ML23257A302

San Luis Obispo Mothers for Peace and Friends of the Earth Corrected Request for Hearing on NRC Staff Decision Effectively Amending Diablo Canyon 1 License, Request for Emergency Order Requiring Shutdown, and Errata

ML23257A302
Person / Time
Site:	Diablo Canyon
Issue date:	09/14/2023
From:	Curran D, Templeton H Friends of the Earth, Harmon, Curran, Harmon, Curran, Spielberg & Eisenberg, LLP, San Luis Obispo Mothers for Peace (SLOMFP)
To:	NRC/OCM
SECY RAS
References
RAS 56783, 50-275
Download: ML23257A302 (0)

Text

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 ERRATA TO REQUEST FOR HEARING ETC. BY SAN LUIS OBISPO MOTHERS FOR PEACE AND FRIENDS OF THE EARTH Petitioners respectfully submit the following errata to their REQUEST TO THE NRC COMMISSIONERS BY SAN LUIS OBISPO MOTHERS FOR PEACE AND FRIENDS OF THE EARTH ETC. (Sept. 14, 2014):

Page Line Correction 1 11 Delete or exceed 9 note 23 Change criterial to criteria 27 6 Insert can after Commission 28 3 Change tensile stress/strain to to wedge opening loading (WOL)

Respectfully submitted,

__/signed electronically by/___

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

1725 DeSales Street N.W., Suite 500 Washington, D.C. 20036 240-393-9285 dcurran@harmoncurran.com September 14, 2023

CORRECTED SEPT. 14, 2023 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 REQUEST TO THE NRC COMMISSIONERS BY SAN LUIS OBISPO MOTHERS FOR PEACE AND FRIENDS OF THE EARTH FOR A HEARING ON NRC STAFF DECISION EFFECTIVELY AMENDING DIABLO CANYON UNIT 1 OPERATING LICENSE TO EXTEND THE SCHEDULE FOR SURVEILLANCE OF THE UNIT 1 PRESSURE VESSEL AND REQUEST FOR EMERGENCY ORDER REQUIRING IMMEDIATE SHUTDOWN OF UNIT 1 PENDING COMPLETION OF TESTS AND INSPECTIONS OF PRESSURE VESSEL, PUBLIC DISCLOSURE OF RESULTS, PUBLIC HEARING, AND DETERMINATION BY THE COMMISSION THAT UNIT 1 CAN SAFELY RESUME OPERATION Submitted 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 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 September 14, 2023

Table of Contents I. INTRODUCTION AND

SUMMARY

.1 II. DESCRIPTION OF PETITIONERS.4 III. BACKGROUND6 A. Role and Importance of the Reactor Vessel..6 B. Pressurized Thermal Shock7 C. Regulations Governing Safety of the Reactor Vessel 8 D. History of Diablo Canyon Unit 1 Reactor Vessel .9

1. Licensing of Unit 110

a. Reactor vessel surveillance program.10

b. Supplemental surveillance program...11

2. License amendment to recover thirteen-year construction period.13

3. Withdrawal and testing of Capsule V14

4. License amendment to recover three-year low-power testing period.15

5. Capsule B withdrawal re-purposed to serve license renewal at PG&Es discretion.17 E. PETITIONERS ARE ENTITLED TO A HEARING BECAUSE THE 7/20/23 EXTENSION ORDER EFFECTIVELY AMENDED PG&ES OPERATING LICENSE FOR UNIT 1 20 IV. CONTENTION 1.....22 V. CONTENTION 2 ....24 VI. REQUEST FOR SHUTDOWN ORDER AND REMEDIAL MEASURES..26 A. Exercise of Commissions Discretionary Supervisory Authority is Warranted...26 B. Unit 1 Must be Shut Down to Protect Public Health and Safety and Should Not be Reopened Until PG&E Has Conducted Adequate Tests and Inspections, Disclosed Their Data and Results, and Subjected Them to Expert Review and a Public Hearing...27 VII. CONCLUSION............................................................................................................29 ii

Table of Authorities Judicial Decisions Citizens Awareness Network v. NRC, 59 F.3d 284, 295 1st Cir. 1995).21 In re Three Mile Island Alert, 771 F.2d 720 (3d Cir. 1985)..21 Massachusetts v. NRC, 878 F.2d 1516 (1st Cir. 1989)..21 San Luis Obispo Mothers for Peace v. NRC, 751 F.2d 1287 (D.C. Cir. 1985)..21 Statutes Atomic Energy Act Section 189a, 42 U.S.C. § 2239(a).1, 20, 21, 29 National Environmental Policy Act.................................................................................24, 25, 29 Regulations 10 C.F.R. § 2.309.1 10 C.F.R. § 2.309(b)(4)(i)3 10 C.F.R. § 50.55a....10, 22 10 C.F.R. § 50.61...22 10 C.F.R. § 50.61(b)1, 9 10 C.F.R. § 50.61(b)(2)9 10 C.F.R. § 50.61(c)(2)9 10 C.F.R. Part 50, Appendix G14, 22 10 C.F.R. Part 50, Appendix G, § IV.A.19 10 C.F.R. Part 50, Appendix H14, 16-17, 22 10 C.F.R. Part 50, Appendix H, Section III.B.3 10 C.F.R. § 51.20...24 10 C.F.R. § 51.30.24, 25 iii

10 C.F.R. § 51.70.25 Administrative Decisions Cleveland Electric Illuminating Co. (Perry Nuclear Power Plant, Unit 1),

CLI-96-13, 44 N.R.C. 315 (1996).20, 21 Pacific Gas and Electric Co. (Diablo Canyon Nuclear Power Plant, Units 1 and 2), LBP-92-27, 36 N.R.C. 196 (1992)10 Yankee Atomic Electric Co. (Yankee Rowe Nuclear Power Station),

CLI-91-11, 34 N.R.C. 3 (1991)..3, 4, 8, 26 Federal Register Notices Final Rule, Analysis of Potential Pressurized Thermal Shock Events, 50 Fed. Reg. 29,937 (July 23, 1985)9 Final Rule, Fracture Toughness Requirements for Light Water Reactor Pressure Vessels, 60 Fed. Reg. 65,456 (Dec. 19, 1995).3, 8 Notice of Exemption Issuance, 88 Fed. Reg. 14,395 (March 8, 2023).27 Notice of license amendment issuance, 71 Fed. Reg. 46,945 (Aug. 15, 2006).17 Proposed No Significant Hazards Consideration Determination for Diablo Canyon license amendment, 57 Fed. Reg. 32,575 (July 22, 1992)14 Miscellaneous ASTM E 1829, 11, 15, 16, 19 NUREG-1801, GALL Report (date unknown).18 iv

I. INTRODUCTION AND

SUMMARY

Pursuant to 10 C.F.R. § 2.309 and Section 189a of the Atomic Energy Act, San Luis Obispo Mothers for Peace (SLOMFP) and Friends of the Earth (FoE) (hereinafter Petitioners) request the Commissioners of the U.S. Nuclear Regulatory Commission (NRC or Commission) to convene a hearing on a license amendment effectively issued by the NRC staff to Pacific Gas and Electric Co. (PG&E) by letter of July 20, 2003, extending the schedule for conducting surveillance of the Diablo Canyon Unit 1 pressure vessel until 2025.1 As demonstrated in the attached supporting expert declaration of Dr. Digby Macdonald, the extension is unjustified and poses an unreasonable risk to public health and safety in light of data from 2003 tests of surveillance capsules installed in the Unit 1 pressure vessel indicating that Unit 1 would approach embrittlement criteria in 10 C.F.R. § 50.61(b) by the end of the initial 1

Letter from Jennifer L Dixon-Herrity, NRC to Paula Gerfen, PG&E re: Diablo Canyon Nuclear Power Plant, Unit 1 - Revision to the Reactor Vessel Material Surveillance Capsule Withdrawal Schedule (EPID L-2023-LLL-0012) (NRC 7/20/23 Extension Decision) (ADAMS Accession No. ML120330497).

The 7/20/23 Extension Decision approves a schedule under which PG&E would withdraw Capsule B from the Unit 1 pressure vessel either during the upcoming 24 th refueling outage (1R24) in October 2023 or the 25th refueling outage in the spring of 2025 (1R25). Id., enclosed Safety Evaluation at 4-5. See also PG&E Letter DCL-23-038 from Paula Gerfen to NRC re:

Docket No. 50-275, OL-DPR-80, Diablo Canyon Unit 1, Revision to the Unit 1 Reactor Vessel Material Surveillance Program Withdrawal Schedule at 2 and Table 5.2-22 (May 15, 2023)

(PG&E Letter DCL-23-038) (ADAMS Accession No. ML23135A217).

operating license term.2 PG&E incorrectly discarded these data as not credible.3 In addition, Dr. Macdonalds own separate and independent analysis of a different set of 2003 surveillance data, deemed credible by PG&E, shows that the Unit 1 pressure vessel could reach an unacceptable level of embrittlement relatively early in the license renewal term (43.8 effective full power years (EFPY) with an estimated uncertainty of +/- 10 EFPY).4 Taking into account the level of uncertainty of +/- 10 EFPY, an unacceptable degree of embrittlement could be reached as early as 33.8 EFPY, or late 2023.5 These indications of embrittlement should have caused PG&E to seek additional data for an adequate understanding of the condition of the pressure vessel. Instead, over the past twenty years, PG&E has repeatedly postponed additional surveillance and testing of the pressure vessel such that withdrawal and testing of Capsule B coupons is now delayed from 2009 to potentially 2025 and ultrasound inspection of reactor beltline welds is now delayed from 2015 to 2025.6 As stated by Dr. Macdonald, PG&Es decades of neglect, coupled with serious 2

Attachment 1, Declaration of Digby Macdonald, Ph.D in Support of Hearing Request and Request for Emergency Action, § V.A.1 ¶ 1 (September 14, 2023) (Macdonald Declaration)

(quoting PG&E 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) (PG&E Letter DCL-03-052) (ADAMS Accession No. ML14230A618)).

Dr. Macdonald is Professor in Residence at the University of California at Berkeley in the Departments of Nuclear Engineering and Materials Science and Engineering and an expert in electrochemistry, thermodynamics and corrosion science, including corrosion cracking and fatigue in nuclear reactor materials. He has been nominated for a Nobel Prize for his work on the phenomenon of passivity in metals and was recently nominated for the prestigious Enrico Fermi Award for introducing electrochemistry into describing corrosion phenomena in the primary coolant system of light water reactors.

3 Id., § V.A.1.

4 Id., § V.A.2.

5 See PG&E Letter DCL-23-038, Table 4, which states that IR24 (October 2024) will occur at 33.58 EFPY and IR25 (spring 2025) will occur at 34.97 EFPY.

6 Macdonald Declaration, § V.D.

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indications of embrittlement, render Unit 1 unsafe to operate.7 Petitioners seek a hearing on the serious safety and regulatory issues raised by PG&Es and the Staffs decades of neglect.8 The safety concerns raised by Dr. Macdonald and by this Petition are extremely grave, given the status of the reactor vessel as perhaps the most important single component in the reactor coolant system.9 As the receptacle that maintains cooling water on the highly radioactive core without any redundant backup, the pressure vessel must be protected against the risk of fracture and failure, which could lead to core melt and catastrophic consequences. The risk is all the greater because Diablo Canyon is located in a high-seismicity zone.10 And the safety and regulatory issues raised by Dr. Macdonald go to a comprehensive failure by PG&E and the Staff, on multiple fronts, to monitor and respond to the development of embrittlement in the Unit 1 vessel.

Accordingly, in addition to demanding the hearing to which they are entitled, Petitioners request the Commissioners to exercise their discretionary supervisory jurisdiction to order the immediate closure of Diablo Canyon pending the completion of a series of remedial actions.11 7

Macdonald Declaration, § III, ¶ 11; § VI.

8 Pursuant to 10 C.F.R. 2.309(b)(4)(ii), a hearing request must be submitted not later than the latest of . . . [s]ixty (60) days after the requestor receives actual notice of a pending application, but not more than sixty (60) days after agency action on the application. This hearing request is timely because it is being submitted within 60 days of receiving notice of the NRCs 7/20/23 Extension Order.

9 Final Rule, Fracture Toughness Requirements for Light Water Reactor Pressure Vessels, 60 Fed. Reg. 65,456, 65,457 (Dec. 19, 1995) (RPV Rule). See also Macdonald Declaration,

§ IV.A.

10 Macdonald Declaration, § IV.

11 As discussed in Section VII.A below, these circumstances pose the safety and regulatory significance previously recognized by the Commissioners as warranting their supervisory involvement. See Yankee Atomic Electric Co. (Yankee Rowe Nuclear Power Station), CLI 11, 34 N.R.C. 3, 12 (1991) (Yankee Rowe) (exercising supervisory review over safety and regulatory issues relating to the condition of the Yankee Rowe pressure vessel).

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These actions include comprehensive testing and inspection of the Unit 1 reactor vessel, including removal and testing of all coupons in Capsule B and other capsules that PG&E has removed since 2003; a comprehensive ultrasound inspection of the reactor beltline welds; and nano-indentation tests as advised by Dr. Macdonald in Section V.E of his declaration. In addition, all test results should be provided to the NRC, the Advisory Committee on Reactor Safeguards, and the public; and finally, a public hearing should be held before Unit 1 is allowed to resume operation.

Due to the gravity of the safety and environmental risks presented by PG&Es and the Staffs failure to provide adequate care or oversight of the Unit 1 pressure vessel, Petitioners seek expedited consideration of their claims on an emergency basis. Petitioners also note that prompt consideration is warranted by the fact that PG&E is scheduled to begin a maintenance outage next month in October. Using a scheduled shutdown to address significant safety issues regarding the pressure vessel, and maintaining the shutdown until the issues are resolved, is consistent with the approach taken by the Commissioners in the Yankee Rowe proceeding, see 34 N.R.C. at 17-19.

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 the safety of operating Unit 1 is not jeopardized by a delay in PG&Es schedule for removing and testing samples from the Unit 1 pressure vessel.

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 4

and social justice, and renewable energy. SLOMFP has participated in NRC licensing cases involving the Diablo Canyon reactors since 1973.

FoE is a tax exempt, nonprofit environmental advocacy organization dedicated to improving the environment and creating a more healthy and just world. 12 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 244,600 members in all 50 states and the District of Columbia, approximately 35,500 of whom are in California. In addition to formal members, FoE has more than 6.6 million online activist supporters across the country. FoE also has office space in Berkeley, California.

Together, SLOMFP and FoE have many members who live, work, and own property within 50 miles of the Diablo Canyon reactors. Their health and safety, and the health of their environment, could be catastrophically damaged by an accident at the Diablo Canyon reactors.

They are concerned that the extension of PG&Es schedule for removing and testing the Capsule B samples from the Unit 1 reactor vessel will deprive PG&E and the NRC of information that is necessary to determine whether Unit 1 can be operated safely. They are also concerned that PG&E has failed to collect any data on the condition of the Unit 1 pressure vessel for the past twenty years. Therefore, as stated in the attached declarations of SLOMFP and FoE members Kaoru Hisasue, Lucy Jane Swanson, and Jill ZamEk, they have authorized SLOMFP to request a hearing on the 7/20/23 Extension Decision, an order by the Commissioners to close Unit 1, and a range of remedial actions to ensure that Unit 1 will be not be allowed to re-open 12 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.

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without a comprehensive set of tests and inspections of its condition that is subject to full transparency and a public hearing. 13 III. BACKGROUND A. Role and Importance of the Reactor Vessel 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. 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 is heated to approximately 500 oF by the water entering the vessel.14 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 kept on the core at all times to prevent the core from overheating and possibly melting down even during shutdown because of the decay heat from the spontaneous decay of unstable isotopes. 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 lethal levels radiation outside the plant, as occurred at Chernobyl, for example.15 13 See Attachment 2A, Declaration of Kaoru Hisasue (Sept. 7, 2023); Attachment 2B, Declaration of Lucy Jane Swanson (Sept. 9, 2023); and Attachment 2C, Declaration of Jill ZamEk (Sept. 8, 2023).

14 Macdonald Declaration., § IV.A.

15 Id.

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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 that uncovered the reactor core, a nuclear meltdown may occur. 16 B. Pressurized thermal shock 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, cooling water is pumped 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 the reactor 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.17 Over the course of a pressurized water reactors operating life, the steel plates and welding materials used in fabricating the pressure vessel become increasingly embrittled or weakened by intense neutron radiation from the core. As the Commission has described the phenomenon:

The fracture resistance of reactor vessel material is initially very high, and thus PTS events are generally not expected to cause vessel failure. However, the fracture resistance of the vessel decreases over the life of the vessel as it is exposed to fast neutron radiation from the core of the reactor. The rate of decrease is dependent on the chemical composition of the vessel wall and weld materials. If the fracture resistance of the vessel is reduced sufficiently by neutron radiation, severe PTs events could cause small flaws that might exist near the inner surface of the vessel to propagate through the wall, thereby 16 Id.

17 Macdonald Declaration., § IV.A.

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threatening the integrity of the vessel, and ultimately the capability of the core cooling systems to cool the fuel in the vessel. 18 The range of temperatures at which the steel changes from brittle to ductile is called the reference temperature for nil ductility transition or RT NDT. In a new vessel, the RTNDT is in the range of 0 to 40OF. However, as the vessel materials are bombarded by high energy (>1 Mev) neutrons during the life of the plant, the RTNDT gradually increases. Thus, the safety margin between the temperature at which the vessel exhibits brittle characteristics, and the temperature to which the vessel will be cooled in the event of an accident, decreases.

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 RT NDT.19 C. Regulations Governing the Safety of the Reactor Vessel As the NRC has recognized, given the singular importance of a nuclear reactors pressure vessel, [m]aintaining the structural integrity of the reactor pressure vessel . . . is a critical concern related to the safe operation of nuclear power plants. 20 The concern is critical not only for the key role played by the reactor vessel in cooling the core, but also for the fact that there is no way to back up the reactor vessel. Unlike many nuclear power plant safety systems, which are designed according to the principle of defense-in-depth to have a redundant, robust and independent double that will function in the event the first system fails, there is only one pressure vessel. Because there is no backup safety system to protect the public in the event of pressure 18 Yankee Rowe, 34 N.R.C. at 8.

19 Macdonald Declaration., § IV.A.

20 RVP Rule, 60 Fed. Reg. at 65,456.

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vessel failure, the Commissions regulations establish design and performance standards that are intended to assure for each plant that the probability of pressure vessel rupture is extremely low.21 NRC regulations in 10 C.F.R. Part 50, Appendix G, § IV.A.1 require all reactor vessel beltline materials to have a Charpy upper-shelf energy (USE) of no less than 75 ft-lb initially and 50 ft-lb throughout the life of the plant. And 10 C.F.R. § 50.61(b)(2) establishes a PTS screening criterion of 270oF for all plates, forgings, and axial weld materials and 300 oF for circumferential weld materials. Requirements for PTS surveillance programs are found in 10 C.F.R. Part 50, Appendix H and 10 C.F.R. § 50.61. Pursuant to 10 C.F.R. § 50.61(c)(2),

evaluations of compliance with 10 C.F.R. § 50.61(b)(2) must include consideration of plant-specific information. The surveillance program must include designation of appropriate locations for surveillance specimen capsules (Appendix H,Section III.B.2) and an NRC-approved withdrawal and testing schedule (id.,Section III.B.3). Surveillance capsules must also contain coupons to measure tensile stress/strain, which are indicative of embrittlement. 22 In order to obtain plant-specific information, the regulations require licensees to conduct reactor-specific surveillance in conformance with the relevant industry guidance of the American Society for Testing of Materials, ASTM E 182.23 21 Final Rule, Analysis of Potential Pressurized Thermal Shock Events, 50 Fed. Reg. 29,937, 29,941 (July 23, 1985).

22 Macdonald Declaration., § IV.B.

23 Licensees must use the version of ASTME E 182 that was in effect at the time the surveillance program was adopted, but may be changed to a later standard.10 C.F.R. § 50.61(b). ASTM E 182 provides licensees with the criterial for determining both the minimum number of surveillance capsules that need to be installed within the reactor vessel at the start of the plants life, and when in the plants life - measured in effective full-power years - a capsule should be withdrawn for evaluation. Appendix H,Section III.B.1.

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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, industry codes incorporated in 10 C.F.R. § 50.55a require that every ten years, licensees must conduct ultrasound testing (UT) inspections of the most vulnerable parts of the reactor vessel, the welds around the beltline, to examine for flaws and cracks.24 D. History of Diablo Canyon Unit 1 Reactor Vessel

1. Licensing of Unit 1 The NRC originally licensed the Diablo Canyon reactors to operate for forty years beyond the issuance dates of their construction permits. 25 Unit 1, which received a construction permit in 1968, was licensed to operate until April 23, 2008; and Unit 2, which received a construction permit in 1970, was licensed to operate until December 9, 2010. 26

a. Reactor vessel surveillance program In the 1970s, while construction was underway, PG&E established separate reactor vessel surveillance programs for the operating license terms Units 1 and 2. The Unit 1 surveillance program consisted of three Type II capsules - Capsules S, Y, and V -- which contained the limiting beltline weld metal, limiting shell plate, and weld heat affected zone (HAZ) from an 24 Macdonald Declaration., § IV.B.

25 See Letter from Gregory M. Rueger, PG&E to NRC re: License Amendment Request 92-04 40-Year Operating License Application (July 9, 1992) (ADAMS Accession No. ML17083C429)

(Rueger Letter).

26 Id. See also Pacific Gas and Electric Co. (Diablo Canyon Nuclear Power Plant, Units 1 and 2),

LBP-92-27, 36 N.R.C. 196, 197 (1992).

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intermediate shell plate.27 PG&E subsequently noted that three Type II capsules had not been enough to satisfy the then-applicable industry standard, ASTM E 182-70, which required five capsules; but nevertheless, the NRC Staff had approved the program. 28

b. Supplemental surveillance program In 1992, PG&E applied to supplement the Unit 1 surveillance program by adding Capsules A, B, C, and D.29 While they did not include the Type II constituents, the new capsules contained the intermediate shell plate 4107-1, which is the limiting base metal at 48 EFPY. 30 The purpose of the supplemental surveillance program was to provide sufficient embrittlement data on the limiting materials to permit effective management of vessel embrittlement during the entire operating life of the vessel. 31 The supplemental surveillance program also had three goals of providing embrittlement data for 48 EFPY or 60 years of operation (i.e., supporting a single license renewal term), providing a standby capsule that could be held in reserve for future use, and providing the necessary data to demonstrate the effects of annealing, should it be needed in the future. 32 To carry out the purpose and goals, PG&E stated that the four capsules would be inserted at EOC [end of cycle] 5 and tested according to the following schedule:

27 This description was provided by PG&E in 1992, when it sought to supplement the program.

PG&E Letter DCL-92-072, Enclosure at 1 and Table 4. While the surveillance program also included other capsules, they were not Type II, i.e., they did not contain the limiting weld metal, base metal, and HAZ specimens that were required by the applicable ASTM standard, ASTME 185-73. Id.

28 Id, Enclosure at 1 and Table 4.

29 PG&E Letter DCL-92-072.

30 Id, Enclosure at 3.

31 Id, Enclosure at 2.

32 Id.

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Capsule B will be tested at approximately 19.2 EFPY 33 after it has accumulated the fluence equivalent to the vessel inside surface at 48 EFPY; Capsule A will remain in the vessel throughout the vessel lifetime as a standby capsule.

Capsule C will be tested at approximately 14.8 EFPY after it has accumulated the fluence equivalent to the vessel inside surface at 32 EFPY; and Capsule D will be removed from the vessel at approximately 14.8 EFPY after it has accumulated the fluence equivalent to the vessel inside surface at 32 EPFY and will be annealed and reinserted into the vessel and removed at approximately 19.2 EFPY after it has accumulated the fluence equivalent to the vessel inside surface at 32 EPFY. 34 In a 1992 Safety Evaluation, the NRC Staff approved the supplemental surveillance program, including the schedule for withdrawal of Capsules B, C, and D and the standby status of Capsule A.35 The Safety Evaluations conclusions included a finding that the changes proposed by PG&E will provide additional data on the limiting reactor vessel materials.36 33 Based on subsequent correspondence, Petitioners estimate that 19.2 EFPY occurred around 2007 in the 14th RFO. See Attachment 3 for a table showing the estimated timing of this and other actual or planned capsule withdrawals.

34 PG&E Letter DCL-92-072, Enclosure at 4. PG&E also proposed to move some of the capsules in the existing program upon insertion of the new capsules.

35 Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Supplemental Reactor Vessel Radiation Surveillance Program, Pacific Gas and Electric Company, Diablo Canyon Power Plant, Unit 1, Docket No. 50-275 at 3 (Sept. 4, 1992) (ADAMS Accession No. ML16341G685) (NRC Safety Evaluation for Supplemental Surveillance Program).

36 Id.

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2. License amendment to recover thirteen-year construction period In July 1992, before the NRC had approved PG&Es supplemental surveillance program, PG&E cited the supplemental surveillance program in support of a license amendment application to recapture the thirteen-year construction period for Unit 1 by changing the expiration dates of the Unit 1operating license from April 23, 2008 to September 22, 2021. 37 In the application, PG&E stated that its existing surveillance program will effectively monitor vessel embrittlement throughout the requested license period. 38 And PG&E asserted that:

In addition to those required surveillance programs, a supplemental surveillance program will be implemented for Unit 1 beginning with Cycle 6 in 1992. The supplemental program consists of four new surveillance capsules that will provide additional data to better manage vessel embrittlement issues during the plant operating life.39 These four new capsules included Capsule B. Further, PG&E asserted that for both reactors:

The overall program to monitor reactor vessel beltline materials is thorough and comprehensive. It meets all applicable regulatory requirements and will yield continuous information relevant to determining the degree of embrittlement of beltline materials over the proposed 40-year operating license terms.40 Nowhere in the license amendment application did PG&E state that the supplemental surveillance program was related to license renewal. Instead, PG&E took credit for the supplemental surveillance program in seeking to extend the original operating license for Unit by thirteen years.

37 PG&E Letter DCL-92-154 from Gregory M. Rueger, PG&E to NRC re: License Amendment Request 92-04, 40-Year Operating License Application (July 9, 1992) (PG&E Letter 92-04)

(ADAMS Accession No. ML16341G621). PG&E also applied to extend the Unit 2 operating license expiration date from December 9, 2010 to April 26, 2025.

38 Id., Attachment A (License Amendment Application) at 14.

39 Id., Attachment A at 15.

40 Id., Attachment A at 15. As discussed in the Macdonald Declaration, § V.A.1, this conclusion was erroneous.

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The NRC Staff approved the license amendment, citing, inter alia, PG&Es comprehensive vessel material surveillance program [that] is maintained in accordance with 10 CFR Part 50, Appendix H that ensures the fracture toughness requirements of Appendix G are met.41 The Staff did not mention license renewal. The license amendment was noticed in the Federal Register.42

3. Withdrawal and testing of Capsule V In 2002, PG&E withdrew Capsule V from the Unit 1 pressure vessel and conducted Charpy tests for PTS reference temperature and USE.43 PG&E subsequently reported that it had calculated a limiting RTPTS value of 250.9 oF for the limiting weld 3-442C.44 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 10 oF below the screening limit of 270 oF.

Taking into consideration a reasonable margin of error of about +/- 10 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. 45 Nevertheless, PG&E discounted the data as not . . .

credible.46 Instead of crediting the data it had gathered from Unit 1, PG&E substituted generic 41 Letter from Melanie A. Miller, NRC, to Gregory M. Rueger, PG&E, re: Issuance of Amendments for Diablo Canyon Nuclear Power Plant, Unit No. 1 (TAC No. M84006) and Unit No. 2 (TAC No. M84007), enclosed Safety Evaluation at 2 (March 1, 1995) (1995 License Amendment) (ADAMS Accession No. ML022340183).

42 See, e.g., 57 Fed. Reg. 32,575 (July 22, 1992) (proposed No Significant Hazards Consideration Determination).

43 PG&E Letter DCL-038.

44 Id.

45 Macdonald Declaration, § III.

46 PG&E Letter DCL-038 at 1.

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data and data from other reactors.47 But PG&E gave no indication of intending to rely on generic data and data from other reactors for a significant length of time. Instead, PG&E asserted that Capsule V is not the last planned capsule to be evaluated in the [Diablo Canyon Unit 1]

surveillance program.48

4. License amendment to recover three-year low-power testing period In 2005, citing a new NRC policy to allow the recovery of time spent on low-power testing of nuclear reactors, PG&E again applied to extend the Unit 1 operating license term, this time by three years.49 PG&E clarified that the proposed license amendment does not constitute license renewal.50 Like PG&Es 1992 license amendment application for recovery of construction time, its 2005 license amendment application for recovery of low-power testing time asserted that the original surveillance program for Unit 1 complies with ASTME E-185-70, the standard in effect when the vessel was designed and will ensure vessel embrittlement is effectively monitored throughout the requested license period.51 And like PG&Es 1992 license amendment application, the 2005 license amendment application took credit for the supplemental surveillance program for the three-year recovery period, asserting that it will provide additional data to better assess and manage vessel embrittlement issues during the plant operating life.52 47 Macdonald Declaration, § III.

48 PG&E Letter DCL-038 at 2.

49 PG&E Letter 05-098 from David H. Oatley to NRC re: License Amendment Request 05-03, Request for Amendment to Recapture Low-Power Testing Time (Aug. 23, 2005) (PG&E Letter DCL-05-03) (ADAMS Accession No. ML05240441).

50 Id., Enclosure 1 at 4.

51 Id.

52 Id., Enclosure 1 at 5 (emphasis added).

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In 2006, the NRC Staff approved the license amendment.53 Among the conclusions listed by the Staff in support of the license amendment was the Staffs determination that:

The RV [reactor vessel] surveillance schedules for DCPP-1/2 [Diablo Canyon Units 1 and 2]

remain in compliance with the requirements of 10 CFR Part 50, Appendix H, and the ASTM E 185 version of record for the units.54 Providing additional detail regarding this conclusion, the Staff asserted:

The licensee stated that the adjustments of the EOL neutron fluences for the RV beltline materials at the clad-to-base metal locations of the RVs do not require the RV material surveillance capsule withdrawal schedules for DCPP-1/2 to be altered. The NRC staff reviewed the limiting neutron fluence values reported in PG&E Serial Letter No. DCL-06-045 for the clad-to-base metal location of the RVs, in order to determine whether the revised fluence values would impact the RVMSP withdrawal schedules for DCPP-1/2.

The ASTM E185 version of record for DCPP-1 is ASTM E185-70. The most recent RVMSP withdrawal schedule for DCPP-1 was requested in PG&E Serial Letter No.

DCL-92-072, dated March 31, 1992. . . . This RVMSP [reactor vessel material surveillance program] withdrawal schedule was approved in an SE [Safety Evaluation] to PG&E dated September 4, 1992 . . . . In the SE, the NRC staff concluded the supplemental RVMSP withdrawal schedule met the criteria of ASTM E185-70 and constituted an acceptable withdrawal schedule for implementation under 10 CFR Part 50, Appendix H. Under this supplemental program, four capsules, Capsule S, Y, V, and B, were designated for removal from the DCPP-1 RV. Capsules S, Y, and V have been removed and tested in accordance with the licensees program.

The request to recover the testing time for DCPP-1 amends the projected withdrawal for Capsule B to approximately 20.7 EFPY, when the capsule is projected to achieve a neutron fluence of 2.9 x 1019 n/cm2 (E > 1.0 MeV). Therefore, the capsule will achieve a neutron fluence approximately equal to twice the projected limiting inside RV fluence for DCPP-1 at the EOL (i.e., approximately 2

• 1.43 x 1019 n/cm2 [E > 1.0 MeV]). This complies with the criterion in ASTM E185-82 for withdrawal of the final capsule of a four capsule withdrawal program. This is acceptable because 10 CFR Part 50, Appendix H, permits the licensees (sic) to meet the RVMSP withdrawal criteria of more recent versions of ASTM E185, inclusive of E185-82. Therefore, the NRC staff concludes that 53 Letter from Alan Wang, NRC, to John S. Keenan, PG&E, re: Diablo Canyon Power Plant, Unit Nos. 1 and 2 - issuance of Amendments re: Request for Recovery of Low-Power Testing Time-Impact on the Reactor vessel Integrity Assessments (TAC Nos. MC8206 and MC 8207)

(July 17, 2006) (2006 License Amendment) (ADAMS Accession No. ML062260278).

54 Id., enclosed Safety Evaluation at 6.

16

the adjustments to the withdrawal time and projected neutron fluence for Capsule B will still be in compliance with 10 CFR Part 50, Appendix H.55 Thus, the Staff viewed Capsule B as part of a four-capsule program that also included Capsules S, Y, and V, which were included in PG&Es original surveillance program. And PG&Es proposed schedule for withdrawal of Capsule B at 20.7 EFPY was a condition for the Staffs approval of PG&Es license amendment application.56 The license amendment was noticed in the Federal Register.57 Accordingly, the Staff relied on PG&Es supplemental surveillance schedule - including removal and testing of Capsule B between 2007 and 2009 -- in approving two separate license amendments that added a total of sixteen years to the term of PG&Es original full-power operating license. And in each case, the public was informed of the change to PG&Es operating license by publication of a notice in the Federal Register.

5. Capsule B withdrawal re-purposed to serve license renewal at PG&Es discretion Starting in 2008, PG&E and the Staff exchanged no less than four sets of correspondence requesting and approving extensions to the schedule for removing and Capsule B, from 2009 to 2010, from 2010 to 2012, from 2012 to 2022, and then from 2022 to 2023 or 2025. This correspondence differed from PG&E Letter DCL-03-052 and the NRCs license amendment decisions in two fundamental respects:

First, both PG&E and the Staff began to assert that the surveillance program for the original license term had been completed with the withdrawal of Capsule V in 2002 and 55 Id., enclosed Safety Evaluation at 5.

56 As shown in Attachment 3, 20.7 EFPY is approximately calendar year 2009.

57 71 Fed. Reg. 46,945 (Aug. 15, 2006) (notice of license amendment issuance).

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that the supplemental surveillance program - including removal of Capsule B -- related to license renewal. Thus, they reasoned that the surveillance program for the original license term was complete, and withdrawal of Capsule B could be scheduled with great and forward-looking flexibility for the sole purpose of meeting PG&Es requirements for license renewal. On these entirely new grounds, PG&E repeatedly sought and was granted extensions of the schedule for removing Capsule B, farther and farther into the future until it stretched beyond the original 2024 retirement date for Unit 1.

Second, unlike the 1995 and 2006 license amendments, the Staffs subsequent approvals of extensions of the surveillance schedule were hidden from the public eye, with no notice published in the Federal Register.

The origin of this fundamental re-casting of the nature and purpose of the supplemental surveillance schedule can be found in a 2008 PG&E letter informing the Staff that PG&E was currently performing a License Renewal Feasibility Study to decide whether to apply for license renewal for the Diablo Canyon reactors. 58 According to PG&E, its current surveillance program did not satisfy the NRCs license renewal guidance because PG&E did not have a vessel material coupon that has fluence exposure equivalent to 60 years of operation. 59 But the guidance would be satisfied by removing Capsule B at approximately 21.9 EFPY. 60 The NRC Staff approved the requested extension, pivoting sharply away from the position underlying the 1995 and 2006 license amendments that withdrawal of Capsule B around 58 Letter DCL-08-012 from James R. Becker to NRC, re: Revision to the Unit 1 Reactor Vessel Material Surveillance Withdrawal Schedule, Enclosure 1 at 1 (March 12, 2008) (ADAMS Accession No. ML080850564).

59 Id. (citing NUREG-1801, Generic Aging Lessons Learned (GALL) Report).

60 Id. at 2. As shown in Attachment 3, a removal time of 21.9 EFPY is about 2010 in calendar years.

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19-20 EFPY was essential to the extension of PG&Es operating license by sixteen years. For the first time, the Staff asserted that the removal of Capsule V in 2002 had fulfilled the third and final recommendation of ASTM E 185-70 for the current [Diablo Canyon Unit 1] operating license.61 By the same token, the Staff also asserted for the first time that removal of Capsule B was not required during the current operating license term, and thus the proposed delayed removal of Capsule B does not deviate from the licensees current RPV materials surveillance program requirements.62 In other words, no deviation had occurred because the surveillance program for Unit 1 no longer existed. Because the removal and testing of Capsule B was not required by PG&Es current license, it could be re-scheduled as needed to be useful for PG&Es license renewal plans.63 After seeking and obtaining the extension sought in PG&E Letter DCL-08-012, PG&E subsequently sought and obtained three additional extensions. These letters repeat and amplify the themes of PG&Es Letter DCL-08-12 and the NRCs response, i.e., that the withdrawal of Capsule B is not part of the pressure vessel surveillance program for the current operating license term, which has now concluded; and that Capsule B relates only to license renewal and its withdrawal can be scheduled to help PG&E satisfy license renewal requirements.64 61 Letter from Alan Wang, NRC, to John Conway, PG&E, re: Diablo Canyon Power Plant, Unit No. 1 - Approval of Proposed Reactor Vessel Material Surveillance Capsule Withdrawal Schedule (TAC No. MD8371), enclosed Safety Evaluation at 2 (Sept. 24, 2008) (ADAMS Accession No. ML082380306) (emphasis added).

62 Id.

63 Id., enclosed Safety Evaluation at 2.

64 See the following:

PG&E Letter DCL-10-141 from James R. Becker to NRC re: Revision 1 to the Unit 1 Reactor Vessel Material Surveillance Withdrawal Schedule (Oct. 25, 2010) (ADAMS Accession No. ML102990079) and Letter from Carl F. Lyon, NRC to John T. Conway, 19

As a result of these delays, by the time Capsule B is removed more than twenty years will have passed since PG&E last withdrew and tested a surveillance capsule from the Unit 1 pressure vessel.65 And while the NRC has issued to PG&E an exemption that allows it to operate Unit 1 indefinitely under the current license, the NRC Staff no longer considers that it has a surveillance program that could be enforced against PG&E in this operating license term. As a result of the Staffs change of position, it now considers withdrawal of Capsule B a discretionary task that PG&E may undertake on its own schedule.

IV. PETITIONERS ARE ENTITLED TO A HEARING BECAUSE THE 7/20/23 EXTENSION ORDER EFFECTIVELY AMENDED PG&ES OPERATING LICENSE FOR UNIT 1 While the NRC Staff did not characterize the 7/20/23 Extension Order as a license amendment, the Order meets the judicial standard adopted by the Commission in Cleveland Electric Illuminating Co. (Perry Nuclear Power Plant, Unit 1), CLI-96-13, 44 N.R.C. 315 (1996)

(Cleveland Electric):

In evaluating whether challenged NRC authorizations effected license amendments within the meaning of section 189a, courts repeatedly have considered the same key factors: did the PG&E (Oct. 29, 2010) (requesting and granting an extension from 2010 to 2012)

(ADAMS Accession No. ML03010159);

PG&E Letter DCL-11-122 from James R. Becker to NRC re: Revision to the Unit 1 Reactor Vessel Material Surveillance Program Withdrawal Schedule (Nov. 21, 2011)

(ADAMS Accession No. ML113260072) and Letter from Joseph M. Sebrosky, NRC to John T. Conway, PG&E re: Diablo Canyon Power Plant, Unit No. 1: Safety Evaluation for Request to Revise the Reactor Vessel Material Surveillance Withdrawal Program TAC ME7615) (March 2, 2012) (ADAMS Accession No. ML120330497) (requesting and granting an extension from 2012 to 2022);

PG&E Letter DCL-23-038 and NRC 7/20/23 Extension Order (requesting and granting an extension from 2022 to 2025).

65 Capsule V probably was withdrawn in 2002 and was tested in 2003. See PG&E Letter DCL-03-052.

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challenged approval grant the licensee any greater operating authority, or otherwise alter the original terms of a license?66 These circumstances meet the Cleveland Electric test because the 1995 and 2006 license amendments for recapture of thirteen years of construction and three years of low-power operation were conditioned on PG&Es surveillance schedule, including the supplemental surveillance plan. PG&E got greater operating authority, i.e., authority to operate the Unit 1 reactor for a much longer period, as a result of its commitment to carry out the supplemental surveillance schedule as described. Id., 44 N.R.C. at 326. In exchange for that greater operating authority, the Staff required that PG&E must provide a more robust surveillance program than before, by adding Capsule B to Capsules S, Y, and V. As stated in the 2006 Safety Evaluation,

[u]nder this supplemental program, four capsules, Capsule S, Y, V, and B, were designated for removal from Diablo Canyon Unit 1.67 As a result of the Staffs reliance on the supplemental surveillance program to justify extended operation, the supplemental surveillance program became a part of PG&Es license that may not be changed without notice and the offer of an opportunity for a hearing, as required by Section 189a the Atomic Energy Act. Cleveland Electric, 44 N.R.C. at 327 (citing Massachusetts

v. NRC, 878 F.2d 1516 (1st Cir. 1989)). The Staffs subsequent issuance of effective license amendments in 2010, 2012, and 2023 does not preclude Petitioners from challenging the most recent of these effective license amendments, because none was issued with public notice or an opportunity to participate.

66 Id., 44 N.R.C. at 326 (quoting, respectively, In re Three Mile Island Alert, 771 F.2d 720, 729 (3d Cir. 1985); San Luis Obispo Mothers for Peace v. NRC, 751 F.2d 1287, 1314 (D.C. Cir.

1985). See also id., 44 N.R.C. at 327 (quoting Citizens Awareness Network v. NRC, 59 F.3d 284, 295 1st Cir. 1995) holding that an NRC regulatory action that undeniably supplement[ed] the original license constituted licensing action) (emphasis in original)).

67 2006 License Amendment, Safety Evaluation at 6.

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V. CONTENTION 1 (Safety)

A. Statement of Contention 1 PG&Es request to postpone the withdrawal and testing of Capsule B until 2025 should be denied, and the Staffs decision to approve it should be reversed, because it is inconsistent with NRC safety regulations 10 C.F.R. Part 50, Appendices G and H and 10 C.F.R. §§ 50.55a and 50.61 and poses an unacceptable risk to public health and safety in violation of NRC regulations and the Atomic Energy Act. Moreover, neither PG&E nor the Staff has any legal grounds for claiming that withdrawal of Capsule B relates only to license renewal and is unnecessary to maintain safety in the current license term.

B. Basis for contention.

Petitioners first basis for this contention is the attached Macdonald Declaration, which sets forth a comprehensive set of legal and technical grounds for reaching three primary conclusions: (1) that PG&E is operating Unit 1 in violation of NRC regulations for reactor vessel safety; (2) it is posing a serious safety risk to the public and the environment; and (3) it should be required to immediately resume the pressure vessel surveillance measures that it has postponed since 2023, namely the removal and testing of Capsule B. Petitioners adopt and incorporate by reference his declaration. To briefly summarize his points, PG&E has ignored credible data showing that embrittlement may be approaching legal limits, thus warranting more testing, not less. In addition, Dr. Macdonald has performed an independent analysis that confirms this concern. Further, PG&E has relied for far too long on generic data and data from sister reactors to justify the safety of continued operation without additional testing. Finally, PG&E has also postponed another critically important test of pressure vessel integrity, UT inspection of reactor beltline welds. As a result, for a twenty-year period between 2005 and 2025, PG&E has no 22

updated data on the prevalence of voids and cracks in these welds; and even the data it has collected are suspect for their paucity of results. 68 Thus, by postponing both the withdrawal and testing of Capsule B and UT inspection of the beltline welds, PG&E has deprived itself and the NRC of any updated Unit 1-specific information regarding the condition of the pressure vessel.

These lapses are particularly serious in light of Diablo Canyons proximity to a web of significant earthquake faults and its defective chemical composition. 69 Second, Petitioners rely on the language in the 1995 License Amendment and the 2006 License Amendment which establishes that withdrawal of Capsule B is required by those license amendments as a condition for operating Unit 1 during the current license term. Further, Capsule B may not be treated solely as a prospective matter that is relevant only to the proposed license renewal term. See also discussion above in Section III.D.5, which is incorporated by reference into this basis statement.

C. Demonstration That the Contention is Within the Scope of the Proceeding This contention is within the scope of the proceeding for the change to PG&Es reactor vessel surveillance schedule because it raises concerns about whether the change will comply with NRC safety standards or pose an undue risk to public health and safety.

D. Demonstration That the Contention is Material to the Findings NRC must make to Approve the Proposed Schedule Change.

This Contention is material to the findings NRC must make regarding the proposed schedule change because the NRC may not issue a license amendment without first concluding that it complies with NRC regulations and poses no undue risk to public health and safety.

68 Macdonald Declaration, § V.B.

69 Dr. Macdonalds concerns about the proposed extension of the deadline for removing and testing Capsule B are summarized in Sections III and V.C of his declaration.

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E. Concise statement of the facts or expert opinion supporting the contention, along with appropriate citations to supporting scientific or factual materials The facts supporting Petitioners contention are set forth in the Basis Statement in Subsection B above, in official PG&E and government documents as cited in the Statement of the Contention and Basis Statement, and in the attached Macdonald Declaration.

VI. CONTENTION 2 (Environmental)

A. Statement of Contention 2 PG&Es request to postpone the withdrawal and testing of Capsule B until 2025 should be denied, and the Staffs decision to approve it should be reversed, because the extension is not supported by an analysis of its environmental impacts that complies with the National Environmental Policy Act (NEPA) or NRC implementing regulations in 10 C.F.R. §§ 51.20 and 51.30. These regulations require the NRC to evaluate the environmental impacts of its proposed actions, including license amendments, before going forward.

B. Basis for contention.

Petitioners rely on the attached Macdonald Declaration, which sets forth a comprehensive set of technical grounds for concluding that the proposed extension of the schedule for withdrawing and testing Capsule B from Unit 1 poses an unacceptable risk to human health and the environment. As Dr. Macdonald asserts in Section IV.A of his declaration, the pressure vessel is a uniquely important part of a reactor coolant system, because it holds the highly radioactive core under water and because it has no backup if it should fail. The consequences of a core melt accident caused by reactor vessel failure could be catastrophic. The NRC should perform an environmental analysis that thoroughly considers the current state of knowledge about the condition of the Unit 1 pressure vessel, its potential to cause a significant radiological 24

accident, and alternatives for mitigating or avoiding those impacts. See 10 C.F.R. § 51.70 for the NRCs general requirements for an environmental impact statement and 10 C.F.R. § 51.30 for the NRCs requirements for an environmental assessment.

C. Demonstration That the Contention is Within the Scope of the Proceeding This contention is within the scope of the proceeding for the change to PG&Es reactor vessel surveillance schedule because it raises concerns about the NRC Staffs lack of compliance with NEPA and NRC implementing regulations.

D. Demonstration That the Contention is Material to the Findings NRC must make to Approve the Proposed Schedule Change.

This Contention is material to the findings NRC must make regarding the proposed schedule change because the NRC may not issue a license amendment without evaluating its environmental impacts, as required by NEPA and the NRCs implementing regulations.

E. Concise statement of the facts or expert opinion supporting the contention, along with appropriate citations to supporting scientific or factual materials The facts supporting Petitioners contention are set forth in the Basis Statement in Subsection B above, in official PG&E and government documents as cited in the Statement of the Contention and Basis Statement, and in the attached Macdonald Declaration.

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VII. REQUEST FOR SHUTDOWN ORDER AND REMEDIAL MEASURES A. Exercise of Commissions Discretionary Supervisory Authority is Warranted.

This matter warrants Commission involvement for three important reasons. First, as recognized by the Commission in Yankee Rowe, the Commission has the ultimate responsibility for the safe operation of the facilities that it licenses.70 The safety concerns raised by decades of PG&Es evasion of its responsibilities for monitoring the condition of the pressure vessel are among the gravest that the Commission can encounter, given the vulnerability of the pressure vessel to embrittlement, and given the lack of any backup if it should fail. In the case of Diablo Canyon, both the reactors proximity to a web of earthquake faults and its inherently defective composition exacerbate the risks caused by PG&Es avoidance and neglect of its responsibilities.

Here, Dr. Digby Macdonald, a highly experienced and respected expert in the field of materials in nuclear reactors, has closely investigated the Diablo Canyon situation and found that PG&E has disregarded credible evidence of embrittlement and systematically avoided testing that would shed light on the reactor vessels condition. Dr. Mcdonalds own calculations, using data established as credible by PG&E, independently confirmed a serious risk of embrittlement.

This situation would never have occurred if PG&E and the Staff had dealt with the problems instead of continually ignoring them and postponing necessary tests and inspections. Given these failures by both PG&E and the Staff, the Commission must step in to provide the reasonable assurance that has been so conspicuously lacking for decades.

Second, the Commission should take review of the regulatory shell game played by PG&E with Capsule B to avoid surveillance testing for two decades. When it was convenient for PG&E to credit the withdrawal of Capsule B to the surveillance program for the current 70 34 N.R.C. at 12.

26

operating license, PG&E did so and thereby won approval of license extensions in 1995 and 2006. Then when it was more convenient to credit the withdrawal of Capsule B to license renewal, PG&E shifted its stance and starting kicking the Capsule B can down the road towards the license renewal term and finally into it. There is only one Capsule B, it has yet to be removed for any purpose, and it is not clear when it will be removed, if ever. Given the Staffs key role as an enabler of this shell game (see Section III.D.5 above), only the Commission can end it.

Finally, PG&Es shell game has particularly egregious risk and regulatory implications with respect to the particular circumstances of Diablo Canyon. Now that the Commission has exempted PG&E from the timely renewal rule, 71 PG&E no longer has an end date to its current operating license. Operation could go on for years - potentially decades -- while the NRC reviews PG&Es license renewal application, leaving Petitioners and other members of the public in limbo between the current operating license - for which the NRC Staff has declared that the surveillance of the Unit 1 pressure vessel has ended - and the license renewal term, for which the requirements for a surveillance program have yet to be determined.

B. Unit 1 Must be Shut Down to Protect Public Health and Safety and Should not Be Reopened Until PG&E Has Conducted Adequate Tests and Inspections, Disclosed Their Data and Results, and Subjected Them to Expert Review and a Public Hearing.

As set forth in Section IV of the Macdonald Declaration, in order to fulfill its statutory responsibility to protect health and safety, the Commission must order the immediate shutdown of the Unit 1 reactor. It must also order the reactor to remain in a shutdown condition until the set of actions listed in Section IV of Dr. Macdonalds declaration have been satisfied. These actions include:

71 Notice of Exemption Issuance, 88 Fed. Reg. 14,395 (March 8, 2023).

27

a) Withdrawal and analysis of the contents of Capsule B as well as other capsules previously withdrawn but not analyzed; b) Evaluation and analysis of wedge opening loading (WOL) specimens contained in Capsule B, C and D and 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.

In addition to the technical demands above, Petitioners wish to emphasize their procedural demand for transparency and public participation in this process. Throughout their review of the record set forth here and in Dr. Macdonalds declaration, Petitioners and their expert consultant have found a disturbing lack of transparency, including the difficulty or impossibility of obtaining some documents that were key to understanding PG&Es and the Staffs actions. It also became clear to Petitioners that they could not rely on either PG&E or the government for robust implementation or enforcement of NRC regulations and regulatory standards. Thus, Petitioners engaged Dr. Macdonald and worked with him for weeks to understand what has happened - or not happened - at Diablo Canyon in the last twenty years.

This pleading and Dr. Macdonalds declaration, the fruit of Petitioners labors, reflect a substantial investment of time and resources to do what appears to be the work of the government.

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We now hand this fully investigated matter back to the highest officials of the agency, with a demand for accountability for the government lapses and inaction that are documented here. Before Unit 1 may be permitted to resume operation, this accountability must be provided in a transparent and rigorous public hearing process.

VIII. CONCLUSION For the foregoing reasons, Petitioners request the NRC Commissioners to grant their hearing request, as required by Section 189a of the Atomic Energy Act and NRC implementing regulations. Petitioners also request the Commission to exercise their supervisory authority to order the immediate shutdown of Unit 1, pending completion of the remedial measures, a thorough NEPA analysis, public disclosures and the hearing process set forth in Section VII above.

29

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 September 14, 2023 30

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 CERTIFICATE OF SERVICE I certify that on September 14, 2023, I posted on the NRCs Electronic Information Exchange the following documents:

REQUEST TO THE NRC COMMISSIONERS BY SAN LUIS OBISPO MOTHERS FOR PEACE AND FRIENDS OF THE EARTH FOR A HEARING ON NRC STAFF DECISION EFFECTIVELY AMENDING DIABLO CANYON UNIT 1 OPERATING LICENSE TO EXTEND THE SCHEDULE FOR SURVEILLANCE OF THE UNIT 1 PRESSURE VESSEL (Sept. 14, 2014)

AND REQUEST FOR EMERGENCY ORDER REQUIRING IMMEDIATE SHUTDOWN OF UNIT 1 PENDING COMPLETION OF TESTS AND INSPECTIONS OF PRESSURE VESSEL, PUBLIC DISCLOSURE OF RESULTS, PUBLIC HEARING, AND DETERMINATION BY THE COMMISSION THAT UNIT 1 CAN SAFELY RESUME OPERATION (Sept. 14, 2023);

Attachment 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 (Sept. 14, 2023);

Attachment 2A, Declaration of Kaoru Hisasue (Sept. 7, 2023); Attachment 2B, Declaration of Lucy Jane Swanson (Sept. 9, 2023); and Attachment 2C, Declaration of Jill ZamEk (Sept. 8, 2023); and Attachment 3, Table of Estimated Dates of Capsule Withdrawals ERRATA TO REQUEST TO THE NRC COMMISSIONERS (Sept. 14, 2023)

___/signed electronically by/__

Diane Curran 31

ATTACHMENT 1 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

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 ii

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 iii

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 iv

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

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 2

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 3

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 H 2O 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.

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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 10 o below the screening limit of 270 o

F. Taking into consideration a reasonable margin of error of about +/- 10 oF (as estimated 5

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 RT PTS 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.

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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.

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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.

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

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.

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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.

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 RTNDT to shift by about 68 o

C (154 oF) and the USE to be reduced by 61 J.

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 RT NDT. 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 RT NDT at 41 J (30 ft-lb) recommended by the ASME Pressure Vessel Code and adopted uncritically by the NRC. Both the RT NDT 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 RT NDT 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 RT NDT 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.

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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 (ART NDT) 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 11

specimen yields a measure of the true fracture toughness, K IC 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 K IC 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 (RT NDT) 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 200 o F. PG&E 12

(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 ART NDT 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 (#/cm 3), 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.

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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 RT PTS 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 o

F) 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 RT PTS 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 14

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.

15

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.

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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 RT NDT data; second, because they are unencumbered with corrections, such as the chemistry factor, margin, and the fluence factor that are required to correct RT NDT 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 17

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.

EoE vs. Fluence 0.6 0.5 0.4 Weld EoE 0.3 Plate HAZ 0.2 EoE = 5E-21f+ 0.4204 SRM 0.1 R² = 0.8235 Linear (Weld) 0 0 2E+19 Fluence /n.cm-2 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 1.37x10 19 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.

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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 cm 3 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 (#/cm 3), 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 19

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 (EoE crit) 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/cn 2, 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 RT NDT,PoI is a more fundamentally-based and hence superior metric for defining the state of embrittlement than is RTNDT,30.

USE vs EoE for all Materials from 160 Capsules S, Y, and V, Diablo Canyon, Unit 140 1.

USE /ft-lb 120 100 80 60 40 USE = 9.4378EoE-2.59 20 R² = 0.9976 0

0.3 0.35 0.4 0.45 0.5 0.55 EoE 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 (R 2 = 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 R 2 = 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.

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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.

Fluence vs. EFPYs 2.50E+19 Fluence n/cm2 2.00E+19 1.50E+19 1.00E+19 F = 5E+18ln(EFPY) + 2E+18 5.00E+18 R² = 0.9939 0.00E+00 0 10 20 30 40 50 EFPY 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 21

(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 cm 3 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 H 2-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.

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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)

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 RT NDT 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.

23

Figure 7: Comparison of RTNDT distributions in the same plate analysis in Grizzly and FAVOR. After Spencer et.al. (2016).

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 24

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 RT NDT) 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 25

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 RT NDT 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 26

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 RT NDT 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 27

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 (UTS unirr), 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 28

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 29

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.

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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) 31

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.

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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 33

APPENDIX A: Curriculum Vitae 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.

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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).

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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).

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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.

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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.

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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).

PG&E 2011. PG&E Letter DCL-11-136 from James R. Becker to NRC re: 10 C.F.R. 54.21(b) annual Update to the DCPP License Renewal Application Amendment Number 45 (Dec. 21, 2011) (ADAMS Accession No. ML12009A070).

PG&E 2014. Letter DCL-14-074 from Barry S. Allen to NRC re: ASME Section XI Inspection Program Request for Alternative RPV-U1-Extension to Allow Use of Alternate Reactor Inspection Interval (August 18, 2014) (ADAMS Accession No. ML14230A618).

PG&E 2023. PG&E Letter DCL-23-038 from Paula Gerfen to NRC re: Docket No. 50-275, OL-DPR-80, Diablo Canyon Unit 1, Revision to the Unit 1 Reactor Vessel Material Surveillance Program Withdrawal Schedule (May 15, 2023) (ADAMS Accession No. ML14230A618),

Shi, J; Wang, J; Macdonald, D.D, 2015. Prediction of primary water stress corrosion crack growth rates in Alloy 600 using artificial neural networks, Corros. Sci., 92, 217-227.

Spencer, B.; Backman, M.; Chakraborty, P.; and Hoffman, W, 2015. "Reactor Pressure Vessel Fracture Analysis Capabilities in Grizzly", INL/EXT-15-34736, Idaho National Laboratory, Idaho Falls, ID, Mar. 2015.

Spencer, B.W.; Backman, M.; Williams, P. T.; Hoffman, W. M.; Alfonsi, A.; Dickson, T. L.;

Bass, R.; and Klasky, H. B., 2016. Probabilistic Fracture Mechanics of Reactor Pressure Vessels with Populations of Flaws, http://www.inl.gov/lwrs.

U.S. NRC 1988. Regulatory Guide 1.99, Radiation Embrittlement of Reactor Vessels, Rev. 2.

(ADAMS Accession No. ML003740284).

U.S. NRC 1998. Generic Letter 92-01 and RVP Integrity Assessment, Status, Schedule and Issues (Feb. 12, 1998) (ADAMS Accession No. ML110070570).

U.S. NRC 2006. Letter from Alan Wang, NRC, to John S. Keenan, PG&E, re: Diablo Canyon Power Plant, Unit Nos. 1 and 2 - issuance of Amendments re: Request for Recovery of Low-Power Testing Time-Impact on the Reactor vessel Integrity Assessments (TAC Nos. MC8206 and MC 8207) (July 17, 2006) (ADAMS Accession No. ML062260278).

U.S. NRC 2007. NUREG-1806, Technical Basis for Revision of the Pressurized Thermal Shock (PTS) Screening Limit in the PTS Rule (10 CFR 50.61): Summary Report (Aug. 2007) (ADAMS Accession No. ML072830074).

U.S. NRC 2011. Letter from Robert A. Nelson, NRC, to W. Anthony Nowinowski, PG&E, re:

Revised Final Safety Evaluation by the office of Nuclear Reactor Regulation Regarding Pressurized Water Reactor Owners Group Topical report WCAP-16168-NP-A, Revision 2, Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval (July 26, 2011)

(ADAMS Accession No. ML111600295).

U.S. NRC 2015. Letter from Michael T. Markey, NRC, to Edward D. Halpin, PG&E, Re: Diablo Canyon Power Plant, Unit No. 1 - Request for Alternative RPV-U1-Extension to Allow Use of Alternate Reactor Inspection Interval Requirements (TAC No. MF4678) (June 19, 2015)

(ADAMS Accession No. ML15168A024.

Westinghouse 2003. WCAP-15958, Revision 0, Analysis of Capsule V from Pacific Gas and Electric Company Diablo Canyon Unit 1 Reactor Vessel Radiation Surveillance Program (Jan.

2003). https://mothersforpeace.org/wp-content/uploads/2023/09/2003-WCAP-15958-Rev.-0.pdf Westinghouse 2011. WCAP-17315-NP, Revision 0, Diablo Canyon Units 1 and 2 Pressurized Thermal Shock and Upper-Shelf Energy Evaluations (July 2011).

https://mothersforpeace.org/wp-content/uploads/2023/09/2011-WCAP-17315-NP-Rev.-0.pdf.

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ATTACHMENT 2 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE COMMISSION In the matter of Pacific Gas and Electric Company Docket Nos. 50-275, 50-373 Diablo Canyon Nuclear Power Plant Units 1 and 2 STANDING DECLARATION OF KAORU HISASUE IN SUPPORT OF REQUEST FOR HEARING AND EMERGENCY ORDER Under penalty of perjury, Kaoru Hisasu declares as follows:

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

2. I live at 2837 Clark Valley Road, Los Osos, California. My home is located approximately six miles from the Diablo Canyon Unit 1 and Unit 2 nuclear reactors.

3. It is my understanding that embrittlement of a nuclear reactor pressure vessel may make it vulnerable to fracture and a core melt accident. Therefore, I am very concerned that the U.S. Nuclear Regulatory Commission (NRC) has granted Pacific Gas and Electric Co.

(PG&E) multiple extensions of the schedule for evaluating conditions inside the reactor pressure vessel in Unit 1 of the Diablo Canyon nuclear plant by withdrawing and testing Capsule B. The latest extension, granted on July 20, 2023, would extend the time for withdrawing Capsule B until as late as spring 2025. PG&E has not collected or tested any samples from the Unit 1 reactor pressure vessel since 2003, and those results were inconclusive.

4. I believe PG&Es ongoing lack of knowledge regarding the condition of the Unit 1 reactor pressure vessel poses an unacceptable risk to my health and safety and the environment. Therefore, I have authorized SLOMFP and FOE to represent my interests by seeking a hearing and emergency order by the Commissioners.

Kaoru Hisasue August __, 2023

ATTACHMENT 3 ATTACHMENT 3 - TABLE OF ESTIMATED DATES OF CAPSULE WITHDRAWALS RF Date Capsule withdrawn, scheduled, EFPY EFPY (Projected)

O a empted, or skipped (Actual when removed) 1 1986(?) Capsule S removed and tested(a) 1.25(a) 2 3

4 5 1992(?) Capsule Y removed and tested(a) 5.86(a) 6 7

8 9

10 11 2002 Capsule V removed and tested(a) 14.3(a) 12.9(b) 12 ? Capsules C and D removed but not 15.9(a) 14.8(b) tested(a) 13 14 2007 Capsule B scheduled(b) 19.2(b) 15 2009 Capsule B re-scheduled(c) 20.7(c) 16 2010 Capsule B scheduled but removal failed(d) 17 2012 Capsule B re-scheduled(d) 18 19 20 21 22 23 5/2022 Capsule B scheduled but removal 33.0(a) skipped because PG&E had w/drawn license renewal applica on(e) 24 10/2023 Capsule B re-scheduled(e) 33.58(e) 25 Spring 2025 Capsule B re-scheduled(e) 34.97(e) 1

ATTACHMENT 3 - TABLE OF ESTIMATED DATES OF CAPSULE WITHDRAWALS Endnotes:

(a) 2023.05.13 PG&E le er DCL-23-038, Encl. 2 (2021 UFSAR Rev. 7, Table 5.2-22)

(b) 1992.03.31 PG&E Applica on for Supp. Surveillance Program (c) 2006.07.17 NRC Safety Evalua on for Low Power Tes ng Recapture License Amendment at 5.

(d) 2010.10.29 NRC Safety Evalua on (e) 2023.07.20 NRC Safety Evalua on 2