ML24067A088

Email Hearing Request from San Luis Obispo Mothers for Peace, Friends of the Earth, and Environmental Working Group

ML24067A088
Person / Time
Site:	Diablo Canyon
Issue date:	03/04/2024
From:	Curran D Harmon, Curran, Harmon, Curran, Spielberg & Eisenberg, LLP
To:	Bessette P, Matthews T Morgan Lewis, NRC/SECY
References
50-275-LR-2, 50-323-LR-2
Download: ML24067A088 (1)

Text

From:

Diane Curran To:

Docket, Hearing; Bessette, Paul M.; ryan.lighty@morganlewis.com; Timothy J. Matthews (timothy.matthews@morganlewis.com); Jeremy Wachutka; Catherine Kanatas; Adam Gendelman Cc:

Templeton, Hallie; Caroline Leary

Subject:

[External_Sender] Filing of hearing request in Diablo Canyon license renewal proceeding, Docket Nos. 50-275-LR and 50.323-LR Date:

Monday, March 04, 2024 9:50:51 PM Attachments:

2024.03.04 SLOMFP-FoE-EWG Hearing Request re DCPP LRA_opt.pdf

Dear NRC Secretary and parties,

On behalf of San Luis Obispo Mothers for Peace, Friends of the Earth, and Environmental Working Group, I am filing a hearing request regarding PG&Es application for renewal of the Diablo Canyon operating licenses.

I am sending the hearing request by email because I was not able to request access to the electronic docket in time for this filing. I will post the hearing request to the electronic docket as soon as I get access.

In the meantime, I am sending the hearing request in three separate emails out of concern that the exhibits may be too large to send in one message:

This email has the hearing request itself with Exhibit 1 attached.

The second email will have Exhibits 2 and 4.

The third email will have Exhibit 3.

Sincerely, Diane Curran Counsel for SLOMFP

Diane Curran Harmon, Curran, Spielberg & Eisenberg LLP Contact (240) 393-9285 1725 DeSales Street, NW, Suite 500, Washington, DC 20036 Email DCurran@harmoncurran.com

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE COMMISSION In the matter of Pacific Gas and Electric Company Docket Nos. 50-275-LR, 50-323-LR Diablo Canyon Nuclear Power Plant March 4, 2024 Units 1 and 2 REQUEST BY SAN LUIS OBISPO MOTHERS FOR PEACE, FRIENDS OF THE EARTH AND ENVIRONMENTAL WORKING GROUP FOR HEARING ON PACIFIC GAS & ELECTRIC COMPANYS LICENSE RENEWAL APPLICATION FOR THE DIABLO CANYON NUCLEAR PLANT I. INTRODUCTION In accordance with the U.S. Nuclear Regulatory Commissions (NRCs) hearing notice, 88 Fed. Reg. 87,817 (Dec. 19, 2023), and pursuant to 10 CFR 2.309(f), San Luis Obispo Mothers for Peace (SLOMFP), Friends of the Earth (FoE) and Environmental Working Group (EWG)

(collectively Petitioners) hereby request the Commissioners of the U.S. Nuclear Regulatory Commission (NRC or Commission) to grant a hearing on Pacific Gas & Electric Companys (PG&Es) application for renewal of the operating licenses for the Diablo Canyon nuclear power plant Units 1 and 2 (DCPP).

As demonstrated below in Section II, Petitioners have organizational standing to represent the interests of their members and supporters in this proceeding. Petitioners contentions are set forth in Section III below. Those contentions are:

CONTENTION 1: CONTINUED OPERATION OF DCPP UNDER A RENEWED LICENSE POSES AN UNACCEPTABLE SAFETY RISK AND SIGNIFICANT ADVERSE ENVIRONMENTAL IMPACT OF SEISMIC CORE DAMAGE ACCIDENTS CONTENTION 2: PG&E FAILS TO PROVIDE AN ADEQUATE PLAN TO MONITOR AND MANAGE THE EFFECTS OF AGING ON UNIT 1 REACTOR PRESSURE VESSEL.

CONTENTION 3: PG&E FAILS TO DEMONSTRATE COMPLIANCE WITH THE COASTAL ZONE MANAGEMENT ACT.

2 II. DESCRIPTION OF PETITIONERS Petitioners are non-profit organizations with a longstanding record of concern about the safety and economic viability of the Diablo Canyon reactors. They seek a hearing in order to ensure that DCPP is not permitted to operate another twenty years unless and until the NRC addresses the significant safety and environmental risks that are now apparent and that would continue.

A. San Luis Obispo Mothers for Peace Located in San Luis Obispo, California, SLOMFP is a non-profit membership organization concerned with the dangers posed by Diablo Canyon and other nuclear reactors, nuclear weapons, and radioactive waste. SLOMFP also works to promote peace, environmental and social justice, and renewable energy. SLOMFP has participated in NRC licensing cases involving the Diablo Canyon reactors since 1973. SLOMFPs representational standing to participate in this proceeding is demonstrated by the attached Declarations of SLOMFP members Sherry Lewis, Julie Mansfield-Wells, Linda Seeley, Lucy Jane Swanson, and Jill ZamEk.1 B. Friends of the Earth FoE is a tax exempt, nonprofit environmental advocacy organization dedicated to improving the environment and creating a more healthy and just world.2 The organization was founded in 1969 by David Brower in part to protest safety and environmental issues at the newly emerging Diablo Canyon. FoE has more than 282,000 members in all 50 states and the District of Columbia, approximately 42,600 of whom are in California. In addition to formal members, FoE has more than 8.6 million online activist supporters across the country. FoE also has office space 1 See Exhibits 1(A) - 1(E).

3 in Berkeley, California. FoEs representational standing to participate in this proceeding is demonstrated by the attached Declarations of FoE members Sherry Lewis, Julie Mansfield-Wells, Linda Seeley, Jane Swanson, and Jill ZamEk.3 C. Environmental Working Group EWG is a 501(c)(3) non-profit, non-partisan organization that works to empower people to live healthier lives in a healthier environment. EWG does this, in part, by creating and sharing research reports and consumer guides that help people educate themselves about the food they consume, the products they buy, and the companies they support, so that everyone can make safer and more informed choices. In furtherance of its mission, EWG engages in research and policy advocacy on a broad range of issues related to state and federal energy policy, climate change, renewable energy, toxic chemicals, food and agriculture, water and air pollution, and public health.

EWGs work on energy policy has included rate design and public policy issues related to consumer use of and access to solar.

While EWG is a national organization, the group has a strong presence in California. Out of a nationwide group of 410,000 active supporters, more than 58,000 Californians have signed up to receive monthly EWG emails that seek to learn about supporters concerns and opinions in order to guide EWG in establishing organizational goals.

Over several decades, EWG has developed public education information and has submitted formal testimony about radiological risks posed by reactors and facilities for nuclear waste transportation, storage and disposal. Since 2002, for example, EWG has examined and assisted 2 Friends of the Earth is a part of Friends of the Earth International, a federation of grassroots groups working in 74 countries on today's most urgent environmental and social issues. Friends of the Earth International is the worlds largest grassroots environmental federation.

3 See Exhibits 1(A) - 1(E).

4 the public in understanding the transportation implications of nuclear waste routes that could be utilized to transport deadly radioactive material from around the United States to California nuclear sites. EWGs President Ken Cook testified on the crucial issues surrounding the licensing process for the proposed facility for long-term storage of lethal, long-lived nuclear waste at Yucca Mountain in Nevada. Additionally, as ionizing radiation is known to cause cancer in humans, EWG provides educational and policy advocacy on radiation in drinking water. EWGs Tap Water Database reports that more than a dozen different radioactive elements are detected in American tap water. The most common are beryllium, radon, radium, strontium, tritium and uranium. EWG data show that radioactive contaminants are detected in water serving 165 million people across the U.S. Continuing Diablo Canyons operation increases the risk that even more people will be exposed to cancer-causing radioactive elements in their drinking water. EWG has also participated in proceedings for the advocacy of alternatives to dangerous nuclear reactor operation. In California, EWG was found to have party status to intervene in a net energy metering tariff proceeding due to its interest in developing a tariff that supports sustainable growth of rooftop solar.4 EWG is highly concerned about -- and its supporters would be directly impacted by --

continued operation of the aging Diablo Canyon nuclear plant because of its high cost to taxpayers and extreme safety and environmental hazards. EWG and its supporters are concerned that continuing its operation hurts the states shift to safe, renewable energy and prolongs the risk of a disaster at the plant. Californians are saddled with the cost of continuing to prop up the unnecessary and unsafe nuclear power plant. Allowing Diablo Canyon to keep operating will 4 CA NEM Proceeding: R20-08-020.

5 enable and compel PG&E and other companies with aging and uneconomic reactors to slow-walk the transition to those renewable and far less expensive energy sources.

While EWG is not technically a membership organization, it represents the interests of its supporters in a manner that satisfies the standard for organizational standing recently set forth by the U.S. Supreme Court in Students for Fair Admissions, Inc. v. President and Fellows of Harv.

Coll., 143 S.Ct. 2141, 600 U.S. 181 (2003). First, EWG seeks to participate in this license renewal proceeding in a good faith effort to ensure that its supporters interest in protecting public health and safety and the environment from radiological contamination and risks are represented in the NRCs decision-making process.5 Second, as stated in the attached declarations of EWG supporters June Cochran, Patricia Kohlen, and Linda Parks, they voluntarily support EWG and support its mission; they regularly receive emails from EWG seeking their input on its programs and activities; and EWG keeps them informed of the status of legal proceedings in which their interests are represented.6 [F]urther scrutiny into how EWG operates is not required.7 Thus, EWG has organizational standing in this proceeding.

In the event the NRC determines that EWG does not have organizational standing to represent its supporters interests, EWG seeks discretionary recognition of its standing to advocate for the interests of its supporters.8 EWG respectfully submits that it satisfies the standard for discretionary intervention, as set forth in Pebble Springs, as follows:

5 Students for Fair Admissions, 143 S.Ct. at 2158 (requiring demonstration that organization represents its members in good faith). See also attached Declaration of Ken Cook, Exhibit 1(F)

(attesting to good faith representation of supporters interests).

6 See Exhibit 1(G) - 1(I). See also Students for Fair Admissions, 143 S.Ct. at 2158.

7 Students for Fair Admissions, 143 S.Ct. at 2158.

8 10 C.F.R. § 2.309(e); Portland General Electric Co. (Pebble Springs Nuclear Plant, Units 1 &

2), CLI-76-27, 4 N.R.C. 610, 616 (1976) (Pebble Springs).

6 First, EWG will significantly contribute to the development of a sound record through its expertise on public health and environmental issues in California; that its thousands of supporters have significant property, financial, public health and safety and environmental interests in the state and the region; and that these interests will be affected by the NRCs decision in this proceeding.

Further, EWG has no means other than this proceeding for protecting its interests. And SLOMFP and FoE may not be able to fully represent EWGs interests without the assistance and participation of EWG staff.

Finally, EWGs participation will not broaden this proceeding because EWG seeks to participate regarding the same contentions as SLOMFP and FoE. In addition, as demonstrated above, EWGs interest and activities fall within the zone of interest protected by the Atomic Energy Act and NEPA.9 9 Students for Fair Admissions, 143 S.Ct. at 2157 (requiring that interests sought to be protected must be germane to the organizations purpose). See also EWG Mission Statement at ewg.org:

EWG's team of scientists, policy experts, lawyers and communications and data experts work tirelessly to reform our nation's broken chemical safety and agricultural laws. We push industries to adopt our standards and stand against chemicals of concern.

7 CONTENTIONS CONTENTION 1: CONTINUED OPERATION OF DCPP UNDER A RENEWED LICENSE POSES AN UNACCEPTABLE SAFETY RISK AND SIGNIFICANT ADVERSE ENVIRONMENTAL IMPACT OF SEISMIC CORE DAMAGE ACCIDENTS A. Statement of Contention The NRC should deny PG&Es license renewal application for DCPP because continued operation of the reactors poses an unacceptable risk of core damage accidents due to earthquakes.

Therefore, renewal of PG&Es operating license would not satisfy the statutory standard set by the Atomic Energy Act that operation of DCPP will provide adequate protection to the health and safety of the public.10 By the same token, continued operation of DCPP also poses significant or LARGE adverse environmental impacts, not SMALL impacts as asserted by PG&E in its Environmental Report.11 As required by the National Environmental Policy Act (NEPA), 42 U.S.C. § 4332(c)(iii), the Environmental Report should weigh the costs and benefits of the alternative that would avoid these impacts: closing DCPP on the reactors current 2024/2025 retirement dates.12 10 42 U.S.C. § 2232. The unacceptable risk of an earthquake-related core damage accident at DCPP is current and ongoing. Therefore, in conjunction with this Hearing Request, Petitioners have filed a separate petition with the Commissioners seeking immediate closure of the DCPP reactors unless and until the risk of an earthquake-related core damage accident can be significantly reduced.

11 Environmental Report at 4-61 (asserting that PG&Es seismic risk analysis did not change the conclusion of the 1996 License Renewal Generic Environmental Impact Statement (1996 License Renewal GEIS) that the environmental impacts of severe accidents are SMALL.). While the NRC retains ultimate responsibility for independently evaluating environmental impacts and alternatives under NEPA, the NRC also relies on the information presented in environmental reports. 10 C.F.R. § 51.41.

12 As discussed below, the risk posed by continued operation of DCPP is almost two orders of magnitude higher than the SMALL risk assumed by PG&E in the Environmental Report.

Therefore, the Severe Accident Mitigation Alternatives (SAMAs) listed in Appendix G of the

8 Quantification of the seismic risks and environmental impacts can be found in PG&Es two most recent public reports on earthquake risks - its 2018 Seismic Probabilistic Risk Assessment (SPRA) and its 2023 Environmental Report. In the SPRA, PG&E estimates a value of 3x10-5 per year for seismic core damage frequency.13 In the Environmental Report, PG&E asserts a similar value (2.96x10-5) and characterizes their environmental significance as SMALL. 14 As set forth in the attached Declaration of Peter Bird, Ph.D (March 4, 2024), (Bird Declaration), however, PG&Es SCDF estimate is too low by a factor of 47~70.15 PG&Es significant underestimate of SCDF arises principally from its assumption that the majority of large earthquakes affecting DCPP are strike-slip earthquakes and its disregard of the significant contribution of thrust-faulting earthquake sources under the DCPP site and in the adjacent Irish Hills. But the January 2024 occurrence of the Noto Peninsula earthquakes on analogous faults in Japan now demonstrates in no uncertain terms that PG&Es assumption is both unfounded and dangerous. As discussed below, these thrust-faulting earthquakes produce strong Environmental Report are grossly inadequate to address the magnitude of the environmental impacts involved. Where the risk of significant impacts is as high as for DCPP, denial of PG&Es license renewal application and closure of the reactors is a reasonable alternative that should be considered.

13 PG&E Letter DCL-18-027 re: Seismic Probabilistic Risk Assessment for the Diablo Canyon Power Plant, Units 1 and 2 - Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding the Seismic Aspects of Recommendation 2.1: Seismic of the (sic) Near-Term Task force Review of Insights from the Fukushima Dai-ichi Accident, Encl. 1 at 52 (Apr. 24, 2018) (NRC Accession No. ML18120A201) (SPRA).

The SPRA relies in turn on PG&Es Seismic Source Characterization for the Diablo Canyon Power Plant, San Luis Obispo County, California; report on the results of SSHAC level 3 study, (Rev. A, March 2018) (Available online at http://www.pge.com/dcpp-ltsp) (SSC); and PG&E Letter DCL-15-035 re: Response to NRC Request for Information Pursuant to 10 CFR 50.54(f)

Regarding the Seismic Aspects of Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident: Seismic Hazard and Screening Report (Mar. 11, 2015) (NRC Accession No. ML15071A045).

14 Environmental Report at 4-62.

9 shaking that leads to a much higher chance of seismic core damage than the strike-slip faults assumed by PG&E to predominate at DCPP.16 Applying the experience of the Noto earthquakes to the thrust-faulting earthquakes at and near DCPP, a reasonable SCDF estimate could be as high as 1.4x10-3/year.17 As stated by Dr. Bird:

In the 2024 Noto Peninsula earthquake, we have the advantage of the finite-fault solution (USGS, 2024), which maps the amount of coseismic slip onto the active fault plane. This study showed maximum slip of 3.7 m under the center of the Noto Peninsula, with a mean slip that I visually estimate as 2.0 m (or 2000 mm) in the seismogenic depth range.

Dividing this mean slip of 2000 mm by the long-term tectonic slip-rate of 2.8 mm/a in the Irish Hills, the inferred recurrence rate for Noto-type earthquakes under the Irish Hills is 715 years. In other words, the inferred probability of Noto Peninsula-type earthquakes under the Irish Hills is the inverse of this, which is 1.4x10-3 /yr.

Again, reasonably presuming that the Noto Peninsula earthquake is a characteristic earthquake for this tectonic setting (shared by the Irish Hills in California), PGA values of 1.0~2.3 g (see section 1 above) must be expected with probability 1.4x10-3 /yr.

However, in the 2015 SSC (specifically, in Figure 2.3.7-1 of PG&E, 2015L), we see that this outdated modeling associated this probability level with a PGA of only 0.32 g.

Consequently, it appears that the 2015 SSC severely underestimated (by a factor of 3~7) the severity of shaking (PGA) that must be resisted every ~715 years.18 In other words, as asserted by Dr. Bird, the severe accident that PG&E asserts will occur only once in 33,000~50,000 years may actually occur every ~715 years. That means that a license extension 15 Dr. Birds Declaration is attached as Exhibit 2.

16 Bird Declaration, ¶¶ 14(5), 18-21.

17 Id., ¶¶ 4, 6, 29-30. As stated in ¶¶ 32-34 of Dr. Birds Declaration, his SCDF estimate is based on information provided in the SPRA, for which some questions about the meaning of PG&Es terminology exist. And there may be differences of opinion about the appropriate interpretation of Noto Peninsula seismographs that should be resolved by further study. In the meantime, for purposes of evaluating PG&Es Environmental Report, it is reasonable to assume that the levels of shaking seen in the Noto Peninsula earthquake will cause seismic core damage at DCPP if and when they occur in the Irish Hills of California.

18 Id., ¶¶ 4, 6, 29-30.

10 for 20 years would incur a ~2.8% probability of a severe accident.19 Under NRC guidance, such a high core damage frequency calls for immediate regulatory action to maintain the plant in a safe condition.20 In this context, it requires the denial of PG&Es license renewal application.21 B. Basis Statement This contention is supported by the Bird Declaration, Exhibit 2. Dr. Bird, Professor Emeritus of Geophysics and Geology at the University of California at Los Angeles (UCLA), is highly qualified through 46 years of training and experience in the fields of geology and geophysics. His focus is on technophysics and seismicity, including plate motion and plate deformation. And Dr. Bird has done extensive work on the geology of California, including a number of academic papers on computer modeling methods and applications, including studies of the ongoing (neotectonic) deformation in California. He has also been a member or officer of several professional organizations relating to his expertise, including the Geological Society of America, the American Geophysical Union and the Southern California Earthquake Center. The former two organizations have recognized Dr. Birds work with two fellowships and an award.

Dr. Bird has a detailed understanding of the seismicity at DCPP. In 2012, he participated in a Senior Seismic Hazards Analysis Committee (SSHAC) review sponsored by PG&E and run by Lettis Consultants International, regarding seismic hazard at DCPP. He presented results on both strike-slip and compressional deformation rates affecting the region, which were derived from his computer models of neotectonics. These models were prepared for the Southern 19 Id., ¶¶ 14(6), 4, 6, 29-30.

20 NRC Office Instruction LIC-504, Integrated Risk-Informed Decision-Making Process for Emergent Issues at 4 (Rev. 6, Sept. 7, 2023) (ML23165A117) (LIC-504).

21 Id. at 4-5 (including orders to shut down reactors among the scenarios that could result from application of LIC-504).

11 California Earthquake Centers project Unified California Earthquake Rupture Forecast, and also for the US Geological Surveys 2013 Update to the National Seismic Hazard Model.

In Section IV of his Declaration, Dr. Bird sets forth in detail the basis for the data and analyses supporting his expert opinion and this contention. As summarized in Section IV.A (Abstract):

(1) The Noto Peninsula earthquake in Japan (2024.01.01, m7.5, 10 km deep) produced peak ground accelerations (PGA) of 1.0~2.3 g (that is, 100~230% of gravity) at 5 modern digital strong-motion seismometers as far as 42 km from the rupture.

(2) This strong shaking occurred in the Noto Peninsula, which is part of the hanging-wall (upper block) of two en-echelon thrust faults that run parallel to its two coasts.

(3) The Irish Hills, San Luis Range, and DCPP site in California are at risk for similar earthquakes and similar shaking because they are underlain by similar thrust faults, including the inland Los Osos thrust fault and the Inferred Coastline thrust running along the shore by DCPP.22 (4) The expected recurrence interval between such events at DCPP can be roughly estimated by dividing the expected fault slip (averaging 2 m in the Noto earthquake, according to the USGS finite-fault solution) by the total heave rate of the thrust faults under DCPP, which is about 2.8 mm/year.... The result is 715 years. The inverse of this is the rate:

1.4x10-3 /yr.

(5) In the existing SSC (PG&E, 2015; 2015L), the intensity of shaking at this return period of 715 years has been underestimated by a factor of 3~7. This means that the chance of seismic core damage is much higher when thrust-faulting earthquake sources are included.

(6) Applying Dr. Birds analysis to these facts, the probability of a severe accident of earthquake origin at DCPP has been underestimated by a factor of (1.4x10-3 /yr) /

(2~3x10-5 /yr) = 47~70. In other words, the severe accident that PG&E asserts will occur only once in 33,000~50,000 years may actually occur every ~715 years. That 22 Inferred Coastline thrust is Dr. Birds own term for a distinct fault surface whose trace follows the coastline opposite DCPP. Unlike the Shoreline fault in the same area, the Inferred Coastline thrust dips at a gentle angle beneath DCPP and has the up-dip rake of a thrust fault.

12 means that a license extension for 20 years would incur a ~2.8% probability of a severe accident.

This contention is also supported by two NRC guidance documents which demonstrate that the SCDF of 1.4x10-3/yr estimated by Dr. Bird poses a significant safety and environmental risk. First, LIC-405 characterizes the risk impact from external events as high and therefore warrants immediate regulatory action to place or maintain the facility in safe condition if:

Conditional core damage frequency (CCDF) (i.e., CDF because of the issue) is high (e.g.,

greater than or on the order of 1x10-3/year).23 Second, NRC Regulatory Guide 1.174, Rev. 1, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis (Jan.

2018) states that if total CDF is considerably greater than 10-4 per reactor year, the NRCs focus in considering licensing actions should be on finding ways to decrease rather than increase the risk.24 Extending the DCPP license by 20 years clearly increases the risk to the public.

In this case, the significant risk and adverse impact posed by continued operation of DCPP can and should be eliminated by closing the reactors no later than their current retirement dates of 2024 (Unit 1) and 2025) Unit 2, i.e. implementing the no action alternative under NEPA.25 Finally, this contention is supported by the NRCs correspondence with PG&E, which demonstrates that the NRC has found that DCPP poses a high seismic risk level compared to other reactors and required PG&E to perform a seismic risk analysis and seismic PRA - and yet failed to 23 Id. at 4 (NRC Accession No. ML23165A117).

24 Id. at 28 (NRC Accession No. ML17317A256).

25 Because the unacceptable seismic risks also apply during the current license term, Petitioners have submitted a separate request to the Commissioners to take immediate enforcement action.

13 make Atomic Energy Act-based safety findings in the course of reviewing those studies.26 As noted in NRC guidance, the word safety is used in NRC review documents to connote that an evaluation measured compliance with the adequate protection standard of the Atomic Energy Act and NRC regulations.27 That statutory language does not appear in either the NRC 12/21/16 Letter or the NRC 01/22/19 Letter. Therefore, in the absence of any Atomic Energy Act-based safety evaluation by the NRC of PG&Es SSC and SPRA, there is no precedential value in this license renewal review.

Accordingly, it is both necessary and appropriate to review PG&Es seismic risk analyses in light of the new information provided by the Noto Peninsula earthquake and assessed by Dr.

Bird, with respect to whether continued operation of DCPP poses an unacceptable safety risk to public health and safety and/or environmental risks that can only be avoided by denying PG&Es license renewal application.

C. Demonstration that the Contention is Within the Scope of the Proceeding This contention is within the scope of this license renewal proceeding by virtue of the 26 See Letter from Frankie Vega, NRC, to Edward D. Halpin, PG&E, re: Diablo Canyon Power Plant, Unit Nos. 1 and 2 - Staff Assessment of Information Provided Under Title 10 of the Code of Federal Regulations Part 50, Section 50.54(f), Seismic Hazard Reevaluation for Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident (CAC Nos. MF5275 and MF5276), Encl. at 1 (Dec. 21, 2016) (NRC 12/21/16 Letter) (NRC Accession No. ML16341C057); Letter from Louise Lund, NRC, to James M. Welsh, PG&E, re: Diablo Canyon Power Plant, Unit Nos. 1 and 2 - Staff Review of Seismic Probabilistic Risk Assessment Associated With Reevaluated Seismic Hazard Implementation of the Near-Term Task Force Recommendation 2.1 Seismic (EPID No. L-2018-JLDE-0006) (Jan. 22, 2019) (NRC 1/21/19 Letter) (NRC Accession No. ML20093B934).

27 See Office Instruction LIC-101, License Amendment Review Procedures (Rev. 6, July 31, 2020)

(NRC Accession No. ML19248C539):

[L]icensing actions typically require findings of reasonable assurance that operation of the facility can be conducted without endangering the health and safety of the public and will not be inimical to the health and safety of the public (e.g., 10 CFR 50.57(a)).

14 commitment made by NRC Chairman Christopher T. Hanson on behalf of the full Commission, whose members sat with him at the hearing, to California Senator Alex Padilla to review seismic safety and environmental risks as part of the proceeding. In an April 19, 2023, hearing before the Senate Committee on Environment and Public Works, Senator Padilla questioned Chairman Hanson regarding the NRCs plans for ensuring that DCPP is operationally safe with specific concern about seismic risk.28 Mr. Hanson responded that in addition to ongoing safety oversight:

Were going to be looking at updated safety information as part of that license renewal process. We did require all plants to take a look at the enhanced... you know to relook at their risks after Fukushima; Diablo, of course did look at their seismic risk again, and well take another look at that as part of the license renewal process... 29 Commissioner Hansons statements are also consistent with the NRCs authority and obligation under the Atomic Energy Act to ensure that its licensing decisions pose no undue risk to public health and safety.30 In addition, Commissioner Hansons commitment on behalf of the full Commission is consistent with NEPAs requirement that all federal agencies must take a hard look at the environmental impacts of their proposed actions.31 Because the Environmental Report is the preliminary environmental risk evaluation on which the NRC will eventually rely for its EIS, its analysis must be rigorous and thorough.32 Petitioners challenge the adequacy of the Environmental 28 A recording of the hearing is posted on the Committees website at:

https://www.epw.senate.gov/public/index.cfm/hearings?ID=DD1B6EC6-588A-4A56-9961-F9961BE12270. Sen. Padillas question can be found at approximately 1:45:26.

29 Id. (emphasis added). Chairman Hansons response can be found at approximately 1:45:55.

30 42 U.S.C. § 2232 (requiring NRC to ensure that licensed activities will provide adequate protection to the health and safety of the public.). See also Yankee Atomic Electric Co. (Yankee Rowe Nuclear Power Station), CLI-91-11, 34 N.R.C. 3, 12 (1991) (The Commission has the ultimate responsibility to ensure the safe operation of the facilities that it licenses.).

31 Sierra Club v. Fed. Energy Regulatory Commn, 867 F.3d 1357, 1367 (D.C. Cir. 2017).

32 10 C.F.R. § 51.41.

15 Report and PG&Es most recent seismic studies that purport to find that the safety and environmental risks of operating DCPP in a renewed operating license term are acceptable or insignificant. Chairman Hanson has committed the agency to review these issues in the course of the license renewal proceeding.

Therefore, the consistency of Chairman Hansons commitments on behalf of the full Commission with the NRCs statutory responsibilities for its licensing decisions under the Atomic Energy Act establish that Petitioners contention is within the scope of the proceeding.

D. Demonstration that the Contention is Material to the Findings NRC Must Make to Renew PG&Es Operating License Petitioners contention is material to the finding that Chairman Hanson and his fellow Commissioners have committed to make in the license renewal proceeding, i.e., whether continued operation of Diablo Canyon is operationally safe with respect to seismic risks. This commitment is also material to the NRCs statutory responsibility under the Atomic Energy Act to ensure that continued operation of DCPP will provide adequate protection to the health and safety of the public.33 Finally, and also consistent with the Chairmans commitment on behalf of the full NRC to Senator Padilla, Petitioners contention is material to the analysis required by NEPA regarding the significance of environmental impacts and reasonable alternatives for avoiding or mitigating those impacts.34 33 See note 30, supra.

34 See note 31, supra.

16 E. Concise Statement of the Facts or Expert Opinion Supporting the Contention, Along with Appropriate Citations to Supporting Scientific or Factual Materials The facts, NRC guidance, and expert opinion relied on in Petitioners contention are provided in the Statement of the Contention, the Basis Statement, and the attached Bird Declaration.

CONTENTION 2: PG&E FAILS TO PROVIDE AN ADEQUATE PLAN TO MONITOR AND MANAGE THE EFFECTS OF AGING ON UNIT 1 REACTOR PRESSURE VESSEL.

A. Statement of Contention PG&Es license renewal application does not include an adequate plan to monitor and manage the effects of aging due to embrittlement of the Unit 1 reactor pressure vessel (RPV) or an adequate time-limited aging analysis (TLAA), as required by 10 C.F.R. § 54.21.

B. Basis Statement As set forth in Exhibit 3, the Declaration of Digby Macdonald, Ph.D., the pressure vessel is a uniquely important and vulnerable component in a nuclear reactor, because it holds water on the highly radioactive reactor core, and because it has no backup if it should crack and lose water during an accident. Therefore, compliance with NRC requirements for monitoring the condition of the plant-specific pressure vessel is essential.

PG&Es proposed aging management program for the RPV relies heavily upon and perpetuates the preexisting and inadequate surveillance program that PG&E has used during the decades-old initial operating license period. That program is fundamentally deficient due to serious indications of an unacceptable degree of embrittlement, coupled with a lack of information to establish otherwise. These deficiencies are summarized in Section V of Dr. Macdonalds

17 Declaration (Exhibit 2) with reference to his more detailed analysis of September 14, 2023.35 Deficiencies include failure to consider serious indications of embrittlement that existed in 2003 and failure to conduct further monitoring of the Unit 1 RPV in the subsequent two decades, including removal of specimens for Charpy testing and ultrasound inspection of reactor beltline welds. In Section VI of his Declaration, Dr. Macdonald identifies additional, more recent concerns, including the erroneous assumption that embrittlement accrues in a non-Markovian manner, the appearance of Extrema in Capsule V CGraphs and Tables, the general failure by both PG&E and the NRC to address the significance of errors in PG&Es analyses, and PG&Es failure to address the potentially significant role of hydrogen in the embrittlement/crack propagation process.

Taking all of these deficiencies into account, Dr. Macdonald concludes that the NRC must reject PG&Es license renewal application because it relies on this outdated preexisting program without addressing or resolving its multiple serious inadequacies.

C. Demonstration that the Contention is Within the Scope of the Proceeding Contention 2 is within the scope of this proceeding because it challenges the adequacy of PG&Es aging management program with respect to the Unit 1 RPV, a component that is explicitly covered by the NRCs Part 54 regulations. As provided by10 C.F.R. § 54.4(a)(3), the scope of the regulations includes [a]ll systems, structures, and components relied on in safety analyses or plant evaluations to perform a function that demonstrates compliance with the Commission's regulations for... pressurized thermal shock (10 CFR 50.61).

35 Declaration of Digby Macdonald, Ph.D. in Support of Hearing Request and Request for Emergency Order by San Luis Obispo Mothers for Peace and Friends of the Earth (Sept. 14, 2023)

(9/14/23 Declaration) (NRC Accession No. ML23257A302). A copy of my 9/14/23 Declaration is attached to this Declaration as Attachment 1.

18 D. Demonstration that the Contention is Material to the Findings NRC Must Make to Renew PG&Es Operating License The issue of whether PG&E has proposed an adequate plan for monitoring and management of RPV embrittlement is material to this relicensing proceeding because the NRC must make certain findings as to the public health and safety; and if serious risks are found, it must either deny the application or impose significant modifications on the applicants operations.

E. Concise Statement of the Facts or Expert Opinion Supporting the Contention, Along with Appropriate Citations to Supporting Scientific or Factual Materials The facts and expert opinion supporting this contention are set forth in the Statement of Contention, the Basis Statement, and Dr. Macdonalds expert Declaration.

CONTENTION 3: PG&E FAILS TO DEMONSTRATE COMPLIANCE WITH THE COASTAL ZONE MANAGEMENT ACT.

A. Statement of Contention The NRC may not approve renewal of PG&Es operating licenses for DCPP because PG&E has not demonstrated compliance with the Coastal Zone Management Act (CZMA), 16 U.S.C. § 1451, et seq. For the same reason, PG&Es Environmental Report also fails to satisfy the requirements of NRCs own regulations mandating the content of environmental reports.36 B. Basis Statement While Appendix F of PG&Es Environmental Report contains a document entitled Coastal Zone Consistency Certification (Consistency Certification), by itself the Consistency Certification is insufficient to support license renewal. Before the NRC may grant license renewal, PG&Es Consistency Certification must be sanctioned by the State of California.37 In addition, the 36 10 C.F.R. §§ 51.45(b), (c), and (d).

37 16 U.S.C. § 1456(c)(3); 15 C.F.R. § 930.50 et. seq. See also NRC Reg. Guide 4.2, Preparation of Environmental Reports for Nuclear Power Stations at 21 (Rev. 3, Sept. 2018) (NRC Accession No. ML18071A400) (emphasis added):

19 State must grant any necessary coastal development permits (CDPs).38 Neither of these crucial approvals have occurred.

NRC regulations reflect these requirements at 10 C.F.R. §§ 51.45(b) and (c), which together requires license renewal applicants to address the environmental impacts of their proposed actions as well as alternatives for avoiding or mitigating them. Pursuant to 10 C.F.R. § 51.45(d), a license renewal applicant must also catalogue required approvals that must be obtained prior to license renewal.39 Thus, through approvals such as CZMA consistency and permitting, the license renewal application must incorporate State-imposed restrictions for addressing, avoiding and mitigating adverse environmental impacts of the proposed action.

In this case, the State agency responsible for implementation of the CZMA, the California Coastal Commission (CCC), has formally rejected PG&Es certification as incomplete and insufficient on multiple grounds (CCC Letter, Exhibit 4).40 These grounds include PG&Es failure to address effects on coastal uses and resources for the full 20-year period covered by PG&Es license renewal application instead of only five years; effects on fragile coastline areas and recreational activities of a proposed reduction in the size of the security area around DCPP; the substantial loss of marine life productivity from continued reliance on DCPPs once-through cooling system; effects of proposed dredging activities at DCPP; effects of continued operation on Applicants must submit to both the NRC and to the State a certification that the proposed activity complies with the enforceable policies of the States program. If the Coastal Zone Management Act applies to the project, the NRC cannot issue its license or permit until the State has concurred with the applicants certification of a coastal consistency determination.

38 California Public Resources Code § 30600.

39 10 C.F.R. § 51.45(b).

40 Letter from Tom Luster, CCC, to Tom Jones, PG&E re: Pacific Gas & Electric Companys Requested Nuclear Regulatory Commission License Renewal for Diablo Canyon Power Plant, San Luis Obispo County - Incomplete Consistency Certification at 3-8 (Dec. 7, 2023).

20 Environmentally Sensitive Habitat Areas, and effects of continued operation on traffic and circulation.41 In addition, the CCC Letter notes that PG&E may be required to obtain a CDP if the CCC finds that some of its activities meet the definition of development as provided by CCC policies for implementation of the CZMA:

The proposed extension of the operating license would affect a variety of coastal uses and resources beyond those acknowledged by PG&E in its consistency certification, and the Commission must evaluate those effects for consistency with the applicable policies of the CCMP [California Coastal Management Program.] Moreover, as described below, the extended intake and discharge of seawater that would be carried out as part of license renewal constitutes development within the meaning of Section 30106 [of CCMP policies.] It is also reasonably foreseeable that the proposed license extension includes other uses and activities that likely constitute development within the meaning of Section 30106 - for example, intake cove dredging and a new or expanded spent fuel facility.42 Accordingly, given the CCCs unequivocal rejection of PG&Es Coastal Certification as incomplete and inadequate, and given the potential requirement to obtain one or more CDPs, the NRC may not approve PG&Es license renewal application. Moreover, PG&Es license renewal application is incomplete because it does not address this issue.

C. Demonstration that the Contention is Within the Scope of the Proceeding Contention 3 is within the scope of this proceeding because the CZMA requires that PG&E must comply with its requirements before license renewal can be approved. In addition, NEPA and NRC regulations require that PG&E must accurately and reasonably describe the environmental 41 Id. at 3-8.

42 Id. at 2. See also id. at 3 n.6 (clarifying that the CCC has not determined that there would be no development requiring a CDP for a license extension).

21 impacts of its actions and alternatives to avoid or mitigate those impacts.43 And PG&E must demonstrate compliance with all relevant federal and State permitting requirements.44 D. Demonstration that the Contention is Material to the Findings NRC Must Make to Renew PG&Es Operating License Contention 3 is material to the findings NRC must make to renew PG&Es operating license because the NRCs Environmental Impact Statement for renewal of PG&Es operating license will rely on the Environmental Report to comprehensively address the environmental impacts of PG&Es renewed operation and alternatives for avoiding or mitigating those impacts. As discussed above in Section B, the Environmental Report is inadequate to fulfill that purpose with respect to the consistency certification and development permits that are required or may be required. As also discussed above in Section B, the NRC must also find that the State has confirmed PG&Es compliance with the CZMA before re-licensing Diablo Canyon.

E. Concise Statement of the Facts or Expert Opinion Supporting the Contention, Along with Appropriate Citations to Supporting Scientific or Factual Materials The facts supporting this contention are described above in the Statement of Contention and Basis Statement. Further support is also provided in Exhibit 3, the attached CCC Letter.

43 10 C.F.R. §§ 51.45(b), (c). See also State of New York v. NRC, 681 F.3d 471, 476 (D.C. Cir.

2012) (renewal of reactor operating licenses constitutes a major federal action requiring evaluation of environmental impacts and alternatives under NEPA).

44 10 C.F.R. §§ 51.45(b), (c), and (d).

22 III. CONCLUSION For the foregoing reasons, the NRC should find that Petitioners have standing and grant their hearing request.

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 Counsel to San Luis Obispo Mothers for Peace

__/signed electronically by/___

Hallie Templeton Friends of the Earth 1101 15th Street, 11th Floor Washington, DC 20005 434-326-4647 htempleton@foe.org Counsel to Friends of the Earth

__/signed electronically by/___

Caroline Leary Environmental Working Group 1250 I St N.W.

Washington, DC 20005 202-667-6982 cleary@ewg.org Counsel to Environmental Working Group March 4, 2023

23 CERTIFICATE OF SERVICE I certify that on March 4, 2023, I sent copies of the foregoing REQUEST BY SAN LUIS OBISPO MOTHERS FOR PEACE, FRIENDS OF THE EARTH AND ENVIRONMENTAL WORKING GROUP FOR HEARING ON PACIFIC GAS & ELECTRIC COMPANYS LICENSE RENEWAL APPLICATION FOR THE DIABLO CANYON NUCLEAR PLANT and Exhibits 1, 2, and 3 to the Secretary of the Commission and counsel for PG&E and the NRC Staff as follows:

NRC hearing docket, Hearing.Docket@nrc.gov NRC Office of the Secretary, Paul Bessette, paul.bessette@morganlewis.com Ryan Lighty, ryan.lighty@morganlewis.com Timothy Matthews, timothy.matthews@morganlewis.com Jeremy Wachutka, jeremy.wachutk@nrc.gov Catherine Kanatas, catherine.kanatas@nrc.gov Adam Gendelman, adam.gendelman@nrc.gov I have requested access to the EIE docket and will upload the hearing request to the docket as soon as I am granted access.

Electronically signed by Diane Curran

EXHIBIT 1 - STANDING DECLARATIONS Exhibit 1(A) Declaration of Sherry Lewis Exhibit 1(B) Declaration of Julie Mansfield-Wells Exhibit 1(C) Declaration of Linda Seeley Exhibit 1(D) Declaration of Lucy Jane Swanson Exhibit 1(E) Declaration of Jill ZamEk Exhibit 1(F) Declaration of Ken Cook Exhibit 1(G) Declaration of June Cochran Exhibit 1(H) Declaration of Patrician Kohlen Exhibit 1(I) Declaration of Linda Parks

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE SECRETARY In the matter of Pacific Gas and Electric Company Docket Nos. 50-275-LR, 50-373-LR Diablo Canyon Nuclear Power Plant Units 1 and 2 DECLARATION OF SHERRY LEWIS Under penalty of perjury, Sherry Lewis declares as follows:

1. My name is Sherry Lewis. I am a member of San Luis Obispo Mothers for Peace (SLOMFP) and Friends of the Earth (FOE).

2. I live at 209 Longview Lane, San Luis Obispo, CA 93405.

3. My home is located within the 50-mile ingestion pathway zone of Diablo Canyon Unit 1 and Unit 2 nuclear reactors (DCPP). I am aware that the licensee, Pacific Gas and Electric Company (PG&E), has requested the U.S. Nuclear Regulatory Commission (NRC) to renew the DCPP operating licenses for 20 years.

4. Based on my knowledge that the Diablo Canyon nuclear reactors are operating with aging equipment, including an embrittled reactor pressure vessel that has not been inspected for over 20 years; and based on the location of the Diablo Canyon reactors on or near a web of significant earthquake faults, I believe that the continued operation of the Diablo Canyon reactors for a renewed operating license term poses an unacceptable radiological accident risk that jeopardizes my health and safety and the environment.

5. Therefore, I have authorized SLOMFP and FOE to represent my interests by submitting a hearing request in this license renewal proceeding.

Executed in Accordance with 10 C.F.R. § 2.304(d) by Sherry Lewis Date: February 24, 2024

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE SECRETARY In the matter of Pacific Gas and Electric Company Docket Nos. 50-275-LR, 50-373-LR Diablo Canyon Nuclear Power Plant Units 1 and 2 DECLARATION OF JULIE MANSFIELD-WELLS Under penalty of perjury, Julie Mansfield-Wells declares as follows:

1. My name is Julie Mansfield-Wells. I am a member of San Luis Obispo Mothers for Peace (SLOMFP) and Friends of the Earth (FOE).

2. I live at 509 Mar Vista Drive, Los Osos, CA.

3. My home is located within the 50-mile ingestion pathway zone of Diablo Canyon Unit 1 and Unit 2 nuclear reactors (DCPP). I am aware that the licensee, Pacific Gas and Electric Company (PG&E), has requested the U.S. Nuclear Regulatory Commission (NRC) to renew the DCPP operating licenses for 20 years.

4. Based on my knowledge that the Diablo Canyon nuclear reactors are operating with aging equipment, including an embrittled reactor pressure vessel that has not been inspected for over 20 years; and based on the location of the Diablo Canyon reactors on or near a web of significant earthquake faults, I believe that the continued operation of the Diablo Canyon reactors for a renewed operating license term poses an unacceptable radiological accident risk that jeopardizes my health and safety and the environment.

5. Therefore, I have authorized SLOMFP and FOE to represent my interests by submitting a hearing request in this license renewal proceeding.

Executed in Accordance with 10 C.F.R. § 2.304(d) by Julie Mansfield-Wells Date: February 24, 2024

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE SECRETARY In the matter of Pacific Gas and Electric Company Docket Nos. 50-275-LR, 50-373-LR Diablo Canyon Nuclear Power Plant Units 1 and 2 DECLARATION OF LINDA M. SEELEY Under penalty of perjury, Linda M. Seeley declares as follows:

1. My name is Linda M. Seeley. I am a member of San Luis Obispo (SLOMFP) and Friends of the Earth (FOE).

2. I live at 1512 7th Street, Los Osos, CA 93402.

3. My home is located within the 50-mile ingestion pathway zone of Diablo Canyon Unit 1 and Unit 2 nuclear reactors (DCPP). I am aware that the licensee, Pacific Gas and Electric Company (PG&E), has requested the U.S. Nuclear Regulatory Commission (NRC) to renew the DCPP operating licenses for 20 years.

4. Based on my knowledge that the Diablo Canyon nuclear reactors are operating with aging equipment, including an embrittled reactor pressure vessel that has not been inspected for over 20 years; and based on the location of the Diablo Canyon reactors on or near a web of significant earthquake faults, I believe that the continued operation of the Diablo Canyon reactors for a renewed operating license term poses an unacceptable radiological accident risk that jeopardizes my health and safety and the environment.

5. Therefore, I have authorized SLOMFP and FOE to represent my interests by submitting a hearing request in this license renewal proceeding.

Executed in Accordance with 10 C.F.R. § 2.304(d) by Linda M. Seeley Date: February 24, 2024

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE SECRETARY In the matter of Pacific Gas and Electric Company Docket Nos. 50-275-LR, 50-373-LR Diablo Canyon Nuclear Power Plant Units 1 and 2 DECLARATION OF LUCY JANE SWANSON Under penalty of perjury, Lucy Jane Swanson declares as follows:

1. My name is Lucy Jane Swanson. I am a member of San Luis Obispo Mothers for Peace (SLOMFP) and Friends of the Earth (FOE).

2. I live at 313 Presidio Place, San Luis Obispo, CA 93401.

3. My home is located within the 50-mile ingestion pathway zone of Diablo Canyon Unit 1 and Unit 2 nuclear reactors (DCPP). I am aware that the licensee, Pacific Gas and Electric Company (PG&E), has requested the U.S. Nuclear Regulatory Commission (NRC) to renew the DCPP operating licenses for 20 years.

4. Based on my knowledge that the Diablo Canyon nuclear reactors are operating with aging equipment, including an embrittled reactor pressure vessel that has not been inspected for over 20 years; and based on the location of the Diablo Canyon reactors on or near a web of significant earthquake faults, I believe that the continued operation of the Diablo Canyon reactors for a renewed operating license term poses an unacceptable radiological accident risk that jeopardizes my health and safety and the environment.

Executed in Accordance with 10 C.F.R. § 2.304(d) by Lucy Jane Swanson Date: February 24, 2024

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE SECRETARY In the matter of Pacific Gas and Electric Company Docket Nos. 50-275-LR, 50-373-LR Diablo Canyon Nuclear Power Plant Units 1 and 2 DECLARATION OF JILL ZAMEK Under penalty of perjury, Jill ZamEk declares as follows:

1. My name is Jill ZamEk. I am a member of San Luis Obispo Mothers for Peace (SLOMFP) and Friends of the Earth (FOE).

2. I live at 1123 Flora Road, Arroyo Grande, CA.

3. My home is located within the 50-mile ingestion pathway zone of Diablo Canyon Unit 1 and Unit 2 nuclear reactors (DCPP). I am aware that the licensee, Pacific Gas and Electric Company (PG&E), has requested the U.S. Nuclear Regulatory Commission (NRC) to renew the DCPP operating licenses for 20 years.

4. Based on my knowledge that the Diablo Canyon nuclear reactors are operating with aging equipment, including an embrittled reactor pressure vessel that has not been inspected for over 20 years; and based on the location of the Diablo Canyon reactors on or near a web of significant earthquake faults, I believe that the continued operation of the Diablo Canyon reactors for a renewed operating license term poses an unacceptable radiological accident risk that jeopardizes my health and safety and the environment.

5. Therefore, I have authorized SLOMFP and FOE to represent my interests by submitting a hearing request in this license renewal proceeding.

Executed in Accordance with 10 C.F.R. § 2.304(d) by Jill ZamEk Date: February 24, 2024

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE SECRETARY In the matter of Pacific Gas and Electric Company Docket Nos. 50-275-LR, 50-373-LR Diablo Canyon Nuclear Power Plant Units 1 and 2 DECLARATION OF KEN COOK Under penalty of perjury, Ken Cook declares as follows:

1. My name is Ken Cook. I am President and co-founder of Environmental Working Group (EWG).

2. EWG is a 501(c)(3) non-profit, non-partisan organization that works to empower people to live healthier lives in a healthier environment. EWG does this, in part, by creating and sharing research reports and consumer guides that help people educate themselves about the food they consume, the products they buy, and the companies they support, so that everyone can make safer and more informed choices. In furtherance of its mission, EWG engages in research and policy advocacy on a broad range of issues related to state and federal energy policy, climate change, renewable energy, toxic chemicals, food and agriculture, water and air pollution, and public health. EWGs work on energy policy has included rate design and public policy issues related to consumer use of and access to solar.

3. While EWG is a national organization, the group has a strong presence in California. Out of a nationwide group of 410,000 active supporters, more than 58,000 Californians have signed up to receive monthly EWG emails that seek to learn about supporters concerns and opinions in order to guide EWG in establishing organizational goals. EWG also uses these emails to keep our supporters informed of the status of legal proceedings in which their interests are represented.

4. Over several decades, EWG has developed public education information and has submitted formal testimony about radiological risks posed by reactors and facilities for nuclear waste transportation, storage and disposal. Since 2002, for example, EWG has examined and assisted the public in understanding the transportation implications of nuclear waste routes that could be utilized to transport deadly radioactive material from around the United States to California nuclear sites. For instance, I testified on the crucial issues surrounding the licensing process for the proposed facility for long-term storage of lethal, long-lived nuclear waste at Yucca Mountain in Nevada.

5. Additionally, as ionizing radiation is known to cause cancer in humans, EWG provides educational and policy advocacy on radiation in drinking water. EWGs Tap Water Database reports that more than a dozen different radioactive elements are detected in

2 American tap water. The most common are beryllium, radon, radium, strontium, tritium and uranium. EWG data show that radioactive contaminants are detected in water serving 165 million people across the U.S. Continuing Diablo Canyons operation increases the risk that even more people will be exposed to cancer-causing radioactive elements in their drinking water. EWG has also participated in proceedings for the advocacy of alternatives to dangerous nuclear reactor operation. In California, EWG was found to have party status to intervene in a net energy metering tariff proceeding due to its interest in developing a tariff that supports sustainable growth of rooftop solar.1

6. EWG and its supporters are highly concerned about continued operation of the aging Diablo Canyon nuclear plant because of its high cost to taxpayers and extreme safety and environmental hazards. We are also concerned that continuing Diablo Canyons operation hurts the states shift to safe, renewable energy and prolongs the risk of a disaster at the plant. Californians are saddled with the cost of continuing to prop up the unnecessary and unsafe nuclear power plant. Allowing Diablo Canyon to keep operating will enable and compel PG&E and other companies with aging and uneconomic reactors to slow-walk the transition to those renewable and far less expensive energy sources.

7. EWG seeks to participate in this license renewal proceeding in a good faith effort to represent our supporters interest in protecting public health and safety and the environment from radiological accidents and contamination.

Executed in Accordance with 10 C.F.R. § 2.304(d) by Ken Cook Date: ___

1 CA NEM Proceeding: R20-08-020.

Kearcook 2 Feb2024

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE SECRETARY In the matter of Pacific Gas and Electric Company Docket Nos. 50-275-LR, 50-373-LR Diablo Canyon Nuclear Power Plant Units 1 and 2 DECLARATION OF JUNE COCHRAN Under penalty of perjury, June Cochran declares as follows:

1. My name is June Cochran. I am a supporter of Environmental Working Group (EWG).

As an EWG supporter, I receive monthly EWG emails that provides me with public education information about health risks posed by toxic and radiological contamination of consumer products and the environment. In these emails, EWG regularly asks about supporters concerns and opinions in order to guide it in establishing organizational goals. I am pleased with EWGs work and believe EWG generally represents my interests and concerns.

2. I live at 2622 Barcelona in Pismo Beach, California 93449. My home is located within the 50-mile ingestion pathway zone of Diablo Canyon Unit 1 and Unit 2 nuclear reactors (DCPP).

3. I am aware that the licensee, Pacific Gas and Electric Company (PG&E), has requested the U.S. Nuclear Regulatory Commission (NRC) to renew the DCPP operating licenses for 20 years.

4. Based on my knowledge that the Diablo Canyon nuclear reactors are operating with aging equipment, including an embrittled reactor pressure vessel that has not been inspected for over 20 years; and based on the location of the Diablo Canyon reactors on or near a web of significant earthquake faults, I believe that the continued operation of the Diablo Canyon reactors for a renewed operating license term poses an unacceptable radiological accident risk that jeopardizes my health and safety and the environment.

5. Therefore, I have authorized EWG to represent my interests by submitting a hearing request in this license renewal proceeding.

Date: March 1, 2024 June Cochran

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE SECRETARY In the matter of Pacific Gas and Electric Company Docket Nos. 50-275-LR, 50-373-LR Diablo Canyon Nuclear Power Plant Units 1 and 2 DECLARATION OF PATRICIA KOHLEN Under penalty of perjury, Patricia Kohlen declares as follows:

1. My name is Patricia Kohlen. I am a supporter of Environmental Working Group (EWG).

As an EWG supporter, I receive monthly EWG emails that provides me with public education information about health risks posed by toxic and radiological contamination of consumer products and the environment. In these emails, EWG regularly asks about supporters concerns and opinions in order to guide it in establishing organizational goals. I am pleased with EWGs work and believe EWG generally represents my interests and concerns.

2. I live at 359 Los Cerros Drive, San Luis Obispo, CA 93405. My home is located within the 50-mile ingestion pathway zone of Diablo Canyon Unit 1 and Unit 2 nuclear reactors (DCPP).

3. I am aware that the licensee, Pacific Gas and Electric Company (PG&E), has requested the U.S. Nuclear Regulatory Commission (NRC) to renew the DCPP operating licenses for 20 years.

4. Based on my knowledge that the Diablo Canyon nuclear reactors are operating with aging equipment, including an embrittled reactor pressure vessel that has not been inspected for over 20 years; and based on the location of the Diablo Canyon reactors on or near a web of significant earthquake faults, I believe that the continued operation of the Diablo Canyon reactors for a renewed operating license term poses an unacceptable radiological accident risk that jeopardizes my health and safety and the environment.

5. Therefore, I have authorized EWG to represent my interests by submitting a hearing request in this license renewal proceeding.

Executed in Accordance with 10 C.F.R. § 2.304(d) by Patricia Kohlen Date: Feb. 24, 2024

TINITED STATES OF AMERICA NUCLEAR REGULATORY COMMI S SION BEFORETHE SECRETARY In the matter of Pacific Gas and Electric Company Diablo Canyon Nuclear Power Plant Units I and?

Docket Nos. 5A-275-L& 5A-373-LR DECLARATION OF LINI}A PARKS Under penalty of perjury, Linda Parks declares as follows:

t. My name is Linda Parks. I am a supporter of Environmental Working Group (EWG). As an EWG supporter, I receive monthly EWG emails that provides me with public education information about health risks posed by toxic and radiological contamination of consumer products and the environment. In these emails, EWG regularly asks about supporters' concerns and opinions in order to guide it in establishing organizational goals. I am pleased with EWG'S work and believe EWG generally represents my interests and concerns.

2. I live at 242Butte Drive, Los Osos. My home is located within the SO-mile ingestion pathway zone of Diablo Canyon Unit I and Unit 2 nuclear reactors (DCPP).

3. I am aware that the licensee, Pacific Gas and Electric Company (PG&E), has requested the U.S. Nuclear Regulatory Commission (NRC) to renew the DCPP operating licenses for 20 years.

4. Based on my knowledge that the Diablo Canyon nuclear reactors are operating with aging equipment, including an embrittled reactor pressure vessel that has not been inspected for over 20 years; and based on the location of the Diablo Canyon reactors on or neax a web of significant earthquake faults, I believe that the continued operation of the Diablo Canyon reactors for a renewed license term poses an unacceptable radiological accident risk that jeopardizes my health and safety and the environment.

5. Therefore,I have authorized EWG to represent my interests by submitting ahearing request in this license renewal proceeding.

Evec.uted in Acc"wdance with lQ Q.f,fi,, {2.30.4.(d,\ -b}t Date: March 1,2024

From:

Diane Curran To:

Docket, Hearing; Bessette, Paul M.; ryan.lighty@morganlewis.com; Timothy J. Matthews (timothy.matthews@morganlewis.com); Jeremy Wachutka; Catherine Kanatas; Adam Gendelman Cc:

Templeton, Hallie; Caroline Leary

Subject:

[External_Sender] RE: Filing of hearing request in Diablo Canyon license renewal proceeding, Docket Nos. 50-275-LR and 50.323-LR Date:

Monday, March 04, 2024 9:52:56 PM Attachments:

2024.03.04 Exhibit 2 Bird Declaration.pdf Exhibit 4 CCC Letter.pdf Attached please find Exhibits 2 and 4 to Petitioners hearing request.

From: Diane Curran <dcurran@harmoncurran.com>

Sent: Monday, March 4, 2024 9:50 PM To: Docket, Hearing <Hearing.Docket@nrc.gov>; Bessette, Paul M.

<paul.bessette@morganlewis.com>; ryan.lighty@morganlewis.com; Timothy J. Matthews (timothy.matthews@morganlewis.com) <timothy.matthews@morganlewis.com>; Jeremy Wachutka (Jeremy.Wachutka@nrc.gov) <Jeremy.Wachutka@nrc.gov>; Catherine E. Kanatas (catherine.kanatas@nrc.gov) <catherine.kanatas@nrc.gov>; Gendelman, Adam

<Adam.Gendelman@nrc.gov>

Cc: Templeton, Hallie <HTempleton@foe.org>; Caroline Leary <cleary@ewg.org>

Subject:

Filing of hearing request in Diablo Canyon license renewal proceeding, Docket Nos. 50-275-LR and 50.323-LR

Dear NRC Secretary and parties,

On behalf of San Luis Obispo Mothers for Peace, Friends of the Earth, and Environmental Working Group, I am filing a hearing request regarding PG&Es application for renewal of the Diablo Canyon operating licenses.

I am sending the hearing request by email because I was not able to request access to the electronic docket in time for this filing. I will post the hearing request to the electronic docket as soon as I get access.

In the meantime, I am sending the hearing request in three separate emails out of concern that the exhibits may be too large to send in one message:

This email has the hearing request itself with Exhibit 1 attached.

The second email will have Exhibits 2 and 4.

The third email will have Exhibit 3.

Sincerely, Diane Curran Counsel for SLOMFP

Diane Curran Harmon, Curran, Spielberg & Eisenberg LLP

Contact (240) 393-9285 1725 DeSales Street, NW, Suite 500, Washington, DC 20036 Email DCurran@harmoncurran.com

EXHIBIT 2 - DECLARATION OF PETER BIRD

EXHIBIT 4 - CCC LETTER

STATE OF CALIFORNIA - NATURAL RESOURCES AGENCY GAVIN NEWSOM, GOVERNOR CALIFORNIA COASTAL COMMISSION ENERGY, OCEAN RESOURCES AND FEDERAL CONSISTENCY 455 MARKET STREET, SUITE 300 SAN FRANCISCO, CA 94105-2421 VOICE (415) 904-5200 FAX (415) 904-5400 December 7, 2023 Mr. Tom Jones Senior Director - Regulatory, Environmental and Repurposing Pacific Gas & Electric Company Diablo Canyon Power Plant P.O. Box 56 Avila Beach, CA 93424 RE: Pacific Gas & Electric Companys Requested Nuclear Regulatory Commission License Renewal for Diablo Canyon Power Plant, San Luis Obispo County - Incomplete Consistency Certification

Dear Mr. Jones:

Thank you for your submittal of the above-referenced consistency certification for the Diablo Canyon nuclear power plant license renewal that Pacific Gas and Electric Company (PG&E) is seeking from the Nuclear Regulatory Commission. The California Coastal Commission (Commission) received your consistency certification on November 8, 2023.

Our review shows the certification is not yet complete, for the reasons provided below.

Pursuant to the Coastal Zone Management Acts (CZMAs) implementing regulations at 15 CFR 930.58, we will need the information requested herein to allow adequate consideration of the likely coastal effects of the proposed federal action. Accordingly, and pursuant to 15 CFR 930.60(a), the Commissions six-month review period has not commenced and will not commence until we receive the missing necessary data and information.1 Pursuant to requirements of the CZMA, we have identified below the information needed for PG&Es consistency certification to be deemed complete. We are happy to discuss the information requests and review process identified in this letter, and we recommend you coordinate with us on the requests herein prior to your next submittal.

Required Additional Information Requests and Analyses Standard of review: Pursuant to CZMA Section 307(c)(3), the Commissions federal consistency review is to consider whether the proposed federal action would affect any land or water use or natural resource of the coastal zone and whether the activity would be consistent with the enforceable policies of the states approved California Coastal 1 In addition, SB 846, with which the state authorized and supported Diablo Canyons proposed extended operations, requires submittal of a complete application. It states, in relevant part: Notwithstanding any other law, the state agency shall take final action on the application or request to extend the operations of the Diablo Canyon powerplant within 180 days of submission of a complete application or request. [emphasis added]

Proposed PG&E NRC license renewal for Diablo Canyon Power Plant Notice of Incomplete Consistency Certification - December 7, 2023 2

Management Program (CCMP). The main enforceable policies of the CCMP are the Chapter 3 policies of the Coastal Act. Cal. Public Resources Code § 30008. When a project requires a federal license or permit, for purposes of the CZMA, the Commission must consider the impacts of the project as a whole under the Chapter 3 policies. Section 307(c)(3)(A) further requires the applicant to provide to a state agency a copy of the consistency certification, with all necessary information and data, for the purpose of state concurrence with or objection to the applicant's certification.

An action by the NRC to approve the proposed license extension would affect several coastal uses and resources that are components of the CCMP; however, as discussed below, PG&Es consistency certification does not adequately address these effects.

Additionally, the certification states that particular CCMP enforceable policies2 are not applicable to the proposed federal action because PG&Es license extension does not include development, as that term is defined in in the CCMP (Coastal Act section 30106).3 However, the threshold for the Commissions consistency review for projects requiring a federal license or permit is not development, but is any activity... affecting any land or water use or natural resource of the coastal zone. 16 U.S.C. § 1456(c)(3)(A)

[emphasis added]. The proposed extension of the operating license would affect a variety of coastal uses and resources beyond those acknowledged by PG&E in its consistency certification, and the Commission must evaluate those effects for consistency with the applicable policies of the CCMP. Moreover, as described below, the extended intake and discharge of seawater that would be carried out as part of license renewal constitutes development within the meaning of Section 30106. It is also reasonably foreseeable that the proposed license extension includes other uses and activities that likely constitute development within the meaning of Section 30106 - for example, intake cove dredging and a new or expanded spent fuel facility.4 Please amend the certifications analyses to include assessments of the proposed extended operations consistency with all relevant CCMP policies and for all CCMP policies applicable to all other known or reasonably anticipated development resulting from license extension (as further described below).

Additionally, the certification applies standards of review to several specific CCMP policies differently than those prescribed by the policies. For example, and as detailed below, the certification states that the entrainment of marine life from Diablo Canyons use of 2 These include, for example, CCMP Sections 30212, 30240, 30250, 30251, 30252, and 30253.

3 Section 30106 states, in relevant part: "Development means, on land, in or under water, the placement or erection of any solid material or structure; discharge or disposal of any dredged material or of any gaseous, liquid, solid, or thermal waste; grading, removing, dredging, mining, or extraction of any materials; change in the density or intensity of use of land, including, but not limited to, subdivision pursuant to the Subdivision Map Act (commencing with Section 66410 of the Government Code), and any other division of land, including lot splits, except where the land division is brought about in connection with the purchase of such land by a public agency for public recreational use; change in the intensity of use of water, or of access thereto; construction, reconstruction, demolition, or alteration of the size of any structure, including any facility of any private, public, or municipal utility; and the removal or harvesting of major vegetation other than for agricultural purposes, kelp harvesting, and timber operations which are in accordance with a timber harvesting plan submitted pursuant to the provisions of the Z'berg-Nejedly Forest Practice Act of 1973 (commencing with Section 4511)...

4 These are examples and not intended to be exhaustive.

Proposed PG&E NRC license renewal for Diablo Canyon Power Plant Notice of Incomplete Consistency Certification - December 7, 2023 3

seawater does not adversely affect coastal resources because entrainment does not result in population-level effects; however, the standard of review for CCMP policies 30230 and 30231 in determining if coastal resources are affected due to entrainment is based on whether there is a loss or reduction of marine life productivity. Further, the certification in some instances relies on other standards that do not necessarily apply to this review for CCMP consistency. For example, it references NRCs Generic Environmental Impact Statement (GEIS) to describe some of the expected environmental effects of its proposed extended operations;5 however, the standards used in the GEIS are generally not relevant to determine conformity with the applicable policies of the CCMP.

Please revise the certification to apply the appropriate standards of review in the relevant analyses - i.e., whether results of the proposed federal action would affect any coastal use or resource and whether other known or anticipated development associated with the proposed license extension would be consistent with relevant CCMP policies.

Scope of review: The federal action being evaluated by the NRC is a 20-year license extension for operating Diablo Canyon Power Plant. Our review will therefore consider effects on coastal uses and resources for that 20-year period. PG&Es certification states in several places that it intends to operate for about five years and appears to evaluate the projects effect over just that shorter operating period (see, for example, the alternatives analysis described below). Where necessary, please augment the certification with analyses that consider the effects on coastal resources and uses over the full 20-year license period requested in PG&Es application to the NRC.

Development related to proposed license extension: PG&E states that it plans no activities that constitute development under the Coastal Act and states that certain CCMP policies (such as Sections 30212, 30240, 30250, 30251, 30252, and 30253) specifically referring to development are inapplicable.6 However, as noted above, some activities resulting from the proposed federal action would constitute development under the Coastal Act and additional development activities may also be triggered by the extension of power plant operations.7 Therefore, please describe and evaluate any known or potential development activities that would result from relicensing and extended operations. This may include installation of new or expanded structural components, equipment, or infrastructure to assure safe and reliable operations during the extension period - e.g., upgraded seismic safety components based on results of upcoming safety 5 Generic Environmental Impact Statement for License Renewal of Nuclear Plants. NRC. 2013a. NUREG-1437, Vols. 1. Revision 1. ADAMS Accession No. ML13106A241. June 2013.

6 In its consistency certification, PG&E argues that material presented by Commission staff at a February 10, 2023 Listening Session implied that staff determined that there is no development associated with any license extension. This misconstrues the material that Commission staff presented. That material discussed the process and timeline specifically for the Commissions consistency certification review process, which is different than the Commissions CDP review process. Commission staff did not determine that there would be no development requiring a CDP for a license extension.

7 We will discuss with you separately the requirements and process for any CDP that may be needed, which could include a combined review process by which federal consistency and permitting review are conducted simultaneously with the same hearing and in the same 180-day review period established in SB 846.

Proposed PG&E NRC license renewal for Diablo Canyon Power Plant Notice of Incomplete Consistency Certification - December 7, 2023 4

and reliability reviews. Please include these types of development in the requested description and analyses.

Additional documentation: We have requested below several studies and analyses that we will need to evaluate the proposed license extension for CCMP conformity. In addition to those requested documents, we plan to rely in part on information contained in the 2023 San Luis Obispo County Draft Environmental Impact Report (DEIR) on Diablo Canyon decommissioning. That document provides detailed descriptions and maps of existing site conditions, areas of sensitive habitats, wildlife species, and other baseline information we will use in our evaluation of this certification.

Filing fee: In 2008, the CCMP was amended to require filing fees for consistency certifications.

As provided in that amendment, [t]he filing fees for federal consistency certifications constitute necessary data and information within the meaning of 15 C.F.R. Sections 930.58(a)(2) and 930.76(a)(3). The Commissions fees for consistency review are determined in the same manner as for coastal development permit applications. We understand from our conversation earlier today that PG&E will soon be submitting the necessary fee of $367,750 based on project costs as described in the current Coastal Commission Filing Fee Schedule, Section 13055(a)(5)(B), available at this link:

https://documents.coastal.ca.gov/assets/cdp/CDP_Application_Form_Energy.pdf Procedural Controls: Section 9.6 of the Environmental Report accompanying PG&Es certification states that DCPP would have procedural controls in place to protect sensitive resources during operation, including the following: Best Management Practices (BMPs),

avoidance/minimization measures, and environmental reviews prior to conducting any ground-disturbing activities. Please provide all information related to these procedural controls.

Requirements of Other Resource Agencies: Please provide anticipated timelines for receiving approval of all permits, permissions, or approvals required by resource agencies for license renewal. Additionally, please provide any studies or supplemental information specifically prepared as part of any application to another resource agency for license renewal that was not included in your consistency certification submittal to the Commission.

Public Access and Recreation (CCMP Sections 30210 through 30224)

The certification states that license renewal would be consistent overall with the CCMPs public access and recreation policies because it does not involve development and would be a continuation of existing operations. As noted above, however, the standard of review is not just whether there is development, as defined by the CCMP but also whether the federal action would affect any coastal use or resource. For this proposed federal action, approval of a license extension would eliminate substantial public access and recreation opportunities that would otherwise occur but for the license extension. For example, and as noted in the certification, Diablo Canyon includes an Owner Controlled Area (OCA) around the facility and a security zone extending 2,000 yards (or just over one mile) offshore in coastal waters adjacent to the facility. But for this proposed license renewal, that security zone would be eliminated or reduced in size much sooner than otherwise and would allow public access and recreation within those coastal waters. Additionally, and as noted in the above-referenced DEIR for Diablo Canyon decommissioning, PG&E states that it intends to reduce the size of that OCA by about 2034 when decommissioning is complete to encompass only the remaining

Proposed PG&E NRC license renewal for Diablo Canyon Power Plant Notice of Incomplete Consistency Certification - December 7, 2023 5

facilities - primarily the spent fuel facility - pursuant to NRC requirements. With the proposed license extension, decommissioning and the opportunities for public coastal access and recreation will be delayed.

The certification also concludes without adequate analysis that it would not be appropriate to require the public access needed for CCMP conformity due to concerns about public safety, fragile coastal resources, or agriculture. However, as the Coastal Commission has determined in previous CDP and consistency reviews at Diablo Canyon, public access and recreation is both allowed and required on Diablo Canyon lands to ensure conformity with these policies, and PG&E has successfully accommodated public access and recreation while taking into account concerns about public safety and protection of fragile coastal resources and agriculture.

Therefore, for those policies where PG&E has made this conclusion of consistency, please analyze public access and recreation opportunities to ensure that the proposed relicensing will be conducted in a manner consistent with the [CCMP], as required under CZMA Section 307(c)(3)(A).

Protection of Marine Biological Resources (CCMP Sections 30230 and 30231)

PG&Es certification states, in part, that license renewal would be consistent with Section 30230 because it will not increase the capacity or output of the Plant or involve any change in operations compared to existing conditions, such that OTC entrainment, impingement, and thermal impacts would not increase, and the protection of the beneficial uses of the receiving water will continue during the LR period.8 The certification also states that license renewal will be consistent with Section 30231 because existing operations will remain unchanged and will be governed by a wastewater discharge permit, as are current operations at the plant.

These statements do not fully acknowledge the requirements of Sections 30230 and 30231.

For example, Section 30230 requires that marine resources be maintained, enhanced, and where feasible restored. It also requires that uses of the marine environment be conducted in a manner that sustains biological productivity in coastal waters. A key aspect of those policies applicable to the proposed license renewal is that the biological productivity of coastal waters be maintained and, where feasible, restored by minimizing the adverse effects of entrainment.

The proposed license renewal would result in substantial losses of marine life productivity that, but for the proposed extension, would not occur. The certification cites PG&Es most recent entrainment study showing that Diablo Canyons use of seawater for cooling results in a loss of productivity equal to almost 700 acres of rocky reef habitat each year.9 However, the actual loss is substantially higher, as that study used a calculation based on a 50% confidence level instead of the 95% confidence level now used by the Coastal Commission and other state agencies to determine the type and extent of entrainment impacts resulting from seawater intakes. Applying the 95% confidence level results in a loss of productivity during each year of operations roughly equal to that provided by well over 1,000 acres of reef habitat.

8 The certification further states in this regard that SB 846 determined that Diablo Canyon is considered an existing facility under 14 CCR 15301; however, this regulation relates to review under the California Environmental Quality Act, not the Coastal Commissions CZMA review.

9 PG&E conducted entrainment studies in 1996-1999 and in 2008-2009.

Proposed PG&E NRC license renewal for Diablo Canyon Power Plant Notice of Incomplete Consistency Certification - December 7, 2023 6

The certification also references a mitigation measure PG&E has used to partially address this impact - i.e., participation in the interim mitigation program that is part of the State Water Resources Control Boards (Water Boards) Once-Through Cooling (OTC) Policy. The certification states that PG&E demonstrates compliance with that Water Board program by paying an annual in-lieu fee that is then directed to be used for various projects. It is not clear, however, that those projects adequately address the extent of entrainment losses noted above or provide suitable compensatory mitigation.10 Additionally, that interim mitigation program was developed to address the relatively short-term impacts that occur during the few years power plants need for transitioning to cooling methods that do not rely on seawater, not the 20 years that could result from PG&Es proposed license extension.11 Finally, the Commission retains its independent authority to ensure consistency with the policies of the CCMP.

We plan to use the above-referenced 2008-09 entrainment study to evaluate the license extensions expected adverse effects on marine life productivity and to identify mitigation measures necessary to conform to CCMP Sections 30230 and 30231. Please inform us if PG&E would instead prefer to collect additional entrainment data and conduct an updated entrainment study. As part of our mitigation assessment, please also specify the type and scope of mitigation projects have been implemented using PG&Es in-lieu fee funds and provide an assessment of the compensation for the above-referenced annual productivity losses provided by those programs. Please include with that assessment other feasible mitigation measures PG&E would consider implementing to address any mitigation shortfalls.

We understand, too, that PG&E and other resource agencies have identified that Diablo Canyons thermal discharge is having adverse effects on nearby populations of black abalone, a species of special biological significance protected under the Endangered Species Act.

Please provide relevant studies that describe those effects and identify the mitigation measures PG&E has implemented, or will implement, to avoid and reduce those effects.

Finally, CCMP Section 30231 requires protection and productivity of wetlands. The Environmental Report accompanying PG&Es certification states, in Section 3.7.2.4.1, that two wetland delineations have been conducted in the vicinity of the DCPP site. However, those delineations appear to show a significantly smaller area of CCMP-designated wetlands than wetlands designated by other resource agencies (e.g., the U.S. Army Corps of Engineers and California State and Regional Water Boards). This appears to be incorrect, as wetlands delineated pursuant to the CCMP are based on the presence of at least a single parameter (vegetation, soils, or hydrology) while wetlands delineated pursuant to other resource agencies 10 See the Water Boards Draft Determination to Approve Mitigation Measures for the Water Quality Control Policy on the Use of Coastal and Estuarine Waters for Power Plant Cooling (Once-Through Cooling Policy:

Diablo Canyon, n.d. This document also identified expected the one-time cost (in 2006 dollars) for creating or restoring 690 acres of this habitat at $86.25 million. For comparison, it appears that PG&E has paid less than half this amount into the interim mitigation program since it was implemented.

11 As noted in the certification, the State Water Resources Control Board is considering a proposal to extend PG&Es compliance date under the OTC Policy from the end of Diablo Canyons current NRC licenses to October 31, 2030. However, this is still in draft form and does not encompass the full 20-year license extension period being considered by the NRC.

Proposed PG&E NRC license renewal for Diablo Canyon Power Plant Notice of Incomplete Consistency Certification - December 7, 2023 7

require evidence of all three parameters. As such, an area with wetlands usually includes more CCMP-delineated wetlands than those designated by other agencies. Please provide additional information on how the wetlands were delineated and revise the results of the delineation as necessary to reflect the CCMP single-parameter method for wetland delineations.

Dredging (CCMP Section 30233)

Section 30233 of the CCMP includes several criteria applicable to proposed dredging activities in coastal waters. As noted above, the certification acknowledges that during the license extension period, PG&E may need to conduct intake cove dredging. While PG&E has not needed to dredge during previous Diablo Canyon operations, it appears that oceanographic conditions may have changed in a manner that will result in additional intake cove sedimentation and the need for periodic dredging to support operations during the license extension period. Please provide any analyses that PG&E has completed describing recent sedimentation rates in the intake cove and expected sedimentation rates during the license extension period. Please also provide any analyses conducted to help determine expected dredging needs and timing during the license extension.

Section 30233 also requires that there be no feasible and less damaging alternatives to the activities such as dredging that may be permitted by this CCMP policy. The alternatives analysis PG&E provided in Section 2.6 of the Environmental Report accompanying the consistency certification appears to consider various alternatives based on just a potential five-year operating period rather than the full 20-year license extension period. Please provide a revised alternatives analysis that incorporates activities expected over the full 20-year period.

Commercial and recreational fishing (CCMP Section 30234.5)

PG&Es certification states that this CCMP section does not apply to the proposed license extension because there are no fishing or boating industries in the area. However, Section 30234.5 requires that fishing activities overall be acknowledged and protected. As noted above, the license extension would extend for up to 20 years the security zone in coastal waters near Diablo Canyon that, but for the extension, would be reduced or eliminated. This would likely make this area available for commercial and recreational fishing. Please provide an evaluation of fishing opportunities that would otherwise be available near Diablo Canyon and describe what measures PG&E would implement to protect fishing opportunities during an extended licensed operating period.

Environmentally Sensitive Habitat Areas (CCMP Section 30240)

Environmentally Sensitive Habitat Areas (ESHA) are areas in which plant or animal life or their habitats are either rare or especially valuable because of their special nature or role in an ecosystem and which could be easily disturbed or degraded by human activities. CCMP Section 30240(a) states that ESHA shall be protected against any significant disruption of habitat values, and only uses dependent on those resources shall be allowed within those areas.

CCMP Section 30240(b) states that development in areas adjacent to ESHA shall not degrade those areas or be incompatible with their continued presence.

PG&Es certification states that the license renewal is consistent with Section 30240 because there is no development associated with the extension. However, the Environmental Report states that PG&E currently employees a Vegetation Management Plan and a Nesting Bird

Proposed PG&E NRC license renewal for Diablo Canyon Power Plant Notice of Incomplete Consistency Certification - December 7, 2023 8

Management Plan to ensure that plant operations minimize adverse impacts to vegetation and nesting birds. As such, it is reasonably foreseeable that extended plant operations have the potential to affect vegetation and bird species. Please provide a copy of the Vegetation Management Plan and the Nesting Bird Management Plan.

Traffic and Circulation (CCMP Sections 30252 and 30253)

CCMP Section 30253(d) requires development to minimize energy consumption and vehicle miles traveled while Section 30252 requires development to facilitate transit services and non-automotive circulation. The typical workforce at the DCPP site consists of approximately 1,633 employees, including 1,222 permanent full-time employees and an additional 411 supplemental staff who support plant operations. Approximately 84 percent of the permanent employees reside in the County of San Luis Obispo, with 11 percent of employees residing in Santa Barbara County and the remaining 5 percent residing in various other locations. During refueling outages, which usually last approximately 35 days per unit, there are typically an additional 500 to 600 temporary contract employees onsite, but there can be as many as 1,000 additional workers depending on the scope and conditions at the time of the outage.

Section 3.9.6 of the Environmental Report discusses transportation in the DCPP region. In 2021 PG&E performed a transportation assessment for DCPP to identify level of service (LOS) based on average daily traffic counts in the immediate area and found that all roadway segments were operating at a LOS of C, which according to the County of San Luis Obispo (County) is acceptable for rural areas. However, the section concludes by stating that the transportation assessment was related to decommissioning, including construction activities and employment changes. Please clarify how the LOS determined in the decommissioning transportation study relates to anticipated LOS during license renewal, including refueling outages, and whether any roadway segments would operate below an acceptable LOS as a result of license renewal and/or refueling outages.

Minimizing risks (CCMP Section 30253)

The proposed extended operations would take place at a location and facility subject to relatively high levels of seismic and other hazards. PG&Es certification states that CCMP Section 30253 does not apply to the proposed license renewal because there is no associated development. As noted above, however, it is reasonably foreseeable that the license extension includes development needed to ensure reliable operations that will result in effects on coastal uses and resources. To allow for adequate review of potential effects, please provide the following hazards-related documents and analyses:

• Seismic Hazards o Please provide the document referenced in hazards discussions in Appendix E: PG&E.

2021a. Diablo Canyon Power Plant Units 1 and 2, Updated Final Safety Analysis Report, Revision 26. October 2021. ADAMS Accession Package No. ML21306A142 and any relevant studies to seismic hazards citied therein.

o Please provide an updated analysis or reevaluation of seismic hazards at DCPP which includes new information and science which has emerged since PG&Es last comprehensive evaluation. The analysis should include an assessment of whether the plant design basis continues to protect against seismic hazards, and what additional development (if any) would be necessary to protect plant facilities and operations against

Proposed PG&E NRC license renewal for Diablo Canyon Power Plant Notice of Incomplete Consistency Certification - December 7, 2023 9

seismic hazards over the proposed relicensing period (at a minimum). The ground-shaking analysis should consider at least the 2,475-yr recurrence interval event (or a deterministic equivalent). Staff understands that a Seismic Review Update required by SB846 is currently underway; subject to staff review, this document may fulfil all or part of the Commissions information needs related to seismic hazards.

• Tsunami Hazards:

o Please provide the document referenced in hazards discussions in Appendix E: PG&E.

2021a. Diablo Canyon Power Plant Units 1 and 2, Updated Final Safety Analysis Report, Revision 26. October 2021. ADAMS Accession Package No. ML21306A142 and any relevant studies to tsunami hazards citied therewithin o Please provide a summary of tsunami analyses completed for PG&E and/or NRC focused on Diablo Canyon site completed since 2000.

o Please provide an analysis of impacts (e.g., how operations would be affected, the potential for risk to personnel onsite) from Maximum Considered Tsunami (at least 2,475-year recurrence interval or deterministic equivalent) that also considers the expected sea level rise scenarios over the project life (i.e., relicensing period) and consideration of flow speed and flow depth in analysis of potential impacts to operations and recovery.

• Waves & Coastal Storms o Please provide the document referenced in hazards discussions in Appendix E: PG&E.

2021a. Diablo Canyon Power Plant Units 1 and 2, Updated Final Safety Analysis Report, Revision 26. October 2021. ADAMS Accession Package No. ML21306A142 and any relevant studies to waves, coastal storms, and seiche hazards citied therein.

o Please provide an analysis of impacts (e.g., how operations would be affected, the potential for risk to personnel onsite) from both 100-year and 500-year total water levels and wave overtopping (if applicable) that also considers the potential for sea level rise over the project life (relicensing period).

• Coastal Erosion o Please provide the document referenced in hazards discussions in Appendix E: PG&E.

2021a. Diablo Canyon Power Plant Units 1 and 2, Updated Final Safety Analysis Report, Revision 26. October 2021. ADAMS Accession Package No. ML21306A142 and any relevant studies to erosion or landslide hazards citied therein o Please provide an analysis of impacts (e.g., how facilities and operations would be affected, the potential for risk to personnel onsite) from both short-term episodic erosion (e.g., landslides or storm-related) and long-term erosion over the proposed relicensing period (at a minimum) that also considers the potential for accelerated erosion from potential sea level rise.

Proposed PG&E NRC license renewal for Diablo Canyon Power Plant Notice of Incomplete Consistency Certification - December 7, 2023 10 Closing Thank you for your attention to these requests. As noted previously, we are happy to discuss these with you and answer any questions you may have.

Sincerely, Tom Luster Energy, Ocean Resources, and Federal Consistency Division

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE COMMISSION AND BEFORE THE SECRETARY In the matter of Pacific Gas and Electric Company Docket Nos. 50-275-LR Diablo Canyon Nuclear Power Plant 50-323-LR Units 1 and 2 DECLARATION OF PETER BIRD, Ph.D Submitted to the U.S. Nuclear Regulatory Commission By San Luis Obispo Mothers for Peace, Friends of the Earth, and Environmental Working Group March 4, 2024

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i TABLE OF CONTENTS I.

EXPERT QUALIFICATIONS..1 II.

PURPOSE AND DESCRIPTON OF MY DECLARATION.......2 III.

BACKGROUND REGARDING PG&E AND NRC SEISMIC STUDIES AND ENVIRONMENTAL DOCUMENTS.....3 A. PG&Es Public Seismic Risk Studies......3 B. Environmental Documents.......3 IV.

SCIENTIFIC ANALYSIS.....4 A. Abstract4 B. Detailed Scientific Argument.......5 (1) Accelerations in the 2024 Noto Peninsula earthquake.....5 (2) Factors responsible for unusually strong shaking.5 (3) Tectonic analogy between the Noto Peninsula and the Irish Hills of California.....6 (4) Thrust-fault slip-rates and earthquake recurrence intervals.....8 (5) Susceptibility of DCPP to seismic core damage..9 (6) Risk of external seismic severe accidents at DCPP has been grossly underestimated...10 C. Figure 1..11 V.

ADDITIONAL OBJECTIONS TO APPLICANTS ENVIRONMENTAL REPORT12 A. Regarding adequacy of existing and planned deformation models.12 B. Regarding status of witness models in the seismicity/hazard communities...13 VI.

REFERENCES..15 A. References cited in Sections I-IV..15 B. References cited in Section V16

ii VII.

CURRICULUM VITAE..18 VIII. LIST OF PUBLICATIONS.20

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE COMMISSION AND BEFORE THE SECRETARY In the matter of Pacific Gas and Electric Company Docket Nos. 50-275-LR Diablo Canyon Nuclear Power Plant 50-323-LR Units 1 and 2 DECLARATION OF PETER BIRD, Ph.D Under penalty of perjury, I, Peter Bird, declare as follows:

I.

EXPERT QUALIFICATIONS

1.

My name is Peter Bird. For over 46 years, I have been a Professor of Geophysics and Geology at the University of California at Los Angeles (UCLA). I now serve as Professor of Geophysics and Geology, Emeritus at UCLA. I am qualified by training and experience as an expert in the fields of geology and geophysics with a focus on tectonophysics and seismicity, including plate motion and plate deformation. A copy of my curriculum vitae is included here as Attachment 1.

2.

I have a Ph.D. in Earth and Planetary Sciences from the Massachusetts Institute of Technology (1976) and a B.A. in Geological Sciences from Harvard College (1972).

Over the past 48 years, I have published 76 academic papers, mostly about tectonics and seismicity, including the tectonics and seismicity of California. And I have authored or contributed to a number of academic papers on computer modeling methods and applications, including studies of the ongoing (neotectonic) deformation in California. I have also been a member or officer of several professional organizations relating to my expertise, including the Geological Society of America, the American Geophysical Union and the Southern California Earthquake Center.

The former two organizations have recognized my work with two fellowships and an award.

3.

In 2012, I participated in a Senior Seismic Hazards Analysis Committee (SSHAC) workshop sponsored by Pacific Gas & Electric Co. (PG&E) and run by Lettis Consultants International, regarding seismic hazard at the Diablo Canyon Power Plant. I presented results on both strike-slip and compressional deformation rates affecting the region, which were derived from my latest computer models of neotectonics. These models were prepared for the Southern California Earthquake Centers project Unified California Earthquake Rupture Forecast version 3, and also for the US Geological Surveys 2013 Update to the National Seismic Hazard Model.

2

4.

On April 28, 2023, on behalf of San Luis Obispo Mothers for Peace (SLOMFP), I prepared a declaration setting forth my criticism of the seismic risk analysis for DCPP that was presented by the U.S. Nuclear Regulatory Commission (NRC) in its Draft Generic Environmental Impact Statement for License Renewal of Nuclear Plants (NUREG-1437, Rev. 2, Feb. 2023) (Draft GEIS) (NRC 2023). SLOMFP submitted my declaration with its comments on the Draft GEIS on May 2, 2023. My declaration can be accessed on the NRCs Agencywide Data Access and Management System (ADAMS) at ML23123A410. My declaration in that rulemaking proceeding is relevant to this DCPP license renewal proceeding because the NRC relied heavily on PG&Es seismic analyses for its conclusion that the environmental impacts of an earthquake-induced or related accident at DCPP are SMALL. This matter is discussed in more detail below. I continue to stand by the facts and expert opinions expressed in my declaration.

II.

PURPOSE AND DESCRIPTION OF MY DECLARATION

4. The purpose of my declaration is to explain why, in my expert opinion, the Environmental Report by applicant PG&E significantly underestimates the likelihood of a severe earthquake at DCPP, i.e., an earthquake that could cause substantial damage to the reactor core. [ER p.

4-61]. PG&Es 2018 estimate and 2023 revision of the long-term rate of seismic core damage as 2~3x10-5 /yr fail to take into account current information or to deploy a technically-defensible seismicity model that show the seismic severe accident rate is about 47~70 times higher, or ~1.4x10-3/ year.

5. The fundamental problem with PG&Es seismic risk analysis is not any error in computations, but the use of incomplete deformation models to support the 2015 Seismic Source Characterization (SSC). These incomplete deformation models also biased PG&Es 2018 seismic probabilistic risk assessment (SPRA). PG&E mistakenly decided that strike-slip faulting is the only important kind of neotectonic activity in the vicinity of DCPP.1 As I have previously discussed, these deformation models do not meet basic scientific standards for objectivity and reliability because are not geometrically self-consistent, nor are they consistent with GPS and regional stress directions. Instead, they appear to be custom-built to minimize seismic hazard at DCPP.

6. In my expert opinion, thrust-faulting (due to horizontal compression of the crust) is an equal contributor to overall seismicity in this area. More importantly, it implies a far greater increase in expected SCDF at DCPP due to the extreme accelerations that occur in hanging-walls of thrust faults, especially near their tips.

7. The basis for my expert opinion is set forth below, first briefly, and then in detail, following a necessary Background section.

3 III.

BACKGROUND REGARDING PG&E AND NRC SEISMIC STUDIES AND ENVIRONMENTAL DOCUMENTS A. PG&Es Public Seismic Risk Studies

8. PG&Es public seismic risk studies are the post-Fukushima SSC (PG&E, 2015; 2015L) and the resulting SPRA (PG&E, 2018). According to the SPRA: The SPRA performed for DCPP shows that the point-estimate mean SCDF [seismic core damage frequency] is 2.8x10-5 per year (page 52).

9. The seismic model presented in the SSC (PG&E, 2015 SSHAC Level-3) is notable for deformation models that focus almost exclusively on strike-slip faults, neglecting to consider thrust faults under DCPP as dangerous seismic sources.1 This significant omission is addressed in (Bird, 2023) and will be discussed later in my declaration.

B. Environmental Documents

10. PG&Es SCDF estimate was accepted by NRC in the Draft License Renewal GEIS (NRC, 2023). Table E.3-11, entitled Seismic (Full Power) Core Damage Frequency Comparison, lists expected severe seismic accident rates for every nuclear plant in the country. In the row labeled Diablo Canyon 1, 2 the value for the metric SAMA SCDF(a) is 1.3x10-5 /yr, and the value for the metric SPRA Mean SCDF(b) is 2.8x10-5 /yr. The mean of these two metrics is 2x10-5 /yr.

11. Both the Draft License Renewal GEIS and Applicants Environmental Report (PG&E, 2023) describe the expected rate of severe accidents of external seismic origin as SMALL.2 In the Draft GEIS, this characterization can be found at page E-34 (The NRC staff concludes that... external event risk is being effectively addressed and reduced by the various NRC Orders and other initiatives, and that, therefore external event risk is not expected to challenge the 1996 LR GEIS 95th percentile UCB [upper confidence bound] risk metrics during the initial LR [license renewal]... period.) Also see page E-1 (The 1996 LR GEIS concluded that the probability-weighted consequences were small compared to other risks to which the populations surrounding nuclear power plants are routinely exposed.)

12. In the Environmental Report, this characterization can be found in Section 4.15 Postulated Accidents / Section 4.15.2 Severe Accidents, on pages 4-61 (PDF page 455). The more specific statement of SCDF in PG&E (2023) is: As shown in Attachment G, Section G.2.1.17, the DCPP application model used for the SAMA analysis has an internal fire CDF of 4.6 x 10-5 and a seismic CDF of 2.96 x 10-5 which are less than the bounding CDFs in 1 Technically, a few of PG&Es 2015 deformation models did include thrust faults; however, they were uniformly parameterized as steeply-dipping, slow-slipping, not passing below DCPP, limited to low maximum-magnitudes, and/or low-weighted on the logic tree(s). Thus, their net impact on PG&Es SSC and SCDF estimates was insignificant.

2 In my understanding, the term SMALL is equivalent to insignificant from the standpoint of the severity of environmental impacts.

4 Tables E.3-10 and E.3-11. Consistent with NRC's conclusions, these lower fire and seismic CDFs are also not significant compared to the previous LR GEIS revisions. (page 4-62; PDF page 456).

13. For brevity in this Declaration, I will refer to this old estimate as a seismic core damage frequency of 2~3x10-5 /yr; that is, one severe accident of seismic origin per 33,000~50,000 years.

IV.

SCIENTIFIC ANALYSIS A. Abstract

14. The following is an abstract of my scientific analysis:

(1)

The Noto Peninsula earthquake in Japan (2024.01.01, m7.5, 10 km deep) produced peak ground accelerations (PGA) of 1.0~2.3 g (that is, 100~230% of gravity) at 5 modern digital strong-motion seismometers as far as 42 km from the rupture.

(2)

This strong shaking occurred in the Noto Peninsula, which is part of the hanging-wall (upper block) of two en-echelon thrust faults that run parallel to its two coasts.

(3)

The Irish Hills, San Luis Range, and DCPP site in California are at risk for similar earthquakes and similar shaking because they are underlain by similar thrust faults, including the inland Los Osos thrust fault and the Inferred Coastline thrust running along the shore by DCPP.3 (4)

The expected recurrence interval between such events at DCPP can be roughly estimated by dividing the expected fault slip (averaging 2 m in the Noto earthquake, according to the USGS finite-fault solution) by the total heave rate of the thrust faults under DCPP, which is about 2.8 mm/year (as I will justify below). The result is 715 years. The inverse of this is the rate: 1.4x10-3 /yr.

(5)

In the existing SSC (PG&E, 2015; 2015L), the intensity of shaking at this return period of 715 years has been underestimated by a factor of 3~7. This means that the chance of seismic core damage is much higher when thrust-faulting earthquake sources are included.

(6)

Applying my analysis to these facts, the probability of a severe accident of earthquake origin at DCPP has been underestimated by a factor of (1.4x10-3 /yr) / (2~3x10-5 /yr)

= 47~70. In other words, the severe accident that PG&E asserts will occur only once in 33,000~50,000 years may actually occur every ~715 years. That means that a license extension for 20 years would incur a ~2.8% probability of a severe accident.

3 Inferred Coastline thrust is my own term for a distinct fault surface whose trace follows the coastline opposite DCPP. Unlike the Shoreline fault in the same area, the Inferred Coastline thrust dips at a gentle angle beneath DCPP and has the up-dip rake of a thrust fault.

5 B. Detailed Scientific Argument

15. In the following pages, I will demonstrate that PG&Es SCDF estimate is too low, by almost two orders of magnitude. PG&Es error lies in the subjective [i.e., committee-based, not algorithm-based] creation of deformation models that served as the basis for the 2015 SSHAC Level-3 SSC, and their almost total exclusion of shallow thrust faults under DCPP as dangerous seismic sources. While my previous criticisms of PG&Es seismic risk analyses (Bird, 2023) remain valid, it will not be necessary to evaluate every feature of the 2015 SSC here; rather, it will only be necessary to consider the kind of seismic source that was excluded.

(1) Accelerations in the 2024 Noto Peninsula earthquake

16. On 1 January 2024, at 07:10 UTC, a very large earthquake occurred beneath the Noto Peninsula on the northwest coast of Ishikawa Prefecture, Japan. Its magnitude was 7.6 on the moment-magnitude scale used by the Japan Meteorological Agency, and 7.5 on the moment-magnitude scale used by USGS. This thrust-faulting shock achieved a maximum JMA seismic intensity of Shindo 7 and Modified Mercalli intensity of IX (Violent) (Wikipedia, 2024). These intensities are very high.

17. Professor Shinji Toda of Tohoku University collected digital seismograms from the many strong-motion seismograph stations on and around the Noto Peninsula and reported them in Toda and Stein (2024). In their Figure 2, it can be seen that one station 42 km from the rupture experienced peak ground acceleration (PGA) of 230% of g; the next 4 highest PGA values observed were 150%, 140%, 120%, and 100% of gravity.4 Toda & Stein noted that, in general, PGA values for this earthquake were about 4x greater than those anticipated by the well-known USGS ShakeMap algorithm at the same distances.

(2) Factors responsible for unusually strong shaking

18. According to the finite-fault solution computed by the U.S. Geological Survey (USGS, 2024), these high PGA sites were all located in the hanging-wall (upper block) of a thrust fault with SE dip. The reasons why unusually strong shaking should be expected in the hanging-wall of a thrust are well-understood, at least in qualitative terms:

19. First, it is common for thrust-fault ruptures to begin in the zone of highest stress-drop, near the base of the seismogenic zone at ~10 km depth. As the rupture expands up-dip, each 4 PGA, or Peak Ground Acceleration, is obtained from a seismogram either directly (if it is an accelerogram), or by taking the first time-derivative (if it is a velocity seismogram), or by taking the second time-derivative (if it is a displacement seismogram). Either way, it is a seismic acceleration in units of m/s2. However, a common practice in this field of seismic hazard assessment is to normalize PGA by dividing it by the everyday (non-seismic) acceleration of gravity on the surface of the Earth, g = 9.8 m/s2. After this normalization, PGA is expressed in units of g.

6 increment of slip adds its seismic energy to a directivity-pulse of strong shear (S) waves.

Second, this shear-wave energy cannot escape into the atmosphere, because it is perfectly reflected by the free surface. Third, along the active fault at the base of the upper block, shear waves are also partially reflected upward by the low-velocity layer of fault gouge.

Where the fault is actively slipping, higher reflection coefficients are caused by temporary coseismic increases of pore pressure in this gouge layer, and by the fact that the fault has left the elastic domain and is in a state of frictional plasticity. Thus, the shear-wave seismic energy propagating up-dip in the upper block is largely confined to a wedge whose thickness and mass decrease towards its tip (at the fault trace). Fourth, conservation of energy then requires seismic wave amplitude, velocity, and acceleration to increase to high values. In fact, there is a loose analogy to the behavior of shear waves in a whip, where the tip is intended to reach supersonic velocities.

20. A necessary step in every seismic source characterization probabilistic seismic hazard assessment study is the use of ground-motion prediction equations (GMPEs) to estimate shaking from earthquake magnitude, distance, and other geometric factors. One of the most respected sources of GMPEs in the next-generation literature is Campbell & Bozorgnia (2014). This source recognizes the special hazard in the hanging-wall of a thrust; the Abstract states (in part): In addition to those terms included in our now-superseded 2008 GMPE, we include a more-detailed hanging wall model, scaling with hypocentral depth and fault dip,. Below, in their text: The hanging wall term was updated in part by empirically constraining the hanging wall model developed by Donahue and Abrahamson (2013, 2014) from ground motion simulations. In their equation (1), term fhng describes additional intensity for observers in a hanging-wall location. This term is itself the product of 6 factors defined by equations (7-16). Thus, modern practice provides ways to estimate the hanging-wall effect, although these were apparently not used in the 2015 SSC study.

21. Notably, high PGA above a thrust-fault has been observed in California, in the 1971.02.09 San Fernando (or Sylmar) earthquake of m6.6, which had a maximum Mercalli intensity of XI (Extreme). A strong-motion seismogram installed on a bedrock base next to the Pacoima Dam observed PGA of 125% of g (Cloud & Hudson, 1975).

(3) Tectonic analogy between the Noto Peninsula and the Irish Hills of California

22. According to Japanese geological sources summarized by Toda & Stein (2024), the Noto Peninsula is a crustal block that is being uplifted from beneath the Sea of Japan by the joint action of conjugate SE-dipping thrust faults just offshore its NW coast and NW-dipping thrust faults just offshore its SE coast. The driving force comes from horizontal convergence (estimated as ~10 mm/yr) between the island of Honshu and the Eurasia plate (or, more precisely, between the Amur and Okhotsk plates in the PB2002 global model of Bird, 2003).

23. The Irish Hills, San Luis Range, and DCPP site in California occupy a closely analogous tectonic setting, with a SW-dipping active thrust fault (Los Osos thrust) on the NE side, and the NE-dipping Inferred Coastline thrust [my proposed name for purposes of this Declaration] on the southwest side. This basic structure was mostly ignored by PG&E in creating deformation models for the 2015 SSC (PG&E, 2015).

7

24. The Irish Hills and the San Luis Range are a dextral-transpressional orogen that has formed since ~3.5 million years (or mega annus, Ma) [Page et al., 1998], or possibly since 7.8~6 Ma

[Atwater & Stock, 1998; Bird & Ingersoll, 2022] when the motion of the Pacific plate changed its direction to become more compressional relative to North America. This means that the region can be expected to be cut by a number of both strike-slip and thrust (compressional) faults.

25. Evidence of compressional tectonic structures in the region includes the following eight significant elements:

a. The Pismo syncline is the primary structural feature exposed in the Irish Hills [Pacific Gas & Electric, 2014]. Here beds have been rotated ~45, which angle is supported by both mapped surface dips in outcrops (geologic map, ibid), and by the overall dip of unit Tmo Obispo Formation in the borehole-controlled cross-section of Figure 13-17 of the SSC for DCPP. This folding began after deposition of the youngest strata in the core of the fold (Tmpm), and prior to deposition of the Squire Member of the (Pliocene) Pismo Formation (Tpps), probably ~5 Ma. This folding implies upper-crustal strains of ~0.8, and mean strain-rates of ~0.8 / 5 Ma = 5x10-15 per second (/s). This is ~10x faster than rates of off-modeled-fault (or continuum) deformation that are typical in the long-term neotectonics of the western US [5x10-16 /s per Bird, 2009]. This high rate of permanent straining implies a high rate of faulting and of earthquakes, even if the relevant thrust fault traces are not always exposed.

b. According to the geologic map [PG&E, 2014] and associated cross-section C-C in its Fig. 13-17, the apparent throw (vertical offset) of stratigraphic unit Tmo Obispo Formation is 1.6~2.2 km across the Shoreline fault trace. (This measurement is illustrated in my own Figure 1.) None of this can be explained by strike-slip on the Shoreline fault because its slip-rate is very low and because regional strikes of bedding are roughly parallel to it. Instead, the simplest explanation is thrust-faulting on the Inferred Coastline thrust that shares the complex, braided surface trace of the Shoreline fault. Assuming a typical thrust-fault dip of 25, the amount of slip required to create this throw is (1.6~2.2 km) / sin(25) = 3.8~5.2 km. Then, assuming this occurred since ~5 Ma, the mean rate of slip on the Inferred Coastline thrust has been 0.76~1.04 mm/a. To the northwest of section C-C the throw of unit Tmo becomes much less, but the area of neotectonic uplift of the Irish Hills (Figure 7-4 in PG&E, 2015) continues to the northwest; so there the thrust fault probably does not terminate but merely deforms unit Tmo into a fault-initiation anticline above it. (In this area, complex older deformation associated with intrusions of Tmod diabase obscures the Pliocene-Quaternary structure, and makes balanced-section methods inapplicable.) In my professional judgment, this Inferred Coastline thrust fault continues, with the same rake and offset, northwest to the Hosgri fault.

c. The neotectonic uplift rate of the whole Irish Hills region is uniform at 0.2 mm/a (Fig. 7-4 in PG&E, 2015). Because the Franciscan Complex basement is weak, and because

8 there is no large isostatic gravity anomaly over the Irish Hills [Simpson et al., 1986], this uplift process should be modeled with Airy isostasy. The implied rate of crustal thickening is then about 6 times larger, or about 1.2 mm/a. If this crustal thickening is occurring on a single thrust fault of dip 25, then its rate of slip should be (1.2 mm/a) /

sin(25) = 2.8 mm/a. Or, if the crustal thickening is driven by two oppositely-vergent and overlapping thrust faults (as in my schematic section, Figure 1 at the end of this testimony), then each should have a slip-rate of ~1.4 mm/a. Obviously, more complex models with more thrust faults can be devised, but the implication for total strain and seismicity due to thrust-faulting will remain unchanged.

d. The southwestern front of the Irish Hills is a topographic scarp with a smooth arcuate shape, mirroring the slightly-lower scarp on the northeast which has been formed by slip on the Los Osos thrust fault. This suggests that the Inferred Coastline thrust is present under the southwestern front, at or near the coastline.

e. The 2003 San Simeon m6.6 and 1983 Coalinga m6.2 earthquake both had thrust mechanisms [Global Centroid Moment Tensor Catalog, Ekstrm et al., 2012]. This is evidence of highly-compressive horizontal stresses in the Coast Ranges region, suggesting a likelihood of seismic thrust-faulting in other locations as well.

f. SSW-NNE directions of most-compressive stress shown by data in the World Stress Map

[Mueller et al., 1997; Heidbach et al., 2008, 2016], and by interpolation of stress directions using the method of Bird & Li [1996], are almost perpendicular to the traces of the regional fault grain (Shoreline, Inferred Coastline, San Luis Bay, and Los Osos fault traces). This strongly suggests that currently these faults are either purely or dominantly thrust faults.

g. Closer to DCPP, two recent small earthquakes had thrust-faulting mechanisms with the expected SSW-NNE direction of maximum horizontal compression: 2023.12.27 m3.1 at 6.2 km depth under the Irish Hills, and 2024.01.01 m5.4 slightly offshore from the NW end of the Irish Hills (D. J. Weisman, pers. comm., 2024.01.02). This shows that the regional stress regime and orientation documented above also apply in the immediate vicinity of DCPP.

h. Models of neotectonic deformation, informed and guided by GPS velocity data, include such long-term compression. Specifically, Shen & Bird [2022] computed a suite of kinematic finite-element (F-E) models of neotectonics across the western US based on geodetic, geologic, & stress data with program NeoKinema. Their preferred model, which has been incorporated into the 2024 update of the USGS National Seismic Hazard Model, shows convergence of crustal blocks on both sides of the Irish Hills/San Luis Range region at velocities of ~1 mm/a, for a total of ~2 mm/a of local horizontal convergence rate.

9 (4) Thrust-fault slip-rates and earthquake recurrence intervals

26. The paragraphs above contain multiple arguments for horizontal convergence at ~2.0 mm/yr in the Irish Hills area, and for total thrust-fault slip rates of ~2.8 mm/yr. In addition, paragraph 25(b) shows that the slip-rate of the Inferred Coastline thrust must be 0.76~1.04 mm/yr. Therefore, deformation models like some of PG&Es in their 2015 SSC that attribute all uplift and shortening to the Los Osos fault are not defensible.

27. In SSC and PSHA studies that include fault seismic sources with very incomplete information, it is traditional to assume a periodic characteristic earthquake model. While this is only an approximation of the chaotic earthquake dynamics in the real Earth, it has the advantage of allowing simple arithmetical conversions between the triad of basic parameters:

slip, slip-rate, and recurrence interval. For example, to compute the recurrence interval for large characteristic thrust-faulting earthquakes under the Irish Hills (either on the Los Osos or Inferred Coastline thrust), it is sufficient to divide the mean coseismic slip by the long-term tectonic slip-rate.

28. In the 2024 Noto Peninsula earthquake, we have the advantage of the finite-fault solution (USGS, 2024), which maps the amount of coseismic slip onto the active fault plane. This study showed maximum slip of 3.7 m under the center of the Noto Peninsula, with a mean slip that I visually estimate as 2.0 m (or 2000 mm) in the seismogenic depth range.

29. Dividing this mean slip of 2000 mm by the long-term tectonic slip-rate of 2.8 mm/a in the Irish Hills, the inferred recurrence rate for Noto-type earthquakes under the Irish Hills is 715 years. In other words, the inferred probability of Noto Peninsula-type earthquakes under the Irish Hills is the inverse of this, which is 1.4x10-3 /yr.

30. Again, reasonably presuming that the Noto Peninsula earthquake is a characteristic earthquake for this tectonic setting (shared by the Irish Hills in California), PGA values of 1.0~2.3 g (see section 1 above) must be expected with probability 1.4x10-3 /yr. However, in the 2015 SSC (specifically, in Figure 2.3.7-1 of PG&E, 2015L), we see that this outdated modeling associated this probability level with a PGA of only 0.32 g. Consequently, it appears that the 2015 SSC severely underestimated (by a factor of 3~7) the severity of shaking (PGA) that must be resisted every ~715 years.

(5) Susceptibility of DCPP to seismic core damage

31. This raises the question of whether PGA of 1.0~2.3 g will cause seismic core damage (SCD) at Diablo Canyon Units 1 & 2. Answering this question quantitatively becomes technical and difficult, given that spectral accelerations critical to individual component failures are typically twice as large as PGA; that is, perhaps 2.0~4.6 g at vibration frequencies of 5~10 Hz in the Noto Peninsula case.

32. The 2018 SPRA (PG&E, 2018) is the most recent available to me. Within this document, Table 5.4-4 (page 65) shows how the overall SCDF of 2.8x10-5 /yr was obtained. In principle, it should be possible to use this information to estimate the probability of SCD at

10 each level of shaking. My interpretation of the table is that the probability of SCD is ~6% at 2 g, rising to ~73% at 3 g and to >98% at 4 g. The problem is that the acceleration levels quoted in this table are not clearly identified; are they PGAs or (more likely) spectral accelerations? The context in this SPRA report suggests that they are spectral accelerations:

the introductory section 3.1.3 Seismic Hazard Analysis Results and Insights only discusses 5 Hz spectral accelerations, and the primary graphs that it refers to (Figure 3 Reference Rock Hazard by Source for 5 Hz Spectral Acceleration and Figure 3 5 Hz Control Point Mean and Fractiles Horizontal Hazard) are plots of 5 Hz spectral acceleration.

33. Therefore, my interpretation of these reports is that a PGA event of 1.0 g would produce 5 Hz spectral accelerations of ~2 g, and incur ~6% of SCD. However, a PGA event of 1.5 g would produce 5 Hz spectral accelerations of ~3 g, and incur a ~73% chance of SCD. And the peak Noto-earthquake observation of PGA of 2.3 g would produce spectral accelerations of ~4.6 g, and incur >98% chance of SCD.

34. It will probably be controversial exactly which of the Noto Peninsula seismograms give the median and worst-case forecasts of shaking at DCPP. The paragraph above shows that this is a critical point. Clearly these questions need to be resolved by independent experts, preferably in a revised SSC study followed by a revised SPRA study. In the meantime, for purposes of evaluating PG&Es Environmental Report, it is reasonable to assume that the levels of shaking seen in the Noto Peninsula earthquake will cause seismic core damage at DCPP if and when they occur in the Irish Hills of California.

(6) Risk of external seismic severe accidents at DCPP has been grossly underestimated

35. The combined implication of the above-cited facts and analysis is that the probability of a severe accident of earthquake origin at DCPP has been underestimated by a factor of (1.4x10-3 /yr) / (2~3x10-5 /yr) = 47~70. In other words, the severe accident that PG&E asserts will occur only once in 33,000~50,000 years may actually occur every ~715 years.

That means that a license extension for 20 years would incur a ~2.8% probability of a severe accident.

11 C. Figure 1 Figure 1. Revised geologic section through the Irish Hills near DCPP. The base for this figure is Figure 13-17 of the Seismic Source Characterization for DCPP (PG&E, 2015). Note that the fault dips suggested by black lines in their figure were not based on data, but were constrained by PG&Es (2015) a priori assumption that only strike-slip tectonics is active in the area. In red, I have suggested more plausible 25 dips for the Los Osos thrust (at right/North) and the Inferred Coastline thrust (at left/South). The upper-left portion of this figure is also edited to show the throw (vertical offset) of map unit Tmo across the Inferred Coastline thrust, discussed in my text paragraph IV.B.25(b).

12 V.

ADDITIONAL OBJECTIONS TO APPLICANTS ENVIRONMENTAL REPORT A. Regarding adequacy of existing and planned deformation models

36. In my previous Declaration (2023.04.28) to NRC regarding their Draft Generic EIS (NRC, 2023), and in my Testimony (2023.06.30) to the California Public Utilities Commission regarding DCPP, I raised objections to the methodology of the SSC for DCPP (PG&E, 2015):

The 2015 SSC for DCPP was deficient and biased in 3 ways: (1) Fault slip-rates were selected subjectively and in isolation, without modern deformation-modeling (as used by USGS) to guarantee that all fault slip-rates and rates of distributed permanent deformation are self-consistent, and also consistent with geodetic-velocity and stress-direction data; (2) Seismicity from unexpected, undetected, and/or subterranean ruptures between the known faults was modeled based on projection of a few decades of microseismicity, ignoring globally-calibrated relationships between long-term tectonic strain-rate and (typically higher) long-term-mean seismicity which includes seismic crises; and (3) Despite several arguments and proposals for a thrust fault at shallow depths under DCPP with slip-rate of ~1 mm/a, no such seismic source was included.

Point (3) has been expanded in Section I of this Testimony, above.

37. However, I wish to restate my objections (1) and (2) above, because both systematic defects in deformation-modeling have the potential to seriously bias the estimated seismic hazard.

38. The response from PG&E appears in the following paragraph on page G-27 of Attachment G to Applicants Environmental Report (PG&E, 2023):

New or updated seismic methodologies and models developed since preparation of the SSC model will be considered as part of the SB-846-required seismic update. The DCPP seismic analyses, however, include a variety of well-established and vetted models rather than a single method. Therefore, additions or changes in data input from a single model typically result in slight to moderate changes in hazard calculations. If proposed new methods or models are determined to be viable and reliable, they will be integrated with other models so the impact of any single change is not expected to result in a significant change in the resulting seismic hazard.

39. The strong implication here is that PG&E intends to keep their old deformation models from 2015, and perhaps add one or two alternative deformation models (probably with small logic-tree weights), so that there is no material change in net seismic hazard. Actually, in a public presentation to the Diablo Canyon Independent Safety Committee of the California Public Utilities Commission on 23 February 2024, the PG&E presenters indicated that there would be no new deformation models, and the geometry of the old deformation models would be unchanged. As discussed above, I consider this unscientific and unacceptable because the

13 old deformation models were not internally self-consistent, and were not consistent with GPS data, and also because they appeared to be custom-built to minimize seismic hazard at DCPP.

40. In this regard, I advise that NRC should apply strong scrutiny to this planned SB-846-required seismic update (if and when it is released), and also carefully consider the anticipated reviews offered by the 3 outside experts of UCLAs Garrick Risk Institute, and also the anticipated opinions of the Diablo Canyon Independent Safety Committee of the California Public Utilities Commission, informed by their Independent Peer Review Panel.

C. Regarding status of witnesss models in the seismicity/hazard communities

41. Attachment G, page G-27 of Applicants Environmental Report (PG&E, 2023) contains a description of how the Technical Integration (TI) Team and the Participatory Peer Review Panel (PPRP) of the SSHAC Level-3 SSC program (2012-2015) considered a presentation I made at the November 2012 San Luis Obispo workshop, and decided to use some elements (rates of strike-slip) and decided to exclude other elements (rates of horizontal compression; computer algorithms for objective creation of optimal deformation models; global calibrations for converting long-term strain-rates to seismicity). The paragraph I object to is this:

Dr. Bird's modeling of off-fault deformation and alternative methods to calculate seismicity rates were not considered mature enough by the Tl Team at the time of the SSHAC to include in the SSC model. This is consistent with exclusion of these models and model elements from the Uniform California Earthquake Rupture Forecast (ver. 3) which is the basis for the 2014 update to the United States Geological Survey Seismic National Seismic Hazard Map (References 111 & 113)

42. The first problem is a misleading implication of the phrase, exclusion of these models. My deformation model, obtained with my dynamic finite-element code NeoKinema, was used by the USGS in their 2014 Update to the National Seismic Hazard Model (Field et al., 2013). It was assigned a weight of 0.3 in the logic tree, and no other deformation model had a higher weight. The necessary distinction is that USGS finally decided to use only the computed fault slip-rates, and not the self-consistent off-fault deformation field.

43. Second, the repetition of this criticism, not. mature enough, probably written in 2012, in the new Applicants Environmental Report (PG&E, 2023) written 11 years later is also misleading. My NeoKinema code for creation of deformation models was used again in the 2024 Update to the National Seismic Hazard Model (Shen & Bird, 2022), with a logic-tree weight of 0.32. (Again, no other deformation model had a higher weight.)

44. Also, my global-calibration method (Bird & Kagan, 2004; Bird & Liu, 2007) for converting long-term strain-rates to shallow seismicity has been developed into 3 global seismicity models of increasing sophistication (Bird et al., 2010; Bird & Kreemer, 2015; Bird et al.,

2015). These models have been registered with the Collaboratory for the Study of Earthquake Predictability (CSEP) and have proven successful in prospective tests by

14 independent experts (Strader et al., 2018; Bayona et al., 2023). The third of these models, named GEAR1, is currently the global standard.

Under penalty of perjury, I declare that the foregoing statements of fact are true and correct to the best of my knowledge and that the statements of opinion expressed above are based on my best professional judgment.

Executed in Accord with 10 CFR 2.304(d) by Peter Bird Date: March 4, 2024

15 VI.

REFERENCES A. References Cited in Sections I-IV Atwater, T., and J. Stock [1998] Pacific-North America plate tectonics of the Neogene southwestern United States: An update, Int. Geol. Rev., 40, 375-402.

Bird, P., 2003. An updated digital model of plate boundaries, Geochemistry Geophysics Geosystems, 4(3), 1027, doi:10.1029/2001GC000252.

Bird, P. [2009] Long-term fault slip rates, distributed deformation rates, and forecast of seismicity in the western United States from fitting of community geologic, geodetic, and stress direction datasets, J. Geophys. Res., 114(B11403), doi: 10.1029/2009JB006317.

Bird, P. [2023] Declaration of Peter Bird, submitted to U.S. Nuclear Regulatory Commission in support of Comments by San Luis Obispo Mothers for Peace on Proposed Rule and Draft Generic Environmental Impact Statement for Renewing Nuclear Power Plant Licenses, Docket No. 2018-0296, May 2, 2023 (NRC ADAMS Accession No. ML23123A410, https://adamswebsearch2.nrc.gov/webSearch2/main.jsp?AccessionNumber=ML23123A410).

Bird, P., and R. V. Ingersoll [2022] Kinematics and paleogeology of the western United States and northern Mexico computed from geologic and paleomagnetic data: 0 to 48 Ma, Geosphere, 18(5), 1563-1599, https://doi.org/10.1130/GES02474.1.

Bird, P., and Y. Li [1996] Interpolation of principal stress directions by nonparametric statistics:

Global maps with confidence limits, J. Geophys. Res., 101(B3), 5435-5443.

Campbell, K. W., and Y. Bozorgnia, 2014. NGA-West2 ground motion model for the average horizontal components of PGA, PGV, and 5% damped linear acceleration response spectra, Earthquake Spectra, 30(3), 1087-1115.

Cloud, W. K., and D. E. Hudson, 1975. Strong motion data from the San Fernando, California, earthquake of February 9, 1971: San Fernando, California, Earthquake of 9 February 1971, Bulletin 196, California Division of Mines and Geology, pp. 273-303.

Ekstrm, G., M. Nettles, and A. M. Dziewonski [2012] The Global CMT project 2004-2010:

Centroid moment tensors for 13,017 earthquakes, Phys. Earth Planet. Int., 200/201, 19.

Mueller, B., V. Wehrle, and K. Fuchs [1997] The 1997 release of the World Stress Map, http://www-wsm.physik.uni-karlsruhe.de/pub/Rel97/wsm97.html.

Heidbach, O., M. Tingay, A. Barth, J. Reinecker, D. Kurfe, and B. Müller [2008] The World Stress Map database release 2008, doi:10.1594/GFZ.WSM.Rel2008.

Heidbach, O., M. Rajabi, K. Reiter, M.O. Ziegler, and the WSM Team [2016] World Stress Map Database Release 2016, doi:10.5880/WSM.2016.001.

Pacific Gas and Electric Company [2014] Geologic Map of the Irish Hills and Adjacent Area, 1:32,000, DCPP Geologic Mapping Project, Ch9.GEO.DCPP.TR.14.01 R0, https://www.pge.com/includes/docs/pdfs/safety/systemworks/dcpp/report/Ch9.GEO.DCPP.TR.1 4.01_R0_Plates.pdf, NRC ADAMS Accession No. ML14260A068.

16 Pacific Gas and Electric Company (PG&E), 2015. Seismic Source Characterization for the Diablo Canyon Power Plant, San Luis Obispo County, California; report on the results of SSHAC level 3 study, Rev. A, March; 652 pages plus Appendices. Available online at http://www.pge.com/dcpp-ltsp; downloaded 2023.05.11.

Pacific Gas and Electric Company (PG&E), 2015L. Letter DCL-15-035 re: Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding the Seismic Aspects of Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident: Seismic Hazard and Screening Report (Mar. 11, 2015) (NRC Accession No. ML15071A045).

Pacific Gas and Electric Company (PG&E), 2018. Letter DCL-18-027 re: Seismic Probabilistic Risk Assessment for the Diablo Canyon Power Plant, Units 1 and 2 - Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding the Seismic Aspects of Recommendation 2.1: Seismic of the (sic) Near-Term Task force Review of Insights from the Fukushima Dai-Ichi Accident (Apr. 24, 2018) (NRC Accession No. ML18120A201).

Pacific Gas and Electric Company (PG&E), 2023. Appendix E: Applicants Environmental Report, Operating License Renewal Stage, Diablo Canyon Power Plant Units 1 and 2, November 2023: NRC ADAMS Accession No. ML23311A154, dated 2023.11.07, 1,137-page PDF file, accessed 2024.02.

Page, B. M., G. A. Thompson, and R. G. Coleman [1998] Late Cenozoic tectonics of the central and southern Coast Ranges of California, Geol. Soc. Am. Bull., 110(7), 846-876.

Shen, Z.-K., and P. Bird [2022] NeoKinema deformation model for the 2023 update to the U.S.

National Seismic Hazard Model, Seismol. Res. Lett., 93, 3037-33052, doi: 10.1785/0220220179.

Simpson, R. W., R. C. Jachens, R. J. Blakely, and R. W. Saltus [1986], A new isostatic residual gravity map of the conterminous United States with a discussion on the significance of isostatic residual anomalies, J. Geophys. Res., 91(B8), 8348-8372 & 8407-8410.

Toda, S., and Stein, Ross S., 2024. Intense seismic swarm punctuated by a magnitude 7.5 Japan shock, Temblor, http://doi.org/10.32858/temblor.333 United States Geological Survey (USGS), 2024. Finite Fault model for 1 Jan 2024 M 7.5 earthquake, https://earthquake.usgs.gov/earthquakes/eventpage/us6000m0xl/finite-fault Wikipedia, 2024. Noto earthquake, https://en.wikipedia.org/wiki/2024_Noto_earthquake, accessed 2024.02.16.

B. References Cited in Section V Bayona, J. A., W. H. Savran, P. Iturrieta, M. C. Gerstenberger, K. M. Graham, W. Marzocchi, D.

Schorlemmer, and M. J. Werner [2023] Are Regionally Calibrated Seismicity Models More Informative than Global Models? Insights from California, New Zealand, and Italy, The Seismic Record, 3(2), 86-95, doi: 10.1785/0320230006.

17 Bird, P., and Y. Y. Kagan [2004] Plate-tectonic analysis of shallow seismicity: Apparent boundary width, beta, corner magnitude, coupled lithosphere thickness, and coupling in seven tectonic settings, Bull. Seismol. Soc. Am., 94(6), 2380-2399, plus electronic supplement.

Bird, P., and C. Kreemer [2015] Revised tectonic forecast of global shallow seismicity based on version 2.1 of the Global Strain Rate Map, Bull. Seismol. Soc. Am., 105(1), 152-166 plus electronic supplements, doi: 10.1785/0120140129.

Bird, P., and Z. Liu [2007] Seismic hazard inferred from tectonics: California, Seismol. Res.

Lett., 78(1), 37-48.

Bird, P., C. Kreemer, and W. E. Holt [2010] A long-term forecast of shallow seismicity based on the Global Strain Rate Map, Seismol. Res. Lett., 81(2), 184-194, doi:10.1785/gssrl.81.2.184.

Bird, P., D. D. Jackson, Y. Y. Kagan, C. Kreemer, & R. S. Stein [2015] GEAR1: a Global Earthquake Activity Rate model constructed from geodetic strain rates and smoothed seismicity, Bull. Seismol. Soc. Am., 105(5); 2538-2554, doi: 10.1785/0120150058.

Field, E. H., G. P. Biasi, P. Bird, T. E. Dawson, K. R. Felzer, D. D. Jackson, K. M. Johnson, T.

H. Jordan, C. Madden, A. J. Michael, K. R. Milner, M. T. Page, T. Parsons, P. M. Powers, B. E.

Shaw, W. R. Thatcher, R. J. Weldon, II, and Y. Zeng [2013] Unified California Earthquake Rupture Forecast, version 3 (UCERF3)-The time-independent model, U.S. Geol. Surv. Open-File Rep., 2013-1165 (and Cal. Geol. Surv. Spec. Rep. 228, and Southern California Earthquake Center Pub. 1792), 97 pages (main report) + 20 Appendices; http://pubs.usgs.gov/of/2013/1165/

Nuclear Regulatory Commission (NRC) [2023] Draft Generic Environmental Impact Statement for License Renewal of Nuclear Plants: NUREG-1437, Rev. 2, Feb. 2023.

Pacific Gas and Electric Company (PG&E) [2023] Appendix E: Applicants Environmental Report, Operating License Renewal Stage, Diablo Canyon Power Plant Units 1 and 2, November 2023: NRC ADAMS Accession No. ML23311A154, dated 2023.11.07, 1,137-page PDF file, accessed 2024.02.

Shen, Z.-K., and P. Bird [2022] NeoKinema deformation model for the 2023 update to the U.S.

National Seismic Hazard Model, Seismol. Res. Lett., 93(6), 3037-3052, https://doi.org/10.1785/0220220179, 16 pages.

Strader, A., M. Werner, J. Bayona, and D. Schorlemmer [2018] Prospective evaluation of global earthquake forecast models: Two years of observations support merging smoothed seismicity with geodetic strain rates, Seismol. Res. Lett., 89(4), 1262-1271, doi: 10.1785/0220180051.

18 VII.

CURRICULUM VITAE CURRICULUM VITAE OF PETER BIRD Department of Earth, Planetary, and Space Sciences Mail Code 156704 University of California Los Angeles, CA 90095-1567 e-mail: pbird@epss.ucla.edu website: http://peterbird.name EDUCATION Massachusetts Institute of Technology: Ph.D. in Earth and Planetary Sciences, 1976 Harvard College: B.A. in Geological Sciences, 1972 EMPLOYMENT University of California, Los Angeles:

Professor Emeritus, 2011-Professor of Geophysics and Geology, 1985-2011 Vice-chairman, Dept. of Earth and Space Sciences, 1994-2002 Associate Professor of Geophysics and Geology, 1981-85 Assistant Professor of Geophysics and Geology, 1976-81 HONORS Woollard Award, Geological Society of America, 2013 Fellow, American Geophysical Union, 1990 Fellow, Geological Society of America, 1989 RESEARCH AREAS (CHRONOLOGICAL FROM 1973)

Lateral refraction and attenuation of surface waves 1973-1977 Marine paleomagnetism and seafloor spreading 1974-1975 Thermal modeling with finite differences 1975-1977 Dynamic modeling with finite elements 1975-Tectonophysics of continental collisions 1975-Formation of marginal basins 1976-1977 Stress and temperature in subduction zones 1976-2009 Continental delamination 1977-1982 Neotectonic models of California 1978-Hydration state and friction of montmorillonite clays 1979-1984 Mechanism of Laramide orogeny 1982-Mechanism of Basin/Range taphrogeny 1986-Solution transfer experiments on quartz 1986-1993 Lateral extrusion of lower crust 1987-1991 Regional neotectonic models: Africa, Alaska, Asia, Europe,...

1989-Global dynamic lithosphere models with plates & driving forces 1992-Inverse or kinematic tectonic models from geologic & paleomag data 1994-Global long-term seismicity forecasts from geodesy & plate tectonics 2000-Long-term seismicity forecasts for Europe, especially Italy 2009-

19 CONSULTING EXPERIENCE ON SEISMIC HAZARD (FROM 2009 TO PRESENT)

GeoPentech, Lettis Consultants International, FM Global, Temblor, San Luis Obispo Mothers for Peace UNPAID AFFILIATIONS Southern California Earthquake Center (2000-present; Board member 2004-2012)

Collaboratory for the Study of Earthquake Predictability (model contributor, 2015)

20 VIII. PUBLICATIONS (CHRONOLOGICAL FROM 1975; OMITTING MOST ABSTRACTS)

Bird, P., and J. D. Phillips [1975] Oblique spreading near the Oceanographer Fracture, J.

Geophys. Res., 80, 4021-4027.

Bird, P., M. N. Toksoz, and N. H. Sleep [1975] Thermal and mechanical models of continent-continent convergence zones, J. Geophys. Res., 80, 4405-4416.

Toksoz, M. N., and P. Bird [1977] Modeling of temperatures in continental convergence zones, Tectonophysics, 41, 181-193.

Bird, P., and M. N. Toksoz [1977] Strong attenuation of Rayleigh waves in Tibet, Nature, 266, 161-163.

Toksoz, M. N., and P. Bird [1977] Formation and evolution of marginal basins and continental plateaus, in: M. Talwani and W. C. Pitman, III (Ed.), Island Arcs, Deep Sea Trenches, and Back Arc Basins, Maurice Ewing Series 1, Am. Geophys. Union, Washington, 379-394.

Bird, P. [1978a] Initiation of intracontinental subduction in the Himalaya, J. Geophys. Res.,

83, 4975-4987.

Bird, P. [1978b] Finite-element modeling of lithosphere deformation: The Zagros collision orogeny, Tectonophysics, 50, 307-336.

Bird, P. [1978c] Stress and temperature in subduction shear zones: Tonga and Mariana, Geophys. J. R. Astron. Soc., 55, 411-434.

Bird, P. [1979] Continental delamination and the Colorado Plateau, J. Geophys. Res., 84, 7561-7571.

Bird, P., and D. A. Yuen [1979] The use of the minimum-dissipation principle in tectonophysics, Earth Planet. Sci. Lett., 45, 214-217.

Bird, P., and K. Piper [1980] Plane-stress finite-element models of tectonic flow in southern California, Phys. Earth Planet. Int., 21, 158-175.

Bird, P., and J. Baumgardner [1981] Steady propagation of delamination events, J. Geophys.

Res., 86, 4891-4903.

Bird, P. [1982] Reply re: Initiation of intracontinental subduction in the Himalaya, J.

Geophys. Res., 86, 9323-9324.

Bird, P., and J. Baumgardner [1983] 3-D Finite element modeling of the Earth's free oscillations (abstract), Eos, 64, 754.

21 Bird, P. [1984] Hydration-phase diagrams and friction of montmorillonite under laboratory and geologic conditions, with implications for shale compaction, slope stability, and strength of fault gouge, Tectonophysics, 107, 235-260.

Bird, P., and J. Baumgardner [1984] Fault friction, regional stress, and crust-mantle coupling in southern California from finite element models, J. Geophys. Res., 89, 1932-1944.

Bird, P., and R. Rosenstock [1984] Kinematics of present crust and mantle flow in southern California, Geol. Soc. Am. Bull., 95, 946-957.

Bird, P. [1985] Laramide crustal thickening event in the Rocky Mountain foreland and Great Plains, Tectonics, 3, 741-758.

Bird, P. [1986] Tectonics of the terrestrial planets, in: M. G. Kivelson (Ed.), The Solar System: Observations and Interpretations, Rubey Volume 4, Prentice Hall, Englewood Cliffs, New Jersey, 176-206.

Bird, P. [1988] Formation of the Rocky Mountains, western United States: a continuum computer model, Science, 239, 1501-1507.

Bird, P. [1989] New finite element techniques for modeling deformation histories of continents with stratified temperature-dependent rheologies, J. Geophys. Res., 94, 3967-3990.

Bird, P., and A. J. Gratz [1990] A theory for buckling of the mantle lithosphere and Moho during compressive detachments in continents, Tectonophysics, 177, 325-336.

Bird, P., and D. R. Williams [1990] Lack of lateral extrusion on Venus limits thickness of the crust (abstract), Eos, 71, 1423.

Gratz, A. J., P. Bird, and G. B. Quiro [1990] Dissolution of quartz in aqueous basic solution, 106-236 C: Surface kinetics of "perfect" crystallographic faces, Geochimica et Cosmochimica Acta, 54, 2911-2922.

Bird, P. [1991] Lateral extrusion of lower crust from under high topography, in the isostatic limit, J. Geophys. Res., 96, 10,275-10,286.

Bird, P. [1992] Deformation and uplift of North America in the Cenozoic era, in: K. R.

Billingsley, H. U. Brown, III, and E. Derohanes (eds.), Scientific Excellence in Supercomputing: the IBM 1990 Contest Prize Papers, Baldwin Press, Athens, Georgia,

v. 1, pp.67-105.

Kemp, D. V., and P. Bird [1992] Bending and dynamic support of subducted slabs (abstract), Eos Trans. AGU, 73(43), Fall Meeting Suppl., 386.

Gratz, A. J., and P. Bird [1993a] Quartz dissolution: Negative crystal experiments and a rate law, Geochimica et Cosmochimica Acta, 57, 965-976.

22 Gratz, A. J., and P. Bird [1993b] Quartz dissolution: Theory of rough and smooth surfaces, Geochimica et Cosmochimica Acta, 57, 977-989.

Bird, P. and X. Kong [1994] Computer simulations of California tectonics confirm very low strength of major faults, Geol. Soc. Am. Bull., 106(2), 159-174.

Bird, P. [1994] Isotopic evidence for preservation of Cordilleran lithospheric mantle during the Sevier-Laramide orogeny, western United States: Comment, Geology, 22 (7), 670-671.

Bird, P. [1995] Lithosphere dynamics and continental deformation, Rev. Geophys.,

Supplement: U.S. National Report to IUGG 1991-94, 379-383.

Kong, X., and P. Bird [1995] SHELLS: A thin-shell program for modeling neotectonics of regional or global lithosphere with faults, J. Geophys. Res., 100, 22,129-22,131.

Bird, P., and Yao Li [1996] Interpolation of principal stress directions by nonparametric statistics: Global maps with confidence limits, J. Geophys. Res., 101, 5435-5443.

Bird, P. [1996] Computer simulations of Alaskan neotectonics, Tectonics, 15, 225-236.

Kong, X., and P. Bird [1996] Neotectonics of Asia: Thin-shell finite-element models with faults, in: An Yin and T. M. Harrison (ed.s), The Tectonic Evolution of Asia, Cambridge University Press, p. 18-34.

Bird, P. [1998a] Testing hypotheses on plate-driving mechanisms with global lithosphere models including topography, thermal structure, and faults, J. Geophys. Res., 103, B5, 10,115-1,129.

Bird, P. [1998b] Kinematic history of the Laramide orogeny in latitudes 35-49 N, western United States, Tectonics, 17, 780-801.

Bird, P. [1999] Thin-plate and thin-shell finite element programs for forward dynamic modeling of plate deformation and faulting, Computers & Geosciences, 25, 383-394.

Bird, P. and Z. Liu [1999] Global finite-element model makes a small contribution to intraplate seismic hazard estimation, Bull. Seismol. Soc. Am., 89(6), 1642-1647.

Jimenez-Munt, I., P. Bird, and M. Fernandez [2001] Thin-shell modeling of neotectonics in the Azores-Gibraltar region, Geophys. Res. Lett., 28(6), 1083-1086.

Jiménez-Munt, I., M. Fernndez, M. Torne, and P. Bird [2001] The transition from linear to diffuse plate boundary in the Azores-Gibraltar region: Results from a thin sheet model, Earth Planet. Sci. Lett., 192, 175-189.

Bird, P. [2002] Stress-direction history of the western United States and Mexico since 85 Ma, Tectonics, 21, doi: 10.1029/2001TC001319.

23 Bird, P., Y. Y. Kagan, and D. D. Jackson [2002] Plate tectonics and earthquake potential of spreading ridges and oceanic transform faults, in: S. Stein and J. T. Freymueller (editors), Plate Boundary Zones, Geodynamics Series, 130, 203-218.

Negredo, A. M., P. Bird, C. Sanz de Galdeano, and E. Buforn [2002] Neotectonic modeling of the Ibero-Maghrebian region, J. Geophys. Res., 107(B11), 2292, doi:

10.1029/2001JB000743.

Liu, Z., and P. Bird [2002a] Finite element modeling of neotectonics in New Zealand, J.

Geophys. Res., 107(B12), 2328, doi: 10.1029/2001JB001075.

Liu, Z., and P. Bird [2002b] North America plate is driven westward by lower mantle flow, Geophys. Res. Lett., 29(24), 2164, doi: 10.1029/2002GL016002.

Bird, P. [2003] An updated digital model of plate boundaries, Geochemistry Geophysics Geosystems, 4(3), 1027, doi: 10.1029/2001GC000252.

Bird, P., and Y. Y. Kagan [2004] Plate-tectonic analysis of shallow seismicity: Apparent boundary width, beta, corner magnitude, coupled lithosphere thickness, and coupling in seven tectonic settings, Bull. Seismol. Soc. Am., 94(6), 2380-2399.

Liu, Z., and P. Bird [2006] Two-dimensional and three-dimensional finite element modelling of mantle processes beneath central South Island, New Zealand, Geophys. J.

Int., 165, 1003-1028.

Bird, P., Z. Ben-Avraham, G. Schubert, M. Andreoli, and G. Viola [2006] Patterns of stress and strain rate in southern Africa, J. Geophys. Res., 111(B8), B08402, doi:

10.1029/2005JB003882.

Bird, P., and Z. Liu [2007] Seismic hazard inferred from tectonics: California, Seismol. Res.

Lett., 78(1), 37-48.

Bird, P. [2007] Uncertainties in long-term geologic offset rates of faults: General principles illustrated with data from California and other western states, Geosphere, 3(6), 577-595; doi: 10.1130/GES00127.1, + 9 digital file appendices.

Liu, Z., and P. Bird [2008] Kinematic modelling of neotectonics in the Persia-Tibet-Burma orogen, Geophys. J. Int., 172(2), 779-797, doi: 10.1111/j.1365-246X.2007.03640.x.

Bird, P., Z. Liu, and W. K. Rucker [2008] Stresses that drive the plates from below:

Definitions, computational path, model optimization, and error estimates, J. Geophys.

Res., 113, B11406, doi: 10.1029/2007JB005460, plus digital appendices.

Bird, P. [2009] Long-term fault slip rates, distributed deformation rates, and forecast of seismicity in the western United States from joint fitting of community geologic, geodetic, and stress direction data sets, J. Geophys. Res., 114, B11403, doi: 10.1029/

2009JB006317.

24 Bird, P., Y. Y. Kagan, D. D. Jackson, F. P. Schoenberg, and M. J. Werner [2009] Linear and nonlinear relations between relative plate velocity and seismicity, Bull. Seismol. Soc.

Am., 99(6), 3097-3113, doi: 10.1785/ 0120090082.

Kagan, Y. Y., P. Bird, and D. D. Jackson [2010] Earthquake patterns in diverse tectonic zones of the globe, Pure Appl. Geophys., 167(6/7; Frank Evison volume), doi:

10.1007/s00024-0075-3.

Bird, P., C. Kreemer, and W. E. Holt [2010] A long-term forecast of shallow seismicity based on the Global Strain Rate Map, Seismol. Res. Lett., 81(2), 184-194, doi:

10.1785/gssrl.81.2.184.

Howe, T. M., and P. Bird [2010] Exploratory models of long-term crustal flow and resulting seismicity across the Alpine-Aegean orogen, Tectonics, 29, TC4023, doi:

10.1029/2009TC002565.

Austermann, J., Z. Ben-Avraham, P. Bird, O, Heidbach, G. Schubert, and J. M. Stock [2011]

Quantifying the forces needed for the rapid change of Pacific plate motion at 6 Ma, Earth Planet. Sci. Lett., 307, 289-297, doi: 10.1016/j.epsl.2011.04.043.

Chu, A., F. P. Schoenberg, P. Bird, D. D. Jackson, and Y. Y. Kagan [2011] Comparison of ETAS parameter estimates across different global tectonic zones, Bull. Seismol. Soc.

Am., 101(5), 2323-2339, doi: 10.1785/0120100115.

Field, E. H., G. P. Biasi, P. Bird, T. E. Dawson, K. R. Felzer, D. D. Jackson, K. M. Johnson, T. H. Jordan, C. Madden, A. J. Michael, K. R. Milner, M. T. Page, T. Parsons, P. M.

Powers, B. E. Shaw, W. R. Thatcher, R. J. Weldon, II, and Y. Zeng [2013] Unified California Earthquake Rupture Forecast, version 3 (UCERF3)-The time-independent model, U.S. Geol. Surv. Open-File Rep., 2013-1165 (Cal. Geol. Surv. Spec. Rep. 228, and Southern California Earthquake Center Pub. 1792), 97 pages (main report) + 20 Appendices; http://pubs.usgs.gov/of/2013/1165/.

Petersen, M. D., Y. Zeng, K. M. Haller, R. McCaffrey, W. C. Hammond, P. Bird, M.

Moschetti, Z. Shen, J. Bormann, and W. Thatcher [2014] Geodesy-and geology-based slip-rate models for the Western United States (excluding California) national seismic hazard maps, U.S. Geol. Surv. Open-File Rep., 2013-1293, 38 pages (main report) + 5 Appendices; http://dx.doi.org/10.3133/ofr20131293.

Curren, I. S., and P. Bird[2014] Formation and suppression of strike-slip fault systems, Pure Appl. Geophys., 171(11), 2899-2918, doi: 10.1007/s00024-014-0826-7.

Bird, P., and C. Kreemer [2015a] Revised tectonic forecast of global shallow seismicity based on version 2.1 of the Global Strain Rate Map, Bull. Seismol. Soc. Am., 105(1),

152-166, doi: 10.1785/0120140129.

Bird, P., D. D. Jackson, Y. Y. Kagan, C. Kreemer, and R. S. Stein [2015] GEAR1: A Global Earthquake Activity Rate model constructed from geodetic strain rates and smoothed seismicity, Bull. Seismol. Soc. Am., 105(5), 2538-2554, doi: 10.1785/0120150058.

25 Carafa, M., S. Barba, and P. Bird [2015] Neotectonics and long-term seismicity in Europe and the Mediterranean region, J. Geophys. Res., 120(7), 5311-5342, doi:

10.1002/2014JB011751.

Rong, Y., P. Bird, and D. D. Jackson [2016] Earthquake potential and magnitude limits inferred from a geodetic strain-rate model of southern Europe, Geophys. J. Int., 205(1),

509-522, doi: 10.1093/gji/ggw018.

Bird, P., and M. Carafa [2016] Improving deformation models by discounting transient signals in geodetic data, 1: Concept and synthetic examples, J. Geophys. Res., 121(7),

5538-5556, doi: 10.1002/2016JB013056.

Carafa, M. M. C., and P. Bird [2016] Improving deformation models by discounting transient signals in geodetic data, 2: Geodetic data, stress directions, and long-term strain rates in Italy, J. Geophys. Res., 121(7), 5557-5575, doi: 10.1002/2016JB013038.

Carafa, M. M. C., G. Valensise, and P. Bird [2017] Assessing the seismic coupling of shallow continental faults and its impact on seismic hazard estimates: a case-study from Italy, Geophys. J. Int., 209, 32-47, doi: 10.1093/gji/ggx002.

Bird, P. [2017] Stress field models from Maxwell stress functions: southern California, Geophys. J. Int., 210(2), 951-963, doi: 10.1093/gji/ggx207.

Tunini, L., I Jimenez-Munt, M. Fernandez, J. Verges, and P. Bird [2017] Neotectonic deformation in central Eurasia: A geodynamic model approach, J. Geophys. Res.,

122(11), 9461-9484, doi: 10.1002/2017JB014487.

Carafa, M. M. C., V. Kastelic, P. Bird, F. Maesano, and G. Valensise [2018] A geodetic gap in the Calabrian Arc: Evidence for a locked subduction megathrust?, Geophys.

Res. Lett., 45, 1794-1804, doi: 10.1002/2017GL076554.

Bird, P. [2018] Ranking some global forecasts with the Kagan information score, Seismol.

Res. Lett., 89(4), 1272-1276, doi: 10.1785/0220180029.

Carafa, M. M. C., A. Galvani, D. Di Naccio, V. Kastelic, C. Di Lorenzo, S. Miccolis, V.

Sespe, G. Pietrantonio, C. Gizzi, A. Massucci, G. Valensise, and P. Bird [2020]

Partitioning the ongoing extension of the Central Apennines (Italy): Fault slip rates and bulk deformation rates from geodetic and stress data, J. Geophys. Res., 125, e2019JB018956, https://doi.org/10.1029/2019JB018956.

Bird, P., and R. V. Ingersoll [2022] Kinematics and paleogeology of the western United States and northern Mexico computed from geologic and paleomagnetic data: 0 to 48 Ma, Geosphere, 18(5), 1563-1599, doi: org/10.1130/GES02474.1.

Shen, Z.-K., and P. Bird [2022] NeoKinema deformation model for the 2023 update to the U.S. National Seismic Hazard Model, Seismol. Res. Lett., 93(6), 3037-3052, https://doi.org/10.1785/0220220179.

26 Carafa, M. M. C., D. Di Naccio, C. Di Lorenzo, V. Kastelic, and P. Bird [2022] A meta-analysis of fault slip rates across the Central Apennines, J. Geophys. Res., 127, e2021JB023252; https://doi.org/10.1029/2021JB023252.

From:

Diane Curran To:

Docket, Hearing; Bessette, Paul M.; ryan.lighty@morganlewis.com; Timothy J. Matthews (timothy.matthews@morganlewis.com); Jeremy Wachutka; Catherine Kanatas; Adam Gendelman Cc:

Templeton, Hallie; Caroline Leary

Subject:

[External_Sender] RE: Filing of hearing request in Diablo Canyon license renewal proceeding, Docket Nos. 50-275-LR and 50.323-LR Date:

Monday, March 04, 2024 9:55:15 PM Attachments:

2024.03.04 Exhibit 3 Macdonald Declaration_opt2.pdf Attached please find Exhibit 3 to Petitioners hearing request.

From: Diane Curran <dcurran@harmoncurran.com>

Sent: Monday, March 4, 2024 9:53 PM To: Docket, Hearing <Hearing.Docket@nrc.gov>; Bessette, Paul M.

<paul.bessette@morganlewis.com>; ryan.lighty@morganlewis.com; Timothy J. Matthews (timothy.matthews@morganlewis.com) <timothy.matthews@morganlewis.com>; Jeremy Wachutka (Jeremy.Wachutka@nrc.gov) <Jeremy.Wachutka@nrc.gov>; Catherine E. Kanatas (catherine.kanatas@nrc.gov) <catherine.kanatas@nrc.gov>; Gendelman, Adam

<Adam.Gendelman@nrc.gov>

Cc: Templeton, Hallie <HTempleton@foe.org>; Caroline Leary <cleary@ewg.org>

Subject:

RE: Filing of hearing request in Diablo Canyon license renewal proceeding, Docket Nos. 50-275-LR and 50.323-LR

Attached please find Exhibits 2 and 4 to Petitioners hearing request.

From: Diane Curran <dcurran@harmoncurran.com>

Sent: Monday, March 4, 2024 9:50 PM To: Docket, Hearing <Hearing.Docket@nrc.gov>; Bessette, Paul M.

<paul.bessette@morganlewis.com>; ryan.lighty@morganlewis.com; Timothy J. Matthews (timothy.matthews@morganlewis.com) <timothy.matthews@morganlewis.com>; Jeremy Wachutka (Jeremy.Wachutka@nrc.gov) <Jeremy.Wachutka@nrc.gov>; Catherine E. Kanatas (catherine.kanatas@nrc.gov) <catherine.kanatas@nrc.gov>; Gendelman, Adam

<Adam.Gendelman@nrc.gov>

Cc: Templeton, Hallie <HTempleton@foe.org>; Caroline Leary <cleary@ewg.org>

Subject:

Filing of hearing request in Diablo Canyon license renewal proceeding, Docket Nos. 50-275-LR and 50.323-LR

Dear NRC Secretary and parties,

On behalf of San Luis Obispo Mothers for Peace, Friends of the Earth, and Environmental Working Group, I am filing a hearing request regarding PG&Es application for renewal of the Diablo Canyon operating licenses.

I am sending the hearing request by email because I was not able to request access to the electronic docket in time for this filing. I will post the hearing request to the electronic docket as soon as I get access.

In the meantime, I am sending the hearing request in three separate emails out of concern that the exhibits may be too large to send in one message:

This email has the hearing request itself with Exhibit 1 attached.

The second email will have Exhibits 2 and 4.

The third email will have Exhibit 3.

Sincerely, Diane Curran Counsel for SLOMFP

Diane Curran Harmon, Curran, Spielberg & Eisenberg LLP Contact (240) 393-9285 1725 DeSales Street, NW, Suite 500, Washington, DC 20036 Email DCurran@harmoncurran.com

EXHIBIT 3 - DECLARATION OF DIGBY MACDONALD

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE SECRETARY In the matter of Pacific Gas and Electric Company Docket Nos. 50-275-LR Diablo Canyon Nuclear Power Plant 50-323-LR Units 1 and 2 DECLARATION OF DIGBY MACDONALD, Ph.D Under penalty of perjury, I, Digby Macdonald, declare as follows:

I.

PURPOSE AND DESCRIPTION OF MY DECLARATION

1. I have been retained by San Luis Obispo Mothers for Peace (SLOMFP), Environmental Working Group (EWG) and Friends of the Earth (FOE) to evaluate Pacific Gas and Electric Companys (PG&Es) license renewal application of November 7, 2023 with respect to its program for surveillance of the Diablo Canyon Unit 1 reactor pressure vessel. The purpose of my declaration is to explain why, in my expert opinion, PG&Es license renewal application does not contain an adequate plan to monitor and manage the effects of aging due to embrittlement of the Unit 1 reactor pressure vessel (RPV).1

2. As discussed below, PG&Es aging management program for the Unit 1 RPV is based upon and continues the surveillance program that PG&E has used during the initial operating license period. In the summer and fall of 2023, I evaluated PG&Es program and found that it was deficient in multiple significant respects, to the point that operation of Unit 1 poses an unreasonable risk to public health and safety due to serious indications of an unacceptable degree of embrittlement, coupled with a lack of information to establish otherwise. In a declaration submitted to the NRC on September 14, 2023, I recommended to the U.S. Nuclear Regulatory Commission (NRC) that Unit 1 should be closed until PG&E obtains and analyzes additional information regarding its condition.2

3. I continue to hold the opinions stated in my 9/14/23 Declaration, which I hereby adopt and incorporate by reference into this Declaration. As discussed below, I am concerned that the significant defects in PG&Es current RPV surveillance program are perpetuated in the LRA without being addressed or corrected. Therefore, the LRA fails to 1 This declaration pertains only to the Unit 1 RPV. I have not studied the Unit 2 RPV in any detail and therefore have not formed an opinion about PG&Es program for managing the aging of that reactor.

2 Declaration of Digby Macdonald, Ph.D in Support of Hearing Request and Request for Emergency Order by San Luis Obispo Mothers for Peace and Friends of the Earth (Sept. 14, 2023) (9/14/23 Declaration) (NRC Accession No. ML23257A302). A copy of my 9/14/23 Declaration is attached to this Declaration as Attachment 1.

2 demonstrate that the effects of aging on the Unit 1 RPV will be managed in a way that is adequate to protect public health and safety. In my expert opinion, Unit 1 should be closed pending withdrawal of Capsule B and other testing to provide reasonable assurance that the Unit 1 RPV is safe to operate now and when renewed operation commences.

4.Section II below provides a statement of my expert qualifications.Section III provides background information on the role of a pressure vessel in a nuclear reactor and NRC guidance and regulations.Section IV provides background information regarding PG&Es existing RPV surveillance program and its license renewal application.Section V summarizes the technical analysis of my 9/14/23 declaration as it applies to PG&Es LRA application. In Section VI, I will address additional concerns that have arisen since then.

II.

EXPERT QUALIFICATIONS

5. I am Professor in Residence at the University of California at Berkeley (UC Berkeley), in the Departments of Nuclear Engineering and Materials Science and Engineering, one of the worlds preeminent nuclear engineering programs. I hold a Ph.D. in Chemistry from the University of Calgary in Canada and B.Sc. and M.Sc. degrees also in Chemistry from the University of Auckland in New Zealand. I am a qualified expert in the field of materials science with an emphasis on materials in nuclear power reactors (fission and fusion). My areas of expertise include electrochemistry, thermodynamics, applied fracture mechanics, and corrosion science, with emphasis on the growth and breakdown of passive films, chemistry of high temperature aqueous solutions, electro-catalysis, advanced batteries and fuel cells, stress corrosion cracking and corrosion fatigue, materials for nuclear power reactors, and the deterministic prediction of corrosion damage.

6. My expert qualifications are described in detail in Section II of my 9/14/23 Declaration.

And a copy of my curriculum vitae is attached as Appendix A to my 9/14/23 Declaration.

III.

BACKGROUND ON ROLE OF PRESSURE VESSEL AND NRC REGULATORY REQUIREMENTS AND GUIDANCE

7. As discussed in Section IV of my 9/14/23 Declaration, the pressure vessel is a uniquely important and vulnerable component in a nuclear reactor, because it holds water on the highly radioactive reactor core, and because it has no backup if it should crack and lose water during an accident. Therefore, compliance with NRC requirements for monitoring the condition of the plant-specific pressure vessel is essential.

8. For pressure vessels, these regulatory requirements are three-fold and complementary:

First, through Charpy testing of samples taken from the reactor vessel, the licensee must demonstrate that the change in the reference temperature for pressurized thermal shock (RTPTS) is below a threshold of 270oF for axially oriented welds and

3 300oF for circumferential welds. RTPTS is the temperature at which fracture morphology of the pressure vessel changes from ductile to brittle as its temperature drops from the addition of cooling water during a loss of coolant accident (LOCA).

Data for the fracture energy vs. test temperature are determined from Charpy testing of standard specimens (ASTM 185-82) that had been irradiated in capsules located between the reactor core and the inner surface of the RPV and are compared with the unirradiated material and the change from that state is deermined. The capsules are withdrawn at more-or-less equally spaced intervals (typically, every ten calendar years) throughout the reactor life of 32 EFPY (40 calendar years).

Second, also through Charpy testing, the licensee must demonstrate that the pressure vessel is strong enough to withstand the transient stresses induced by thermal shock of the rapidly changing temperature caused by the addition of cooling water, i.e., that the upper shelf energy (USE) will remain above 50 ft-lb.

Finally, every ten years, the licensee must conduct ultrasound testing (UT) inspections of the most vulnerable part of the reactor vessel, the welds around the beltline, to examine for flaws and cracks. NRC guidance appropriately provides that the schedules for these inspections may be relaxed only upon a verifiable demonstration that safety will not be jeopardized.

9. These three types of tests and inspections are complementary in three significant respects.

First, each of the measured phenomena makes a distinct and significant contribution to determining the vulnerability of a pressure vessel to cracking. Second, while the reference temperature and USE calculations are both derived from the same Charpy tests, the method of analysis for each is different; and of course, the UT inspections involve completely different methods of acquiring and analyzing data. Third, each type of test or inspection has a different level of reliability. As discussed in Section V.A.2 of my 9/14/23 Declaration, my calculations show that Charpy tests are not particularly sensitive to the extent of embrittlement. Therefore, their results should not be substituted for UT inspections, nor should they be used to justify an extension of the schedule for UT inspections. The three types of data must be considered in unison because they convey important, complementary information on the safety of the RPV.

10. Adequate monitoring of the condition of the pressure vessel is particularly important in the case of Diablo Canyon Unit 1 because the composition of the welds in the pressure vessel was found to be defective at the time it was installed by having excessive copper and nickel. Not surprisingly, in 2006, the NRC identified the Unit 1 pressure vessel among the most embrittled, with only 14 of 72 PTS reference temperatures as high as or higher than Diablo Canyon Unit 1. U.S. NRC 2007. And today, half of those 14 reactors are closed.

11. To obtain NRC approval of a renewed operating license, NRC regulation 10 C.F.R. 54.21 requires that licensees must demonstrate that that the effects of aging due to embrittlement of the RPV and associated internals will be adequately monitored and managed during the entire license renewal term. Required measures for aging

4 management include time-limited aging analyses (TLAAs). As defined in the regulations, TLAAs are calculations and analyses that (1) Involve systems, structures, and components within the scope of license renewal, as delineated in § 54.4(a); (2) Consider the effects of aging; (3) Involve time-limited assumptions defined by the current operating term, for example, 40 years; (4) Were determined to be relevant by the licensee in making a safety determination; (5) Involve conclusions or provide the basis for conclusions related to the capability of the system, structure, and component to perform its intended functions, as delineated in § 54.4(b); and (6) Are contained or incorporated by reference in the CLB. Thus, TLAAs depend significantly on the calculations and analyses developed in the initial license renewal term.

IV.

BACKGROUND REGARDING PG&ES LICENSE RENEWAL APPLICATION

12. In Section 4.2, PG&Es license renewal application provides a reactor vessel neutron embrittlement analysis. [p. 4.2-1] Here and elsewhere in the LRA, PG&E makes several statements that relate its current RPV monitoring program to the monitoring program in the license renewal period. For instance, at page 4.2-1 the LRA states that current calculations regarding RTNDT and USE values are updated for the license renewal application:

For DCPP Units 1 and 2, the reactor vessel material RTNDT and USE values, calculated on the basis of 32 effective full power years (EFPY) neutron fluence, are determined as part of the CLB and support safety determinations and TS operating limits. Therefore, these calculations are TLAAs [time-limited aging analyses]. For LR, these must be updated to account for the fluence expected to occur during 60 years of plant operation (54 EFPY).

13. Similarly, on page 4.2-2 the LRA states:

The current license period reactor vessel embrittlement analyses that evaluate reduction of fracture toughness of the DCPP Units 1 and 2 reactor vessel beltline materials are based on predicted 40-year EOLE fluence values. The fluence analysis and the neutron embrittlement analyses that are based upon the fluence analysis are TLAAs as defined by 10 CFR 54.21(c) that must be evaluated for the increased neutron fluence associated with 60 years of operation. These TLAAs include the analyses for neutron fluence, PTS USE, adjusted reference temperature (ART), and pressure-temperature limits including low temperature over pressure protection analysis.

14. On pages 4.2 4.2-3 the LRA states:

The last capsule withdrawn and tested from Unit 1 was Capsule V at the EOC 11 in 2002. At that point, Unit 1 Capsule V had an exposure equivalent to 32.25 EFPY of operation. The results were documented in WCAP-15958 (Reference 4.9.1).

5 This exposure is less than that expected at EOLE. Therefore, to obtain capsule data for a neutron fluence of between one and two times the peak reactor vessel wall neutron fluence at EOLE, in PG&E Letter DCL-23-038, dated May 15, 2023 (Reference 4.9.2), PG&E requested NRC approval of the Unit 1 Capsule B withdrawal schedule. NRC approved the requested withdrawal schedule by letter dated July 20, 2023 (Reference 4.9.3).

In the context of this discussion, the LRA also states that PG&E will submit the results of the tested surveillance capsule to NRC in accordance with 10 CFR 50, Appendix G and H. Id. Thus, the previous schedule for withdrawal of Capsule B is relied on by PG&E for its plans. However, I am unable to locate any commitment by PG&E to a deadline for removing and testing Capsule B.

15. On page 4.3-3, the LRA also references the current monitoring program in stating:

For LR, updated fluence projections based upon 54 EFPY were prepared for use as inputs in the neutron embrittlement analyses for 60 years of operation.

The reactor vessel beltline neutron fluence values for 60 years of operation were calculated for DCPP Units 1 and 2 reactor vessel beltline material in WCAP-17299-NP (Reference 4.9.5). These fluence data tabulations include fuel cycle specific power distributions through the end of Cycle 16 for Units 1 and 2, as well as fluence projections at several intervals out to 54 EFPY. The analysis methods used to calculate the predicted 60-year DCPP Units 1 and 2 vessel fluence values satisfy the requirements set forth in RG 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," Revision 0. These methodologies have been approved by the NRC and are described in WCAP-14040-NP-A, Revision 4.

In addition, in accordance with 10 CFR 50, Appendix H, any materials exposed to a neutron fluence exceeding 1.0 x 1017 n/cm2 (E > 1.0 MeV) must be monitored to evaluate changes in fracture toughness. Reactor vessel materials that are not traditionally thought of as being plant limiting were also evaluated in WCAP-17299-NP (Reference 4.9.5) to determine their cumulative fluence values at 54 EFPY. Fluence calculations were performed for the DCPP Units 1 and 2 reactor vessels to determine if the fluence at specific locations will exceed 1.0 x 1017 n/cm2 (E > 1.0 MeV) at 54 EFPY. The materials exposed to fluences that exceed this threshold are referred to as the extended beltline materials.

16. On page B.2-33, the LRA program for monitoring the loss of fracture toughness to the existing program, which provides a baseline:

The AMP will not directly monitor for loss of fracture toughness that is induced by thermal aging or neutron irradiation embrittlement. Instead, the impact of loss of fracture toughness on component integrity will be indirectly managed by: (1)

6 using visual or volumetric examination techniques to monitor for cracking in the components, and (2) in cases where cracking is detected in the components and is extensive enough to necessitate a supplemental flaw growth or flaw tolerance evaluation, applying applicable reduced fracture toughness properties, including reductions for thermal aging or neutron embrittlement. The AMP will use physical measurements to monitor for any dimension changes due to void swelling or distortion.

However, I am unable to locate any mention in the LRA of how PG&Es general commitment to conduct ultrasonic testing of beltline welds relates to the schedule for UT inspections that is discussed below in par. 19(e).

17. At page B.2-95, the LRA states:

The DCPP Reactor Vessel Surveillance AMP provides guidance for removal and testing or storage of material specimen capsules. All capsules that have been withdrawn, tested, and not otherwise donated, were stored in conformance with 10 CFR 50 requirements. For Unit 1, the last capsule is expected to be withdrawn and tested having accumulated 1-2 times the peak reactor vessel neutron fluence at 60 years of operation (NRC approved Unit 1 Capsule B withdrawal at approximately 96.19 - 101.01 EFPY). The remaining four standby capsules in Unit 1 have low lead factors, will remain inside the vessel throughout the vessel lifetime, and will be available for future testing. There are no capsules remaining in the Unit 2 vessel.

18. In Enclosure 3, WCAP-17315-NP, the LRA also asserts that PG&E meets the criteria in 10 C.F.R. § 50.61 and therefore does not need to use alternative less restrictive criteria in 10 C.F.R. § 50.61a. Id. at 6-1.

V.

SCIENTIFIC ANALYSIS

19. As demonstrated above, the LRA incorporates and depends heavily on previous tests and analyses of RPV embrittlement at DCPP and other reactors for its conclusion that (a) the Unit 1 RPV is entering the period of license renewal in a reasonably safe condition that complies with NRC regulations and that (b) its condition can be adequately managed throughout the license renewal term. In my expert opinion, however, these conclusions are not justified. To summarize my views:

a. As discussed in Section V.A of my 9/14/23 Declaration, in 2002 PG&E withdrew and tested coupons or weld samples from the Unit 1 pressure vessel and conducted Charpy tests for PTS reference temperature and USE. PG&E (2003). In 2003, PG&E reported that it had calculated a limiting RTPTS value of 250oF for the limiting weld 3-442C. Id. Thus, PG&E predicted that in 2021 (the expected retirement date for Unit 1 at that time), the reference temperature for Unit 1 would be slightly more than 10o below the screening limit of 270 oF. Taking into consideration a reasonable margin of

7 error of about 12 oF (as estimated by inspection of the Charpy curves), PG&Es test showed that Unit 1 would be approaching the limit at the end of its operating life.

b. Nevertheless, PG&E discounted the data as not credible. Id. However, PG&E may have found that the data were credible if it had applied standard scientific and NRC guidance for its evaluation. U.S. NRC (1998). PG&Es failure to apply this well-established and reasonable guidance is both inexplicable and gravely concerning, given that the RTPTS data indicated a serious degree of embrittlement. The NRC Staffs approval of PG&Es disregard of the data is also puzzling, given that PG&E had ignored the agencys guidance.

c. Instead of crediting the data it had gathered from Unit 1, PG&E substituted generic data and data from other reactors. As discussed in Section V.C of my 9/14/23 Declaration, PG&Es reliance on substitute data from other reactors was also unreasonable, especially for a period that stretched across decades. Regardless of their initial similarities, all nuclear reactors soon become individualized by unique operating conditions and histories. At the very least, PG&E should have applied a larger error band to any reference temperature calculations that were based on generic data or data from so-called sister reactors. Instead, PG&E is doubling down on its reliance on data from sister reactors.3

d. As also discussed in Sections V.C and V.D of my 9/14/23 Declaration, the results of the 2003 evaluation of the Charpy tests should have motivated PG&E to speed up its schedules for obtaining more data to get a better sense of the pressure vessels condition. At the very least, PG&E should have adhered to its approved schedule for the next capsule extraction and Charpy test in approximately 2009.

e. As also discussed in Sections V.C and V.D of my 9/14/23 Declaration, PG&E should have ensured that the most recent (2005) UT inspection -- which identified one indication... in the beltline region (PG&E (2014)) -- would be followed on schedule with another beltline inspection in 2015. Yet, PG&E repeatedly sought and obtained extensions of time for these measures: the next Charpy test has now been rescheduled from 2009 to 2023 or 2025, depending on whether PG&E is able to withdraw the capsule in 2023 (U.S. NRC (2023)); and the next UT inspection is scheduled for 2025 (U.S. NRC (2015)).

3 In 2011, eight years after informing the NRC that the data from Capsules S, Y, and V were not credible (PG&E (2003)), PG&E relied on data from another reactor to assert that Unit 1 can be safely operated to the end of a 20-year renewal period. PG&E (2011). See Table 4.2-4, showing that the limiting weld 3-442C does not meet or approach the regulatory limit of 270 oF until 54 EFPY, the equivalent of 60 years of operation. The reference document for this prediction is WCAP-17315-NP (Westinghouse (2011)), which relies in part on data from the Palisades reactor to project RTPTS values for the end of the Unit 1 license term.

8

f. In both cases, the extensions leave an unacceptable gap of 20 years between the tests or inspections. In my professional opinion, two decades is an unacceptable amount of time, for two reasons. First, there was no reason for PG&E to rely on questionable generic data or data from so-called sister reactors for more than a short time after the 2003 evaluation. PG&E could have and should have obtained more plant-specific data by now. Second, the condition of the pressure vessel may change significantly over a single decade. See Section V.C. of my 9/14/23 Declaration.

g. In addition, the fact that PG&Es 2005 UT inspection of the pressure vessel were essentially identical to an inspection done 10 years earlier and yielded only one indication of cracking (PG&E (2014)) should have prompted PG&E to evaluate whether the UT inspection was faulty and needed to be repeated. It is reasonable to expect many more indications of voids and cracks, and that they would increase over time. See Section V.B. of my 9/14/23 Declaration.

20. Under these circumstances, it is my expert opinion that the NRC currently lacks an adequate basis to conclude that Diablo Canyon Unit 1 can be operated safely. The NRC Staffs recent decision to allow PG&E to postpone the next Charpy test for Unit 1 until 2025 (U.S. NRC (2023)) is unjustified. As in my 9/14/23 Declaration, I continue to believe that in order to protect the public from the unacceptable risk of a core meltdown accident caused by pressure vessel cracking and fracture during a loss of coolant accident (LOCA), the NRC should (a) order the immediate closure of the reactor by accelerating a maintenance shutdown now scheduled for October, (b) require that the reactor must remain closed pending completion of the next scheduled Charpy tests, (c) ensure that any coupons or capsules that have been withdrawn but were not tested are subject to Charpy tests, (d) account for the data provided by the wedge opening loading (WOL) specimens and the tensile specimens that were scheduled to be contained in the capsules, and (e) ensure that any remedial steps taken by PG&E to address the condition of the Unit 1 reactor pressure vessel are subjected to rigorous review by the NRC Staff, the Advisory Committee on Reactor Safeguards (ACRS), and the general public. See Section VI.A of my 9/14/23 Declaration.

21. For the same reasons, it is also my expert opinion that the NRC lacks a reasonable basis to approve PG&Es license renewal application. Unless and until the NRC establishes that the Unit 1 pressure vessel can operate with a reasonable degree of safety, it has no basis to permit continued operation in a license renewal term.

22. Finally, in Section V.E of my 9/14/23 Declaration, I have offered information that I believe will improve the accuracy of the RTPTS value significantly. In my professional opinion, the newly developed method of nano-indentation promises to be capable of far more extensive results from a single specimen than the conventional Charpy Impact Test methods prescribed by NRC regulations. The more extensive data will permit rigorous statistical analysis, something that is not possible with Charpy because of the limited number of specimens that can be accommodated in a capsule. Importantly, this method has already been applied by Professor Peter Hosemann of the Department of Nuclear Engineering, University of California, Berkeley and found to be sensitive to the change in physical properties of PWR RPV steels brought about by radiation embrittlement.

9 Accordingly, in my professional opinion, the technique requires further application in the field to define and quantify its advantages.

VI.

ADDITIONAL CONCERNS

23. The following additional concerns have arisen since I prepared my 9/11/23 Declaration.

24. Additional concern 1. PG&Es strategy (and the strategy of all other reactor operators) for formulating the lead factor used to estimate the fluence to a certain level of embrittlement has relied on the assumption that embrittlement damage accrues in a non-Markovian manner. A Markovian process is a stochastic model describing a sequence of possible events, in which the probability of each event depends only on the state attained in the previous event; that is, what happens in the present depends on what happened in the immediate past. But this assumption is unproven and likely erroneous. Thus, it is tacitly assumed when carrying out fluence calculations that the reactor is operating at 80

% of the full-power output, thereby fulfilling the condition for a non-Markovian process a priori. However, what happens if the initial fluence damage was accrued at a power level that is significantly below the 80 % level, as is the case for the DCPP during low power testing? Furthermore, the accrual of radiation damage is a highly non-linear phenomenon as evidenced by the shape of the ETC, Equation (1), shown schematically in Figure 1.

Figure 1: Expected variation of RTNDT,41J (T41J) with fluence for Charpy specimens in a surveillance capsule from DCPP, Unit-1 as specified by the Embrittlement Trend Curve (ETC),

Equation (1) according to RG1.99-Rev. 2. The broken lines at any given fluence correspond to the mean value of RTNDT +/-, where is the standard deviation established from the surveillance data for many PWRs in the US fleet. = standard deviation of 9.4 oC (17 oF) for base metal and 15.6 oC (28 oF) for weld metal as depicted in the figure.

25. These data are plotted on a single graph, shown schematically in Figure 1 and it is assumed that ARTNDT,41J (equal to ) should vary with fluence (f) following the RG1.99-R2 ETC (Embrittlement Trend Curve) given by Equation (1)

10

= (..)

(1) where CF is the chemistry factor that adjusts (.., ) elements in the alloy being irradiated that predispose the substrate to neutron irradiation embrittlement (primarily Cu and Ni).

26. Thus, this ETC shows that for a given increment in fluence, the damage accrued, as measured by the shift in the ARTNDT by decreasing amounts from left to right, thereby satisfying the conditions for a non-Markovian process. The capsules are in specific locations in the core between the inner surface of the vessel and the thermal shield, as displayed in Figure 2 for DCPP, Unit 1.

27. This issue is further complicated by the fact that because of the cylindrical geometry of the core, the neutron flux decreases with 1/r2, where r is the radial distance from the core centerline. Thus, the fluence experienced by a capsule depends on its exact radial location with the highest fluence being experienced by the location closest to the core center line. Thus, as the reactor power level changes, not only does the fluence experienced in a capsule also change because of the decrease in the neutron flux (E > 1 MeV),but that at the RPV wall changes by a different amount because of the lead factor.

Because of this, a capsule will generally experience a time-varying fluence that is greater than the inner surface of the RPV with the ratio defining the lead factor as =

/, where is the neutron flux at the capsule and is the flux at the RPV inner wall, but one that has a different time-variance than that at the wall. Lead factors Figure 1: Arrangement of surveillance capsules A, B, C, D, U, V, W, and X in Diablo Canyon Unit 1 NPP. From PG&E, Capsule V Report, WCAP15958.

11 typically range from 2 to 4 in value. This allows reactor operators to look into the future by rearranging the above expression to read = /. Thus, for = 2, for example, and the end of Life (EoL) is 32 EFPYs (40 calendar years), extraction of the capsule at 16 EFPYs will yield the fluence at the inner RPV wall equivalent to that at the EoL (32 EFPYs). Likewise, if the capsule was extracted at 27 EFPYs, the fluence at the inner RPV wall would be equivalent to that at 48 EFPYs or the end of the first 20-year life extension of the reactor.

28. Additional Concern 2: The appearance of Extrema in Capsule V CGraphs and Tables is troubling. Given the discovery of the extrema in ARTNDT,30J, and USE data (maximum in the reference temperature shift and a minimum in USE) and recalling the fact that PG&E rushed to deem all of its pre-2004 surveillance data as not credible in its interpretation of RG1.99 Criterion 3, this issue should be revisited. The DCPP Unit 1 surveillance data were credible but were summarily dismissed in 2003 when it became clear that the 2003 Capsule V results were not favorable (RTPTS=250.9°F). Although these data were also confirmed by the Capsule S Report (RTPTS=258°F), PG&E argues that after they developed a new best-fit curve procedure (CV Graph, a new ETC?), the Capsule Y data that they had previously stated was within RG1.99 predictions was now no longer within the +/-1-sigma standard deviation allowed under RG1.99 Criterion 3.

29. In reality, this one data point for the most limiting weld was only 2 °F out of range. Yet this fairly minor outlier was used as the basis for discrediting all of their prior data sets through a cascading argument that is entirely contrary to NRC procedures for dealing with outliers and how to deem data not credible as per GL9201. To my knowledge, PG&E never actually conducted a detailed analysis of the outlier datum as required by regulations. One must ask, Why is it assumed that regulations were precisely followed, on what appears to be a biased supposition?

30. Additional Concern 3: For all the analyses that I have seen for RG1.99-R2 or 10 CFR 50.61/50.61(a), a serious discussion of errors is seldom included. Thus, beltline components that have a regulatory limit of 270 oF, for example, are deemed to have passed even though they display an ADT,40J of 260 oF. However, given the empirical nature of the field, it is important to determine on a well-founded statistical basis whether the difference of 10 oF is significant. Although error assessment is best done on a statistical basis, particularly if the failure mechanism is poorly defined and the errors cannot be calculated deterministically. A back-of-the-envelope calculation suggests that an error in ARTNDT,40J of +/- 30 oF may not be unreasonable. If so, can we state categorically that a range in ARTNDT,40J of 60 oF, with the upper end of this range being at 270 oF, does not indicate that the vessel has not reached the regulatory limit? This issue should be addressed and resolved before license renewal is approved.

31. Additional Concern 4: I am concerned that PG&E has not addressed the potentially significant role of hydrogen in the embrittlement/crack propagation process, either in the LRA or previous documents on which PG&E relies. Thermally-embrittled ferritic and martensitic steels are notoriously susceptible to hydrogen embrittlement (HE) and hydrogen-induced cracking (HIC) in many other industrial-scale systems, such as the oil

12 and gas production, transportation, and storage systems that they are identified as being principal modes of failure of metallic structures. That HIC occurs in reactor internals, such as those components that are manufactured from Alloy 600 and 182, for example, appears to be indisputable given the work of Totsuka and Smialowska.4 And although it is often claimed that no evidence exists for the role of corrosion in the failure of RPVs, great caution must be exercised because that was precisely that position of many concerning the ID cracking of steam-generator tubing until they were proven wrong by the work of Totsuka and Smialowska in 1987.

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

By Digby Macdonald March 4, 2024 4 N. Totsuka and Z. Szklarska-Smialowska, Effect of Electrode Potential on the Hydrogen-Induced IGSCC of Alloy 600 in an Aqueous Solution at 350 C, CORROSION (1987) 43 (12):

734-738.

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE COMMISSION In the matter of Pacific Gas and Electric Company Docket No. 50-275 Diablo Canyon Nuclear Power Plant, Unit 1 DECLARATION OF DIGBY MACDONALD, Ph.D IN SUPPORT OF HEARING REQUEST AND REQUEST FOR EMERGENCY ORDER BY SAN LUIS OBISPO MOTHERS FOR PEACE AND FRIENDS OF THE EARTH September 14, 2023

ii Table of Contents I.

INTRODUCTION..............................................................................................1 II.

STATEMENT OF PROFESSIONAL QUALIFICATIONS..............................1 III.

SUMMARY

OF EXPERT OPINION................................................................4 IV.

BACKGROUND ON PRESSURE VESSEL AND REGULATORY REQUIREMENTS..............................................................................................7 A. Importance of pressure vessel integrity in a pressurized water reactor........7 B. Importance of reactor-specific surveillance and inspection programs to assess and maintain safe operation.10 V.

DISCUSSION..13 A. PG&E failed to consider credible data showing that Unit 1 is now approaching PTS temperature screening criteria13 A.1 Unit 1 RTPTS surveillance data obtained in 2003, erroneously characterized by PG&E as not credible, show that Unit 1 could approach NRCs threshold for remedial action as early as 202413 A.2 My separate and independent analysis of 2003 Charpy Impact Test data that were deemed credible by PG&E shows that the Unit 1 pressure vessel could reach an unacceptable level of embrittlement at 43.8 +/-10 EFPY..16 B. The most recent ultrasound inspection of reactor vessel beltline welds (2005) does not provide reasonable assurance that Unit 1 is safely operating..21 C. PG&E has obtained no embrittlement data for Unit 1 for 18-20 years, at a significant risk to public health and safety23 D. The NRCs extension of the deadline for beltline ultrasound inspections is not supported by adequate data25 E. Alternative testing methods would provide far more accurate results..26 IV.

CONCLUSION AND RECOMMENDATIONS.32 APPENDIX A: CURRICULUM VITAE APPENDIX B: REFERENCE LIST

iii GLOSSARY OF ACRONYMS ACRS Advisory Committee on Reactor Safeguards AECL Atomic Energy of Canada Ltd ANN artificial neural network ARTNDT Adjusted Nil Ductility Transition Temperature ASME American Society of Mechanical Engineers ASTM American Society of Testing and Materials BWR Boiling Water Reactor CANDU CANada Deuterium Uranium CEFM Coupled Environment Fracture Model CECFM Coupled Environment Corrosion Fatigue Model CGR crack growth rate CIT Charpy Impact Test CRUD Chalk River Unidentified Deposit ECCS emergency core cooling system ECP electrochemical corrosion potential EoE extent of embrittlement EOL end of operating life FAVOR Fracture Analysis of Vessels FoE Friends of the Earth HAZ heat affected zone HIC hydrogen-induced cracking HLNW high-level nuclear waste IGSCC inter granular stress corrosion cracking INL Idaho National Laboratory J

Joules, SI unit of energy MPM Mixed Potential Model NPP nuclear power plant NRC U.S. Nuclear Regulatory Commission

iv ORNL Oak Ridge National Laboratory PG&E Pacific Gas and Electric Company PTS pressurized thermal shock PWR pressurized water reactor RFO refueling outage RRE rate of radiation embrittlement RoA reduction of area upon fracture RPV reactor pressure vessel RTNDT Nil Ductility Transition Temperature RTPTS Reference Temperature for Pressurized Thermal Shock SCC stress corrosion cracking SCK CEN Belgian Nuclear Research Centre SG steam generator SLOMFP San Luis Obispo Mothers for Peace SRM Standard Reference Material SS stainless steel SSM Swedish Radiation Safety Authority TWCF through-wall cracking frequency STP standard temperature and pressure USE upper shelf energy UT ultrasonic testing VP vice president VPM void pressurization model WOL wedge opening loading YS yield strength

I.

INTRODUCTION

1. I have been retained by San Luis Obispo Mothers for Peace (SLOMFP) and Friends of the Earth (FOE) to evaluate changes in Pacific Gas and Electric Companys (PG&Es) program for surveillance of the Diablo Canyon Unit 1 reactor pressure vessel and the adequacy of the justifications provided by the U.S. Nuclear Regulatory Commission (NRC) in support of those changes. My analysis, provided below, supports the Hearing Request and Request for Emergency Action submitted by SLOMFP and FoE to the NRC.

2. The purpose of my declaration is to explain the reasons why, in my professional opinion, the current operation of Diablo Canyon Unit 1 poses an unreasonable risk to public health and safety due to serious indications of an unacceptable degree of embrittlement, coupled with a lack of information to establish otherwise. Therefore, the reactor should be closed until PG&E obtains and analyzes additional data regarding its condition.

II.

STATEMENT OF PROFESSIONAL QUALIFICATIONS

1. I am Professor in Residence at the University of California at Berkeley (UC Berkeley), in the Departments of Nuclear Engineering and Materials Science and Engineering, one of the worlds preeminent nuclear engineering programs. I hold a Ph.D. in Chemistry from the University of Calgary in Canada and B.Sc. and M.Sc. degrees also in Chemistry from the University of Auckland in New Zealand. A copy of my curriculum vitae is attached as Appendix A.

2. I am a qualified expert in the field of materials science with an emphasis on materials in nuclear power reactors (fission and fusion). My areas of expertise include electrochemistry, thermodynamics, applied fracture mechanics, and corrosion science, with emphasis on the growth and breakdown of passive films, chemistry of high temperature aqueous solutions, electro-catalysis, advanced batteries and fuel cells, stress corrosion cracking and corrosion fatigue, materials for nuclear power reactors, and the deterministic prediction of corrosion damage. My experience with the study of corrosion damage includes a wide range of damaging events, including stress corrosion cracking of thermally-embrittled reactor pressure vessel steels and of thermally (weld)-sensitized austenitic stainless steel components in the coolant circuits of water-cooled nuclear power reactors. Radiation embrittlement is often mimicked in the laboratory by using thermal embrittlement to the same physical properties (hardness, yield strength, etc.).

That is common practice when access to a nuclear reactor or another high energy neutron (E > 1 MeV) source is not available, which is often the case in academia. Since completing my Ph.D. in 1969, I have held multiple positions related to nuclear engineering and materials science, which are listed in my curriculum vitae. Most recently, from 2003 to 2012, I was Distinguished Professor of Material Science and Engineering Director for the Center for Electrochemical Science and Technology at Penn State University, again with an emphasis on materials in nuclear power reactors.

3. I have written over 1,000 papers and four books, and I hold eleven patents. My book Transient Techniques in Electrochemistry was the foundational text in the study of electrochemical systems using current and voltage perturbation techniques. These

2 techniques have been used to study certain corrosion-related phenomena in nuclear materials, such as the hydrogen embrittlement of high strength steels and alloys. In 2003, during my tenure at Penn State, I received the U.R. Evans Award, the highest award in the field of corrosion science and engineering, from the Institute of Corrosion in the United Kingdom. In 2011, I was also nominated for a Nobel Prize in chemistry for my work in the passivity of metals in reactive environments and for explaining how such metals (iron, chromium, nickel, copper, zinc, aluminum, zirconium, titanium, etc.) can form the basis of our reactive metals-based civilization. In fact, I reduced that issue to a single mathematical inequality.

4. Regarding nuclear reactors, I developed the Coupled Environment Fracture Model (CEFM) and the Coupled Environment Corrosion Fatigue Model (CECFM) to deterministically model stress corrosion and corrosion fatigue crack growth rate (CGR) in both boiling water reactor (BWR) and pressurized water reactor (PWR) primary coolant circuits. In the case of BWR coolants, a student and I performed an artificial intelligence analysis (using an artificial neural network) of CGR data from both field and laboratory sources. For the CGR in sensitized Type 304 stainless steel (SS), we showed that the CEFM could predict CGR at least as accurately as it can be measured and a similar result was obtained for the CECFM. To my knowledge, the CEFM and the CECFM are the only deterministic models that are currently available for accurate, first principles calculation of CGR in BWR primary coolant circuits. I have used the CEFM to model the evolution of inter granular stress corrosion cracking (IGSCC) damage in 14 operating BWRs worldwide and where comparison with plant data can be made, the agreement between calculated and observed damage is excellent.

5. For PWR primary coolant circuits, I have concentrated on addressing the Alloy 600 steam generator issues by developing the Void Pressurization Model (VPM), a fully deterministic model, to calculate hydrogen-assisted SCC in Alloy 600 that is in contact with primary coolant. Comparison with experimental CGR data again shows that the VPM is also capable of accurately predicting CGR in mill-annealed Alloy 600 under PWR primary coolant conditions. I and a student then developed a Mixed Potential Model (MPM) and demonstrated that because of (a) the large amount of hydrogen that is added to the coolant [25 cc (STP) H2/kg H2O)] and (b) the pH vs fuel burnup protocol commonly employed (the Coordinated Water Chemistry Protocol), the corrosion potential drops below the critical potential for hydrogen-induced cracking (HIC) in the alloy, thereby rendering crack growth spontaneous with the eventual failure of the component (e.g., steam generator tube). We further demonstrated that to maintain the corrosion potential above the critical cracking potential throughout a fuel cycle and thereby address the problem of primary side cracking in steam generator (SG) tubing, the solution is to tailor the coolant hydrogen concentration and/or to modify the pH vs fuel burnup trajectory (by controlling the Li content of the coolant). The MPM is also applicable to analyzing the embrittlement of highly cold-worked Type 316 SS baffle bolts and high alloy hold-down spring in the core structure, for example. Fracture of these, and other components like them (e.g., radiation embrittled RPVs), might be inhibited by the

3 judicious tailoring of the primary water chemistry to ensure that the corrosion potential always remains more positive than the critical potential for HIC in these components throughout the fuel cycle. Coolant-side chemical and electrochemical effects to the cracking of embrittled RPVs are all but ignored in the current NUREGs.

6. At the beginning of my career (1971 - 73), I was employed by Atomic Energy of Canada Ltd (AECL) and became heavily involved in resolving the activity transport problem at the Douglas Point CANDU prototype. In this capacity, in 1971 (est.), I proposed a redox shock strategy for removing the activated CRUD (Chalk River Unidentified Deposit) from the boilers so it could be collected on the filters that are designed to hold activated corrosion products. This resulted in an immediate reduction in the -photon radiation field in the boiler room thereby (as expressed to me by a site VP of AECL) saving the CANDU program. For this accomplishment, I received in 1993 the prestigious W.B.

Lewis Memorial Lecture from Atomic Energy of Canada, Ltd., in recognition of [his]

contributions to the development of nuclear power in the service of mankind. I was only the sixth awardee, with four previous winners being Nobel Laureates. To my knowledge, the redox shock strategy was the first example of electrochemical control in an operating nuclear power plant (NPP).

7. I have been heavily involved as an expert consultant on various reactor issues, including hot-shortness cracking in the Perry Unit 1 BWR suppression pool, flow-assisted corrosion at Surry Unit 1, out-of-specification water chemistry at Calvert Cliffs, and others. Additionally, a colleague and I raised a concern with the continued operation of the Doel-3 and Tihannge-2 PWRs in Belgium, which both contain hydrogen flakes in the pressure vessels. Bogearts (2022). Ultrasonic testing (UT) examination over the years indicated that both the number density and the sizes of the flakes had increased with time, but it was argued by our opponents (primarily from Electrobel and its subcontractors) that perhaps the change reflected enhanced sensitivity of the UT and that the flakes had been present at the manufacture of the vessels. We raised the concern that embrittlement had reduced the fracture toughness so that even a smaller flake could eventually initiate a crack at a lower stress level than would be the case for a non-embrittled steel. We also found that hydrogen flakes had the potential to grow to a dimension that, if properly orientated with respect to the principal stress axis, would have a stress intensity factor exceeding the fracture toughness of the RPV steel. This phenomenon could result in an unstable crack growth rate and failure of the vessel. Given the large size of some existing flakes (> 1-cm), in our opinion the continued operation of the reactors created accidents waiting to happen. Nevertheless, our argument was rejected, and the plants have continued operating.1 1 The NRC, the staff of which are primarily mechanical/nuclear engineers, do not consider hydrogen embrittlement (HE) or hydrogen-induced cracking of radiation-embrittled RPVs in their repertoire of failure mechanisms even though it is considered to be the primary cause of failure of embrittled steels (e.g., of welds in carbon steels) in the oil and gas industry. This

4

8. During the last ten years, I have striven to introduce determinism into corrosion science to accurately predict the evolution of corrosion damage in nuclear systems. Macdonald (2023). For example, under sponsorship of ONDRAF-NIRAS of Belgium, I predicted the evolution of general corrosion and pitting corrosion to carbon steel canisters for the disposal of high-level nuclear waste (HLNW) in Boom Clay repositories over a 100,000-year disposal period, yielding realistic results. Under sponsorship of the Swedish Radiation Safety Authority (SSM), I performed similar work on copper canisters in granitic rock repositories. Prior to that, I was heavily involved in predicting corrosion damage in canisters for the now-defunct Yucca Mountain program and demonstrated that pitting corrosion might lead to the failure of the Alloy 22 corrosion resistant alloy outer layer of the canister. Using the CEFM, I and a student also calculated the CGR in Alloy 22 under Yucca Mountain environmental conditions where the CGR was so low (< 10-11 cm/s) that it cannot be measured experimentally without the imposition of a ripple load (low R-ratio fatigue loading). Our calculations were judged to be realistic and showed that SCC is not a threat to canister integrity.

9. Since the early 1970s, when I was employed by AECL, I have worked to introduce electrochemistry into reactor coolant technology. For that effort, I was recently nominated for the Enrico Fermi Award, perhaps the premier award in nuclear science and engineering.

10. I am familiar with NRC regulations and industry guidance for pressure vessel maintenance and surveillance and the record of PG&Es surveillance program and NRC reviews.

III.

SUMMARY

OF EXPERT OPINION

1. As discussed below in Section IV, the pressure vessel is a uniquely important and vulnerable component in a nuclear reactor, because it holds water on the highly radioactive reactor core, and because it has no backup if it should crack and lose water during an accident. Therefore, compliance with NRC requirements for monitoring the condition of the plant-specific pressure vessel is essential.

2. For pressure vessels, these regulatory requirements are three-fold and complementary:

First, through Charpy testing of samples taken from the reactor vessel, the licensee must demonstrate that the reference temperature for pressurized thermal shock (RTPTS) is below a threshold of 270oF for axially oriented welds and 300oF for circumferential welds. RTPTS is the temperature at which fracture morphology of the pressure vessel changes from ductile to brittle as its temperature drops from the addition of cooling water during a loss of coolant accident (LOCA). Data for the oversight is greatly concerning when it is noted that on the solution side of the RPV is a coolant, a solution of boric acid and lithium hydroxide containing 25-35 cc(STP)/kg H2O of molecular hydrogen. The, n, and radiolysis of the coolant produces a large amount of atomic hydrogen, some of which enters the RPV and further embrittles the steel.

5 fracture energy vs. test temperature are determined from Charpy testing of standard specimens (ASTM 185-82) that had been irradiated in capsules located between the reactor core and the inner surface of the RPV. The capsules are withdrawn at more-or-less equally spaced intervals (typically, every ten calendar years) throughout the reactor life of 32 EFPY (40 calendar years).

Second, also through Charpy testing, the licensee must demonstrate that the pressure vessel is strong enough to withstand the transient stresses induced by thermal shock of the rapidly changing temperature caused by the addition of cooling water, i.e., that the upper shelf energy (USE) will remain above 50 ft-lb.

Finally, every ten years, the licensee must conduct ultrasound testing (UT) inspections of the most vulnerable part of the reactor vessel, the welds around the beltline, to examine for flaws and cracks. NRC guidance appropriately provides that the schedules for these inspections may be relaxed only upon a verifiable demonstration that safety will not be jeopardized.

3. These three types of tests and inspections are complementary in three significant respects.

First, each of the measured phenomena makes a distinct and significant contribution to determining the vulnerability of a pressure vessel to cracking. Second, while the reference temperature and USE calculations are both derived from the same Charpy tests, the method of analysis for each is different; and of course, the UT inspections involve completely different methods of acquiring and analyzing data. Third, each type of test or inspection has a different level of reliability. As discussed below in Section V.A.2, my calculations show that Charpy tests are not particularly sensitive to the extent of embrittlement. Therefore, their results should not be substituted for UT inspections, nor should they be used to justify an extension of the schedule for UT inspections. The three types of data must be considered in unison because they convey important, complementary information on the safety of the RPV.

4. As discussed below in Section IV.B., adequate monitoring of the condition of the pressure vessel is particularly important in the case of Diablo Canyon Unit 1 because the composition of the welds in the pressure vessel was found to be defective at the time it was installed by having excessive copper and nickel. Not surprisingly, in 2006, the NRC identified the Unit 1 pressure vessel among the most embrittled, with only 14 of 72 PTS reference temperatures as high as or higher than Diablo Canyon Unit 1. U.S. NRC 2007.

And today, half of those 14 reactors are closed.

5. As discussed below in Section V.A, in 2002, PG&E withdrew and tested coupons or weld samples from the Unit 1 pressure vessel and conducted Charpy tests for PTS reference temperature and USE. PG&E (2003). In 2003, PG&E reported that it had calculated a limiting RTPTS value of 250oF for the limiting weld 3-442C. Id. Thus, PG&E predicted that in 2021 (the expected retirement date for Unit 1 at that time), the reference temperature for Unit 1 would be slightly more than 10o below the screening limit of 270 oF. Taking into consideration a reasonable margin of error of about +/- 10 oF (as estimated

6 by inspection of the Charpy curves), PG&Es test showed that Unit 1 would be approaching the limit at the end of its operating life.

6. Nevertheless, PG&E discounted the data as not credible. Id. But PG&E may have found that the data were credible if it had applied standard scientific and NRC guidance for its evaluation. U.S. NRC (1998). PG&Es failure to apply this well-established and reasonable guidance is both inexplicable and gravely concerning, given that the RTPTS data indicated a serious degree of embrittlement. The NRC Staffs approval of PG&Es disregard of the data is also puzzling, given that PG&E had ignored the agencys own guidance.

7. Instead of crediting the data it had gathered from Unit 1, PG&E substituted generic data and data from other reactors. As discussed in Section V.C, PG&Es reliance on substitute data from other reactors was also unreasonable, especially for a period that stretched across decades. Regardless of their initial similarities, all nuclear reactors soon because individualized by unique operating conditions and histories. At the very least, PG&E should have applied a larger error band to any reference temperature calculations that were based on generic data or data from so-called sister reactors. Instead, PG&E is doubling down on its reliance on data from sister reactors.2

8. As also discussed in Sections V.C and V.D, the results of the 2003 evaluation of the Charpy tests should have motivated PG&E to speed up its schedules for obtaining more data in order to get a better sense of the pressure vessels condition. At the very least, PG&E should have adhered to its approved schedule for the next capsule extraction and Charpy test in approximately 2009. And PG&E should have ensured that the most recent (2005) UT inspection -- which identified one indication... in the beltline region (PG&E (2014)) -- would be followed on schedule with another beltline inspection in 2015. Yet, PG&E repeatedly sought and obtained extensions of time for these measures:

the next Charpy test has now been rescheduled from 2009 to 2023 or 2025, depending on whether PG&E is able to withdraw the capsule in 2023 (U.S. NRC (2023)); and the next UT inspection is scheduled for 2025 (U.S. NRC (2015)).

9. In both cases, the extensions leave an unacceptable gap of 20 years between the tests or inspections. In my professional opinion, two decades is an unacceptable amount of time, for two reasons. First, there was no reason for PG&E to rely on questionable generic data or data from so-called sister reactors for more than a short time after the 2003 2 In 2011, eight years after informing the NRC that the data from Capsules S, Y, and V were not credible (PG&E (2003)), PG&E relied on data from another reactor to assert that Unit 1 can be safely operated to the end of a 20-year renewal period. PG&E (2011). See Table 4.2-4, showing that the limiting weld 3-442C does not meet or approach the regulatory limit of 270 oF until 54 EFPY, the equivalent of 60 years of operation. The reference document for this prediction is WCAP-17315-NP (Westinghouse (2011)), which relies in part on data from the Palisades reactor to project RTPTS values for the end of the Unit 1 license term.

7 evaluation. PG&E could have and should have obtained more plant-specific data by now.

Second, the condition of the pressure vessel may change significantly over a single decade. See Section V.C below.

10. In addition, the fact that PG&Es 2005 UT inspection of the pressure vessel were essentially identical to an inspection done 10 years earlier and yielded only one indication of cracking (PG&E (2014)) should have prompted PG&E to evaluate whether the UT inspection was faulty and needed to be repeated. It is reasonable to expect many more indications of voids and cracks, and that they would increase over time. See Section V.B below.

11. Under these circumstances, it is my expert opinion that the NRC currently lacks an adequate basis to conclude that Diablo Canyon Unit 1 can be operated safely. And the NRC Staffs recent decision to allow PG&E to postpone the next Charpy test for Unit 1 until 2025 (U.S. NRC (2023)) is unjustified. In order to protect the public from the unacceptable risk of a core meltdown accident caused by pressure vessel cracking and fracture during a loss of coolant accident (LOCA), the NRC should (a) order the immediate closure of the reactor by accelerating a maintenance shutdown now scheduled for October, (b) require that the reactor must remain closed pending completion of the next scheduled Charpy tests, (c) ensure that any coupons or capsules that have been withdrawn but were not tested are subject to Charpy tests, (d) account for the data provided by the wedge opening loading (WOL) specimens and the tensile specimens that were scheduled to be contained in the capsules, and (e) ensure that any remedial steps taken by PG&E to address the condition of the Unit 1 reactor pressure vessel are subjected to rigorous review by the NRC Staff, the Advisory Committee on Reactor Safeguards (ACRS), and the general public. See Section VI.A.

12. Finally, in the spirit of 10 C.F.R. § 50.51(c)(3), I will offer information that I believe will improve the accuracy of the RTPTS value significantly. In my professional opinion, the newly developed method of nano-indentation promises to be capable of far more extensive results from a single specimen than the conventional Charpy Impact Test methods prescribed by NRC regulations. See Section V.E. The more extensive data will permit rigorous statistical analysis, something that is not possible with Charpy.

Importantly, this method has already been applied by Professor Peter Hosemann of the Department of Nuclear Engineering, University of California, Berkeley and found to be sensitive to the change in physical properties of PWR RPV steels brought about by radiation embrittlement. Accordingly, in my professional opinion, the technique requires further application in the field to define and quantify its advantages.

8 IV.

BACKGROUND ON PRESSURE VESSEL AND REGULATORY REQUIREMENTS A. Importance of pressure vessel integrity in a pressurized water reactor

1. At Diablo Canyon and other pressurized water reactors, the reactor fuel core is contained within the pressure vessel, a massive steel structure approximately 30 feet tall and ten feet in diameter, with a wall thickness of approximately 10 inches. A cut-away view of the RPV of a typical Westinghouse PWR is displayed in Figure1. The pressure vessel is normally completely filled with water to keep the core covered and is kept under pressure to prevent the cooling water from boiling at the high temperatures under which the reactor is operated. During normal operation, the pressure vessel and its contents are heated to approximately 550 oF by the nuclear fissioning of 235U92 and toward the end of the core life by fissioning of various isotopes of plutonium such as 239Pu94 and 241Pu94. The region of principal concern in the petition is the beltline region, which is the region of the RPV that is immediately opposite to the core and is depicted in Figure 1 as the 150 active core length. It is this region that experiences the greatest fast neutron flux (E > 1 MeV) and hence fluence and which becomes the most radiation embrittled. Of principal concern is the embrittlement of limiting materials, such as welds and heat-affected zones (HAZ) that are envisioned to be the weakest components when embrittled and hence are those that will likely fail first.

9

2. The reactor pressure vessel, together with the reactor coolant piping connected to it, form the reactor coolant pressure boundary which holds the reactor cooling water. Reactor cooling water must be always kept on the core to prevent the core from overheating and possible melting down even during shutdown because of the decay heat from the spontaneous decay of unstable isotopes (fission products). The melting of the core, should it occur, could release a large quantity of radioactivity into the reactors containment. Should the containment building also fail, this would probably result in the release of significant levels of radiation outside the plant, potential causing deaths, illness, environmental damage, and economic injuries. The Chernobyl accident is illustrative of the scale of potential health and environmental effects and costs, although that reactor did not have containment of the type in Western reactors.

3. Unlike most other reactor safety components, the pressure vessel has no redundant and independent backup system that can be called upon if it should crack or fracture and lose essential cooling water. In the event of water loss from the pressure vessel and uncovering of the reactor core, a nuclear meltdown may occur.

Figure 1: Cut-away schematic of the core of a typical Westinghouse PWR.

10

4. Pressurized thermal shock (PTS) is a reactor pressure vessel condition that can occur during an accident when high pressure combines with sudden decrease in temperature. If core cooling water is lost during a break in the pressure boundary, a loss of coolant accident (LOCA) may occur. In response to such an event, the emergency core cooling system (ECCS) responds by pumping cold water into the vessel. The rapid decrease in the temperature at the vessel wall compared with that further into the wall generates thermal stresses, which together with the stresses induced by the operating pressure of ca.

2250 psi, may act upon a suitably oriented flaw such that the stress intensity factor (KI) exceeds the fracture toughness, KIc. This may result in the rapid propagation of a through wall crack in the embrittled vessel and in the failure of the vessel.

5. If the ductile to brittle transition temperature of the embrittled steel, as characterized by the nil ductility transition temperature or RTNDT, is sufficiently high compared with the unirradiated, non-embrittled steel, the vessel may fail by brittle fracture because of the sudden reduction in the fracture toughness as the temperature moves below RTNDT. This is indicated in Figure 2 where RTNDT is depicted by the inflection points (indicated by the blue arrows) in the hyperbolic tangent dependence of the fracture (Absorb) energy on temperature for both the unirradiated steel and the irradiated steel. These values are quite different from the arbitrarily defined values for RTNDT at 41 J (30 ft-lb) recommended by the ASME Pressure Vessel Code and adopted uncritically by the NRC. Both the RTNDT and the USE are used to judge the susceptibility of the RPV to PTS but the NRC defines RTNDT as that temperature corresponding to a fracture energy of 30 ft-lb (41 J), as indicated by the red-dotted line in Figure 2. These values are significantly different from those indicated by the inflection points.

6. Thus, while it is readily understood as to why RTNDT was defined this way by ASME, ASTM, and the NRC in that it yielded a definite metric corresponding to the intersection of two lines, the more fundamental RTNDT corresponding to the inflection point is also readily determined from the hyperbolic tangent function that is used to fit to the Charpy fracture energy (FE) vs. test temperature data with minimal mathematical manipulation.

Figure 2: Effect of neutron irradiation on the Charpy impact test results for a fluence of 1020 n/cm2 (E > 1 MeV) for A508-3 RPV steel.

After Lin, et.al. Note that irradiation cause the value of RTNDT to shift by about 68 oC (154 oF) and the USE to be reduced by 61 J.

RTNDT

11 It is generally good scientific practice to choose the more fundamentally defined metric if they can all be determined with comparable precision.

B. Importance of reactor-specific surveillance programs to assess and maintain safe operation

1. NRC standards for the condition of reactor vessels are found in 10 C.F.R. Part 50 Appendix G and 10 C.F.R. § 50.61(b). These standards establish two general sets of requirements: for fracture toughness as demonstrated by Charpy upper shelf energy (USE) and the shift in the adjusted nil ductile to brittle transition (ARTNDT) temperature of the embrittled (neutron irradiated) steel microstructure compared with the un-embrittled (unirradiated) microstructure and the fracture resistance to pressurized thermal shock (PTS). Appendix G sets a limit of 50-ft-lbs for the USE in a pressure vessel.

Section 50.61(b)(2) establishes a screening criterion of 270 oF for (RTPTS) for axial welds and 300 oF for circumferential welds, where RTPTS is the reference temperature at the end of a reactors operating life (EOL). If a reactor vessel is predicted to exceed the screening criterion, 10 C.F.R. § 50.61(b)(3) requires that flux reduction measured must be employed. Both sets of requirements must be satisfied.

2. The purpose of a surveillance program is to expose in situ samples of limiting materials

[e.g., plates, welds, heat-affected zones (HAZ), and standard reference materials (SRM)]

in the beltline region in the reactor pressure vessel (RPV) under identical conditions to those experienced by the RPV itself. Because the neutron flux varies with radial distance

() from the core axis roughly as

(), >, where is the radius of the core, the placement of the capsule at a specific radial distance enables the end of life (EOL) fluence to be simulated for an exposure time of less than the design life of the reactor (typically 32 EFPYs or 40 calendar years). This lead factor, which is the ratio of the neutron flux at the capsule and that at the vessel inner surface, is important in the design of an effective surveillance program because it enables the fluence future to be foretold within certain constraints, provided various factors (e.g., operating conditions) remain the same into the future as they were in the immediate past.

3. Equally important is the capsule withdrawal schedule, which typically specifies that one capsule must be withdrawn every 10 years for a four-capsule surveillance program. This is so because a regular withdrawal schedule allows the evolution of radiation embrittlement to be followed and hence to provide consistency in the EOL radiation damage estimates (from all capsules depending on the lead factors). As discussed below in Section V.D, PG&E has postponed this surveillance to such an extent that it completely skipped the withdrawal and testing of Capsule B as originally scheduled for 2007, and now proposes to withdraw the capsule in 2023 or 2025. As a result, PG&E lacks fundamentally important data regarding the condition of the Unit 1 pressure vessel.

4. The regulations also require tensile and fracture mechanics (WOL, wedge opening loading) to be exposed in each capsule along with the Charpy specimens. The tensile specimens are used to measure ex situ the yield stress (YS) and the ultimate tensile stress/strain, both of which are indicative of the state of embrittlement, while the WOL

12 specimen yields a measure of the true fracture toughness, KIC from the crack length upon removal of the capsule and the compliance of the specimen. This is important, because the fracture toughness measured by the Charpy tests is not the same as KIC that is used to determine if a suitably oriented flaw (with respect to the stress axis) in the vessel will grow unstably and possibly initiate a LOCA. Although PG&E appears to have performed the tensile tests, I cannot find any analysis of the WOL specimens. In my opinion, this is an unacceptable omission from the surveillance program for Diablo Canyon Unit 1.

5. Because the strength and fracture resistance of a reactor vessel change over time as the vessel is exposed to radiation and changing temperatures, NRC regulations in Appendix H and 10 C.F.R. § 50.61 Subsection I(2) requires licensees to have a material surveillance program with a schedule for removal and testing of surveillance capsules that conforms to industry standard ASTM E 185. NRC regulation 10 C.F.R. § 50.61I(2)(i) further requires all licensees to integrate the results of their plant-specific surveillance programs into the estimate of reference temperature (RTNDT) for the reactor vessel material.

6. In my professional opinion, the reactor-specific surveillance data required by the NRCs regulations is key to ensuring that a reactor operates in compliance with NRC safety limits. As contemplated by the regulations, generic data and data from so-called sister reactors should not be relied on unless and until the options for obtaining reactor-specific data have been exhausted. In any complex industrial system (nuclear reactor, chemical plant, aircraft, etc.) the judgment that the system is safe to operate must be based on plant-specific data in the same way that a health professional judges the viability of a person to operate successfully in life. That decision cannot be made upon the basis of the health of a sibling, even if that sibling was an identical twin. So it is for a nuclear reactor.

It is for that reason that the NRC mandates a plant-specific surveillance program.

7. In the case of Diablo Canyon Unit 1, obtaining surveillance data specific to that pressure vessel is particularly important because the reactor weld chemistry was deemed defective when the pressure vessel was installed, because of excessive copper and nickel content that render it more vulnerable to embrittlement. The excessive copper (approx. 0.2 %)

arises from the corrosion protective copper coating on the weld wire employed and the excessive nickel content of approx. 1 % originates from the composition of the weld wire itself. The deleterious impact of both copper and nickel in the radiation embrittlement of welds in ferritic steels has been established by numerous laboratory and field studies.

After Diablo Canyon Unit 1 was completed, the error was realized, and Unit 2 did not contain excessive Cu and Ni in the welds.

8. The number of capsules needed for a reactor vessel surveillance program is established with reference to the ASTM standard. In the case of Diablo Canyon, to satisfy the requirements of ASTM E 185-73, PG&E started with a five-capsule program based on the estimated shift in the adjusted nil ductility reference temperature above 200o F. PG&E

13 (1992).3 In 2006, for unexplained reasons, the NRC re-characterized the surveillance program as a four capsule program. U.S. NRC (2006). Whether characterized as a 4 or 5-capsule program, each program was designed for the current license term and included a schedule for removal of Capsule B about midway through the current license term (EFPY 19.2 or EFPY 20.7, RFO 14 and RFO 15 in the period 2007-09).4

9. The data collected by a reactor vessel surveillance program is useful both for assessing the current integrity of the reactor vessel and for projecting its condition in the future.

Thus, for example, PG&Es surveillance program, as approved by the NRC in a 2006 license amendment for recapture of the low-power testing period, required removal of Capsule B at 20.7 EFPY. U.S. NRC (2006). This timing would allow PG&E to obtain data about the current condition of the vessel. It would allow provide information about the fluence of the vessel at the end of the license renewal term, or approximately twice the projected limiting inside RV fluence for DCPP-1 [Diablo Canyon Unit 1] at the EOL (i.e., approximately 2

• 1.43 x 1019 n/cm2 (E > 1.0 MeV]. U.S. NRC (2006).

10. And while the number of capsules inserted into a pressure vessel cannot be changed (other than by adding more of them for future assessment), the schedule can be adjusted to accommodate the demands of the surveillance program. For instance, if a set of surveillance data from a particular capsule turns out not to be credible, the licensee may remove other capsules if the altered schedule change is consistent with the industry standard.

11. In my professional opinion, the most important reason for changing a surveillance schedule, other than adjusting to new information regarding vessel fluence, would be to provide additional data where available data had proven to be insufficient. It would not be reasonable, however, to change a capsule removal schedule for any other purpose if the change would leave the surveillance program with a gap of ten or more years.

12. The measurement of RTNDT and USE is only part of the story in assessing whether an embrittled RPV is in danger of rupture particularly under pressurized thermal shock (PTS) conditions resulting from the injection of cold water to compensate for loss of coolant from the rupture of the pressure boundary elsewhere. While ARTNDT and USE are appropriate monitors of the state of embrittlement, the probability of crack nucleation is a question that must be addressed by probabilistic fracture mechanics that requires the assessment of the population, size, and orientation of flaws close to the cladding/steel interface. Therefore, UT is used to evaluate flaw volume density (#/cm3), flaw size, and flaw orientation so as to determine if any flaw is characterized by a stress intensity factor (KI) that exceeds KIC for the embrittled steel. The American Society of Mechanical 3 PG&E inserted Capsule B into the Unit 1 pressure vessel and the NRC approved a schedule for withdrawing and testing it when the reactor achieved 19.2 EFPY. Id. See also Table 4. In 2006, in approving a license amendment for recapture of the three years of low-power testing of Unit 1, the NRC approved a change in the withdrawal schedule to 20.7 EFPY. U.S. NRC 2006.

4 This schedule can be derived from PG&E (1992), Enclosure at 3-4, Table 4; U.S. NRC (2006),

Safety Evaluation at 5; and PG&E (2023), Enclosure 2.

14 Engineers (ASME) code that is incorporated by NRC regulation 10 C.F.R. § 50.55a requires that an UT inspection must be performed every ten years.

V.

DISCUSSION A. PG&E failed to consider credible data showing that Unit 1 is now approaching PTS temperature screening criteria.

A.1.

Unit 1 RTPTS surveillance data obtained in 2003, erroneously characterized by PG&E as not credible, show that Unit 1 could approach NRCs threshold for remedial action as early as 2024.

1. In my professional opinion, PG&E has incorrectly discredited the data it obtained from Unit 1 in Capsules S, Y and V for the purpose of calculating RTPTS values. PG&E should have been concerned that these data showed that Unit 1 could approach the PTS temperature screening limit by the end of the reactors initial license term and should have investigated the reasons for anomalies in the data. Yet, in disregard of common scientific practice methods and NRC guidance, PG&E claimed the data were not credible. PG&E (2003).

2. In 2003, PG&E tested data from recently withdrawn Capsule V. According to PG&E Letter DCL-03-052, at Unit 1s EOL date of 32 EFPY (which at that time was 2021), the limiting RTPTS value calculated by PG&Es contractor, Westinghouse, for the limiting weld 3-442C was 250.9 oF. PG&E (2023), Westinghouse (2003). This calculation should have concerned PG&E because it was approaching the PTS screening criterion of 270 oF for plates, forgings and axial weld materials and within a reasonable margin of error of about +/- 10 oF (as estimated by inspection of the Charpy curves), resulting in an overlap of uncertainties in the screening criterion (270 oF) and the Westinghouse estimate (250.9 oF) for weld 3-442C. In addition, as further explained in Section V.A1, the fact that the measured RTNDT for Capsule V (201.07 oF) was lower than the value for Capsule Y that had been removed ten years earlier at 1R5 (232.59 oF) (Westinghouse (2003), Table D-2) indicated a reasonable possibility that one of those tests was erroneous, because it unlikely that continued exposure to radiation would heal the metal. If the value of Capsule V was erroneous and the value of Capsule Y was correct, then the limiting RTPTS value Unit likely was even closer to the PTS screening criterion than calculated by PG&E.

3. Despite these concerning results, PG&E discredited all of the data it had obtained from Unit 1 in Capsules S, Y and V, based on a determination that the best fit curve between the Capsule V data and data from earlier-withdrawn Capsules S and Y contained scatter values for two data points that exceeded the criteria in Regulatory Guide (RG) 1.99, Rev.

2, Criterion 3 (U.S. NRC 1988)). According to RG 1.99, the scatter values for data normally should be less than 28oF for welds and 17oF for base metal PG&E (2003),

Westinghouse (2003). This is equivalent to +/- 1 Sigma. Therefore, PG&E declared that all

15 the data from Capsules S, Y and V were not credible for the purpose of calculating limiting RTPTS values. PG&E (2003).5

4. PG&Es methodology for assessing the credibility of the data is inconsistent with NRCs own guidance for performing credibility assessments. U.S. NRC (1998). At page 11, the guidance states as follows:

A. If there exists an identified and recorded deficiency in a datapoint - a duplicate or untraceable record, a record which identifies an atypical condition or sample location, or B. If a datapoint is identified as a statistical outlier and a physical basis exists for believing the datapoint to be atypical -

All data not excluded in (A.) should be used as the dataset A priori exclusion of some data based on inconsistency with expected norms should not be used before analysis for statistical outliers is conducted.

(Italics mine). In violation of the NRC guidance, PG&E excluded not just inconsistent data but all of the data a priori, without conducting an analysis for statistical outliers.

5. In addition, the rejection of all the data because one datum did not fall within the bounds by a narrow margin does not conform with accepted scientific and engineering practice.

In analyzing scattered data, it is common to find points that lie outside of a preconceived scatter band. If the scatter band has been established via the analysis of a significant population of historical data for identical samples from the same system (reactor) and it is established that the data follow a normal distribution, it is possible to define the width of the scatter band in terms of the standard deviation with the next sample having a 68 %

probability of falling within the mean +/- one standard deviation or a 96 % probability for falling within a +/- two standard deviations,and so forth. However, there is a finite probability that future values of RTNDT and USE will lie outside of these limits (32 % and 4 %, respectively). That is the inherent nature of experimental data.6 For a system as critical as a beltline weld, for example, a margin of error of the mean +/- one standard 5 As discussed in Section V.A.2 below, separately, PG&E found that the USE data from Capsule V do not indicate excessive embrittlement. USE remains above 50 ft-lbs to the reactors end of life (EOL) or 32 EFPY, as required by 10 C.F.R. Part 50, Appendix G. My own analysis of the USE data, however, demonstrates that Unit 1 may reach an unacceptable level of embrittlement at 43.8 EFPY or earlier.

6 If the data from a single reactor are insufficient, it is possible to examine data from another reactor to evaluate whether the distribution is normal. But if the data are not from the same system, a systematic error will likely be introduced, the magnitude of which could vary widely from one data set to another from different reactors. If sufficient data were available from two sister reactors it is unlikely that they follow the same standard normal distribution since each reactor is unique because of unique operating conditions and histories. Under these circumstances, defining the uncertainty in terms of a standard deviation becomes problematic.

16 deviation is too tight and in my professional judgement the probability and consequences of failure are too high.

6. Even if the use of the standard deviation is correct and I had established the correlation with three data points (as is the case for Diablo Canyon Unit 1) and found the distribution to be normal, and I added one more datum that was from the same population, there is a 0.32x3 = 0.96 (= 1) probability that the datum will fall outside the mean +/- one standard deviation for no obvious reason. Thus, the observation that one point in the Diablo Canyon Unit 1 correlation fell outside the error band is statistically insignificant (bordering on the nonsensical) and calls into serious question the invalidation of the Capsule S, Y, and V data by PG&E.

7. PG&E also departed from standard scientific practice in failing to plot the data it relied on, relying instead on a narrative. Nowhere can I find the actual graphical presentation of the correlation of RTNDT with fluence so that I can judge for myself the validity of PG&Es non-credibility claim. Given the safety significance of PG&Es rejection of the Unit 1 surveillance data, its failure to fully disclose the quantitative data on which it relies constitutes a serious violation of normal scientific and engineering practice. Furthermore, I can find no attempt by PG&E to establish the assumption that the data follow a standard normal distribution, which must support any analysis and specification of a standard deviation. Many physical phenomena follow a lognormal distribution that could significantly change the conclusions arrived at by PG&E.7

8. Accordingly, for any point that does lie outside of the limits, especially far outside the limits, the first course of action should be to ascertain whether there is a valid physicochemical reason for the anomalous result. If a valid reason can be found, such as an experimental error, then that datum is treated as an outlier and can be excluded from the analysis of the remaining data. Importantly, where outliers exist, they do not provide a valid reason for discrediting the data that do meet the criteria for credibility.

9. It is also unreasonable to reject otherwise plausible data out of hand when the entire available data set is so small. The only reasonable solution to the problem that the scatter values exceeded the NRCs criteria was to gather more data and compare it to the existing data. Had PG&E collected and tested more data, then the appropriate placement of the best fit curve in the correlation would have been more reliably established and it would have been more difficult to throw the data out. Gathering the data from Capsule B and testing those data along with Capsule C is an essential step toward improving the size of the data pool and thereby the quality of the analysis.

10. Had PG&E appropriately credited its own data, it would have had to take remedial measures to ensure the integrity of the pressure vessel, as required by Section 10 C.F.R.

50.61a. Instead, as discussed below in Sections V.C and V.D, PG&E relied for an 7 Underlying this whole issue is the paucity of data from the Charpy test. See Section V.A.2 above.

17 extended period on data from other reactors to justify continued operation and postponed any further testing or inspection of the reactor vessel.

A.2 My separate and independent analysis of 2003 Charpy Impact Test data that were deemed credible by PG&E shows that the Unit 1 pressure vessel could reach an unacceptable level of embrittlement at 43.8 +/-10 EFPY.

1. The paucity of plant-specific data from 14.27 EFPY (when the Capsule S was withdrawn and tested (PG&E (2023)), to the EOL EFPY of 32 is a problem of the utmost seriousness, particularly when one realizes that data from one or both of Capsules Y and V are suspect for reasons speculated upon elsewhere in this Declaration. Leaving aside for the moment PG&Es unjustified attempt to exclude all plant-specific data, the paucity of data could stretch from 5.87 EFPY or even from 1.25 EFPY to the EOL at 32 EFPY.

This is an intolerable situation that essentially means that neither PG&E nor the NRC have a defendable estimate of the time that it will take for the weld to achieve the critical condition of USE = 50 ft-lb. This deficiency is addressed below in my reanalysis of PG&Es Charpy data using completely new methodology for analyzing those data. Using that methodology, I calculate that the critical condition will be reached at 43.8 EFPY with an estimated uncertainty of +/- 10 EFPY.

2. Given PG&Es failure in 2003 to present any Unit 1-specific evidence regarding the rate of embrittlement over time, I developed a model that would use the Charpy Impact Test (CIT) data deemed credible by PG&E to determine the Extent of Embrittlement (EoE) over the life of Diablo Canyon Unit 1.

3. USE measurements or CIT data for nuclear reactor pressure vessels provide a direct experimental quantification of the degree of embrittlement over time. For the 2003 USE evaluation, PG&E and Westinghouse determined that the CIT data were credible. PG&E (2003), Westinghouse (2003). For my own review, I have consulted the CIT data for three reasons: first, because PG&E deemed them credible in contrast to the RTNDT data; second, because they are unencumbered with corrections, such as the chemistry factor, margin, and the fluence factor that are required to correct RTNDT to a specific material in a specific plant; and third, because the USE is more directly related to the degree of embrittlement than is the adjusted RTNDT.

4. By mathematically deriving an expression for the EoE from coefficients (A, B, C, and T0) obtained for the symmetric hyperbolic tangent function (= +. tanh [()/)

that is used by PG&E to optimize on the fracture energy (FE) vs test temperature CIT data, I have calculated = 1 +

/2 and = (,)/ where

, is the transition temperature that is defined for a fracture energy of 30 ft-lb (41 J). The EoE are plotted as a function of fluence in Figure 3. The expression for EoE tacitly assumes that the EoE also follows the hyperbolic tangent function given above where the point of inflection,=. By my reasoning,, is a much better definition of the nil-ductility transition temperature than is the arbitrarily defined

,, as noted above. Note that at the point of inflection (PoI), the EoE = 0.5 indicating that the fracture is 50 % brittle and 50 % ductile. As we will see below, this

18 ratio of brittle vs. ductile fracture is close to the ratio (= 1.1) at the critical condition defined by the NRC of 50 ft-lb.

Figure 3: Values for EoE derived from the CIT data of PG&E for metal specimens from Capsules S, Y, and V that were exposed in Diablo Canyon Unit 1.

5. As we see from Figure 3, the EoE for the weld metal is significantly greater than that of the plate, HAZ, and SRM samples showing that the weld is the most susceptible of the samples contained in Capsules S, Y, and V that were exposed in Diablo Canyon Unit 1.

6. This difference is addressed as follows. When choosing a technique to monitor a selected phenomenon in a well-designed experiment, it is essential that the dependent variable (the measure of the phenomenon, e.g., the EoE) have a high sensitivity to the principal independent variable, in this case, the fluence. Figure 3 reveals that the CIT has different levels of sensitivity for different materials. For the plate, HAZ, and SRM, the CIT is not very sensitive to the extent of embrittlement, with EoE changing by no more than 3 %

over the first 14.27 EFPY operating life of the reactor. In contrast, for the weld metal, the EoE changes by about 8 %. Of course, the lack of sensitivity may also reflect that the plate, HAZ, and SRM do not embrittle rapidly, at least up to a fluence of 1.37x1019 n/cm2. Fortunately, the CIT does effectively detect the embrittlement of the limiting weld material.8 8 In my opinion, the CIT should be replaced, or at least complemented by another technique that does meet that standard of high sensitivity of the dependent variable on the principal independent variable. Such a technique might be nano indentation that is recognized by the NRC (U.S. NRC (1988) and currently being further developed by Prof. Peter Hosemann in the Department of Nuclear Engineering at the University of California at Berkeley (see below). While indentation is recommended by the NRC as an optional technique, in my opinion it should be made mandatory in reactor surveillance programs.

EoE = 5E-21f+ 0.4204 R² = 0.8235 0

0.1 0.2 0.3 0.4 0.5 0.6 0

2E+19 EoE Fluence /n.cm-2 EoE vs. Fluence Weld Plate HAZ SRM Linear (Weld)

19

7. As demonstrated by my methodology, the EoE for the plate, HAZ, and SRM changes by no more than 3 % over the entire 14.27 EFPY at the withdrawal of Capsule V from the reactor while that for the weld metal changes by about 8 %; and (b) The final issue of the time that it will take to achieve the critical condition of the USE being reduced to 50 ft-lb has not so much to do with the CIT, itself, as it has to do with PG&Es analysis of the data obtained using the CIT.

8. It is also important to note that my methodology differs from the traditional approach of assessing USE changes over time. I have observed that most, if not all engineers and scientists skilled in the science of radiation embrittlement accept the view that whatever metric is adopted for monitoring the progression of radiation embrittlement (,,

,, USE) the metric should change monotonically with increasing fluence and approach a plateau asymptotically at very high fluence. However, by all metrics examined by me, the extent of embrittlement as determined from PG&Es Charpy data passes through a maximum (,,,) or a minimum (USE) with increasing fluence, which is at odds with theoretical expectation. The rationale for my expectation of monotonic change is that the metal displacement reaction can be written as +

+ where n is the concentration of high energy neutron in 1 cm3 of the metal in their transit from the entrance to the exit face of the metal cube and,, and are the concentrations of metal atoms, metal interstitials, and metal vacancies, respectively in the same volume. The rate of formation of displaced atoms (i.e., interstitials) can be written from chemical rate theory as:

/

= (1 )

[

] where

[] is the concentration of displaced metal atoms (#/cm3), is the fluence at the 1 cm2 input face of the metal cube, and is the neutron absorption coefficient in the metal.

Note that the thickness of the cube of metal is 1 cm. At steady state and at limitingly high fluence

/

0 and we obtain

(1 ). This corresponds to the steady state initiation of damage as measured by the concentration of displaced metal atoms alone.9

9. Using the assumptions and methods set forth above, I now proceed with calculating when the beltline weld material will become unacceptably embrittled as reflected by the USE dropping below 50 ft-lb (41 Joules (J)). Thus, a plot of USE vs. EoE for all materials in Capsules S, Y, and V is displayed in Figure 4.10 All the data are found to follow a single 9 This simple model is incomplete in that it does not consider cascading, in which the displaced atom moves through the lattice and induces further displacements. But the model provides a reasonable physical account of the initial events in the embrittlement phenomenon. In addition, the equation is first order in fluence and cannot predict an extremum (maximum or minimum).

That would require at least a second order dependence on fluence, i.e., of the form mi = Af2 + Bf

+ C, where A, B, and C are constants.

10 I note here that the measured USE data passes through a minimum, indicating that, somehow, the damage heals with increasing fluence from Capsules Y to V. This seems unlikely if not

20 locus that is represented by the equation = 9.4378. with the plot being characterized by R2 = 0.9976, indicating a high goodness of fit. Substitution of USE =

50 ft-lb yields the critical extent of embrittlement (EoEcrit) of 0.525; that is, the fracture is predicted to comprises 52.5 % of brittle fracture (47.5 % ductile fracture) when the USE is reduced to the NRC-imposed lower limit of 50 ft-lb (41 J). This critical condition is shown as the orange data point in Figure 4. From the correlation shown in Figure 3, the critical EoE will be reached at a fluence of 2.09e19 n/cn2, E > 1 MeV. Note that the ratio of brittle vs. ductile facets on the fracture surface (ratio = 1.1) is close to that defined by RTNDT,PoI (ratio = 1) thereby supporting my conclusion that RTNDT,PoI is a more fundamentally-based and hence superior metric for defining the state of embrittlement than is RTNDT,30.

Figure 4: Plot of USE vs. EoE for all materials from Capsules S, Y, and V, Diablo-Canyon, Unit 1 NPP.

10. In Figure 5, I plot the fluence vs the EFPYs when Capsules S, Y, and V were withdrawn from the reactor. The data, although of significant paucity, are adequately represented by the equation given in the figure as shown by the high goodness of fit (R2 = 0.9939).

Extrapolation of the data to the critical fluence of 2.09e19 yield the time at which the USE of the weld (24702) in the beltline equals the 50-ft-lb limit. That time is calculated as 43.8 EFPYs and is represented by the last datum on the right side of Figure 5.

Inclusion of this point in the fitting yields the same equation but with R2 = 0.9911. Thus, the weld is predicted to meet the regulatory minimum USE in about 55 calendar years after the original, adjusted startup date or 2039. Upon consideration of these various impossible based on current knowledge, and may have resulted from discrepancies in the testing methods over time - or possibly by transposing the results from Capsules Y and V. This issue should be carefully examined by PG&E. Nevertheless, PG&E initially accepted the data as being credible.

USE = 9.4378EoE-2.59 R² = 0.9976 0

20 40 60 80 100 120 140 160 0.3 0.35 0.4 0.45 0.5 0.55 USE /ft-lb EoE USE vs EoE for all Materials from Capsules S, Y, and V, Diablo Canyon, Unit 1.

21 contributions to the total uncertainty, I estimate that the uncertainty in the time taken for the weld to reach fracture criticality is about +/- 10 EFPY. The uncertainty band appears to be dominated by the asymptotic nature of the curves (blue points) USE vs. EoE and Fluence vs. EFPY, as plotted in Figures 4 and 5, respectively. As a result, fracture criticality could be reached as soon as 33.8 EFPY, which is soon after the EOL of 32 EFPY, or as long as 53.8 EFPY, but safety prudence dictates that the lower number of 33.8 EFPY should be adopted. In my opinion, the uncertainty could have been reduced significantly had PG&E adhered to the capsule withdrawal schedule that was initially accepted from the NRC and had they followed the accepted scientific analytical method, as sanctioned by the NRC for the exclusion of identified problematic data.

Figure 5: Plot of fluence vs the EFPYs when Capsules S, Y, and V were withdrawn from the reactor.

11. There is uncertainty in this projection, arising from four sources: (a) the inherent uncertainty in the data themselves; (b) the lack of any capsule surveillance data after 14.27 EFPYs; (c) the shape of the curves, particularly those in Figures 4 and 5, and (d)

The length of the extrapolation, which is really a consequence of (b) above. Regarding the accuracy of USE, examination of the Charpy Impact Test data in WCAP-15958 suggests that the data are accurate to about +/- 5 ft-lb. This number is important is determining the time at which the weld reaches the critical condition because, as shown in Figure 4, the USE vs. EoE plot approaches a limit asymptotically indicating that any uncertainty in USE becomes an increasingly larger uncertainty in EoE as the fluence increases. Thus, from Figure 5, this error is propagated into a corresponding uncertainty in the critical fluence that, in turn, is transferred to an uncertainty in the EFPY at which the critical condition is reached.

12. This analysis does not predict that the radiation embrittlement damage passes through an extremum (maximum or minimum) as is shown by PG&Es data (see, for example, the two highest fluence points in Figure 3), as that would require the expression for F = 5E+18ln(EFPY) + 2E+18 R² = 0.9939 0.00E+00 5.00E+18 1.00E+19 1.50E+19 2.00E+19 2.50E+19 0

10 20 30 40 50 Fluence n/cm2 EFPY Fluence vs. EFPYs

22 (given immediately above) to be a quadratic in the Fluence at the least. It seems more likely that the extrema simply reflect erroneous experimental technique and/or data analysis or that the data from Capsules Y and V were somehow transposed. Regardless of the speculated reason, if PG&E followed accepted scientific practice, they should have immediately inquired as to the reason for this anomalous result, but I can find no evidence that this was ever done. It is likely that this apparent sloppiness is responsible for the outliers that caused PG&E to reject all the data from Capsules S, Y, and V and leave them with no plant-specific data for Diablo Canyon Unit 1. Had they found the cause and identified the specific points in error, normal scientific practice would have justified rejection of those data while retaining the rest. As discussed in Section V.C, PG&E should have obtained more data by withdrawing and testing Capsule B, by testing other capsules that had already been withdrawn, by adding tensile strength testing, and by conducting a thorough ultrasound inspection.11 B. The most recent ultrasound inspection of reactor vessel beltline welds in 2005 does not have credible results and therefore does not support a finding that Unit 1 is safe to operate.

1. I am concerned by PG&Es 2014 statement that the results of its 2005 UT inspection of the pressure vessel were essentially identical to an inspection done 10 years earlier and yielded only one indication of voiding/cracking. PG&E (2014). It is reasonable to expect many more indications of voids and cracks, and that they would increase over time. For instance, in UT examinations of the Doel-3 and Tihannge-2 PWRs in Belgium conducted in 2012, up to 40 indications per cm3 were detected in the Doel-3 reactor for a total of 7,776. Bogaerts et.al. (2022). Additional tests conducted in 2014 with adapted equipment detection parameters, revealed 13,047 voids and cracks in Doel-3 and 3,149 voids and cracks in Tihannge-2. Indications were found at depths ranging from 30 to 120 mm measured from the primary water side. Note that the thickness of the stainless-steel cladding is 7 mm, so that the indications occurred at 23 to 113 mm from the cladding/RPV steel interface. The indications were concentrated in the bottommost and upper core shell and were located in base metal, outside of the weld regions. These features can be correlated to steel microstructure and thermo-mechanical history (theoretical modeling) according to SCK-CEN, the Belgian Nuclear Research Centre.

These indications were identified as hydrogen flakes and were postulated by Electrobel as having formed via excess humidity at the time of casting of the steel. However, the number of indications appear to be increasing with time which indicates that atomic hydrogen is entering from the primary side via the radiolysis of the H2-rich primary side coolant (the PSC contains about 25 ccSTP) of hydrogen per kg of water), diffusing to and recombining in voids (e.g., clusters of metal vacancies), so as to pressurize the voids and causing the voids to grow on number and in size with some eventually transitioning into cracks.

11 While we are aware that Capsule B apparently did not contain and beltline weld specimens, testing nevertheless would provide useful data.

23

2. As shown by Bogaerts et.al. (2015), the microstructure contains both brittle (red arrows) and ductile (blue arrows) features, Figure 1, indicating mixed mode cracking not unlike that observed in other RPVs. Spencer and coworkers at INL have modeled RPV embrittlement within the Grizzly and FAVOR [Fracture Analysis of Vessels] codes.

Spencer et.al. (2015, 2016). These are computer algorithms that were developed at Idaho National Laboratory (INL) and Oak Ridge National Laboratory (ORNL), respectively, for modeling the embrittlement and physical changes to RPVs under neutron irradiation.

Typical distributions of the number of flaws in a RPV with respect to RTNDT as predicted by FAVOR and Grizzly are shown in Figure 7. FAVOR, which was developed at the ORNL, is acknowledged as providing an accurate prediction of the number and distribution of flaws in a PWR RPV and Grizzly are found to be in excellent agreement except for at the tail for RTNDT < 120 oF.

Figure 6: Typical hydrogen flake cracking in carbon or low-alloy steel. Typical features of hydrogen-induced brittle fracture are: micro-quasi-cleavage fracture, pores and fine hair-lines (indicating ductile fracture on a micro-scale). After Bogaerts et.al. (2015)

24

3. Accordingly, it is difficult to accept and understand PG&Es claim of detecting only one indication in the 2005 UT examination of beltline materials at Diablo-Canyon, Unit 1, when Figure 7 indicates thousands as determined by summing the number of indications for each bar. In my professional opinion, therefore, the anomalous results of the 2005 UT inspection should have prompted PG&E to evaluate whether the UT inspection was faulty and needed to be repeated. Instead, PG&E sought and obtained a ten-year extension of the 2015 deadline for the next UT inspection, until 2025. PG&E (2014),

U.S. NRC (2015). See also Section V.D. below.

C. PG&E has obtained no embrittlement data for Unit 1 for 18-20 years, at a significant risk to public health and safety.

1. In my opinion, PG&Es failure to obtain embrittlement data since 2003 (Charpy test) and 2005 (UT inspections), plus the questionable quality of those tests and inspection, and on top of indications that embrittlement was occurring at a significant rate, raises serious questions that should be addressed immediately.

2. My concern stems in part from the complex nature of radiation embrittlement, which is idiosyncratic to individual reactors and may change unexpectedly over time, including periods of time less than a decade. Radiation embrittlement is a progressive phenomenon that increases with fluence, but which also depends on temperature. Thus, as the metal component of interest, is irradiated with high energy neutrons (E > 1 MeV), the fluence increases monotonically. The fluence, which is the neutron flux multiplied by the time of irradiation is, itself, independent of temperature but the rate of accumulation of damage in the metal is temperature dependent. This is because the various processes that contribute to the accumulation of damage, including the displacement of atoms into interstitial positions, the diffusion of the vacancies and interstitials through the lattice, the multiplication of the interstitial/vacancy pairs through cascading, the condensation of vacancies into clusters at impurities in the lattice that may grow into microscopic voids Figure 7: Comparison of RTNDT distributions in the same plate analysis in Grizzly and FAVOR. After Spencer et.al. (2016).

25 and eventually form the macroscopic defects at which unstable cracks may nucleate under PTS conditions, and the recombination of interstitial/vacancy pairs, are thermally activated processes whose rates are temperature dependent.

3. Thus, while the fluence may be determined from the flux and the irradiation time regardless of the temperature, that is not the case for the irradiation damage.

Westinghouse/PG&E calculate the fluence as though the reactor operates at full power for 80 % of the calendar years with the remaining 20 % accounting for downtime such as refueling. The resulting effective full power years (EFPYs) is therefore independent of whether the reactor operated at reduced power for periods (and hence reduced temperature) throughout the cycle or whether it operated at full power provided the end fluence was the same. However, this is not the case for the accumulated damage because the processes that contribute to the net damage are all thermally activated whose rates are temperature dependent. Because of this, the accumulation of damage depends upon the temperature history of the component, i.e., on the power level history. Thus, the case can be made that specifying RTPTS at a critical fluence would be better recast as RTPTS at a critical level of accumulated damage as measured by hardness, for example. This would appear, then, to fairly consider the effects of both temperature and fluence on the EFPYs required to achieve critical conditions.

4. I am also concerned by PG&Es reliance on data from so-called sister reactors that supposedly have similar characteristics. While this may be permissible as a stop-gap measure, PG&E has relied on data from other reactors for decades, instead of obtaining more data from Unit 1. As I have discussed above, complex industrial systems begin to differ in their characteristics almost as soon as they begin to operate. As has been noted by me and others, even if two nuclear plants are identical in every respect (and sister nuclear reactors never are), each soon becomes individualized by unique operating conditions and histories. Accordingly, in establishing correlations between accumulated damage (e.g., as measured by USE and/or RTNDT) and fluence or EFPYs from many sister plants, this uniqueness must be recognized and built into the correlation.

5. Thus, if the sister plants were identical even after unique operating histories and the damage was normally distributed with respect to EFPY (a significant and poorly established assumption), a 1 sigma scatter band would yield a probability of only 68.2% that an additional datum added to the correlation would fall within that band (Figure 3). In my professional opinion as a scientist and an engineer, that probability is too low to be used for judging the probability of embrittlement in the Diablo Canyon Unit 1 vessel. However, because the sister plants and Diablo Canyon Unit 1 do have unique operating histories a larger uncertainty (standard deviation) should be assigned that would significantly increase the width of the scatter band. Given the above, it is my opinion, that the 2-sigma scatter band, corresponding to a roughly 95.4 % probability that an additional plant (e.g., Diablo Canyon Unit 1), and as specified in RG1.99, would fall within that band and would be more appropriate. By that standard, any legitimacy to

26 PG&Es decision to discredit the results from Capsules S, Y, and V collapses.

Figure 8. The normal distribution function displaying the probability of an additional observation falling within, where n = 1,2,3,..

6. Many uncertainties, including the memory effect arising from different operating histories arise in describing the evolution of radiation embrittlement damage that are not explicitly accounted for in the evaluation of correlation between RTNDT and fluence.

Thus, numerous studies on the rupture of pipes in NPPs have established that the underlying statistics are Markovian, which specifies that what happens now depends on what happened in the past. I refer to this as the memory effect and, when applied to radiation embrittlement of NPP RPVs indicates that the rate of radiation embrittlement (RRE) in the present depends on the factors that controlled the RRE at some past time.

For example, it is well established that the RRE is a function of temperature because the recombination of displaced (interstitial) atoms and vacancies, among other factors, is a thermally activated process and hence depends on the temperature.

7. Thus, the vessel, with respect to RRE, remembers past excursions in temperature, such as those associated with past shutdowns and restarts, and this factor contributes to the individualization of each plant. This also negates the application of strictly stochastic statistical methods in which the distribution can be defined in terms of a completely random distribution function such as the standard normal distribution. This is important, because in their fluence calculation, PG&E assumes that the neutron flux at the source (the core) is a constant when, in fact, the flux changes with the power level of the reactor and that may induce a memory effect that is not captured by defining operation in terms of EFPYs.

D. The NRCs extension of the deadline for beltline ultrasound inspections is not supported by adequate data

1. In my professional opinion, both PG&E and the NRC Staff have created an unacceptable safety risk by extending the deadline for removing and testing Capsule B a number of times from its originally scheduled removal in 2007 or 2009, to the point that PG&E does not plan to remove the capsule until the fall of 2023 or as late as the spring of 2025. As a result, PG&E has operated Unit 1 for two decades without essential information on the condition of the pressure vessel. And the gap is all the more concerning given the

27 indications of embrittlement in 2003 and further indications that some of the data were erroneous. Instead of postponing the next scheduled withdrawal and testing of a capsule, the Staff should have required PG&E to hasten the removal of Capsule B, and also to test whatever other capsules had been removed, using all available testing protocols, such as tensile (WOL) testing. Using all available protocols is especially important in light of the fact that Capsule B does not contain the limiting weld material that was in Capsules S, Y and V.

2. For several reasons, it is also my professional opinion that PG&E should conduct a UT inspection of beltline welds as soon as possible, preferably in the next refueling outage, rather than postponing it until 2025. First, as previously discussed, the UT inspection is both different and more reliable than the Charpy tests in that it detects and characterizes flaws that potentially could initiate unstable crack growth in the RPV under PTS conditions. Because it detects events that occur after the initial radiation embrittlement phenomenon, it has an independent value. Second, once PG&E had declared the Charpy data from Capsules S, Y, and V showed that Unit 1 was approaching regulatory limits and yet found the data not to be credible, it was incumbent on PG&E to acquire and evaluate as much additional data as possible, not to postpone obtaining it. Finally, PG&E inappropriately relied on reference temperature data from a sister reactor as input to the calculation of through-wall cracking frequency (TWCF). PG&E (2014), Enclosure at 6.

As discussed above, reference temperature data from generic data bases or sister reactors should not have been relied on more than ten years after the 2003 Charpy tests for any purpose. Certainly, they should not be relied on to evade a UT inspection of the Unit 1 reactor vessel. The data is suspect and the reasoning is circular.

E. Alternative testing methods would provide far more accurate results.

1. 10 C.F.R. § 50.51(c)(3) requires licensees to offer information that will improve the accuracy of the RTPTS value significantly. The regulation doesnt apply only to CIT, which obtains one result per sample, and hence yields too few data to be statistically significant for a reasonable confidence level, but I am aware of the newly developed method of nano-indentation that is capable of obtaining many more replicate data than the conventional fracture mechanics methods prescribed by NRC regulations. The nano-indentation technique has been used for many years to assess embrittlement in steels and other alloys as reflected in a change in hardness. Briefly, a sharp point is pressed into a material under a known load and the dimensions of the indentation (width and depth) are measured. Thus, with increasing hardness, the depth and width of the indent become smaller. However, the relationship between hardness and RTNDT and USE still need to be established for this technique to replace the Charpy Impact Test. Nevertheless, I believe that can be done by using an Artificial Neural Network (ANN) to analyze the large body of information on RTNDT and USE vs. degree of embrittlement that is available from PWRs operating within the US and abroad.

2. I note that ASTM185-82 recommends indentation as an optional method for assessing the extent of embrittlement but it appears that too few plants have exercised that option to judge the viability of the method. However, the failed Charpy specimens are archived so

28 that the NRC could require each operator to measure the hardness using a suitable indenter and compile the results with as many independent variables (IVs) as possible.

3. The variables should include indentation width (pw), indentation depth (pd), fluence (f),

temperature of irradiation (Tirr), copper content [Cu], nickel content [Ni], unirradiated yield strength (YS), unirradiated ultimate tensile strength (UTSunirr), reduction of area upon fracture (RoA) and possibly others. The data should then be analyzed using artificial intelligence in the form of an artificial neural network (ANN) as presented in Figure 6. The independent variables would make up the input vector in the ANN as shown in the figure. This is the same ANN that I used to analyze the very large body of data from both the field and the laboratory on IGSCC in sensitized Type 304 SS in developing the CEFM. Shi, Wang, and Macdonald (2015). The net comprised one input layer, one output layer, and three hidden layers, each containing as many neurons as the data contained in each input layer. All of the neurons in any given hidden layer are connected to all of the neurons in the preceding and following layers by interconnections of specific weights recognizing the bias associated with them. Establishment of the weights essentially imbues the net with memory and enables the relationships between the output and input layers to be established. The data collected from both laboratory and field studies are divided randomly into two groups; a training set and an evaluation set.

The first set is used to train the net in a supervised, back propagation manner by incrementally adjusting the weights until the difference between the ANN predicted output and the known outputs satisfies some criterion such as the sum of the squares of that difference being minimal. Typically, this occurs after a few thousand to a few tens of thousands of iterations or about a few seconds of execution time on a laptop computer.

Figure 9: Artificial Neural Network for establishing relationships between the dependent variables (RTNDT and USE) and the vector of the Input Variables (pw, pd, f, Tirr, [Cu], [Ni], YSunirr, UTSunirr, RoA). Note that the neuron sums the values of the inputs from all preceding neurons and then applies a transfer function that determines

29 how the information is passed on to each of the neurons in the following layer with the amount of the information passed being determined by the weight of the connection between the two neurons.

4. It is important to note that no preconceived relationship between the output and the input is employed and the net has no physical theoretical basis. This extraordinarily powerful technique will define those relationships for us, with the result that we do not need to develop a theoretical physical model for the system. Once the ANN is trained and evaluated for accuracy using the evaluation data set, the net can be used to predict RTNDT and USE or some other parameter that measures the state of embrittlement of the RPV steel for any given indentation parameters. Because nano-indentation (or even classical indentation for that matter) requires very little material (< 2 mm2), many sets of parameters can be obtained from each broken Charpy specimen (for example) thereby allowing the statistical basis of the RTNDT and USE to be explored in a manner that is not possible with the Charpy Impact Test method. The indentation method is quick (a few minutes per measurement) so that large databases of RTNDT and USE vs. the IVs can be developed without interfering with reactor operation. Furthermore, the addition of new data to the net represents continual retraining and refinement of the uncovered relationships between the dependent variables (RTNDT and USE) and the IVs. I suggest that this technology be developed and employed in a complementary manner until its advantages over the CIT have been established.

5. Professor Peter Hosemann, the developer of the nano indentation method at UC Berkeley and my fellow faculty in the Department of Nuclear Engineering kindly contributed the following material that describes the method in greater depth that my account given above and outlines some of his work on using it to characterize the radiation embrittlement of RPV steels. Any additions/clarifications other than correcting grammatical errors, such as missing articles, etc. that I have made to Prof. Hosemanns account are identified in italics.

6. In many nuclear applications there is simply not sufficient sample material available to provide a statistically sound and comprehensive dataset assessing a material mechanical property. In most instances, only a limited number of samples can be tested due to limited reactor space or the hazardous nature of the material. Nanoindentation is a technique assessing a materials hardness using an indenter that quantifies the force and the depth as a load is applied. Both force and displacement-controlled tools are available today.

Assessing the force and displacement in-situ allows for a fully instrumentalized hardness measurement. Traditionally, a three-sided pyramid indenter (Berkovich) is used to perform the measurement that is calibrated against fused silica. The Oliver and Pharr method allows one to establish hardness and elastic property values. Other approaches utilize spherical indenters that are not self-similar but have the advantage of generating flow curves more directly.

7. Dynamic measurements (CSM, DMA, etc.) allow one to assess hardness as a function of indentation depth. Of course, hardness by itself is not a measure of yield strength or ductility at all but the properties measured using an instrumented hardness test or

30 nanoindentation allows them to be strongly correlated with these more engineering approaches. The real strength of nanoindentation originates with the fact that no elaborate sample preparation and shaping is required but only a nicely polished surface is needed.

Furthermore, many datapoints can be collected within a matter of minutes and hours on a sample allowing one to assess local microstructures and provide statistics.

8. In recent years, scientists have spent significant effort to correlate and calculate more relevant engineering data from simple nano hardness measurements and utilize the benefits of large data numbers from indentation experiments. Several approaches emerged from these efforts allowing one to quantify yield strength as a function of irradiation conditions. Figure 10 shows one approach originally developed by Hosemann et al. and adopted and modified by Zinkle and others. In this approach, the nano hardness is used to calculate a macro hardness (corrected for pile up) which then in turn is used to calculate yield strength [Figure 10 (a)]. A blind test conducted over different reactor irradiated materials compares tensile test and shear punch test generated data to data obtained from nano hardness. As one can see there is a clear agreement between these very different measurements [Figure 10 (b)] again with the benefit that no elaborate sample preparation is needed while always collecting more than 15 datapoints per sample. Therefore, each datapoint is an average of 15 measured datapoints. The large number of datapoints allows the distribution function to be determined and the appropriate error to be specified (e.g., the standard deviation) with an accuracy that is not possible using Charpy analysis.

31 Figure 10: (a) Roadmap of nano indentation techniques. (b) Correlation between tensile test and shear punch test generated data to data obtained from nano hardness.

9. Of course, neither the yield strength nor the nanoindentation-obtained yield strength can make a direct statement about the strain to failure or embrittlement. However, the correlation investigating the temperature shift obtained by tensile testing with other more conventional methods such as Charpy or fracture toughness allows a comparison to be made. However, elevated temperature nanoindentation experiments are rare and not very common today but will need to be carried out in the future.

10. Other techniques such as spherical indentation have taken a slightly different approach.

There the indentation can generate a direct measurement of yield strength from a single experiment. A direct comparison between different mechanical test techniques was made in the literature (Figure 11)

32 Figure 11: (a) Different micromechanical measurement tools; (b) Yield strength as a function of distance from a weld fusion line; and (c) True fracture stress vs plastic strain for irradiated and unirradiated RPV steel as measured using to micromechanical techniques depicted in (a).

11. Again, the key advantage of performing indentation in addition to other more conventional tests is the fact that one can conduct a near limitless number of measurements on the sample since the material is rather small not needing to cut specific sample geometries.

12. As matters currently stand, PG&E has no credible, plant specific data except for the 2005 UT examination, which PG&E claims (improbably) shows only one indication, to assess the state of embrittlement of the RPV of Diablo Canyon Unit 1 with which to assure the public of the reactors safety. Given this, PG&E should be required to measure the hardness of the fractured Charpy specimens using the indentation method. These measurements should be performed of the actual weld metal, the HAZ, and the plate and be assessed against the unirradiated material. The method of analysis can follow that specified in RG1.99 and the critical hardness may be defined by plotting hardness vs, ARTNDT and extrapolating the plot to the critical value of ARTNDT for the weld dependent upon its orientation.

33 VI.

CONCLUSION AND RECOMMENDATIONS

1.

For the reasons stated above, it is my professional opinion that the continued operation of Diablo Canyon Unit 1 poses an unreasonable risk to public health and safety and the environment.

2. Therefore, I recommend that the NRC Commissioners order the immediate closure of the reactor and that it must remain closed pending the completion of the following measures:

a) Withdrawal and analysis of the contents of Capsule B as well as Capsules C and D (previously withdrawn but not analyzed);

b) Evaluation and analysis of the WOL specimens contained in Capsules B, C and D and the archived capsules; c) Performance of nano indentation studies on the fractured remnants of the Charpy specimens from Capsules S, Y, and V; d) A comprehensive UT inspection of reactor vessel beltline welds; e) publication of the data from the 2015 UT inspection of reactor vessel beltline welds; f) A robust re-evaluation of the credibility of data from Capsules S, Y, and V that fully complies with NRC guidance and scientific principles:

g) Any follow-up steps that may be appropriate for a finding of credibility of the data from Capsules S, Y, and V, including compliance with 10 C.F.R. 50.61a; h) Provision to the NRC, the ACRS, and the general public of all data and analyses that are obtained or performed, and a description of any remedial steps taken by PG&E to address the condition of the Unit 1 reactor pressure vessel; and i) A decision by the NRC Commissioners regarding the safety of continued operation that is informed by the outcome of a proceeding for public participation in the decision-making process.

3. In my professional opinion, nothing short of these steps can provide a reasonable level of assurance that Diablo Canyon Unit 1 is safe to operate - either currently or in a license renewal term.

Under penalty of perjury, I declare that the foregoing facts are true and correct to the best of my knowledge and that the opinions expressed herein are based on my best professional judgment.

Executed in Accord with 10 CFR 2.304(d) by Digby Macdonald Digby Macdonald September 14, 2023

APPENDIX A: Curriculum Vitae

1 DIGBY D. MACDONALD Professor in Residence, Departments of Nuclear Engineering and Materials Science and Engineering University of California at Berkeley 4151 Etcheverry Hall Berkeley, CA 94720 (814) 360-3858, macdonald@berkeley.edu EDUCATIONAL BACKGROUND B.Sc. (1965) and M.Sc. (1966) in Chemistry, University of Auckland (New Zealand);

Ph.D. in Chemistry (1969), University of Calgary (Canada).

PROFESSIONAL EXPERIENCE (past 52 years)

• Professor in Residence, Departments of Nuclear Engineering and Materials Science and Engineering, University of California at Berkeley, 1/2013 - present.

• Distinguished Professor of Materials Science and Engineering, Penn. State Univ.,6/2003 -

12/2012.

• Chair, Metals Program, Penn. State Univ., 6/2001 - 6/2003

• Director, Center for Electrochemical Sci. & Tech., Penn. State Univ., 7/99 - 12/2012.

• Vice President, Physical Sciences Division, SRI International, Menlo Park, CA, 1/98 -

7/99

• Director, Center for Advanced Materials, Penn. State Univ., 7/91-3/2000

• Professor, Materials Science and Engineering, Penn. State Univ., 7/91 - 6/03.

• Deputy Director, Physical Sciences Division, SRI International, Menlo Park, CA, 4/87 -

7/91

• Laboratory Director, Mat. Research Lab., SRI International, Menlo Park, CA, 4/87 - 7/91

• Laboratory Director, Chemistry Laboratory, SRI International, Menlo Park, CA, 3/84 -

4/87

• Director and Professor, Fontana Corrosion Center, Ohio State University, 3/79 - 3/84

• Sr. Metallurgist, SRI International, Menlo Park, CA, 3/77 - 3/79.

• Sr. Research Associate, Alberta Research Ltd/University of Calgary, Canada, 3/75 - 3/77.

• Lecturer in Chemistry, Victoria University of Wellington, New Zealand, 4/72 - 3/75.

• Assist. Research Officer, Whiteshell Nuclear Research Establishment, Atomic Energy of Canada Ltd., Pinawa, Manitoba, Canada, 9/69 - 4/72.

CONSULTING ACTIVITIES (Partial list for the last twenty years).

OLI Systems Electric Power Research Institute SRI International Stone & Webster Engineering Co.

Canadian Auto Preservation, Inc.

Numerous oil and gas companies.

SSM, Sweden.

2 PATENTS

1. D. D. Macdonald and A. C. Scott, Pressure Balanced External Reference Electrode Assembly and Method, US Patent 4,273,637 (1981).

2. D. D. Macdonald, Apparatus for Measuring the pH of a Liquid, US Patent 4,406,766 (1983).

3. S. C. Narang and D. D. Macdonald, Novel Solid Polymer Electrolytes, US Patent 5,061,581 (1991).

4. S. Hettiarachchi, S. C. Narang, and D. D. Macdonald, Synergistic Corrosion Inhibitors Based on Substituted Pyridinium Compounds, US Patent 5,132,093 (1992).

5. S. Hettiarachchi, S. C. Narang, and D. D. Macdonald, Reference Electrode Assembly and Process for Constructing, US Patent, 5,238,553 (1993).

6. D. D. Macdonald, et al, Conducting Polymer for Lithium/Aqueous Syst., US Prov. Pat.

60/119,360 (1998).

7. D. D. Macdonald, et al, Polyphosphazenes as Proton Conducting Membranes, US Pat. Appl.

09/590,985 (1999).

8. D. D. Macdonald, et al, Impedance/Artificial Neural Network Method, US Prov. Pat.

60/241,871 (1999)

9. D. D. Macdonald, Electrochemical Conditioning of Wine, US Prov. Pat. 60/295,080 (2001).

10. D. D. Macdonald, et.al., Silicon Air Battery, Int. Patent WO2011/061728A1, May 26, 2011.

11. D. D. Macdonald, et.al., Silicon Air Battery, US Patent, 8,835,060 B2, Sept. 16, 2014.

RELEVANT PUBLICATIONS (from a total of 1000).

1. D. D. Macdonald and G. R. Engelhardt, Predictive Modeling of Corrosion. In:

Richardson J A et al. (eds.), Shriers Corrosion, 2, 1630-1679 (2010). Amsterdam:

Elsevier.

2. J. Qiu, A. Wu, J. Yao, Y. Xu, Y. Li, R. Scarlat, D.D. Macdonald, Kinetic study of hydrogen transport in graphite under molten fluoride salt environment. Electrochim. Acta, 2020, 136459 (2020).

3. J. Yao, J. Qiu, F. Carotti, R. Scarlat, D.D. Macdonald, Kinetic study of the hydrogen charging reaction on the graphite in aqueous solution and in room temperature ionic liquid (RTIL), Electrochimica Acta, 330, 135291 (2000).

4. D Kovalov, B Fekete, G R Engelhardt, D D Macdonald, Prediction of corrosion fatigue crack growth rate in alloys. Part I: General corrosion fatigue model for aero-space aluminum alloys, Corrosion Science, 141, 22-29 (2018).

5. D Kovalov, B Fekete, G. R Engelhardt, D. D Macdonald, Prediction of Corrosion Fatigue Crack Growth Rate in Alloys. Part II: Effect of Electrochemical Potential, NaCl Concentration, and Temperature on Crack Propagation in AA2024-T351, Corrosion Science, 152. 130-139 (2019).

6. P. C. Lu, D. D. Macdonald, M. Urquidi-Macdonald and T. K. Yeh. Theoretical Estimation of Crack Growth Rates in Type 304 Stainless Steel in BWR Coolant Environments.

Corrosion, 52(10), 768-785 (1996).

7. G. R. Engelhardt, M. Urquidi-Macdonald, and D. D. Macdonald. A Simplified Method for Estimating Corrosion Cavity Growth Rates. Corros. Sci., 39(3), 419-441 (1997).

3

8. S.-K. Lee, P. Lv, and D. D. Macdonald, Customization of the CEFM for Predicting Stress Corrosion Cracking in Lightly Sensitized Al-Mg alloys in Marine Applications, J. Solid State Electrochem., 17(8), 2319-2332 (2013).

9. J Shi, J Wang, D D Macdonald, Prediction of crack growth rate in Type 304 stainless steel using artificial neural networks and the coupled environment fracture model, Corrosion Science, 89, 69-80 (2014).

10. J Shi, J Wang, D D Macdonald, Prediction of primary water stress corrosion crack growth rates in Alloy 600 using artificial neural networks, Corrosion Science, 92, 217-227 (2015).

11. G. R. Engelhardt and D.D. Macdonald. Modeling the Crack Propagation Rate for Corrosion Fatigue at High Frequency of Applied Stress, Corros. Sci., 52(4), 1115-1122 (2010).

12. M. P. Manahan, D. D. Macdonald, and A. J. Peterson, Jr. Determination of the Fate of the Current in the Stress-Corrosion Cracking of Sensitized Type 304SS in High Temperature Aqueous Systems. Corros. Sci., 37(1), 189-208 (1995).

13. G. R. Engelhardt, and D. D. Macdonald, Deterministic Prediction of Pit Depth Distribution, Corrosion, 54, 469-479 (1998).

14. D. D. Macdonald, M. Al-Rafaie and G. R. Engelhardt, New Rate Laws for the Growth and Reduction of Passive Films, J. Electrochem. Soc., 148(9), B343 - B347 (2001).

15. D. D. Macdonald, Stress Corrosion Cracking in Reactor Coolant Circuits - An Electrochemists Viewpoint, Power Plant Chemistry, 6, 731-747 (2004).

16. L. G. Million, A. Sun, D. D. Macdonald, and D. A. Jones, General Corrosion of Alloy 22:

Experimental Determination of Model Parameters from Electrochemical Impedance Spectroscopy Data, Met. Trans. A, 36A, 1129 (2005).

17. D. D. Macdonald, Internal/External Environment Coupling in Stress Corrosion Cracking, J. Corr. Sci. Eng., 6, Paper C065 (2005).

18. D. D. Macdonald, On the Existence of our Metals-Based Civilization: I. Phase Space Analysis, J. Electrochem. Soc., 153(7), B213 (2006).

19. D. D. Macdonald and G. R. Engelhardt, The Point Defect Model for Bi-Layer Passive Films, ECS Trans, 28(24), 123 - 144 (2010).

20. D. Kong, A. Xu, C. Dong, F. Mao, K. Xiao, X. Li, D. D. Macdonald, Electrochemical investigation and ab initio computation of passive film properties on copper in anaerobic sulphide solutions, Corros. Sci., 116, 34-43 (2017).

21. S K Lee, D D Macdonald, Theoretical aspects of stress corrosion cracking of Alloy 22, J.

Nucl. Mat., 503, 124-139 (2018).

22. E. Huttunen-Saarivirta, E. Ghanbari, F. Mao, P. Rajala, L. Carpén, and D. D. Macdonald, (2019). Erratum: Kinetic properties of the passive film on copper in the presence of sulfate-reducing bacteria (Journal of the Electrochemical Society (2018) 165 (C450)

23. J. Qiu, D. D. Macdonald, Y. Xu, L. Sun, General corrosion of carbon steel in a synthetic concrete pore solution, Mat. Corros., 72(1-2), 107-119 (2021).

24. Z Ghelichkhah, FK Dehkharghani, S Sharifi-Asl, IB Obot, DD Macdonald, K Farhadi, M Avestan, A Petrossians, The inhibition of type 304LSS general corrosion in hydrochloric acid by the New Fuchsin compound, Corros. Sci. 178, 109072 (2021).

25. Y Zhu, DD Macdonald, J Yang, J Qiu, GR Engelhardt, The Corrosion of Carbon Steel in Concrete. Part II: Literature Survey and Analysis of Existing Data on Chloride Threshold, Corros. Sci., 109439 (2021).

4

26. K Liivand, M Kazemi, P Walke, V Mikli, M Uibu, DD Macdonald, I Kruusenberg, Spent Li-Ion Battery Graphite Turned Into Valuable and Active Catalyst for Electrochemical Oxygen Reduction, ChemSusChem 14 (4), 1103-1111 (2021).

27. F Carotti, E Liu, DD Macdonald, RO Scarlat, An electrochemical study of hydrogen in molten 2LiF-BeF2 (FLiBe) with addition of LiH, Electrochim. Acta, 367, 137114 (2021).

PROFESSIONAL ASSOCIATIONS AND HONORS Research Award, College of Engineering, Ohio State University, 1983.

Selector of the Kuwait Prize for Applied Sciences, 1985.

The 1991 Carl Wagner Memorial Award from The Electrochemical Society.

The 1992 Willis Rodney Whitney Award from The National Association of Corrosion Engineers.

Chair, Gordon Research Conference on Corrosion, New Hampshire, 1992.

W.B. Lewis Memorial Lecture by Atomic Energy of Canada, Ltd., 1993, in recognition of [his] contributions to the development of nuclear power in the service of mankind.

Elected Fellow, NACE-International, 1994.

Member, USAF Scientific Advisory Board, Protocol Rank: DE-4 (Lieutenant General equivalent), 1993-1997 Elected Fellow, The Electrochemical Society, 1995.

Elected Fellow, Royal Society of Canada, 1996. (National Academy of Canada).

Wilson Research Award, College of Earth and Minerals Sciences, Pennsylvania State University, 1996.

Elected Fellow, Royal Society of New Zealand, 1997. (National Academy of New Zealand).

H. H. Uhlig Award, Electrochemical Society, 2001.

U. R. Evans Award, British Corrosion Institute, 2003.

Elected Fellow, Institute of Corrosion (UK), 2003.

Appointed Adjunct Professor, Massey University, New Zealand, 2003.

Appointed Adjunct Professor, University of Nevada at Reno, 2003.

Elected Fellow, World Innovation Foundation, 2004.

Elected Fellow, ASM International, 2005.

Elected Fellow, International Society of Electrochemistry, 2006.

Khwarizmi International Award Laureate in Fundamental Science, Feb. 2007.

Trustee, ASM International, 2007-2010.

Appointed SABIC Visiting Chair Professor, King Fahd University of Petroleum and Minerals, Dhahran, Saudi Arabia, 2010.

Recipient, Lee Hsun Research Award, Chinese Academy of Sciences, China, 2010.

Inducted Doctuer Honoris Causa by INSA-Lyon, Lyon, France, 2011.

Nominated for the 2011 Nobel Prize in Chemistry for work on passivity.

Awarded the Faraday Memorial Trust Gold Medal, 2012.

Awarded the Gibbs Award in Thermodynamics by IAPWS, 2013 Awarded Frumkin Medal, ISE, 2014.

Awarded the OLIN Palladium Medal by the Electrochemical Society, 2015.

Received the Ad Augusta Award from Auckland Grammar School, 2016.

5 Plenary Lecturer, Corrosion2019, Nashville, TN, 2019.

Plenary Lecturer, Mexican Electrochemical Society, 2019.

Elected Member of the EU Academy of Science, 2019.

FLOGEN Fray International Sustainability Award for distinguished work in corrosion science.

Signed. Digby D. Macdonald.

September 13,2023.

APPENDIX B: Reference List

APPENDIX B: REFERENCE DOCUMENTS Bogearts, W; Macdonald, D.D.; Zheng, J.H.; and Jovanovic, A.S., Feb. 2022. Hydrogen and NPP Life Management: Doel 3 and Tihange 2, Technical Report, KU Leuven, Belgium.

Bogaerts, W.F.,; Zheng, J.H.; Jovanovic A.S.; and Macdonald, D.D., 2015. "Hydrogen-induced Damage in PWR Reactor Pressure Vessels", CORROSION 2015, Research in Progress Symposium, Corrosion in Energy Systems, Dallas, March 15-19, 2015.

Macdonald, D.D., 2023. The Role of Determinism in the Prediction of Corrosion Damage.,

Corros. Mater. Degrad. 2023, 4, 212-273. https:// doi.org/10.3390/cmd4020013.

PG&E 1992. Letter DCL-92-072 to NRC re: Diablo Canyon Supplemental Surveillance Program, Enclosure at 4 and Table (Mar. 31, 1992) (PG&E Letter DCL-92-072) (ADAMS Accession No. ML16341G504).

PG&E 2003. Letter DCL-03-052 from David H. Oatley to NRC re: Diablo Canyon Reactor Vessel Material Surveillance Program Capsule V Technical Report (May 13, 2003) (ADAMS Accession No. ML031400334).

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