ML24088A028

Hearing Request and Petition to Intervene by Beyond Nuclear and Sierra Club

ML24088A028
Person / Time
Site:	North Anna
Issue date:	03/28/2024
From:	Curran D Beyond Nuclear, Harmon, Curran, Harmon, Curran, Spielberg & Eisenberg, LLP, Sierra Club
To:	NRC/SECY
SECY RAS
References
RAS 56972, 50-338 SLR-2, 50-339 SLR-2
Download: ML24088A028 (0)

Text

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE SECRETARY

)

In the Matter of )

Virginia Electric Power Co. ) Docket Nos. 50-338/339 SLR North Anna Power Station Units 1 & 2 ) March 28, 2024

___________________________________ )

HEARING REQUEST AND PETITION TO INTERVENE BY BEYOND NUCLEAR AND THE SIERRA CLUB

I. INTRODUCTION

Pursuant to 10 C.F.R. § 2.309, and the hearing notice published at 89 Fed. Reg. 960 (Jan.

8, 2024), Petitioners Beyond Nuclear, Inc. (Beyond Nuclear) and the Sierra Club, Inc. (Sierra

Club) hereby request the U.S. Nuclear Regulatory Commission (NRC or Commission) to

grant a hearing on new information discussed in the Draft Supplemental Environmental Impact

Statement (Draft SEIS) prepared by the NRC to inform its review of an application for

subsequent license renewal (SLR) of the operating license for the North Anna Units 1 and 2 nuclear power station (NAPS). 1 If VEPCOs application is granted, it will be allowed to

operate North Anna Units 1 and 2 for an additional twenty years beyond its current renewed operating license term, or until 2058 (Unit 1) and 2060 (Unit 2), for an aggregate of 80 years. 2

Petitioners contend that the NRC should not approve subsequent renewal of VEPCOs

operating license because the Draft SEIS fails to support its conclusion that the environmental

1 The Draft North Anna EIS is entitled: Site-Specific Environmental Impact Statement for License Renewal of Nuclear Plants, Supplement 7a, Subsequent License Renewal for North Anna Power Station Units 1 and 2, Draft Report for Comment (NUREG-1437, Supplement 7a, Dec. 2023).

2 The NRC issued original operating licenses for North Anna in 1978 and 1980, with expiration dates of 2018 (Unit 1) and 2020 (Unit 2). In 2003, the NRC renewed both licenses for 20 years, with new expiration dates of 2038 and 2040. If renewed again, the North Anna licenses would expire in 2058 and 2060.

impacts of accidents are SMALL or insignificant. In particular, the Draft SEIS fails to address

environmental significance of 2011 Mineral Earthquake; provides incomplete, inadequate,

incorrect or misleading data and analyses in support of its general conclusion that severe accident

impacts are small; and fails to address the effects of climate change on accident risk. Each of

these categories of deficiencies is significant in its own right. Taken together, they show a level

of inadequacy that is grossly unacceptable.

Petitioners contentions are supported by the expert declaration of Jeffrey T. Mitman, a nuclear engineer with a significant level of expertise in risk analysis. 3

The remainder of this Hearing Request is organized as follows: Section II contains a

demonstration that Petitioners Beyond Nuclear and the Sierra Club each has organizational

standing to participate in this proceeding.Section III presents the legal framework for

Petitioners Hearing Request.Section IV presents Petitioners Contentions.Section V contains

Petitioners Conclusion.

II. PETITIONERS HAVE STANDING TO REQUEST A HEARING.

Pursuant to 10 C.F.R. § 2.309(d), a request for a hearing must address: (1) the nature of

the petitioners right under the Atomic Energy Act to be made a party to the proceeding, (2) the

nature and extent of the petitioners property, financial, or other interest in the proceeding, and

(3) the possible effect of any order that may be entered in the proceeding on the petitioners

interest. The Atomic Safety and Licensing Board (ASLB) has summarized these standing

requirements as follows:

In determining whether a petitioner has sufficient interest to intervene in a proceeding, the Commission has traditionally applied judicial concepts of standing. Contemporaneous judicial standards for standing require a petitioner to demonstrate that (1) it has suffered

3 Declaration of Jeffrey T. Mitman (March 27, 2024) (Mitman Declaration). Mr. Mitmans Declaration is attached as Attachment 1.

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or will suffer a distinct and palpable harm that constitutes injury-in-fact within the zone of interest arguably protected by the governing statutes ( e.g., the Atomic Energy Act of 1954 and the National Environmental Policy Act of 1969); (2) the injury can fairly be traced to the challenged actions; and (3) the injury is likely to be redressed by a favorable decision. An organization that wishes to intervene in a proceeding may do so either in its own right by demonstrating harm to its organizational interests, or in a representational capacity by demonstrating harm to its members. To intervene in a representational capacity, an organization must show not only that at least one of its members would fulfill the standing requirements, but also that he or she has authorized the organization to represent his or her interests. 4

As demonstrated below, each of the Petitioners has standing by virtue of organizational

interests that fall within the zone of interests protected by the Atomic Energy Act and the

National Environmental Policy Act (NEPA). By intervening in this proceeding, Petitioners

seek to protect their members health and safety, as well as protection of the environment. They

wish to ensure that VEPCOs operating license is not approved for a second renewal term unless

and until VEPCO demonstrates full compliance with NEPAs requirements for protection of

public health and the environment.

In addition, as also demonstrated below, each Petitioner organization has members and/or

staff who live and/or work within 50 miles of North Anna Units 1 and 2, whose interests in

protecting their own health and the health of the environment would be adversely affected by

extended operation of North Anna Units 1 and 2 under an additional SLR term, and who have

authorized Petitioners to represent their interests in this proceeding. Therefore, Petitioners have

presumptive standing by virtue of the location of their members residences and property within 50 miles of the North Anna reactors. 5

4 Pacific Gas & Electric Co. (Diablo Canyon Power Plant Independent Spent Fuel Storage Installation), LBP-02-23, 56 N.R.C. 413, 426 (2002) ( petition for review denied, CLI-03-12, 58 N.R.C. 185 (2003)).

5 Diablo Canyon, 56 N.R.C. at 426-27 (citing Florida Power & Light Co. (Turkey Point Nuclear Generating Plant, Units 3 and 4), LBP-01-06, 53 N.R.C. 138, 146, affd, CLI-01-17, 54 N.R.C. 3 (2001)).

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A. Standing of Beyond Nuclear

Beyond Nuclear is a nonprofit, nonpartisan membership organization that aims to educate

and activate the public about the connections between nuclear power and nuclear weapons and

the need to abolish both to protect public health and safety, prevent environmental harms, and

safeguard our future. Beyond Nuclear advocates for an end to the production of nuclear waste

and for securing the existing reactor waste in hardened on-site storage until it can be permanently

disposed of in a safe, sound, and suitable underground repository. For more than fifteen years,

Beyond Nuclear has worked toward its mission by regularly intervening in NRC licensing,

relicensing, and other proceedings related to nuclear safety matters.

Beyond Nuclears representational standing to participate in this proceeding is

demonstrated by the attached declarations of its members: Declaration of Declaration of Glen

Besa (March 23, 2024) (Attachment 2A); Declaration of Erica Gray (March 23, 2024)

(Attachment 2B); and Declaration of Jerry Rosenthal (March 24, 2024); (Attachment 2C).

B. Standing of the Sierra Club

Founded in 1892, the Sierra Club is a national environmental organization with more than 3.8

million members across the United States. The purposes of the Sierra Club are to explore, enjoy,

and protect the wild places of the earth; to practice and promote the responsible use of the earths

ecosystems and resources; to educate and enlist humanity to protect and restore the quality of the

natural and human environment; and to use all lawful means to carry out these objectives.

The Sierra Clubs representational standing to participate in this proceeding is demonstrated

by the attached declarations of its members: Declaration of Barbara Cruikshank (March 23,

2024) (Attachment 2D); Declaration of John Cruikshank (March 22, 2024) (Attachment 2E);

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Declaration of Diana Johnson (March 23, 2024) (Attachment 2F); Declaration of William J.

Johnson (March 23, 2024) (Attachment 2G).

III. LEGAL FRAMEWORK: ATOMIC ENERGY ACT AND NEPA

The NRCs regulation and licensing of reactors is governed by two statutes: the Atomic Energy Act, 42 U.S.C. § 2011, et seq.; and NEPA, 42 U.S.C. §§ 4321-4370h. While the

substantive concerns of these statutes overlap, Citizens for Safe Power v. NRC, 524 F.2d 1291,

1299 (D.C. Cir. 1975), they impose independent procedural obligations. Limerick Ecology Action

v. NRC, 869 F.2d 719, 729-31 (3rd Cir. 1989). Even where the NRC purports to have resolved

safety issues through its Atomic Energy Act-based regulatory process, it must nevertheless

comply with NEPAs procedural obligations for addressing those issues in its decision-making processes. 6

A. Atomic Energy Act and NRC Safety Regulations

Under § 103(d) of the Atomic Energy Act, the NRC may not issue an operating license

for a nuclear plant if it would be inimical to the common defense and security or to the health

and safety of the public. 42 U.S.C. § 2133(d). Section 161 of the Atomic Energy Act also

empowers the NRC to set standards to protect health or to minimize danger to life or property,

inter alia. 42 U.S.C. § 2201(b).

Among the many regulatory standards promulgated by the NRC for the safe construction

and operation of nuclear power reactors, the General Design Criteria (GDCs) in Appendix A to

10 C.F.R. Part 50 are fundamentally important, because they establish minimum requirements

6 Limerick Ecology Action, 869 F.2d at 729-31. See also State of New York v. NRC, 681 F.3d 471, 478 (D.C. Cir. 2012) (a finding that reasonable assurance exists that sufficient mined geologic repository capacity will be available when necessary... does not describe a probability of failure so low as to dismiss the potential consequences of such a failure.).

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for the principal design criteria for water-cooled nuclear power plants. Id., Introduction. These

principal design criteria, in turn, establish:

the necessary design, fabrication, construction, testing, and performance requirements for structures, systems, and components important to safety; that is, structures, systems, and components that provide reasonable assurance that the facility can be operated without undue risk to the health and safety of the public.

Id. General Design Criterion (GDC) 2, Design Bases for Protection Against Natural

Phenomena) requires that [s]tructures, systems, and components important to safety shall be

designed to withstand the effects of natural phenomena such as earthquakes, tornadoes,

hurricanes, floods, tsunami, and seiches without loss of capability to perform their safety

functions. As the NRCs ASLB has recognized, SSCs must be able to withstand an earthquake and other natural disasters within the design basis of the plant. 7 Design-basis structures that

must remain functional in the event of a safe shutdown earthquake are referred to as Category I structures. 8 Category I safety structures and components (SSCs) encompass a broad array of

equipment and structures, including the pressure vessel internals, the reactor coolant pressure boundary, the steam generators, and the emergency core cooling system. 9

B. NEPA General Requirements

NEPA implements a broad national commitment to protecting and promoting environmental

quality. Louisiana Energy Services, L.P. (Claiborne Enrichment Center), CLI-98-3, 47 N.R.C.

77, 87 (1998) (quoting Robertson v. Methow Valley Citizens Council, 490 U.S. 332, 348 (1989)

and citing 42 U.S.C. § 4331). NEPA has two key purposes: to ensure that the agency will have

7 NextEra Energy Seabrook, LLC (Seabrook Station, Unit 1), LBP-17-07, 86 N.R.C. 59, 79 (2017) (citing 10 C.F.R. Part 50, App. A, GDC 2).

8 Regulatory Guide, 1.29, Rev. 5, Seismic Design Classification for Nuclear Power Plants at 5 (July 2016) (ML16118A148) (Reg. Guide 1.29).

9 Id. at 5-6.

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available, and will carefully consider, detailed information concerning significant environmental

impacts before it makes a decision; and to guarantee that the relevant information will be made

available to the larger audience that may also play a role in the decision-making process and implementation of that decision. Robertson, 490 U.S. at 349. In fulfilling NEPAs second

purpose of public participation, the agencys environmental analysis must be published for

public comment to permit the public a role in the agencys decision-making process.

Robertson, 490 U.S. at 349-50; Hughes River Watershed Conservancy v. Glickman, 81 F.3d 437,

443 (4th Cir. 1996).

In fulfilling NEPAs first purpose of evaluating the environmental impacts of its

decisions, NEPA requires a federal agency to take a hard look at potential environmental

consequences by preparing an EIS prior to any major Federal action[] significantly affecting the quality of the human environment. Robertson, 490 U.S. at 349; 42 U.S.C. § 4332(c). The

hallmarks of a hard look are thorough investigation into environmental impacts and forthright

acknowledgment of potential environmental harms. National Audubon Society v. Dept of Navy,

422 F.3d 174, 185 (4th Cir. 2005). The requirement to analyze environmental impacts in a draft

EIS is codified in NRC regulation 10 C.F.R. § 51.71 (requiring that a draft EIS must include a

preliminary analysis that considers and weighs the environmental effects, including any

cumulative effects, of the proposed action; the environmental impacts of alternatives to the

proposed action; and alternatives available for reducing or avoiding adverse environmental

effects.).

C. Reasonably Foreseeable Harms Covered by NEPA Include Climae Change

A NEPA analysis must address harms that are reasonably foreseeable, even if they are

indirect or unlikely. State of New York, 681 F.3d at 476, 482. The analysis must address both

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the probabilities of potentially harmful events and the consequences if those events come to

pass. State of New York, 681 F.3d at 482 (rejecting environmental analysis of spent fuel pool

fire risks because it did not consider consequences as well as probability of fires in spent fuel

storage pools).

The Presidents Council on Environmental Quality (CEQ) has concluded that climate

change is a fundamental environmental issue, and its effects fall squarely within NEPAs purview. 10 Among the climate-related environmental impacts that CEQ advises agencies to

consider are the reasonably foreseeable effects of climate change on infrastructure investments. 11 As stated by the CEQ:

The effects of climate change observed to date and projected to occur in the future include more frequent and intense heat waves, longer fire seasons and more severe wildfires, degraded air quality, increased drought, greater sea-level rise, an increase in the intensity and frequency of extreme weather events, harm to water resources, harm to agriculture, ocean acidification, and harm to wildlife and ecosystems. The IPCC [Intergovernmental Panel on Climate Change] Assessment Report reinforces these findings by providing scientific evidence of the impacts of climate change driven by human-induced GHG emissions, on our ecosystems, infrastructure, human health, and socioeconomic makeup. 12

Consistent with this policy, multiple federal agencies have established programs for

assessing the effects of climate change on critical infrastructure such as power plants,

transmission systems, and dams. For instance, the Department of Defense has initiated a

Climate Risk Analysis to address the implications for U.S. national security and defense of

[i]ncreasing temperatures; changing precipitation patterns; and more frequent, intense, and unpredictable extreme weather conditions caused by climate change. 13 The Federal Emergency

10 National Environmental Policy Act Guidance on Consideration of Greenhouse Gas Emissions and Climate Change, 88 Fe3d. Reg. 1,196, 1,197 (Jan. 9, 2023).

11 Id.

12 Id. at 1,200 (emphasis added).

13 See https://www.defense.gov/spotlights/tackling-the-climate-crisis/. (last visited 3/27/24).

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Management Agency has declared that the Changing Climate is a Priority for Emergency

Managers because the changing climate is a force multiplier - increasing the number of

storms, floods, fires, and extreme temperatures that threaten the well-being of people across our nation. 14 The Critical Infrastructure Security Agency analyzes extreme weather and its impacts

to critical infrastructure to develop strategies for resilience. 15

IV. CONTENTIONS

Contention 1: Draft SEIS Fails to Address Environmental Significance of the 2011 Mineral Earthquake

A. Statement of Contention

The Draft SEIS fails to satisfy NEPA or NRC implementing regulation 10 C.F.R. § 51.71

because it does not address the environmental significance of the 2011 Mineral Earthquake,

whose epicenter was a short distance from the two reactors and whose ground motion exceeded

the design basis levels for both reactors. By exceeding the reactors design basis, the earthquake

disproved the assumption underlying the NRCs issuance of operating licenses in 1978 (for Unit

1) and 1980 (for Unit 2) and renewal of those licenses 2003, that the reactors could be operated

safely and without significant adverse environmental impacts because their SSCs were built to a

design basis of sufficient rigor to protect against likely earthquakes. This assumption can also be

found in the 2013 License Renewal GEIS and the Draft SEIS for the North Anna SLR application. 16

14 See https://www.fema.gov/fact-sheet/fema-and-changing-climate#:~:text=The%20Changing%20Climate%20is%20a%20Priority%20for%20Emergency

%20Managers&text=When%20emergency%20managers%20plan%20for,recovery%20starts%2 0sooner%20for%20survivors. (last visited 3/27/24).

15 See https://www.cisa.gov/topics/critical-infrastructure-security-and-resilience/extreme-weather-and-climate-change. (last visited 3/27/24).

16 Mitman Declaration, ¶ 25 and Draft SEIS as cited therein.

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Because that assumption has been proven wrong, the NRC must explicitly address the

question of whether the environmental impacts of operating North Anna Units 1 and 2 in non-

compliance with its design basis for an additional twenty years will have significant impacts. As

discussed in the attached Mitman Declaration, the NRC fails to acknowledge it or explain the

fundamental difference between a finding of no significant or small impact that is based on a

deterministic analysis and a finding of no significant impact that is based on a probabilistic

analysis. In Mr. Mitmans expert opinion, the deterministic analysis is more conservative

because it requires a robust design that provides reasonable assurance that an external event like

an earthquake will not harm necessary safety systems. A probabilistic analysis, in comparison,

does not assume safety related equipment will perform as designed and then calculates the

likelihood of an accident occurring. The NRC should explain the difference and how its assessment of risk has changed as a result of the Mineral Earthquake. 17 As asserted by Mr.

Mitman, the NRC should also explain what it has done to evaluate the potential that safety

systems, which are assumed to survive a beyond-design-basis earthquake only once will be able to perform their safety functions when the next earthquake occurs. 18

Further, the Draft SEIS does not address, let alone reconcile, the significant disparity

between the results of the seismic risk analyses for Unit 3 and Units 1 and 2. In both cases, the

NRC and VEPCO were responding to the very same earthquake. Yet, while the NRC required

seismic upgrades for Unit 3, it required no seismic upgrades for Units 1 and 2 which required

only a set of nonpedigree commercial-grade FLEX components with significantly lower

reliability. The NRC should explain the reason for this disparate result. If the NRC considered

17 See Mitman Declaration, ¶ 27.

18 Id.

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significant safety grade improvements necessary for adequate protection of Unit 3, the obvious

conclusion is that it thought the safety and environmental impacts of an earthquake were significant. Why did it make a different finding for Units 1 and 2? 19

B. Basis Statement

Petitioners rely for this Contention on Sections B and C.1 of Mr. Mitmans Declaration.

Petitioners also rely on the legal authorities cited above in Section III. In particular, Petitioners

rely on Citizens for Safe Power, 524 F.2d at 1299 (substantive concerns of Atomic Energy Act

and NEPA overlap); Limerick Ecology Action, 869 F.2d at 729-31 (despite overlap, the Atomic

Energy Act and NEPA impose independent procedural obligations); and State of New York v.

NRC, 681 F.3d at 478 (reasonable assurance findings do not excuse NEPA compliance unless

probability of impacts is so low as to dismiss the potential consequences of such a failure.).

Even where the NRC purports to have resolved safety issues through its Atomic Energy Act-

based regulatory process, it must nevertheless comply with NEPAs procedural obligations for

addressing those issues in its decision-making processes.

C. Demonstration that the Contention is Within the Scope of the Proceeding

This Contention is within the scope of the SLR proceeding for North Anna Units 1 and 2

because it seeks compliance by the NRCs environmental review with NEPA and NRCs

implementing regulations. The subject matter of the Contention falls within the scope of new information as described in the hearing notice 20 because it concerns a new reactor-specific

accident analysis in the Draft SEIS that takes the place of a previous environmental analysis for

19 See Mitman Declaration, ¶ 28.

20 89 Fed. Reg. at 962.

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which the NRC had unlawfully relied on the 2013 License Renewal Generic Environmental Impact Statement (GEIS). 21

D. Demonstration that the Contention is Material to the Findings NRC Must Make to Renew VEPCOs Operating License

This Contention is material to the findings NRC must make regarding the environmental

impacts of re-licensing North Anna Units 1 and 2 for a second renewed license term, because it

challenges the adequacy of the Draft SEIS to support the NRCs proposed findings that the

environmental impacts of re-licensing NAPS are SMALL.

E. Concise Statement of the Facts or Expert Opinion Supporting the Contention, Along with Appropriate Citations to Supporting Scientific or Factual Materials

The facts supporting Petitioners Contention are stated in the Contention itself and in the

attached Mitman Declaration.

CONTENTION 2: Draft SEIS does not contain a complete or adequately rigorous evaluation of accident risks.

A. Statement of Contention

The Draft SEIS does not contain a complete or adequately rigorous evaluation of accident

risks because essential data are missing and important analytical assertions are erroneous or

misleading. Therefore, the NRC lacks an adequate basis for concluding that the environmental impacts of accidents during a license renewal term are SMALL. 22 In particular, and as set forth

in detail in Section C.2 of Mr. Mitmans Declaration:

21 See Duke Energy Carolinas, L.L.C. (Oconee Nuclear Station, Units 1, 2, and e), CLI-22-3, 95 N.R.C. 40 (2022) (CLI-22-03). See also Florida Power & Light Co. (Turkey Point Nuclear Generating Units 3 and 4), LBP-24-03, slip op. at 13-16 (March 7, 2024) (LBP-24-03)

(allowing petitioners to submit arguments that pre-date the draft SEIS for Turkey Point if they were based on the draft SEIS).

22 Id. at 3-169 170.

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The Draft SEIS is inadequate as a general matter for making broad generalizations about

external event core damage frequency (CDF) based on extrapolations from internal

event CDF values and limited actual plant-specific values for external event CDF.

In finding that the environmental impacts of severe accidents are SMALL, the NRC

ignores its own data regarding seismic and fire core damage frequency (CDF) that

indicate these impacts are significant. The NRC also disregards the fact that the

occurrence of the 2011 Mineral Earthquake, by itself, increased the risk of an earthquake

severe enough to damage safety equipment.

The Draft SEIS assertion at page F-26 that there has been a substantial decrease in

internal event CDF is erroneous. This error affects other estimates such as the estimate

of population dose risk.

The Draft SEIS fails to demonstrate consideration of external flooding risk with

subsequent ingress of water into the turbine building. As demonstrated by Mr. Mitmans

Declaration, flooding poses a significant accident risk that has not been addressed in the

Draft SEIS.

The Draft SEIS makes misleading statements about the NRCs review of Fukushima-

related information relevant to North Anna and risk improvements obtained by NRC and

license efforts after September 2001.

The Draft SEIS takes inappropriate credit for reductions in environmental risk that are

not reflected in the PRA for NAPS.

The Draft SEIS fails to demonstrate consideration of uncertainties with respect to the

conclusion that severe accident impacts are small.

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The Draft SEIS does not address the environmental impacts of concurrent multi-unit

accidents.

The Draft SEIS severe accident mitigation alternatives (SAMA) analysis is deficient

in multiple respects, including failure to consider SAMAs that meet criteria for

consideration, and failure to provide documentation of an NRC audit relied on to

conclude that VEPCOs approach to its SAMA analysis was methodical and reasonable.

B. Basis Statement

Petitioners rely for this Contention on Section C.2 of Mr. Mitmans Declaration. Petitioners

also rely on the legal authorities cited above in Section III. In particular, Petitioners rely on Robertson, 490 U.S. at 349 (requiring hard look at potential environmental consequences) and

National Audubon Society, 422 F.3d at 185 (hallmarks of a hard look are thorough

investigation into environmental impacts and forthright acknowledgment of potential

environmental harms.). In addition, Petitioners rely on the NRC guidance for preparation and use of probabilistic risk assessments (PRAs) in both safety and environmental documents. 23

C. Demonstration that the Contention is Within the Scope of the Proceeding

This Contention is within the scope of the SLR proceeding for North Anna Units 1 and 2

because it seeks compliance by the NRCs environmental review with NEPA and NRCs

implementing regulations. The subject matter of the Contention falls within the scope of new information as described in the hearing notice 24 because it concerns a new reactor-specific

accident analysis in the Draft SEIS that takes the place of a previous environmental analysis for

23 See Mitman Declaration, § C.2, pars. 43 - 44 and notes 43 -49.

24 89 Fed. Reg. at 962.

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which the NRC had unlawfully relied on the 2013 License Renewal Generic Environmental Impact Statement (GEIS). 25

D. Demonstration that the Contention is Material to the Findings NRC Must Make to Renew VEPCOs Operating License

This Contention is material to the findings NRC must make regarding the environmental

impacts of re-licensing North Anna Units 1 and 2 for a second renewed license term, because it

challenges the adequacy of the Draft SEIS to support the NRCs proposed findings that the

environmental impacts of re-licensing NAPS are SMALL.

E. Concise Statement of the Facts or Expert Opinion Supporting the Contention, Along with Appropriate Citations to Supporting Scientific or Factual Materials

The facts supporting Petitioners Contention are stated in the Contention itself and in the

attached Mitman Declaration.

CONTENTION 3: Draft SEIS fails to address the effects of climate change on accident risk.

A. Statement of Contention

The Draft SEIS fails to satisfy NEPA or NRC implementing regulation 10 C.F.R. § 51.71

because it does not address the effects of climate change on accident risk. No such discussion

can be found in Section 3.11.6.9 or Appendix F. To the contrary, the NRC asserts that the effects of climate change are outside the scope of the NRC staffs SLR review. 26 In support of

this assertion, the NRC claims to consider climate-related information in its licensing reviews and ongoing oversight. 27 But this is exactly the kind of blindered reasoning that was rejected in

25 See CLI-22-03; LBP-24-03.

26 Id. at 3-194.

27 Id.

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State of New York. The fact that NRC plans to address climate change risks in the future does not

excuse the agency from addressing the risks as they are understood at this time. Only if the NRC

can say that the effects of climate change are so small as to be remote and speculative can it avoid addressing those effects in its environmental review. 28 And the Executive Branch of the

U.S. government, including CEQ and other federal agencies, has stated in no uncertain terms that

climate change poses a current and future threat to critical infrastructure that should be addressed now in NEPA reviews and all other decision-making processes. 29

Further, as set forth in Mr. Mitmans Declaration, the Draft SEIS failure to address climate

change impacts on accident risk constitutes a significant deficiency because climate change

demonstrably affects the frequency and intensity of some external events and therefore has the

potential to significantly increase accident risks. Moreover, the frequency and intensity of climate change effects are increasing over time. 30

Mr. Mitman also presents an illustration of how the reasonably foreseeable increase in the

frequency and volume of flooding could significantly increase the risk of a serious accident at NAPS.31 This is just one example of the increased accident risk that can be reasonably expected

due to climate change and that should be addressed in the Draft SEIS.

B. Basis Statement

Petitioners rely for this contention on Mr. Mitmans Declaration and the legal authorities

cited above in Section III. In particular, Petitioners rely on State of New York v. NRC, 681 F.3d at

478 (reasonable assurance findings do not excuse NEPA compliance unless probability of

28 681 F.3d at 478.

29 See discussion above in Section III.C.

30 Mitman Declaration, ¶ 48.

31 Mitman Declaration, ¶ 51.

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impacts is so low as to dismiss the potential consequences of such a failure.). In addition,

Petitioners rely on the CEQ guidance discussed above in Section III.C. While this guidance is

not binding on the NRC, it should be given substantial deference. State of New York v. NRC, 681

F.3d at 476 (citing Andrus v. Sierra Club, 442 U.S. 347, 358 (1979).

C. Demonstration that the Contention is Within the Scope of the Proceeding

This Contention is within the scope of the SLR proceeding for North Anna Units 1 and 2

because it seeks compliance by the NRCs environmental review with NEPA and NRCs

implementing regulations. The subject matter of the Contention falls within the scope of new information as described in the hearing notice 32 because it concerns a new reactor-specific

accident analysis in the Draft SEIS that takes the place of a previous environmental analysis for

which the NRC had unlawfully relied on the 2013 License Renewal Generic Environmental Impact Statement (GEIS). 33

D. Demonstration that the Contention is Material to the Findings NRC Must Make to Renew VEPCOs Operating License

This Contention is material to the findings NRC must make regarding the environmental

impacts of re-licensing North Anna Units 1 and 2 for a second renewed license term, because it

challenges the adequacy of the Draft SEIS to support the NRCs proposed findings that the

environmental impacts of re-licensing NAPS are SMALL.

E. Concise Statement of the Facts or Expert Opinion Supporting the Contention, Along with Appropriate Citations to Supporting Scientific or Factual Materials

The facts supporting Petitioners Contention are stated in the Contention itself and in the

attached Mitman Declaration.

32 89 Fed. Reg. at 962.

33 See CLI-22-03; LBP-24-03.

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V. CONCLUSION

For the foregoing reasons, Petitioners Hearing Request and Waiver Request should be granted.

Respectfully submitted,

__/signed electronically by/___

Diane Curran Harmon, Curran, Spielberg, & Eisenberg, L.L.P.

1725 DeSales Street N.W., Suite 500 Washington, D.C. 20036 240-393-9285 dcurran@harmoncurran.com

March 28, 2024

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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE SECRETARY

)

In the Matter of )

Virginia Electric Power Co. ) Docket Nos. 50-338/339 SLR North Anna Power Station, Units 1 and 2 )

___________________________________ )

CERTIFICATE OF SERVICE

I certify that on March 28, 2024, I posted HEARING REQUEST AND PETITION TO INTERVENE BY BEYOND NUCLEAR AND THE SIERRA CLUB, including Attachment 1 and Attachments 2A-2G, on the NRCs Electronic Information Exchange.

___/signed electronically by/__

Paul Gunter

19

ATTACHMENT 1

Declaration of Jeffrey T. Mitman UNITED STATES OF AMERICA BEFORE THE SECRETARY

)

In the Matter of )

Virginia Electric Power Co. ) Docket Nos. 50-338/339 SLR North Anna Power Station Units 1 & 2 )

___________________________________ )

DECLARATION OF JEFFREY T. MITMAN

Under penalty of perjury, I, Jeffrey T. Mitman declare:

1. My name is Jeffrey T. Mitman. I have been retained by Beyond Nuclear, Inc. (Beyond Nuclear) and the Sierra Club, Inc. (Sierra Club) to evaluate the U.S. Nuclear Regulatory Commissions (NRCs) draft supplemental environmental impact statement (EIS) for subsequent license renewal (SLR) of the operating licenses for North Anna Power Station, Units 1 and 2 (hereinafter North Anna Draft SEIS). 1

A. Expert Qualifications and Relevant Experience

2. By education and experience, I am a nuclear engineer, with a significant level of expertise in nuclear power and risk analysis.

3. As set forth in my attached Curriculum Vitae (Exhibit 1), I have more than 40 years of experience in the nuclear industry and 16 years as a regulator with the U.S. Nuclear Regulatory Commission (NRC). My experience includes 16 years on the technical staff of the NRC as a Reliability and Risk Analyst. During my last 15 years at the NRC, I served as Senior Reliability and Risk Analyst, with significant responsibility for managing a number of risk analysis projects and teams.

4. During my employment in the nuclear industry and the NRC, I became very familiar with NRC regulations and guidance regarding nuclear power plant safety, and with the application of risk analysis to reactor safety analysis. I am also generally familiar with the NRCs conceptual approach to the analysis of Severe Accident Mitigation (SAMA) alternatives in the context of reactor license renewal.

5. As an NRC Staff member, I participated in NRC safety reviews and performed risk analysis on U.S. nuclear reactors. I also worked on the NRCs restart readiness review for North Anna Units 1 and 2 (NAPS) after the Mineral Virginia earthquake in 2011. I also participated in reviews related to the risk to Oconee Nuclear Plant, Units 1, 2 and 3

1 The Draft North Anna EIS is entitled: Site-Specific Environmental Impact Statement for License Renewal of Nuclear Plants, Supplement 7a, Subsequent License Renewal for North Anna Power Station Units 1 and 2, Draft Report for Comment (NUREG-1437, Supplement 7a, Dec. 2023).

(Oconee) posed by potential failure of the upstream Jocassee Dam. In addition, I coauthored a generic study by NRC of dam failure risk, with particular application to Oconee.

6. In 2021, I retired from the NRC and became a private consultant.

7. In 2023, on behalf of Beyond Nuclear and the Sierra Club, I prepared a written report regarding my evaluation of the accident risk analysis in the NRCs Draft Generic Environmental Impact Statement for License Renewal of Nuclear Plants (NUREG-1437, Rev. 2, Feb. 2023). (Draft License Renewal GEIS). Beyond Nuclear and the Sierra Club submitted my report to the NRC in support of their comments on the Draft License Renewal GEIS. 2

2 Comments by Beyond Nuclear And The Sierra Club on Proposed Rule And Draft Generic Environmental Impact Statement For Renewing Nuclear Power Plant Licenses (May 2, 2023; corrected May 12, 2023); attached Declaration of Jeffrey T. Mitman (May 2, 2023); attached Mitman report, Technical Review of U.S. Nuclear Regulatory Commissions Draft License Renewal GEIS With Respect to Section 4.9.1.2 (Environmental Consequences of Postulated Accidents) and Appendix E (Environmental Impact of Postulated Accidents) (May 2, 2023)

(hereinafter Technical GEIS Review) (NRC ADAMS Accession No. ML23123A411).

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B. Background Information Regarding NAPS

B.1.1 Seismic Design of NAPS

8. The NAPS reactors were designed and built to an array of closely related and interdependent deterministic requirements, i.e., design rules, promulgated by NRC under the Atomic Energy Act. The design rules are intended to ensure the safety of nuclear reactors - and therefore their low environmental impacts - by establishing requirements for quality in design and construction, demonstrated ability to withstand likely external forces such as earthquakes, and floods. 3

9. Probabilistic considerations did not play a role in the development of the design rules, other than in the course of evaluating the likelihood of the hazards such as earthquakes that must be defended against by engineered safety features; nor do probability considerations play a role in the way the design rules are applied in licensing proceedings. Satisfaction of these rigorous and deterministic safety requirements is required, regardless of the probability that compliance or non-compliance will affect the safe operation of a particular reactor.

10. The NRC issued operating licenses for North Anna Units 1 and 2 in 1978 and 1980, respectively. As required by GDC 2, VEPCO based the reactors seismic safety design on the most severe earthquake historically reported for the site and the area surrounding the site, an earthquake that had occurred in 1875. 4 Based on this historic earthquake, VEPCO established a design-basis earthquake with ground motions of 0.12g horizontal and 0.08g vertical for structures founded on rock. 5 VEPCO also identified a set of Category I safety systems and components (SSCs) that must remain functional in the event of a design-basis earthquake. 6 The Updated Final Safety Analysis Report (UFSAR) for North Anna Units 1 and 2 represents that Class I piping systems are qualified to withstand a total of five operational-basis earthquake (OBE) (one-half safe-shutdown earthquake) and one design-basis earthquake (DBE).). 7 Thus, the NRCs reasonable assurance/no undue risk finding for North Anna Unit 1 was based on the assumption that an earthquake more severe than the 1875 earthquake would not occur, and that an earthquake with the severity of the 1875 earthquake would occur only once.

B.1.2 Seismic Design of Proposed North Anna Unit 3

11. In 2007, VEPCO applied to the NRC for a combined license (COL) to build and operate an Economic Simplified Boiling Water Reactor (ESBWR) at the North Anna site. 8 The new reactor was later canceled.

B.1.3 Mineral Earthquake

12. On August 23, 2011, while North Anna Units 1 and 2 were operating at full power, a Magnitude 5.8 earthquake occurred in Mineral Virginia, about ten miles from the reactors.9 After the earthquake, the North Anna site lost offsite power, both reactors automatically tripped, and four emergency diesel generators were activated. 10 As

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described by the NRC, the Mineral Earthquake exceeded the reactors design basis in two respects: spectral and peak ground accelerations for the Operating Basis (OBE) and Design Basis Earthquakes (DBE). 11

B.1.4 VEPCO and NRC Response to Mineral Earthquake for NAPS

13. On August 26, the licensee declared all safety-related SSCs of Units 1 and 2 inoperable and issued a 10 CFR 72 Notification 12

14. Following the 2011 beyond design basis earthquake, VEPCO conducted a restart evaluation by inspecting piping, concrete structures and other SSCs affected by the earthquake and found they were operable. 13 VEPCO also added some measures to

3 See, e.g., 10 C.F.R. Part 50, Appendix A, General Design Criteria (GDC) 1, 2, 4, and 5. See also 10 C.F.R. Part 50, Appendix B, containing the NRCs quality assurance standards.

4 Updated Final Safety Analysis Report at 2.5-4 (Rev. 56, Sept. 30, 2020) (ML20309A602).

5 Id. at 2.5-5. The Operating Basis Earthquake (OBE) was set at 0.06g horizontal and 0.04g vertical (50% of the design-basis earthquake values). Id.

6 Category I SSCs encompass a broad array of equipment and structures, including the reactor vessel, the reactor coolant system pressure boundary, steam generators, the pressurizer, piping, pump casings, valve bodies, the core shroud, component supports, pressure retaining boundaries, heat exchangers, ventilation ducts, the containment, the containment liner, electrical and mechanical penetrations, equipment hatches, and seismic Category I structures. North Anna Power Station Units 1 and 2, Application for Subsequent License Renewal at 2-29 (August 2020)

(ML20246G696) (SLR Safety Application).

7 Id. at 3.7-35.

8 Letter from David A. Christian, VEPCO, to NRC (Nov 26, 2007) (ML073320913).

9 Technical Evaluation by the Office of Nuclear Reactor Regulation Related to Plant Restart after the Occurrence of an Earthquake Exceeding the Level of the Operating Basis and Design Basis Earthquakes, Virginia Electric and Power Company, North Anna Power Station, Unit Nos. 1 and 2 Renewed Facility Operating License Nos. NPF-4 and No. NPF-7, Docket Nos. 50-338 and 50-339, Executive Summary at 1 (November 11, 2011) (ML11308B406) (NRR Technical Evaluation).

10 Id.

11 Id.

12 Presentation to the Commissioners North Anna Nuclear Power Plant Seismic Event, at Slide 3 (August 30 and September 1, 2011) (ML112420551) 13 NRR Technical Evaluation at 31.

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improve its earthquake response, such as upgrades to its seismometers. 14 Based on VEPCOs evaluation and responsive measures, the NRC Staff approved restart. 15

15. At the same time that VEPCO and the NRC were responding to the 2011 earthquake, they were also responding to a catastrophic 2011 accident at the Fukushima Daiichi nuclear power plant complex in Japan, initiated by an earthquake and a tsunami. The Fukushima Daiichi accident led NRC to require further investigations and modifications of all reactors, including North Anna. 16

16. Because the Fukushima Daiichi accident was considered beyond the design basis for U.S. reactors, the NRC did not require changes to the rigorously qualified design bases to which the reactors had been built; nor did they require upgrades to SSCs or installation of SSCs. Instead, the NRC and the nuclear industry developed plans to install FLEX equipment for the purpose of providing diverse and flexible mitigation capacity. 17 As the Nuclear Energy Institute (NEI) explained, the FLEX equipment did not have to meet the NRCs strict standards for SSCs:

Plant equipment relied upon to support FLEX implementation does not need to be qualified to all extreme environments that may be posed, but some basis should be provided for the capability of the equipment to continue to function. 18

17. Thus, while FLEX equipment was intended to play a safety role, it need not satisfy the quality standards, environmental qualification, and other measures mandated to assure the reliability of SSCs.

18. The NRC further explained the distinction between SSCs and FLEX equipment in its 2019 rule imposing post-Fukushima mitigation requirements (including FLEX strategies) on reactor licensees to protect against beyond design basis accidents:

The NRC also clarifies the confusion that appears to stem from the application of the reasonable protection standard to safety-related SSCs that have both design-basis and beyond-design-basis functions. Safety-related SSCs that function initially in response to beyond-design-basis external events have two sets of functions: safety-related functions and beyond-design-basis functions. The NRC imposes extensive, special treatment requirements on these SSCs for their safety-related functions for design-basis events. This framework produces an increased

14 Id. at 15.

15 Id., Executive Summary at 2.

16 See, e.g., Order EA-12-049 to All Power Reactor Licensees et al., re: Issuance of Order to Modify Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events (Mar. 12, 2012) (ML12054A735).

17 See NEI-12-06, Diverse and Flexible Coping Strategies (FLEX) Implementation Guide, Rev. 5 (April 2018) (ML18120A300).

18 Id. at 16.

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level of assurance that the SSCs will perform those safety-related functions during and/or following the design-basis events as applicable. (See Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors; Final Rule (69 FR 68008; November 22, 2004)).

Through this final rule, the NRC places fewer regulatory requirements associated with the beyond-design-basis functions that dual-function SSCs perform to maintain or restore core cooling, containment, and SFP [spent fuel pool] cooling capabilities, as compared to their safety-related, design-basis functions. The reasonable protection standard is a means for enabling greater flexibility for addressing external hazards, and in the process, enabling a beyond-design-basis regulatory framework that establishes an appropriate level of assurance. The fundamental applicability of the reasonable protection requirement is to equipment that is relied on for the mitigation strategies for beyond-design-basis events without regard to whether the equipment is FLEX equipment as defined in NEI 12-06 or plant equipment as that term is used in NEI 12-06.

Accordingly, the set of requirements that are applicable, and by direct extension, the resulting level of regulatory assurance required is directly linked to whether the SSC or equipment is performing a design-basis function or a beyond-design-basis function. The level of assurance is established by the function performed by the SSC, not by the equipment or SSC alone. 19

19. In responding to the NRCs post-Fukushima orders, VEPCO noted that the 2011 Mineral earthquake was somewhat unique, and undertook a Senior Safety Hazard Analysis Committee (SSHAC) Level 2 evaluation that:

[F]ocused on the implications of the Mineral, VA earthquake, an update to the earthquake catalog that is the basis for the estimate of earthquake recurrence rates, and evaluation of new information available in the literature to determine if there was a basis for making revisions to the SSC model or the addition of new, local seismic sources that would contribute to the ground motion hazard at North Anna. 20

20. But VEPCO found no basis to revise or amend the SSC model for the North Anna PSHA [probabilistic seismic hazard analysis]. 21 Instead, VEPCO provided non-safety-grade FLEX equipment such as a portable diesel generator and a portable RCS pump. 22

19 Final Rule, Mitigation of Beyond-Design-Basis Events, 84 Fed. Reg. 39,684, 39,690 (Aug. 9, 2019) (emphasis added) (Accident Mitigation Rule).

20 Letter from Daniel G. Stoddard to NRC re: Virginia Electric and Power Company North Anna Power Station Units 1 and 2, Response to March 12, 2012 Information Request: Seismic Probabilistic Risk Assessment for Recommendation 2.1 at 84 (Mar. 28, 2018) (ML18093A445)

(Dominion Letter re SSHAC Review).

21 Id.

22 Id. at 69.

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21. The NRC Staff concluded that the SPRA for NAPS was adequate to show that no further response or regulatory actions ( e.g., modifications to the plants seismic design basis) were needed for adequate protection or compliance with existing requirements. 23

B.1.5 VEPCO and NRC Response to Mineral Earthquake for Proposed Unit 3

22. The Mineral Earthquake also affected the site of North Anna Unit 3. After the earthquake, Dominion reviewed the adequacy of its COL application and proposed to make safety-related design changes such as the arrangement of steel reinforcements and shear ties and increasing the size of certain welds, anchor bolts, and a steel girder. 24 The NRC approved those design changes. 25 But VEPCO had to obtain an exemption from the ESBWR certified design in order to make the seismic upgrades. 26 The seismic upgrades were considered necessary to ensure Unit 3s compliance with General Design Criterion 2, which requires that reactors must be adequately designed against likely earthquake hazards. 27

C. Evaluation of North Anna Draft SEIS

23. In 2024, on behalf of Beyond Nuclear and the Sierra Club, I evaluated the North Anna Draft SEIS with particular attention to Section 3.11.6.9 and Appendix F. I also became generally familiar with the entire Draft SEIS and with the applicants Environmental Report. My evaluation is documented in the attached expert report, entitled: Technical Review of U.S. Nuclear Regulatory Commissions Draft Site-Specific Environmental Impact Statement for Subsequent License Renewal of North Anna Power Station Units 1 and 2 With Respect to Accident Analysis (Feb. 22, 2014) (hereinafter Technical NAPS Review).

23 Letter from L. Lund, NRC, to D. Stoddard, VEPCO, re North Anna Power Station, Units 1 and 2 - Staff Review of Seismic Probabilistic Risk Assessment Associated with Reevaluated Seismic Hazard Implementation of the Near-Term Task Force Recommendation 2.1: Seismic (EPID No. L-2018-JLD-0003), Enc. 1 at 34 (Apr. 25, 2019) (ML19052A522) 24 Dominion Virginia Power (North Anna Power Station, Unit 3) CLI-17-08, 85 N.R.C. 157, 178 (2017) 25 Id.

26 SECY-17-0009, Memorandum to the Commissioners from Victor M. McCree re: Staffs Statement in Support of the Uncontested Hearing for Issuance of a Combined License for North Anna Power Station Unit 3 (Jan. 18, 2017) (ML15341A103). SECY-17-0009 was submitted as Exhibit NRC-001 in the mandatory hearing on Dominions COL application. In addition, the Staff submitted Exhibit NRC-007, the Final Safety Evaluation Report for Dominions COL application (SER) (March 16, 2017) (ML17075A474). See North Anna, CLI-17-08, 85 N.R.C.

at 176-177, notes 113 and 116.

27 SECY-17-0009 at 24.

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24. I hereby adopt and incorporate by reference my Technical NAPS Review. As set forth there, in my expert opinion, the North Anna Draft SEIS conclusion that the environmental impacts of reactor accidents at NAPS during an SLR term are SMALL lacks adequate supporting factual data and/or qualitative information and does not reflect the use of sufficiently rigorous or up-to-date analytical methods with respect to three important subject areas: the environmental significance of the 2011 Mineral Earthquake, new and significant information regarding accident risks, and the effects of climate change on accident risk and their environmental impacts. Below, in Subsections C.1, C.2, and C.3, I will discuss those issues in more detail.

C.1 Failure to Address Environmental Significance of 2011 Mineral Earthquake

25. The Draft SEIS claim that the environmental impacts of design-basis accidents is SMALL is based on the incorrect assumption that NAPS is operating and will continue to operate in compliance with its design basis. On page 3-169, the Draft SEIS states:

The UFSAR design-basis accident analysis forms the technical bases for the North Anna Technical specifications for operation. In Section F.1.1 (Page F-2 at 41), the Draft SEIS also states: [T]he licensee is required to maintain the acceptable design and performance criteria (which includes withstanding design-basis accidents) throughout the operating life of the nuclear power plant, including any license-renewal periods of extended operation. And at page F-3, the Draft SEIS state: Because the requirements of the existing design basis and any necessary aging management programs will be in effect for SLR, the environmental impacts of design-basis accidents as calculated for the original operating license application should not differ significantly from the environmental impacts of design-basis accidents during other periods of plant operations, including during the initial license renewal and SLR periods.

26. In making these assertions, the NRC disregards the occurrence of a beyond design-basis earthquake that caused NAPS to exceed its licensed Design Basis Analysis. As a result, NAPS has not been maintained with its design basis or its current licensing basis. Thus, the Draft SEIS is factually incorrect. The error is significant for two reasons.

27. First, the NRC fails to acknowledge it or explain the fundamental difference between a finding of no significant or small impact that is based on a deterministic analysis and a finding of no significant impact that is based on a probabilistic analysis. In my expert opinion, the deterministic analysis is more conservative because it requires a robust design that provides reasonable assurance that an external event like an earthquake will not harm necessary safety systems. A probabilistic analysis, in comparison, does not assume safety related equipment will perform as designed and then calculated the likelihood of an accident occurring. The NRC should explain the difference and how its assessment of risk has changed as a result of the Mineral Earthquake. It should also explain what it has done to evaluate the potential that safety systems, which are assumed to survive a beyond-design-basis earthquake only once (see par. 10 above), will be able to perform their safety functions when the next earthquake occurs.

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28. Second, the Draft SEIS does not address, let alone reconcile, the significant disparity between the results of the seismic risk analyses for Unit 3 and Units 1 and 2. In both cases, the NRC and VEPCO were responding to the very same earthquake. Yet, while the NRC required seismic upgrades for Unit 3, it required no seismic upgrades for Units 1 and 2 which required only a set of nonpedigree commercial-grade FLEX components with significantly lower reliability (i.e., higher failure probably). The NRC should explain the reason for this disparate result. If the NRC considered significant safety grade improvements necessary for adequate protection of Unit 3, the obvious conclusion is that it thought the safety and environmental impacts of an earthquake were significant. Why did it make a different finding for Units 1 and 2?

29. This issue is also discussed at pages 1-2 and 3-4 of my Technical NAPS Review.

30. In summary, due to the deficiencies discussed above, I do not believe the NRC has a reasonable basis for concluding that the environmental impacts of re-licensing NAPS are SMALL or insignificant with respect to earthquake risks.

C.2 Inadequate Support for Finding That Severe Accident Impacts Are Small

31. As discussed at pages 2 - 7 of my Technical NAPS Review, the Draft SEIS discussion of accident risks is seriously deficient in multiple respects. First, the Draft SEIS is inadequate as a general matter for making broad generalizations about external event CDF based on extrapolations from internal event CDF values and limited actual plant-specific values for external event CDF. Appendix F looks explicitly at external events focusing exclusively on seismic issues. It ignores other external events such as flooding, external fires (e.g., forest and wildfires), tornadoes, etc. 28 These additional external events, especially flooding, may not be ignored without serious underestimating environmental impacts. Further, as discussed below in Section C.3, increased frequency and severity of weather-induced external events is now a certainty, and therefore must be considered.

32. In finding that the environmental impacts of severe accidents are SMALL, the NRC ignores its own data regarding seismic and fire core damage frequency (CDF) that indicate these impacts are significant. 29 The NRC also disregards the fact that the occurrence of the 2011 Mineral Earthquake, by itself, increased the risk of an earthquake severe enough to damage safety equipment. This increased risk should be quantified and added to the Draft SEIS risk estimate. For these reasons, the Draft SEIS assertion at page F-26 that there has been a substantial decrease in internal event CDF is erroneous and should be corrected both factually and with respect to the Draft SEIS finding that accidents are small.

28 Draft SEIS at Page F-10 Lines 24 to 37 for the absence of a discussion of flooding, fires, high winds, etc.

29 See Technical NAPS Review at 5.

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33. Appendix F looks explicitly at external events focusing exclusively on seismic issues. It ignores other external events such as flooding, external fires (e.g., forest and wildfires),

high wind events (e.g., tornadoes), etc. To have a complete and comprehensive understanding to the risks the analysis should look at all hazards. The following paragraphs 34 - 37 provide an illustration of why the Draft SEIS omission has such a significant effect on the quality and reliability of the NRCs environmental impact evaluation.

34. A review of the licensees Flood Hazard Reevaluation Report (FHRR) shows an increase in the calculated Local Intense Precipitation [LIP] Protected Area, Local Intense Precipitation West Basin Area, and Flooding in Streams and Rivers. 30 These values are evaluated and confirmed in the NRCs corresponding Staff Assessment and reevaluated values are reproduced here. 31

30 Flooding Hazard Reevaluation Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Flooding, Attachment 1 March 2013, (ML13318A106). See Table 3.0-1 Current and Reevaluation Flood Elevations, Page 3.0-4.

31 Staff Assessment Related to Flooding Hazard Reevaluation Report Near-Term Task Force Recommendation 2.1 Related to the Fukushima Dai-ichi Accident North Anna, Units 1 and 2, Sept. 25, 2015 (ML15238A844) at Page 41.

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35. Table 4.0-2 Footnote 3 shows that a LIP event causes water to flow into the Turbine Building. The total storage volume available in the West Basin area and in the Turbine Building basement below the crest of the flood protection wall is 274,131 ft 3. Thus, in both cases, the maximum flood levels during a LIP storm event causes water to flow into the Turbine Building and flow over the top of the flood protection wall. 32 This will overtop the Emergency Switchgear Room (ESGR) flood protection wall incapacitating the ESGR and the Emergency Core Cooling Systems (ECCS) and containment cooling systems.

36. This is a significant and new finding not addressed in either the ER nor the NRCs DEIS.

Instead of using this information to inform the associated external event CDF values the NRC relies on using average external event flooding information carried over from the 2013 GEIS.

32 Flooding Hazard Reevaluation Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Flooding, March 2013 (ML13318A106), at Page 2.1-6.

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37. In addition, the NRC should correct other estimates that currently depend on a CDF estimate that is too low, for instance the estimate of population dose risk. 33

38. In other respects, the Draft SEIS is incomplete. For example, the Draft SEIS fails to demonstrate consideration of internal flooding risks. 34

39. The Draft SEIS is grossly misleading when it states: [T]he NRC completed its review of Fukushima-related information relevant to North Anna and concluded that no further regulatory actions were needed to ensure adequate protection or compliance with regulatory requirements, thereby reconfirming the acceptability of North Annas design basis. 35 The DEIS references an NRC letter to Dominion. 36 This letter makes no reference to adequate protection or a finding of adequate protection. In addition, it makes no reference to confirming the design basis.

40. The post-Fukushima work looked at only two hazards, external flooding and earthquakes.

If the NRC believes that the design basis or a portion of the design basis was reconfirmed, they should supply a citation to that reconfirmation. Absent such documentation this erroneous conclusion should be removed.

41. The Draft SEIS is similarly misleading with respect to its discussion of risk improvements obtained by NRC and license efforts after September 2001. It is reasonable to assume that the PRA for NAPS that was prepared in 2020 captured these measures, and therefore it is inappropriate to double-count them. 37

42. The Draft SEIS also takes inappropriate credit for reductions in environmental risk. For example, the Draft SEIS states that security is beyond the scope of license renewal, 38 and yet takes credit for these security enhancements. 39 And the Draft SEIS takes credit for supposedly reducing risk estimates in VEPCOs PRA by taking new factors into consideration quantitatively, such as reinforcement of defense capabilities, better control of sensitive information, enhancements in emergency preparedness, and implementation of mitigation strategies to deal with postulated events potentially causing loss of large areas of the plant due to explosions or fires, including aircraft crashes. 40 But none of

33 Technical NAPS Review at 4.

34 Technical NAPS Review at 2

35 Draft SEIS at F-3 at Lines 10 - 13 (emphasis added).

36 NRC Letter from R.J. Bernardo to D.G. Stoddard, June 9, 2020, regarding North Anna Power Station, Units 1 And 2Documentation of the completion of required actions taken in response to the lessons learned from the Fukushima Dai-Ichi accident, (ML20139A077).

37 Technical NAPS Review at 6.

38 Draft SEIS at F-20.

39 Draft SEIS Section F.4.1 at F-19

40 Draft SEIS at F F-20

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these elements were actually incorporated in the base PRA, and thus credit may not be taken for changing the quantitative values. At the most, the Draft SEIS should have asserted that the base PRA underestimated the risk. 41

43. The Draft SEIS fails to demonstrate consideration of uncertainties with respect to the conclusion that severe accident impacts are small. 42 The NRCs PRA Policy Statement states The Commission's safety goals for nuclear power plants and subsidiary numerical objectives are to be used with appropriate consideration of uncertainties in making regulatory judgments on the need of proposing and backfitting new generic requirements on nuclear power plant licensees (emphasis added). 43 Consideration of uncertainties is integral component of PRA 44 and an NRC required component. 45 The NRCs NUREG-1855 Rev. 1 supplies extensive guidance on how to perform uncertainty analysis and how to use uncertainty analysis in risk-informed decision making. 46 This is because the uncertainties show the degree to which the NRC has confidence in its predictions. 47

44. With respect to PRAs, the NRC expects that appropriate consideration of uncertainties will be given in the analyses used to support the decision and the interpretation of the findings of those analyses. 48 Because PRAs are integral to reactor risk analyses in EISs,49 the requirement for uncertainty analysis is equally important to an environmental analysis as to a safety analysis. An adequate probabilistic risk analysis would include parametric uncertainty data on all input parameters and calculate the corresponding CDF and Large Early Release Frequencies (LERF) with uncertainty bounds (e.g. a CDF or LERF of 1E-5 per year with a 90% confidence band of 1E-6 to 5E-5 per year). The

41 Technical NAPS Review at 6.

42 Id. at 2-3, 4.

43 Final Policy Statement, Use of PRA Methods in Nuclear Regulatory Activities, 60 Fed. Reg.

42,622 (Aug. 16, 1995),

44 American Society of Mechanical Engineers /American Nuclear Society Standard ASME/ANS RA-Sa-2009, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, Addendum A to RA-S-2008, ASME, Feb.

2009.

45 NRC Regulatory Guide 1.200 Rev. 3 Dec. 2020 (ML120238B871). See Table 2 Summary of Technical Characteristics and Attributes of a Level 1, Internal Events PRA for the At-Power Operating Mode at Page 17 and Section C.1.2.11 Technical Elements for the Interpretation of Results (Including Uncertainty Analysis) at Page 34.

46 NUREG-1855, Rev. 1, Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making, March 2017 (ML17062A466).

47 Id. at 1.

48 Id. at iii.

49 See Standard Review Plans for Environmental Reviews for Nuclear Power Plants, Supp. 1, Operating License Renewal at 5-3, 5-5, 5-7 (NUREG-1555, Supp. 1, Draft for Comment, Feb.

2023).

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analysis would then propagate those CDF and LERF values with their uncertainty bands through the Level 2 and Level 3 PRA evaluations ending with estimates of both prompt and latent cancer fatalities with uncertainty bands. Finaly, the analysis would compare that calculated values with their corresponding uncertainties against the decision thresholds, i.e., safety goals. But this DEIS is deficient and unacceptable as it never calculates probabilistic uncertainties. Therefore, it cannot use them in comparison to a goal or set of criteria for its decisions - i.e., whether environmental impacts are small, medium or large.

45. The Draft SEIS does not address the environmental impacts of concurrent multi-unit accidents. This is a significant omission, given the well-recognized independent contribution that multi-unit accidents make to accident risk. As discussed in a 2013 paper by NRC Staff member Suzanne Schroer and University of Maryland Professor Dr.

Mohammed Modarres, multi-unit site risk is neither formally nor adequately considered in either the regulatory or the commercial nuclear environment [citations omitted] despite the fact that the questions of multi-unit accident is not one of possibility, but of probability. 50 This issue is discussed in detail in my Technical NAPS Report at pages 9

- 10.

46. The SAMA analysis is deficient in multiple respects, including failure to consider SAMAs that meet criteria for consideration, and failure to provide documentation of an NRC audit relied on to conclude that VEPCOs approach to its SAMA analysis was methodical and reasonable. 51

47. In summary, in my professional opinion, the Draft SEIS does not reflect a complete or adequately rigorous evaluation of accident risks, because essential data are missing and important analytical assertions are erroneous or misleading. Therefore, the NRC lacks an adequate basis for concluding that the environmental impacts of accidents during a license renewal term are SMALL.

50 Schroer and Modarres, An Event Classification Schema for Evaluating Site Risk in a Multi-Unit Nuclear Power Plant Probabilistic Risk Assessment, p. 1 (2013) (ML13217A335).

51 Technical NAPS Review at 7.

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C.3 Inadequate Discussion Effects of Climate Change on Accident Risk

48. Various sections of the Draft SEIS address climate change. However, the Draft SEIS does not address climate changes impacts on accident risks in Section 3.11.6.9 or Appendix F. This omission constitutes a significant deficiency in the Draft EIS because climate change demonstrably affects the frequency and intensity of some external events and therefore has the potential to significantly increase accident risks. Moreover, the frequency and intensity of climate change effects are increasing over time. Given that the NRC is proposing to rely on the Draft SEIS for decisions that could affect reactor safety decades from now, the Draft SEIS must address these changing effects over the entire licensed lifetime of reactors, which may end 4 decades from now.

49. Climate change has already started to increase the frequency and intensity of these events. 52

50. As discussed above in Section C.2, the Draft SEIS is already inadequate as a general matter for making broad generalizations about external event CDF based on extrapolations from internal event CDF values and limited actual plant-specific values for external event CDF.

51. Climate change will increase the frequency and severity of these types of events, including floodings events. The Turbine Building flooding resulting from a LIP event discussed in Paragraphs 34 - 37 above will be significantly exacerbated in both frequency and severity by climate change. This is another example of the inadequacies of the DEIS.

52. The NRC is well-aware of the issues of climate change and its impact on nuclear plant safety. After the Fukushima meltdowns, the NRC Office of Research initiated a research program to develop tools to assist in probabilistic and deterministic assessments of external hazards including seismic, high winds and flooding with a consideration of climate change. 53 In addition, climate change has been a topic of discussion at the NRCs Regulatory Information Conference (RIC) in recent years. 54

52 See, for example, Climate change is probably increasing the intensity of tropical cyclones, March 31, 2021 NOAA, https://www.climate.gov/news-features/understanding-climate/climate-change-probably-increasing-intensity-tropical-cyclones ; Climate Change Indicators: Weather and Climate, EPA, https://www.epa.gov/climate-indicators/weather-climate ; Global Warming and Hurricanes, NOAA Geophysical Fluid Dynamics Laboratory, April 11, 2023, https://www.gfdl.noaa.gov/global-warming-and-hurricanes/.

53. See NRC Probabilistic Flood Hazard Assessment Research Program Overview,, February 22 - 25, 2021 (ML21064A418) and Potential Impacts of Accelerated Climate Change, PNNL-24868, May 2016 (ML16208A282)).

54 See Climate Change Impact on the Safety of Nuclear Installations, March 8-10, 2022 (ML22140A312)) & Observations on Extreme Weather and Impacts on Nuclear Power Plants, EPRI ML22140A320, 2022).

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53. The effects of climate change on accident risk are and will continue to be site-specific and not subject to generalization. For example, the three reactors at the Oconee plant --

for which the NRC is now considering an application for subsequent license renewal --

lie downstream of two large dams. The design of the dams includes consideration of the maximum probable flood induced by the maximum probable precipitation (i.e., storm).

Climate change has the potential to significantly increase the amount of precipitation falling on watersheds above the dams. Will the dams be able to pass these higher intensity storms and the resulting floods? In order to conduct a well-informed licensing review, the NRC will need to address this question.

54. Another example is the Turkey Point plant, located in a low-lying coastal area of South Florida. With climate change the already-occurring, sea level change will continue and possibly accelerate during the SLR period. Likewise, hurricane intensity, i.e., wind speed, rain fall and storm surge, will intensify.

55. As discussed below, the Duane Arnold plant in Iowa was prematurely and permanently shuttered after being hit with a Derecho with wind speeds exceeding 100mph. Climate change has been implicated in the severity of this extreme weather event (Hints of a derecho-climate change link, ten years after 2012 storm, Washington Post, June 29, 2022, https://www.washingtonpost.com/climate-environment/2022/06/29/derecho-climate-change-severe-storm/ )

56. Therefore, in order to provide a reasonably thorough and complete analysis of accident risks during the license renewal term, the Draft SEIS must address the continuing and growing contribution of climate change to accident risks at nuclear plants. And this evaluation must be conducted on a site-specific basis.

57. Climate change affects risk in two ways. First, it increases the likelihood or initiating event frequency of events. For example, increased storm frequency can lead to higher initiating event frequency for losses of offsite power (LOOPs). Second, climate change can increase the probability of failure of design features or mitigation equipment. A 2020 severe windstorm at the Duane Arnold plant (ML21139A091) illustrates this phenomenon. While the storm may or may not be directly attributable to climate change, it is a reasonable example of the type of severe weather effects that climate change can cause today and will cause in the future. In that case, a severe windstorm caused a loss of offsite power (LOOP). As a result of the LOOP, debris accumulated at the suction of the service water systems, which are necessary to cool the emergency diesel generators (EDGs) and the emergency core cooling system (ECCS) heat exchangers. The NRCs risk analysis of the event showed an increase in the failure probabilities of the service water system, the EDGs and the ECCS due to this climate-related external event.

Consideration of these risks in an EIS would provide important information regarding climate-related accident risk as well as identification of mitigation measures to address those risks.

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58. A third way that climate change affects risk analysis, which is unique to flooding risk, is the cliff edge effect. With most hazards if the severity is increased slightly, the stress on the system is increased somewhat proportionately. However, with many flood-related issues, a small increase in the hazard can cause a dramatic and often overwhelming impact on a structure. For example, a small increase in wave height could raise the flood height sufficiently to overtop a floodwall inundating the equipment the floodwall is designed to protect. Risk analyses for climate change-related flooding must look carefully at this cliff-edge phenomenon.

59. PMP is a significant input into the design of critical infrastructure such as dam and reactor safety analysis directly and indirectly through its impact on probable maximum flood (PMF). The National Academies under sponsorship of the National Oceanic and Atmospheric Administration (NOAA) has started a project to modernize the probable maximum precipitation (PMP) methodology. 55 The NRC is well aware of this effort as they have already participated in at least one of the initial project workshops. PMP and PMF also impact reactor safety directly via their impact on local intense precipitation (LIP). This project will consider approaches for estimating PMP in a changing climate, with the goal of recommending an updated approach, appropriate for decision-maker needs. This project is clear evidence that the Federal Government and the NRC understands the significance and severity of climate change on critical infrastructure.

Waiting for the project completion is unnecessary and inappropriate. Climate change is here, the NRC and the licensee know it, steps should be taken now to protect the plant and the public from its effects.

60. In summary, in my professional opinion, the Draft SEIS does not reflect a complete or adequately rigorous evaluation of all external hazards, does not consider uncertainties and does not address the reasonably foreseeable effects of climate change on the risks of accidents at North Anna. Given these serious deficiencies, the NRC cannot claim to have a reasonable basis for concluding that the environmental impacts of accidents during a license renewal term are SMALL.

D. Conclusion

61. In my professional opinion, the deficiencies in the Draft SEIS with respect to any one of the above three categories -- failure to address environmental significance of the 2011 Mineral Earthquake, inadequate support for finding that severe accident impacts are small

, and inadequate discussion effects of climate change on accident risk - are significant and must be addressed before the Draft SEIS can be deemed to constitute an adequately complete and rigorous environmental analysis. Taken together, all three categories of deficiencies show a level of inadequacy that is grossly unacceptable.

55 See https://www.nationalacademies.org/our-work/modernizing-probable-maximum-precipitation-estimation#sectionSponsors ).

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I declare that the foregoing statements of fact are true and correct to the best of my knowledge and that the statements of opinion are based on my best professional judgment.

s/Jeffrey T. Mitman Executed in Accord with 10 C.F.R. §2.304(d)

March 27, 2024

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CURRICULUM VITAE JEFFREY T. MITMAN Poolesville, MD February 22, 2024

QUALIFICATIONS Reliability and risk analyst with more than 40 years experience in the nuclear industry. Skills include evaluation and modeling of probabilistic risk analyses (PRA) and management of PRA projects and teams.

Highly experienced in low power and shutdown (LPSD) risk modeling issues. Solid record of bringing projects in on schedule and budget.

MAJOR ACCOMPLISHMENTS Transitioned NRC to detailed PRA models for LPSD significance determinations process evaluations.

Guided development of and managed industrys first configuration risk management software tool.

Obtained regulatory approval of EPRIs risk informed in-service inspection (RI-ISI) methodology.

Managed first PRA of bolted spent fuel storage cask.

EXPERIENCE PRIVATE CONSULTANT (Poolesville, MD)

Nuclear risk analyst 2021-Present Reviewed Oconee Subsequent License Renewal application and prepared technical report on adequacy of environmental and safety analyses to address flooding risks.

Reviewed and submitted comments on NRCs draft (2023) Generic Environmental Impact Statement (NUREG-1437 Revision 2).

US NUCLEAR REGULATORY COMMISSION (Rockville, MD) 2005 - 2021 Senior Reliability and Risk Analyst (NRC Office of Nuclear Reactor Regulation)

Conducted Significance Determination Process (SDP) evaluations of reactor events including development and/or modification of required models.

Served as lead analyst for LPSD event issues and concerns.

Guided development of shutdown Standardized Plant Analysis Risk (SPAR) models.

Conducted Human Reliability Analysis (HRA).

Evaluated external event risk from dam failures.

Served on NRCs Japan Team (part of USAID disaster assistance response team for Fukushima Daiichi accident), providing technical advice and support through the U.S. Ambassador to Japanese government.

Participated in post NRCs Fukushima Near Term Task Force (NTTF) flooding guidance development.

Developed NRCs guidance on crediting FLEX in risk-informed regulatory applications.

Advised NRC National Fire Protection Association (NFPA) 805 team on issues related to shutdown fire risk.

Performed evaluations of risk informed license applications.

Reliability and Risk Analyst (NRC Office of Nuclear Regulatory Research)

Project Manager for the development of shutdown SPAR models ERIN ENGINEERING AND RESEARCH, INC. (Walnut Creek, CA) 2004 - 2005 Lead Senior Engineer Prepared configuration risk management evaluation of at-power fire risk.

Prepared configuration risk management evaluation of loss of offsite power.

ABE STAFFING SERVICES (Palo Alto, CA) 2003 - 2005 Consultant to EPRI Brought project to closure involving Dry Cask Storage PRA project and team, involving Transnuclear bolted cask containing PWR fuel.

Jeffrey T. Mitman l Page 2

EPRI (Palo Alto, CA) 1998 - 2003 Project Manager Outage Risk Assessment and Management (ORAM-Sentinel)

- Grew first of a kind software application for performing configuration risk management in nuclear power plants.

- Conducted research in low power and shutdown risk; shutdown initiating event and event frequency derivation.

- Delivered multiple versions (including alpha, beta & production), testing and full documentation.

- Administered utility user group, marketing, contract preparation, technology transfer, technical report publication and training.

- Actively managed both development and application contracts with multiple suppliers and customers.

Managed annual $1M budget.

Dry Cask Storage PRA : Initiated innovative analysis of Transnuclear cask containing PWR fuel.

- Managed unique team with diverse experience in both cask design and PRA backgrounds.

Risk Informed In-service Inspections Project (RI-ISI): Lead team in obtaining regulatory approval of methodology to safely reduce piping weld inspection requirements using combination of probabilistic and degradation analysis.

- Responsible for methodology finalization and acceptance by industry and U.S. NRC.

- Conducted marketing, sales, contract preparation, technology transfer, training and technical report publication.

- Actively managed both development and application contracts with both suppliers and customers.

Managed annual $1M budget.

Human Reliability Analysis Project : Managed project to bring consistency to on industry use of HRA methods.

- Responsible for EPRI HRA area, including development of HRA Calculator software and establishment of associated users group.

ERIN ENGINEERING AND RESEARCH, INC. (Palo Alto, CA) 1992 - 1998 Lead Senior Engineer Collaborated with EPRI ORAM-SENTINEL Project Manager in project development and administration, user group administration, contract preparation, technology transfer workshops, technical report generation and editing. Performed ORAM analysis of the Diablo Canyon plant. Performed ORAM Probabilistic Analysis of Perry spent fuel pool. Drafted and edited ORAM V2.0 Users Manual. Assisted in ORAM-SENTINEL software design, performed software debugging. Principle researcher and author of BWR outage contingency report. Prepared marketing and training, materials.

ABB IMPELL CORPORATION (King of Prussia, PA) 1990 - 1992 Lead Senior Engineer Design Basis Documentation : directed team of three engineers to review PECO Feedwater System Design. Wrote Design Basis Documentation reports for Limerick and Peach Bottom power plants, identifying licensing and design concerns by reviewing the system design as documented in drawings, calculations, vendor manuals, Technical Specifications, UFSAR, SER, SRP, 10CFR50.59 safety evaluations etc. and by interfacing with utility engineering personnel. Prepared Engineering Change Requests as necessary.

Shift Outages : during Limerick Nuclear Power Plant refueling / maintenance outage. Coordinated all shift maintenance work and testing. Collaborated with all groups in power plant, allocating resources as needed to maintain schedule and reporting to senior plant outage management. Performed system reviews prior to placing them back in service. Conducted shift outage meetings. Tracked work group performance against schedule. Advised utility management on techniques for schedule and outage organizational improvements.

Jeffrey T. Mitman l Page 3

GENERAL ELECTRIC COMPANY (San Jose, CA) Experience Prior to 1990 Startup-Test Engineer Shift Startup Engineer: During power ascension phase coordinated all system testing on shift and startup interface with operations. During preoperational phase, acted as operations shift supervisor responsible for coordinating all system testing and flushing on shift from main control room. Updated senior utility management daily on testing status.

Additional positions: Shift Technical Advisor, Test Engineer, Lead QC / Welding Inspector

EDUCATION / PROFESSIONAL DEVELOPMENT BSE, Nuclear Engineering, University of Michigan, Ann Arbor, MI Introductory VBA class, University of California, Berkeley, CA Misc. business courses at various colleges and universities Senior Reactor Operator Certified GE Station Nuclear Engineering Effective Utilization of PSA, ERIN Engineering & Research, Walnut Creek, CA.

PROFESSIONAL ASSOCIATIONS American Nuclear Society (ANS) member since 1978.

ANS Nuclear elected member of Installation Safety Division Executive Committee 2015 to 2021.

ANS Risk Informed Standards Committee (RISC).

ANS/ASME Risk Informed Standards Writing Group on Shutdown PRA Standard.

ASME Section XI, Working Group on Implementation of Risk Based Examination.

MIT Professional Summer Programs Guest Lecturer at Risk-Informed Operational Decision Management Course.

PAPERS

1. Technical Challenges Associated with Shutdown Risk when Licensing Advanced Light Water Reactors, PSAM12 2014. Co-author.

2. Comparing Various HRA Methods to Evaluate Their Impact on the results of a Shutdown Risk Analysis during PWR Reduced Inventory, PSAM11 2012. Co-author.

3. Uncertainty Analysis for Large Dam Failure Frequencies Based on Historical Data, PSAM11 2012. Co-author.

4. An Assessment of Large Dam Failure Frequencies Based on US Historical Data, PSA 2011. Co-author.

5. Generic Failure Rate Evaluation for Jocassee Dam, US NRC (ML13039A084), 2010. Co-author.

6. Development of PRA Model for BWR Shutdown Modes 4 and 5 Integrated in SPAR Model, to be presented at PSAM10 2010. Co-author.

7. Development of Standardized Probabilistic Risk Assessment Models for Shutdown Operations Integrated in SPAR Level 1 Model, PSAM9 2008. Co-author.

8. PRA of Bolted Dry Spent Fuel Storage Cask, Presented at ICONE12. 2004. Co-author.

9. Low Power and Shutdown Risk Assessment Benchmarking, Presented at PSA 02 2002. Co-author.

10. EPRI Human Reliability Analysis Guidelines, Presented at PSA 02 2002. Co-author.

11. Derivation of Shutdown Initiating Event Frequencies, Presented at PSAM5 2000. Co-author.

12. Quantitative Assessment of a Risk Informed Inspection Strategy for BWR Weld Overlays, Presented at ICONE 8 2000. Co-author.

13. EPRI RI-ISI Methodology and the Risk Impacts of Implementation, Presented at SMiRT 11 1999. Co-author.

14. Application of Markov Models and Service Data to Evaluate the Influence of Inspection on Pipe Rupture Frequencies published. PVP 1999. Co-author.

15. Progress in Risk Evaluation of Outages, International Conference on the Commercial and Operational Benefits of PSA. 1997. Co-author.

16. Control of Reactor Vessel Temperature/Pressure during Shutdown, GE SIL 357. June 1981. Co-author.

Jeffrey T. Mitman l Page 4

SOFTWARE

1. HRA Calculator Version 2.0, EPRI 2003. 1003330. Project Manager (PM).

2. ORAM-Sentinel Version 3.4, EPRI 2001. 1002958. PM and co-author.

REPORTS / STANDARDS

1. Requirements for Low Power and Shutdown PRA - ANS/ASME-58.22-2014 (Trial Use Standard).

2. Probabilistic Risk Assessment (PRA) of Bolted Storage Casks: Quantification and Analysis Report, EPRI 2003. 1002877. PM.

3. Low Power and Shutdown Risk Assessment Benchmarking Study, EPRI, Palo Alto, CA and U.S. DOE.

2002. 1003465. PM and principal investigator.

4. Dry Cask Storage PRA Scoping Study, EPRI 2002. 1003011. PM.

5. Guidance for Incorporating Organizational Factors into Nuclear Power Plant Risk Assessments: Phase 1 Workshop. EPRI and U.S. DOE 2002. 1003322. PM.

6. An Analysis of Loss of Decay Heat Removal Trends and Initiating event Frequencies (1989-2000):

EPRI 2001. 1003113. PM.

7. Piping System Failure Rates and Rupture Frequencies for Use in Risk Informed In-service Inspection Applications: TR-111880-NP, EPRI 2000. 1001044. PM.

8. Application of Risk-Informed Inservice Inspection Alternative Element Selection Criteria. EPRI, Charlotte NC: 2000. TE-11482. PM.

9. Revised Risk-Informed Inservice Inspection Evaluation Procedure, EPRI 1999. TR-112657 Revision B-A. PM & co-author.

10. Piping System Failure Rates and Rupture Frequencies for Use in Risk Informed In-service Inspection Applications, EPRI 1999. TR-111880. PM.

11. Comparison between EDF and EPRI of Pipe Inspection Optimization Methods, EPRI Palo Alto, CA; Electricite de France, Paris, France: 1999. TR-113315. PM.

12. Economic Feasibility Study of Implementing RBISI at 2-loop PWR, EPRI 1998. TR-107613. PM.

13. Evaluation of Pipe Failure Potential via Degradation Mechanism Assessment, EPRI Palo Alto, CA:

1998. TR-110157. PM.

14. Piping Failures in U.S. Nuclear Power Plants: 1961-1997, EPRI 1998.TR-110102. PM.

15. Piping System Reliability Models and Database for used in Risk Informed Inservice Inspection Applications, EPRI 1998. TR-110161. PM.

16. Use of Risk Informed Inspection Methodology for BWR Class 1 Piping, EPRI 1998. TR-110701. PM.

17. ORAM v4.0 Functional Specification Outline, EPRI 1999. TR-111652. PM.

18. Survey on the Use of Configuration Risk and Safety Management Tools at NPPs, EPRI, 1998.

TR-102975. PM.

19. ORAM-SENTINEL Demonstration at Diablo Canyon, EPRI 1998. TR-110739. PM.

20. ORAM-SENTINEL Development at Indian Point 3, EPRI 1999, TR-110716. PM.

21. ORAM-SENTINEL Development and ORAM Integration at Oconee, EPRI 1998. TR-111207. PM.

22. ORAM-SENTINEL Development at Fitzpatrick, EPRI 1998. TR-110505. PM.

23. ORAM-SENTINEL Demonstration at Sequoyah, EPRI 1998. TR-110771. PM.

24. SENTINEL Technical Basis Report for Limerick, EPRI 1998. TR-108953. PM.

25. Outage Risk Assessment and Management Implementation at Fermi 2, EPRI 1997. TR-109013. Co-author.

26. Contingency Strategies for BWRs during Potential Shutdown Operations Events, EPRI 1993.

TR-102973. Principal investigator.

27. Generic Outage Risk Management Guidelines for BWRs, EPRI 1993. TR-102971. Co-principal investigator.

Technical Review of U.S. Nuclear Regulatory Commissions Draft Site-Specific Environmental Impact Statement for Subsequent License Renewal of North Anna Power Station Units 1 and 2 With Respect to Accident Analysis

Submitted by Jeffrey T. Mitman to the U.S. Nuclear Regulatory Commission on behalf of Beyond Nuclear, Inc.

February 22, 2024

Introduction

This report presents my technical review of the accident risk analyses presented by the U.S. Nuclear Regulatory Commission (NRC) in the Draft Site-Specific Environmental Impact Statement for License Renewal of Nuclear Plants, Supplement 7a, Second Renewal Regarding Subsequent License Renewal for North Anna Power Station Units 1 and 2 (Dec. 2023) (ML2339A047) (Draft SEIS). My review focuses on Section 3.11.6.9 (Postulated Accidents), and Appendix F (Environmental Impact of Postulated Accidents).

I am submitting this technical review on behalf of Beyond Nuclear, Inc. (Beyond Nuclear) and the Sierra Club, who have participated in multiple environmental and safety proceedings regarding Virginia Electric Power Companys/Dominions (VEPCOs or Dominions) 2020 application for subsequent license renewal (SLR) for the North Anna reactors (NAPS). In 2023, I also prepared a technical review for Beyond Nuclear and the Sierra Club regarding the Draft License Renewal GEIS. Beyond Nuclear and the Sierra Club submitted my expert declaration and technical review in support of their comments on the Draft GEIS, Comments by Beyond Nuclear and the Sierra Club on Proposed Rule and Draft Generic Environmental Impact Statement for Renewing Nuclear Power Plant Licenses (May 2, 2023, corrected May 19, 2023)

(ML23123A411).

As set forth in the expert declaration submitted with Beyond Nuclears and the Sierra Clubs comments, I am a nuclear engineer with a significant level of expertise in risk analysis. I have more than 40 years of experience in the nuclear industry and 16 years of experience with the NRC. While at the NRC, I served as Senior Reliability and Risk Analyst, with significant responsibility for managing a number of risk analysis projects and teams. A copy of my curriculum vitae is attached to my comments here.

Comments on Section 3.11.6.9 Postulated Accidents

According to the Draft SEIS at Page 3-169 (lines 35-36): For design-basis accidents, site-specific analysis of design-bass accidents is in the North Anna Updated Final Safety Analysis Report (UFSAR). Because the UFSAR is part of the current licensing basis and also the subject of the NRC oversight program for operation during PEO [period of extended operation], the impacts of design-basis accidents are SMALL. But this claim is not consistent with the history of operation of NAPS, because the reactors experienced a beyond-design-basis event in 2011, the Mineral earthquake. While the Mineral Virginia

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earthquake is discussed in the Draft SEIS, the NRC does not describe whether or how it brought the plant back into compliance with the design basis. My review of NRC-VEPCO correspondence indicates that nothing has been done in this regard, i.e., the earthquake has led to no changes in the NAPS design-basis accident (DBA) analysis, the design basis (DB) or its current licensing basis (CLB). Thus, the Draft SEIS lacks support for its claim that environmental impacts of operating NAPS are SMALL because the reactors are operating in compliance with their design basis. I would also note that the Mineral Earthquake is not discussed in any accident risk analysis in the Draft SEIS, either here or in Appendix F.

Comments on Appendix F Environmental Impacts of Postulated Accidents

Section F.1.1 Design-Basis Accidents (Page F-2 line 41) states: [T]he licensee is required to maintain the acceptable design and performance criteria (which includes withstanding design-basis accidents) throughout the operating life of the nuclear power plant, including any license-renewal periods of extended operation. On August 23, 2011 the NAPS experience the Mineral Virginia earthquake. This earthquake caused NAPS to exceed its licensed DBA. Thus, NAPS has not been maintained with its DB or its CLB.

At Page F-3 (lines 10 - 13) the Draft SEIS states: [T]he NRC completed its review of Fukushima-related information relevant to North Anna and concluded that no further regulatory actions were needed to ensure adequate protection or compliance with regulatory requirements, thereby reconfirming the acceptability of North Annas design basis. This assertion is incorrect. The correspondence documenting the NRCs review of Fukushima -related information relevant to North Anna did not reconfirm the acceptability of the entire NAPS design basis as claimed. At best, the correspondence confirmed the elements of the design of NAPS regarding seismic and flooding hazards. Nothing was said about the acceptability of North Annas design basis.

At Page F-3 (lines 42 - 46) the Draft SEIS states: Because the requirements of the existing design basis and any necessary aging management programs will be in effect for SLR, the environmental impacts of design-basis accidents as calculated for the original operating license application should not differ significantly from the environmental impacts of design-basis accidents during other periods of plant operations, including during the initial license renewal and SLR periods. This statement is incorrect. As stated in the previous comment, NAPS experienced a beyond design basis earthquake in August of 2011.

Thus, NAPS was not maintained within its design basis. Accordingly, the conclusion of the Draft SEIS regarding the impacts of design basis accidents is unsupported.

In Section F.3.1 New Internal Events Information (Section E.3.1 of the 2013 LR GEIS), starting at Page F-9 (line 28), the Draft SEIS presents a discussion of internal events risks. This discussion makes comparisons between the 2013 GEIS and the current Dominion PRA model for internal events. Typically, a reactor risk analysis of internal events considers internal flooding along with other internal events risks. In this case, however it is unclear from the text whether the Draft SEIS includes an analysis of internal flooding. If it has been excluded, the omission is significant and should be corrected.

At Page F-10 (lines 10 - 14) the Draft SEIS states: [T]he NRC staff concludes that no new and significant information exists for North Anna during the SLR term concerning the offsite consequences of severe accidents initiated by internal events that would alter the conclusion that the probability weighted consequences of severe accidents would be SMALL reached in the 1996 LR GEIS, the 2013 LR GEIS, and the North Anna initial LR SEIS. This conclusion is stated after reiterating the corresponding 2013 GEIS

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CDF values for NAPS and the current NAPS internal events values as supplied by Dominion. However, the analysis underlying the conclusion does not comply with NRC guidance requiring that risk-informed decision-making must include consideration of uncertainties. See NUREG-1855 Rev. 1 "Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making," March 2017.

At page F-10 (line 32) of Section F.3.2 External Events (Section E.3.2 of the 2013 LR GEIS), the Draft SEIS reports that the seismic CDF value for NAPS is 6E-5 per year. (This is consistent with the NRCs 2019 letter to Dominion giving a value of 6.3E-5 per year. See letter from NRC to Dominion re: North Anna Power Station, Units 1 And 2 - Staff Review of Seismic Probabilistic Risk Assessment Associated with Reevaluated Seismic Hazard Implementation of the Near -Term Task Force Recommendation 2.1:

Seismic, April 25, 2019 (ML19052A522 )). In contrast, Table E.3-12 in the Draft License Renewal GEIS (Page E-33) shows a mean all hazards CDF (i.e., including both internal events and external events) for PWRs of 6.1E-5 per year. Thus, the Draft SEIS value for seismic alone at NAPS is greater than the NRCs calculated average for all hazards combined in the Draft License Renewal GEIS. This discrepancy should be addressed.

In addition, Table E.3-10 of the Draft License Renewal GEIS (Page E-28) lists the mean fire CDF for all reactors as 4.5E-5 per year. For NAPS the sum of internal events (1.36E-6), seismic (6.3E-5) and fire risk (4.5E-5) totals 1.1E-4 per year. This value (1.1E-4) is significantly above both the Draft License Renewal GEIS all hazards value of 6.1E-5 and the 8.4E-5 internal events value used in the original 1996 License Renewal GEIS to make its decisions. In fact, per RG-1.174 Rev. 3 An Approach for Using PRA in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, January 2018 (ML17317A256), a value of 1.1E-4 per year places NAPS in a risk region: Applications that result in increases to CDF above 1E-5 per reactor year would not normally be considered (Page 28 of RG 1.174).

In this regard, it should be noted that Dominion has not completed a fire PRA on NAPS. Therefore, the above combined internal and external CDF sum of 1.1E-4 per year is based on an industry mean fire CDF of 4.5E-5 per year which corresponds to a median fire CDF of 4.6E-5 per year (from the Draft License Renewal GEIS, Table E.3-10 (Page E-28)). It should be remembered that half of all values are above the median value. Therefore, there is a 50% probability that the NAPS fire CDF is greater than the median value of 4.6E-5. Draft NUREG-1437 Rev. 2 Table E.3-10 reports fire CDFs for 68 reactors or about 75% of the current fleet. In this regard, I would also note that a quarter of the fleet has either not performed a fire PRA or has not reported their fire PRA results to the NRC. Because the NRC has only partial fire CDF data, it is very possible that a plant that has not performed a fire PRA has a fire CDF higher or even significantly higher than the highest results reported (i.e., Turkey Point Unit 3 with a reported fire CDF of 8.7E-5 per year - Draft License Renewal GEIS Appendix E Table E-3.10 at Page E-28). That is, NAPS could have a fire CDF significantly higher than the value used to make this important regulatory decision. With the tools available today, this is not acceptable.1

Page F-11 (lines 16 - 18) the Draft SEIS states: The staff also noted that the actions taken by Dominion and experience gained after the 2011 Mineral earthquake provide additional assurance regarding North Annas ability to handle a beyond-design-basis seismic event. The Draft SEIS gives no explanation for

1 At Page F-11 (lines 24 - 32) the Draft SEIS uses ratios and percentages to make arguments why margin exists between the SLR decision today and the original decisions for LR in 1996 GEIS. As documented by the illustration above, there is no need to use these ratio surrogates when actual CDF values are or should be available.

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what actions taken or experienced gained has anything to do with the quantitative risk reduction or the resolving the design basis exceedance in 2011.

At Page F-11 (lines 28 - 29) the Draft SEIS states that the population dose risk was calculated in the initial SAMA analysis to be 50 person-rem per reactor year (RY). According to the NRC: This provides a ratio of the North Anna 1996 LR GEIS 95 percent upper confidence bound predicted population dose to North Anna initial license renewal total population dose risk of 30. This considerable margin offsets any increases in external events since the previous SAMA analysis. (lines 29 - 31). The NRC argues that a ratio of 30 ( 1,496 / 50 = 30 ) between the 1996 LR GEIS and the initial license renewal SAMA total population does risk supplies assurance that risks are low enough. But it appears that the initial LR 50 person-rem/RY calculation was at least in part based on an unrealistically low seismic CDF, given that the Dominion-reported seismic CDF is 6.3E-5 per year. The current base estimate analysis shows a ratio of external event risk to internal event risk of 75 [ = ( 3.9E-5 (fire CDF) + 6.3E-5 (seismic CDF) ) / 1.36E-6 (internal events CDF) ]. Thus, the external to internal events ratio of 75 increase swamps the 30 ratio reduction argued by the Draft SEIS.

In Section F.3.3 New Source Term Information (Section E.3.3 of the 2013 LR GEIS) Page F-12 (lines 17 -

18, the Draft SEIS) states: The NRC staff expects to incorporate the information gleaned from the SOARCA project in future revisions of the LR GEIS (NRC 2013-TN2654). NUREG-1935, State-of-the-Art Reactor Consequence Analyses (SOARCA) Report was published in November 2012. One would think that the subsequent decade would supply sufficient time to incorporate any insights gained from the SOARCA work into the Draft GEIS Rev. 2 or this Draft SEIS.

In Section F.3.5 Higher Fuel Burnup Information (Section E.3.5 of the 2013 LR GEIS) at Page F-14 (lines 1 -

4) the Draft SEIS reports the NRC Staffs conclusion that: [N]o new and significant information exists for North Anna SLR concerning offsite consequences due to higher fuel burnup that would alter the conclusions reached in the 1996 LR GEIS and 2013 LR GEIS or the North Anna initial LR SEIS. But on Page F-13 (lines 33 - 36) the Draft SEIS states: According to the 2013 LR GEIS, increased peak fuel burnup from 42 to 75 gigawatt days per metric ton uranium (GWd/MTU) for PWRs resul ts in small to moderate increases (up to 38 percent) in population dose in the event of a severe accident (emphasis added). It then goes on to say (Page F-13 lines 39 - 40): Dominion indicated that the average burnup level of the peak rod is not planned to exceed 60,000 MWd/MTU during the proposed SLR operating term. The Draft SEIS contradicts itself as it says that a 42,000 to 75,000 MWd/MTU can lead to a moderate increase and that Dominion may operate at the 60,000 MWd/MTU range. It appears that this should be a moderate increase and not a no new and significant information conclusion.

At Page F-14 (lines 26 - 32) Section F.3.6 Low Power and Reactor Shutdown Event Information (Section E.3.6 of the 2013 LR GEIS) references industry initiatives described in SECY 97-168 for the proposition that: [T]he offsite consequences of severe accidents, considering low power and reactor shutdown events, during the North Anna SLR term would not exceed the impacts predicted in either the 1996 LR GEIS or 2013 LR GEIS. SECY 97-168 discusses improvements in outage conduct achieved in the early 1990s. The benefits achieved by these industry changes were known before the 1996 initial LR GEIS was promulgated and were incorporated into the 1996 initial LR GEIS. Any residual improvements post the 1996 initial LR GEIS were obtained and certainly understood long prior to the 2013 GEIS update. Thus, the NRC should not double-count those industry initiatives as it does here. To the contrary, the NRC should consider expert industry evidence that the number of outage events increased during the decade between 2000 and 2010, see INPO SOER 10-2 September 7, 2010.

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Section F.3.7 Spent Fuel Pool Accident Information (Section E.3.7 of the 2013 LR GEIS) (Pages F-14 to F-15) fails to capture new information documented in NUREG-2161 "Consequence Study of a Beyond-Design-Basis Earthquake Affecting the Spent Fuel Pool for a U.S. Mark I Boiling Water Reactor,"

September 2014. While this is a BWR analysis it is relevant to PWRs also. The authors of the Draft SEIS appear to be unaware of the work. This section also fails to capture and address the fact that utilities are packing more spent fuel with higher burnup into spent fuel pools (SFPs), while at the same time not increasing the heat removal capacity of the SFP cooling systems. These facts are not considered in this analysis.

At page F-16 (lines 11 - 14) Section F.3.9 Uncertainties (Section E.3.9 of the 2013 LR GEIS) presents the NRC Staffs conclusion in the 2013 LR GEIS, the NRC staff concluded that the reduction in environmental impacts resulting from the use of new information (since the 1996 LR GEIS analysis) outweighs any increases in impact resulting from the new information. With a close reading of the sentence, it can be reduced to The staff concluded the reduction in impacts resulting from the use of new information outweighs any increases from the new information. This makes no sense.

This section is also seriously deficient because it purports to address risk uncertainties without complying with NUREG-1855 Rev. 1 "Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making," March 2017 (ML17062A466). As stated in NUREG-1855, that document provides guidance on how to treat uncertainties associated with PRAs used by a licensee or applicant to support a risk-informed application to NRC. The NRC Commissioners themselves identified

[t]reatment of uncertainty as an important issue for regulatory decisions in NRC Final Policy Statement Use of PRA Methods in Nuclear Regulatory Activities, Federal Register, Vol. 51, p. 42622 (51 FR 42622), Washington, D.C., 1995.) As they explained: Uncertainties exist... from knowledge limitations... A probabilistic approach has exposed some of these limitations and provided a framework to assess their significance and assist in developing a strategy to accommodate them in the regulatory process. Id. The Draft SEIS should be revised consistent with the Commissions direction.

Section F.3.10 Summary and Conclusion (Section E.5 of the 2013 LR GEIS) at Pages F-17 to F-18 reiterates areas of risk reduction but fail to mention the increase in risk identified by new seismic analysis. See discussion above.

Page F-18 (lines 3 - 4) states: The LR GEIS estimated the net increase from the five areas listed above would be (in a simplistic sense) approximately an increase by a factor of 4.7. The Draft SEIS does not specify where the factor of 4.7 comes from in the LR GEIS. The Draft SEIS then goes on to combine this 4.7 factor risk increase with a factor of 25 risk reduction in the internal events CDF to compute a total risk reduction by a factor of 20.3 (25 minus 4.7) (at 4). Assuming that the increase factor of 4.7 and the decrease factor of 25 are correct, the math applied is wrong. Instead of subtraction, the two factors should be divided, i.e., 4.7 divided by 25 yielding a net reduction by a factor of 0.2 or 5. This reduction is significantly less than the factor of 20.3 claimed.

Section F.4.1 10 CFR 50.54(hh)(2) Requirements Regarding Loss of Large Areas of the Nuclear Power Plant Caused by Fire or Explosions starting at Page F-19 and continuing to Page F-20 discusses the risk improvements obtained from efforts by the NRC and licensees post September 2001. However, any risk reductions obtained by these efforts should have been incorporated into the NAPS initial license renewal in 2002 and certainly by the 2013 GEIS Revision 1 report. The NAPS PRA used as the basis of Dominion Environmental Report (ER) supplied as part of the SLR application in 2020 should have

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captured any of these risk reductions. Assuming that the NAPS 2020 PRA does capture these risk reductions, the Draft SEIS should not be attempting to credit them again.

Starting at Page F-19 (at 46) and continuing on Page F-20 (lines 1 - 4) the Draft SEIS states: These enhancements included significant reinforcement of the defense capabilities for nuclear facilities, better control of sensitive information, enhancements in emergency preparedness, and implementation of mitigating strategies to deal with postulated events potentially causing loss of large areas of the nuclear power plant due to explosions or fires, including those that an aircraft impact might create. None of these enhancements can be quantified in the PRA as these risks were never incorporated in the base PRA to begin with. Thus, to imply that the quantified risk has been improved is misleading at best. A more accurate characterization is that the base PRA understated the risks. Take for example the stated risk reduction from aircraft impact improvements, the aircraft impact hazard was not incorporated into the base PRA. If a risk reduction is to be credited, then the risk should be acknowledged and added into the base risk analysis before any enhancement is credited.

At Page F-20 (lines 14 - 16) the Draft SEIS states: NRC requirements pertaining to nuclear power plant security are subject to NRC oversight on an ongoing basis under a nuclear power plants current operating license and are beyond the scope of license renewal. If security is indeed beyond the scope of LR and SLR, then the Draft SIES should not argue for risk reductions for enhance security.

At Page F-21 (line 15) Section F.4.3 Fukushima-Related Activities, the Draft SEIS states that [T]here was a partial meltdown of fuel in three of the reactors (emphasis added). Fukushima Dai-ichi Units 1, 2 and 3 melted enough fuel to generate at least three powerful explosions. Unit 1 has fuel debris below the reactor pressure vessel. These should not be characterized as partial meltdowns.

Starting at Page F-21 (lines 41 - 42) Section F.4.4 Operating Experience states: Section E.2 of the 2013 LR GEIS mentions the considerable operating experience that supports the safety of U.S. nuclear power plants. The 2013 GEIS in Section E.2 discusses operating experience that has led to improved performance. This discussion includes topics on: IPE/IPEEE, aging monitoring improvements, generic safety issue (GSI) 191 on sump performance, and the September 2011 terrorist attacks. All but GSI 191 have already been credited in Appendix F of this Draft SEIS. This Draft SEIS section is suggesting that there are some other risk improvements not previously captured in Appendix F. However, the 2013 GEIS examples are not new and have already been credited in this Draft SEIS. The NRC should not double-count factors that have already been considered.

At Page F-24 (lines 17 - 19) Section F.5.3 Dominions Evaluation of 1 Unimplemented North Anna Phase 2 SAMAs states: SAMAs related to creating a containment vent were screened out because this nuclear power plant modification has been evaluated industrywide and explicitly found to not be cost effective in Westinghouse large/dry containments. The Draft SEIS should supply a reference documenting this bald assertion.

Starting at Page F-24 (lines 45 - 46) and continuing onto Page F-25 (lines 1 - 2) Section F.5.4 Dominions Evaluation of SAMAs Identified as Potentially Cost-Beneficial at Other U.S. Nuclear Power Plants that Are Applicable to North Anna states: Of the results presented in Table E4.15-2 [of the NAPS ER], one case (labeled as emergency diesel generator (EDG)) yielded an internal events LLRF (Large Late Release Frequency) reduction of 57 percent. However, Dominion explained that the total change in the Maximum Benefit for the EDG case is well below 50 percent. The SAMA methodology has a threshold for continued evaluation of a 50% reduction. The case identified here has a risk reduction of 57% and

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thus, exceeds the 50% threshold. The NRC should explain: Why wasnt this case explored in more detail as required by the SAMA methodology?

At Page F-25 (lines 7 - 8) the Draft SEIS states: The NRC staff reviewed North Annas onsite information and its SAMA Stage 1 process during an in-office audit at NRC headquarters (NRC 2020 -TN8100 see Appendix D). The supplied reference (NRC 2020 -TN8100) goes to ML20351A388. This reference document is a four-page letter from NRC to Dominion documenting the occurrence of the in -office audit. It does not document the audit findings. There is no Appendix D to this letter and thus, no documentation that: The staff found that Dominion had used a methodical and reasonable approach to identify any SAMAs that might reduce the maximum benefit by at least 50 percent and therefore could be considered potentially significant (Page F-25 lines 8 - 11). Thus, it is not possible to evaluate this claim. The NRC should supply the missing information.

Section F.5.6 Conclusion at Page F-26 (lines 29 - 33) the Draft SEIS states: The NRC staff reviewed Dominions new and significant information analysis for severe accidents and SAMAs at North Anna during the SLR period and finds Dominions analysis and methods to be reasonable. As described above, Dominion evaluated a total of 334 SAMAs for North Anna SLR and did not find any SAMAs that would reduce the maximum benefit by 50 percent or more. This conclusion is inaccurate. As discussed above Dominion found a EDG SAMA with a Phase 1 risk reduction of 57%.

At Page F-26 (lines 37 - 41) the Draft SEIS states: [T]he NRC staff did not otherwise identify any new and significant information that would alter the conclusions reached in the previous SAMA analysis for North Anna. Therefore, the NRC staff concludes that there is no new and significant information that would alter the conclusions of the SAMA analysis performed for North Annas initial license renewal. In the Draft SEIS, the NRC has not documented any effort to look for new and significant information beyond the work presented by Dominion or document in previous versions the GEIS or the NAPS supplemental EISs. This is not acceptable. The NRC should at a minimum review its own Generic Issues Program and the Office of Researchs ongoing research plan for relevant new information.

At Page F-26 (lines 42 - 43) the Draft SEIS states: In addition, given the low residual risk at North Anna, the substantial decrease in internal event CDF (emphasis added) Here the NRC fails to acknowledge the identified increase in seismic risk and the complete reliance on industry average fire risk evaluations as Dominion has not published any fire risk results. See discussion above.

Climate Change

Various sections of the Draft SEIS address climate change. However, the Draft SEIS does not address climate changes impacts on accident risks in Section 3.11.6.9 or Appendix F. This omission constitutes a significant deficiency in the Draft EIS because climate change demonstrably affects the frequency and intensity of some external events and therefore has the potential to significantly increase accident risks.

Moreover, the frequency and intensity of climate change effects are increasing over time. Given that the NRC is proposing to rely on the Draft SEIS for decisions that could affect reactor safety decades from now, the Draft SEIS must address these changing effects over the entire licensed lifetime of reactors, which may end 4 decades from now.

As discussed above, the Draft SEIS is already inadequate as a general matter for making broad generalizations about external event CDF based on extrapolations from internal event CDF values and

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limited actual plant-specific values for external event CDF. Appendix F looks explicitly at external events focusing exclusively on seismic issues. It ignores other external events such as flooding, external fires (e.g., forest and wildfires), tornadoes, etc. Climate change has already started to increase the frequency and intensity of these events. See, for example, Climate change is probably increasing the intensity of tropical cyclones, March 31, 2021 NOAA, https://www.climate.gov/news-features/understanding-climate/climate-change-probably-increasing-intensity-tropical-cyclones; Climate Change Indicators:

Weather and Climate, EPA, https://www.epa.gov/climate-indicators/weather-climate; Global Warming and Hurricanes, NOAA Geophysical Fluid Dynamics Laboratory, April 11, 2023, https://www.gfdl.noaa.gov/global-warming-and-hurricanes/.

The NRC is well-aware of the issues of climate change and its impact on nuclear plant safety. After the Fukushima meltdowns, the NRC Office of Research initiated a research program to develop tools to assist in probabilistic and deterministic assessments of external hazards including seismic, high winds and flooding with a consideration of climate change. See NRC Probabilistic Flood Hazard Assessment Research Program Overview,, February 22 - 25, 2021 (ML21064A418) and Potential Impacts of Accelerated Climate Change, PNNL-24868, May 2016 (ML16208A282)). In addition, climate change has been a topic of discussion at the NRCs Regulatory Information Conference (RIC) in recent years. See Climate Change Impact on the Safety of Nuclear Installations, March 8-10, 2022 (ML22140A312)) &

Observations on Extreme Weather and Impacts on Nuclear Power Plants, EPRI ML22140A320, 2022).

Accident risk evaluations for climate change must be site-specific

The effects of climate change on accident risk are and will continue to be site-specific and not subject to generalization. Some reactors already have been identified as vulnerable to climate impacts and others

- like NAPS - have not been evaluated for their vulnerability. Given what we know about some U.S.

reactors, it is unacceptable not to provide a comparable analysis for NAPS.

Reactors for which climate vulnerability has been demonstrated or poses an unusual risk include Oconee, Turkey Point, and Duane Arnold. For example, the three reactors at the Oconee plant -- for which the NRC is now considering an application for subsequent license renewal -- lie downstream of two large dams. The design of the dams includes consideration of the maximum probable flood induc ed by the maximum probable precipitation (i.e., storm). Climate change has the potential to significantly increase the amount of precipitation falling on watersheds above the dams. Will the dams be able to pass these higher intensity storms and the resulting floods? See the attached declaration NRC Relicensing Crisis at Oconee Nuclear Station: Stop Duke from Sending Safety Over the Jocassee Dam for a thorough analysis.

Another example is the Turkey Point plant, located in a low-lying coastal area of South Florida. With climate change the already-occurring, sea level change will continue and possibly accelerate during the SLR period. Likewise, hurricane intensity, i.e., wind speed, rain fall and storm surge, will intensify.

As discussed below, the Duane Arnold plant in Iowa was prematurely and permanently shuttered after being hit with a Derecho with wind speeds exceeding 100mph. Climate change has been implicated in the severity of this extreme weather event (Hints of a derecho-climate change link, ten years after 2012 storm, Washington Post, June 29, 2022, https://www.washingtonpost.com/climate -

environment/2022/06/29/derecho -climate-change-severe-storm/)

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Therefore, in order to provide a reasonably thorough and complete analysis of accident risks during the license renewal term, the Draft SEIS must address the continuing and growing contribution of climate change to accident risks at nuclear plants. And this evaluation must be conducted on a site-specific basis.

Effects of climate change considerations on Probabilistic Analysis

Climate change affects risk in two ways. First, it increases the likelihood or initiating event frequency of events. For example, increased storm frequency can lead to higher initiating event frequency for losses of offsite power (LOOPs). Second, climate change can increase the probability of failure of design features or mitigation equipment. A 2020 severe windstorm at the Duane Arnold plant (ML21139A091) illustrates this phenomenon. While the storm may or may not be directly attributable to climate change, it is a reasonable example of the type of severe weather effects that climate change can cause today and will cause in the future. In that case, a severe windstorm caused a loss of offsite power (LOOP). As a result of the LOOP, debris accumulated at the suction of the service water systems, which are necessary to cool the emergency diesel generators (EDGs) and the emergency core cooling system (ECCS) heat exchangers. The NRCs risk analysis of the event showed an increase in the failure probabilities of the service water system, the EDGs and the ECCS due to this climate -related external event. Consideration of these risks in an EIS would provide important information regarding climate -related accident risk as well as identification of mitigation measures to address those risks.

A third way that climate change affects risk analysis, which is unique to flooding risk, is the cliff edge effect. With most hazards if the severity is increased slightly, the stress on the system is increased somewhat proportionately. However, with many flood-related issues, a small increase in the hazard can cause a dramatic and often overwhelming impact on a structure. For example, a small increase in wave height could raise the flood height sufficiently to overtop a floodwall inundating the equipment the floodwall is designed to protect. Risk analyses for climate change-related flooding must look carefully at this cliff-edge phenomenon.

Finally, the National Academies under sponsorship of the National Oceanic and Atmospheric Administration (NOAA) has started a project to modernize the probable maximum precipitation (PMP) methodology (https://www.nationalacademies.org/our -work/modernizing-probable-maximum-precipitation-estimation#sectionSponsors). This project will consider approaches for estimating PMP in a changing climate, with the goal of recommending an updated approach, appropriate for decision -maker needs. PMP is a significant input into the design of critical infrastructure such as dam and reactor safety analysis directly and indirectly through its impact on probable maximum flood (PMF). The NRC is well aware of this effort as they have already participated in at least one of the initial project workshops.

PMP and PMF also impact reactor safety directly via their impact on local intense precipitation (LIP). The Draft SEIS is silent on this and is thus deficient. As this process is likely to take several years, if the SEIS cannot wait for resolution, then any plant issued a SLR prior to its resolution should be required to revisit the issue once the update is completed.

Multiunit Impacts

The Draft SEIS does not address the environmental impacts of concurrent multi -unit accidents. This is a significant omission, given the well-recognized independent contribution that multi -unit accidents make to accident risk. As discussed in a 2013 paper by NRC Staff member Suzanne Schroer and University of

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Maryland Professor Dr. Mohammed Modarres, multi-unit site risk is neither formally nor adequately considered in either the regulatory or the commercial nuclear environment [citations omitted] despite the fact that the questions of multi-unit accident is not one of possibility, but of probability. Schroer and Modarres, An Event Classification Schema for Evaluating Site Risk in a Multi-Unit Nuclear Power Plant Probabilistic Risk Assessment, p. 1 (2013) (ML13217A335).

As recognized by Schroer and Modarres, the events at Fukushima Daiichi in 2011 underlined the significance and importance of accident events involving multiple units. Id, p. 1. And indeed, the NRC has been discussing how to address the issue of multi-unit nuclear power plant PRAs for many years, including a lessons learned report after the Chernobyl accident. Id. The Near Term Task Force Review of Insights from the Fukushima Dai-ichi Accident (July 12, 2011) (ML111861807) specifically identified multi-unit accidents as an issue that should be investigated and addressed. As noted by the Task Force:

The accident at Fukushima has shown that prolonged SBO [station black out] and multiunit events are realities that must be addressed as part of EP [emergency planning]. While of low probability, these events have the potential for severe consequences that require an effective EP response. The Task Forces evaluation in this section focuses on a licensees capability to respond during these types of events. Currently, the United States has 29 single-unit sites, 33 dual-unit sites, and 3 triple-unit sites. The agency is currently reviewing new reactor applications that may add units to existing sites; however, no applicant has requested to bring the total number of units at a single site to more than four. In most cases, proposed quadruple-unit sites have physical separation between the two existing and the two proposed units.

Id., p. 51. While the NTTF focused its recommendations on safety improvements related to emergency planning, its conclusion that multi-unit accidents are realities with potentially severe consequences demonstrates their relevance to risk analysis for environmental impact studies.

The differences between single-unit PRAs and multi-unit PRAs are well-understood, as is the risk-significance of failing to address interdependent multi-unit events. As described in a recent paper by Taotao Zhou, Mohammad Modarres, Enrique López Droguett:

Conventional PRA studies have traditionally been restricted to single reactor units and are referred to as single-unit PRAs (SUPRAs). The SUPRAs include accident scenarios exclusive to one reactor unit, assuming the effects of other units are not critical. Hence, SUPRAs only consider the dependencies between the structures, systems, and components (SSCs) within a single reactor unit. These dependencies, referred to as intra-unit dependencies, are likely to induce multiple failure events that may overcome redundancies or diversities and ultimately lead to a class of SSC failures called dependent failures. Although these dependent events are usually much less frequent than the independent events, they have proven to be the most critical contributors to the likelihood of reactor core damage, environmental radioactive exposure, and overall plant risk. Typically, the influence of these dependencies is explicitly modeled in the PRA event tree and fault tree logics or implicitly treated as the type of dependencies commonly referred to as common cause failure events.

Multi-unit nuclear power plant probabilistic risk assessment: A comprehensive survey, Taotao Zhou, Mohammad Modarres, Enrique López Droguett, Reliability Engineering & System Safety, Volume 213, September 2021 (emphasis added).

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(https://www.sciencedirect.com/science/article/abs/pii/S0951832021003070) As further explained by Profs. Zhou, Modarres, and Droguett:

These inter-unit dependencies can play critical roles in nuclear accident risks with the possibility of multiple core damages, including damages to the spent fuel pool and other radioactive waste storage facilities. Proper characterization of these site-level dependencies is thus critical to obtain an accurate risk profile of a nuclear power plant site. Examples of these inter-unit dependencies include certain initiating events simultaneously occurring in multiple units, a transient event in one unit affecting some or all of the other units, the proximity of the units to each other, shared structures, components (e.g., shared batteries and diesel generators),

common operation practices, and substantial procedural and other organizational similarities.

(emphasis added). Three important conclusions can be drawn from the study of multi -unit accidents by the NRC and independent researchers. First, multi-unit accident risks - including risks to reactors and fuel storage pools - are well-understood as reasonably foreseeable. Second, multi-unit accidents have unique characteristics that are not bounded by single -unit accident risk studies. Finally, the risks of multi-unit accidents are unique to reactor sites, and must consider the relative location of reactor units, fuel storage pools, and other onsite facilities. Therefore, multi-unit accident risks must be independently evaluated for each separate reactor site for which license renewal is considered.

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CURRICULUM VITAE JEFFREY T. MITMAN Poolesville, MD February 22, 2024

QUALIFICATIONS Reliability and risk analyst with more than 40 years experience in the nuclear industry. Skills include evaluation and modeling of probabilistic risk analyses (PRA) and management of PRA projects and teams.

Highly experienced in low power and shutdown (LPSD) risk modeling issues. Solid record of bringing projects in on schedule and budget.

MAJOR ACCOMPLISHMENTS

• Transitioned NRC to detailed PRA models for LPSD significance determinations process evaluations.

• Guided development of and managed industrys first configuration risk management software tool.

• Obtained regulatory approval of EPRIs risk informed in-service inspection (RI-ISI) methodology.

• Managed first PRA of bolted spent fuel storage cask.

EXPERIENCE PRIVATE CONSULTANT (Poolesville, MD)

Nuclear risk analyst 2021-Present

• Reviewed Oconee Subsequent License Renewal application and prepared technical report on adequacy of environmental and safety analyses to address flooding risks.

• Reviewed and submitted comments on NRCs draft (2023) Generic Environmental Impact Statement (NUREG-1437 Revision 2).

US NUCLEAR REGULATORY COMMISSION (Rockville, MD) 2005 - 2021 Senior Reliability and Risk Analyst (NRC Office of Nuclear Reactor Regulation)

• Conducted Significance Determination Process (SDP) evaluations of reactor events including development and/or modification of required models.

• Served as lead analyst for LPSD event issues and concerns.

• Guided development of shutdown Standardized Plant Analysis Risk (SPAR) models.

• Conducted Human Reliability Analysis (HRA).

• Evaluated external event risk from dam failures.

• Served on NRCs Japan Team (part of USAID disaster assistance response team for Fukushima Daiichi accident), providing technical advice and support through the U.S. Ambassador to Japanese government.

• Participated in post NRCs Fukushima Near Term Task Force (NTTF) flooding guidance development.

• Developed NRCs guidance on crediting FLEX in risk-informed regulatory applications.

• Advised NRC National Fire Protection Association (NFPA) 805 team on issues related to shutdown fire risk.

• Performed evaluations of risk informed license applications.

Reliability and Risk Analyst (NRC Office of Nuclear Regulatory Research)

• Project Manager for the development of shutdown SPAR models ERIN ENGINEERING AND RESEARCH, INC. (Walnut Creek, CA) 2004 - 2005 Lead Senior Engineer

• Prepared configuration risk management evaluation of at-power fire risk.

• Prepared configuration risk management evaluation of loss of offsite power.

ABE STAFFING SERVICES (Palo Alto, CA) 2003 - 2005 Consultant to EPRI

• Brought project to closure involving Dry Cask Storage PRA project and team, involving Transnuclear bolted cask containing PWR fuel.

Jeffrey T. Mitman l Page 2

EPRI (Palo Alto, CA) 1998 - 2003 Project Manager

• Outage Risk Assessment and Management (ORAM-Sentinel)

- Grew first of a kind software application for performing configuration risk management in nuclear power plants.

- Conducted research in low power and shutdown risk; shutdown initiating event and event frequency derivation.

- Delivered multiple versions (including alpha, beta & production), testing and full documentation.

- Administered utility user group, marketing, contract preparation, technology transfer, technical report publication and training.

- Actively managed both development and application contracts with multiple suppliers and customers.

Managed annual $1M budget.

• Dry Cask Storage PRA: Initiated innovative analysis of Transnuclear cask containing PWR fuel.

- Managed unique team with diverse experience in both cask design and PRA backgrounds.

• Risk Informed In-service Inspections Project (RI-ISI): Lead team in obtaining regulatory approval of methodology to safely reduce piping weld inspection requirements using combination of probabilistic and degradation analysis.

- Responsible for methodology finalization and acceptance by industry and U.S. NRC.

- Conducted marketing, sales, contract preparation, technology transfer, training and technical report publication.

- Actively managed both development and application contracts with both suppliers and customers.

Managed annual $1M budget.

• Human Reliability Analysis Project: Managed project to bring consistency to on industry use of HRA methods.

- Responsible for EPRI HRA area, including development of HRA Calculator software and establishment of associated users group.

ERIN ENGINEERING AND RESEARCH, INC. (Palo Alto, CA) 1992 - 1998 Lead Senior Engineer Collaborated with EPRI ORAM-SENTINEL Project Manager in project development and administration, user group administration, contract preparation, technology transfer workshops, technical report generation and editing. Performed ORAM analysis of the Diablo Canyon plant. Performed ORAM Probabilistic Analysis of Perry spent fuel pool. Drafted and edited ORAM V2.0 Users Manual. Assisted in ORAM-SENTINEL software design, performed software debugging. Principle researcher and author of BWR outage contingency report. Prepared marketing and training, materials.

ABB IMPELL CORPORATION (King of Prussia, PA) 1990 - 1992 Lead Senior Engineer

• Design Basis Documentation: directed team of three engineers to review PECO Feedwater System Design. Wrote Design Basis Documentation reports for Limerick and Peach Bottom power plants, identifying licensing and design concerns by reviewing the system design as documented in drawings, calculations, vendor manuals, Technical Specifications, UFSAR, SER, SRP, 10CFR50.59 safety evaluations etc. and by interfacing with utility engineering personnel. Prepared Engineering Change Requests as necessary.

• Shift Outages: during Limerick Nuclear Power Plant refueling / maintenance outage. Coordinated all shift maintenance work and testing. Collaborated with all groups in power plant, allocating resources as needed to maintain schedule and reporting to senior plant outage management. Performed system reviews prior to placing them back in service. Conducted shift outage meetings. Tracked work group performance against schedule. Advised utility management on techniques for schedule and outage organizational improvements.

Jeffrey T. Mitman l Page 3

GENERAL ELECTRIC COMPANY (San Jose, CA) Experience Prior to 1990 Startup-Test Engineer

• Shift Startup Engineer: During power ascension phase coordinated all system testing on shift and startup interface with operations. During preoperational phase, acted as operations shift supervisor responsible for coordinating all system testing and flushing on shift from main control room. Updated senior utility management daily on testing status.

• Additional positions: Shift Technical Advisor, Test Engineer, Lead QC / Welding Inspector

EDUCATION / PROFESSIONAL DEVELOPMENT

• BSE, Nuclear Engineering, University of Michigan, Ann Arbor, MI

• Introductory VBA class, University of California, Berkeley, CA

• Misc. business courses at various colleges and universities

• Senior Reactor Operator Certified

• GE Station Nuclear Engineering

• Effective Utilization of PSA, ERIN Engineering & Research, Walnut Creek, CA.

PROFESSIONAL ASSOCIATIONS

• American Nuclear Society (ANS) member since 1978.

• ANS Nuclear elected member of Installation Safety Division Executive Committee 2015 to 2021.

• ANS Risk Informed Standards Committee (RISC).

• ANS/ASME Risk Informed Standards Writing Group on Shutdown PRA Standard.

• ASME Section XI, Working Group on Implementation of Risk Based Examination.

• MIT Professional Summer Programs Guest Lecturer at Risk-Informed Operational Decision Management Course.

PAPERS

1. Technical Challenges Associated with Shutdown Risk when Licensing Advanced Light Water Reactors, PSAM12 2014. Co-author.

2. Comparing Various HRA Methods to Evaluate Their Impact on the results of a Shutdown Risk Analysis during PWR Reduced Inventory, PSAM11 2012. Co-author.

3. Uncertainty Analysis for Large Dam Failure Frequencies Based on Historical Data, PSAM11 2012. Co-author.

4. An Assessment of Large Dam Failure Frequencies Based on US Historical Data, PSA 2011. Co-author.

5. Generic Failure Rate Evaluation for Jocassee Dam, US NRC (ML13039A084), 2010. Co-author.

6. Development of PRA Model for BWR Shutdown Modes 4 and 5 Integrated in SPAR Model, to be presented at PSAM10 2010. Co-author.

7. Development of Standardized Probabilistic Risk Assessment Models for Shutdown Operations Integrated in SPAR Level 1 Model, PSAM9 2008. Co-author.

8. PRA of Bolted Dry Spent Fuel Storage Cask, Presented at ICONE12. 2004. Co-author.

9. Low Power and Shutdown Risk Assessment Benchmarking, Presented at PSA 02 2002. Co-author.

10. EPRI Human Reliability Analysis Guidelines, Presented at PSA 02 2002. Co-author.

11. Derivation of Shutdown Initiating Event Frequencies, Presented at PSAM5 2000. Co-author.

12. Quantitative Assessment of a Risk Informed Inspection Strategy for BWR Weld Overlays, Presented at ICONE 8 2000. Co-author.

13. EPRI RI-ISI Methodology and the Risk Impacts of Implementation, Presented at SMiRT 11 1999. Co-author.

14. Application of Markov Models and Service Data to Evaluate the Influence of Inspection on Pipe Rupture Frequencies published. PVP 1999. Co-author.

15. Progress in Risk Evaluation of Outages, International Conference on the Commercial and Operational Benefits of PSA. 1997. Co-author.

16. Control of Reactor Vessel Temperature/Pressure during Shutdown, GE SIL 357. June 1981. Co-author.

Jeffrey T. Mitman l Page 4

SOFTWARE

1. HRA Calculator Version 2.0, EPRI 2003. 1003330. Project Manager (PM).

2. ORAM-Sentinel Version 3.4, EPRI 2001. 1002958. PM and co-author.

REPORTS / STANDARDS

1. Requirements for Low Power and Shutdown PRA - ANS/ASME-58.22-2014 (Trial Use Standard).

2. Probabilistic Risk Assessment (PRA) of Bolted Storage Casks: Quantification and Analysis Report, EPRI 2003. 1002877. PM.

3. Low Power and Shutdown Risk Assessment Benchmarking Study, EPRI, Palo Alto, CA and U.S. DOE.

2002. 1003465. PM and principal investigator.

4. Dry Cask Storage PRA Scoping Study, EPRI 2002. 1003011. PM.

5. Guidance for Incorporating Organizational Factors into Nuclear Power Plant Risk Assessments: Phase 1 Workshop. EPRI and U.S. DOE 2002. 1003322. PM.

6. An Analysis of Loss of Decay Heat Removal Trends and Initiating event Frequencies (1989-2000):

EPRI 2001. 1003113. PM.

7. Piping System Failure Rates and Rupture Frequencies for Use in Risk Informed In-service Inspection Applications: TR-111880-NP, EPRI 2000. 1001044. PM.

8. Application of Risk-Informed Inservice Inspection Alternative Element Selection Criteria. EPRI, Charlotte NC: 2000. TE-11482. PM.

9. Revised Risk-Informed Inservice Inspection Evaluation Procedure, EPRI 1999. TR-112657 Revision B-A. PM & co-author.

10. Piping System Failure Rates and Rupture Frequencies for Use in Risk Informed In-service Inspection Applications, EPRI 1999. TR-111880. PM.

11. Comparison between EDF and EPRI of Pipe Inspection Optimization Methods, EPRI Palo Alto, CA; Electricite de France, Paris, France: 1999. TR-113315. PM.

12. Economic Feasibility Study of Implementing RBISI at 2-loop PWR, EPRI 1998. TR-107613. PM.

13. Evaluation of Pipe Failure Potential via Degradation Mechanism Assessment, EPRI Palo Alto, CA:

1998. TR-110157. PM.

14. Piping Failures in U.S. Nuclear Power Plants: 1961-1997, EPRI 1998.TR-110102. PM.

15. Piping System Reliability Models and Database for used in Risk Informed Inservice Inspection Applications, EPRI 1998. TR-110161. PM.

16. Use of Risk Informed Inspection Methodology for BWR Class 1 Piping, EPRI 1998. TR-110701. PM.

17. ORAM v4.0 Functional Specification Outline, EPRI 1999. TR-111652. PM.

18. Survey on the Use of Configuration Risk and Safety Management Tools at NPPs, EPRI, 1998.

TR-102975. PM.

19. ORAM-SENTINEL Demonstration at Diablo Canyon, EPRI 1998. TR-110739. PM.

20. ORAM-SENTINEL Development at Indian Point 3, EPRI 1999, TR-110716. PM.

21. ORAM-SENTINEL Development and ORAM Integration at Oconee, EPRI 1998. TR-111207. PM.

22. ORAM-SENTINEL Development at Fitzpatrick, EPRI 1998. TR-110505. PM.

23. ORAM-SENTINEL Demonstration at Sequoyah, EPRI 1998. TR-110771. PM.

24. SENTINEL Technical Basis Report for Limerick, EPRI 1998. TR-108953. PM.

25. Outage Risk Assessment and Management Implementation at Fermi 2, EPRI 1997. TR-109013. Co-author.

26. Contingency Strategies for BWRs during Potential Shutdown Operations Events, EPRI 1993.

TR-102973. Principal investigator.

27. Generic Outage Risk Management Guidelines for BWRs, EPRI 1993. TR-102971. Co-principal investigator.

ATTACHMENT 2 A: Declaration of Declaration of Glen Besa (March 23, 2024)

B: Declaration of Erica Gray (March 23, 2024)

C: Declaration of Jerry Rosenthal (March 24, 2024)

D: Declaration of Barbara Cruikshank (March 23, 2024)

E: Declaration of John Cruikshank (March 22, 2024)

F: Declaration of Diane Johnson (March 23, 2024)

G: Declaration of William J. Johnson (March 23, 2024)

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION

BEFORE THE SECRETARY

)

In the Matter of )

Virginia Electric Power Co. ) Docket Nos. 50-338/339 SLR North Anna Power Station Units 1 & 2 )

____________________________________)

DECLARATION OF GLEN BESA

Under penalty of perjury, Glen Besa declares as follows:

1) My name is Glen Besa. I am a member of Beyond Nuclear, Inc. (Beyond Nuclear).

2) I live at 4869 Burnham Road, North Chesterfield, VA 23234.

3) My home is located within the 50-mile emergency planning zone Ingestion Pathway Zone of the North Anna Nuclear Power Station, for which Virginia Electric Power Company (VEPCO) has submitted an application to the US Nuclear Regulatory Commission for the Subsequent License Renewal of its operating licenses. Both North Anna units have previously received a 20-year license extension on their original 40-year operating licenses, which now expire in 2038 and 2040, respectively.

4) Based on the historical experience of nuclear power stations, I believe that these reactors are inherently dangerous. Continued operations of North Anna Nuclear Power Station Units 1 and 2 for an additional 20 years beyond 2038 and 2040 could cause a severe nuclear accident in the reactor(s) and/or irradiated fuel storage pool(s), thereby causing death, injury, illness, dislocation, and economic damage to me and my family. A severe reactor accident could also cause devasting environmental damage.

5) I believe that VEPCOs application to extend operations of North Anna Units 1 and 2 from 60 to 80 years is inadequate to reasonably assure the protection of my health, safety and the environment.

Therefore, I have authorized Beyond Nuclear to represent my interests in this proceeding.

Executed in Accordance with 10 C.F.R. § 2.304(d) by

Glen Besa Date: March 23, 2024 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE SECRETARY

)

In the Matter of )

Virginia Electric Power Co. ) Docket Nos. 50-338/339 SLR North Anna Power Station Units 1 & 2 )

___________________________________ )

DECLARATION OF ERICA GRAY Under penalty of perjury, Erica Gray declares as follows:

1) My name is Erica Gray. I am a member of Beyond Nuclear, Inc. (Beyond Nuclear).

2) I live at 406 Glendale Drive, Henrico, Virginia 23229.

3) My home is located within the 50-mile emergency planning zone Ingestion Pathway Zone of the North Anna Nuclear Power Station, for which Virginia Electric Power Company (VEPCO) has submitted an application to the US Nuclear Regulatory Commission for the Subsequent License Renewal of its operating licenses. Both North Anna units have previously received a 20-year license extension on their original 40-year operating licenses, which now expire in 2038 and 2040, respectively.

4) Based on the historical experience of nuclear power stations, I believe that these reactors are inherently dangerous. Continued operations of North Anna Nuclear Power Station Units 1 and 2 for an additional 20 years beyond 2038 and 2040 could cause a severe nuclear accident in the reactor(s) and/or irradiated fuel storage pool(s), thereby causing death, injury, illness, dislocation, and economic damage to me and my family. A severe reactor accident could also cause devasting environmental damage.

5) I believe that VEPCOs application to extend operations of North Anna Units 1 and 2 from 60 to 80 years is inadequate to reasonably assure the protection of my health, safety and the environment. Therefore, I have authorized Beyond Nuclear to represent my interests in this proceeding.

Executed in Accordance with 10 C.F.R. § 2.304(d) by Erica Gray Date: 3/23/2024

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE SECRETARY

)

In the Matter of )

Virginia Electric Power Co. ) Docket Nos. 50-338/339 SLR North Anna Power Station Units 1 & 2 )

___________________________________ )

DECLARATION OF JERRY ROSENTHAL Under penalty of perjury, Jerry Rosenthal declares as follows:

1) My name is Jerry Rosenthal. I am a member of Beyond Nuclear, Inc. (Beyond Nuclear).

2) I live at 877 Holland Creek Road, Louisa, Virginia 23093. I also have a home at 1213 Belleview Avenue, Charlottesville, Virginia 22091.

3) Both of my homes are located within the 50-mile emergency planning zone Ingestion Pathway Zone of the North Anna Nuclear Power Station, for which Virginia Electric Power Company (VEPCO) has submitted an application to the US Nuclear Regulatory Commission for the Subsequent License Renewal of its operating licenses. Both North Anna units have previously received a 20-year license extension on their original 40-year operating licenses, which now expire in 2038 and 2040, respectively.

4) Based on the historical experience of nuclear power stations, I believe that these reactors are inherently dangerous. Continued operations of North Anna Nuclear Power Station Units 1 and 2 for an additional 20 years beyond 2038 and 2040 could cause a severe nuclear accident in the reactor(s) and/or irradiated fuel storage pool(s), thereby causing death, injury, illness, dislocation, and economic damage to me and my family. A severe reactor accident could also cause devasting environmental damage.

5) I believe that VEPCOs application to extend operations of North Anna Units 1 and 2 from 60 to 80 years is inadequate to reasonably assure the protection of my health, safety and the environment. Therefore, I have authorized Beyond Nuclear to represent my interests in this proceeding.

Executed in Accordance with 10 C.F.R. § 2.304(d) by Jerry Rosenthal Date: March 24, 2024

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE SECRETARY

)

In the Matter of )

Virginia Electric Power Co. ) Docket Nos. 50-338/339 SLR North Anna Power Station Units 1 & 2 )

___________________________________ )

DECLARATION OF BARBARA CRUICKSHANK Under penalty of perjury, Barbara Cruickshank declares as follows:

1) My name is Barbara Cruickshank. I am a member of the Sierra Club.

2) I live at 700 Spring Lake Drive, Earlysville, Virginia 22936.

3) My home is located within the 50-mile emergency planning zone Ingestion Pathway Zone of the North Anna Nuclear Power Station, for which Virginia Electric Power Company (VEPCO) has submitted an application to the US Nuclear Regulatory Commission for the Subsequent License Renewal of its operating licenses. Both North Anna units have previously received a 20-year license extension on their original 40-year operating licenses, which now expire in 2038 and 2040, respectively.

4) Based on the historical experience of nuclear power stations, I believe that these reactors are inherently dangerous. Continued operations of North Anna Nuclear Power Station Units 1 and 2 for an additional 20 years beyond 2038 and 2040 could cause a severe nuclear accident in the reactor(s) and/or irradiated fuel storage pool(s), thereby causing death, injury, illness, dislocation, and economic damage to me and my family. A severe reactor accident could also cause devasting environmental damage.

5) I believe that VEPCOs application to extend operations of North Anna Units 1 and 2 from 60 to 80 years is inadequate to reasonably assure the protection of my health, safety and the environment. Therefore, I have authorized the Sierra Club to represent my interests in this proceeding.

Executed in Accordance with 10 C.F.R. § 2.304(d) by Barbara Cruickshank Date: March 23, 2024

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE SECRETARY

)

In the Matter of )

Virginia Electric Power Co. ) Docket Nos. 50-338/339 SLR North Anna Power Station Units 1 & 2 )

___________________________________ )

DECLARATION OF JOHN CRUICKSHANK Under penalty of perjury, John Cruickshank declares as follows:

1) My name is John Cruickshank. I am a member of the Sierra Club.

2) I live at 700 Spring Lake Drive, Earlysville, Virginia 22936.

3) My home is located within the 50-mile emergency planning zone Ingestion Pathway Zone of the North Anna Nuclear Power Station, for which Virginia Electric Power Company (VEPCO) has submitted an application to the US Nuclear Regulatory Commission for the Subsequent License Renewal of its operating licenses. Both North Anna units have previously received a 20-year license extension on their original 40-year operating licenses, which now expire in 2038 and 2040, respectively.

4) Based on the historical experience of nuclear power stations, I believe that these reactors are inherently dangerous. Continued operations of North Anna Nuclear Power Station Units 1 and 2 for an additional 20 years beyond 2038 and 2040 could cause a severe nuclear accident in the reactor(s) and/or irradiated fuel storage pool(s), thereby causing death, injury, illness, dislocation, and economic damage to me and my family. A severe reactor accident could also cause devasting environmental damage.

5) I believe that VEPCOs application to extend operations of North Anna Units 1 and 2 from 60 to 80 years is inadequate to reasonably assure the protection of my health, safety and the environment. Therefore, I have authorized the Sierra Club to represent my interests in this proceeding.

Executed in Accordance with 10 C.F.R. § 2.304(d) by John Cruickshank Date: March 22, 2024

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE SECRETARY

)

In the Matter of )

Virginia Electric Power Co. ) Docket Nos. 50-338/339 SLR North Anna Power Station Units 1 & 2 )

___________________________________ )

DECLARATION OF DIANA JOHNSON Under penalty of perjury, Diana Johnson declares as follows:

1) My name is Diana Johnson. I am a member of the Sierra Club.

2) I live at 40 Winsome Lane, Fredericksburg, Virginia 22406.

3) My home is located within the 50-mile emergency planning zone Ingestion Pathway Zone of the North Anna Nuclear Power Station, for which Virginia Electric Power Company (VEPCO) has submitted an application to the US Nuclear Regulatory Commission for the Subsequent License Renewal of its operating licenses. Both North Anna units have previously received a 20-year license extension on their original 40-year operating licenses, which now expire in 2038 and 2040, respectively.

4) Based on the historical experience of nuclear power stations, I believe that these reactors are inherently dangerous. Continued operations of North Anna Nuclear Power Station Units 1 and 2 for an additional 20 years beyond 2038 and 2040 could cause a severe nuclear accident in the reactor(s) and/or irradiated fuel storage pool(s), thereby causing death, injury, illness, dislocation, and economic damage to me and my family. A severe reactor accident could also cause devasting environmental damage.

5) I believe that VEPCOs application to extend operations of North Anna Units 1 and 2 from 60 to 80 years is inadequate to reasonably assure the protection of my health, safety and the environment. Therefore, I have authorized the Sierra Club to represent my interests in this proceeding.

Executed in Accordance with 10 C.F.R. § 2.304(d) by Diana Johnson Date: March 23, 2024

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE SECRETARY

)

In the Matter of )

Virginia Electric Power Co. ) Docket Nos. 50-338/339 SLR North Anna Power Station Units 1 & 2 )

___________________________________ )

DECLARATION OF WILLIAM J. JOHNSON Under penalty of perjury, William J. Johnson declares as follows:

1) My name is William J. Johnson. I am a member of the Sierra Club.

2) I live at 40 Winsome Lane, Fredericksburg, Virginia 22406.

3) My home is located within the 50-mile emergency planning zone Ingestion Pathway Zone of the North Anna Nuclear Power Station, for which Virginia Electric Power Company (VEPCO) has submitted an application to the US Nuclear Regulatory Commission for the Subsequent License Renewal of its operating licenses. Both North Anna units have previously received a 20-year license extension on their original 40-year operating licenses, which now expire in 2038 and 2040, respectively.

4) Based on the historical experience of nuclear power stations, I believe that these reactors are inherently dangerous. Continued operations of North Anna Nuclear Power Station Units 1 and 2 for an additional 20 years beyond 2038 and 2040 could cause a severe nuclear accident in the reactor(s) and/or irradiated fuel storage pool(s), thereby causing death, injury, illness, dislocation, and economic damage to me and my family. A severe reactor accident could also cause devasting environmental damage.

5) I believe that VEPCOs application to extend operations of North Anna Units 1 and 2 from 60 to 80 years is inadequate to reasonably assure the protection of my health, safety and the environment. Therefore, I have authorized the Sierra Club to represent my interests in this proceeding.

Executed in Accordance with 10 C.F.R. § 2.304(d) by William J. Johnson Date: March 23, 2024