ML24338A248

(Redacted) Hearing Request and Petition to Intervene by Beyond Nuclear and Sierra Club and Petition for Waiver

ML24338A248
Person / Time
Site:	Oconee
Issue date:	09/02/2021
From:	Curran D Beyond Nuclear, Harmon, Curran, Harmon, Curran, Spielberg & Eisenberg, LLP, Sierra Club
To:	NRC/SECY
Shared Package
ML24338A251	List: ML24338A248
References
50-269-SLR, 50-270-SLR, 50-287-SLR, ASLBP 22-973-01-SLR-BD01, RAS 56245, Subsequent License Renewal
Download: ML24338A248 (73)

Text

This document is a publicly available redacted version. The original file is non-public and designated as Critical Energy/Electric Infrastructure Information (CEII) in ADAMS under ML21270A250.

Most of the redactions were approved by the Federal Energy Regulatory Commission (FERC) on August 13, 2024 (ML24324A040). However, beyond the redactions approved on that date, additional redactions were made by NRC staff to address CEII missed in the August 13, 2024, FERC-approved version. The additional redacted values were identified as CEII and subsequently approved by FERC on November 20, 2024, (ML24326A159) in the following documents:

NRC Staff Attachments A-D REDACTED (ML24326A364)

NRC Staff Attachments 1 and 2 REDACTED (ML24326A366)

Furthermore, to ensure consistency with the FERC-approved documents mentioned above, the units of measure accompanying each CEII value were also redacted in this document.

For clarity, the excerpt below highlights the additional values redacted throughout this document that were missed in the August 13, 2024, FERC approval, but were redacted by NRC staff and are consistent with the November 20, 2024, FERC-approved documents.

and; NRC staff actions were taken in accordance with:

The Memorandum of Understanding Between NRC and FERC Regarding Treatment of Critical Energy/Electric Infrastructure Information found at: https://www.nrc.gov/reading-rm/doc-collections/memo-understanding/2024/index.html.

The FERC definition of CEII found at: https://www.ferc.gov/ceii, and, https://

www.ferc.gov/enforcement-legal/ceii/designation-incoming-dam-safety-documents.

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE SECRETARY In the Matter of

)

Duke Energy Carolinas, LLC

) Docket Nos. 50-269/270/287 SLR Oconee Nuclear Station,

)

Units 1, 2 & 3

)

HEARING REQUEST AND PETITION TO INTERVENE BY BEYOND NUCLEAR AND SIERRA CLUB AND PETITION FOR WAIVER OF 10 C.F.R. §§ 51.53(c)(3)(i),

51.53(c)(3)(ii)(L), 51.71(d), 51.95(c)(1),

AND 10 C.F.R. PART 51 SUBPART A, APPENDIX B, TABLE B-1 TO ALLOW CONSIDERATION OF CATEGORY 1 NEPA ISSUES I.

INTRODUCTION AND

SUMMARY

OF CONTENTIONS Pursuant to 10 C.F.R. § 2.309, and the hearing notice published at 86 Fed. Reg. 40662(July 28, 2021), 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 Duke Energy Corp.s (Dukes) application for subsequent license renewal (SLR) of the operating license for the Oconee Units 1, 2 and 3 nuclear power plant.1 If Dukes application is granted, it will be allowed to operate Oconee Units 1, 2 and 3 for an additional twenty years beyond its current renewed operating license term, or until 2053-54.2 Petitioners contend that the NRC should not approve subsequent renewal of Dukes operating 1 The portion of Dukes SLR application that is relevant to this Hearing Request is Appendix E, Dukes Environmental Report are Oconee Power Station Units 1 and 2, Application for Subsequent License Renewal (March 2021) (NRC ADAMS Accession Nos. ML21158A195 (Part 1) and ML21158A196 (Part 2)).

2 The NRC issued operating licenses for Oconee Units 1, 2, and 2 in 1973 and 1974, and renewed those licenses in 2000. Environmental Report at 1-1. Currently, the operating licenses for Units 1 and 2 will expire in 2033, and the operating license for Unit 3 will expire in 2034. Id.

2 license because Dukes Environmental Report fails to comply with the requirements of the National Environmental Policy Act (NEPA) and NRC implementing regulations. In Section IV below, Petitioners contentions set forth three significant concerns:

First, Duke has submitted a truncated Environmental Report, which avoids discussing a range of potentially significant impacts of a second license renewal term, including severe accidents. Duke claims to rely on a set of regulatory exemptions in 10 C.F.R. § 51.53(c)(3)(i) and Table B-1 of 10 C.F.R. Part 51, Subpart A, Appendix B. Duke also claims to rely on 10 C.F.R. § 51.53(c)(3)(ii)(L), which excuses it from preparing a SAMA analysis if it prepared one previously. But these exemptions, by their own terms, apply only to initial license renewal applicants, not to subsequent license renewal applicants such as Duke.3

Second, even assuming for purposes of argument that Duke could rely on the regulatory exemptions in Table B-1, it has failed to address new and significant information showing that the past environmental analyses on which Duke relies are now demonstrably wrong, and that the risk of a core melt accident during a second license renewal term is significant and must be addressed.

In particular, Dukes own risk analysis, with corrected initiating event frequencies, shows that the likelihood of a core melt accident caused by a random failure of the Jocassee Dam, which lies twelve miles above Oconee, is significantly higher than presented in Dukes Environmental Report.4 In addition, Duke failed to take other factors into account that further raise the likelihood of a core melt accident caused by Jocassee 3 See 10 C.F.R. § 51.53(c)(3).

4 Mitman Report at 22 (Equation 5).

3 Dam failure, including seismically-induced dam failure and dam overtopping. This new and significant information demonstrates that Duke erred by concluding that operation of Oconee for an additional license term will have no significant environmental impacts. It also demonstrates that Dukes analysis of Severe Accident Mitigation Alternatives (SAMAs) is incorrect and should be done again using reasonable and up-to-date assumptions.

Finally, Duke fails to address the environmental significance of the Staffs 2011 Safety Evaluation that the potential for a random (i.e., sunny day) Jocassee Dam failure constitutes an adequate protection issue and therefore must be addressed by new measures to protect against flooding of essential safety equipment that would inevitably cause a reactor meltdown.5 Because Duke has failed to implement these measures, the outstanding and unresolved safety issue now constitutes a significant environmental issue that must be addressed in the Environmental Report. Duke must explain why it has yet to implement the flood mitigation measures ordered by the NRC in 2011, and also explain how that failure affects its SAMA analysis.

By failing to address the environmental impacts of flooding caused by failure of the Jocassee Dam in light of newly available information demonstrating the significance of the environmental risk, or to re-evaluate SAMAs in light of this new information, Duke fails to comply with the requirements of NEPA and NRC implementing regulations.

Petitioners contentions and Waiver Petition are supported by the attached declaration and Expert Report of Jeffrey T. Mitman. See Declaration of Jeffrey T. Mitman in Support of 5 2011.01.28 Safety Evaluation on Confirmatory Action Letter to Address External Flooding Concerns, (ML110280153) (2011 NRC Safety Evaluation Letter).

4 Beyond Nuclears and Sierra Clubs Hearing Request (Sept. 27, 2021) (Mitman Declaration)

(Attachment 1 to this Hearing Request); and Mr. Mitmans Expert Report, NRC Relicensing Crisis at Oconee Nuclear Station: Stop Duke From Sending Safety Over the Jocassee Dam (Sept.

2021) (Mitman Report or Expert Report) (Exhibit 1 to Mitman Declaration).

The remainder of this Hearing Request is organized as follows: Section II contains a demonstration that Petitioners Beyond Nuclear and the Sierra Club have organizational standing to participate in this proceeding.Section III sets forth the legal framework for Petitioners Hearing Request.Section IV presents Petitioners Contention.Section V contains a Petition for Waiver of 10 C.F.R. §§ 51.53(c)(3)(i), 51.53(c)(3)(ii)(L), 51.71(d), 51.95(c)(1), and 10 C.F.R. Part 51 Subpart A, Appendix B, Table B-1 to Allow Consideration of Category 1 NEPA Issues.

Section VI 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 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

5 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.6 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 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 Dukes operating license is not approved for a second renewal term unless and until Duke 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 within 50 miles of Oconee Units 1, 2 and 3, whose interests in protecting their own health and the health of the environment would be adversely affected by extended operation of Oconee Units 1, 2 and 3 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 Oconee reactors.7 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 6 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).

7 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)).

6 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 fourteen 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 Declaration of Jane F. Powell (Sept. 21, 2021) (Attachment 2A).

B. Standing of the Sierra Club Founded in 1892, the Sierra Club is a national environmental organization with 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 Frank M. Powell (Sept. 21, 2021)

(Attachment 2B) and Declaration of Rosellen Aleguire (Sept. 23, 2021) (Attachment 2C).

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.8 Even where the NRC purports to have resolved safety issues through its Atomic Energy Act-based regulatory process, 8 Limerick Ecology Action v. NRC, 869 F.2d 719, 729-31 (3rd Cir. 1989).

7 it must nevertheless comply with NEPAs procedural obligations for addressing those issues in its decision-making processes.9 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.10 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). Under the NRCs Part 50 regulations, a nuclear reactor may not be licensed absent a finding that the operation will not be inimical to public health and safety, will provide reasonable assurance that public health and safety will not be endangered; and NRCs Part 54 regulations require confirmation that those reasonably safe practices will continue.11 9 Id.

10 42 U.S.C. § 2133(d).

11 See, e.g., 10 C.F.R. § 50.40(c) (requiring NRC to find that [t]he issuance of a construction permit, operating license, early site permit, combined license, or manufacturing license to the applicant will not, in the opinion of the Commission, be inimical to the common defense and security or to the health and safety of the public); 10 C.F.R. § 50.57(a)(3) (conditioning the issuance of a reactor operating license on a finding of reasonable assurance (i) that the activities authorized by the operating license can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the regulations in this chapter); 10 C.F.R. § 54.29(a) (conditioning reactor license renewal on a finding that

[a]actions have been identified and have been or will be taken... such that there is reasonable assurance that the activities authorized by the renewed license will continue to be conducted in accordance with the [current licensing basis].).

See also Maine Yankee Atomic Power Co. (Maine Yankee Atomic Power Station), ALAB-161, 6 A.E.C. 1000, 1009 (1973) (equating the reasonable assurance that the operation of a facility will not endanger the health and safety of the public with the no undue risk standard).

8 B. NEPA

1. Statutory requirements NEPA implements a broad national commitment to protecting and promoting environmental quality.12 NEPA has two key purposes: to ensure that the agency will have 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.13 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.14 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.15 The hallmarks of a hard look are thorough investigation into environmental impacts and forthright acknowledgment of potential environmental harms.16The analysis must address both the probabilities of potentially harmful events and the 12 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).

13 Robertson, 490 U.S. at 349-50.

14 Robertson, 490 U.S. at 349-50; Hughes River Watershed Conservancy v. Glickman, 81 F.3d 437, 443 (4th Cir. 1996).

15 Robertson, 490 U.S. at 349; 42 U.S.C. § 4332(c).

16 National Audubon Society v. Dept of Navy, 422 F.3d 174, 185 (4th Cir. 2005).

9 consequences if those events come to pass.17 Impacts must be quantified if possible, but may be addressed qualitatively, with consideration of uncertainties.18 In addition, the agency must rigorously explore and objectively evaluate the projected environmental impacts of all reasonable alternatives for completing the proposed action.19 The alternatives analysis is the heart of an EIS.20

2. NRC Regulations and Guidance for Implementation of NEPA

a. Regulations NRC regulation 10 C.F.R. § 51.45(a) requires that a reactor license applicant must submit an environmental report with its application. Specific requirements for license renewal applications are set forth in 10 C.F.R. § 51.53(c). Section 51.53(c)(2) establishes general requirements for reactor license renewal applicants, and § 51.53(c)(3) establishes requirements for applicants seeking an initial renewed license.

Section 51.53(c)(2) requires an operating license renewal applicant (other than an applicant for initial license renewal) to describe, inter alia, the affected environment around the plant, the environmental impacts of alternatives, and any other matters described in § 51.45(a). Section 51.45(a), requires, in turn, that the Environmental Report must include the following information:

Analysis. The environmental report must include an analysis that considers and balances the environmental effects of the proposed action, the environmental impacts of 17 State of New York v. NRC, 681 F.3d 471, 479, 482 (D.C. Cir. 2012) (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).

18 Limerick Ecology Action, 869 F.2d at 744 (lack of quantifiability does not necessarily preclude further consideration under NEPA.).

19 Van Ee v. EPA, 202 F.3d 296, 309 (D.C. Cir. 2000).

20 City of Alexandria, Va. v. Slater, 198 F.3d 862, 866 (D.C. Cir. 1999) (quoting 40 C.F.R. § 1502.14).

10 alternatives to the proposed action, and alternatives available for reducing or avoiding adverse environmental effects.... The analyses for environmental reports shall, to the fullest extent practicable, quantify the various factors considered. To the extent that there are important qualitative considerations or factors that cannot be quantified, those considerations or factors shall be discussed in qualitative terms. The environmental report should contain sufficient data to aid the Commission in its development of an independent analysis.

Section 51.53(c)(3) excuses initial license renewal applicants from addressing environmental impacts that have been classified as Category 1 generic impacts in 10 C.F.R. Part 51, Subpart A, Appendix B, Table B-1. The NRC made these generic impact findings in the Generic Environmental Impact Statement for License Renewal of Nuclear Plants, NUREG-1437, Vols. 1-2 (May 1996) (ML040690705, ML040690738) (1996 License Renewal GEIS).21 At the same time the NRC issued the 1996 License Renewal GEIS, it codified Table B-1s generic findings and made them applicable in license renewal proceedings via 10 C.F.R. §§ 51.53(c)(3)(i), 51.71(d), and 51.95(a).22 The generic findings in Table B-1 include a finding that the environmental impacts of reactor accidents are small.

Pursuant to 10 C.F.R. § 51.53(c)(3)(ii)(L), initial operating license renewal applicants must evaluate alternatives to mitigate severe accidents, if they have not already done so.

Where an initial operating license applicant has previously considered environmental impacts, 10 C.F.R. § 51.53(c)(3)(iv) requires the applicant to contain any new and significant information regarding the environmental impacts of license renewal of which the applicant is aware.

Interested members who wish to challenge Category 1 environmental determinations and SAMA 21 In 2013, the NRC updated the 1996 License Renewal GEIS in the Generic Environmental Impact Statement for License Renewal of Nuclear Plants, NUREG-1437, Rev. 1 (June 2013)

(ML13106A241) (2013 Revised License Renewal GEIS).

22 Final Rule for Environmental Review for Renewal of Nuclear Power Plant Operating Licenses, 61 Fed. Reg. 28,467 (June 5, 1996).

11 analyses made under 10 C.F.R. §§ 51.53(c)(3)(i) and 51.53(c)(ii)(L) must submit a waiver petition pursuant to 10 C.F.R. § 2.335(b).23 IV.

CONTENTIONS CONTENTION 1: FAILURE TO COMPLY WITH 10 C.F.R. §§ 51.53(c)(2).

A. Statement of Contention Dukes Environmental Report fails to satisfy 10 C.F.R. § 51.53(c)(2) because it fails to fulfill that provisions requirement to discuss the environmental impacts of alternatives and any other matters described in [10 C.F.R.] § 51.45. In particular, Duke incorrectly relies on 10 C.F.R. § 51.53(c)(3) to excuse it from discussing significant environmental impacts classified as Category 1 in 10 C.F.R. Part 51, Part A, Appendix B.24 By its own terms, however, 51.53(c)(3) applies only to applicants seeking an initial renewed license, and therefore does not apply to SLR applicants. Pursuant to 10 C.F.R. §§ 51.53(c)(2) and 51.45(a), Duke must discuss all significant environmental impacts of the proposed approval of Dukes SLR application, including the environmental impacts of reactor accidents and alternatives for avoiding or mitigating those impacts. The impacts that must be considered include the environmental impacts of a core melt accident caused by failure of the Jocassee Dam. Relevant information that must be considered is set forth in Contentions 2 and 3, below, which are hereby adopted and incorporated by reference.

B. Basis Statement In preparing its Environmental Report, Duke has assumed that the content of the Environmental Report is governed by 10 C.F.R. § 51.53(c)(3), which excuses initial license 23 Exelon Generation Co. (Limerick Generating Station, Units 1 and 2), CLI-12-19, 76 N.R.C.

377, 386 (2012) 24 See Environmental Report, Table 1.1-1 at 1-4 and 1-5.

12 renewal applicants from addressing a set of Category 1 issues that were evaluated in 1996 License Renewal GEIS. Environmental Report at E-4-2. These excluded environmental impacts include the probability and consequences of design basis accidents and beyond design basis accidents. Petitioners respectfully submit that Dukes reliance on 10 C.F.R. § 51.53(c)(3) is legally erroneous, because § 51.53(c)(3) - by its plain language - applies only to applicants for initial license renewal, not SLR applicants. Nor does the regulatory history of NRCs NEPA regulations or the content of the Commissions generic environmental impact statements support Dukes reading of the regulations. Instead, Dukes Environmental Report is governed by 10 C.F.R. § 51.53(c)(2) and the other regulations it cross-references.

Nevertheless, Petitioners recognize that the Commission has upheld Dukes interpretation of NRC NEPA regulations in two other subsequent license renewal cases, Florida Power and Light Co. (Turkey Point Nuclear Generating Plant Units 3 and 4), CLI-20-03, 91 N.R.C. 33, 141-45 (2020) (Turkey Point - CLI-20-03) and Exelon Generation Co., L.L.C. (Peach Bottom Atomic Power Station, Units 2 and 3), CLI-20-11, _ N.R.C. _, slip op. at 11-12 (Nov. 12, 2020)

(Peach Bottom - CLI-20-11). At this time, therefore, those opinions govern this proceeding.

Nevertheless, Petitioners rely on and incorporate by reference the dissenting opinion of Commissioner Baran in Turkey Point-CLI-20-03 and the dissenting opinion of Commissioners Baran and Hanson in Peach Bottom CLI-20-11. In addition, Petitioners rely on and incorporate by reference the rationale and citations in the concurring and dissenting opinion of ASLB Judge Abreu in Florida Power and Light Co. (Turkey Point Nuclear Generating Plant Units 3 and 4),

LBP-19-03, 89 N.R.C. 245, 303 (2019).

13 C. Demonstration that the Contention is Within the Scope of the Proceeding This Contention is within the scope of the SLR proceeding for Oconee because it seeks compliance by Dukes Environmental Report with NEPA and NRC implementing regulations.

D. Demonstration that the Contention is Material to the Findings NRC Must Make to Renew Dukes Operating License This Contention is material to the findings NRC must make regarding the environmental impacts of re-licensing Oconee for a second renewed license term, because it challenges the adequacy of Dukes Environmental Report to support those findings.

E. Concise Statement of the Facts or Expert Opinion Supporting the Contention, Along with Appropriate Citations to Supporting Scientific or Factual Materials The concise statement of law supporting this contention is provided in Section C above.

In addition, Petitioners rely on Contentions 2 and 3 for examples of environmental impacts that must be considered in an Environmental Report that fully complies with 10 C.F.R. §§ 51.53(c)(2) and 51.45. Contentions 2 and 3 are hereby adopted and incorporated by reference for that purpose.

CONTENTION 2: FAILURE TO CONSIDER NEW AND SIGNIFICANT INFORMATION REGARDING SIGNIFICANT IMPACTS OF REACTOR ACCIDENTS CAUSED BY FAILURE OF JOCASSEE DAM A. Statement of Contention Even assuming for purposes of argument that Dukes SLR application is governed by 10 C.F.R. § 51.53(c)(3) and the categorical exclusions of 10 C.F.R. Part 51, Part A, Appendix B, Duke has violated NEPA and 10 C.F.R. § 51.53(c)(3)(iv) by failing to address new and significant information regarding the environmental impacts of license renewal of which [Duke]

is aware. The Commission should waive 10 C.F.R. § 51.53(c)(3) and the categorical exclusions

14 of 10 C.F.R. Part 51, Part A, Appendix B, and require Duke to address those impacts in a complete environmental impact analysis, as set forth in 10 C.F.R. § 51.45.

The new and significant information of which Duke is aware, and that is not addressed in the Environmental Report, consists of the following:

Dukes own risk analyses show that the likelihood of a core melt accident and containment failure caused by a random failure of the Jocassee Dam is significantly higher than presented in Dukes Environmental Report. And even this higher estimate of Jocassee Dam failure frequency is too low, given Dukes failure to consider the additional credible contributors to Jocassee dam failure frequency of seismic events and dam overtopping.

Duke fails to address the environmental significance of a 2011 Safety Evaluation, in which the NRC Staff determined that the potential for a random (i.e., sunny day)

Jocassee Dam failure constitutes an adequate protection issue requiring Duke to implement additional measures to protect against flooding of essential safety equipment and thereby prevent a reactor meltdown. By establishing the risk of a core melt accident with an associated containment failure due to failure of the Jocassee Dam as an adequate protection issue, the NRC effectively established it as a significant environmental issue as well.

B. Basis Statement Petitioners rely on NEPA and the requirements of 10 C.F.R. § 51.45 for this contention. In addition, Petitioners rely on the factual assertions and technical analysis in the attached Mitman Report. As discussed by Mr. Mitman, Duke has failed to consider new and significant information showing that its probability estimates for a random sunny day Jocassee Dam failure

15 are wrong, and that they fail to take into account additional significant contributors to dam failure risk such as seismically-induced dam failure and dam overtopping.25 Duke also fails to consider the risk contribution from shutdown operations. A reasonable best estimate for core damage frequency from a Jocassee Dam failure is 2.8E-4 per year, which is significantly larger than the probability of 5.9E-06, as presented in Dukes SAMA analysis.26 The significance of this new and significant information is highlighted by the fact that a core damage frequency of 2.8E-04 per year is larger than the total core damage frequency form all Oconee internal events of 2.4E-05 per years, and also larger than Dukes estimate of 6.3E-05 per year for all external events.27 Dukes failure to account for this information is profoundly significant, because a significant flooding event at Jocassee would inevitably lead to containment failure and a core melt accident.28 C. Demonstration that the Contention is Within the Scope of the Proceeding This Contention is within the scope of the SLR proceeding for Oconee because it seeks compliance by Dukes Environmental Report with NEPA and NRC implementing regulations.

D. Demonstration that the Contention is Material to the Findings NRC Must Make to Renew Dukes Operating License This Contention is material to the findings NRC must make regarding the environmental impacts of re-licensing Oconee for a second renewed license term, because it challenges the adequacy of Dukes Environmental Report to support those findings.

25 Id., Section 3 (Analysis).

26 Mitman Report at 21 (citing 1998 SAMA Analysis, Page 10). (Duke relied on the same probability estimates it had used in its first license renewal application in 1998.)

27 Id.

28 Id. at 22.

16 E. Concise Statement of the Facts or Expert Opinion Supporting the Contention Along with Appropriate Citations to Supporting Scientific or Factual Materials A concise statement of the facts and expert opinion supporting Contention 2 is set forth in Mr. Mitmans Expert Report, which in turn relies on documents generated by Duke and the NRC.

CONTENTION 3: FAILURE TO CONSIDER NEW AND SIGNIFICANT INFORMATION AFFECTING DUKES ANALYSIS OF SEVERE ACCIDENT MITIGATION ALTERNATIVES C. Statement of Contention Even assuming for purposes of argument that Dukes SLR application is governed by 10 C.F.R. § 51.53(c)(ii)(L) and therefore need not conduct a new SAMA analysis for its second license renewal application, Duke does not comply with § 51.53(c)(3)(iv) because it has failed to consider new and significant information that would affect the outcome of its SAMA analysis.

The new and significant information of which Duke is aware and that is not addressed in the Environmental Report consists of the following:

As discussed above in Contention 2, Dukes own risk analyses show that the likelihood of a core melt accident caused by a random failure of the Jocassee Dam is significantly higher than presented in Dukes Environmental Report. And even this higher estimate of Jocassee Dam failure frequency is too low, given Dukes failure to consider the additional credible contributors to Jocassee dam failure frequency of seismic events and dam overtopping. Contention 2 is hereby adopted and incorporated by reference.

The NRCs 2011 Safety Evaluation, discussed above in Contention 2, required Duke to implement certain measures for protection against a random (i.e., sunny day) Jocassee Dam failure as a matter of providing adequate protection to public health and safety.

By deeming these measures necessary, the NRC established them as SAMAs worthy of

17 consideration. Indeed, in order to exclude a necessary safety measure, Duke would have a very high burden of justification. Yet, these measures are not discussed or implemented in the Environmental Report.

The Environmental Report fails to consider additional mitigative measures that may well be cost-effective at Oconee, given the increased likelihood of a core melt accident. These measures include preemptively shutting down the reactors when reservoir water levels get too high, lowering the water levels in the lake behind the Jocassee and Keowee Dams, or lowering the crest elevation of some of the surround earthworks such that they overtop before the Jocassee Dam, thus lowering the flood impacts at ONS.29 D. Basis Statement Petitioners rely for this contention on the assertions in Contention 2, which is hereby adopted and incorporated herein by reference. In addition, they rely on Mr. Mitmans discussion of SAMAs in Section 3 of his Expert Report.

C. Demonstration that the Contention is Within the Scope of the Proceeding This Contention is within the scope of the SLR proceeding for Oconee because it seeks compliance by Dukes Environmental Report with NEPA and NRC implementing regulations.

D. Demonstration that the Contention is Material to the Findings NRC Must Make to Renew Dukes Operating License This Contention is material to the findings NRC must make regarding the environmental impacts of re-licensing Oconee for a second renewed license term, because it challenges the adequacy of Dukes Environmental Report to support those findings.

29 Mitman Report at 24.

18 E. Concise Statement of the Facts or Expert Opinion Supporting the Contention, Along with Appropriate Citations to Supporting Scientific or Factual Materials A concise statement of the facts and expert opinion supporting Contention 3 is set forth in Mr. Mitmans Expert Report, which in turn relies on documents generated by Duke and the NRC.

V.

PETITION FOR WAIVER OF 10 C.F.R. §§ 51.53(c)(3), 51.53(c)(3)(ii)(L), 51.71(d),

51.95(c)(1), AND 10 C.F.R. PART 51, SUBPART A, APPENDIX B, TABLE B-1 TO ALLOW CONSIDERATION OF CATEGORY 1 NEPA ISSUES As discussed above in Contention 1, Petitioners respectfully submit that the Category 1 exclusions do not apply to environmental documents prepared in support of SLR applications, and that the Commissions rulings in CLI-20-03 (Turkey Point) and CLI-20-11 (Peach Bottom) are legally erroneous. Nevertheless, in order to pursue every avenue available to them for obtaining consideration of their environmental concerns in this SLR proceeding, in the alternative, Petitioners request a waiver under 10 C.F.R. § 2.335(b) of the set of NRC regulations that would bar consideration of their claims: s10 C.F.R. §§ 51.53(c)(3)(i), 51.53(c)(3)(ii)(L),

51.71(d), 51.95(c)(1), and 10 C.F.R. Part 51, Subpart A, Appendix B, Table B-1.

Petitioners seek a waiver of 10 C.F.R. §§ 51.53(c)(3)(i) is necessary in order to allow admission of their Contentions 2 and 3, which challenge the adequacy of Dukes Environmental Report to satisfy NEPA and NRC implementing regulation 10 C.F.R. § 51.53(c)(2) on the ground that it fails to consider the environmental implications of a core melt accident caused by failure of the Jocassee Dam. See Section IV above. In reasonable anticipation that the same issues will arise upon publication by the NRC Staff of the draft and final Supplemental Environmental Impact Statements (SEISs) for subsequent renewal of Dukes operating license, Petitioners also seek a waiver of 10 C.F.R. § 51.71(d) (which would apply the Category 1

19 exclusion to the Draft SEIS) and 10 C.F.R. § 51.95(c)(1) (which would apply the Category 1 to the Final SEIS).

As required by 10 C.F.R. § 2.335(b), this Waiver Petition is supported by the Curran Declaration. See Attachment 3.

A. Standard for Issuance of Waivers NRC regulation 10 C.F.R. § 2.335(b) allows submission of requests to waive or make an exception to the application of a particular regulation, on the sole ground that special circumstances with respect to the subject matter of the particular proceeding are such that the application of the rule or regulation (or a provision of it) would not serve the purposes for which the rule or regulation was adopted. The petition must be supported by an affidavit or declaration that identifies, with particularity, the specific aspect or aspects of the subject matter of the proceeding as to which the application of the rule or regulation (or provision of it) would not serve the purposes for which the rule or regulation was adopted. Id. The Commission has also interpreted § 2.335(b) to require the following four-step demonstration:

1. the rules strict application would not serve the purposes for which it was adopted;

2. special circumstances exist that were not considered, either explicitly or by necessary implication, in the rulemaking proceeding leading to the rule sought to be waived;

3. those circumstances are unique to the facility rather than common to a large class of facilities; and

4. waiver of the regulation is necessary to reach a significant safety or environmental problem.30 30 Duke Nuclear Connecticut, Inc. (Millstone Nuclear Power Station, Units 2 and 3), CLI-05-24, 62 N.R.C. 551, 559-560 (2005); Exelon Generation Co. (Limerick Generating Station, Units 1 and 2), CLI-13-07, 78 N.R.C. 199, 207-08 (2013).

20 B. Petitioners Satisfy the Commissions Four-Part Test for Issuance of a Waiver.

As demonstrated below, Petitioners satisfy the NRCs four-part standard for issuance of a waiver.

1. Application of the Category 1 exclusion would not serve the purpose for which it was promulgated.

The Commissions purposes in making and codifying Category 1 environmental impact findings are set forth at the beginning of the preamble to the 1996 License Renewal Rule:

The amendment improves regulatory efficiency in environmental reviews for license renewal by drawing on the considerable experience of operating nuclear power reactors to generically assess many of the environmental impacts that are likely to be associated with license renewal.

The increased regulatory efficiency will result in lower costs to both the applicant in preparing a renewal application and to the NRC for reviewing plant-specific applications and better focus of review resources on significant case specific concerns.

The results should be a more focused and therefore a more effective NEPA review for each license renewal. The amendment will also provide the NRC with the flexibility to address unreviewed impacts at the site-specific stage of review and allow full consideration of the environmental impacts of license renewal.

61 Fed. Reg. at 28,467. Thus, the Commissions purposes in creating and codifying Category 1 generic impacts may be summarized as increasing efficiency, saving costs, and improving the quality of both generic and site-specific environmental analyses. It is important to note that nothing about this statement of purpose suggested that the Commission would sacrifice NEPA compliance to goals of efficiency or cost-savings. To the contrary, the Commission clearly thought that by improving the efficiency of the environmental review process through generic determinations, it would raise the quality of its NEPA review processes across the board.

Application of the Category 1 exclusions in this proceeding would defeat the Commissions purpose of ensuring NEPA compliance and improving the quality of site-specific license renewal reviews, by barring consideration of new and significant information regarding

21 the environmental impacts of operating Oconee Units 1, 2 and 3 in a subsequent license renewal term. The new and significant information shows that the potential for a core melt accident caused by failure of the Jocassee Dam is significantly higher than estimated by Duke in its SAMA analysis. This new and significant information should be considered in order to ensure that the NRCs subsequent license renewal decision for Oconee Units 1, 2 and 3 is fully informed and that appropriate alternatives are implemented that would prevent or mitigate adverse environmental impacts from accidents that are caused by or contributed to by earthquakes.

The information is new because it has not previously been considered in any environmental document by Duke or the NRC. The information is significant because the difference in risk estimates is high enough to present a seriously different picture of the risk of a core melt accident at Oconee, and affects the cost-benefit analysis of SAMAs to avoid or mitigate those impacts.31

2. Special circumstances exist that were not previously considered.

This Waiver Petition raises special circumstances that were not previously considered, because no previous environmental impact statement has considered the new and significant information presented in Mr. Mitmans Expert Report and Petitioners Contentions 2 and 3.

3. The special circumstances raised by this Waiver Petition are unique to Oconee Units 1, 2 and 3.

This Waiver Petition raises circumstances that are unique to Oconee Units 1, 2 and 3, because the risk assessment relied on by Duke for Oconee Units 1, 2 and 3 is unique to the circumstances of that nuclear plant, i.e., a large 3-unit nuclear reactor complex built beneath two 31 Mitman Expert Report, Pag 24.

22 large dams (Jocassee and Keowee). Oconee is also unique because there is no NRC generic environmental impact statement that addresses the environmental impacts of operating another reactor in the shadow of a large dam, in consideration of the risk information presented in the Mitman Report.

4. Waiver of the regulation is necessary to reach a significant environmental problem.

Petitioners respectfully submit that waiver of the regulations is necessary to address significant adverse environmental impacts that have not previously been considered, as well as SAMAs that may be cost-effective in preventing or mitigating those impacts but that have not previously been considered. These SAMAs include measures deemed necessary by NRC under its Atomic Energy Act-based regulations, but which were not implemented to Petitioners knowledge. The NRCs identification of these measures as necessary, and Dukes failure to implement them, constitute information that is relevant to environmental risk from operation of Oconee and therefore warrants consideration under NEPA.

VI.

CONCLUSION For the foregoing reasons, Petitioners Hearing Request, Petition to Intervene and Waiver Petition 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 September 27, 2021

23 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE SECRETARY In the Matter of

)

Duke Energy Carolinas, LLC

) Docket Nos. 50-269/270/287 SLR Oconee Nuclear Station,

)

Units 1, 2 & 3

)

CERTIFICATE OF SERVICE I certify that on September 27, 2021, I posted on the NRCs Electronic Information Exchange HEARING REQUEST AND PETITION TO INTERVENE BY BEYOND NUCLEAR AND SIERRA CLUB AND PETITION FOR WAIVER OF 10 C.F.R. §§ 51.53(c)(3)(i), 51.71(d),

AND 51.95(c)(1) TO ALLOW CONSIDERATION OF CATEGORY 1 NEPA ISSUES, including:

(Declaration of Jeffrey T. Mitman) o Exhibit 1 to Mr. Mitmans Declaration (his Expert Report, NRC Relicensing Crisis at Oconee Nuclear Station: Stop Duke From Sending Safety Over the Jocassee Dam (September 2021))

o Exhibit 2 to Mr. Mitmans Declaration (curriculum vitae)

A (Declaration of Jane Powell)

B (Declaration of Frank Powell)

C (Declaration of Rosellen Aleguire)

(Declaration of Diane Curran, Esq.)

___/signed electronically by/__

Diane Curran

PETITIONERS HEARING REQUEST AND WAIVER PETITION ATTACHMENT 1

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE SECRETARY In the Matter of Duke Energy Carolinas, LLC Oconee Nuclear Station, U ni t s l, 2 & 3

)

)

Docket Nos. 50-269/270/287 SLR

)

)

DECLARATION OF JEFFREY T. MITMAN IN SUPPORT OF BEYOND NUCLEAR AND SIERRA CLUB HEARING REQUEST Under penalty of perjury, I, Jeffrey T. Mitman declare:

1. My name is Jeffrey T. Mitman. By education and experience, I am a nuclear engineer, with a significant level of expertise in risk analysis.

2.

I have been retained by Beyond Nuclear, Inc. and the Sierra Club to evaluate the Environmental Report submitted by Duke Energy Corp. in connection with its subsequent license renewal (SLR) application to the U.S. Nuclear Regulatory Commission (NRC) for 20-year extension of the license terms for the Oconee Nuclear Plant, Units 1, 2 and 3 (Oconee). My expert report on the application, entitled NRC Relicensing Crisis at Oconee Nuclear Station: Stop Duke From Sending Safety Over the Jocassee Dam (September 2021 ), is attached to my declaration as Exhibit 1. I understand that Beyond Nuclear and the Sierra Club intend to submit a hearing request and waiver petition in this proceeding, based on my expert report.

3.

As set forth in my attached Curriculum Vitae (Exhibit 2), I have more than 40 years of experience in the nuclear industry and 16 years as a regulator with the NRC. My experience includes 16 years on the technical staff of the NRC as a Reliability and Risk Analyst. For the past 15 years, I have 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 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 NRC's 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 some NRC safety reviews and performed risk analysis for Oconee, including reviews related to the risk to Oconee posed by potential failure of the upstream Jocassee Dam. I also participated in a generic study by NRC of dam failure risk, with particular application to Oconee.

6.

I am familiar with correspondence between Duke Energy Corp. (Duke) and the NRC regarding Oconee's design and operation, including NRC Staff technical reports. I am

Beyond Nuclear and Sierra Club Hearing Request also familiar with the Environmental Reports submitted by Duke in connection with its initial license renewal application in 1998, and its subsequent license renewal (SLR) application in June 2021.

7. The factual statements in my expert report are true and correct to the best of my knowledge, and the opinions stated therein are based on my best professional judgment.

%lot 2

MITMAN DECLARATION EXHIBIT 1

i NRC Relicensing Crisis at Oconee Nuclear Station:

Stop Duke from Sending Safety Over the Jocassee Dam Jeffrey T. Mitman September 2021 Submitted to U.S. Nuclear Regulatory Commission on Behalf of Beyond Nuclear, Inc. and the Sierra Club, Inc.

In Subsequent License Renewal Proceeding for Oconee Nuclear Power Plant, Units 1, 2, and 3

ii TABLE OF CONTENTS Table of Contents..i List of Acronyms....iv

1. INTRODUCTION.....1

2. BACKGROUND...2 2.1 Integrated Design and Operation of Oconee Nuclear Reactors and Upstream Dams..2 2.2 Jocassee and Keowee Dam Characteristics.....3 2.3 Oconee Nuclear Plant Design and Construction...3 2.3.1 NRC Safety Requirements for Nuclear Plant Design and Construction...3 2.3.2 Design and Construction of Oconee Units 1, 2 and 3..4 2.3.3 Post-construction Addition of Safe Shutdown Facility...4 2.4 Flood Risk Studies.....5 2.4.1 1983 Flood Study for FERC and Construction of Wall around SSF....5 2.4.2 NSAC-60 Probabilistic Risk Assessment....5 2.4.3 IPE/IPEEE.......6 2.4.4 1992 Flood Study for FERC....7 2.5 Initial Oconee License Renewal and Severe Accident Mitigation Alternatives Analysis...7 2.6 Updated Dam Failure and Flood Routing Evaluations and Related Regulatory Actions...9 2.6.1 NRC 2006 Significance Determination Process on Oconee Flooding Issue....9 2.6.2 NRC 50.54(f) Letter......9 2.6.3 Dukes Response to 50.54(f) Letter.12 2.6.4 2011 NRC Safety Evaluation....13 2.7 Fukushima - Lessons Learned 50.54(f) Letter and Staff Assessment.15 2.8 Dukes 2021 Subsequent License Renewal Application and SAMA Analysis.16

iii

3. ANALYSIS..18

4. CONCLUSION..25

iv LIST OF ACRONYMS AEC Atomic Energy Commission CCDP conditional core damage probability CDF core damage frequency CFR Code of Federal Regulations FERC Federal Energy Regulatory Commission GDC General Design Criterion GL Generic Letter ECCS emergency core cooling system EAP Emergency Action Plan EPRI Electric Power Research Institute FOIA Freedom of Information Act HEC-RAS U.S. Army Corps of Engineers River Analysis System IEF Initiating Event Frequency IPE Individual Plant Examination IPEEE Individual Plant Examination for External Events kv kilovolt LERF large early release frequency LOCA loss of coolant accident LPI low pressure injection MSL mean sea level NRC U.S. Nuclear Regulatory Commission NSAC Nuclear Safety Analysis Center NTTF Near Term Task Force ONS Oconee Nuclear Station ORNL Oak Ridge National Laboratory PMP Probable Maximum Precipitation PRA probabilistic risk assessment ROP Reactor Oversight Process

v SAMA Severe Accident Mitigation Alternatives SDP Significance Determination Process SLR subsequent license renewal SSF Safe Shutdown Facility

1

1. INTRODUCTION This purpose of this report is to explain and provide the basis for my expert opinion, as a nuclear engineer and risk analyst, regarding the safety of Duke Energy Corporations (Dukes) current operation of Oconee Units 1, 2 and 3, and its proposal to extend the reactors operating license terms by 20 years until 2053 (Units 1 and 2) and 2054 (Unit 3). The report is based on my extensive experience as a nuclear engineer and safety regulator with the U.S. Nuclear Regulatory Commission (NRC), including evaluation of Oconees safety in relation to potential failure of the upstream Jocassee Dam. My curriculum vitae is attached.

In my expert opinion, and as discussed in more detail below, Oconees current operation, and proposed operation under an additional twenty-year subsequent license renewal (SLR) term, pose an unacceptable risk to public health and safety, due to Dukes failure to fully implement flood-protective measures required by the NRC in a 2011 Safety Evaluation.1 The NRC deemed those flood protection measures necessary to protect against a core melt accident in the event the Oconee site becomes inundated by failure of the Jocassee Dam.

While the NRC has not sought to force Duke to implement those flood protection measures, neither has it withdrawn or repudiated the 2011 Safety Evaluation in which it found those measures were necessary to provide adequate protection to public health and safety. Instead, the regulatory agency has kept silent with respect to the Safety Evaluation for the past ten years.

Now that Dukes SLR application has come before the NRC, it is time for the agency to break its silence and address the significant safety and environmental issues raised by Dukes bid for another 20 years of unprotected operation. The NRC should not accept Dukes erroneous and outdated risk assessment, and require Duke to provide a thorough and accurate estimate of the core melt risk posed by Jocassee Dam failure. In addition, the NRC should require Duke to implement the flood protection measures required ten years ago by NRC.

In my professional opinion as a recently retired regulator, the NRC has the authority to impose these requirements in order to protect public health and safety. In the event the NRC fails to do so, my report is intended to assist two environmental organizations, Beyond Nuclear and the Sierra Club, to force an accounting by Duke and the NRC under the National Environmental Policy Act (NEPA). As required by NEPA, Duke and the NRC should fully and accurately evaluate the environmental risks of continuing to operate Oconee in spite of the accident risk, which is now known to be higher than what Duke has presented in its SLR application.

A note about secrecy: A significant portion of the information relied on in this report was not available publicly until members of the public forced NRC to release it by requesting it under 1 2011.01.28 Safety Evaluation on Confirmatory Action Letter to Address External Flooding Concerns, (ML110280153) (2011 NRC Safety Evaluation Letter).

2 the Freedom of Information Act (FOIA). I am grateful to Jim Riccio for FOIA Request FOIA/PA-2012-0325 (submitted on behalf of Greenpeace) and Dave Lochbaum for FOIA Request FOIA/PA-2018-0010 (submitted on behalf of the Union of Concerned Scientists), which generated some of the key information relied on this report. The NRC never attempted to justify withholding this critical, damming, and now-public safety information from the public eye, nor is any justification evident.

While Duke and the NRC have continued to withhold some information relevant to this report, the information now in the public record is more than sufficient to show that Duke has failed to provide the public with an accurate, up-to-date, and thorough risk analysis of the potential for a serious core melt accident at Oconee Units 1, 2, and 3 during the second license renewal term.

In addition, publicly available information is more than sufficient to show that for the past ten years, the NRC has considered the risk of a core melt accident caused by Jocassee Dam failure to implicate the adequacy of protection to public health and safety and require significant measures to prevent catastrophe. By assembling this information into a single document, the author seeks to ensure a measure of accountability by Duke and the NRC that they previously eluded through secrecy.

Finally, while some nonpublic documents are cited in the footnotes to this report, the report does not rely on the content of any of those nonpublic documents. Citations of those documents are provided for completeness of the record, not for their content. When the content of the IPEEE or any other nonpublic document is described in this report, that description is taken from descriptions in publicly available documents.

2. BACKGROUND 2.1 Integrated Design and Operation of Oconee Nuclear Reactors and Upstream Dams Duke Energy Corp.s three-unit Oconee Nuclear Plant is located in the mountains of northwestern South Carolina, at the confluence of the Keowee and Little Rivers. Licensed by the U.S. Nuclear Regulatory Commission (NRC) in 1973 and 1974. Oconee is uniquely designed as a pumped storage facility: at the same time the reactors were built, Duke also built two upstream dams, for the purpose of generating additional hydro-powered electricity. When demand for electricity from the reactors was low, the plant could be used to pump water into Jocassee Lake behind the Jocassee Dam. When demand was high, Duke would then allow flow through hydroelectric generators in the dam generating power.

The Jocassee Dams tailwaters were dammed by the Keowee Dam, below which Duke built the Oconee reactor complex. Two hydrogenerators, built into the side of the Keowee Dam, were designed to provide the nuclear plant with an emergency power supply in the event of a loss of offsite power. The Oconee design did not and does not include diesel-powered emergency generators, which are at every other U.S. nuclear power plants.

3 Thus, the Jocassee Dam and the Keowee Dam, as well as the lakes behind them, constitute an integral part of the Oconee nuclear power plant, including its backup emergency power supply.

2.2 Jocassee and Keowee Dam Characteristics Completed in 1971 and licensed by the Federal Energy Regulatory Commission (FERC), the Keowee Dam is a 170 foot-high rock-filled earthen dam about 3,500 feet in length. The Oconee nuclear power plant complex is built into the side of the dam, which contains two hydroelectric generators with a combined output of about 150 MW.2 These hydroelectric generators provide emergency power to Oconee.

The Keowee Dam lies about 14 miles downstream of the Jocassee Dam. It impounds about one million acre-feet of water and has a surface area of about 18,000 acres. The top of the dam is at 815 feet above mean sea level (MSL). Full pond or normal operating level of Keowee Lake is at 800 ft. Construction of the Keowee Dam was completed in 1971.

Completed in 19751 and also licensed by FERC, Jocassee Dam is a rock-filled earthen dam 385 feet high and about 1,000 feet long. It also impounds about a million acre-feet of water in the Jocassee Lake, with an area of 7,565 acres. The lakes pumped storage capability is supplied by four hydroelectric turbines that can be reversed to pump water from below the Jocassee Dam to above the dam.

The top of the Jocassee dam is at 1,125 ft. Full pond operating level of Jocassee Lake (i.e.,

normal operating level) is 1,110 ft.

2.3 Oconee Nuclear Plant Design and Construction 2.3.1 NRC Safety Requirements for Nuclear Plant Design and Construction All nuclear power plants constructed after 1973 are required to meet 10 Code of Federal Regulations (CFR) Part 50 Appendix A General Design Criteria for Nuclear Power Plants, including Criterion 2 - Design Bases for Protection Against Natural Phenomena. General Design Criterion (GDC) 2 states in part:

Structures, 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.

2 https://www.duke-energy.com/community/lakes/hydroelectric-relicensing/keowee-toxaway/keowee-toxaway-project.

4 While Oconee was built prior to 1973 and therefore was not required to meet GDC 2, it was required to meet a similar draft version of the criterion.3 2.3.2 Design and Construction of Oconee Units 1, 2 and 3 During initial Oconee licensing, Duke convinced the Atomic Energy Commission (the AEC was the predecessor to the NRC) that a Jocassee Dam failure was not credible. Duke has repeatedly stated that they believe a Jocassee Dam failure is not credible.4 Thus, at the earliest point of design and construction, the NRC did not require Duke to protect Oconee from a Jocassee Dam failure. For instance, the turbine building, located at a grade of 796 feet mean sea level MSL), houses portions of the emergency core cooling system (ECCS) and other safety related and important to safety systems, including the service water systems and the 4kv emergency buses. But the NRC did not require Duke to build the turbine building as a watertight structure. As a result, 2.3.3 Post-construction addition of Safe Shutdown Facility Sometime prior to 1983, in order to address other Oconee design weakness not related to flooding, Duke decided to install additional equipment to improve Oconees safety. Duke completed the installation of the safe shutdown facility (SSF) prior to 1983. The SSF is designed to address events including fire, sabotage, turbine building floods, station blackouts and tornado missile events. It contains a single diesel generator capable of supplying sufficient power only for the SSF equipment. It contains pumps capable of supply water to the steam generators and to the reactor coolant systems of all three units and a service water pump only 3 The pre-GDC 2 version for Oconee provided that:

Those systems and components of reactor facilities which are essential to the prevention of accidents which could affect the public health and safety or to mitigation of their consequences shall be designed, fabricated, and erected to performance standards that will enable the facility to withstand, without loss of the capability to protect the public, the additional forces that might be imposed by natural phenomena such as earthquakes, tornadoes, flooding conditions, winds, ice, and other local site effects. The design bases so established shall reflect: (a) appropriate consideration of the most severe of these natural phenomena that have been recorded for the site and the surrounding areas and (b) an appropriate margin for withstanding forces greater than those recorded to reflect uncertainties about the historical data and their suitability as a basis for design.

FOIA-2013-0239, Oconee Non-concurrence (ML13340A179), Applicable Regulatory Guidance, Page 2 4 2008.09.26, Duke Response to 50.54(f) Request (ML082750106), (2008 Duke 50.54(f) Response Letter) Attachment 2, Page 6. 2009.04.30, NRC Letter to Duke Evaluation of Duke Responses to NRC Letter Dated August 15, 2008, Related to External Flooding at Oconee (ML090570779), Page 2.

5 capable of cooling the SSF loads. None of this equipment is safety related, single failure proof or redundant. It is manually controlled and operated only locally from the SSF itself. The SSF is at a grade of 796 ft.5 because the SSF was not intended to be used for external floods, it was not protected from them.

2.4 Flood Risk Studies 2.4.1 1983 Flood Study for FERC and Construction of Wall around SSF Duke in a 1983 hydrological analysis determined as follows:

[T]he impacts of flooding from a postulated sunny-day failure of the Jocassee Dam. The results of the study indicated an estimated peak flood elevation of

) MSL at Keowee Dam, and a resulting ONS powerblock flood depth of

). In order to reduce the risk of flooding, the licensee erected walls around the entrances to the Standby Shutdown Facility (SSF) with average wall height of

(

). The construction of the walls was not part of the design-basis.6 Thus, by 1983 Duke recognized that external flooding was possible and that if Oconee experienced a flood above grade, the flood would incapacitate the ECCS. In that event, Duke would have no way to mitigate the flood.

2.4.2 NSAC-60 Probabilistic Risk Assessment In the late 1970s and early 1980s Duke initiated one of the first industry-conducted nuclear power plant probabilistic analyses (PRA). The study was prepared by the Nuclear Safety Analysis Center 7 and was called NSAC-60. NSAC-60 was a full-scope PRA, meaning it included both internal hazards such as loss of coolant accidents (LOCA) and external events such as earthquakes. It included an analysis of core damage frequency (referred to as a Level 1 analysis), containment failure frequency (referred to as a Level 2 analysis), and impacts on the surrounding population (referred to as a Level 3 analysis).

The NSAC-60 analysis included contributions to core melt frequency by failures of the Jocassee and Keowee Dams.8 As described by Duke, the study determined the failure frequency for the 5 2018.06.18 NRC Staff Assessment Related to Focused Evaluation for Oconee, Page 3 (ML18141A755).

6 2016.04.14 NRC Staff Assessment by the Office of NRR Related to flooding Hazard Reevaluation Report NTTF Recommendation 2.1 Oconee (ML16273A128).

7 NSAC initially was a separate legal entity, collocated with the Electric Power Research Institute (EPRI).

About 1990 it was folded into EPRI.

8 Nuclear Safety Analysis Center, NSAC-60, A Probabilistic Risk Assessment of Oconee Unit 3, June 1984.

6 Jocassee Dam by compiling data for dams with similar attributes. It considered three time periods and derived three median annual failure frequencies for causes other than earthquakes and overtopping:

1900 to 1981 2.3x10-5 per year

1940 to 1981 1.6x10-5 per year

1960 to 1981 1.4x10-5 per year9 2.4.3 IPE/IPEEE for Severe Accident Vulnerabilities In 1988 the NRC issued Generic Letter (GL) 88-20, requesting all reactor licensees submit a systematic examination in order to identify any plant-specific vulnerabilities to severe accidents and report the results to the Commission.10 Initially, GL-88-20 requested licensees to analyze only internal events such as loss of coolant accidents (LOCA) and transients. The NRC subsequently issued 5 revisions. Among other changes, the revisions, expanded the scope to include external events such as tornados, seismic events and external floods.

In response, in December 1990, Duke submitted an Individual Plant Examination (IPE) that evaluated internal events.11 In December 1995, Duke submitted an Individual Plant Examination for External Events (IPEEE) that evaluated external events.12 In 1997, in a nonpublic document, Duke updated the IPE and IPEEE and resubmitted the results.13 In the 1995 IPEEE, Duke considered whether and how to evaluate the risks of external flooding at Oconee. First, Duke considered evaluating the risk of a probable maximum precipitation (PMP) event at the Oconee site, i.e., a large storm in the direct vicinity of the plant. But Duke screened out a PMP event based on the large size of the reservoirs above the Keowee and Jocassee Dams.

Duke also considered whether to evaluate a Jocassee Dam failure in the IPEEE. In making this evaluation, Duke focused on three types of dam failures: seismic dam failure, random (i.e.,

sunny day) dam failure, and a dam failure caused by a PMP above the Jocassee Dam that overtopped the dam (i.e., a dam breach caused by water flowing over the top of the dam).

9 US NRC Information Notice 2012-02, Potentially Nonconservative Screening Value for Dam Failure Frequency in PRA, March 5, 2012, Page 2, (ML090510269).

10 1988.11.23, NRC Generic Letter 88-20, Individual Plant Examinations for Severe Accident Vulnerabilities.

11 1990.12, Duke IPE (nonpublic). As discussed above in my Note on Secrecy, the IPEE is cited here for purposes of identification. This report does not rely directly on the content of the IPEEE, or any other nonpublic document. When the content of the IPEEE or any other nonpublic document is described in this report, it is taken from descriptions in publicly available documents.

12 1995.12.21, Duke IPEEE (nonpublic).

13 1996.12, Duke Oconee Nuclear Station PRA Revision 2 Summary Report (ML080780111) (nonpublic).

7 The IPEEE found that a seismic failure of Jocassee Dam was a dominant contributor to the total Oconee CDF, and calculated the contribution to core damage frequency from a seismic failure of Jocassee at 7.2E-6 per year (i.e., 20% of the total seismic CDF of 3.6E-5).14 In evaluating a random or sunny day failure, the IPEEE found a CDF of 7.0E-6 per year.15 In making this estimate, Duke used a dam failure frequency of 1.3E-5 per year, an insignificant decrease from the values derived and used in NSAC-60. 16 With respect to a PMP-induced Jocassee Dam failure, Duke concluded that such a failure was not credible.17 Therefore, Duke did not evaluate a PMP-induced Jocassee Dam failure.

2.4.4 1992 Flood Study for FERC In 1992, Duke performed an inundation study to meet a FERC requirement for formulating an emergency action plan in the event that the Jocassee Dam failed. This study showed that approximately of water would inundate the yard area surrounding the SSF, thereby rendering inoperable Oconees all systems necessary to shut down and maintain all three units in a safe and stable condition.18 2.5 Initial Oconee License Renewal and Severe Accident Mitigation Alternatives Analysis In July 1998, Duke submitted a license renewal application to NRC, requesting an extension of the Oconee reactors licenses terms by 20 years. The NRC renewed Dukes licenses in May 2000.19 Dukes Environmental Report for the license renewal application included a Severe Accident Mitigation Alternatives (SAMA) analysis, containing a review of potential design alternatives along with any procedural, non-hardware, alternatives. 20 For its risk estimates, 14 1995.12.05, Oconee IPEEE Submittal Report (nonpublic). See also 2008 Duke 50.54(f) Response Letter; 1996.07.08, NRC Letter: Draft Reports Related to the Keowee Hydro Station Emergency Electrical System Supply to Oconee (ML15118A442). Total seismic CDF is 3.6E-5 per year (see Page 106) while 20% of this is from a Jocassee Dam failure (Page 107), i.e., 3.6E-5 x 0.2 = 7.2E-6 per year.

15 2000.03.15, NRC Letter: Oconee Review of IPEEE (ML003694349), Staff Evaluation at Page 2.

16 FOIA Response 2012-0325 at Page 17 of 308, (ML15156A702) (FOIA Response 2012-0325). See also 1996.07.08, NRC Letter: Draft Reports Related to the Keowee Hydro Station Emergency Electrical System Supply to Oconee (ML15118A442), at Page 110.

17 1996.07.08, NRC Letter: Draft Reports Related to the Keowee Hydro Station Emergency Electrical System Supply to Oconee (ML15118A442) Section 6.4.1, Page 110. See also FOIA Response 2012-0325.

18 While the inundation study is not a public document, the NRC described it in its 2011 NRC Safety Evaluation Letter.

19 2000.05.23, NRC Renews License of Oconee for an Additional 20 Years (ML003718834).

20 1998.04, Environmental Report, Application for Renewed Operating Licenses, Oconee Nuclear Station, Units 1, 2, and 3, Attachment K, Page 1 (1998 SAMA Analysis). This document appears to be the

8 the SAMA analysis relied on the IPE/IPEEE risk analyses, as well as a non-public revised IPE/IPEEE submitted in December of 1997, also referred to as Oconee PRA Revision 2 and Oconee PRA/IPE Revision 2.21 The SAMA analysis started with the total core damage frequency from the IPE/IPEEE of 8.9E-5 per year, with 2.6E-5 per year (29%) from internal events and 6.3E-5 per year (71%) from external events. The external events were broken down as follows:

CDF from External Events22 Frequency (per reactor-year)

Initiating Events Seismic 3.9E-05 Tornado 1.4E-05 External Flood 5.9E-06 Fire 4.5E-06 Total External 6.3E-05 The SAMA analysis considered flooding hazards from a Jocassee Dam failure, apparently in reliance on the NSAC-60 and IPEEE studies.23 The discussion about a Jocassee Dam failure describes it in the context of random failures.24 Based on this statement, it is reasonable to assume that Duke only considered random sunny-day dam failures, ignoring seismic and overtopping, failures. This approach of excluding seismic and overtopping-related dam failures was consistent with the IPEEE.

But the SAMA analysis differed from the IPEEE in the respect that it estimated the external flooding contribution at 5.9E-6 per year, whereas the IPEEE estimated the external flooding contribution at 7E-6. The SAMA analysis did not address or explain this difference.

The SAMA analysis evaluated two alternatives that would impact Jocassee Dam failure consequences. The first alternative was to staff the SSF continuously with a trained operator, and the second was to increase the height of the -

wall protecting the SSF from floods to

.25 But Duke determined these alternatives were not cost-effective.26 Duke also identified a third alternative: strengthening the Jocassee Dam and thus lowering the random document identified in Reference 2.8 of the Environmental Report (Page 31), although it is not clear.

Reference 2.8 is not a public document.

21 1998 SAMA Analysis, Pages 4, 9, 10.

22 1998 SAMA Analysis, Page 10.

23 1998 SAMA Analysis, Pages 7, 15, 19.

24 1998 SAMA Analysis, Page 15.

25 1998 SAMA Analysis, Page 16.

26 1998 SAMA Analysis, Page 28.

9 failure frequency. But Duke rejected this alternative without evaluating it, on the ground that the cost would far exceed the benefit of core damage frequency reduction.27 The NRC reviewed the SAMA analysis and concluded: Based on its review of SAMAs for ONS (Oconee Nuclear Station), the staff concurs that none of the candidate SAMAs are cost beneficial.28 This included the two evaluated alternatives addressing a Jocassee Dam failure.

2.6 Updated Dam Failure and Flood Routing Evaluations and Related Regulatory Actions 2.6.1 NRC 2006 Significance Determination Process on Oconee Flooding Issue In November 2006, the NRC completed a Significance Determination Process (SDP) evaluation related to a performance deficiency involving a missing covering in the wall protecting the SSF.29 NRC characterized the missing flood barrier as a violation and determined its significance as a White finding.30 After Dukes appeals of the finding, the NRC affirmed the finding.31 Dukes repeated appeals prompted the NRC to re-evaluate the flooding risk at Oconee from a Jocassee Dam failure. While Duke had previously estimated the dam failure rate in the range of 2.3E-5 to 1.4E-5 per year (NSAC-60) and had revised it to 1.3E-5 per year (IPEEE), the NRC found these estimates of failure frequency of the Jocassee dam were too low. In the SDP appeal process the NRC calculated a Jocassee Dam failure rate of 1.8E-4 per year.32 2.6.2 NRC 50.54(f) Letter In 2008, in light of its new understanding from the previously discussed SDP that the Jocassee Dam failure frequency was significantly larger than what Duke had previously represented, NRC issued Duke a 10CFR50.54(f) letter requesting additional information.33 First, the 50.54(f) letter 27 1998 SAMA Analysis, Page 15.

28 1999.12, Generic Environmental Impact Statement for License Renewal of Nuclear Plants Supplement 2, Regarding Oconee, NUREG-1437, Page 5-19.

29 2006.11.22, NRC Final Significance Determination for White Finding and Violation (ML063260282)

(2006 NRC White Finding). The SDP is part of the NRCs reactor oversight process (ROP). The ROP is the NRCs program to inspect, measure and assess the safety performance of operating plants. The SDP is the NRCs process for assessing the significance of findings identified in the ROP.

30 2006 White Finding, at Page 1.

31 2007.11.20, NRC Reconsideration of Final Significance Determination Associated with SSF Flood Barrier White Finding, (ML073241045) (2007 NRC Reconsideration of Significance Determination).

32 2007 NRC Reconsideration of Significance Determination, Page 1.

33 2008.08.15, NRC letter to Duke: Information Request Pursuant to 10CFR50.54(f) Related to External Flooding Including Failure of the Jocassee Dam at Oconee (ML081640244) (2008 NRC 50.54(f) Letter).

10 laid out the regulatory requirements applicable to Oconee, and described the status of Dukes flood protection measures:

Section 3.1.2 of the UFSAR, Criterion 2 - Performance Standards (Category A), states, Those systems and components of reactor facilities which are essential to the prevention of accidents which could affect public health and safety or to mitigation of their consequences shall be designed, fabricated and erected to performance standards that will enable the facility to withstand, without loss of the capability to protect the public, the additional forces that might be imposed by natural phenomena such as earthquakes, tornadoes, flooding conditions, winds, ice, and other local site effects.

The current UFSAR discusses walls that are used for flood protection at the SSF.

However, it does not include the effects of a Jocassee Dam failure, nor does it include the flood protection features to mitigate the consequences of such an event. We further note that in the mid-1990s, the UFSAR was revised by removing the reference to the Jocassee Dam failure and postulated wave height of in the yard at the Oconee site.34 The letter also references the flood heights calculated from the 1992 FERC analysis. This letter characterizes the 1992 FERC analysis as

,35 while the previous discussion of the FERC analysis characterized the same analysis as having a flood height.36 In addition, the NRCs letter requested Duke to address three specific issues:

1) Explain the bounding external flood hazard at Oconee and the basis for excluding consideration of other external flood hazards, such as those described in the Inundation Study, as the bounding case.

2) Provide your assessment of the Inundation Study (the 1992 study conducted for FERC) and why it does or does not represent the expected flood height following a Jocassee Dam failure.

3) Describe in detail the nuclear safety implications of floods that render unavailable the SSF and associated support equipment with a concurrent loss of all Alternating Current power.37 In subsequent discussions with Duke, the NRC compared the Jocassee Dam hazard with other hazards considered in the design and licensing basis. It observed that a Jocassee Dam failure frequency of about 2E-4 per year was less than the hazard from general transients, losses of 34 2008 NRC 50.54(f) Letter, Page 1, 9 (emphasis added).

35 2008 NRC 50.54(f) Letter, Page 2.

36 2011 Safety Evaluation Letter, Safety Evaluation, Page 1.

37 2008 NRC 50.54(f) Letter, Page 2.

11 offsite power, etc., but greater than the hazard from medium and large break LOCAs (see Figure 1 below).38 It should be noted here that for all of the other hazards listed, Oconee -- as well as every other US nuclear power plant -- is required to have safety grade, fully redundant, single failure proof ECCS capable of responding. For the Jocassee Dam failure, Oconee had the SSF which is non-safety grade, has no redundancy, is not single failure proof and is not part of the ECCS. Even if the original Jocassee Dam failure rate of 1.3E-5 was correct, this is still an order of magnitude greater than the large LOCA rate of 2E-6 per year which is in the design basis and requires the ECCS to protect the public. At this point in time (2008) the SSF was protected from a Jocassee Dam failure by a wall that Duke from its previous analysis knew was inadequate because the hydraulic analysis showed that there was a potential for over of water at the SSF.

Figure 1 Oconee Hazard (or Initiating Event) Frequency Comparison Credible Events 39 38 2008.08.28, NRC Presentation Oconee Flood Protection and Jocassee Dam Hazard (2008 NRC Presentation Oconee), Slide 8 (ML082550290).

39 2008 NRC Presentation Oconee), Slide 8.

12 2.6.3 Dukes Response to 50.54(f) Letter Duke responded to NRCs 50.54(f) letter that: Duke considers a random sunny day failure of the Jocassee dam not credible because of the nature of its design, its construction, the inspections conducted during its construction, and those periodic inspections that have occurred, and continue to occur, since its construction.40 Duke further argued that the higher flood elevations posited by NRC in the 50.54(f) letter were not applicable to Oconee, because they came from the 1992 study Duke had conducted for FERC to establish an Emergency Action Plan (EAP) for the population downstream of Jocassee, and thus was intended to provide a worst case analysis rather than credible flood levels.41 After considerable discussion with Duke, the NRC sent a letter in the spring of 2009. This letter states in part: The NRC staffs position is that a Jocassee Dam failure is a credible event and needs to be addressed deterministically.42 The letter clearly articulates that the NRC is concerned about adequate protection. For example, it states: When the inundation study and sensitivity analyses are completed, the NRC staff will evaluate the results to determine whether further regulatory actions are necessary to ensure there is adequate protection against external flooding at Oconee.43 Finally, the NRC states its expectation of receiving analyses which would establish an adequate licensing basis for external flooding... 44 In response to the NRCs concerns, and after further analysis, Duke decided to raise the wall height protecting the SSF by to a total height of

. It completed this work in February of 2009.45 Duke also responded to the NRCs inquiries by performing an additional hydrological analysis of the failure of Jocassee Dam and propagating the resulting flood onto the Keowee Lake and Dam and then onto Oconee. Building on the model in the 1992 study for FERC, Duke modified it and increased the level of detail. Duke reported its preliminary results to the NRC in a presentation on October 28, 2009.46 Duke had expected the flood heights to decrease by using the new 40 2008.09.26, Duke Letter in Response to 10CFR50.54(f) Request, Attachment 2 Page 3 (ML082750106)

(2008 Duke 50.54(f) Response Letter).

41 2008 Duke 50.54(f) Response Letter, Attachment 2 Page 3.

42 2009.04.09, NRC letter to Duke Evaluation of Duke September 26, 2008, Response to NRC Letter Dated August 15, 2008, Related to External Flooding at Oconee (ML090570779), Page 2 (2009 NRC External Flooding Letter).

43 2009 NRC External Flooding Letter, Page 3.

44 2009 NRC External Flooding Letter, Page 3.

45 2009.05.11, Duke Presentation on Oconee External Flood (ML091380424).

46 2009.10.28, Duke Presentation on Oconee External Flood with Initial HEC-RAS Results (ML093080034)

(2009 Duke Presentation with Initial HEC-RAS Results).

13 software and model. However, flood heights increased. The new model, using a conservative but not worst-case scenario, predicted a flood height at the SSF of about

.47 To resolve this adequate protection issue, NRC required Duke to re-perform the Jocassee Dam failure analysis using conservative input parameters (i.e., assumptions) and methods.48 In response, Duke revised its 1D and 2D analysis. And Duke committed to protecting the SSF based on results from its revised analysis.49 Protective measures would include increasing the height of the flood barriers protecting the site, protecting an offsite power line from the expected flood conditions and other improvements.50 2.6.4 2011 NRC Safety Evaluation In January of 2011, the NRC transmitted to Duke a Safety Evaluation confirming Dukes approach to the issue resolution. This safety evaluation concluded:

The NRC staff evaluated the information provided by Duke in their August 2, 2010, letter. The unmitigated Case 2 dam breach parameters that were used in the flooding models, provided by Duke for the ONS site, demonstrated that the licensee has included conservatisms of the parameters utilized in the dam breach scenario. These conservatisms provide the staff with additional assurance that the above Case 2 scenario will bound the inundation at ONS, therefore providing reasonable assurance for the overall flooding scenario at the site. This new flooding scenario is based on a random sunny-day failure of the Jocassee Dam. This Case 2 scenario will be the new flooding basis for the site.51 The NRCs Safety Evaluation required Duke to protect the Oconee site from random sunny day failures of the Jocassee Dam to a flood depth of in order to ensure adequate protection.52 The requirement was based on conservative deterministic analysis.53 However, the Safety Evaluation was silent to other relevant Jocassee Dam failure mechanisms including seismic and overtopping even though these mechanisms had been constantly discussed both internally within the NRC and with Duke.

47 2009 Duke Presentation with Initial HEC-RAS Results, Slide 26.

48 2010.01.29, NRC letter Evaluation of Duke Response to Related to External Flooding at Oconee (ML100271591), See Enclosure. Using a conservative approach would supply margin and account for uncertainty, and is the norm for design basis and licensing basis issues -- which this adequate protection issue had become.

49 2009 Duke Presentation with Initial HEC-RAS Results. Duke had presented examples of these preliminary results in its previous meeting with NRC.

50 2010.11.29, Duke Letter: Oconee Response to CAL, Page 2 (ML103490330).

51 2011 NRC Safety Evaluation Letter.

52 2011 NRC Safety Evaluation Letter, Safety Evaluation at Page 12.

53 2011 NRC Safety Evaluation Letter.

14 In 2010, NRC finalized its own generic dam failure frequency analysis for the Jocassee Dam. 54 The staff estimated generic dam failure rates for large rock-fill dams, which it considers applicable to the Jocassee Dam, as 2.8E-4 per year.55 The authors of that analysis and other members of the NRC Staff subsequently performed additional analyses exploring and confirming those results.56 In 2012, because the demonstrably erroneous NSAC-60 dam failure rate was widely referenced and used throughout the nuclear industry at that time, the NRC issued an information notice warning of the inadequacies in the dam failure rate found in the NSAC-60 report.57 According to Information Notice 2012-02, NSAC-60 provide(d) an insufficient basis for estimating site-specific dam failure frequency.58 In 2019, in a more detailed study commissioned by the NRC, Oak Ridge National Laboratory (ORNL) stated:

NRC has estimated the likelihood of failure of Jocassee dam, upstream of the Oconee Nuclear Station in South Carolina, at approximately 2.8x10-4 per year. This estimate aligns with historical dam failure rates found in literature.59 In addition, Dukes own risk analysis calculated that the SSF had a failure probability of about 0.27 or 27%.60 This is a very high failure probability, orders of magnitude greater than the failure probability estimated by Duke for safety related equipment found in the ECCS.

Thus, the outcome of the multi-year NRC safety evaluation was to increase the flood protection from a Jocassee Dam failure from approximately f to a new licensing basis height of about 54 2010.03.15, Generic Failure Rate Evaluation for Jocassee Dam (ML13039A084). The NRCs generic analysis was published internally and subsequently released via a Freedom of Information Act request.

(2010 Generic Failure Rate Evaluation for Jocassee Dam).

55 2010 Generic Failure Rate Evaluation for Jocassee Dam, Page 6.

56 2013.07.17, Uncertainty Analysis for Large Dam Failure Frequencies Based on Historical Data (ML13198A170); Ferrante, et al., An Assessment of Large Dam Failure Frequencies Based on Us Historical Data ANS PSA 2011 International Topical Meeting on Probabilistic Safety Assessment and Analysis, March 13-17, 2011, Wilmington, NC, USA.

57 NRC Information Notice 2012-2.

58 NRC Information Notice 2012-02, Page 4.

59 2019.12.14, Current State-of-Practice in Dam Safety Risk Assessment, https://www.osti.gov/servlets/purl/1592163/.

60 FOIA Response 2012-0325 at Pages 110, 115 of 308.

15 Ten years later, the 2011 Safety Evaluation and the safety requirements it imposed remain in effect. Duke has not appealed the 2011 Safety Evaluation, nor has the NRC retracted or repudiated it. Yet, there is no record that Duke has completed the required modifications to protect the plant to a flood depth of Nor has the NRC sought to ensure its completion.

2.7 Fukushima - Lessons Learned 50.54(f) Letter and Staff Assessment In 2011, the Fukushima Dai-chi disaster occurred, with waves as high as 45 feet, leading to core damage and containment failures at three of the six nuclear power plants on the site. A year later, the NRC issued 10CFR50.54(f) letters to all licensees, requesting them to reevaluate the flooding hazards at their sites against present-day regulatory guidance and methodologies being used for early site permits and combined license reviews.61 After Duke submitted responses in 2013 and 2015, the NRC issued a Staff Assessment.62 By titling the document a Staff Assessment rather than a Safety Evaluation, the NRC Staff indicated that the document did not have the regulatory equivalence of safety findings. And indeed, the conclusions of the Staff Assessment do not measure Dukes submittal against an NRC safety standard of reasonable assurance of adequate protection or no undue risk.

Instead, the Staff measured Dukes submittal against a reasonableness standard. The Staff, for instance, found that [s]eismically-induced failure of the Jocassee Dam is not a reasonable mode of failure based on current information, present-day methodologies and regulatory guidance.63 Similarly, the Staff found that [o]vertopping-induced failure of the Jocassee Dam is not reasonable model of failure based on current information, present-day methodologies and regulatory guidance. 64 The NRC also approved Dukes conclusion that a random sunny-day failure was an unlikely, although reasonable, failure mode.65 Because the Staff Assessment did not repudiate or even address the 2011 Safety Evaluations issues, because it applied the distinctly different and weaker (albeit undefined) standard of reasonableness rather than adequate protection, and because it did not even purport to be a 61 2012.03.12 Letter from NRC to all Power Reactor Licensees and Construction Permit Holders re:

REQUEST FOR INFORMATION PURSUANT TO TITLE 10 OF THE CODE OF FEDERAL REGULATIONS 50.54(f)

REGARDING RECOMMENDATIONS 2.1.2.3, AND 9.3, OF THE NEAR-TERM TASK FORCE REVIEW OF INSIGHTS FROM THE FUKUSHIMA DAI-ICHI ACCIDENT (ML12053A340), Enclosure 2, Page 1 (NRC Post-Fukushima 50.54(f) Letter).

62 2016.04.14, NRC letter re Oconee Staff Assessment of Response to Request for Information Pursuant to 50.54(f) Flood-Causing Mechanisms Reevaluation (ML15352A207), enclosing NRC Staff Assessment by the Office of NRR Related to flooding Hazard Reevaluation Report NTTF Recommendation 2.1 (M16273A128) (2016 NRC Letter re 50.54(f) Response). A redacted version of this document was released in Interim Response 3 to FOIA-2018-0010 on October 26, 2017.

63 2016 NRC Letter re 50.54 Response, Enclosure 2, Page 3.

64 2016 NRC Letter re 50.54 Response, Enclosure 2, Page 3.

65 2016 NRC Letter re 50.54 Response, Enclosure 2, Page 3.

16 Safety Evaluation, the Staff Assessment cannot be compared to the 2011 Safety Evaluation or presumed to override it in any way.

2.8 Dukes 2021 Subsequent License Renewal Application and SAMA Analysis In March 2021, Duke submitted a subsequent license renewal application to NRC, requesting an extension of each of the three Oconee reactors operating licenses terms by an additional 20 years. Like Dukes initial license renewal application in 1998, Dukes SLR application relied on its PRA to look for insights into whether there was new and significant information that would provide a significantly different picture of the impacts from severe accidents during the second license renewal period.66 Duke first discussed the question of whether it had new and significant information regarding design-basis accidents, and concluded that it identified no new and significant information that would change the conclusion of the 2013 Revised License Renewal Generic Environmental Impact Statement (GEIS) that impacts of design-basis accidents are not significant because a licensee is required to maintain the plant within acceptable design and performance criteria, including during any license renewal term.67 Next, Duke discussed the question of whether it had new and significant information about severe accident impacts.68 According to Duke, [p]eriodic updates to the Oconee PRA have ensured that the PRA includes relevant changes to the plant design, operation and maintenance practices. In addition, PRA updates include updates to the plant-specific initiating events and equipment data utilized, and improvements in state-of-the-art analysis of severe accidents.69 Duke also asserted that it had considered developments since the initial license renewal decision which could have changed the assumptions made concerning severe accident consequences after SAMAs were previously evaluated.70 Duke then provides a list of six areas of developments that it reviewed, including internal events, external events, and risk-beneficial changes in response to the NRCs Fukushima Daiichi Near Term Task Force 66 Oconee Environmental Report, Section 4.15, Severe Accident Mitigation Alternatives Analysis (2021 SAMA Analysis, Page 4-75.

67 2021 SAMA Analysis, Page 4-78, citing Generic Environmental Impact Statement for License Renewal, Rev. 1 (NUREG-1437, 2013).

68 2021 SAMA Analysis, Page 4-75. In this context, Duke stated that it interprets the term significance to relate to both the cost-effectiveness of SAMAs and their potential to significantly reduce risk to the public. Id.

69 2021 SAMA Analysis, Page 4-74.

70 2021 SAMA Analysis, Page 4-76.

17 (NTTF).71 According to Duke, [n]o new and significant information was determined for any of these areas.72 Duke provided additional explanation with respect to each of these three areas. With respect to internal events, Duke characterized its updated PRA results as decreasing total CDF for internal events by a small 8% to a value of 2.4E-5 per year.73 But Duke did not provide any references for this decreased risk estimate.

With respect to external events, Duke asserted that ONS fire, seismic, high winds, and external flood PRA models have been developed and have been utilized in the quantitative PRA calculation that demonstrated the absence of any potentially significant SAMAs.74 But Duke supplied no quantitative information regarding the value of the assertedly insignificant change in the external events CDF, nor did it cite any references.

With respect to risk-beneficial changes in response to the NRCs Fukushima Daiichi NTTF, Duke asserted that changes it has made in response to the NTTF recommendations have not been credited in ONS PRA models.75 Thus, according to Duke, no further analysis is needed of the NRCs conclusion in the NRCs 2013 Revised License Renewal GEIS that the probability-weighted consequences from severe accidents are small and no further analysis is needed.76 Duke also described its methodology for evaluating whether new and significant information existed that would affect its SAMA analysis. According to Duke, it looked for changes such as identification of new hazards and updated plant risk models using as an example the fire PRA that replaces the IPEEE analysis.77 Duke also asserted that it determined which changes were significant by using the internal and external Oconee PRA. In addition, the Oconee fire, seismic, external flood and high wind models are capable of determining impacts to the CDF and (large early release frequency) LERF.78 According to Duke, the SAMA analysis evaluated 283 industry SAMAs. All but 45 were qualitatively screened out. ONS-specific SAMAs further were reviewed to determine if they are 71 2021 SAMA Analysis, Page 4-76.

72 2021 SAMA Analysis, Page 4-76.

73 2021 SAMA Analysis, Page 4-77.

74 2021 SAMA Analysis, Page 4-77.

75 2021 SAMA Analysis, Page 4-78.

76 2021 SAMA Analysis, Page 4-78, citing Generic Environmental Impact Statement for License Renewal, Rev. 1 (NUREG-1437, 2013).

77 2021 SAMA Analysis, Page 4-79.

78 2021 SAMA Analysis, Page 4-80.

18 still applicable.79 Ultimately, all SAMAs were screened out either qualitatively or quantitatively, and therefore the Level 3 PRA was not updated.80

3. ANALYSIS Dukes operating licenses for Oconee Units 1, 2 and 3 will expire in 2033 (Units 1 and 2) and 2034 (Unit 3) unless the NRC approves Dukes SLR application. In that case, Duke will be allowed to operate Oconee until 2053 (Units 1 and 2) and 2054 (Unit 3). In my expert opinion as a nuclear engineer and risk analyst, Duke is now operating Oconee at an unacceptable risk to public health and safety, due to its failure to fully implement flood-protective measures required by NRC in its 2011 Safety Evaluation. The NRC deemed those flood protection measures to be necessary to protect against a core melt accident in the event the Oconee site becomes inundated by failure of the Jocassee Dam. While the NRC has not sought to force Duke to implement those measures, neither has it withdrawn or repudiated the 2011 Safety Evaluation in which it found those measures were necessary to provide adequate protection to public health and safety.

In my years of experience as a NRC safety regulator, this is one of the most serious safety issues I have encountered. Yet, it is my understanding that NRC regulations for license renewal exclude it from the scope of safety issues that may be reviewed, because it does not relate to the aging of Oconees safety equipment. However, the NRC must also review Dukes SLR application under the National Environmental Policy Act (NEPA), which requires NRC to fully evaluate the environmental impacts of its proposed actions, including the environmental impacts of reasonably foreseeable accidents. Duke must also evaluate the relative costs and benefits of Severe Accident Mitigation Alternatives. Therefore, I have applied my expert skills as a risk analyst to evaluate whether Duke has taken into account all relevant data regarding the likelihood and consequences of a core melt accident caused by failure of the Jocassee Dam. I have also evaluated the adequacy of Dukes SAMA analysis to consider all relevant data.

Evaluation of accidents under NEPA, including SAMA analysis, requires the evaluation of the frequency of severe accidents, the consequences of those severe accidents and the evaluation of potential cost-effective mitigation strategies to deal with those consequences.

Level 1 PRA is used to evaluate the frequency of severe accidents while Level 2 and 3 PRA is used to evaluate the consequences. To perform the Level 1 analysis the basic PRA Equation is used:

CDF (/yr.)

=

IEF (/yr.)

X CCDP 81

[Eq. 1]

79 2021 SAMA Analysis, Page 4-81.

80 2021 SAMA Analysis, Page 4-79.

81 Oconee Nuclear Site Adequate Protection Backfit Documented Evaluation (circa 2010), Page 6 (ML14058A015).

19 Where CDF is the core damage frequency (in events per year), IEF is the initiating event frequency (in events per year) and CCDP is the conditional core damage probability (all probabilities are unit-less). PRA is always intended to be a best estimate analysis.

Typical PRA projects start with evaluation of IEF. In the case of external flooding, a thorough analysis would include flooding from all sources. Each hazard (e.g., local intense storms, dam failures, etc.) would be characterized with a hazard curve that supplies a range of intensities (e.g., flood height and flood inundation timing) and the corresponding frequency (in some reports it is characterized as annual exceedance probabilities). An example of a flooding hazard curve is shown in Figure 2.

Figure 2 Best Estimate and Approximate 90% Uncertainty Bounds of Peak River Level on the Kankakee River at the Nuclear Plant Site 82 Duke supplied no flooding hazard information in its SAMA analysis. It simply referred back to the 1998 SAMA, which in turn refers back to the IPEEE. The 1998 SAMA, however, supplied a single value, in contrast to more detailed example hazard curve illustrated in Figure 2. The single value supplied is for a Jocassee Dam failure with a rate of 1.3E-5 per year.83 That is the only information supplied by Duke about flooding initiating events. But this one data point is insufficient information to obtain any insights from the likelihood of dam failure events.

82 2014.08 EPRI Riverine Probabilistic Flooding Hazard Analysis, Figure 8-10, Page 8-9 (3002003013).

83 FOIA Response 2012-0325 at Page 17 of 308.

20 Equally important, the limited initiating event information provided in Dukes SAMA analysis is wrong. While Duke presents a Jocassee Dam failure rate of 1.3E-5 per year, NRC calculated a best generic failure rate for Jocassee of 2.8E-4 per year - more than twenty times greater.84 This information is well-known to Duke, because NRC shared its conclusions with Duke in 2008 and followed up with an Information Notice to alert the industry in 2012.85 This differs from the Duke value by over a factor of 20.

A middle ground between a single point estimate and a comprehensive analysis for each flood hazard would derive a range of flood hazards that would explore the possible spread of risks.

Duke could have (but did not) evaluate the range that would capture the spread of postulated outcomes as follows:

1. Flood depths that do not come above grade have the least impact and there is the probability that much of the ECCS will be available for mitigation.

2. Flood depths that come above grade but stay below the top of the protective SSF wall have an intermediate mitigation impact as the ECCS will be incapacitated but SSF should survive and then assigned a random failure probability based on the best available equipment database.

3. Flood depths that rise above the SSF wall have the most severe impact as these floods incapacitate all permanently install mitigation equipment.

After deriving IEF information, thorough PRA practices evaluate a range of mitigating capabilities for each and every previously identified initiating event sequence. In PRA terminology these mitigating strategies are characterized as conditional core damage probabilities (CCDP). The CCDP evaluates the probabilities of each combination of equipment available to cope with the associated hazard. For example, for a large LOCA the associated CCDP would evaluate the failure probabilities of both division of low pressure injection (LPI). It would cover all the combinations that would fail both trains. A few examples combinations that would fail both trains of LPI are:

LPI Train 1 Fails LPI Train 2 Fails Train 1 pump fails Train 2 injection valve fails Train 1 pump motor fails Train 2 pump suction valve fails Train 1 power fails Train 2 injection valve fails If Duke had evaluated the three initiating event scenarios described above, it would have derived CCDPs for each scenario. For the first scenario, with Jocassee flooding below grade, 84 Generic Failure Rate Evaluation for Jocassee Dam.

85 NRC Information Notice 2012-2.

21 Duke could have evaluated the failure probability of the ECCS, the SSF and any other equipment that might be available. For the middle scenario where the flood waters come above grade but not to the top of the SSF wall, the ECCS fails (and is given a failure probability of 1.0), but the SSF would not be incapacitated by the flood and thus would be assigned a random failure rate based on historical data. In the final scenario where the flood water come above the SSF wall, the SSF also fails and it would be given a failure probability of 1.0.

Neither Dukes 2021 SAMA analysis nor its 1998 SAMA analysis supplied any information about mitigating equipment failure probabilities. In fact, neither SAMA analysis supplies any CCDP information at all.

However, a minimal amount of CCDP information can be extracted from the limited amount of information that Duke supplied. Equation 1 from above (reproduced below) can be used as a starting point to extract the composite CCDP.

CDF (/yr.)

=

IEF (/yr.)

X CCDP

[Eq. 1]

Solving for the CCDP gives us Equation 2:

CCDP

= CDF (/yr.)

/

IEF (/yr.)

[Eq. 2]

From the 1998 SAMA analysis, we know that Duke used a flooding external event IEF value of 1.3E-586 per year. The corresponding external event flooding CDF is also supplied by the 1998 SAMA analysis in the table reproduced below:

CDF from External Events 87 Frequency (per reactor-year)

Initiating Events Seismic 3.9E-05 Tornado 1.4E-05 External Flood 5.9E-06 Fire 4.5E-06 Total External 6.3E-05 Plugging the external flooding IEF and CDF into Equation 2 allows us to find the associated CCDP:

CCDP

= CDF (/yr.)

/

IEF (/yr.)

[Eq. 3]

4.5E-1

=

5.9E-6

/

1.3E-5 86 FOIA Response 2012-0325 at Page 17 of 308.

87 1998 SAMA Analysis, Page 10.

22 Thus, Dukes CCDP for external flooding is 4.5E-1.

If we assume this composite CCDP is correct, we can calculate a corrected best estimate CDF for external flooding events using this CCDP and the NRCs best estimate IEF of 2.8E-4 per year and Equation 1.88 CDF (/yr.)

=

IEF (/yr.)

X CCDP

[Eq. 4]

1.3E-4

=

2.8E-4 X

4.5E-1 Thus, a corrected external flooding event CDF has value of 1.3E-4 per year, which is more than 20 times higher than Dukes wrong value of 5.9E-6 per year. It should be noted that the data used as input into the NRCs generic Jocassee Dam failure rate calculation does include failures from seismic and overtopping. See Section 2.7. Thus, my calculation includes seismic and overtopping contributions.

But the CDF of 1.3E-4 per year assumes that the CCDP of 4.5E-1 derived from the Duke analysis is appropriate. However, in 2008, Duke told the NRC that based on the 1992 inundation study, if the dam fails:

[T]he predicted flood would reach ONS in approximately

, at which time the SSF walls are overtopped. The SSF is assumed to fail, with no time delay, following the flood level exceeding the height of the SSF wall. The failure scenario results are predicted such that core damage occurs in about following the dam break and containment failure in about

. When containment failure occurs, significant dose to the public would result.89 Hidden in this statement is the fact that even Duke believes that if the SSF walls are overtopped, all mitigation fails, including the SSF -- thus resulting in core damage and containment failure. In other words, Duke is saying that the conditional core damage probability (CCDP) given a Jocassee Dam failure which overtops the SSF wall is a given, or has a value of 1.0, not the value of 4.5E-1. If we use this CCDP, i.e., a value of 1.0 then the CDF from a Jocassee Dam failure is equal to the Jocassee Dam failure rate or from Equation 1:

CDF (/yr.)

=

IEF (/yr.)

X CCDP

[Eq. 5]

2.8E-4

=

2.8E-4 X

1.0 Revisiting the Jocassee Dam failure rate, we can compare it to other initiating events. The NRC calculated a Jocassee Dam failure rate of 2.8E-4 per year. This value is in the range of LOCAs.

88 Generic Failure Rate Evaluation for Jocassee Dam.

89 2008 Duke 50.54(f) Response Letter, Attachment 2.

23 Even the Duke value of 1.3E-5 per year is larger than the value for large LOCA (see Figure 1 above).

Therefore, a reasonable best estimate CDF from a Jocassee Dam failure is 2.8E-4 per year based on the available PRA information, i.e., information supplied by Duke and NRC. This CDF is larger than the total CDF from all Oconee internal events of 2.4E-5 per year.90 It is also larger than Dukes estimate of 6.3E-5 per year for all external events.91 In fact, the CDF from Jocassee Dam failure is greater than the sum all Oconee internal and external events of 8.7E-5 per year reported by Duke.

In addition, Duke has ignored the risk contribution from shutdown operations. It is widely understood in the nuclear industry and by the NRC that the risks from shutdown are comparable to those during power operations. But this factor is not addressed in Dukes environmental analysis. Again, this omission significantly undercuts the credibility of the risk analysis.

Like the 2021 SAMA CDF information, the 2021 SAMA analysis supplied almost no information on the large early release frequency (LERF) analysis. But again, we can use the information supplied by Duke elsewhere. In the same quote from the 2008 Duke letter that is provided above, Duke also supplied relevant information about LERF:

(T)he predicted flood would reach ONS in approximately

, at which time the SSF walls are overtopped. The SSF is assumed to fail, with no time delay, following the flood level exceeding the height of the SSF wall. The failure scenario results are predicted such that core damage occurs in about following the dam break and containment failure in about

. When containment failure occurs, significant dose to the public would result.92 It is important to note that Duke presented containment failure as inevitable after the SSF walls are overtopped. Duke did not say the containment might fail, nor did it estimate the probability of containment failure. Duke is telling the NRC that the conditional failure probability of containment given a flood induced core damage event is 1.0. This is PRA language for a LERF multiplier of 1.0. Multiplying the CDF by the LERF multiplier gives us the LERF. With a LERF multiplier of 1.0, the LERF is equal to the CDF.93 Thus, not only is the CDF from an external flooding event 2.8E-4 per year but the LERF from an external flooding event is 2.8E-4 per year.

90 2021 SAMA Analysis, Page 4-77.

91 1998 SAMA Analysis, Page 10.

92 2008 NRC 50.54(f) Letter (emphasis added).

93 2013.09.23 NRC letter, NMP1 Integrated Inspection Report and Preliminary Greater than Green.

Finding, Page A-8 (ML13266A237).

24 All of the preceding impact discussion is based on Dukes 2008 conclusion of core damage in and a flood height at the SSF between which comes from the 1992 inundation analysis performed for FERC.94 However, the NRC required Duke to perform a new dam failure and flood routing analyses. Dukes new analysis increased the flood height at the SSF to about

.95 It is helpful to put these flooding results into perspective. Dukes August 2010 analysis indicated a peak flow across the Keowee Dam and significantly onto the Oconee site, of between 2.3 and 2.8 million cubic feet per second (cfs) and a peak flow across the Oconee intake canal structure of between 0.7 and 0.8 million cfs.96 As a point of reference, the average flow of the Mississippi River at New Orleans is approximately 0.6 million cfs.97 The 2010 Duke analysis also tells us that the Keowee Dam is overtopped to an elevation between msl.98 Bear in mind that the top of the Keowee Dam and the intake dike are at 815 feet msl, thus the dam is overtopped by some 99 This is a lot of water on the Oconee site, a site that was never designed to handle any water on site.

These significantly higher CDFs and LERFs would lead to significantly higher risks to the public and the environment. Yet, there is no evidence that Dukes 2021 Environmental Report has considered this new and significant flooding hazard information, the information from the more current dam failure and flood routing study that concluded with the flood depth or how this would impact the corresponding CDFs or LERFs. Nor has it considered the significant uncertainty on the timing, flood heights and flows, which should be part of any thorough risk assessment.

Furthermore, Dukes SAMA analysis does not reflect any consideration of the extensive work done to incorporate the Jocassee Dam failure and flood routing analysis, even though this work has supplied significant insights into possible additional severe accident mitigating strategies.

For instance, although the NRC required significant flood control measures in the 2011 Safety Evaluation, Duke does not mention them at all - either to take credit for them or, if they have not been installed, to explain why not. Duke has also failed to mention some other obvious ways to reduce the flood hazard from Oconee, such as preemptively shutting down the reactors when reservoir water levels get too high, lowering the water levels in the lake behind the Jocassee and Keowee Dams, or lowering the crest elevation of some of the surround earthworks such that they overtop before the Jocassee Dam, thus lowering the flood impacts at 94 2008 NRC 50.54(f) Letter, Page 1.

95 2011 NRC Safety Evaluation Letter, Page 12.

96 2010.08.02 Duke letter Oconee Response to CAL, Attachment 1, Table 1, Page 4 (ML102170006)

(2010.08.02 Duke Oconee Response to CAL).

97 National Park Service, Mississippi River Facts, https://www.nps.gov/miss/riverfacts.htm.

98 2010.08.02 Duke Oconee Response to CAL, Attachment 1, Table 2, Page 9.

99 2011 NRC Safety Evaluation Letter, Page 12.

25 ONS. PRA is a valuable tool for identifying vulnerabilities (and suggesting associated corrective measures), evaluating the costs and benefits of these measures, and also prioritizing them for their effectiveness. Unfortunately, the public has not benefited from a thorough and comprehensive external events flooding PRA.

Another significant shortcoming of Dukes risk analysis is Dukes failure to consider other Jocassee Dam failure mechanisms besides random sunny-day failures. Duke ignores seismic failures and overtopping failures, although they are both comparable contributors to public and environmental risk. Seismic failure could cause the dam to fail faster and overtopping failures would include additional water volumes behind the Jocassee Dam and potentially the Keowee Dam both scenarios could increase the flood volumes and heights at Oconee.

Therefore, not only has Dukes Environmental Report failed to address new and significant information, of which it is fully aware and which significantly bears on its environmental impact analysis and SAMA analysis but it has failed to correct probability estimates that are demonstrably wrong. Duke should be required to update its Environmental Report, taking in the new and significant information that significantly affects its previous conclusions that the environmental impacts of renewing the Oconee license are insignificant and that no cost-effective mitigative measures exist.

4.

CONCLUSION The history of the NRCs regulation of the Oconee reactors presents grave concerns in several significant respects.

First and foremost, from a regulatory perspective, it is unacceptable that the NRC has allowed Duke to operate for the past ten years without completing flood protection measures that NRC required ten years ago in 2011 to protect the public from the undue risk of a core melt accident caused by failure of the Jocassee Dam.

Second, the NRCs silence on this matter for the past ten years is inexcusable. The NRC should stand by its judgment, which it has never repudiated or withdrawn, that protection of public health and safety requires installation of substantial additional flood protection measures.

Finally, Duke has consistently downplayed the severity of the risk posed by the Jocassee Dam, to the point that it now seeks approval of a second license term for its three Oconee reactors, based on flood risk estimates that are demonstrably incorrect, incomplete, and poorly conducted. Duke has ignored data in its own possession showing that the risk of a core melt accident with subsequent containment failure caused by Jocassee Dam failure is significantly higher than Duke asserts. Duke has also ignored significant additional contributors to core damage frequency, including seismically induced dam failure, overtopping, and outages. Of course, climate change will only make the flood results and effects worse.

26 SLR Proceeding: a moment of crisis and opportunity: The NRCs SLR proceeding provides the agency with an opportunity to restore public confidence in its commitment to ensure public health and safety, by ending its silence regarding the crucially important 2011 Safety Evaluation, and by requiring Duke to complete the flood protection measures required ten years ago. The NRC should also require Duke to prepare a new environmental risk analysis that uses correct, complete, and up-to-date methods and data. Finally, Duke should account for its failure to implement mitigative measures required by the NRC ten years ago for adequate protection, and now ignored in Dukes SLR application.

MITMAN DECLARATION EXHIBIT 2

CURRICULUM VITAE FOR JEFFREY T. MITMAN Rockville, MD Project Management / PRA Position in the Nuclear Industry QUALIFICATIONS Senior Reliability and Risk Analyst with more than 35 years experience in the Nuclear Industry. Responsible for managing risk analysis projects and teams. Solid record of bringing projects in on schedule and budget.

MAJOR ACCOMPLISHMENTS Transitioned NRC to detailed PRA models for low power and shutdown significance determinations process evaluations.

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

Obtained regulatory approval of EPRIs RI-ISI methodology.

Managed first PRA of bolted spent fuel storage cask.

EXPERIENCE 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.

Lead analyst for low power and shutdown 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.

Participated in post NRCs Fukushima NTTF flooding guidance development.

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

Advised NRC 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 Configuration risk management evaluation of at-power fire risk.

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.

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.

Jeffrey T. Mitman Page 2 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.

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

Jeffrey T. Mitman Page 3 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 Risk Informed Standards Committee (RISC)

ANS 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

Page 1 of 2 Reports and Papers by Jeffrey T. Mitman As of September 2021 Papers:

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

2. Potentially Non-conservative Screening Value for Dam Failure Frequency in PRA, US NRC Information Notice 2012-02 (ML090510269). Co-author and technical point of contact.

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

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

5. An Assessment of Large Dam Failure Frequencies Based on US Historical Data, PSA 2011.

Co-author.

6. Generic Failure Rate Evaluation for Jocassee Dam, US NRC (ML13039A084), 2010.

Co-author.

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

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

9. Probabilistic Risk Assessment of Bolted Dry Spent Fuel Storage Cask, Presented at ICONE12. 2004. Co-author.

10. Low Power and Shutdown Risk Assessment Benchmarking, Presented at PSA 02 2002.

Co-author.

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

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

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

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

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

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

17. Control of Reactor Vessel Temperature/Pressure during Shutdown, GE SIL 357. June 1981.

Co-author.

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.

Page 2 of 2 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.

PETITIONERS HEARING REQUEST AND WAIVER PETITION ATTACHMENT 2A

PETITIONERS HEARING REQUEST AND WAIVER PETITION ATTACHMENT 2B

PETITIONERS HEARING REQUEST AND WAIVER PETITION ATTACHMENT 2C

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE SECRETARY ln the Matter of Duke Energy Carolinas, LLC Oconee Nuclear Station, Units 1,2 & 3 Docket Nos. 50-26912701287 SLR DECLARATION OF ROSELLEN ALEGUIRE Under penalty of perjury, Rosellen Aleguire declares as follows:

1) My name is. Rosellen Aleguire. I am a member of the sierra club'

2) I live at145 Gladys Circle, Fair Play, SC 29643'

3) My home is located within the So-mile emergency planning zone (EPZ) of the oconee Nuclear Power station, for which Duke Energy carolinas LLC (Duke) has submitted an application to the U.S, Nuclear Regulatory Commission for the subsequent License Renewal of its operating license. All three oconee units have previously received aZo-year license extension on their original 40-year operating licenses.

4) Based on the historical experience of nuclear power stations, I believe that these facilities are inherenfly dangerous. continued operations of oconee Nuclear Power station for an additional 20 years beyond the three reactors' current license expiration dates could cause a severe nuclear accident in the reactor(s) andlor irradiated fuel storage pool(s) thereby causing death, injury, illness, dislocation, and economic damage to me and my family. lt could also cause devastating environmental damage'

5) I believe that Duke,s application to extend operations of oconee Nuclear station from o'o 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'

)

)

)

)

)

nlfts/ao a t DATE

PETITIONERS HEARING REQUEST AND WAIVER PETITION ATTACHMENT 3

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE SECRETARY In the Matter of

)

Duke Energy Carolinas, LLC

) Docket Nos. 50-269/270/287 SLR Oconee Nuclear Station,

)

Units 1, 2 & 3

)

DECLARATION OF DIANE CURRAN IN SUPPORT OF PETITION FOR WAIVER OF 10 C.F.R. §§ 51.53(c)(3)(i), 51.53(c)(ii)(L), 51.71(d),

51.95(c)(1), AND 10 C.F.R. SUPBART A, APPENDIX B, TABLE B-1 TO ALLOW CONSIDERATION OF CATEGORY 1 NEPA ISSUES AND SAMA ISSUES Under penalty of perjury, I, Diane Curran state as follows:

1. I am an attorney for Petitioners Beyond Nuclear, Inc. and Sierra Club, Inc. in the above-captioned subsequent license renewal proceeding.

2. I am qualified by my legal training and professional experience as an expert on the Atomic Energy Act, the National Environmental Policy Act (NEPA), judicial opinions interpreting those statutes, and regulations and decisions of the U.S. Nuclear Regulatory Commission (NRC) interpreting those statutes. My expertise extends to NRC regulations and decisions regarding reactor license renewal and subsequent license renewal. I also have a general understanding of technical issues related to nuclear reactor safety and environmental impact analysis, at a level that is sufficient for me to make a reasonable evaluation of NRC technical correspondence and reports on safety and environmental issues.

3. My legal training consists of a Juris Doctor degree from the University of Maryland. My professional experience consists of more than 35 years of providing legal representation to citizen groups and state and local governments in NRC licensing and enforcement proceedings, including license renewal proceedings and subsequent license renewal proceedings.

4. The purpose of my declaration is to support Petitioners Petition for Waiver of 10 C.F.R.

§§ 51.53(c)(3)(i), 51.53(c)(ii)(L), 51.71(d), 51.95(c)(1), and 10 C.F.R. Part 51, Subpart A, Appendix B, Table B-1 to Allow Consideration of Category 1 NEPA Issues (Waiver Petition). The Petition is presented in Section V of Hearing Request and Petition to Intervene by Beyond Nuclear and the Sierra Club (Sept. 27, 2021) (Hearing Request),

which has been submitted to the NRC.

5. I am responsible for the contents of the Waiver Petition, including its assertion that application of the regulations from which Petitioners seek an exemption would not serve the purposes for which the rules were adopted. I am also responsible for identifying, with particularity, the specific aspects of the subject matter of this proceeding that should be

2 considered in a full environmental analysis that is compliant with the procedural requirements of the National Environmental Policy Act (NEPA).

6. The factual assertions in the Waiver Petition and Hearing Request are based on the expert opinion of Mr. Jeffrey Mitman, as set forth in his Declaration (Attachment 1 to PetitionersHearing Request) and his Expert Report, NRC Relicensing Crisis at Oconee Nuclear Station: Stop Duke From Sending Safety Over the Jocassee Dam (Sept. 2021)

(Exhibit 1 to Mr. Mitmans Declaration). Mr. Mitmans Expert Report, in turn, is based on publicly available documents generated by Duke Energy Corp. (Duke) and the NRC.

7. I have no reason to question the veracity of the facts recited by Mr. Mitman or the reliability of his expert opinion. Thus, my representation of the content of his Expert Report and the documents he relies on is true and correct to the best of my knowledge.

And the legal opinions expressed in the Waiver Petition are based on my best professional judgment.

September 27, 2021 Diane Curran