rogers McCulloogh, ACRS Chairman, to John A. McCone, NRC Chairman,

Subject:

Proposed Study of the Reactor Hazard and Criteria Problem, December 16, 1959.

29. Advisory Committee on Reactor Safeguards, Letter from T. J. Thompson, Chairman of ACRS, to Glenn T. Seaborg, Chairman, U.S. Atomic Energy Commission,

Subject:

Radiation Damage to Reactor Pressure Vessels, May 20, 1961.

30. Advisory Committee on Reactor Safeguards, Letter from W.D. Manly, Chairman of ACRS, to Glenn T. Seaborg, Chairman, U.S. Atomic Energy Commission,

Subject:

Reactor Pressure Vessels, November 24, 1965.

31. Walker J. S. and T. R. Wellock, "A Short History of Nuclear Regulation, 1946

-2009," NRC , NUREG/BR-0175, 2010. 32. Ergen, W. K. "Emergency Core Cooling: Report of Advisory Task Force on Power Reactor Emergency Cooling," U.S. Atomic Energy Commission (1967).

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Atomic Energy Commission,

Subject:

Report on Advisory Task Force on Power Reactor Emergency Cooling, February 26, 1968.

34. U.S. Nuclear Regulatory Commission, "Reactor Safety Study

-An Assessment of Accident Risks in U.S. Commercial Nuclear Power Plants,"

WASH-1400 (NUREG

-75/014), 1975.

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Nuclear Safety Research Review Committee (NSRRC), September 9, 1997.