ML24089A013

03.25.24 Corrected Petitioners Excerpts of Record Volume 2

ML24089A013
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Case: 23-3884, 03/25/2024, DktEntry: 23.2, Page 1 of 222

Case No. 23-3884

IN THE UNITED STATES COURT OF APPEALS FOR THE NINTH CIRCUIT

SAN LUIS OBISPO MOTHERS FOR PEACE, INC.

AND FRIENDS OF THE EARTH, INC.

Petitioners,

v.

UNITED STATES NUCLEAR REGULATORY COMMISSION and the UNITED STATES OF AMERICA, Respondents,

PACIFIC GAS & ELECTRIC COMPANY, Intervenor

Petition for Review of Final Administrative Action of the United States Nuclear Regulatory Commission

PETITIONERS' EXCERPTS OF RECORD VOLUME 2

DIANE CURRAN RICHARD E. AYRES Harmon, Curran, Spielberg 2923 Foxhall Road, N.W.

& Eisenberg, LLP Washington, D.C. 20016 1725 DeSales Street NW, Suite 500 (202) 744-6930 Washington, D.C. 20036 ayresr@ayreslawgroup.com (240) 393-9285 dcurran@harmoncurran.com

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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 l

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 DeSa1es 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

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Table of Contents

I. INTRODUCTION AND

SUMMARY

1....................................................

II. DESCRIPTION OF PETITIONERS 4.......................................................

III. BACKGROUND 6..............................................................................

A. Role and Importance of the Reactor Vessel .6...........................................

B. Pressurized Thermal Shock 7...............................................................

C. Regulations Governing Safety of the Reactor Vessel 8.................................

D. History of Diablo Canyon Unit l Reactor Vessel 9.....................................

l. Licensing of Unit l 10......................................................

a. Reactor vessel surveillance program 10.........................

b. Supplemental surveillance program .l l..........................

2. License amendment to recover thirteen-year construction period.l3

3. Withdrawal and testing of Capsule V 14.................................

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

5. Capsule B withdrawal re-purposed to serve license renewal at PG&E's discretion 17.......................................................

E. PETITIONERS ARE ENTITLED TO A HEARING BECAUSE THE 7/20/23 EXTENSION ORDER EFFECTIVELY AMENDED PG&E'S 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 Commission's 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 Subj ected Them to Expert Review and a Public Hearing ..27.....................................................................................

VII. CONCLUSION 29............................................................................................................

ii

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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 l89a, 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()(2) 9.......................................................................................

10 C.F.R. Part 50, Appendix G 14, 22........................................................................

10 C.F.R. Part 50, Appendix G, § IV.A.1 9..................................................................

10 C.F.R. Part 50, Appendix H 14, 16-17,22...............................................................

10 C.F.R. Part 50, Appendix H, Section IH.B 3.............................................................

10 C.F.R. § 51.20 ..24...........................................................................................

10 C.F.R. § 51.30 24, 25........................................................................................

iii

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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...................................................................

Pacy'ic Gas and Electric Co.(Diablo Canyon Nuclear Power Plant, Units l 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 oflExemption 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 ASTME 182 9,11,15,16,19.................................................................................

NUREG-1801, GALL Report (date unknown) 18..........................................................

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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 l pressure vessel until 2025.1

As demonstrated in the attached supporting expert declaration of Dr. Dig by 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 l 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 Geffen, PG&E re: Diablo Canyon Nuclear Power Plant, Unit l - Revision to the Reactor Vessel Material Surveillance Capsule Withdrawal Schedule (EPID L-2023-LLL-0012) ("NRC 7/20/23 Extension Decision") (ADAMS Accession fN0.h4Ll20330497)

The 7/20/23 Extension Decision approves a schedule under which PG&E would withdraw "Capsule B" from the Unit l pressure vessel either during the upcoming 24threfueling outage

("lR24") in October 2023 or the 25threfueling outage in the spring of 2025 (lR25). 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 l, Revision to the Unit l 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. ML23l35A2l7).

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operating license term PG&E incorrectly discarded these data as "not credible.793In addition,

Dr. Macdonald's 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 dz 10 EFPY).4Taking into account

the level of uncertainty of dz 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

2025.6As stated by Dr. Macdonald, PG&E's decades of neglect, coupled with serious

2 Attachment 1, Declaration of Dig by Macdonald, Ph.D in Support of Hearing Request and Request for Emergency Action, § V.A.1 '11 (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 ld., § v.A.1.

414, § 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.7Petitioners seek a hearing on the

serious safety and regulatory issues raised by PG&E's 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.799As 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.10And 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 l

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, 1111, § 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 NRC's 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 alsoMacdonald Declaration,

§ IV.A.

10 Macdonald Declaration, § IV.

II 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)("YankeeRowe") (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

nana-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 l is allowed

to resume operation.

Due to the gravity of the safety and environmental risks presented by PG&E's and the Staff" s

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 inthe YankeeRowe 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&E's 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

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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 Blower 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&E's 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 l, 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 world's 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."

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 "F by the water entering the

VSS1.14

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 reactor's 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 ZA, 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.

15Id.

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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, KIck 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 reactor's 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 RTNDT. In a new vessel, the RTNDT is in the

range of 0 to 40°F. 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 RTNDT. 19

c. Regulations Governing the Safety of the Reactor Vessel As the NRC has recognized, given the singular importance of a nuclear reactor's pressure

vessel, 44[m]aintaining the structural integrity of the reactor pressure vessel ... is a critical

concern related to the safe operation of nuclear power plants.7920The 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 Commission's regulations establish design and performance standards that are

intended to assure for each plant that the probability of pressure vessel rupture is extremely

10W.21

NRC regulations in 10 C.F.R. Part 50, Appendix G, § IV.A.l 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 270°13 for all plates, forgings, and axial weld materials and 300°13 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. § 50.6l(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 HI.B.2) and an NRC-

approved withdrawal and testing schedule(id., Section HI.B.3). Surveillance capsules must also

contain coupons to measure tensile stress/strain, which are indicative of embrittlement." 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 oflMaterials, 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 plant's life, and when in the plant's 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." 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 l and 2. The Unit l 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. MLl7083C429)

("Rueger Letter").

26Id. See also Pacu9c Gas and Electric Co.(Diablo Canyon Nuclear Power Plant, Units l and 2),

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

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intermediate shell plate.7927 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.

b. Supplemental surveillance program

In 1992, PG&E applied to supplement the Unit 1 surveillance program by adding Capsules

A, B, C, and D." 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.m30

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.m31The 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.m32To 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 EFPY33after 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

79 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.7934

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.35The Safety Evaluation's 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 14thRFO. 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 l, Docket No. 50-275 at 3 (Sept. 4, 1992) (ADAMS Accession No. MLl634lG685) ("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&E's 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 operating license from April 23, 2008 to September 22, 2021. In37

the application, PG&E stated that its existing surveillance program "will effectively monitor

vessel embrittlement throughout the requested license period.m38And PG&E asserted that:

In addition to those required surveillance programs, a supplemental surveillance program will be implemented for Unit l beginning with Cycle 6 in 1992. The supplemental program consistsoff oar new surveillance capsules that will provide additional data to better manage vessel embrittlement issaes daring the plant operating life."

These "four new capsules" included Capsule B. Further, PG&E asserted that for both reactors:

The overall program tomonitor 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.4°

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&E's "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.m41The

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.43PG&E subsequently reported that it had

calculated a limiting RTpTs value of 250.9 "F for the limiting weld 3-442C.44Thus, 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 "F below the screening limit of 270 °F.

Taking into consideration a reasonable margin of error of about :E 10 °F (as estimated by

inspection of the Charpy curves), PG&E's test showed that Unit 1 would be approaching the

limit at the end of its operating life.45Nevertheless, PG&E discounted the data as "not ...

credible.m46Instead 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. l (TAC No. M84006) and Unit No. 2 (TAC No. M84007), enclosed Safety Evaluation at 2 (March l, 1995) ("1995 License Amendment") (ADAMS Accession No. ML022340183).

42 See, Ag.,57 Fed. Reg. 32,575 (July 22, l 992) (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 l.

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data and data from other reactors.47But 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 l]

surveillance program.m48

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.49PG&E clarified that the proposed license amendment "does not constitute license

renewal.m50Like PG&E's 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.m51And like PG&E's 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 issuesduring the plant operating Izfe.m52

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).

50Id., Enclosure l at 4.

51 Id.

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

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In 2006, the NRC Staff approved the license amendment.53Among 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-l/2 [Diablo Canyon Units l 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.m54

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-l/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-l/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-I RV Capsules S, Y, and V have been removed and tested in accordance with the licensee's program.

The request to recover the testing time for DCPP-l 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 1019n/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 licensee's (sic) to meet the RVMSP withdrawal criteria of more recent versions of ASTM El85, inclusive ofEl85-82. 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).

54Id.,enclosed Safety Evaluation at 6.

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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&E's original surveillance program. And PG&E's

proposed schedule for withdrawal of Capsule B at 20.7 EFPY was a condition for the Staff's

approval of PG&E's license amendment application.56The license amendment was noticed in

the Federal Register."

Accordingly, the Staff relied on PG&E's 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&E's original full-power

operating license. And in each case, the public was informed of the change to PG&E's operating

license by publication of a notice in the Federal Register.

5. Capsule B withdrawal re-purposed to serve license renewal at PG&E's 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 NRC's 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&E's 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 Staff's 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.58According to PG&E, its current surveillance

program did not satisfy the NRC's license renewal guidance because PG&E did not have a

"vessel material coupon that has fluence exposure equivalent to 60 years of operation.7959But 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 l at l (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&E's operating license by sixteen years. For the

first time, the Staff asserted that the removal of Capsule V in 2002 had "fulfilled thethird and

foalrecommendation of ASTM E 185-70 for the current [Diablo Canyon Unit 1] operating

license.m61By 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 licensee's current RPV materials surveillance program requirements. In other words, no deviation had occurred because the surveillancem62

program for Unit 1 no longer existed. Because the removal and testing of Capsule B was not

required by PG&E's current license, it could be re-scheduled as needed to be "useful" for

PG&E's 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&E's Letter DCL-08-12 and the NRC's 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. MD837l), 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-l4l 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. MLl02990079) and Letter from Carl F. Lyon, NRC to John T. Conway,

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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 l

pressure vessel.65And 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&E'S 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 inCleveland

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 l89a, 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&E's surveillance schedule, including the supplemental

surveillance plan. PG&E got "greater operating authority,"i.e.,authority to operate the Unit l

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 Staff" s reliance on the supplemental surveillance program to justify

extended operation, the supplemental surveillance program became a part of PG&E's 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. ClevelandElectric,44 N.R.C. at 327 (citing Massachusetts

v. NRC,878 F.2d 1516 (1st Cir. 1989)). The Staff's 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.3d284, 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&E's 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 oflNRC

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

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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.68Thus, by postponingboth 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 Canyon's proximity to a web of

significant earthquake faults and its defective chemical composition.

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 alsodiscussion 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&E's 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. Macdonald's 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&E's 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. §§ 5 l .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

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accident, and alternatives for mitigating or avoiding those impacts. See 10 C.F.R. § 51 .70 for the

NRC's general requirements for an environmental impact statement and 10 C.F.R. § 51 .30 for

the NRC's 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&E's reactor

vessel surveillance schedule because it raises concerns about the NRC Staff" s 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 NRC's 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 Commission's Discretionary Supervisory Authority is Warranted.

This matter warrants Commission involvement for three important reasons. First, as

recognized by the Commission in YankeeRowe, the Commission has the "ultimate responsibility

for the safe operation of the facilities that it licenses.m70The safety concerns raised by decades of

PG&E's 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 reactor's proximity to a web of earthquake faults and its inherently defective

composition exacerbate the risks caused by PG&E's avoidance and neglect of its responsibilities.

Here, Dr. Dig by 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 vessel's condition. Dr. Mcdonald's 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.

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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&E's 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," 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&E's 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. Macdonald's declaration have been satisfied. These actions

include:

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

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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 nana 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. 50.6la, 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 l 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. Macdonald's 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&E's and the

Staff" s 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. Macdonald's 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.

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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 15*** Street, 11*** Floor Washington, DC 20005 434-326-4647 htempleton@foe.org

Counsel to Friends of the Earth

September 14, 2023

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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 l

CERTIFICATE OF SERVICE

I certify that on September 14, 2023, I posted on the NRC's 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

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ATTACHMENT 1

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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 l

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

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Table of Contents

I. INTRODUCTION a 0 l a 0 . 1..............................................................................................

II. STATEMENT OF PROFESSIONAL QUALIFICATIONS ru .1

III.

SUMMARY

OF EXPERT OPINION o a a o . 4.................................................................

IV. BACKGROUND ON PRESSURE VESSEL AND REGULATORY REQUIREMENTS a 0 l a 0 l 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 criteria 13......................................................

A.l Unit 1 RTpTS surveillance data obtained in 2003, erroneously characterized by PG&E as "not credible", show that Unit l could approach NRC's threshold for remedial action as early as 2024 13.........................................................

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 il0 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 .2 l...................

C. PG&E has obtained no embrittlement data for Unit 1 for 18-20 years, at a significant risk to public health and safety 23.............................................

D. The NRC's extension of the deadline for beltline ultrasound inspections is not supported by adequate data 25...............................................................

E. Alternative testing methods would provide far more accurate results .26.............

IV. CONCLUSION AND RECOMMENDATIONS 32.....................................

APPENDIX A: CURRICULUM VITAE

APPENDIX B: REFERENCE LIST

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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

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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

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I. INTRODUCTION

1. I have been retained by San Luis Obispo Mothers for Peace (SLOMFP) and Friends of the Earth (POE) to evaluate changes in Pacific Gas and Electric Company's (PG&E's) program for surveillance of the Diablo Canyon Unit l 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 l 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 world's 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 > l 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 Electrochemistrywas the foundational text in the study of electrochemical systems using current and voltage perturbation techniques. These

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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 201 l, 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) H;/kg H2O)] and (b) the pH vs fuel burr up 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 burr up 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

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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 "redo 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 y-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 redo 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 l BWR suppression pool, flow-assisted corrosion at Surry Unit l, 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 (> l-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. l

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

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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 (< 10-1l 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&E's 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 270°F for axially oriented welds and 300°F 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 H20 of molecular hydrogen. The y, n, and a 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 l 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 l pressure vessel among the most embrittled, with only 14 of 72 PTS reference temperatures as high as or higher than Diablo Canyon Unit l. 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 l 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 250°F for the limiting weld 3-442C. Id. Thus, PG&E predicted that in 2021 (the expected retirement date for Unit l at that time), the reference temperature for Unit l would be slightly more than 10° below the screening limit of 270

°F. Taking into consideration a reasonable margin of error of about :E 10 °F (as estimated

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by inspection of the Charpy curves), PG&E's 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&E's failure to apply this well-established and reasonable guidance is both inexplicable and gravely concerning, given that the RTPTS data indicated a serious degree of embrittlement. The NRC Staff's approval of PG&E's disregard of the data is also puzzling, given that PG&E had ignored the agency's 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&E's 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 vessel's 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 l 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 "F 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&E's 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 Staff's recent decision to allow PG&E to postpone the next Charpy test for Unit l 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 l 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. § 50.5l(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 nana-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 Figurel. 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 "F by the nuclear fissioning of235U92 and toward the end of the core life by fissioning of various isotopes of plutonium such as 2g9Put and241Put. 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 l as the "150" active core length". It is this region that experiences the greatest fast neutron flux (E > l 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.

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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 reactor's 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, KIck This may result in the rapid propagation of a through wall crack in the embrittled vessel and in the failure of the vessel.

Q I I I L I

.- Figure2: Effect of neutron 4 un-In-aaiunqiilrrudialed irradiation on the Charpy

-I_- 9 4Ls8=6l.;impact test results for a fluence of 1020n/cm2(E > 1

_ i . MeV) for A508-3 RPV steel.

30 After Lin, et.al. Note that

-E 45 irradiation cause the value of

'Eo4 RTNDT RTNDT to shift by about 68 I °C (154 °F) and the USE to 1 be reduced by 61 J.

I EE .ltvlmililinisnl l

-'100 u. i

Test Temperature /°C

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

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

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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. § 50.6l(b). These standards establish two general sets of requirements: for fracture toughness as demonstrated by "Charpy" upper shelf energy (USE) and the shift in the adjusted nil ductile to brittle transition (ARTnDT) temperature of the embrittled (neutron irradiated) steel microstructure compared with the un-embrittled (unirradiated) microstructure and the fracture resistance to pressurized thermal shock (PTS). Appendix G sets a limit of 50-ft-lbs for the USE in a pressure vessel.

Section 50.6l(b)(2) establishes a screening criterion of 270 "F for (RTPTS) for axial welds and 300 "F for circumferential welds, where RTPTS is the reference temperature at the end of a reactor's operating life (EOL). If a reactor vessel is predicted to exceed the screening criterion, 10 C.F.R. § 50.6l(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 exposein situsamples 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 (r) from the core axis roughly as (T_0)2 , r > to, where rO 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 l 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 situthe yield stress (YS) and the ultimate tensile stress/strain, both of which are indicative of the state of embrittlement, while the WOL

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

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 1(2) requires licensees to have a "material surveillance program" with a schedule for removal and testing of surveillance capsules that conforms to industry standard ASTM E 185. NRC regulation 10 C.F.R. § 50.61I(2)(i) further requires all licensees to integrate the results of their plant-specific surveillance programs into the estimate of reference temperature (RTnDT) for the reactor vessel material.

6. In my professional opinion, the reactor-specific surveillance data required by the NRC's 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 l, 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. l % originates from the composition of the weld wire itself. The deleterious impact of both copper and nickel in the radiation embrittlement of welds in ferrite steels has been established by numerous laboratory and field studies.

After Diablo Canyon Unit l 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° F. PG&E

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(1992).3In 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&E's 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-l [Diablo Canyon Unit 1] at the EOL (i.e., approximately 2

• 1.43 X 1019 n/cm2 (E > 1.0 MeV]." U.s. NRC (2006).

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

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

12. The measurement of RTNDT and USE is only part of the story in assessing whether an embrittled RPV is in danger of rupture particularly under "pressurized thermal shock" (PTS) conditions resulting from the injection of cold water to compensate for loss of coolant from the rupture of the pressure boundary elsewhere. While ARTnDT and USE are appropriate monitors of the state of embrittlement, the probability of crack nucleation is a question that must be addressed by probabilistic fracture mechanics that requires the assessment of the population, size, and orientation of flaws close to the cladding/steel interface. Therefore, UT is used to evaluate flaw volume density (#/cm3), flaw size, and flaw orientation so as to determine if any flaw is characterized by a stress intensity factor (KI) that exceeds KIC for the embrittled steel. The American Society of Mechanical

3 PG&E inserted Capsule B into the Unit 1 pressure vessel and the NRC approved a schedule for withdrawing and testing it when the reactor achieved 19.2 EFPY. Id. See alsoTable 4. In 2006, in approving a license amendment for "recapture" of the three years of low-power testing of Unit l, 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 NRC's threshold for remedial action as early as 2024.

1. In my professional opinion, PG&E has incorrectly discredited the data it obtained from Unit 1 in Capsules S, Y and V for the purpose of calculating RTPTS values. PG&E should have been concerned that these data showed that Unit l could approach the PTS temperature screening limit by the end of the reactor's 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 l's EOL date of 32 EFPY (which at that time was 2021), the limiting RTpTS value calculated by PG&E's contractor, Westinghouse, for the limiting weld 3-442C was 250.9 °F. PG&E (2023), Westinghouse (2003). This calculation should have concerned PG&E because it was approaching the PTS screening criterion of 270 "F for plates, forgings and axial weld materials and within a reasonable margin of error of about dz 10 "F (as estimated by inspection of the Charpy curves), resulting in an overlap of uncertainties in the screening criterion (270 °F) and the Westinghouse estimate (250.9

°F) for weld 3-442C. In addition, as further explained in Section V.Al, the fact that the measured RTNDT for Capsule V (201 .07 °F) was lower than the value for Capsule Y that had been removed ten years earlier at lR5 (232.59 °F) (Westinghouse (2003), Table D-2) indicated a reasonable possibility that one of those tests was erroneous, because it unlikely that continued exposure to radiation would "heal" the metal. If the value of Capsule V was erroneous and the value of Capsule Y was correct, then the limiting RTpTs value Unit likely was even closer to the PTS screening criterion than calculated by PG&E.

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

2, Criterion 3 (U.S. NRC 1988)). According to RG 1.99, the scatter values for data "normally should be less than 28°F for welds and l7°F for base metal" PG&E (2003),

Westinghouse (2003). This is equivalent to i l Sigma. Therefore, PG&E declared that all

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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&E's methodology for assessing the credibility of the data is inconsistent with NRC's 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 outlieranda physical basis exists for believing the datapoint to be atypical -

. All data not excluded in (A.) should be used dS 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 i one standard deviation or a 96 % probability for falling within a i 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 i 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 reactor's 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 l 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 .

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deviation is too tight and in my professional judgment 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 l) 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 (= l) probability that the datum will fall outside the mean :E one standard deviation for no obvious reason. Thus, the observation that one point in the Diablo Canyon Unit l 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 ARTNDT with fluence so that I can judge for myself the validity of PG&E's non-credibility claim. Given the safety significance of PG&E's rejection of the Unit l 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 NRC's 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.

50.6la. 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 :t10 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&E's 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&E's 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 i 10 EFPY.

2. Given PG&E's failure in 2003 to present any Unit l-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 l.

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

4. By mathematically deriving an expression for the EoE from coefficients (A, 8, C, and TO) obtained for the symmetric hyperbolic tangent function (FE = A + B. tank [(T - TO) /C) that is used by PG&E to optimize on the fracture energy (FE) VS test temperature CIT data, I have calculated EOE = [1+-Xex;;_3/2 and x ; (RTnDT30 - To)/C where RTnDT30is the transition temperature that is defined for a fracture energy of 30 ft-lb (4 l 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 RTnDT,po1 =TO. By my reasoning, RTnDT,po1is a much better definition of the nil-ductility transition temperature than is the arbitrarily defined RTnDT30,as noted above. Note that at the point of inflection (Pol), the EoE = 0.5 indicating that the fracture is 50 % brittle and 50 % ductile. As we will see below, this

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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

• b 0.4 0' O' uJ II * * * .Weld oUJ0.3 . Plate 0.2 EoE = 5E-21 f+ 0.4204 . HAZ 0.1 R2= 0.8235

• SRM

.........Linear (Weld) 0 0 2E+19 Fluenceln.cm'2

Figure3: 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 l.

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 l.

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

over the first 14.27 EFPY operating life of the reactor. In contrast, for the weld metal, the EoE changes by about 8 %. Of course, the lack of sensitivity may also reflect that the plate, HAZ, and SRM do not embrittle rapidly, at least up to a fluence of 1.37x1019 n/cm2. Fortunately, the CIT does effectively detect the embrittlement of the limiting weld material.8

8 In my opinion, the CIT should be replaced, or at least complemented by another technique that does meet that standard of high sensitivity of the dependent variable on the principal independent variable. Such a technique might be nana 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&E's 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(RTnDT,307 RTnDT,po1,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&E's Charpy data passes through a maximum(RTnDT,307 RTnDT,po1)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 n + m <2 mi + Vm where n is the concentration of high energy neutron in l Cm3 of the metal in their transit from the entrance to the exit face of the metal cube and m, mi, and am 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: dlmd/m?l = k1f(1 - e`") mo - k_1[ ] wherem H i

[mi] is the concentration of displaced metal atoms (#/cm3), f is the fluence at the l cm; input face of the metal cube, and a is the neutron absorption coefficient in the metal.

Note that the thickness of the cube of metal is l cm. At steady state and at limitingly high 0

fluence (")f(1-dlmd/tm] = 0 and we obtain Mi = k 1 e"*)m.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 4.10All 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 extreme (maximum or minimum).

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

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locus that is represented by the equation USE = 9.4378E0E2.59 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 (EoEctit) 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 of2.09el9 n/cn2, E > l MeV. Note that the ratio of brittle VS. ductile facets on the fracture surface (ratio = l.l) is close to that defined by RTNDT,P01 (ratio = l) thereby supporting my conclusion that RTNDT,P0I is a more fundamentally-based and hence superior metric for defining the state of embrittlement than is RTNDT,30.

160 USE vs EoE for all Materials from 140 Capsules S, y, and v, Diablo Canyon, Unit

.Q 1.o ¢ 120 100 Ew: ._

• 80 60 0 4 40 USE = 9.4378EoE'2-59 20 R2= 0.9976 0

0.3 0.35 0.4 0.45 0.5 0.55 EoE

Figure4: Plot of USE VS. EoE for all materials from Capsules S, Y, and V, Diablo-Canyon, Unit l NPP.

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

Extrapolation of the data to the critical fluence of 2.09el9 yield the time at which the USE of the weld (24702) in the beltline equals the 50-ft-lb limit. That time is calculated as 43.8 EFPYs and is represented by the last datum on the right side of Figure 5.

Inclusion of this point in the fitting yields the same equation but with R2= 0.991 l. 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 i 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

NEoC2.00E+19 9

.O.°'.0-° 1.50E+19 I Il-,.¢.100.1l O... .. *. "

GJoCGJ3Ll. 5.00E+18 O.. **

• 1.00E+19 ,.x.r/I,. F = 5E+18ln(EFpy) + 2E+18

x'II R2 = 0.9939 6

0.005+00 0 10 20 30 40 50 EFPY

Figure5: 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 i 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 tum, 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 extreme (maximum or minimum) as is shown by PG&E's data (see, for example, the two highest fluence points in Figure 3), as that would require the expression for Mi

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(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, bull 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 l. 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. l l

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&E's 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 Cm3were detected in the Doel-3 reactor for a total of 7,776. Bogaerts et.al. (2022). Additional tests conducted in 2014 with adapted equipment detection parameters, revealed 13,047 voids and cracks in Doel-3 and 3,149 voids and cracks in Tihannge-2. Indications were found at depths ranging from 30 to 120 mm measured from the primary water side. Note that the thickness of the stainless-steel cladding is 7 mm, so that the indications occurred at 23 to 113 mm from the cladding/RPV steel interface. The indications were concentrated in the bottommost and upper core shell and were located in base metal, outside of the weld regions. These features can be correlated to steel microstructure and thermo-mechanical history (theoretical modeling) according to SCK-CEN, the Belgian Nuclear Research Centre.

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

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

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Figure6: 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 l, indicating mixed mode cracking not unlike that observed in other RPVs. Spencer and coworkers at INL have modeled RPV embrittlement within the Grizzly and FAVOR [Fracture Analysis of Vessels] codes.

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

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

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l. in I M I 44 II.

1000 . FAV5R Figure7: Comparison of 900 .I Grizzly RTNDT distributions in the

800 l same plate analysis in w3LE700 l Grizzly and FAVOR. After Spencer et.al. (2016).

600 I

`6 500 .6.oE3Z 400 !

300 .

200 .

100 l 0-so 0 50 100 150200 250 300 350

RTnDT

3. Accordingly, it is difficult to accept and understand PG&E's claim of detecting only one indication in the 2005 UT examination of beltline materials at Diablo-Canyon, Unit l, 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 alsoSection 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&E's 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 > l 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

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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 RTprS 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&E's 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 l. 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 ARTnDT) 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 l 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 l vessel. However, because the sister plants and Diablo Canyon Unit l do have unique operating histories alarger 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 l), and as specified in RGl .99, would fall within that band and would be more appropriate. By that standard, any legitimacy to

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PG&E's decision to discredit the results from Capsules S, Y, and V collapses.

Qd'

mI64

Nof

>>-4ds

oo'l p-3o p-20 p-o p p4-o l.z+2o 144-30

Figure8. The normal distribution function displaying the probability of an additional observation falling within pt Tr- na, where n = 1,2,3,....oo.

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 ARTnDT 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 NRC's 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 l for two decades without essential information on the condition of the pressure vessel. And the gap is all the more concerning given the

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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 l 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 l 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. § 50.5 l(c)(3) requires licensees to offer "information" that will "improve the accuracy of the RTpTs value significantly." The regulation doesn't 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 nana-indentation that is capable of obtaining many more replicate data than the conventional fracture mechanics methods prescribed by NRC regulations. The nano-indentation technique has been used for many years to assess embrittlement in steels and other alloys as reflected in a change in hardness. Briefly, a sharp point is pressed into a material under a known load and the dimensions of the indentation (width and depth) are measured. Thus, with increasing hardness, the depth and width of the indent become smaller. However, the relationship between hardness and RTNDT and USE still need to be established for this technique to replace the Charpy Impact Test. Nevertheless, I believe that can be done by using an Artificial Neural Network (ANN) to analyze the large body of information on RTNDT and USE vs. degree of embrittlement that is available from PWRs operating within the US and abroad.

2. I note that ASTM185-82 recommends indentation as an optional method for assessing the extent of embrittlement but it appears that too few plants have exercised that option to judge the viability of the method. However, the failed Charpy specimens are archived so

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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 (pa), indentation depth (pd), fluence U),

temperature of irradiation (TW), copper content [Cu], nickel content [Ni], unirradiated yield strength (YS), unirradiated ultimate tensile strength (UTsuni1-f), 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.

Figure9: Artificial Neural Network for establishing relationships between the dependent variables (RTNDT and USE) and the vector of the Input Variables (pa,pdf YM, [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

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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 nana-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 nana 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. Hosemann's 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 material's 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-situallows 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

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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 nana 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 nana hardness is used to calculate a macro hardness (corrected for pile up) which then in tum 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 nana 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 distribationfanction to be determined and the appropriate error to be specy9ed (e.g., the standard deviation) with an accuracy that is not possible asing Charpy analysis.

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I Load- Oliver and Displacement Pharr Method Data

I E-modulus Hardness vs Depth curve vs Depth curve r v r

'\\. \\ s 1 Pile-up Nix and Gao Vickers to correction plot (H2, 1lh) Berkovich Bad datapoint with AFM data -> H' and H0 conversion Low or High or reasonable extraction HV = 0.0945H0 E-modulus or odd assessment ,r ,II shaped curves. 4 Data that may suggest Hvto YS surface effect YS = 2.82 X HV- 114 or inclusions.

+

Compare to Tensile or SP data, (b) 1200 I I I I I I I I I I I

. ..l.

3o 1000 . °o* . ..,Cr

Q.:s 800 .4 ,.l. .

I2E6lE, 400 4"

600 a *I~¢

'cw44N.§u(D .. y=x 200 i' 250 nm

." .. 500 nm G.)is.

0.. I I I I I I 0 200 400 600 800 1000 1200 1400 Tensile or SP oy(MPa, Irradiated)

Figure 10:(a) Roadmap of nana indentation techniques. (b) Correlation between tensile test and shear punch test generated data to data obtained from nana 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 l l)

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Spherical Spherical Vs. Uniaxial

Jarrad Tate d

NIGrainGrain 1 Grain 21 i'l

'l:3am .Damage profileIII lilnirradiated Proton irradiation Distance from Edge [umUlPlastic Strain 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 l with which to assure the public of the reactor's 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 RGl .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 nana 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. 50.6la, 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 l 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 l 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 Dig by Macdonald Dig by Macdonald September 14, 2023

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APPENDIX A: Curriculum Vitae

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DIGBY D. MACDONALD Professor in Residence, Departments of Nuclear Engineering and Materials Science and Engineering University of California at Berkeley 415 l 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 7/9 l 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.

Q)NSULTIN(ACTIITIES(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 Patent4,406,766 (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 W02011/061728A1, 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 z 1000).

1. D. D. Macdonald and G. R. Engelhardt, "Predictive Modeling of Corrosion". In:

Richardson J A et al. (eds.), Sorrier 's 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 A1-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", Corrode.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".Corrode. 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. Electro chem.Soc.,148(9), B343 - B347 (2001).

15. D. D. Macdonald, "Stress Corrosion Cracking in Reactor Coolant Circuits - An Electrochemist's 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",Corrode. 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. Carpen, and D. D. Macdonald, (2019). Erratum: Kinetic properties of the passive film on copper in the presence ofsulfate-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), 1103-1 l II (2021).

27. F Carotti, E Liu, DD Macdonald, RO Scarlat, An electrochemical study of hydrogen in molten 2LiF-BeF; (FLiBe) with addition ofLiH, 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 mail<ind".

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, 201 l.

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.

/44 8"/64/ M Signed. Dig by D. Macdonald.

September 13,2023.

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APPENDIX B: Reference List

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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 .

79 Macdonald, D.D., 2023. "The Role of Determinism in the Prediction of Corrosion Damage.7 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 201 l. PG&E Letter DCL-l1-136 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 Geffen to NRC re: Docket No. 50-275, OL-DPR-80, Diablo Canyon Unit l, Revision to the Unit 1 Reactor Vessel Material Surveillance Program Withdrawal Schedule (May 15, 2023) (ADAMS Accession No. MLl4230A6l8),

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

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

Spencer, B.W., Back ran, M., Williams, P. T., Hofinan, 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).

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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, 201 l)

(ADAMS Accession No. MLl l 1600295).

U.S. NRC 2015. Letter from Michael T. Markey, NRC, to Edward D. Halpin, PG&E, Re: Diablo Canyon Power Plant, Unit No. l - 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 l Reactor Vessel Radiation Surveillance Program (Jan.

2003).https://mothersforpeace.org/wp-content/uploads/2023/09/2003-WCAP-l5958-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).

https1//mothersforpeace.org/wp-content/uploads/2023/09/2011-WCAP- 17315-NP-Rev.-0.pdf.

2

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ATTACHMENT 2

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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 l 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 l 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 l 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 l reactor pressure vessel since 2003, and those results were inconclusive.

4. I believe PG&E's 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

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UNITED STATES oF AMERICA NUCLEARREGULATORY 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 I and 2

T r v STANDING DECLARATION OF LL CY JANE SwAnson IN SUPPORT OF REQUEST F()R HEARING AND EMERGENCY ORDERSr 1 L

Under penalty of perjury, Lucy Jane Swanson declares as follows:

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

2. I live at 313 Presidio Place, San Luis Obispo, CA 93401. My home is located within the 50-mile ingestion pathway zone of Diablo Canyon Unit l and Unit 2 nuclear reactors.

3. lt 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 t`or withdrawing Capsule B until as late as spring 2025. PG&E has not collected or tested any samples from the Unit l reactor pressure vessel since 2003, and those results were inconclusive.

4. I believe PG&E's ongoing lack of knowledge regarding the condition of the Unit l 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 orders by the Commissioners.

____Q ,864/ a~\\4¢o>~.-

Lucy 'S l Swanson Septembers 2023

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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORETHE 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 J7LL ZAMEK IN SUPPORT OF REQUEST FOR HEARING AND EMERGENCY ORDERS

Under penalty of perjury, Jill ZamEk declares as follows:

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

2. I live at 1123 Flora Road, Arroyo Grande, California. My home is located within eighteen miles of 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 l 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. believe PG&E's 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 SLOMPP and FOE to represent my interests by seeking a hearing and emergency orders by the Commissioners.

36/-

Ji1I'2lamEk September 8, 2023

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ATTACHMENT 3

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ATTACHMENT 3 -TABLE OF ESTIMATED DATES OF CAPSULE WITHDRAWALS

RF Date Capsule withdrawn, scheduled, EFPY EFPY (Projected) o attempted, or skipped (Actual when removed) 1 1986(?) Capsule S removed and testedlal 1.25(H) 2 3

4 5 1992(?) Capsule Y removed and testedla) 5.86l"l_

6 7

8 9

10 .

11 2002 'Capsule V removed and testedlal 14.3(a) 12.9(b>

12  ? Capsules C and D removed but not 15.9(a) 14.8(b) to$ted(a) 13 14 2007 Capsule B scheduledlbl 19.2(b) 15 2009 Capsule B re-scheduledl°) 20.7(c) 16 2010 Ca B sch9 l removal failedld) 17 2012 Capsule B re-scheduledldl 18 19 20 21 22 23 5/2022 33.0(a)Capsule B scheduled but removal skipped because PG&E had w/drawn license renewal application e' 24 10/2023 Capsule B re-scheduledlel 33.58(e) 25 Spring 2025 Capsule B re-scheduledlel 34.97(e)

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ATTACHMENT 3 -TABLE OF ESTIMATED DATES OF CAPSULE WITHDRAWALS

Endnotes:

in) 2023.05.13 PG&E letter DCL-23-038, Encl. 2 12021 UFSAR Rev. 7, Table 5.2-22)

(b) 1992.03.31 PG&E Application for Supp. Surveillance Program (c) 2006.07.17 NRC Safety Evaluation for Low Power Testing Recapture License Amendment at 5.

(d) 2010.10.29 NRC Safety Evaluation (e) 2023.07.20 NRC Safety Evaluation

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Pacific Gas and Electric Eu4pmw'

Paula Gerfen Diablo Canyon Power Plant Senior Vine President andP.[J. Box 56 Chief Nuclear SUfficerAvila Beach, CA 93424

805.545.4596 Fax:805.545.4234 Paula.Gerfen@PQ&com

PG&E Letter DCL-23-038

U.S. Nuclear Regulatory Commission 10 CFR 50, App.H Attn: Document Control Desk Washington, DC 20555-0001

Docket No. 50-275, OL-DPR-80 Diablo Canyon Unit 1 Revision to the Unit 1 Reactor vessel Material Surveillance Program Withdrawal Schedule

References:

1. NRC Letter, "Diablo Canyon Power Plant, Unit No. 1: Safety Evaluation for Request to Revise the Reactor Vessel Material Surveillance Program Withdrawal Schedule (TAC ME7615)," dated March 2, 2012 (ML120330497)

2. NRC Letter, "Response to Request to Withdraw the Diablo Canyon Power Plant, Unit Nos. 1 and 2, License Renewal Application," dated April 16, 2018 (ML18093A115)

3. PG&E Letter DCL-22-085, "Request to Resume Review of the Diablo Canyon Power Plant License Renewal Application or, Alternatively, for an Exemption from 10 CFR 2.109(b), Concerning a Timely Renewal Application," dated October 31, 2022 (ML22304A691)

Dear Commissioners and Staff:

Pursuant to 10 CFR 50, Appendix H, Section lll.B.3, Pacific Gas and Electric Company (PG&E) hereby requests approval for a revision to the Unit 1 reactor vessel material surveillance program withdrawal schedule.

In Reference 1, NRC approved the withdrawal schedule for the Diablo Canyon Power Plant (DCPP) Unit 1 Capsule B during the Unit 1 Twenty-Third Refueling Outage (1 R23) conducted in 2022. The approved withdrawal schedule supported obtaining reactor pressure vessel fluence data for the period of extended operation which is only applicable to license renewal. Subsequent to NRC's approval, by Reference 2, NRC granted PG&E's request to withdraw the DCPP license renewal application. Because the Unit 1 Capsule B removal was to support DCPP license renewal, it was not withdrawn in 1 R23.

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Document Control Desk PG&E Letter DCL-23-038 Page 2

In September 2022, the California Governor signed Senate Bill No. 846 (Dodd),

which reversed the prior California Public Utilities Commission decision approving the retirement of DCPP by the expiration of the current operating licenses. By Reference 3, PG&E notified the NRC of the intent to submit a new DCPP license renewal application no later than December 2023. Consequently, Unit 1 reactor pressure vessel fluence data is now needed for license renewal and PG&E requests revision to the Unit 1 reactor vessel material surveillance program withdrawal schedule to allow withdrawal of Capsule B during the Unit 1 Twenty-Fourth Refueling Outage (Fall 2023) or Unit 1 Twenty-Fifth Refueling Outage (Spring 2025).

DCPP has withdrawn and tested three capsules from Unit 1 that meet the three recommendations of ASTM E185-70 and the approved supplemental surveillance capsule withdrawal changes listed in NRC staff Safety Evaluation dated September 4, 1992. The withdrawal and testing of Capsule V during the Unit 1 Eleventh Refueling Outage fulfilled the third and final recommendation of ASTM E185-70 for the current DCPP Unit 1 operating license. Therefore, the proposed removal of Capsule B does not deviate from DCPP's current reactor vessel materials surveillance program requirements. DCPP installed the Unit 1 Capsule B in the vessel in the Unit 1 Fifth Refueling Outage at a vessel exposure of 5.86 effective full power years (EFPY). The proposed withdrawal schedule allows Capsule B to be withdrawn at approximately 33.58 - 34.97 plant EFPY or a fluence of 3.39E+19 - 3.56E+19 n/cm2 at the capsule (approximately 96.19 - 101 .01 EFPY).

This will provide data for PG&E and the NRC to review for license renewal, and can be utilized by the industry and Electric Power Research Institute for further analysis.

provides a description and assessment of the proposed change to the reactor vessel material surveillance program withdrawal schedule.

The Unit 1 reactor material surveillance program withdrawal schedule is provided in the DCPP Updated Final Safety Analysis Report (UFSAR) Table 5.2-22. Enclosure 2 provides a mark-up of the affected DCPP UFSAR page that will be incorporated into the DCPP UFSAR upon approval of this schedule revision.

To support implementation of the revised withdrawal schedule, PG&E requests approval of this proposed change by August 31, 2023. The request for expedient turnaround is to ensure PG&E will have adequate time to mobilize the required special resources and to secure support for removal of the Unit 1 Capsule B in the Unit 1 Twenty-Fourth Refueling Outage, which is scheduled to begin in October 2023.

PG&E makes no new or revised regulatory commitments (as defined by NEI 99-04) in this letter.

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Document Control Desk PG&E Letter DCL-23-038 Page 3

If you have any questions, please contact Mr. Philippe Soenen at (805) 459-3701 .

Sincerely,

f:._ A/ t

Paula Gerfen May 15, 2023 Senior Vice President and Chief Nuclear Officer Date

Enclosures cc: Diablo Distribution cc/enc: Lauren Gibson, License Renewal Branch Chief Brian K. Harris, NRC Senior Project Manager Mahdi o. Hayes, NRC Senior Resident Inspector Samson S. Lee, NRC Senior Project Manager Robert J. Lewis, NRC Acting Region IV Administrator

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Enclosure 1 PG&E Letter DCL-23-038 DESCRIPTION AND ASSESSMENT

1.0 BACKGROUND

Appendix H to Title 10 of the Code of Federal Regulations (10 CFR) Part 50 (Reference 1) requires a material surveillance program to monitor changes in the fracture toughness properties of ferrite material in the reactor vessel beltline region that result from exposure of these materials to neutron irradiation and the thermal environment. Under this program, fracture toughness test data are obtained and analyzed from material specimens exposed in surveillance capsules that are withdrawn periodically from the reactor vessel. Test results must be reported to the NRC within 18 months of the date of the capsule withdrawal. In addition, Section lll.B.3 of Appendix H to 10 CFR 50 requires the NRC to approve the capsule withdrawal schedule prior to implementation.

The design of the surveillance program and the withdrawal schedule must meet the requirements of the edition of ASTM E185 that is current on the issue date of the ASME Code to which the reactor vessel was purchased. The ASTM E185 version of record for Diablo Canyon Power Plant (DCPP) Unit 1 is ASTM E185-70 (Reference 2). A Unit 1 supplemental reactor vessel surveillance capsule program schedule was requested by DCPP in Pacific Gas and Electric Company (PG&E) Letter DCL-92-072, dated March 31, 1992 (Reference 3). The NRC approved this program in a letter to PG&E dated September 4, 1992 (Reference 4).

The schedule to withdraw Capsule B during the Unit 1 Twenty-Third Refueling Outage (1R23) was requested by DCPP in 2011 (Reference 5) and approved by the NRC in 2012 (Reference 6). In its approval letter, NRC noted the "requirements of Appendix H and ASTM E185-70 regarding the initial license period (the first 40 years of service for DCPP, Unit 1) have already been met."

Subsequent to NRC's 2012 approval, the NRC granted PG&E's request to withdraw the DCPP license renewal application (Reference 7). Therefore, Unit 1 Capsule B, which was only scheduled for removal to support license renewal, was not removed in 1 R23 and was designated as a "standby" capsule.

In September 2022, the California Governor signed Senate Bill No. 846 (Dodd),

which reversed the prior California Public Utilities Commission decision approving the retirement of DCPP by the expiration of the current operating licenses. In October 2022, PG&E notified NRC of the intent to submit a new DCPP license renewal application no later than December 2023 (Reference 8).

Consequently, Unit 1 reactor pressure vessel fluence data is now needed for the period of extended operation for license renewal. The objective of the change to the withdrawal schedule is to align with NUREG-1801, Revision 2, "Generic Aging Lessons Learned (GALL) Report," December 2010 (Reference 9) for a capsule withdrawal at a neutron fluence level exceeding, but not greater than twice, the peak reactor vessel neutron fluence at 60 years of operation.

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Enclosure 1 PG&E Letter DCL-23-038

2.0 DESCRIPTION

OF CHANGES TO THE REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM WITHDRAWAL SCHEDULE

The Unit 1 Reactor Vessel Material Surveillance Program withdrawal schedule is located in the DCPP Updated Final Safety Analysis Report (UFSAR)

(Reference 10). The proposed change revises the Unit 1 Capsule B withdrawal schedule from a "standby" capsule to withdrawn at approximately 33.58-34.97 plant effective full power years (EFPY).

Unit 1 Capsule B was installed in the vessel in the Unit 1 Fifth Refueling Outage at a vessel exposure of 5.86 EFPY per UFSAR Table 5.2-22, note (b). The lead factor for Capsule B at the 40 degree location is 3.47 per UFSAR Table 5.2-22.

The equivalent exposure on Capsule B if it is withdrawn during the Unit 1 Twenty-Fourth Refueling Outage (1 R24) is projected to be (33.58 - 5.86) X 3.47 96.19 EFPY (neutron fluence of 3.39E+19 n/cm2) or during the Unit 1 Twenty-Fifth Refueling Outage (1R25) is projected to be (34.97 - 5.86) X 3.47 = 101 .01 EFPY (neutron fluence of 3.56E+19 n/cm2).

3.0 ANALYSIS

10 CFR Part 50

The staff approved the revised capsule withdrawal schedule of 1 R23 in its safety evaluation dated March 2, 2012 (Reference 6). The staff concluded that removing Unit 1 Capsule B during 1 R23 would have met the expectations of NUREG-1801 and the requirements of 10 CFR Part 50, Appendix H. PG&E is requesting approval to revise this surveillance capsule removal until 1 R24 or 1 R25 to support data acquisition for the period of extended operation.

PG&E is asking for approval of withdrawal in a two refueling outage window based on previous experience with attempting to remove Unit 1 Capsule B.

Specifically, in 2010, PG&E could not remove Capsule B due to difficulty with the access plug on the reactor core barrel flange that provides access to the capsule (Reference 11). In 1 R24, PG&E plans to attempt Capsule B withdrawal.

However, in the event Capsule B cannot be removed, PG&E is conducting parallel contingency planning for alternate means of removal in 1R25. The core barrel is scheduled for removal in 1 R25, which allows alternate access paths to the capsule. PG&E concludes that either withdrawal schedule meets the expectations of NUREG-1801, Revision 2, and the requirements of 10 CFR Part 50, Appendix H.

10 CFR Part 54

PG&E plans to submit a new DCPP license renewal application no later than December 2023 (Reference 8) using the expectations of NUREG-1801, Revision 2 (Reference 9). NUREG-1801, Revision 2, Section Xl.M31, Reactor Vessel Surveillance, states the "program withdraws one capsule at an outage in which the capsule receives a neutron fluence of between one and two times the peak

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Enclosure 1 PG&E Letter DCL-23-038 reactor vessel wall neutron fluence at the end of the period of extended operation." The maximum fluence calculated for the license renewal period at DCPP Unit 1 is 2.01E+19 n/cm2(E>1 MeV) (References 6 and 12). With the proposed withdrawal schedule, the fluence for Unit 1 Capsule B will be 1.69x to 1.77x the maximum fluence expectations for the period of extended operation.

Therefore, the proposed withdrawal schedule is consistent with the expectations laid out in ASTM E185-82 and NUREG-1801, Revision 2.

The DCPP Unit 2 surveillance capsule withdrawal program currently meets the expectations of NUREG-1801, Revision 2, therefore, no changes in the DCPP Unit 2 surveillance capsule withdrawal program are needed.

4.0 REFERENCES

1. Code of Federal Regulations, Title 10, Part 50, Appendix H, "Reactor vessel Material Surveillance Program Requirements."

2. American Society of Testing and Materials, "Standard Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels,"

ASTM E185-70.

3. PG&E Letter DCL-92-072, "Supplemental Reactor Vessel Radiation Surveillance Program," dated March 31, 1992 (ML163416505).

4. NRC Letter, "Evaluation of Diablo Canyon Unit 1 Supplemental Reactor Vessel Radiation Surveillance Program (TAC No. M83285)," dated September 4, 1992 (ML16341G686).

5. PG&E Letter DCL-11-122, "Revision to the Unit 1 Reactor vessel Material Surveillance Program Withdrawal Schedule," dated November 21, 2011 (ML113260072).

6. NRC Letter, "Diablo Canyon Power Plant, Unit No. 1: Safety Evaluation for Request to Revise the Reactor Vessel Material Surveillance Program Withdrawal Schedule (TAC ME7615)," dated March 2, 2012 (ML120330497).

7. NRC Letter, "Response to Request to Withdraw the Diablo Canyon Power Plant, Unit Nos. 1 and 2, License Renewal Application," dated April 16, 2018 (ML18093A115).

8. PG&E Letter DCL-22-085, "Request to Resume Review of the Diablo Canyon Power Plant License Renewal Application or, Alternatively, for an Exemption from 10 CFR 2.109(b), Concerning a Timely Renewal Application," dated October 31, 2022 (ML22304A691).

9. NUREG-1801, "Generic Aging Lessons Learned (GALL) Report." Revision 2, December 2010 (ML103490041 ).

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Enclosure 1 PG&E Letter DCL-23-038

10. Diablo Canyon Power Plant Updated Final Safety Analysis Report, Revision 26, September 2021 .

11. PG&E Letter DCL-10-141, "Revision to the Unit 1 Reactor vessel Material Surveillance Program Withdrawal Schedule," dated October 25, 2010 (ML102990079).

12. Westinghouse Report WCAP-18655-NP, "Ex-Vessel Neutron Dosimetry Program for Diablo Canyon Unit 1 Cycle 22," dated August 2021 .

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Enclosure 2 PG&E Letter DCL-23-038

Mark-up of DCPP UFSAR Table 5.2-22

REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM WITHDRAWAL SCHEDULE

UNIT 1

Lead Fluence at Capsule Removal CapsuIe(fx9l Location Factor(d) Center (n/cm2)ldl Time (Plant EFPY)(a)

S 320° 3.48 2.83E+18 1.25 (Tested,1R1)

Y 40° 3.45 1.05E+19 5.86 (Tested, 1 R5)

T 140° 3.45 1.05E+19 5.86 (Removed, 1 R5)

Z 220° 3.45 1.05E+19 5.86 (Removed, 1 R5)

V 320° 2.26 1.36E+19 14.3 (Tested 1R11)

C(b) 140° 3.47 1.22E+19 15.9 (Removed 1R12)

D(b) 220° 3.47 1.22E+19 15.9 (Removed 1R12)

B(b) 40° 3.47 StaF1dby(3.39E+19 - Standby33.58-34.97 3.56E+19 projected) (Planned 1R24/1R25)

A(b) 184° 1.32 Standby Standby U 356° 1.24 Standby Standby X 176°40 1.24 Standby Standby W 1.24 Standby Standby

UNIT 2

Lead Fluence at Capsule Removal Capsule Location Factor(d) Center (n/cm2)(dl Time (EFPY)(a)

U 56° 5.20 3.30E+18 1.02 (Tested, 2R1)

X 236° 5.39 9.06E+18 3.16 (Tested, 2R3)

YW(e) 238.5° 4.56 1.53E+19 7.08 (Tested, 2R6) 124° 5.35 2.78E+19 11.49 (Removed, 2R9)

V(e) 58.5° 4.57 2.38E+19 11.49 (Tested, 2R9) 2(e) 304° 5.35 2.78E+19 11.49 (Removed, 2R9)

(a) Approximate full power years from plant startup.

(b) Four supplemental capsules installed at 5.86 EFPY (EOC5).

(C) Deleted in Revision 16.

(d) Approximate values taken from WCAP-17299-NP (Rev. 0) or WCAP-18655-NP (Rev. 0) for Units 1 and 2.

(e) Capsule EFPY for Unit 2 capsules removed in 2R9, W = 61 .5, v = 52.5, and Z = 61 .5 (f) Unit 1 capsules T, U, W, X, and Z are Type 1 (base metal only)

(9) Unit 1 capsules S, V, and Y are Type 2 (base metal and weld)

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Enclosure 2 PG&E Letter DCL-23-038

Mark-up of DCPP UFSAR Table 5.2-22

REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM WITHDRAWAL SCHEDULE

UNIT 1

Lead Fluence at Capsule Removal Capsule(fl(9) Location Factored) Center (n/cm2)ld) Time (Plant EFPY)la)

S 320° 3.48 2.83E+18 1.25 (Tested,1 R1 )

Y 40° 3.45 1.05E+19 5.86 (Tested, 1 R5)

T 140° 3.45 1.05E+19 5.86 (Removed, 1 R5)

Z 220° 3.45 1.05E+19 5.86 (Removed, 1 R5)

V 320° 2.26 1.36E+19 14.3 (Tested 1R11)

C(b) 140° 3.47 1.22E+19 15.9 (Removed 1R12)

D(b) 220° 3.47 1.22E+19 15.9 (Removed 1R12)

B(b) 40° 3.47 Standby(3.39E+19 - Sta9eUey33.58-34.97 3.56E+19 projected) (Planned 1 R24/1R25)

A(b) 184° 1.32 Standby Standby U 356° 1.24 Standby Standby X 176°40 1.24 Standby Standby W 1.24 Standby Standby

UNIT 2

Lead Fluence at Capsule Removal Capsule Location Factored) Center (n/cm2)ld) Time (EFPY)(al

U 56° 5.20 3.30E+18 1.02 (Tested, 2R1)

X 236° 5.39 9.06E+18 3.16 (Tested, 2R3)

YW(e) 238.5° 4.56 1.53E+19 7.08 (Tested, 2R6) 124° 5.35 2.78E+19 11.49 (Removed, 2R9)

V() 4.57 2.38E+19 11.49 (Tested, 2R9) 2(6) 58.5°304°5.352.78E+19 11.49 (Removed, 2R9)

(a) Approximate full power years from plant startup.

(b) Four supplemental capsules installed at 5.86 EFPY (EOC5).

(C) Deleted in Revision 16.

(d) Approximate values taken from WCAP-17299-NP (Rev. 0) or WCAP-18655-NP (Rev. 0) for Units 1 and 2.

(e) Capsule EFPY for Unit 2 capsules removed in 2R9, W = 61.5, v : 52.5, and Z = 61 .5 (f) Unit 1 capsules T, U, W, X, and Z are Type 1 (base metal only)

(9) Unit 1 capsules S, v, and Y are Type 2 (base metal and weld)

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Pacific Gas and Electric Company James B. Becker Diablo Canyon Power Plant m Site Vice PresidentMail Code 104/5/601

p. o. Box 5B Avila Beach, CA 93424

805.545.3462 November 21, 2011 Internal: 881.3462 Fax: 805.545.5445

PG&E Letter DCL-11-122

U.S. Nuclear Regulatory Commission 10 CFR 50,App. H Attn: Document Control Desk Washington, DC 20555

Docket No. 50-275, OL-DPR-80 Diablo Canyon Unit 1 Revision to the Unit 1 Reactor Vessel Material Surveillance Program Withdrawal Schedule

Dear Commissioners and Staff:

Pursuant to 10 CFR 50, Appendix H, Section Ill.B.3, Pacific Gas and Electric Company (PG&E) hereby requests approval for a revision to the Unit 1 reactor vessel material surveillance program withdrawal schedule.

The Unit 1 reactor material surveillance program withdrawal schedule is provided in the Diablo Canyon Power Plant (DCPP) Final Safety Analysis Report Update (FSARU) Table 5.2-22 and in the DCPP License Renewal Application (LRA),

Appendix 82.1.15, Reactor vessel Surveillance Aging Management Program. The NRC approved withdrawal schedule for Capsule B is during the Unit 1 Seventeenth Refueling Outage (1 R17) and is cited in the Safety Evaluation Report Related to the License Renewal of Diablo Canyon Nuclear Power Plant. DCPP is currently operating in Cycle 17 and 1 R17 is scheduled for May 2012.

EPRI MRP-326, Draft E, "Materials Reliability Program: Coordinated PWR Reactor Vessel Surveillance Program, (CRVSP)", has recommended that Diablo Canyon Unit 1 delay the removal and testing of Capsule B until approximately twice the 60-year fluence. This is estimated to occur during the Unit 1 Twenty Third Refueling Outage (1 R23), which is scheduled for May 2022. The recommended delay has been proposed to support data acquisition for the EPRI CRVSP.

Therefore, PG&E requests revision to the Unit 1 reactor vessel material surveillance program withdrawal schedule to allow withdrawal of Capsule B during 1 R23.

DCPP has withdrawn and tested three capsules from Unit 1 that meet the three recommendations of ASTM E 185-70 and the approved supplemental surveillance capsule withdrawal changes listed in NRC staff Safety Evaluation dated September 4, 1992. The withdrawal and testing of Capsule V during the Unit 1 Eleventh Refueling Outage fulfilled the third and final recommendation of

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Document Control Desk PG&E Letter DCL-11-122 m8 November 21, 2011 Page 2

ASTM E 185-70 for the current DCPP Unit 1 operating license. Therefore, the proposed delayed removal of Capsule B does not deviate from DCPP's current reactor vessel materials surveillance program requirements. DCPP installed the Unit 1 Capsule B in the vessel in the Unit 1 Fifth Refueling Outage at a vessel exposure of 5.86 effective full power years (EFPY). The change in withdrawal schedule allows Capsule B to be withdrawn at a fluence of approximately 93.9 EFPY for the reactor pressure vessel. This will provide reactor pressure vessel fluence data for the period of extended operation for license renewal.

Enclosure 1 provides a description and assessment of the proposed change to the reactor vessel material surveillance program withdrawal schedule. Enclosure 2 provides a mark-up of the affected DCPP FSARU page, If the request to defer is approved, the DCPP LRA will be updated in the 2012 annual update, in accordance with 10 CFR 54.21(b).

To support implementation of the revised withdrawal schedule, PG&E requests approval of this proposed change by December 31, 2011. The request for fast turnaround is to assure that PG&E will have adequate time to mobilize the required special resources and to secure support for machining, repair, and replacement of the stuck cap on the Unit 1 surveillance coupon "B" if the request to defer is denied.

PG&E makes commitments to revise DCPP FSARU Table 5.2-22 to incorporate the change within 60 days of the NRC approval and update the DCPP LRA in the 2012 annual update upon staff approval of this deferral.

If you have any questions, please contact Mr. Tom Baldwin at (805) 545-4720.

Sincerely,\\ \\ ,n

\\ I\\r"'.

James R. Becker Site Vice President

rntt/4231 Enclosures cc: Diablo Distribution cc/enc: Elmo E. Collins, NRC Regional Administrator, Region IV Nathanial B. Ferrer, NRC Project Manager, License Renewal Michael S. Peck, NRC Senior Resident Inspector Alan B. Wang, NRC Project Manager, Nuclear Reactor Regulation

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=

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Enclosure 1 PG&E Letter DCL-11-122

DESCRIPTION AND ASSESSMENT

1.0 BACKGROUND

Appendix H to 10 CFR 50 requires a material surveillance program to monitor changes in the fracture toughness properties of ferrite material in P the reactor vessel beltline region that result from exposure of these materials to neutron irradiation and the thermal environment. Under this program, fracture toughness test data are obtained and analyzed from material specimens exposed in surveillance capsules that are withdrawn periodically from the reactor vessel. Test results must be reported to the NRC within one year of the date of the capsule withdrawal. In addition, Section lll.B.3 of Appendix H to 10 CFR 50 requires the NRC to approve 1

3 the capsule withdrawal schedule prior to implementation. 1

§ The design of the surveillance program and the withdrawal schedule must 1

meet the requirements of the edition of ASTM E 185 that is current on the issue date of the ASME Code to which the reactor vessel was purchased. 3 The ASTM E 185 version of record for Diablo Canyon Power Plant (DCPP) Unit 1 is ASTM E 185-70. A Unit 1 supplemental reactor vessel surveillance capsule program schedule was requested by DCPP in Pacific Gas and Electric Company (PG&E) Letter DCL-92-072, dated March 31 ,

1992. The NRC approved this program in a letter to PG&E dated September 4, 1992, "Evaluation of Diablo Canyon Unit 1 Supplemental Reactor Vessel Radiation Surveillance Program (TAC No. M83285)."

The schedule to withdraw Capsule B during the Unit 1 Seventeenth I Refueling Outage (1R17) was requested by DCPP in PG&E Letter

1 DCL-10-141, dated October 25, 2010. The NRC approved this schedule revision by letter to PG&E dated October 29, 2010, "Diablo Canyon Power Plant, Unit No. 1 -Approval of Proposed Reactor Vessel Material Surveillance Program Withdrawal Schedule (TAC No. ME4924)."

However, PG&E requests approval to defer this schedule to the Unitl Twenty Third Refueling Outage (1 R23) as recommended in EPRI MRP-326, Draft E, "Materials Reliability Program: Coordinated PWR Reactor Vessel Surveillance Program, (CRVSP)." The objective of the CRVSP is to manage the withdrawal schedules of remaining pressurized water reactor (PWR) surveillance capsules to increase the fluence levels of the capsules at withdrawal and to fill the high fluence irradiated Charpy data gaps in the PWR surveillance capsule database that is used to develop 1 embrittlement correlations.

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Enclosure 1 PG&E Letter DCL-11-122

1

2.0 DESCRIPTION

OF CHANGES TO THE REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM WITHDRAWAL SCHEDULE The Unit t Reactor Material Surveillance Program withdrawal schedule is III located in the DCPP Final Safety Analysis Report Update (FSARU). The

3 proposed change revises the schedule to change the removal time for Capsule B from 23.2 effective full power years (EFPY) to 33 EFPY.

Unit 1 Capsule B was installed in the vessel in the Unit 1 Fifth Refueling Outage at a vessel exposure of 5.86 EFPY per FSARU Table 5.2-22. The lead factor for Capsule B at the 40 degree location is 3.46 per FSARU Table 5.2-22. The equivalent exposure on Capsule B at its withdrawal during 1 R23 is projected to be (33 - 5.86) x 3.46 = 93.9 EFPY.

I 3.0 ANALYSIS

10 CFR Part 50 i

The staff approved the revised capsule withdrawal schedule of 1R17 in its safety evaluation dated October 29, 2010. The staff concluded that removing Surveillance Capsule B during the seventeenth refueling outage will meet the expectations of NUREG-1801 and the requirements of I1 10 CFR Part 50, Appendix H. PG&E is requesting approval to defer this surveillance until 1 R23 to support data acquisition for the EPRl CRVSP.

10 CFR Part 54

PG&E submitted its License Renewal Application for DCPP Unit 1 and 2 in PG&E Letter DCL-09-079, dated November 23, 2009. In the Safety Evaluation Report (SER) Related to the License Renewal of Diablo Canyon Nuclear Power Plant, June 2, 2011, the staff reviewed the DCPP aging management programs using the expectations of NUREG-1801 ,

Revision 1. The staff concluded the Reactor Vessel Surveillance Program demonstrated that the effects of aging will be adequately managed so that the intended functions will be maintained consistent with the current licensing basis for the period of extended operation.

The SER concluded the equivalent EFPY value is just over one times the end of license extended (EOLE) reactor vessel fluence (54 EFPY), this meets the expectations of NUREG-1801, Revision 1, Xl.M31, Criterion 6.

The proposed equivalent exposure on Capsule B is projected to be 93.9 EFPY, exceeding the 60-year fluence of NUREG-1801, Revision 1, Xl.M31, Criterion 6. Criterion 7 allows the use of alternative dosimetry to monitor neutron fluence as part of the aging management program for

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Enclosure 1 1

i PG&E Letter DCL-11-122

reactor vessel neutron embrittiement. Unit 1 and Unit 2 are equipped with ex-vessel dosimetry to monitor neutron fluence. Therefore, the exposure of 93.9 EFPY is consistent with the expectations of NUREG-1801 , i Revision 1 .

A removal time of 33 EFPY for Capsule B, corresponding to 93.9 EFPY 3

1 will also continue to satisfy the expectations of NUREG-1801, Revision 2, which is to have a capsule with fluence exposure between one and two times the peak reactor vessel wall neutron fluence at the end of the period of license extension fluence (93.9 EFPY fluence).

The DCPP Unit 2 surveillance capsule withdrawal program currently meets the expectations of NUREG-1801, therefore, no changes in the DCPP Unit 2 surveillance capsule withdrawal program are needed.

4.0 REFERENCES

1. Code of Federal Regulations, Title 10, Part 50, Appendix H, "Reactor Vessel Material Surveillance Program Requirements."

2. American Society of Testing and Materials, "Standard Recommended 3

3 Practice for Surveillance Tests for Nuclear Reactor Vessels," ..........................................................................

ASTM E 185-70.

3. Diablo Canyon Final Safety Analysis Report Update (FSARU),

Revision 19, May 2010.

11

4. NUREG-1801, "Generic Aging Lessons Learned (GALL),

Revision 1, 2005.

5. NUREG-1801, "Generic Aging Lessons Learned (GALL)."

Revision 2, 2010.

6. PG&E Letter DCL-09-079, "License Renewal Application," dated November 23, 2009.

7. EPRl MRP-326, "Materials Reliability Program: Coordinated PWR Reactor Vessel Surveillance Program (CRVSP)," Draft E, dated October 2011 .

8. "Safety Evaluation Report Related to the License Renewal of Diablo Canyon Nuclear Power Plant, Units 1 and 2," dated June 2, 2011 .

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Enclosure 2 PG&E Letter DCL-11-122

MARK-UP of FSARU TABLE 5.2-22

REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM WITHDRAWAL SCHEDULE

UNIT 1 Lead Removal Capsule(fllgl Location Factorldl Time (EFpY)(2)

S 320° 3.46 1.25 (Tested,1R1)

Y 40° 3.44 5.86 (Tested, 1 R5)

T 140° 3.44 5.86 (Removed, 1 R5)

Z 220° 3.44 5.86 (Removed, 1 R5)

V 320° 2.26 14.3 (Tested 1R11)

C(b) 140° 3.46 15.9 (Removed 1R12)

D(b) 220° 3.46 15.9 (Removed 1 R12)

B(b) 40° 3.46 .2 33 (1 R23) I A(b) 184° 1.31 Standby U 356° 1.28 Standby X 176°4o 1.28 Standby W 1.28 Standby UNIT 2 Lead Removal Capsule Location Factored) Time lEFPYl(a)

U 56° 5.15 1.02(Tested,2R1 )

X 236° 5.40 3.16 (Tested, 2R3)

YW() 238.5° 4.58 7.08 (Tested, 2R6)

V() 124° 5.26 11 .49 (Removed, 2R9)

Z(8) 58.5° 4.58 11 .49 (Tested, 2R9) 304° 5.26 11.49 (Removed, 2R9)

(a) Approximate full power years from plant startup (b) Four supplemental capsules installed at 5.86 EFPY (EOC5)

(C) Deleted in Revision 16 (d) Approximate lead factors taken from WCAP-15958 (Rev. 0) and WCAP-15423 (Rev. 0) for Units 1 and 2, respectively (e) EFPY for Unit 2 capsules removed in 2R9, W = 60.4, V = 52.6, and Z = 60.4 (f) Unit 1 capsules T, U, W, X, and Z are Type 1 (base metal only)

(9) Unit 1 capsules S, V, and Y are Type 2 (base metal and weld)

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Pacific Gas and m Electric Company James R. BeckerDiablo Canyon PowerPlant Site Vice PresidentMail Code104/5/501 R 0. Box 56 Avila Beach, CA 93424

805.545.3452 October 25, 2010 Internal: 691.3462 Fax: 805.545.5445 PG&E Letter DCL-10-141 10 CFR 50, APP. H U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

Docket No. 50-275, OL-DPR-80 Diablo Canyon Unit 1 Revision to the Unit 1 Reactor Vessel Material Surveillance Program Withdrawal Schedule

Dear Commissioners and Staff:

Pursuant to 10 CFR 50, Appendix H, Section lll.B.3, Pacific Gas and Electric Company (PG&E) hereby requests approval for a revision to the Unit 1 reactor vessel material surveillance program withdrawal schedule.

The Unit 1 reactor material surveillance program withdrawal schedule is provided in the Diablo Canyon Power Plant (DCPP) Final Safety Analysis Report Update (FSARU) Table 5.2-22. The proposed change would revise that schedule to reflect withdrawal of Capsule B during the Unit 1 Seventeenth Refueling Outage, which is scheduled to begin May 1, 2012. The NRC approved withdrawal schedule for Capsule B is during the Unit 1 Sixteenth Refueling Outage (1R16). DCPP is currently in 1R16 and the schedule for entry into mode 2 is 1900 hours0.022 days <br />0.528 hours <br />0.00314 weeks <br />7.2295e-4 months <br /> on November 2, 2010, which is the next time the specimen would be subjected to additional Huence.

During 1R16, refueling personnel have not been able to remove the Capsule B access plug on the reactor core barrel flange. Removal of the access plug is required to gain access to the specimen capsule. Normally the plug is held in place by its own weight (approximately five pounds). Refueling personnel have applied over 2,000 pounds of force in attempts to remove the plug, The application of additional extraction force may result in damage and prevent the plug from being reinserted after the capsule is removed. If the plug itself is damaged or the hole is deformed, the vendor does not have a spare access plug and is not prepared to machine the hole in the flange of the core barrel during the current refueling outage.

There is also a concern for introducing foreign material to the reactor vessel if personnel damage the plug or tool while attempting to remove the plug.

A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway Comanche Peak >> Diablo.Canyon

• Palo Verde c San Onofre

• South Texas Project Wolf Creek ER-140 Case: 23-3884, 03/25/2024, DktEntry: 23.2, Page 108 of 222

Document Control Desk PG&E Letter DCL-10-141 an8 October 25, 2010 Page 2

Therefore, PG&E requests revision to the Unit 1 reactor vessel material surveillance program withdrawal schedule to allow withdrawal of Capsule B during the Unit 1 Seventeenth Refueling Outage (1R17). Removal of Capsule B during 1R17 will ensure adequate time to allow for the appropriate tooling, materials, and contingency plans to be in place to remove and reinsert or replace the Capsule B access plug.

DCPP has withdrawn and tested three capsules from Unit 1 that meet the three recommendations of ASTM E 185-70 and the approved supplemental surveillance capsule withdrawal changes listed in NRC staff Safety Evaluation dated September 4, 1992. The withdrawal and testing of Capsule v during the Unit 1 Eleventh Refueling Outage fulfilled the third and final recommendation of ASTM E 185-70 for the current DCPP Unit 1 operating license. Therefore, the proposed delayed removal of Capsule B does not deviate from DCPP's current reactor vessel materials surveillance program requirements. DCPP installed the Unit 1 Capsule B in the vessel in the Unit 1 Fifth Refueling Outage at a vessel exposure of 5.86 effective full power years (EFPY). The change in withdrawal schedule allows Capsule B to be withdrawn at a fluence of approximately 60 EFPY for the reactor pressure vessel. This will provide reactor pressure vessel fluence data for the period of extended operation for license renewal.

Enclosure 1 provides a description and assessment of the proposed change to the reactor vessel material surveillance program withdrawal schedule. Enclosure 2 provides a mark-up of the affected DCPP FSARU page,

To support implementation of the revised withdrawal schedule, PG&E requests approval of this proposed change by November 1, 2010. PG&E will revise DCPP FSARU Table 5.2-22 to incorporate the change upon NRC approval.

If you have any questions, please contact Mr. Tom Baldwin at (805) 545-4720.

Sincerely,

'\\ r M r . _ "< -4 James Qs Becker Site Vice President

A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway .Comanche Peak >> Diablo Canyon >> Palo Verde >> San Onofre

• South Texas Project Wolf Creek ER-141 Case: 23-3884, 03/25/2024, DktEntry: 23.2, Page 109 of 222

Document Control Desk PG&E Letter DCL-10-141 48 October 25, 2010 Page 3

prs/6984 Enclosures cc/enc: Elmo E. Collins, NRC Regional Administrator, Region IV Michael S. Peck, NRC Senior Resident Inspector Fred Lyon, NRC Project Manager, Nuclear Reactor Regulation Nathanial Ferrer, NRC Project Manager, License Renewal Kimberly J. Green, NRC Project Manager, License Renewal Alan B. Wang, NRC Project Manager, License Renewal cc: Diablo Distribution

A member ofthe STARS (Strategic Teaming and Resource Sharing) Alliance Callaway . Comanche Peak .Diablo Canyon >> Palo Verde San Onofre . South Texas Project Wolf Creek ER-142 Case: 23-3884, 03/25/2024, DktEntry: 23.2, Page 110 of 222

Enclosure 1 PG&E Letter DCL-10-141 Page 1 of 3

DESCRIPTION AND ASSESSMENT

1.0 BACKGROUND

Appendix H of 10 CFR 50 (Reference 1) requires a material surveillance program to monitor changes in the fracture toughness properties of ferrite material in the reactor vessel beltline region that result from exposure of these materials to neutron irradiation and the thermal environment. Under this program, fracture toughness test data are obtained and analyzed from material specimens exposed in surveillance capsules that are withdrawn periodically from the reactor vessel. Test results must be reported to the NRC within one year of the date of the capsule withdrawal. In addition, Section lll.B.3 of Appendix H to 10 CFR 50 requires the NRC to approve the capsule withdrawal schedule prior to implementation.

The design of the surveillance program and the withdrawal schedule must meet the requirements of the edition of ASTM E 185 that is current on the issue date of the ASME Code to which the reactor vessel was purchased.

The ASTM E 185 version of record for Diablo Canyon Power Plant (DCPP) Unit 1 is ASTM E 185-70. A Unit 1 supplemental reactor vessel surveillance capsule program schedule was requested by DCPP in Pacific Gas and Electric Company (PG&E) Letter DCL-92-072, dated March 31, 1992. The NRC approved this program in a letter to PG&E dated September 4, 1992, "Evaluation of Diablo Canyon Unit 1 Supplemental Reactor Vessel Radiation Surveillance Program (TAC No. M83285)." The schedule to withdraw Capsule B during the Unit 1 Sixteenth Refueling Outage was requested by DCPP in PG&E Letter DCL-08-021, dated March 12, 2008. DCPP requested this schedule revision to allow additional fluence exposure to the Unit 1 Capsule B to satisfy the NUREG-1801 requirement to have a capsule with fluence exposure between one and two times the vessel end of license extension fluence (54 EFPY fluence). The NRC approved this schedule revision by letter to PG&E dated September 24, 2008, "Diablo Canyon Power Plant, Unit No. 1 -Approval of Proposed Reactor Vessel Material Surveillance Capsule Withdrawal Schedule (TAC No. MD8371)."

2.0 DESCRIPTION

OF CHANGES TO THE REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM WITHDRAWAL SCHEDULE

The Unit 1 Reactor Material Surveillance Program withdrawal schedule is located in the DCPP Final Safety Analysis Report Update (FSARU). The proposed change revises the schedule to change the removal time for Capsule B from 21 .9 effective full power years (EFPY) to 23.2 EFPY.

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Enclosure 1 PG&E Letter DCL-10-141 Page 2 of 3

Unit 1 Capsule B was installed in the vessel in the Unit 1 Fifth Refueling Outage at a vessel exposure of 5.86 EFPY per FSARU Table 5.2-22. The lead factor for Capsule B at the 40 degree location is 3.46 per FSARU Table 5.2-22. The equivalent exposure on Capsule B at its withdrawal during the Unit 1 Seventeenth Refueling Outage is projected to be (23.2 - 5.86) X 3.46 = 60.0 EFPY.

3.0 ANALYSIS

NUREG-1801 requires that a licensee pursuing license renewal, and not crediting alternative dosimetry, must have a reactor vessel surveillance program consisting of a vessel material coupon that has fluence exposure equivalent to 60 years of operation. PG&E submitted its License Renewal Application for DCPP Unit 1 and 2 in PG&E Letter dated November 2009, and which is currently undergoing NRC review. The current DCPP withdrawal schedule for Unit 1 meets the NUREG-1801 requirements for license renewal. A removal time of 23.2 EFPY for Capsule B corresponds to 60 EFPY. This will continue to satisfy the NUREG-1801 requirements, which is to have a capsule with fluence exposure between one and two times the vessel end of license extension fluence (54 EFPY fluence).

The DCPP Unit 2 surveillance capsule withdrawal program currently meets the requirements of NUREG-1801, therefore, no changes in the DCPP Unit 2 surveillance capsule withdrawal program are needed.

This request to revise the removal time for DCPP Unit 1 Capsule B does not deviate from DCPP's current reactor pressure vessel materials surveillance program requirements.

4.0 REFERENCES

1. Code of Federal Regulations, Title 10, Part 50, Appendix H, "Reactor Vessel Material Surveillance Program Requirements."

2. American Society of Testing and Materials, "Standard Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels,"

ASTM E 185-70.

3. Diablo Canyon Final Safety Analysis Report Update (FSARU),

Revision 19, May 2010.

11

4. NUREG-1801, "Generic Aging Lessons Learned (GALL),

Revision 1, 2005.

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Enclosure 1 PG&E Letter DCL-10-141 Page 3 of 3

5. PG&E Letter DCL-09-079, "License Renewal Application," dated November 23, 2009.

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Enclosure 2 PG&E Letter DCL-10-141 Page 1 of 1

MARK-UP of TABLE 5.2-22

REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM WITHDRAWAL SCHEDULE

UNIT 1 Lead Removal Capsulelfllg) Location Factor(d) Time (EFPY)l@)

S 320° 3.46 1.25 (Tested,1 R1)

Y 40° 3.44 5.86 (Tested, 1 R5)

T 140° 3.44 5.86 (Removed, 1 R5)

Z 220° 3.44 5.86 (Removed, 1 R5)

V 320° 2.26 14.3 (Tested 1R11)

C(b) 140° 3.46 15.9 (Removed 1R12)

D(b) 220° 3.46 15.9 (Removed 1R12)

B(b) 40° 3.46 24-9-23.2 A(b) 184° 1.31 Standby U 356° 1.28 Standby X 176°40 1.28 Standby W 1.28 Standby UNIT 2 Lead Removal Capsule Location Factored) Time (EFPY)l@)

U 56° 5.15 1.02(Tested,2R1 )

X 236° 5.40 3.16 (Tested, 2R3)

Y 238.5° 4.58 7.08 (Tested, 2R6)

W08) 124° 5.25 11.49 (Removed, 2R9)

V(8) 58.5° 4.58 11.49 (Tested, 2R9)

Z(9) 304° 5.25 11.49 (Removed, 2R9)

(a) Approximate full power years from plant startup (b) Four supplemental capsules installed at 5.86 EFPY (EOC5)

(C) Deleted in Revision 16 (d) Approximate lead factors taken from WCAP-15958 (Rev. 0) and WCAP-15423 (Rev. 0) for Units 1 and 2, respectively (e) EFPY for Unit 2 capsules removed in 2R9, W = , = 52.6, and Z60.4 V 60.4 (f) Unit 1 capsules T, U, and Z are Type 1 (base metal only)W, x, (9) Unit 1 capsules S, V, and Y are Type 2 (base metal and weld)

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i

Pacific Gas and Electric Company

Diablo Canyon Power Plant PO. Box 56 Avila Beach, CA 93424 March 12, 2008 B00.545,50U0

PG&E Letter DCL-08-021

U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

Docket No. 50-275, OL-DPR-80 Diablo Canyon Unit 1 Revision to the Unit 1 Reactor Vessel Material Surveillance Program Withdrawal Schedule

Dear Commissioners and Staff:

I Pursuant to 10 CFR 50 Appendix H, Section lll.B.3, Pacific Gas and Electric Company (PG&E) hereby requests approval for a revision to the Unit 1 reactor vessel material surveillahce program withdrawal schedule.

The Unit 1 reactor material surveillance program withdrawal schedule is provided in Diablo Canyon Power Plant (DCPP) Final Safety Analysis Report Update (FSARU) Table 5.2-22. The proposed change would revise that schedule to reflect withdrawal of Capsule B during the Unit 1 Sixteenth Refueling Outage, which is scheduled to begin October 4, 2010. Capsule B is currently scheduled to be withdrawn during the Unit 1 Fifteenth Refueling Outage, which is scheduled to begin January 26, 2009.

Enclosure 1 provides a description and assessment of the proposed change to the reactor vessel material surveillance program withdrawal schedule.

Enclosure 2 provides a mark-up of the affected DCPP FSARU page.

To support implementation of the revised withdrawal schedule, PG&E requests approval of this proposed change by December 31, 2008. DCPP FSARU Table 5.2-22 will be revised to incorporate the change upon NRC approval.

If you have any questions, please contact Mr. Stan Ketelsen at (805) 545-4720.

Sinc rely,\\ \\\\ > *.

• \\ al'~

James R. ecker SiteVicePresident & Station Director /

A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway Comanche Peak Diablo Canyon Palo Verde

• South Texas Project Wolf Creek 4008 LLP(

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Document Control Desk PG&E Letter DCL-08-021 Ml8 March 12, 2008 Page 2

tcg/4231 Enclosures cc/enc: Elmo E. Collins, NRC Regional Administrator, Region IV \\

Michael S. Peck, NRC Senior Resident Inspector Alan B. Wang, NRC Project Manager, Nuclear Reactor Regulation Diablo Distribution

A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway Comanche Peak Diablo Canyon

• Palo Verde

• South Texas Project >> wolf Creek

• ER-148

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1

_ Enclosure 1 PG&E Letter DCL-08-021

a

DESCRIPTION AND ASSESSMENT \\

1.0 BACKGROUND

\\

Appendix H of10 CFR 50 (Reference 1) requires a material sUrveillance program to monitor changes in the fracture toughness properties of ferritic material in the reactor vessel beltline region which result from exposure of these materials to neutron irradiation and the thermal environment. Under this program, fracture toughness test data are,obtai.ned and analyzed from material specimens exposed in surveillance capsules which arewithdrawn periodically from the reactor vessel. Test results must be reported to the NRC within one year of the data of the capSulewithdrawal. Also,

4 Section lll.B.3 of Appendix H to 10 CFR 50, requires the capsule withdrawal schedule to be approved by the NRC prior to implementation.

The design of the surveillance program and the withdrawal schedule must meet the requirements of the edition of ASTM E 185 that is current on the issue date of the ASME Code to which the reactor vessel was purchased.

The ASTM E 185version of record for Diablo Canyon Power Plant I (DCPP) Unit 1 is ASTM E 185-70. A Unit 1 supplemental reactor vessel surveillance capsule program schedule was requested by DCPP in Pacific Gas and Electric Company (PG&E) Letter DCL-92-072, dated March 31, 1992, This program was approved by the NRC in a Safety Evaluation to PG&E dated September 4, 1992, "Evaluation of Diablo Cahyon Unit 1 Supplemental Reactor Vessel Radiation Surveillance Program (TAC No. M83285)."

2.0 DESCRIPTION

OF CHANGES TO THE REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM WITHDRAWAL SCHEDULE .

The Unit 1 Reactor Material.Surveillance Program withdrawal schedule is

l. located in the DCPP Final Safety Analysis Report Update (FSARU). The proposed change revises the schedule to change the removal time for Capsule B from 20.7 effective full power years (EFPY) to 21 .~9 EFPY.. ,

3.0 ANALYSIS \\ _

NUREG41801 requires that a licehsee pursuing license.renewal, and not crediting alternative dosimetry, must have a reactor vessel surveillance program consisting of a vessel material coupon that has fluence exposure

. equivalent to 60 years of operation. PG&E is currently performing a License Renewal Feasibility Study to determine whether to file a License Renewal Application for DCPP Units 1 and2. The current DCPP withdrawal schedule for Unit 1 does notmeet the NUREG-1801 requirements for license renewal. A change is requested in the removal

1 I

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M Enclosure 1 PG&E Letter DCL-08-021

time for Capsule B to accommodate NUREG-1801 compliance. A removal time of approximately 21.9 EFPY for Capsule B satisfies NUREG-1801 u

The DCPP Unit 2 sLlrveillance capsule withdrawal program currently meets the requirements of NUREG-1801, therefore, no changes in the DCPP Unit 2 surveillance capsule withdrawal program are needed.

The request to revise the removal time for DCPP Unit 1 Capsule B fully complies with the requirements of 10 CFR 50 Appendix H and ASTM E 185-70. In addition, this request remains in full compliance with the requirements set forth in DCPP Technical Specifications.

4.0 REFERENCES

1. Code of Federal Regulations, Title 10, Part 50, Appendix H, "Reactor Vessel Material Surveillance Program Requirements," January 1998.

2. American Society of Testing and Materials, "Standard Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels,"

ASTM E 185-70.

3. Diablo Canyon Final Safety Analysis Report Update (FSARU),

Revision 17, 2006.

4. NUREG-1801, "Generic Aging Lessons Learned (GALL),"

Revision 1, 2005.

8

2 ER-150

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( Enclosure 2 PG&E Letter DCL-08-021

MARK-UPof TABLE 5.2-22

REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM WITHDRAWAL SCHEDULE

1 UNIT 1

Lead Removal CapsUle(fl(gl Location Factorldl Time (EFPY)(a)

S 320° 3.46 1.25 (Tested,1 R1)

Y 40° 3.44 \\ 5.86 (Tested, 1 R5)

T 140° 3.44 5.86 (Removed, 1 R5)

Z 220° 3.44 5.86 (Removed, 1 R5)

V 320° 2.26 14.3 (Tested 1R1 1)

C(b) 140° 3.46 \\ 15.9 (Removed 1R12)

D(b) 220° 3.48 15.9 (Removed 1R12)

B(b) 40° 3.48 2Q=¥21.9 I A(b) 184° 1 .31 Standby U 356° 1 .28 Standby X 176° 1 .28 Standby W 40 1 .28 Standby

UNIT 2 Lead Removal Capsule Location Factored) Time (EFPY)(a)

U 56° 4 5.15 1.~02(Tested,2R1 )'

X 236° 5.40 3.16 (Tested, 2R3)

'Y 238.5° 4.58 7.08 (Tested, 2R6)

W(6) 124° 5.26 11.49 (Removed, 2R9)

V() 58.5° 4.58 11.49 (Tested, 2R9) 2(8) 304° 5.26 11.49 (Removed, 2R9)

(a) Approximate full power years from plant startup.

(b) Four supplemental capsules installed at 5.86 EFPY (EOC5).

(C) Deleted in Revision 16. /

(d) Approximate lead factors taken from WCAP-15958 (Rev. 0) and WCAP-15423 (Rev. 0) for Units 1 and 2, respectively.

(e) EFPY for Unit 2 capsules removed in 2R9, W = 60.4, V = 52.6, and Z 60.4 (f) Unit 1 capsules T, U, W, X, and Z are Type 1 (base metal only)

(9) Unit 1 capsules S, V, and Y are Type 2 (base metal and weld)

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Federal Register/Vol. 71, No. 157/Tuesday, August 15, 2006/Notices 46945

comments received: No. Amendment Nos.:232 and 228.No significant hazards considerationRenewedFacility Operating Licensethan OI` equal to 89 °F, and the UHSverified at least once per hour to be less Energy Operations, Inc., Docket No. 50- Nos. DPR-29 and DPR-30:The temperature does not exceed a 368, Arkansas Nuclear One, Unit No. 2, amendments revised the License.Date of initial notice inFederalmaximum value of 91.4 °F.Date of issuance:August 1, 2006.

Pope County, Arkansas Register:May 23, 2006 (71 FR 29678). Ejective date: As of the date of September 19, 2005, as supplementedDate of application for amendment:The May 17, 2006, supplementissuance, to be implemented within 60 by letters dated May 11 and June 19, contained clarifying information and days.Amendment No.:168.

did not change the NRC staff's initial 2006.Brief description of amendment:The proposed finding of no significant 57: This amendment revised the TSs.Facility Operating License No. NPF-amendment revised the existing steam hazards consideration.The Commission's related evaluationDate of initial notice inFederal generator tube surveillance program to of the amendments is contained in a Register:August 30, 2005 (70 FR be consistent with the U.S. Nuclear Safety Evaluation dated July 24, 2008. 51382].

Regulatory Commission's approved No significant hazards consideration The Commission's related evaluation Technical Specification Task Force commentsreceived: No. of the amendment is contained in a Standard Technical Specification Pacific Gas and Electric Company, Safety Evaluation dated August 1, 2000.

Change Traveler, TSTF-449, "Steam Docket Nos. 50-275 and 50-323, Diablo No signi]9cant hazards consideration Generator Tube Integrity," Revision 4. Canyon Nuclear Power Plant, Unit NOS. comments received: No.

TSTF-449 is part of the consolidated 1 and 2, San Luis Obispo County, R.E. Ginna Nuclear Power Plant, LLC, line item improvement process.Date of issuance:August 2, 2006.California Docket No. 50-244, RE. Ginna Nuclear E]§'ective date:As of the date of Date of application for amendments: PowerPlant, Wayne County, New York issuance to be implemented within 90 August 23, 2005, as supplemented on November 7, 2005, as supplemented onDate of application for amendment:

days from the date of issuance.Amendment No.:266.April 6, 2006.Brief description of amendments:TheMay 5, 2006.

No. NPF-6:Amendment revised theRenewed Facility Operating Licenseamendments extended the licensed livesamendment revises TechnicalBrief description of amendment:The Technical Specifications and Renewed of the Diablo Canyon Power Plant, Unit Specification 3.9.3, "Containment Facility Operating License. Nos. 1 and 2 reactors by the amount of Penetrations," to allow an emergency Date of initial notice inFederal time the licensee had expended to egress door, access door, or roll up door, Register:January 3, 2006 (71 FR 147). perform low-power testing of the as associated with the equipment hatch The supplements dated May 11 and reactors prior to initial startup.Dateofissuancefluly 17, 2006.penetration, to be open, but capable of June 19, 2006, provided additional Effectivedate:As of its date of being closed, during core alterations or information that clarified the issuance and shall be implemented movement of irradiated fuel within application, did not expand the scope of within Q0 days of issuance. containment.

the application as originally noticed, Amendment Nos.:Unit 1-188, Unit EjjL'ective date:As of the date ofDate of issuance:July 26, 2006.

and did not change the staff's original 2-190.Facility Operating License Nos. DPR-proposed no significant hazards issuance to be implemented within 60 consideration determination as revised the Facility Operating Licenses.80 and DPR-82:The amendmentsdays.Amendment No.:98.

published inthe Federal Register.The Commission's related evaluationDate of initial notice inFederalRenewedFacility Operating License of the amendment is contained in a Register:October 11, 2005 (70 FR No. DPR-18:Amendment revised the Safety Evaluation dated August 2, 2006. 59087). The April 6, 2006, supplemental Technical Specifications.

No signi]9cant hazards consideration letter provided additional information Date of initial notice inFederal commentsreceived: No. that clarified the application, and did Register:Ianuary 3, 2006 (71 FR 154).

Echelon Generation Company, LLC, not expand the scope of the application The May 5, 2006, letter provided Docket Nos. 50-254 and 50-265, Quad as originally noticed. additional information that clarified the Cities Nuclear Power Station, Units 1 of the amendments is contained in aThe Commission's related evaluationapplication, did not expand the scope ofthe application as originally noticed, and 2, Rock Island County, Illinois Safety EvaluationdatedJuly 17, 2006. and did not change the staff's original Date of application for amendments: No signijdcant hazards consideration proposed no significant hazards Ianuary 25, 2006, as supplemented by comments received: No. consideration determination as letter dated May 17, 2006.Briefdescription of amendments:ThePSEGNuclear LLC, Docket No.50-354,published inthe Federal Register.The Commission's related evaluation amendment revised the Quad Cities Hope Creek Generating Station, Salem of the amendment is contained in a licensing basis, as described in the County, New Iersey Safety Evaluation dated July 26, 2006.

Updated Final Safety Analysis Report, Date of application for amendment: No significant hazards consideration to allow the use of automatic load tap August 4, 2005, as supplemented by commentsreceived: No.

changers to operate in automatic mode letters dated February 9, July 18, and RE. Ginna Nuclear Power Plant, LLC, on the reserve auxiliary transformers toAugust 1, 2006.Brief description of amendment:TheDocket No. 50-244, RE. Ginna Nuclear compensate for potential offsite power amendment revised Technical PowerPlant, Wayne County, New York voltage fluctuations, in order to ensureSpecification (TS) 3.7.1.3, "Ultimate that acceptable voltage is maintained forHeat Sink," to permit continued plant Date of application for amendment:

safety-related equipment.Dateof issuance:July 24, 2006.operation if the temperature of theNovember 18, 2005.Brief description of amendment:The Effective date:Asof thedateof ultimate heat sink (UHS) exceeds 89 °F, amendment revises the frequency in issuance and shall be implemented provided the UHS temperature averaged Technical Specification Surveillance within 30 days. over the previous 24-hour period is Requirement 3.6.6.15, which verifies

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46946 Federal Register/Vol. 71, No. 157/Tuesday, August 15, 2006/Notices

that each containment spray nozzle is Specification Surveillance in TS 5.65.b, "Core Operating Limits unobstructed. The frequency is changed Requirements to increase the minimum Report (COLR)," to permit the use of an from "10 years" to "following required average ice basket weight, thus,alternate methodology to perform a maintenance which could result in increasing the corresponding total thermal-hydraulic analysis to predict nozzle blockage." weight of the stored ice in the WBN ice the critical heat flux and departure from Dateof issuance:July 31, 2006. condenser. The changes to the ice basket nucleate boiling ratio for the AREVA Effective date:As of the date of and total ice weights are due to the Advanced Mark-BW fuel in the North issuance to be implemented within 60 additional energy associated with the Anna 1 and 2 cores.

days.Amendment No.:QQ. Replacement Steam Generators.Date of issuance:Iuly 25, 2006.Elective date:As of the date ofDateofi55uance:July 21, 2006.

No. DPR-18:Amendment revised theRenewedFacility Operating LicenseEffectivedate:As of the date ofissuance and shall be implemented Technical Specifications and the issuance and shall be implementedprior to Mode 4 at startup to begin Cyclewithin 60 days from the date of License. 8 fuel cycle. issuance.Amendment Nos.:247, 227.

Register:January 3, 2006 (71 FR 154).Date of initial notice inFederalAmendment No. 62.RenewedFacility Operating License The Commission's related evaluation 90: Amendment revises the TechnicalFacility Operating License No. NPF-Nos. NPF-4 and NPF-7:Amendments of the amendment is contained in a Specifications. changed the Licenses and the TSS.Date of initial notice inFederal Safety Evaluation dated July 31, 2006.No significant hazards considerationRegister:February 14, 2006 (71 FRDate of initial notice inFederalRegister:August 16, 2005 (70 FR commentsreceived: No. 7814). The supplemental letter provided 48208]. The supplements dated March clarifying information that was within 30, April 13, and May 11, 2006, Tennessee Valley Authority, Docket No. the scope of the initial notice and did contained clarifying information only 50-259 Browns FerryNuclear Plant, not change the initial proposed no and did not change the initial no Unit 1, Limestone County, Alabama significant hazards consideration significant hazards consideration Date of application for amendment: determination. determination or expand the scope ofthe initial application.

December 6, 2004 (TS 428] as The Commission's related evaluation The Commission's related evaluation supplemented by letter dated Lune 16, of the amendment is contained in a of the amendments is contained in a 2005.Brief description of amendment:The Safety Evaluation dated July 25, 2006. Safety Evaluation dated July 21, 2006.

amendment revised the reactor vessel comments received: No.No 5igni]9cant hazards considerationNo signicant hazards consideration Pressure-Temperature curves depicted comments received: No.

in the Technical Specification (TS) Union Electric Company, Docket No. of August, 2006.Dated at Rockville, Maryland, this 8th day Figure 3.4.9-1 and adds a new TS 50-483,Callaway Plant, Unit 1, Figure 3.4.9-2. Calloway County, Missouri For the Nuclear Regulatory Commission.

Date of issuance:July 26, 2006. Date of application for amendment: Catherine Haney, issuance and shall be implementedEffective date:As of the date ofMarch 28, 2006. Licensing,Officeof Nuclear ReaetorDirector, Division of Operating Reactor within 60 days of issuance. amendment revised TechnicalBrief description of amendment:TheRegulation.

Amendment No.:256. Specification 5.0, "Administrative [FR Doc. 06-6921 Filed 8-14-06; 8:45 am]

33:Amendment revised theTS.Facility Operating License No. DPR-Controls," by changing a position titleBILLING CODE 7590-01-P Date of initial notice inFederal and department name.Date of issuance:July 11, 2006.

Register:January 18, 2005 (70 FR E]§'ective date:As of its date of SECURITIES AND EXCHANGE 2899). The supplement dated June 16, issuance, and shall be implemented COMMISSION 2005, provided additional information within 90 days of the date of issuance. [Release No. 34-54296; File No. SR-lSE-that clarified the application, did not Amendment No.:173. 2006-30]

expand the scope of the application as Facility Operating License No, NPF-originally noticed, and did not change 30: The amendment revised the Self-Regulatory Organizations; the staff's original proposed no Technical Specifications, International Securities Exchange, Inc.,

significant hazards consideration Date of initial notice inFederal Order Approving a Proposed Rule determination as published in the Register:May 9, 2006 (71 FR 27005).The Commission's related evaluationChange, and Amendment No. 1 Federal Register.The Commission's related evaluationof the amendment is contained in aThereto, Increasing the Linkage of the amendment is contained in a Safety Evaluation dated July 11, 2006. Inbound Principal Order Fee Safety Evaluation dated July 26, 2006. No signi]9cant hazards consideration August 9, 2006.

comments received:No.No significant hazards considerationcomments received: No. Securities Exchange, Inc. ("ISE" orOn June 5, 2006, the International Virginia Electric and Power Company, "Exchange"] filed with the Securities Tennessee Valley Authority, Docket No. Docket Nos. 50-338 and 50-39, North and Exchange Commission 50-390, WattsBar Nuclear Plant, Unit 1, Anno Power Station, Units 1 and 2, ("Commission"), pursuant to Section Rhea County, Tennessee Louisa County, Virginia 19(b](1) of the Securities Exchange Act December 15, 2005 (TS-05-09), asDate of application for amendment:July 5, 2005, as supplemented by lettersDate of application for amendment:of 1934 ("Act") 1 and Rule 1eb-4 supplemented by letter datedJune 7, dated March 30, April 13, and May 11, thereunder? a proposed rule change to 2006. 2006. amend its Schedule of Fees in the Brief description of amendment:The Brief description of amendment:The manner described below. On June 29, amendment revises the Watts Bar amendments revised the Technical 1 15 U.S.C. 78s(b)(1].

Nuclear Plant (WEN) Technical Specifications (TSS) to add a reference 217 CFR 240.19b-4.

ER-153 Case: 23-3884, 03/25/2024, DktEntry: 23.2, Page 121 of 222

July 17, 2006

Mr. John S. Keenan Senior Vice President and Chief Nuclear Officer Pacific Gas and Electric Company Diablo Canyon Power Plant P.O. Box 770000 San Francisco, CA 94177-0001

SUBJECT:

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 MC8207)

Dear Mr. Keenan:

The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 188 to Facility Operating License No. DPR-80 and Amendment No. 190 to Facility Operating License No. DPR-82 for the Diablo Canyon Power Plant, Unit Nos. 1 and 2 (DCPP-1/2), respectively. The amendments consist of changes to the Facility Cperating Licenses in response to your application dated August 23, 2005, as supplemented on April 6, 2006.

The amendments extend the licensed lives of the DCPP-1/2 reactors by the amount of time the licensee had expended to perform low-power testing of the reactors prior to initial startup.

A copy of the related Safety Evaluation is enclosed. The Notice of Issuance will be included in the Commission's next regular biweeklyFederal Registernotice.

Sincerely,

/RAI

Alan Wang, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Docket Nos. 50-275 and 50-323

Enclosures:

1. Amendment No. 188 to DPR-80

2. Amendment No. 190 to DPR-82

3. Safety Evaluation

CC w/encls: See next page

ER-154 Case: 23-3884, 03/25/2024, DktEntry: 23.2, Page 122 of 222 July 17, 2006 Mr. John S. Keenan Senior Vice President and Chief Nuclear Officer Pacific Gas and Electric Company Diablo Canyon Power Plant P.O. Box 770000 San Francisco, CA 94177-0001

SUBJECT:

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 MC8207)

Dear Mr. Keenan:

The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 188 to Facility Operating License No. DPR-80 and Amendment No. 190 to Facility Operating License No. DPR-82 for the Diablo Canyon Power Plant, Unit Nos. 1 and 2 (DCPP-1/2), respectively. The amendments consist of changes to the Facility Operating Licenses in response to your application dated August 23, 2005, as supplemented on April 6, 2006.

The amendments extend the licensed lives of the DCPP-1/2 reactors by the amount of time the licensee had expended to perform low-power testing of the reactors prior to initial startup.

A copy of the related Safety Evaluation is enclosed. The Notice of Issuance will be included in the Commission's next regular biweeklyFederal Registernotice.

Sincerely,

/RAI Alan Wang, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-275 and 50-323 DISTRIBUTION:

PUBLIC GHiII (4)

LPLIV Reading RidsNrrDirsltsb (TKobetz)

Enclosures:

1. Amendment No. 188 to DPR-80 RidsNrrDorl (CHaney/CHolden)

2. Amendment No. 190 to DPR-82 RidsNrrDorlLpl4 (DTerao) RidsOgcRp

3. Safety Evaluation RidsNrrPMAWang RidsNrrLALFeizollahi CC w/encls: See next page RidsAcrsAcnwMailCenter RidsRegion4MailCenter (BJones)

RidsNrrDorlDpr JMedhoff, DLR ACCESSION NO.: Pkg ML062260278 (ML061660220, TS Pgs. ML062270019)

OFFICE NRR/LPL4/PM NRR/LPL4/LA ADES/CVIB/BC OGC NRR/LPL4/BC NAME AWang LFeizollahi MMitchell HEW DTerao DATE 6/21/06 6/20/06 5/31/06 7/3/067/12/06 OFFICIAL RECORD COPY

ER-155 Case: 23-3884, 03/25/2024, DktEntry: 23.2, Page 123 of 222

PACIFIC GAS AND ELECTRIC COMPANY

DOCKET NO. 50-275

DIABLO CANYON NUCLEAR POWER PLANT, UNIT NO. 1

AMENDMENT TO FACILITY OPERATING LICENSE

Amendment No. 188 License No. DPR-80

1. The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Pacific Gas and Electric Company (the licensee) dated August 23, 2005, as supplemented on April 6, 2006, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's regulations set forth in 10 CFR Chapter I,

B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission,

c. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations,

D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public, and

E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2. Accordingly, the license is amended by changes to the Facility Operating License No. DPR-80 as indicated in the attachment to this license amendment.

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2

3. This license amendment is effective as of its date of issuance and shall be implemented within 90 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

/RAI

David Terao, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to pages 3 and 9 of Facility Operating License No. DPR-80

Date of Issuance: July 17, 2006

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PACIFIC GAS AND ELECTRIC COMPANY

DOCKET NO. 50-323

DIABLO CANYON NUCLEAR POWER PLANT, UNIT NO. 2

AMENDMENT TO FACILITY OPERATING LICENSE

Amendment No. 190 License No. DPR-82

1. The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Pacific Gas and Electric Company (the licensee) dated August 23, 2005, as supplemented on April 6, 2006, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's regulations set forth in 10 CFR Chapter I,

B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission,

c. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations,

D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public, and

E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2. Accordingly, the license is amended by changes to the Facility Operating License No. DPR-82 as indicated in the attachment to this license amendment.

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2

3. This license amendment is effective as of its date of issuance and shall be implemented within 90 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

/RAI

David Terao, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to pages 3 and 7 of Facility Operating License No. DPR-82

Date of Issuance: July 17, 2006

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ATTACHMENT TO LICENSE AMENDMENT NO. 188

TO FACILITY OPERATING LICENSE NO. DPR-80

AND AMENDMENT NO. 190 TO FACILITY OPERATING LICENSE NO. DPR-82

DOCKET nos. 50-275 AND 50-323

Replace the following pages of the Facility Operating Licenses with the attached revised pages.

The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Facility Operating License No. DPR-80:

REMOVE INSERT

3 3 g g

Facility Operating License No. DPR-82:

REMOVE INSERT

3 3 7 7

ER-160 Case: 23-3884, 03/25/2024, DktEntry: 23.2, Page 128 of 222

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION

RELATED TO AMENDMENT NO. 188 TO FACILITY OPERATING LICENSE NO. DPR-80

AND AMENDMENT NO. 190 TO FACILITY OPERATING LICENSE NO. DPR-82

PACIFIC GAS AND ELECTRIC COMPANY

DIABLD CANYON POWER PLANT, UNITS 1 AND 2

DOCKET nos. 50-275 AND 50-323

1.0 INTRODUCTION

By application dated August 23, 2005, as supplemented on April 6, 2006 (Agencywide Documents Access and Management System Accession Nos. ML052420441 and ML061040224, respectively), Pacific Gas and Electric Company (PG&E, the licensee) requested changes to the Facility Operating Licenses (Facility Operating License Nos. DPR-80 and DPR-82) for the Diablo Canyon Power Plant, Units 1 and 2 (DCPP-1/2).

The proposed amendments would extend the licensed lives of the DCPP-1/2 reactors by the amount of time the licensee had expended to perform low-power testing of the reactors prior to initial startup. Specifically, the proposed changes would revise the licenses to reflect the following:

1. Extend the expiration date for DCPP-1 Operating License No. DPR-80 from September 22, 2021, to November 2, 2024. This revised expiration date equates to 35.2 effective full power years (EFPY) of power operations.

2. Extend the expiration date for DCPP-2 Operating License No. DPR-82 from April 26, 2025 to August 26, 2025. This revised expiration date equates to 35.8 EFPY.

The supplemental letter dated April 6, 2006, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the Nuclear Regulatory Commission (NRC) staff's original proposed no significant hazards consideration determination as published inthe Federal Registeron October 11, 2005 (70 FR 59087).

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2

2.0 REGULATORY EVALUATION

2.1 Requirements for Upper Shelf Energy (USE)

Section lV.A.1 to Title 10,Code of Federal Regulations(10 CFR) Part 50, Appendix G (Reference 4), provides the Commission's requirements for demonstrating that reactor vessels (RVs) in U.S. pressurized-water reactor (PWR), light-water facilities will maintain adequate protection from failure by ductile tearing throughout their service lives. The rule requires RV beltline materials to have USE values equal to or above 75 ft-lb when the materials are in the unirradiated condition and equal to or above 50 ft-lb throughout the licensed life of the reactor.

Regulatory Guide (RG) 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials,II provides an expanded discussion regarding the calculation of USE values and describes two methods for determining USE values for RV beltline materials, depending on whether or not a given RV beltline material is represented in the plant's Reactor Vessel Material Surveillance Program (RVMSP).

2.2 Requirements for Performing RTPTs Calculations

Section 50.61 of 10 CFR (Reference 7) provides the Commission's requirements for demonstrating that RVs in U.S. PWR light-water reactor facilities will have adequate protection against the consequences of pressurized thermal shock (PTS) events throughout their service lives. The rule requires PWR licensees to calculate an adjusted reference temperature value for PTS (i.e., an RTPTSvalue) for each base metal and weld material located in the beltline region of their RVs. The rule sets a screening limit of 270 EF for RTPTSvalues that are calculated for base metals (i.e., forging and plate materials) and axial weld materials, and a screening limit of 300 EF for RTPTSvalues that are calculated for circumferential weld materials. The rule also provides an expanded discussion regarding how the calculations of RTPTSvalues should be performed and describes two methods for determining RTPTSvalues for RV beltline materials, depending on whether or not a given RV beltline material is represented in the plant's RVMSP.

2.3 RVMSP Requirements

The NRC staff's regulatory requirements for the establishment of RVMSPs are given in 10 CFR Part 50, Appendix H (Reference 8).

2.4 Requirements for Pressure-Temperature (P-T) Limits

Section 50.36 of 10 CFR (Reference 5) requires that the Technical Specification (TS) limiting conditions of operation (LCOs) include LCOs on P-T limits for operating reactors.Section IV.A.2 of 10 CFR Part 50, Appendix G, requires that P-T limits for normal operations (including heatups, cooldowns, operation with the core critical, and operation during anticipated operational transients) and pressure test operations of operating reactors be at least as conservative as those that would be generated using the methodology in Appendix G to Section Xl of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Section Xl, Appendix G)-

The guidance in Generic Letter (GL) No. 96-03, "Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits" (Reference 6), permits licensees to relocate the actual P-T limit curves from the LCO's into an owner controlled

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3

pressure-temperature limit report (PTLR). The PTLR is controlled under the TS Administrative Controls Section based on the licensee's use of an NRC-approved P-T limit calculational methodology and is provided the PTLR satisfies particular criteria that are provided in of GL 96-03. NRC approval of a license amendment for a PTLR permits the particular licensee to make further changes of the P-T limits without having to submit them for NRC approval.

3.0 TECHNICAL EVALUATION

By letter dated January 27, 2006 (Reference 2), the NRC staff issued a request for additional information (RAI) requesting PG&E to provide the adjusted reference temperature (RTPTS) value and Charpy-V notch USE value calculations for the DCPP-1/2 RV beltline materials, as assessed for the neutron fluence exposure (in units of n/cm2, E > 1.0 MeV) to the materials at the new expiration dates for the reactor units. The licensee responded to the NRC staff's RAI by letter dated April 6, 2006 (PG&E Serial Letter No. DCL-06-045/Reference 3). In this letter, the licensee provided the following information: (1) the amended neutron fluence values for the RV beltline materials, as assessed at the RV clad-to-base metal interface through the new expiration dates for the units, and (2) the limiting RTPTs values and limiting USE values for the DCPP-1/2 RV beltline materials, as assessed through the new expiration dates for the units. All future references to the licensee's response to the NRC staff's RAI are made in reference to Serial Letter No. DCL-06-045.

The evaluations in Section 3.0 of this safety evaluation (SE) provide the NRC staff's assessment of the impact of the license amendment request on the following mandated RV programs and analyses:

(1) the safety assessment that is required by 10 CFR Section 50.61 to ensure protection ofthe DCPP-1/2 RVs against the consequences of a PTS event,

(2) the safety assessment that is required by 10 CFR Part 50, Appendix G, to ensure that theDCPP-1/2 RV beltline materials will have acceptable levels of USE at the expiration of the operating licenses,

(3) the RV materials surveillance program withdrawal schedule that is mandated by10 CFR Part 50, Appendix H, and

(4) the pressure-temperature limits for the RVs, as mandated in accordance with10 CFR 50.36 and regulated in accordance with the requirements of 10 CFR Part 50,

Appendix G, and Administrative Technical Specification 5.9.6, which controls the licensee's PTLR process.

3.1 USE Assessment

The licensee performed updated USE assessments for the RV beltline materials at DCPP-1/2 and provided the updated limiting USE values in its response to the NRC staff's RAI. The USE assessments were based on the 1/4T neutron fluence values for the RV beltline materials, as assessed for the revised expiration dates of the DCPP-1/2 operating licenses. This is consistent with the recommended methods in RG 1.99, Revision 2, for performing USE calculation of RV beltline materials.

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The NRC staff performed independent calculations of the USE values for the RV beltline materials and applied the 1/4T neutron fluence values for the DCPP-1/2 RVs as its basis for its independent USE calculations. These values were based on the updated inside surface fluences provided in PG&E Serial Letter No. DCL-06-045, as attenuated using the methods of analysis in RG 1.99, Revision 2.

The NRC staff determined that for its USE assessment of DCPP-1, Lower Shell Axial Weld No. 3-442C (Heat No. 27204) is the limiting beltline material for USE. The NRC staff calculated a USE of 64.5 ft-lb for this weld at 35.2 EFPY. This value is in reasonable agreement with the USE value calculated by the licensee for these welds (i.e., 61 .1 ft-lb) at 35.2 EFPY. Both of these values meet the acceptance criterion in 10 CFR Part 50, Appendix G, for maintaining the USE values of the RV beltline materials above 50 ft-lbs throughout the licensed life of the plant.

The NRC staff determined that for DCPP-2, Lower Shell Axial Weld No. 3-201 B (Heat No. 33A277) is the limiting beltline material for USE. The NRC staff calculated a USE of 57.5 ft-lb for this weld at 35.8 EFPY. This value is in good agreement with the USE value calculated by the licensee for this welds (i.e., 57.7 ft-Ib) at 35.8 EFPY. Both of these values meet the acceptance criterion in 10 CFR Part 50, Appendix G, for maintaining the USE values of the RV beltline materials above 50 ft-lbs throughout the licensed life of the plant.

Based on this assessment, the NRC staff concludes the RV beltline materials at DCPP-1/2 will have acceptable remaining margins on USE through the amended expiration dates for the current DCPP-1/2 operating licenses.

3.2 Impact on the PTS Assessments and RTPTSCalculations

The licensee performed updated RTPTScalculations for the RV beltline materials at DCPP-1/2 and provided the updated limiting RTPTs values in its response in PG&E Serial Letter No. DCL-06-045. The licensee's RTPTScalculations were performed in accordance with the methods of calculation in 10 CFR 50.61, as assessed using the updated neutron fluences for the RV beltline materials at the clad-to-based metal interface for the revised expiration dates for the current operating licenses. This is in compliance with 10 CFR 50.61 .

The NRC staff performed independent calculations of the RTPTSvalues for the RV beltline materials at DCPP-1/2 through the revised expiration dates for the DCPP-1/2 operating licenses.

The NRC staff used the methods of analysis in 10 CFR 50.61 as the basis for its independent RTPTScalculations. The NRC staff applied the updated clad-to-base metal neutron fluences that were provided in PG&E Serial Letter No. DCL-06-045.

The NRC staff determined that for DCPP-1, Lower Shell Axial Weld No. 3-442C (Heat No. 27204) is the limiting beltline material for PTS. The NRC staff calculated a RTPTs value of 258.8 EF for this weld at 35.2 EFPY. This value is in agreement with the RTPTs value calculated by the licensee for this weld (i.e., 258.7 EF) at 35.2 EFPY. Both of these values meet (i.e., are lower than) the PTS screening criterion of 270 EF in 10 CFR 50.61 for RV axial weld materials at the end of the operating license (EOL) and are acceptable.

The NRC staff determined that for DCPP-2, Intermediate Shell Plate No. 65454-2 (Heat No. C5168-2) is the limiting beltline material for PTS. The NRC staff calculated a RTPTs value of 214.8 EF for this weld at 35.8 EFPY. This value is in agreement with the RTPTs value

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5

calculated by the licensee for this weld (i.e., 214.8 EF) at 35.8 EFPY. Both of these values meet (i.e., are lower than) the PTS screening criterion of 270 EF in 10 CFR 50.61 for RV axial weld materials at the EOL and are acceptable.

Based on this assessment, the NRC staff concludes the RV beltline materials as DCPP-1/2 will remain in compliance with the PTS screening criteria of 10 CFR 50.61 through the amended expiration dates for the current DCPP-1/2 operating licenses.

3.3 Impact on the RVMSP

Section lll.B.3 of 10 CFR Part 50, Appendix H, requires that designs of RVMSPs and RV surveillance capsule withdrawal schedules be implemented in accordance with the version of American Society for Testing and Materials Standard Practice E185 (ASTM E185

[Reference 9]) that was current at the time the RV was purchased in accordance with Section Ill of the ASME Boiler and Pressure Vessel Code, Division 1. The rule permits more current versions of ASTM E185 to be used inclusive of the 1982 version (ASTM E185-82). The rule also requires that changes to the RV surveillance capsule withdrawal schedule be reviewed and approved by the NRC staff.

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 (Reference 10). This RVMSP withdrawal schedule was approved in an SE to PG&E dated September 4, 1992 (Reference 11). 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 licensee's 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 1019n/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 1019n/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 licensee's to meet the RVMSP withdrawal criteria of more recent versions of ASTM E185, inclusive of E185-82. Therefore, the NRC staff concludes that 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.

The ASTM E185 version of record for DCPP-2 is ASTM E185-73. The most recent RVMSP withdrawal schedule for DCPP-2 was requested in PG&E Serial Letter No. DCL-97-178, dated

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October 22, 1997 (Reference 13). This RVMSP withdrawal schedule was approved in an SE to PG&E dated Feburary 10, 1998 (Reference 14). In the NRC staff's SE, the NRC staff concluded the RVMSP withdrawal schedule provided in PG&E's letter of October 22, 1997, constituted an acceptable RVMSP withdrawal schedule for DCPP-2 that met the withdrawal schedule criteria of ASTM E185-73 and met the intent of the more recent withdrawal schedule criteria of ASTM E185-82. Under this program, four capsules, Capsule U, X, Y, and V were designated for removal from the DCPP-2 RV. PG&E has removed all of these capsules in accordance with the withdrawal schedule criteria that was approved in the NRC staff's SE of February 10, 1998.

PG&E is not required to withdraw any further surveillance capsules from DCPP-2 for the current operating term and the proposed license amendment will not impact continued compliance of the DCPP-2 RVMSP for the current operating period, as assessed against the requirements of 10 CFR Part 50, Appendix H.

3.4 Impact of Reactor Vessel P-T Limits

On May 13, 2004 (Reference 16), License Amendment No. 170 for DCPP-1 Operating License No. DPR-80 and License Amendment No. 171 for DCPP-2 Operating License No. DPR-82 were approved for DCPP-1/2, respectively. These license amendments permitted PG&E to relocate the P-T limits for DCPP-1/2 into a PTLR and to make administrative changes of the P-T limits in accordance with the PTLR process governed by TS 5.9.6. Thus, the licensee is authorized to make any changes of the P-T limits that are necessary to account for the change in the neutron fluence values resulting from the recovery of the power ascension time in accordance with the PTLR process (i.e., in accordance with TS 5.9.6).

3.5 Recovery of the Low-Power Testing Time Period

The NRC staff has reviewed PG&E's license amendment request to recover the low-power testing time that was performed during the initial startups of the DCPP-1/2 reactors. The NRC staff has determined that authorization of the requested license may be granted based on the following conclusions:

(1) The DCPP-1/2 RV beltline materials will remain in compliance with the acceptance criteriaof 10 CFR Part 50, Appendix G, for the remaining USE safety margins.

(2) The revised RTPTs values for the DCPP-1/2 RV beltline materials are below theRTPTs screening criteria. Therefore, the DCPP-1/2 RV beltline materials remain in

compliance with the requirements of 10 CFR 50.61 for demonstrating adequate protection against PTS events.

(3) The RV surveillance capsule withdrawal schedules for DCPP-1/2 remain in compliancewith the requirements of 10 CFR Part 50, Appendix H, and the ASTM E185 versions of record for the units.

(4) The licensee will make any changes that may be necessary to the DCPP-1/2 P-T limits inaccordance with the PTLR process that is administratively mandated and controlled under

TS 5.9.6.

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4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the California State official was notified of the proposed issuance of the amendments. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration and there has been no public comment on such finding on October 11, 2005 (70 FR 59087). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51 .22(c)(9). Pursuant to 10 CFR 51 .22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.

6.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

7.0 REFERENCES

1. PG&E Serial Letter No. DCL-05-098, "License Amendment Request 05-03, Request for Amendment to Recapture Low-Power Testing Time," August 23, 2005.

2. Letter from A. B. Wang (NRC) to J. S. Keenan, "Diablo Canyon Power Plant, Units 1 and 2 - Request for Additional Information on License Amendment Request for Recovery of Low-Power Testing Time - Impact on Reactor Vessel Integrity Assessments (TAC NOS.

MC8206 and MC8207)," January 27, 2006.

3. PG&E Serial Letter No. DCL-06-045, "Response to Request for Additional information Regarding License Amendment Request 05-03, 'Request for Amendment to Recapture Low-Power Testing Time,' " April 6, 2006.

4. Appendix G to Title 10,Code of Federal Regulations,"Fracture Toughness Requirements."

5. Paragraph §50.36 of Part 50 to Title 10,Code of Federal Regulations(10 CFR 50.36),

"Technical specifications."

6. Generic Letter No. 96-03, "Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits."

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8

7. Paragraph §50.61 of Part 50 to Title 10,Code of Federal Regulations(10 CFR 50.61 ),

"Fracture toughness requirements for protection against pressurized thermal shock events."

8. Appendix H to Title 10,Code of Federal Regulations,"Reactor Vessel Material Surveillance Program Requirements."

9. American Society for Testing and Materials Designation E185, "Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels":

1982 Edition (ASTM E185-82), July 1, 1982, 1973 Edition (ASTM E185-73), March 1, 1973, 1970 Edition (ASTM E185-70),

10. PG&E Serial Letter No. DCL-92-072, "Diablo Canyon Unit 1, Supplemental Reactor Vessel Radiation Surveillance Program," March 31, 1992..

11. Letter from H. Rood (NRC) to G. M. Rueger (PG&E), "Evaluation of Diablo Canyon Unit 1 Supplemental Reactor Vessel Radiation Surveillance Program (TAC No. M83285),"

September 4, 1992.

12. WCAP-15858, Revision 0, "Analysis of Capsule V from Pacific Gas and Electric Company Diablo Canyon Unit 1 Reactor Vessel Radiation Surveillance Program," January 2003.

13. PG&E Serial Letter No. DCL-97-178, "Diablo Canyon Unit 2, Revision to Reactor Vessel Surveillance Capsule Withdrawal Schedule," October 22, 1997.

14. Letter from S. D. Bloom (NRC) to G. M. Rueger (PG&E), "Pacific Gas & Electric Company's Revision to the Reactor Vessel Surveillance Capsule Withdrawal Schedule for Diablo Canyon Unit No. 2 (TAC No. M99917)," February 10, 1998.

15. WCAP-15423, Revision 0, "Analysis of Capsule V from Pacific Gas and Electric Company Diablo Canyon Unit 2 Reactor Vessel Radiation Surveillance Program," September 2000.

16. Letter from D. Holland (NRC) to G. M. Rueger (PG&E), "Diablo Canyon Power Plant, Unit 1 (TAC No. MB5796) and Unit 2 (TAC No. MB5797) - Issuance of Amendment Revising Technical Specification 5.6.6 - Reactor Coolant System Pressure Temperature Limits Report," May 13, 2004.

Principal Contributor: James Medhoff

Date: July 17, 2006

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Diablo Canyon Power Plant, Units 1 and 2

cc:

NRC Resident Inspector Richard F. Locke, Esq.

Diablo Canyon Power Plant Pacific Gas & Electric Company c/o U.S. Nuclear Regulatory Commission P.O. Box 7442 P.O. Box 369 San Francisco, CA 94120 Avila Beach, CA 93424 City Editor Sierra Club San Lucia Chapter The Tribune ATTN: Andrew Christie 3825 South Higuera Street PC. Box 15755 P.O. Box 112 San Luis Obispo, CA 93406 San Luis Obispo, CA 93406-0112

Ms. Nancy Culver Director, Radiologic Health Branch San Luis Obispo State Department of Health Services Mothers for Peace P.O. Box 997414, MS 7610 P.O. Box 164 Sacramento, CA 95899-7414 Pismo Beach, CA 93448 Mr. James D. Boyd, Commissioner Chairman California Energy Commission San Luis Obispo County 1516 Ninth Street (MS 31)

Board of Supervisors Sacramento, CA 95814 1055 Monterey Street, Suite D430 San Luis Obispo, CA 93408 Mr. James R. Becker, Vice President Diablo Canyon Operations Mr. Truman Burns and Station Director Mr. Robert Kinosian Diablo Canyon Power Plant California Public Utilities Commission P.O. Box 56 505 Van Ness, Room 4102 Avila Beach, CA 93424 San Francisco, CA 94102 Jennifer Tang Diablo Canyon Independent Safety Field Representative Committee United States Senator Barbara Boxer ATTN: Robert R. Wellington, Esq. 1700 Montgomery Street, Suite 240 Legal Counsel San Francisco, CA 94111 857 Cass Street, Suite D Monterey, CA 93940

Regional Administrator, Region IV U.S. Nuclear Regulatory Commission Harris Tower & Pavillion 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-8064

March 2006

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Pacit7c Gas and anl 8 E/eet/ic Company David H. Dailey Diablo Canyon Power Plant Vice President and EO. Box 56

&neral Manager Avila Beach, CA 93424 August 23, 2005 805.545.4350 Fax:805.545.4234

PG&E Letter DCL-05-098

U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001

Docket. No. 50-275, OL-DPR-80 Docket No. 50-323, OL-DPR-82 Diablo Canyon Units 1 and 2 License Amendment Request 05-03 Request for Amendment to Recapture Low-Power Testing Time

Dear Commissioners and Staff:

In accordance with 10 CFR 50.90, enclosed is an application for amendment to Facility Operating License Nos. DPR-80 and DPR-82 for Units 1 and 2 of the Diablo Canyon Power Plant (DCPP), respectively. The enclosed license amendment request (LAR) would revise the expiration dates of the Units 1 and 2 facility-operating licenses to recapture low-power testing time. Specifically, the expiration date of each unit's full-power operating license (FPOL) would be revised to reflect a 40-year term measured from the date of issuance of the FPOL, as permitted by 10 CFR 50.51.

SECY-98-296, "Agency Policy Regarding Licensee Recapture of Low-power Testing or Shutdown Time for Nuclear Power Plants," dated December 21, 1998, and the associated Commission Voting Record and Staff Requirements Memorandum, dated March 30, 1999, established NRC policy regarding license recapture of low-power testing or shutdown time for nuclear power plants. By establishing this policy, the Commission has acknowledged that recapturing low-power testing time does not involve a significant hazards consideration. The Agency's policy bounds the proposed amendment request since this amendment request is similar to prior license recapture situations as described in SECY-98-296.

Enclosure 1 contains a description of the proposed change, the supporting technical analyses, and the no significant hazards consideration determination. Enclosure 2 contains a markup of the revised portions of the facility operating licenses for Units 1 and 2.

A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway - Comanche Peak >> Diablo Canyon >> Palo Verde >> South Texas Project c Wolf Creek ,OG l

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Document Control Desk PG&E Letter DCL-05-098 m8 August 23, 2005 Page 2

PG&E has determined that this LAR does not involve a significant hazard consideration as determined per 10 CFR 50.92. Pursuant to 10 CFR 51 .22(b), an environmental assessment does not need to be prepared since the proposed change does not involve a significant change in the types or in the amounts of any effluent that may be released offsite, or a significant increase in the individual or cumulative occupational radiation exposure.

The change in this LAR is not required to address an immediate safety concern.

PG&E requests approval of this LAR be assigned a medium priority for review and approval and requests that the amendments be issued no later than March 2006.

PG&E requests the LAR be made effective upon NRC issuance.

If you have any questions or require additional information, please contact Mr. Terence Grebel at (805)545-4160.

Sincerel

/'

David H. Oatley .

Vice President and General Manager

tlg/4160 Enclosures cc: Edgar Bailey, DHS Terry W. Jackson Bruce S. Mallett Diablo Distribution cc/enc: Girija S. Shukla

A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway >> Comanche Peak ¢ Diablo Canyon >> Palo Verde >> South Texas Project

• Wolf Creek

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PG&E Letter DCL-05-098

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION

) Docket No. 50-275 In the Matter of )Facility Operating License PACIFIC GAS AND ELECTRIC COMPANY ) No. DPR-80

. )

Diablo Canyon Power Plant )Docket No. 50-323 Units 1 and 2 )Facility Operating License

) No. DPR-82 P

AFFIDAVIT

David H. Oatley, of lawful age, first being duly sworn upon oath says that he is Vice President and General Manager of Pacific Gas and Electric Company, that he has executed license amendment request LAR 05-03 on behalf of said company with full power and authority to do so, that he is familiar with the content thereof, and that the facts stated therein are true and correct to the best of his knowledge, information, and belief.

David H. Oatley Wce President and General Manager

Subscribed and sworn to before me this 23rd day of August , by David H. Oatley, personally known to me or proved to me on the basis of satisfactory evidence to be the person(s) who appeared beforeme.

Notary public ' (

County of San Luis Obispo Innbsofy runuc - ounwuaa I

State of California 31 ( cemmwon# wwav5($on Lua oospc county

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Enclosure 1 PG&E Letter DCL-05-098 Page 1 of 14

EVALUATION

1.0 DESCRIPTION

This License Amendment Request (LAR) proposes to amend Full Power Operating Licenses (FPOLs) DPR-80 and DPR-82 for Units 1 and 2 of the Diablo Canyon Power Plant (DCPP), respectively.

The proposed changes would revise the Operating Licenses for Units 1 and 2 to recapture the low-power testing period such that expiration of the FPOL would occur 40 years from the date of issuance of the FPOL, as permitted by Title 10 of the Code of Federal Regulations (10 CFR) Part 50.51 .

2.0 PROPOSED CHANGE

The current expiration dates of the Unit 1 and 2 facility operating licenses are measured from the date of issuance of the low-power operating licenses (LPOLS) for each unit. The proposed amendments would revise the expiration date of the Unit 1 and 2 operating licenses to reflect a 40-year term measured from the date of issuance of each facility's FPOL. The Units 1 and 2 facility operating licenses would be revised as follows:

Facilitv ODeratinG Date of FPOL Current Revised License Issuance Expiration Expiration

DPR-80 11/02/1984 09/22/2021 11/02/2024 DPR-82 08/26/1985 04/26/2025 08/26/2025

Thus, the additional operating.period for Unit 1 would be just over 37 months.

The additional operating period for Unit 2 would be 4 months

The revised portions of the Unit 1 and 2 licenses are provided in Enclosure 2.

3.0 BACKGROUND

Section 103.c of the Atomic Energy Act of 1954 (AEA), as amended, provides that a license may be issued for a specific period not to exceed 40 years.

Section 104.b of the AEA does not identify a specific license term. However, 10 CFR 50.51 also specifies that each license will be issued for a fixed period of time not to exceed 40 years from the date of issuance. Also, 10 CFR 50.56 and 50.57 allow the issuance of an operating license pursuant to 10 CFR 50.51 after the construction of the facility has been substantially completed, in conformity with the construction permit and when other provisions specified in 10 CFR 50.57 are met.

I

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J The Commission issued the LPOLs for DCPP Units 1 and 2 on September 22, 1981 (DPR-76), and April 26, 1985 (DPR-81), respectively. In the LPOL for each unit, the licensee was only authorized to operate the respective unit up to 5 percent of rated thermal power.

On November 2, 1984, the Commission issued FPOL Facility Operating License No. DPR-80 for Unit 1, which superceded the Unit 1 LPOL. The FPOL included an expiration date of April 23, 2008. On August 26, 1985, the Commission issued FPOL FaCility Operating License DPR-82 for Unit 2, which superceded the Unit 2 LPOL. The Unit 2 FPOL included an expiration date of December 9, 2010. Both FPOLs were issued underSection 104.b of the AEA.

The initial FPOL term for both units was 40 years, commencing with the issuance of the construction permit on April 23, 1968, for Unit 1 and December 9, 1970, for Unit 2. On March 1, 1995, the Commission issued Amendment No. 97 to Facility Operating License No. DPR-80 and Amendment No. 96 to Facility Operating License No. DPR-82 for DCPP Units 1 and 2, respectively, to extend the operating license dates to September 22, 2021, for Unit 1, and to April 26, 2025, for Unit 2, or 40 years after the date of issuance of the LPOLs. The proposed .

amendments would revise the Units 1 and 2 licenses so that the licenses would expire40years from the date of issuance of the respective FPOLs.

In summary, the proposed amendments to the Units 1 and 2 facility operating licenses recapture the time between issuance of the LPOL and the FPOL for each unit. SECY-98-296, "Agency Policy Regarding Licensee Recapture of Low-Power Testing or Shutdown Time for Nuclear Power Plants," dated December 21, 1998, and the associated Commission Voting Record and Staff Requirements Memorandum (SRM), dated March 30, 1999, established NRC policy regarding license recapture of low-power,testing periods (Reference 7.1).

In the voting record and SRM, the Commission approved the staff's recommendation to allow Grand Gulf Nuclear Station to recover the time spent in low-power testing before their FPOL was issued. The Commission also approved the granting of similar requests from other licensees provided the 40-year license term began with the issu8nce of an LPOL or construction permit and a separate FPOL was subsequently issued.

In the case of DCPP, the 40-year FPOL term for Units 1 and 2 began with the issuance of the LPOLs. Each LPOL was subsequently superceded by the issuance of a FPOL. Therefore, the FPOL for Units 1 and 2 is bounded by the Commission's policy allowing license recapture of low-power testing and permitting a 40-year license term. No exemption from 10 CFR 50.51 is required.

The proposed amendments do not constitute license renewal and are therefore not subject to the requirements of 10 CFR 54.

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4.0 TECHNICAL ANALYSIS

The request for amendments to the present operating licenses is based on the fact that a 40-year service life was considered during the design and construction of the plant. Because some components will foreseeably wear out during the plant's operating lifetime, design features were incorporated in the plant, which provide for inspectability of structures, systems, and components. In particular, surveillance, inspection, and maintenance procedures were implemented in accordance with the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code) for inservice inspection and testing of pumps and valves and in accordance with the plant's Technical Specifications (TSS).

The specific provisions and requirements for ASME Code testing are set forth in 10 CFR 50.55a. in total, these procedures provide assurance that any equipment degradation will be identified and addressed during the operating life of the plant, including the proposed additional operating period.

4.1 Reactor Pressure Vessel

The DCPP reactor pressure vessels for Units 1 and 2 were designed and fabricated in accordance with the 1965 Edition through Summer 1966 Addenda for Unit 1 and the 1968 Edition for Unit 2, of the ASME Boiler and Pressure Vessel Code, Section Ill, "Nuclear Power Plant Components." .They were designed for transients considered to envelope design conditions over a 50-year operating period. To ensure the continued integrity of the vessels during operation, an lnservice Inspection (ISI) Program has been in place since plant startup.

The effects of neutron radiation embrittlement of the vessel beltline region are considered in the design and operation of the units. Compliance with all NRC

'regulations governing vessel integrity has been documented most recently in the

. NRC staff safety evaluation related to Amendment No. 133 to Facility Operating License No. DPR-80 and Amendment No. 131 to Facility Operating License No.

DPR-82, dated May 3, 1999. In addition, PG&E~has instituted an Embrittlement Management Plan to manage reactor vessel embriltlement throughout the entire operating life of DCPP.

Pressurized Thermal Shock

Following Cycle 1 for each unit, the neutron fluence at the reactor vessel inner wall was reduced by installing increasingly lower neutron leakage cores, thus decreasing the reactor vessel rate of embrittlement and prolonging vessel life.

The DCPP reactor pressure vessel beltline materials have been evaluated according to the NRC's Pressurized Thermal Shock (PTS) screening criteria defined in 10 CFR 50.61. The Reference Temperature for Pressurized Thermal Shock (RTpTS). has been calculated for each weld metal and base metal in the

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DCPP beltline regions for neutron fluences corresponding to 40 operating years.

The RTPTS for all materials will not exceed the screening limit of 270°F for base metal and longitudinal welds and 300°F for circumferential welds. The most recent NRC review of DCPP Units 1 and 2 confirming conformance with the current PTS rule, is documented in Reference 7.2 of this LAR. .

Based on a conservative fluence projection for 40 operating years, DCPP will also meet the requirements of 10 CFR 50, Appendix G. Charpy Upper shelf Energies were determined (DCPP Final Safety Analysis Report (FSAR) Update Tables 5.2-19A, 5.2-198, 5.2-21A, and 5.2-218) in accordance with Regulatory Guide (RG) 1.99, Revision 2. All DCPP beltline materials will remain above the 50 ft-lb Charpy Upper Shelf Energy fracture toughness requirement for more than 40 operating years. In addition, reactor vessel pressure-temperature limits will meet 10 CFR 50, Appendix G requirements for 40-year operation without requiring plant modification or imposing operational restraints.

Material Surveillance Program

The toughness properties of the reactor vessel beltline material will be monitored throughout the proposed 40-year operating license terms with a material surveillance program that meets the requirements of 10 CFR 50, Appendix H.

The original surveillance program for DCPP Unit 1 complies with ASME E 185-70, the standard in effect When the vessel was manufactured.

Although the Unit 1 surveillance program was designed prior to the existence of 10 CFR 50, Appendix H, that program does contain the significant features required for later surveillance programs and will ensure vessel embrittlement is effectively monitored throughout the requested license period. The original program includes a total of eight surveillance capsules. Three of the eight capsules contain the limiting weld metal and base metal, correlation monitor material, dosimeters, and thermal monitors. The remaining five capsules contain the limiting base metal, but no weld metal. All base metal charpy specimens in the capsules are longitudinally oriented. '

The Unit 2 surveillance program includes six capsules and conforms to ASTM E 185-73. All capsules contain the limiting weld metal. The base metal specimens in the capsules are not from the limiting plate, but were machined from an adjacent plate with similar chemistry, the same heat treatment, and similar level of embrittlement at plate end-of-life as the limiting plate.

As discussed in the "Pressure and Temperature Limits Report (PTLR) for Diablo Canyon," submitted in PG&E Letter DCL-05-016, dated February 28, 2005 (Reference 7.3), Units 1 and 2 are currently using the same heatup and cooldown limits, which are based on the limiting-unit surveillance program results. To date there have been three surveillance capsules removed and

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analyzed from the Unit 1 reactor vessel and four from the Unit 2 reactor vessel.

The Unit 1 surveillance results are currently limiting since the calculated delta Reference Temperature for Nondestructive Testing (RTNDT) data scatter does not fall within the two standard deviations of the predicted data as required by RG 1.99, Revision 2 criteria for "credible" surveillance data..Therefore, theDCPP heatup and cooldown limits established in the PTLR are currently based on the generic limiting CF values in Tables 1 and 2 of 10 CFR 50.61 and the chemistry values provided by CE Report CE NPSD 1039, Revision 2 (Reference 7.4). If the RG 1.99 credibility criteria are met upon future surveillance capsule withdrawal and evaluation, then the RG 1.99 Position C.2 will be utilized using plant specific surveillance data.

In order to enhance the current surveillance programs, a supplemental surveillance program was implemented for Unit 1 beginning with Cycle 6. The supplemental program consists of four new surveillance capsules that contain the limiting base metal and weld metal specimens that are representative of the Unit 1 limiting weld. This supplemental program will provide additional data to -

better assess and manage vessel embrittlement issues during the plant operating life.

Additional measures to monitor DCPP Units 1 and 2 vessel fluence are provided in the Reactor Cavity Neutron Dosimetry Program. This voluntary program has been in effect since initial criticality and consists of irradiating and evaluating reactor cavity dosimetry, which includes multiple foil sensor sets and axial flux gradient wires attached to the metal reflective insulation surrounding the reactor vessel. Results obtained are used to confirm and complement surveillance capsule data.

The overall program to monitor reactor vessel beltline materials' is thorough and comprehensive. lt meets all applicable regulatory guidance and will provide continuous information relevant todetermining the degree of embrittlement of beltline materials over the proposed 40-year operating license terms. This program provides reasonable assurance that the reactor pressure vessel will be in conformity with the applicable provisions of NRC rules and regulations for the proposed additional operating period.

4.2 Structures

The Category I structures at DCPP were designed and constructed in accordance with 10 CFR 50. The major codes and specifications used in the design and construction of the Category I structures are discussed in the FSAR Update, Chapter 3.0. The design basis, fabrication, construction, and implementation of quality assurance criteria for the plant were reviewed by the NRC staff when the plant was being licensed for low-power operation. Structures and associated protective coatings are periodically inspected and maintained in

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accordance with 10 CFR 50.65, "Requirements for monitoring the effectiveness of maintenance at nuclear power plants," to ensure continued structural integrity.

Industrial experience with Category I structures confirms a service life in excess of 40 years may be anticipated.

Criteria that were used in the analysis, design, and construction of Category I structures account for anticipated loadings and postulated conditions that may be imposed on the structures during their service lifetime, including the proposed additional operating periods.

The above program provides reasonable assurance that the Category I structures will be in conformity with the applicable provisions of NRC rules and regulations for the proposed additional operating period.

4.3 Mechanical Equipment

DCPP mechanical.equipment is designed, licensed, and constructed for a 40-year service life. The reactor coolant system components and support systems were analyzed for the integrated effects of radiation damage and cyclic loadings (with added margin) that could reasonably be expected to occur in a 40-year operating lifetime measured from issuance of the FPOL. Surveillance and maintenance practices were implemented in accordance with the ASME Code for ISI and inservice Testing of Pumps and Valves, a maintenance program satisfying 10 CFR 50.65 requirements, and the TS. The TSs are part of the plant's operating license and have been approved by the NRC, as are all subsequent amendments to the TSs. The specific provisions and requirements for ASME Code testing are set forth in 10 CFR 50.55a.

Surveillance, maintenance, and testing requirements for mechanical equipment are in place at the plant to verify operability or to detect degradation and ensure that the equipment that does degrade is replaced or other corrective actions are taken. In addition, subcomponents such as nonmetallics (e.g., gaskets and o-rings) are inspected and replaced as necessary, as part of routine maintenance in order to ensure the design life of the equipment.

CoMpliance with the codes, standards, and regulatory requirements to which mechanical equipment were analyzed, constructed, tested, and inspected provides adequate assurance that the structural integrity of equipment important to safety will be maintained during the Units 1 and 2 operating lifetime and during the additional period proposed.

4.4 Electrical Equipment

Environmental qualification (EQ) is a rigorous program of testing, analysis, and maintenance to confirm that electrical equipment relied upon in the event of an

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accident will be capable of performing its design safety function, despite exposure to the harsh environment resulting from an accident. The DCPP EQ Program complies with 10 CFR 50.49, "Environmental Qualification of Electric Equipment Important to Safety for Nuclear Power Plants." As applied to DCPP, 10 CFR 50.49 requires electrical equipment important to safety and located in a harsh environment to be environmentally qualified, at a minimum, in accordance with IEEE Trial-Use Standard 323-1971 and Category II positions in .

NUREG-0588 ("For Comment" version, dated December 1979). In accordance with 10 CFR 50.49(1 ), replacement equipment (for equipment that is required to be environmentally qualified) is required to be qualified in accordance with IEEE Standard 323-1974 and Categoryl positions in NUREG-0588 ("For Comment" version, dated December 1979), unless there are sound reasons to the contrary.

The DCPP EQ Program is a continuing program. The master list of equipment to be qualified is maintained as a controlled engineering drawing and is revised as plant design changes are implemented. Detailed EQ files document the results of the testing and analysis that substantiate that the equipment will perform as required in accident environments. Surveillance activities are performed to detect adverse trends in aging or performance. Maintenance procedures assure that the qualified configuration of equipment is restored after maintenance.

Equipment that is not qualified for the entire 40-year operating license term (FPOL) is refurbished or replaced prior to exceeding its qualified life.

Supplements 15 (dated September 1981) and 31 (dated April 1985) to the DCPP Safety Evaluation Report (SER) provide the NRC staffs evaluation of the DCPP EQ Program. In Supplement 31 to the SER, the NRC staff concluded that the DCPP EQ Program is acceptable and that compliance with 10 CFR 50.49 has been demonstrated. Supplement 31 also noted that the DCPP EQ Program had been expanded to include RG 1.97 Category 1 and 2 instrumentation. The NRC staff's findings are premised on the continuing nature of the DCPP EQ Program (e.g., replacement of equipment prior to expiration of its qualified life), without regard to the length of the remaining license period.

In summary, the DCPP EQ Program ensures that electrical equipment important to safety within the scope of 10 CFR 50.49 will be adequately qualified and maintained, and thus will be capable of performing required safety functions throughout the proposed 40-year FPOL terms.

4.5 Quality Assurance and Maintenance Programs

The Units 1 and 2 Quality Assurance (QA) Program continuously assess how programs are implemented, procedures are followed, and operating requirements are met. This oversight includes the maintenance programs, which assure that equipment remains operable or corrective actions are taken. The maintenance programs must be performed in accordance with 10 CFR 50.65, "Requirements

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for monitoring the effectiveness of maintenance at nuclear power plants."' .

Assessments of the QA Program and maintenance programs show that these programs remain acceptable. The QA Program meets the requirements of Appendix B to 10 CFR 50.

Therefore, implementation and use of these programs at DCPP provides reasonable assurance that equipment important to safety will, for the proposed additional operating period, be in conformity with the provisions of the rules and regulations of the NRC, and the DCPP licenses.

5.0 REGULATORY ANALYSIS

5.1 No Significant Hazards Consideration

PG&E has evaluated whether or,not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

I 1. Does the proposed change involve a significant increase in the

probability or consequences of an accident previously evaluated?

Response: No.

The proposed additional operating license periods do not affect the probability or consequences of an accident Previously evaluated since they require ho physical change in the plant equipment or operating procedures and the Final Safety Analysis Report (FSAR) Update safety analyses are based on 40-year full power operation. Surveillance and maintenance practices, as well as other programs such as environmental qualification of equipment, ensure timely identification and correction of any degradation of safety-related plant equipment. The long-term integrity of the reactor vessels has been evaluated using currently acceptable NRC calculational methods and best available Diablo Canyon Power Plant (DCPP) specific data. The evaluation results demonstrate that both reactor vessels are safe for normal operations in excess of40years.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different accident from any accident previously evaluated?

Response: No.

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The possibility of a new or different kind of accident is not created by the proposed additional operating periods since at least 40 years of full power operation was assumed in the design and construction of DCPP Units 1 and 2. The plant maintenance programs are also designed to both maintain and determine the need to replace"safety-related components.

These programs will continue to be applied as they are presently to assure safe operation.

Therefore, the proposed change does not create the possibility of a new or different accident from any accident previously evaluated.

3. Does the proposed change involve.a significant reduction in a margin of safety?

Response: No.

The proposed additional operating periods do not involve a significant reduction in a margin of safety since, as is the case with present operation, degradation of safety-related equipment will be identified and corrected by ongoing surveillance and maintenance practices. Existing programs, routine maintenance, and compliancewithTechnical Specifications assure that an adequate. margin of safety is maintained.

These activities will remain in effect for the duration of the proposed additional operating periods.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above evaluation, PG&E concludes that the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c) and, accordingly, a finding that the amendments involve "no significant hazards consideration" is justified.

5.2 Applicable Requlaton Requirements/Criteria

The proposed change has been evaluated to determine whether applicable regulations and requirements continue to be met. PG&E has determined that the proposed license amendments do not require any exemptions or relief from regulatory requirements and does.not affect conformance with 10 CFR 50, Appendix A, "General Design Criteria for Nuclear Power Plants." Applicable regulatory requirements will continue to be met, adequate defense-in-depth will be maintained, and sufficient safety margins will be maintained. The applicable regulatory guidance of SECY-98-296, "Agency Policy Regarding Licensee Recapture of Low-Power Testing or Shutdown Time for Nuclear Power Plants" is met. The

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applicable regulatory requirements are addressed in the individual sections of the technical analysis.

Based on the considerations discussed above and within the individual sections of the technical analysis: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security of the health and safety of the public.

6.0 ENVIRONMENTAL CONSIDERATION

The following is a summary of the environmental considerations associated with the proposed low power license recapture. There are no significant environmental considerations involved with the proposed action.

The proposed amendments do not affect the design or operation of the plant, do not involve any modifications to the plant or any increase in the licensed power level for the plant, and will not create any new or unreviewed environmental impacts that were not previously considered in the Final Environmental Statement (FES) related to operation of DCPP. The proposed license amendments also will not significantly increase the probability or consequences of accidents, do not involve any changes in the types of effluents that may be released off site, and do not increase occupational or public radiation exposures.

Therefore, there are no significant radiological environmental impacts associated with the proposed action.

The FES fully evaluated the environmental impacts of generating power at DCPP based on 40 years of operation. The FES, in general, assesses various aspects associated with operation of DCPP in terms of annual impacts and balances these against the anticipated annual energy production benefits. This assessment is not changed by the proposed amendments.

Background

As part of the original licensing of DCPP, PG&E prepared and submitted an environmental report to the NRC (Reference 7.5) addressing the potential impact of the operation of DCPP on the surrounding environment. The NRC reviewed this report and issued a FES in 1973 (Reference 7.6). Both the environmental report and the FES concluded that operation of the DCPP would have no significant adverse environmental effects on the areas surrounding DCPP.

PG&E updated this environmental information as part of the construction period recapture LAR (Reference 7.7) and concluded that there were no significant

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adverse environmental effects associated with the construction period recapture.

The NRC review of the construction period recapture also determined that there was no significant adverse environmental impact associated with this additional operating period (Reference 7.8).

Environmental Impacts of the Low Power Testinq Period Recapture

The offsite exposure from releases during postulated accidents has been previously evaluated in the FSAR Update. The results are acceptable when compared with the criteria defined in 10 CFR 100. This conservative design-basis evaluation is a function of four parameters: (1) the type of accident postulated, (2) the radioactivity calculated to be released during the accident, (3) the assumed meteorological conditions at the~site, and (4) the population distribution versus distance from the plant. An environmental assessment is also provided in the FES. The type of accidents and the calculated release does not change as a result of the proposed action. The site meteorology as defined in Chapter 2 of the FSAR Update is essentially constant. The population size and distribution has not changed significantly from that evaluated in the FES.

The expected annual occupational exposure for the proposed extended period of operation does not change previous conclusions presented in the FES for average annual occupational exposure. The actual annual occupational exposure of workers at the plant is reported routinely in the Occupational Radiation Exposure Report submitted to the NRC. Through continued implementation of as low as is reasonably achievable (ALARA) and other programs, projected collective occupational exposure for the plant through the proposed extended period will continue to remain sig niltca ntly below the exposures considered in the FES.

The offsite exposure from releases during routine operations was also previously evaluated in the FES. During the low-power license, the plant was restricted to no more than 5 percent of rated power and the generation of radioactivity at the plant was significantly smaller than would have occurred if the plant were at full-power operation. In addition, routine releases to the environment are governed by 10 CFR 20, which states that such releases should be ALARA. The annual Radioactive Effluent Release Reports provide an annual assessment of radiation dose as a result of efHuents released from the plant. These reports show that actual releases of radioactive liquids and gases have historically been lower than those estimated in the FES. The volume of radwaste generated at the plant from the routine processing of radioactive liquids (filters and resins), and from routine maintenance on equipment, has significantly decreased from values considered in the FES due to waste volume reduction technology improvements.

The plant Annual Radiological Environmental Operating Program is used to monitor the effect of plant operation on the environment. This is accomplished

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. Page.12 of 14

by continuously measuring radiation levels and airborne radioactive materials and periodically measuring amounts of radioactive materials in samples at various locations surrounding the plant. Continued environmental monitoring and surveillance under this program will contin.ue to ensure early detection of any increase in exposures over the proposed additional operation period. Therefore, the proposed amendments do not change previous conclusions presented in the FES on annual public doses.

With regard to the environmental impacts of the uranium fuel cycle, all fuel at DCPP is bounded by the impacts reported in Table S-4 of 10 CFR 51 .52. Thus this generic assessment is bounding for DCPP. To provide for the storage of additional spent fuel assemblies beyond the licensed capacity of the DCPP spent fuel pools, dry cask storage was licensed under a site-specil9c 10 CFR 72 license (Docket No. 72-26). Onsite storage capacity in the spent fuel pools and in dry cask storage will be adequate for the extended period of proposed operation.

With regard to potential nonradiological impacts, there are no impacts beyond these previously considered for 40 years of operation, including potential impacts on historical sites and impacts due to nonradiological plant effluents. Therefore, there are no significant nonradiological environmental impacts associated with the proposed action.

Accordingly, there are no significant environmental impacts associated with the proposed action.

Conclusions

The environmental affects associated with the proposed license amendments are enveloped by the original and recapture environmental reviews (References 7.5 and7.6), since these reviews assumed 40 years of full-power operation. The impacts associated with the additional periods of operation have thus been previously addressed. Furthermore, some of the environmental information related to 40-year plant operation, such as spent fuel storage options and no project alternatives, was updated and confirmed to have no significant adverse environmental effects as part of its Diablo Canyon Independent Spent Fuel Storage Installation (ISFSI) Environmental Report (ER) (Reference 7.9). The lSFSl ER was reviewed by the NRC (Reference 7.10).

PG&E has evaluated the proposed amendment and has determined that the proposed amendment does not involve: (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51 .22(b), no environmental

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impact statement or environmental assessment need be prepared in connection with the proposed amendment.

7.0 REFERENCES

7.1 SECY-98-296, "Agency Policy Regarding Licensee Recapture of Low-Power Testing or Shutdown Time for Nuclear Power Plants," dated December 21, 1998, and the associated Commission Voting Record and Staff Requirements Memorandum, dated March 30, 1999. .

7.2 NRC Letter,Fracture Touqhness/Pressurized Thermal Shock (10 CFR 50.61) (TAC Nos. 59951 and 59952), October 30, 1987.

7_3 PG&E Letter DCL-05-016, Pressure and Temperature Limits Report (pTLR-1). Revision 4, Diablo Canyon Power Plant Units 1 and 2, February 28, 2005.

7.4 CE NPSD-1039, Best Estimate Copper and Nickel Values in CE Fabricated Reactor Vessel Welds, Revision 2, and CE NPSD-1039 Appendix A, CE Reactor Vessel Weld Properties Database Volumes 1 and 2, Revision 2. .

7.5 Environmental Report, Units 1 and 2, Diablo Canyon Site, PG&E, July 1971, and supplements No. 1, November 1971, No. 2, July 1972, and No. 3, August 1972.

7.6 Final Environmental Statement Related to the Nuclear Generating Station, Diablo Canyon Units 1 an.d 2, USAEC, May 1973.

7.7 PG&E License Amendment 92-04,40-Year Operatinq License Application, July 9, 1992.

7.8 NRC Letter to PG&E,License Amendment No. 97 to Facilitv Operating License No. DPR-80 and Amendment No. 96 to Facility Ooeratinu License No. DPR-82, March 1, 1995.

7.9 Diablo Canyon Independent Spent Fuel Storage Installation

'Environmental Report, PG&E, December 2001.

7.10 NRC Letter to PG&E, Environmental Assessment and Findinq of No Sicnificant Impact Related to the Construction and Operation of the Diablo_

Canvon Independent Spent Fuel Storage Installation (TAC NO. L233991, dated October 24, 2003. '

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8.0 Precedent

The proposed license changes are consistent with SECY-98-296, "Agency Policy Regarding Licensee Recapture of Low-Power Testing or Shutdown Time for Nuclear Power Plants," dated December 21, 1998, and the associated Commission Voting Record and Staff Requirements Memorandum, dated March 30, 1999, which established NRC policy regarding license recapture of .

low-power testing or shutdown time for nuclear power plants.

t.

I

/'

s

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Enclosure 2 PG&E Letter DCL-05-098

PACIFIC GAS AND ELECTRIC COMPANY DIABLO CANYON NUCLEAR POWER PLANT. UNIT2 DOCKET N0. 50-323 FACILITY OPERATING LICENSE LiceNse No. DPR-82

I. Term of License

This License is effective as of the date of issuance and shall expire at midnight on A9141 August 26, 2025. I

PACIFIC GAS AND ELECTRIC COMPANY DIABLO CANYON NUCLEAR POWER PLANT. UNIT 1 DOCKET N0. 50-275 FACILITY OPERATING LICENSE License No. DPR-80

I. Term of License

This License is effective as of the date of issuance and shall expire at midnight onSeptember 22, 2021November 2, 2024. I

.I ER-187 Case: 23-3884, 03/25/2024, DktEntry: 23.2, Page 155 of 222

Paci#'c Gas and url8 Electric Company

David H. Oakley Diablo Canyon Power Plant VicePresident and EO. Box 56 Genera! Manager Avila Beach, CA 93424

805.545.4350 May 13, 2003 Fax:805.545.4234

PG&E Letter DCL-03-052

U.S. Nuclear Regulatory Commission ATTN; Document Control Desk Washington, DC 20555-0001

Docket No. 50-275, OL-DPR-80 Diablo Canyon Unit 1 Diablo Canyon Unit 1 Reactor Vessel Material Surveillance Program Capsule V Technical Report

Dear Commissioners and Staff:

On May 13, 2002, surveillance Capsule V was withdrawn from the Diablo Canyon Power Plant (DCPP) Unit 1 reactor vessel and shipped to Westinghouse for testing.

Pursuant to 10 CFR 50 Appendix H, Part IV.A, this submittal provides the Capsule V technical report. Enclosure B is the Westinghouse technical report, WCAP-15958, Revision 0, "Analysis of Capsule V from Pacific Gas and Electric Company Diablo Canyon Unit 1 Reactor Vessel Radiation Surveillance Program."

Pursuant to 10 CFR.50.61(b)(1) and 10 CFR 50 Appendix H, Part IV.C, included herein are the results of: (1) the pressurized thermal shock (PTS) evaluation, (2) the reactor coolant system (RCS) pressure/temperature (P/T) limit curve evaluation, (3) the low temperature overpressure (LTOP) setpoint evaluation, and the upper shelf energy (USE) evaluation. .

Evaluation

Table D-2 ofWCAP-15958 (Enclosure B) summarizes the best-fit surveillance capsule data chemistry factor (CF) evaluation. The Capsule V plate and weld data point resulted in new best-fit curves. As a result, the Capsule S plate data point now has a scatter value that exceeds a one-sigma value of 17°F, and the Capsule Y weld data point now has a scatter value that exceeds a one-sigma value of 28°F.

Therefore, neither the plate nor the weld data meet Regulatory Guide (RG) 1.99, Revision 2, criterion 3, and the data are not deemed to be credible. Also, the plate and weld CF values calculated from the surveillance data are less than the corresponding RG 1.99 position 1.1 values. Thus, the WCAP-15958 Table D-2 CF values derived for the plate and weld metal were not used in this evaluation.

A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway . Comanche Peak Diablo Canyon Palo Verde

• South Texas Project Wolf"Creek. 74008

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For the Unit 1 end of operating license (EOL) at approximately 32 effective full power years (EFPY) on September 22, 2021, the limiting RTprs values calculated and their respective 10 CFR 50.61 screening limits are:

RTpTs(Weid 3-442C) = 250.9°F, which is <270°F plate or axial weld limit

RTpT3(Wid 9-442) = 192.8°F, which is <30D°F circumferential weld limit

Therefore, the PTS screening limits are met at EOL. PG&E performed this evaluation.

Table 1 of Enclosure A shows that the DCPP Unit 1 adjusted reference temperatures projected to 16 EFPY are less than the quarter thickness (T/4) and three quarter thickness (3T/4) values assumed for the existing 16 EFPY P/T limit curves and LTOP setpoint found in the DCPP pressure and temperature limits report (PTLR). The P/T limit curves and LTOP setpoints for 16 EFPY remain bounding and valid. This evaluation was not included in the scope of WCAP-15958.

Instead, PG&E performed this evaluation with the results shown in Table 1.

The Westinghouse USE evaluation for the surveillance capsule materials is provided in Table C-1 ofWCAP-15959. Appendix G of 10 CFR 50 requires that the USE remain > 50 ft-lb throughout the life of the vessel at T/4. In addition, PG&E calculated the USE for the vessel materials not in the capsule. The most limiting (minimum) T/4 USE at EOL, (approximately 32 EFPY), is 61 .9 foot-pounds (ft-lbs). This is predicted to occur for axial weld 3-4420. Thus the 50 ft-Ib minimum requirement is met for all Unit 1 vessel materials at EOL.

Conclusion

In conclusion, the results of the specimen testing show that the limiting vessel beltline plate and weld material are behaving in accordance with previous predictions. Consequently, the results from Capsule V do not indicate any changes needed to the LTOP setpoints or P/T curves currently approved. Capsule v is not the last planned capsule to be evaluated in the DCPP Unit 1 surveillance program.

In PG&E Letter DCL-02-079, "License Amendment Request 02-04, Revision of Technical Specification 5.6.6 - Reactor Coolant System Pressure and Temperature Limits Report," dated July 31, 2002, PG&E requested NRC review and approval of PG&E's proposed application of.the PTLR methodology that will allow PG&E to calculate new P/T and LTOP limits without prior staff approval. As required. by Technical Specification 5.6.6(c), PG&E will also submit the revised PTLR, including data from the Capsule V report, when the PTLR is issued, upon approval of the PTLR methodology.

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Sincerely, 8/V @-

David H. Oatley VicePresident and General Manager- DiabloCanyon

SWH/3664/A0556643 Enclosure

cc: David L. Proulx cc/enc: Ellis W. Merschoff Girija S. Shukla

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March 30, 1999

MEMORANDUM William D. Travers TO: Executive Director for Operations Karen D. Cyr General Counsel FROM : Andrew L. Bates, Acting Secretary /s/

SUBJECT:

STAFF REQUIREMENTS - SECY-98-296 - AGENCY POLICY REGARDING LICENSEE RECAPTURE OF LOW-POWER TESTING OR SHUTDOWN TIME FOR NUCLEAR POWER PLANTS

The Commission has approved the staff's plans to grant the Grand Gulf license amendment to amend the expiration date of the license to recover the time spent in low power testing before receiving the Full Power Operating License (FPOL). The Commission has also approved the granting of similar requests from other licensees provided that the 40-year license term began with issuance of a Low Power Operating License (LPOL) and a separate FPOL was issued.

The Commission has approved the staff's plan to continue to grant license amendment requests to amend the expiration date of the license to recover time spent in construction in cases where the 40-year license term began with the construction permit date, and has approved the staff's recommendation to deny license amendment requests to amend the expiration date of the license to recover time spent in a shutdown condition where the shutdown commenced after the full power license was issued.

The Office of the General Counsel should develop a legal analysis for the Commission regarding those plants in which the LPOL was amended to allow full-power operation and no separate FPOL was issued. Specifically, the OGC analysis should provide the following :

Is it consistent with the Atomic Energy Act (AEA) to conclude that the initial operating license begins to run when the staff authorizes full-power operation, irrespective of whether the grant of that authority was accomplished through a separate license (FPOL) or through an amendment to the LPOL?

What legislative approaches are recommended if the analysis identifies issues for which legislative proposals should be considered ?

(OGC) (SECY Suspense:5/28/99)

CC : Chairman Jackson Commissioner Dicus Commissioner Diaz Commissioner McGaffigan Commissioner Merrifield OGC CIO CFO OCA OIG OPA Office Directors, Regions, ACRS, ACNW, ASLBP (via E-Mail)

PDR DCS

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December 21, 1998 SECY-98-296

FOR: The Commissioners

FROM: William D. Travers /s/

Executive Director for Operations

SUBJECT:

AGENCY POLICY REGARDING LICENSEE RECAPTURE OF LOW-POWER TESTING OR SHUTDOWN TIME FOR NUCLEAR POWER PLANTS

PURPOSE:

To obtain Commission approval of a policy issue concerning recapture of low-power testing or shutdown time for nuclear power plants not in commercial operation because of unusual, unforeseen, or exigent circumstances.

SUMMARY

The Nuclear Regulatory Commission (NRC) has received a request from Entergy Operations, Inc., the licensee for the Grand Gulf Nuclear Station (GGNS), to amend the expiration date of the GGNS license. The proposed amendment would recover time the plant spent in an extended low-power testing condition before receiving a full-power operating license (FPOL).

Because a number of legal, policy, and technical issues are associated with granting the amendment request, the staff wishes to inform the Commission of this license amendment request and other types of recapture situations.

BACKGROUND:

On June 16, 1982, GGNS was issued a low-power operating license (LPOL) to operate up to 5-percent rated power. After a number of technical specification and startup testing issues were resolved, the Commission amended the LPOL on August 31, 1984, to allow operation up to 100-percent rated power. In response to a court challenge to the amendment, the Commission

Contact:

Claudia M. Craig, NRR 301-415-1053

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subsequently directed the staff to issue a separate FPOL to GGNS. Mississippi Power & Light Co., et al. (Grand Gulf Nuclear Station, Unit 1), CLI-84-19, 20 NRC 1055 (1984). On November 1, 1984, the staff issued to GGNS an FPOL with an expiration date of June 16, 2022, 40 years from the date of issuance of the LPOL. Additional background regarding the period GGNS was in low-power testing is available in CLI-84-19 and the Director's Decision Under 10 CFR 2.206 inMississippi Power & Light Co., et al. (Grand Gulf Nuclear Station Unit 1), 20 NRC 788 (1984).

The GGNS amendment requests approximately a 28.5-month extension of the expiration date of the FPOL to November 1, 2024, 40 years from the date of issuance of the FPOL. In its application, the licensee argues that in accordance with the Commission's regulations in 10 CFR 50.51, the FPOL term should be 40 years from the date of issuance of the FPOL, not 40 years from the date of issuance of the LPOL. That regulation states that "[e]ach license will be issued for a fixed period of time to be specified in the license but in no case to exceed 40 years from the date of issuance." It does not differentiate between LPOLs and FPOLs.

Additionally, the licensee stated that the 28.5-month period between issuance of the LPOL and the FPOL was unique and the low-power license period was not intended to be part of the licensed 40-year life of the facility. Rather, this period was intended to confirm design adequacy in an operational setting. After confirming design criteria, the 40-year license period was to begin with full-power licensing and operation.

Licenses have been issued under two separate sections of the Atomic Energy Act of 1954, as amended (AEA): a commercial license under Section 103 and a research and development license under Section 104b. Prior to the 1970 amendments to the AEA, a Section 103 license required a finding of "practical value." The Atomic Energy Commission (AEC) would have based a finding of "practical value" for a type of reactor on a reliable estimate of its economics, based upon a demonstration of the technology and plant performance. In addition, after an AEC finding of "practical value" for a particular type of reactor, licenses issued under Section 103 were subject to a prelicensing review to determine if the proposed license would tend to create or maintain a situation inconsistent with antitrust laws. At that time, the AEC did not believe that it had sufficient information to make the "practical value" finding and all licenses were issued under Section 104b. In 1970, the AEA was amended to abolish the requirement of a finding of practical value and stated that any license issued for a utilization or a production facility for industrial or commercial purposes must be issued under Section 103. Note, however, that the operating license for a facility whose construction permit had previously been issued under Section 104b would likewise be issued under Section 104b, as stated in Section 102b of the AEA. The AEA does not specifically identify a license term for Section 104b licenses, although the AEA does restrict Section 103 licenses to 40 years. Section 50.51 does not distinguish between licenses issued under Sections 103 and 104b.

The Commission's practice with respect to issuance of operating licenses has varied. At one time, the staff issued provisional operating licenses (under Section 104b), followed by a full-term operating license. In addition, the staff, in some cases, issued an LPOL which was amended to allow full-power operation. The current practice is to issue an LPOL followed by a separate FPOL.

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Over the years, the date from which the 40-year license term began has also varied.

Originally, the forty year term began with issuance of the construction permit. Most plants in such circumstances have requested and been authorized to recapture the construction period in their operating licenses. Today, the majority of 40-year license terms begin with issuance of the LPOL. For some plants, however, the 40-year license term begins from the date of issuance of the FPOL. In most cases where the 40-year term begins with the LPOL, the time period between issuance of the LPOL and the FPOL is relatively short (several months),

therefore, applying the 40-year license term to the FPOL date versus the LPOL date usually has limited benefit.

DISCUSSION:

Grand Gulf

The staff has reviewed the GGNS request and identified a number of issues associated with the proposed recapture. These issues are discussed in more detail below.

First, it appears that extending the expiration date for GGNS is not inconsistent with the regulations. GGNS was issued a separate FPOL, and the request for the license term of 40 years from the FPOL does not contravene 10 CFR 50.51 .

Second, the licensee's technical position for the recapture is based on the interpretation that a 40-year service life was considered during the design and construction of the plant. The licensee believes that the original intent of the regulations was to treat the operating license term independently of the startup testing period. Surveillance, inspection, and maintenance practices are implemented in accordance with the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code), Quality Assurance Program, and Technical Specifications and ensure that any degradation in plant safety equipment will be identified and corrected to provide safe operation of the plant for the proposed recapture period. Although there is no regulatory guidance for review of this type of recapture, the staff performed its review on the basis of the effects of aging of safety-related and other structures and components, relative to the licensing basis. The review specifically focused on the adverse effects of aging to ensure that important systems, structures, and components will continue to perform their intended functions during the requested period of recapture. The staff reviewed the effect of the recapture period on the reactor pressure vessel, structures, mechanical equipment, electrical equipment, and quality assurance and maintenance programs, and addressed outstanding safety issues. The staff concluded that no safety issues existed that would preclude an additional 28.5 months of operation. The staff's draft safety evaluation is ahached.

Policy Implications of Granting Recapture

Although the GGNS amendment request is legal and technically acceptable, there are policy issues and implications associated with granting the amendment. The staff wants to make the Commission aware of these issues before it acts on the amendment.

First, a number of licensees, such as GGNS, received separate low-power and full-power

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licenses. As stated previously, a subset of these licensees already have a 40-year term beginning from the date of the FPOL. The GGNS amendment would be consistent with those licenses. Those licensees that were issued separate LPOLs and FPOLs with the 40-year license term beginning with the LPOL date would be eligible to recapture the time at low power without needing an exemption from 10 CFR 50.51 or applying for license renewal under 10 CFR Part 54. Granting the GGNS amendment would set a precedent for granting recaptures for those eligible plants. Although a number of plants would be eligible to apply for recapture, it would probably be advantageous for only a small number of licensees because of the variation in duration between issuance of the LPOL and the FPOL and fuel cycle length.

However, a number of licensees were not issued a separate FPOL and have a 40-year license term dating from the issuance of the LPOL. If these licensees hold Section 104b licenses, they would be able to recapture the time spent at low power with an exemption to 10 CFR 50.51 or by applying for license renewal. A holder of a Section 103 license under such circumstances would be limited to applying for license renewal to recapture time spent at low-power testing due to the 40-year limit in the AEA. This would in essence, be treating some licensees differently than others merely because of differences in agency practice in issuing licenses. However, some licensees have been granted 40 years from the issuance of the FPOL and not granting the amendment to Grand Gulf would be treating Grand Gulf differently from these licensees, although Grand Gulf would be the first plant that the 40 years from FPOL was not granted with the issuance of the FPOL. As a matter of policy, a Commission decision is needed on whether to allow a certain group of licensees i.e., those issued a separate FPOL, the opportunity to recapture low-power testing time via the license amendment process when that opportunity does not exist for another group i.e., those not issued a separate FPOL.

Second, each recapture technical review would have to be customized to some extent to address the variable lengths of proposed recaptures and the level of detail of aging programs.

Other Types of Recapture

Two Section 104b licensees, GPU Nuclear and Tennessee Valley Authority, have expressed interest in recapturing time spent in long shutdowns for the Three Mile Island (TMI) Unit 1 plant and the Browns Ferry Unit 3 plant, respectively. These cases are distinctly different from the Grand Gulf case in that Grand Gulf is specifically requesting a 40-year license term beginning with the date of FPOL. Browns Ferry 3 and TMl's recapture of time spent in long shutdowns, however, could result in the duration of those licenses extending beyond 40 years from the date of the FPOL. In the TMI and Browns Ferry cases, the Commission would be extending the 40-year license as a result of long shutdown conditions. An exemption from 10 CFB 50.51 would be needed. Should Section 103 licensees seek a similar shutdown recapture, they would be precluded from seeking such extensions by statute.

The Commission has generally expressed the view, in the context of license renewal under Part 54, that Section 103 and 104 licensees should be treated similarly. Therefore, the Commission concluded that extended operation of nuclear power plants under renewed licenses, which cannot be accomplished by amendment for Section 103 licensees in accordance with the Act, should also not be accomplished by amendment for plants licensed

under Section 104b. Even though the Commission is not precluded from granting

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amendments for the recapture of shutdown time requested by licensees of plants licensed under Section 104b such as Browns Ferry and TMI, the Commission, as a matter of policy, may wish to treat such licensees similarly to Section 103 licensees in similar circumstances and require that such types of "recapture" be accomplished under Part 54. Other factors associated with shutdown recaptures include whether to allow recapture of voluntary or involuntary shutdown time, unplanned shutdown time resulting from equipment or regulatory concerns, time spent in a shutdown condition because of NRC orders, or planned shutdown time caused by refueling outages or other maintenance outages. The duration of the shutdown would also have to be considered. The staff notes that arguments could be made to recapture shutdown time of any duration, planned or unplanned. The staff does not recommend granting these types of requests, other than through the established license renewal program specified in 10 CFR Part 54.

Although most licensees whose 40-year license term began with issuance of the construction permit have been granted construction period recapture, the staff recommends this practice continue consistent with past Commission policy.

Conclusion

In summary, the issue of whether to allow a licensee to recover (recapture) time spent in an extended low-power testing condition before receiving a full-power operating license is a policy matter. Because of past licensing practices, only licensees in similar circumstances to those of GGNS would be eligible to benefit from a policy that allows a licensee to recover time spent in low-power testing prior to receiving a full-power license. It is unclear how many of these licensees would view such an action as beneficial for the specific length of time spent in low-power operation. A policy issue exists on whether to allow a certain group of licensees the ability to recapture low-power testing time via the license amendment process when that opportunity does not exist for another group. In the Grand Gulf case, recovering the low-power testing period does not contravene the regulations, the evaluation is technically acceptable, and because of the long period of time between the LPOL and FPOL, GGNS has determined that recovering the time spent in low-power testing to be beneficial. Therefore, unless otherwise directed, the staff will issue the GGNS amendment. The Commission should note that by granting the GGNS amendment, a precedent would be set for the staff to grant other recaptures for similarly eligible plants if requested.

A policy issue also exists on whether to allow licensees to recover time spent in a shutdown condition, such as TMI and Browns Ferry. These cases are distinctly different than the GGNS case. Only 104b licensees would be eligible to request recovery of this time via an exemption from 10 CFR 50.51 i.e., without applying for license renewal under Part 54. The staff notes that should these types of requests for recovering time spent in shutdown conditions be granted, arguments could be made to recover time spent in a planned or unplanned shutdown of any duration. It is unclear how many licensees would view this type of action as beneficial.

A policy issue exists whether the duration of licenses should be extended via the exemption process for 104b licensees, while this cannot be done for 103 licensees. In the context of license renewal under 10 CFR Part 54, the Commission has expressed the view that 103 and 104b licensees

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be treated similarly even though a 104b licensee is not precluded from requesting a license amendment that would recover extended shutdown time. The staff recommends not granting these types of requests and requiring these types of recaptures be accomplished under Part 54.

The Commission has established a policy of granting licensee requests to recover time spent in construction in cases where the 40-year license term began with the construction date. The staff recommends continuing this policy.

RESOURCES:

Plant-specific recapture requests will be assigned priorities and reviewed as part of the existing licensing activities of the Office of Nuclear Reactor Regulation.

COORDINATIONz

Input to this paper was provided by the Office of the General Counsel (OGC). Additionally, OGC has reviewed this paper and has no legal objection to its contents.

The Office of the Chief Financial Officer has reviewed this Commission Paper for resource implications and has no objections.

RECOMMENDATION:

The policy issues and staff recommendations are as follows:

1. Grant the Grand Gulf license amendment request to amend the expiration date of the license to recover the time spent in an extended low-power testing condition before receiving the FPOL. Grant similar requests from other licensees provided the 40-year license term began with the LPOL and a separate FPOL was issued.

2. Continue to grant licensee requests to amend the expiration date of the license to recover time spent in construction in cases where the 40-year license term began with the construction date.

3. Deny granting license amendment requests to amend the expiration date of the license to recover time spent in a shutdown condition.

William D. Travers Executive Director for Operations

Attachment:

Draft GGNS Safety Evaluation

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DRAFT

Mr. Joseph J. Hagan Vice President, Operations GGNS Entergy Operations, Inc.

p. o. Box 756 Port Gibson, MS 39150

SUBJECT:

ISSUANCE OF AMENDMENT no. TO FACILITY OPERATING LICENSE no.

NPF GRAND GULF NUCLEAR STATION, UNIT 1 (TAC no. M92993)

Dear Mr. Hagan:

The Nuclear Regulatory Commission has issued the enclosed Amendment No. to Facility Operating License No. NPF-29 for the Grand Gulf Nuclear Station, Unit 1. This amendment is in response to your application dated July 21, 1995.

The amendment extends the expiration date of the operating license from June 16, 2022, to November 1, 2024. The extended date is 40 years from when the full power license was issued on November 1, 1984, in accordance with Section 103.c of the Atomic Energy Act of 1954, and 10 CFR 50.51, 50.56, and 50.57.

A copy of our related Safety Evaluation is also enclosed. A Notice of Issuance will be included in the Commission's next biweeklyFederal Registernotice. The enclosed evaluation would not be sufficient for a request for license renewal under 10 CFR Part 54.

Sincerely,

Jack N. Donohew, Senior Project Manager Project Directorate IV-1 Division of Reactor Projects Ill/IV Office of Nuclear Reactor Regulation

Docket No. 50-416

Enclosures:

1. Amendment No. to NPF-29

2. Safety Evaluation

CC w/encls: See next page

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ENTERGY OPERATIONS, INC.

SYSTEM ENERGY RESOURCES, INC.

SOUTH MISSISSIPPI ELECTRIC POWER ASSOCIATION

ENTERGY MISSISSIPPI, INC.

DOCKET no. 50-416

GRAND GULF NUCLEAR STATION, UNIT 1

AMENDMENT TO FACILITY OPERATING LICENSE

Amendment No.

License No. NPF-29

1. The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Entergy Operations, Inc. (the licensee) dated July 21, 1995, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I,

B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission,

C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations,

D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public, and

E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

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2. Accordingly, the license is amended, as indicated in the attachment to this license amendment, by amending paragraph 2. H. of Facility Operating License No. NPF-29 to read as follows:

H. This license is effective as of the date of issuance and shall expire on November 1 2024.5

3. This license amendment is effective as of its date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

Jack N. Donohew, Senior Project Manager Project Directorate IV-1 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation

Attachment:

Page 17 of the License

Date of Issuance:

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ATTACHMENT TO LICENSE AMENDMENT

AMENDMENT no. FACILITY OPERATING LICENSE no. NPF-29

DOCKET no. 50-416

Revise the above license by removing the page identified below and inserting the enclosed page. The revised page is identified by amendment number and contains a vertical line indicating the area of change.

REMOVE PAGE INSERT PAGE

17 of license 17 of license

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTQR REGULATION

RELATED TO AMENDMENT no. TO FACILITY OPERATING LICENSE no. NPF-29

ENTERGY OPERATIONS, INC.. ET AL.

GRAND GULF NUCLEAR STATION, UNIT 1

DOCKET no. 50-416

1 .0INTRODUCTION

By letter dated July 21, 1995, Entergy Operations, Inc. (the licensee) submitted a request for changes to the operating license for Grand Gulf Nuclear Station, Unit 1 (GGNS). The amendment would extend the expiration date of the operating license from June 16, 2022, to November 1, 2024. The extended date for termination of the operating license would be 40 years after issuance of the full-power license, NPF-29, issued on November 1, 1984. This proposed amendment is not a request for license renewal under 10 CFR Part 54.

2.0 BACKGROUND

Section 103.c of the Atomic Energy Act of 1954, as amended, provides that a license is to be issued for a specific period not to exceed 40 years. The Code of Federal Regulations (CFR) in 10 CFR 50.51 also specifies that each license will be issued for fixed period of time not to exceed 40 years from the date of issuance. Also, 10 CFR 50.56 and 50.57 allow the issuance of an operating license pursuant to 10 CFR 50.51 after the construction of the facility has been substantially completed, in conformity with the construction permit and when other provisions specified in 10 CFR 50.57 are met.

The current licensed term for GGNS ends on June 16, 2022. This is 40 years from the date of the low-power license, which was issued on June 16, 1982. In the low-power license, the licensee was only authorized to operate the plant up to 5 percent of rated power or 191 megawatts thermal.

On August 31, 1984, the Commission amended the low-power license to allow the licensee to operate up to 100 percent rated power or 3833 megawatts thermal. However, in response to a court challenge to the amendment, the Commission issued CLI-84-19 on October 25, 1984, directing the Staff to issue a separate full power license to Grand Gulf. On November 1, 1984, a full power license was issued to Grand Gulf whose expiration date was 40 years from the date of issuance of the low power license. In the full-power license, the licensee was authorized to operate up to 100 percent of rated power.

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The licensee requests an extension to the full-power operating license so that the licensed term will expire on November 1, 2024, or 40 years from the date of issuance of the full-power operating license on November 1, 1984. In the full-power license, the licensee was authorized to operate the plant up to 100 percent rated power or 3833 megawatts thermal.

In its application, the licensee stated that GGNS is fairly unique among licensed commercial nuclear power facilities in having an extended period of low-power operation. The period from the date of issuance of the low-power license to the full-power license is approximately 2.5 years. The licensee proposed to recapture this period of low-power operation by having the 40-year operating license term extended from June 16, 2022 to November 1, 2024. The licensee stated that, for GGNS, the additional license period would allow for at least one additional cycle of operation and perhaps two cycles and, therefore, the economic value of the license extension would be substantial.

The licensee's request for an extension of the operating license is based on the fact that a 40-year service life was considered during the design and construction of the plant.

Although this does not mean that some components will not wear out during the plant's lifetime, design features were incorporated which provide for inspectability of structures, systems, and components during this lifetime. Surveillance, inspectability, and maintenance practices which were implemented in accordance with the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code) for if service inspection and testing of pumps and valves and the plant Technical Specifications (TSs) provide assurance that any degradation in plant safety equipment will be identified and corrected to provide safe operation of the plant for the proposed license extension period. The specific provisions and requirements for ASME Code testing are set forth in 10 CFR 50.55a.

3.0 EVALUATION

The staff has evaluated the environmental and safety issues associated with the proposed amendment which would allow approximately 2.5 years of additional plant operation. The major safety issue is the effects of aging and neutron fluence on plant structures and equipment. This is addressed in Section 3.2.

The staff reviewed the licensee's application, the licensee's Updated Final Safety Analysis Report (UFSAR) for GGNS, the GGNS TSs, and the Safety Evaluation Report (SER, NUREG-0831 and its seven supplements) and Final Environmental Statement (FES, NUREG-0777) related to the operation of Grand Gulf Nuclear Station, Units 1 and 2, which documented the staff's review prior to issuance of a license to GGNS. The two NUREGs were issued in September 1981. The supplements to NUREG-0831 were issued at later dates as identified in the evaluation below.

Unit 2 is not involved with this amendment because it was abandoned in September 1985 and never constructed. The construction permit for Unit 2 was revoked by the Nuclear Regulatory Commission (NRC) Order dated August 7, 1991 .

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3.1 Environmental Assessment

The environmental assessment for the proposed amendment was addressed by the staff in the Notice of Issuance of Environmental Assessment and Finding of No Significant Impact in theFederal Register (62 FR 19144) on April 18, 1997. The conclusion of the staff was that the proposed amendment would result in no significant differences from the environment impacts that were reported in the FES dated September 1981 issued prior to the issuance of the licenses for GGNS on June 16, 1982 (low-power license), and November 1, 1984 (full-power license).

3.2 Safety Assessment

3.2.1 Neutron Damage of the Reactor Pressure Vessel

The reactor pressure vessel was designed and fabricated in accordance with the requirements of Section III, Class 1, of the ASME Code edition, addenda, and Code Cases applicable at the time of design and construction. Operating limitations of the ASME Code and of Appendix G, "Fracture Toughness Requirements," of 10 CFR Part 50 are also applicable. The reactor pressure vessel (RPV) and the reactor coolant system were designed to allow inspections in accordance with Section Xl of the ASME Code. The staff's evaluation approving the programs and their implementation with respect to these structures are contained in NUREG-0831 and its seven supplements. Industry experience with steel structures confirms a service life in excess of 40 years may be anticipated.

Over the operating life of a reactor vessel, ferrite materials exposed to neutron irradiation will undergo changes in material properties and a decrease in fracture toughness. The decrease in fracture toughness is of particular importance because the ability to resist failure caused by the propagation of a crack decreases with increasing irradiation. The fracture toughness of the vessel is monitored by a surveillance program in accordance with the requirements of Appendix H, "Reactor Vessel Materials Surveillance Program Requirements," of 10 CFR Part 50. The purpose of the materials surveillance program is to help ensure vessel integrity by monitoring changes in the fracture toughness properties of the reactor vessel beltline materials. The ferrite materials must meet the fracture toughness properties of Section Ill of the ASME Code and Appendix G to 10 CFR Part 50. This surveillance program will aid in adjusting the operational conditions in order to maintain sufficient safety margin for the prevention of brittle failure of the reactor vessel.

The reactor vessel is discussed in Section 5.3 of the UFSAR. In that section, the following are discussed :

The vessel is designed, fabricated, tested, inspected, and stamped in accordance with the ASME Code,Section III, Class 1 including the addenda in effect at the date of the order of the vessel, Winter 1972 and meets Seismic Category I.

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Shifts in transition temperature caused by irradiation during the vessel life can be accommodated by raising the minimum pressurization temperature, and the predicted value of adjusted reference temperature does not exceed 200 degrees F.

Compliance with Appendices G and H of 10 CFR Part 50.

The reactor vessel was also designed to withstand a variety of transient and cyclic loads which occur throughout the operational life of the plant. Table 3.9-35 of UFSAR provides the cyclic or transient limits for the vessel.

To date one material specimen capsule has been removed from the reactor vessel, however, by letters dated May 2 and 31, 1996, the licensee requested that it be placed back in the vessel because testing of the first capsule at 8 effective full power years (EFPY) may not be useful. The low neutron fluence and good material chemistry for the vessel will result in a minimal shift in the material properties of the specimen in the capsule. A revision to the capsule withdrawal schedule and placing the first capsule back in the vessel was approved by the staff in its letter of August 27, 1996.

On May 19, 1995, the NRC issued Generic Letter (GL) 92-01, Revision 1, Supplement 1, "Reactor Vessel Structural Integrity." In this GL, the NRC requested that licensees perform a review of their reactor vessel structural integrity analyses in order to identify, collect, and report any new data pertinent to the analysis of the vessel structural integrity and to assess the impact of that data on the analysis relative to the requirements of 10 CFR 50.60 (Acceptance criteria for fracture prevention measures for normal operation) and 50.61 (Fracture toughness requirements for protection against pressurized thermal shock), and Appendices G and H. The licensee responded in its letters of August 14 and November 20, 1995, and indicated that it has performed additional reviews and the structural integrity analyses remain valid, however, the licensee also stated that there is an industry initiative to ensure all sources of information pertinent to the reactor vessel are considered in the structural integrity analyses. In its letter of August 22, 1996, the staff concluded that the licensee had completed all of the actions in GL 92-01 and requested that the licensee provide NRC with results of this industry initiative for GGNS within 120 days of receipt of the final generic assessment.

Based on the above, there is reasonable assurance that the RPV will, for the proposed license term extension requested by the licensee, be in conformity with the applicable provisions of the rules and regulations of the Commission, and the GGNS license.

3.2.2 Structures

The concrete and steel Category I structures at GGNS were designed and constructed in accordance with the General Design Criteria of Appendix A, "General Design Criteria for Nuclear Power Plants," to 10 CFR Part 50. This DRAFT 5

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is discussed in Sections 3.1 and 3.2 of the UFSAR. The licensee's design basis, fabrication, construction, and implementation of quality assurance criteria for the plant were reviewed by the staff when the plant was being licensed for low-power operation. The staff's evaluation approving the programs and their implementation with respect to these structures are contained in NUREG-0831 and its seven supplements. Industrial experience with concrete and steel structures confirms a service life in excess of 40 years may be anticipated.

The major codes and specifications used in the design and construction of the Category I concrete and steel structures were, respectively, American Concrete Institute (ACI) 349, "Criteria for reinforced Concrete Nuclear Power Containment Structures," and ACI 318-71 ,

"Building Code Requirements for Reinforced Concrete," and the American Institute of Steel Construction (AISC) specification, "Specification for the Design, Fabrication, and Erection of Structural Steel for Building." The foundations of the seismic Category I structures are reinforced concrete designed to ACI 318-71. Section 3.8 of NUREG-0831 stated that the criteria that were used in the analysis, design, and construction of seismic Category I structures at GGNS account for anticipated loadings and postulated conditions that may be imposed on the structures during their service lifetime which would include the requested 2.5 years of additional power operation. These criteria are in conformance with the established criteria, codes, standards, and specifications acceptable to the staff.

The licensee's use of the indicated codes, standards, and specifications in the plant's design, analyses, and construction and the licensee's quality assurance program required by Appendix B, "Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants," to 10 CFR Part 50, as approved by NUREG-0831 and its supplements, provide reasonable assurance that the concrete and steel structures will, for the proposed license term extension requested by the licensee, be in conformity with the applicable provisions of the rules and regulations of the Commission, and the GGNS license.

3.2.3 Mechanical Equipment

Surveillance, maintenance, and testing requirements for mechanical equipment are in place at the plant to verify operability or to detect degradation and ensure that the equipment that does degrade is replaced or other corrective actions are taken. In addition, subcomponents such as nonmetallics (e.g., gaskets and o-rings) are inspected and replaced as necessary, as part of routine maintenance in order to ensure the design life of equipment. Surveillance, inspection, and testing requirements at GGNS, which will apply during the operating life of the plant, include the following:

ASME Code Section XI: Equipment that is safety-related is ASME Code Class 1, 2, or 3 and is subject to the if service inspection and testing requirements of Section XI and 10 CFR 50.55a, except where relief has been granted in writing from these requirements. These requirements apply throughout the operating life of a plant and will provide reasonable assurance that mechanical components will be properly monitored throughout the plant lifetime.

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Technical Specifications (TSs): 10 CFR 50.36 requires the establishment of limiting conditions for operation (LCOs) for certain equipment. (LCOs are the lowest functional

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capability or performance levels of equipment required for safe operation of the facility). This equipment is subject to the surveillance and testing requirements in the TSs to assure systems are operable. These surveillance requirements include calibration and inspection of systems and components to ensure that operation of the plant will remain in accordance with the limiting conditions for operation.

10 CFR Part 50, Appendix J: Equipment and components associated with containment penetrations, including containment isolation valves, are subject to the leak testing requirements in Appendix J, "Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors." This is for Type B and C testing of valves and penetrations, and Type A testing of the overall containment structure.

The licensee has implemented procedures for maintaining the operability of the mechanical components and, thus, the mechanical equipment the components are a part of. Examples of the procedures are the following: 01-8-07-27 (Rev. 10, dated 8/26/92) Lubricating Oil Sample Program, 01-8-07-35 (Rev.100, 3/20/95) ASME Section Xl System Pressure Test, 01-S-07-39 (Rev. 101, 2/4/97) lnservice Testing, 01-S-17-12 (Rev. 2, 3/11/96) Maintenance Monitoring Program for effectiveness of maintenance on equipment, 01-S-17-18 (Rev. 0, 4/5/90) Predictive Maintenance Program on early detection and diagnosis of equipment problems and degradation before failures, 01 -S-17-20 (Rev. 3, 1/30/97) Reliability Centered Maintenance Program to enhance preventative maintenance to improve reliability, 01 -S 21 (Rev. 1, 2/9/96) Oil/Lubricant Program, 01-S-17-43 (Rev. 0, 2/21/95) Air Operated Valves (AORs) Program for proper maintenance of AORs, 17-S-03-24 (Rev. 1, 8/6/93)

Thermography Program to inspect electrical and mechanical equipment to determine if maintenance is required, 17-S-03-25 (Rev. 1, 10/13/94) and 17-S-03-27 (Rev. 0,6/17/94)

Vibration Monitoring program to establish criteria and vibration limits to prevent equipment failures, 17-8-05-3 (Rev. 1, 11/14/95) Review of Pump lnservice Test Results, 17-S-05-12 (Rev. 100, 3/31/95) Snubber Service Life Program, GGNS-MS-41 (Rev. 3, 4/14/95) and GGNS-MS-46 (Rev. 1, 4/24/94) Monitoring Internal Erosion/Corrosion in High Energy and Moderate Energy Piping Components, and QAP.9.90 (Rev. 3, 2/26/97) Administration of Microbiological Induced Corrosion (MIC) Tracking in Standby Service Water Systems. The Station Information Management System (SIMS) and the repetitive task program (procedure 01-S-17-1 1, Rev. 2, 9/3/93) schedules the repetitive maintenance tasks, such as maintenance, inspections, testing, sampling, and surveillances to assure the tasks are completed as scheduled. This will assure that the preventative maintenance of the mechanical components is performed to ensure the operability and qualification of the mechanical equipment is maintained.

From this evaluation, the staff concludes that compliance with the codes, standards, and regulatory requirements to which mechanical equipment were analyzed, constructed, tested, and inspected provide adequate assurance that the structural integrity of equipment important to safety will be maintained during the operating lifetime of the plant and during the additional period authorized by this amendment. Any significant degradation by such equipment would be discovered and the equipment restored to an acceptable, and operable, condition.

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3.2.4 Electrical Equipment

The licensee has a program in place for the environmental qualification (EQ) of electrical

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equipment. As noted in Appendix H of Supplement 2 (dated June 1982), and Section 3.11 of Supplements 4 (dated May 1983) and 5 (dated August 1984), to NUREG-0831, the staff approved the program and deleted the low-power license conditions related to the program.

The full-power license condition 2.C.(11) required the licensee to qualify the electrical equipment to the EQ requirements in 10 CFR 50.49 by March 31, 1985. By letter dated March 7, 1985, the licensee requested an extension to no later than November 30, 1985 (as was allowed by 50.49) to have all electrical equipment important to safety qualified and in compliance with 10 CFR 50.49, and the staff granted the extension in its letter of March 27, 1985. The NRC Inspection Report 50-416/87-32, of March 25, 1988, documented that the licensee had met 10 CFR 50.49. Although there were 4 deficiencies in the EQ program, the staff concluded that within the scope of the inspection, the program met the requirements of 10 CFR 50.49.

The EQ program at GGNS includes qualification of the electrical equipment through accelerated aging tests. In accordance with 10 CFR 50.49, the program is required during the entire period of the operating license, which will include the term of the proposed license extension requested by the licensee, with approval of this amendment. The program will continue to ensure electrical equipment important to safety will not be used beyond its qualified life. To determine whether the program can and will perform this function, the following GGNS procedures which govern the environmental equipment qualification program were reviewed: (1) Standard ES-19 (Revision 9, September 9, 1996) Engineering Standard for Environmental Equipment Qualification Maintenance," (2) 07-8-01-227 (Revision 6, dated February 19, 1997) "Maintenance Procedure Equipment Qualification Program Safety Related," and (3) Standard ES-21 (Revision 1, May 5, 1988) "Engineering Standard Environmental Qualification Program Safety-Related." The safety information management system (SIMS) provides a computerized method of ensuring that the requirements of maintaining qualification of safety-related equipment are tracked and met.

The maintenance activities for each component which are required to maintain qualification of that component are shown in and scheduled through the SIMS database.

Although the plant's original life was considered to be 40 years, the EQ program will account for operation during the term of the proposed extension requested by the licensee. If a component has a qualified life of less than 40 years, its replacement is scheduled through the maintenance program and the SIMS database. Similarly, if the component has a 40-year qualified life, the replacement of the component is also scheduled through the maintenance program and the SIMS database. Therefore, the EQ program will support the proposed amendment.

3.2.5 Quality Assurance and Maintenance Programs

In licensing GGNS, the staff reviewed the quality assurance (QA) program and the conduct of operations, including the maintenance procedures, at GGNS. The QA program for the plant operations will assess how the plant organization is following procedures and meeting

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requirements for plant operation. This would include the maintenance program at the plant which assures the equipment is operable. In NUREG-0813, the staff concluded that the QA program and maintenance procedures were acceptable.

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Inspections by the staff of the QA and maintenance programs at GGNS show that these programs remain acceptable, although corrections in the programs have been identified.

The QA program meets the requirements of Appendix B to 10 CFR Part 50.

Therefore, the licensee's implementation and use of these programs at GGNS provides reasonable assurance that equipment important to safety will, for the proposed license term extension requested by the licensee, be in conformity with the applicable provisions of the rules and regulations of the Commission, and the GGNS license.

3.2.6 Status of Outstanding Issues, Confirmatory Issues, and License Conditions

At the time the plant was licensed, there were outstanding issues, confirmatory issues, and license conditions discussed in Sections 1.09, 1.10, and 1.11, respectively, of NUREG-0831 and its seven supplements. These issues and license conditions were either resolved prior to issuance of the full-power license or made license conditions. The proposed amendment has no effect on the license conditions in the full-power license.

3.3 Conclusion

Based on the discussion above, on the safety and environmental issues involved with granting an extension to the operating license, there are no safety issues that would preclude an additional 2.5-year period of operation, from June 16, 2022, to November 1, 2024, extending the term of the license. Based on this, the staff concludes that the proposed amendment is acceptable, however, it should be noted that the above evaluation would not be sufficient for license renewal under 10 CFR Part 54.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Mississippi State official was notified of the proposed issuance of the amendment. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

Pursuant to 10 CFR 51 .21, 51 .32, and 51 .35, an environmental assessment and finding of no significant impact has been prepared for the proposed amendment and published in the Federal Registeron April 18, 1997, (62 FR 19144). Accordingly, based upon the environmental assessment, the staff has determined that the issuance of the amendment will not have a significant effect on the quality of the human environment.

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6.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to

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the common defense and security or to the health and safety of the public.

Principal Contributor: Jack Donohew

Date :

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Federal Register / Vol. 60, No. 50 / Wednesday, March 15, 1995 / Notices 14035

Specifications.Public comments Illinois Power Company and Soyland hazards consideration comments requested as to proposed no significant Power Cooperative, Inc., Docket No. 50- received: No.Local Public Document Room hazards consideration: Yes (60 FR 5739, 461, Clinton Power Station, Unit No. 1, location:Maud Preston Palenske dated January 30, 1995). The notice DeWitt County, Illinois Memorial Library, 500 Market Street, St.

provided an opportunity to submit Date of application for amendment: Joseph, Michigan 49085.

comments on the Commission's August 12, 1994, as supplemented on proposed no significant hazards October 14, 1994 and February 6, 1995. Northeast Nuclear Energy Company, et consideration determination. No Brief description of arnendmen t:The al., Docket No. 50-423, comments have been received. The amendment modifies Clinton Power MillstoneNuclear Power Station, Unit notice also provided for an opportunity Station Technical Specification 3.6.5.1, No. 3, New London County, Connecticut to request a hearing by March 1, 1995, "Drywell," to permit a one-time only Date of application for amendment:

but stated that, if the Commission makeschange to forego performance of the May 18, 1994 a final no significant hazards drywell bypass leakage rate test during Brief description of amendment t:The consideration determination, any such the fifth refueling outage scheduled to amendment modifies the operability hearing would take place after issuance begin in March 1995. requirements for the fuel building of the amendments. Date of issuance:March 1, 1995 exhaust filter system. The amendment The Commission's related evaluation Effective date:March 1, 1995 will result in modifications to the of the amendments, finding of exigent Amendment No.: 96 applicability, surveillance requirement, circumstances, and final determination Facility Operating License No.NPF- and bases sections of Technical ono significant hazards consideration 62. The amendment revised the Specification 3/4.9.12, "Fuel Building is contained in a Safety Evaluation Technical Specifications. Exhaust Filter System."

dated February 14, 1995. Date of initial noticeinFederal Dateof issuance:February 22, 1995 Local Public Document Room Register:September 28, 1994 (59 FR Effective date:As of the date of location:Wharton County Junior 49427). The October 14, 1994, and issuance to be implemented within30 College, J. M. Hodges Learning Center, February 6, 1995, submittals consisted days.Amena'mentNo.: 105 911 Boling Highway, Wharton, Texas not change the staffs initial proposedof revisions and clarifications which didFacility Operating License No.NPF-77488 no significant hazards consideration 49. Amendment revised the Technical IES Utilities Inc., Docket No. 50-331, determination or expand the scope of Specifications.Date of initial noticein Federal Duane Arnold Energy Center, Linn the original notice.The Commission's Register:June 22, 1994 (59 FR 32234)

County, Iowa related evaluation of the amendment is The Commission's related evaluation of Date of application for amendment: contained in a Safety Evaluation dated the amendment is contained in a Safety August 15, 1994, as supplemented on March 1, 1995. No significant hazards Evaluation dated February 22, 1995.No December 21, 1994, and January 20, consideration comments received: No significant hazards consideration 1995. The licensee's submittals of location:The Vespasian Warner PublicLocal Public Document Roomcomments received: No.

December 21, 1994, and January 20, Library, 120 West Johnson Street, Local Public Document Room 1995, provided clarification and did notClinton, Illinois 61727. Three Rivers Community-Technicallocation:Learning Resources Center, change the original no significant Indiana Michigan Power Company, College, Thames Valley Campus, 574 hazards consideration. Docket Nos. 50-315 and 50-316, Donald New London Turnpike, Norwich, CT Brief description of amendment:The C. Cook Nuclear Plant, Unit Nos. 1 and 06360.

proposed amendment would revise the 2, Berrien County, Michigan Pacific Gas and Electric Company, Technical Specifications by increasing Date of application for amendments:

the allowable main steam isolation Docket Nos. 50-275 and 50-323, valve (MSIV) leakage and deleting the November 18, 1994Brief description of amendment is:TheDiabloCanyon Nuclear Power Plant, requirements applicable to the MSIV amendments revise Technical Unit Nos. 1 and 2, San Luis Obispo leakage control system. County,(:alifornia Date of issuance:February 22, 1995 Specification 4.0,5 to delete the wording"except where specific written relief hasDate of application for amendments:

Effective date:February 22, 1995 and been granted by the Commission July 9, 1992Brief description of amendment is:The to be implemented within 90 days. pursuant to 10 CFR 50, Section amendments extend the operating A amendmentNo.: 207 50.55a(g)(6)(i)." This change allows the licenses for the Diablo Canyon Nuclear Facility Operating LicenseNo. DPR- licensee to implement certain 10 CFR Power Plant, Units l and 2 to recover or

49. Amendment revised the Technical 50.55a relief requests while the relief recapture the construction period of the Specifications. at the beginning of an updated interval.requests are being reviewed by the NRCreactors. Specifically, the amendments Date of initial notice inFederal Dateof issuance:February 23, 1995 extend the expiration date of the Unit l Register:September 14, 1994 (59 FR Effective date:February 23, 1995 license from April 23, 2008, to 47169) The Commission's related Amendment Nos.:190/176 September 22, 2021, and the expiration evaluation of the amendment is Facility Operating LicenseNos. DPR- date of the Unit 2 license from contained in a Safety Evaluation dated 58 and DPR-74. Amendments revised December 9, 2010, to April 26, 2025.Dateof issuance:March 1, 1995 February 22, l995.No significant the Technical Specifications. Effective date:March 1, 1995 hazards consideration comments Dateof initial notice in Federal Amendment Nos.: 97 and 96 received: No. Register:December 21, 1994 (59 FR Facility Operating LicenseNos. DPR-Local Public DocurnentRoom 65817) The Commission's related 80 and DPR-82: The amendments location:Cedar Rapids Public Library, evaluation of the amendments is revised the license.

500 First Street, S. E., Cedar Rapids, contained in a Safety Evaluation dated Date of initial notice inFederal Iowa 52401. February 23, 1995. No significant Register:July 22, 1992 (57 FR 32575)

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14036 Federal Register / Vol. 60, No. 50 / Wednesday, March 15, 1995 / Notices

The Comlnission's related evaluation of Brief description of amendment is:The Philadelphia Electric Company, Docket the amendments is contained in a Safety amendment revises Technical Nos. 50-352 and 50-353, Limerick Evaluation dated March l, l995.No Specification 4.0.5, which provides the Generating Station, Units 1 and 2, significant hazards consideration requirements for in service inspection M ontgomery C aunty, Pennsylvania comments received: Yes. Comments and testing ofASME Code components, Date of application for amendments:

from the San Luis Obispo Mothers for to conform to Standard Technical August 31, 1994 Peace (MFP) and their contentions were Specifications (NUREG-1433).

admitted into this proceeding. These Date of issuance:February 28, 1995 Brief description of amendment is:

contentions concern the adequacy of the Effective date:February 28, 1995 These amendments address Section 5, licensee's maintenance and surveillance Amendment Nos.: 144 and 113 "Remove Temperature Requirement for program and interim corrective actions Facility Operating LicenseNos. NPF- Operational Condition 5 (TSCR 94 in lieu of Thermo-Lag. The Atomic 14 and NPF-22. The amendments 0), by revising TS Table 1.2 and TS Safety and Licensing Board, in its initialrevised the Technical Specifications. Bases 3/4,9.11 to remove the average decision dated November 4, 1994 (LBP- Date of initial notice inFederal reactor coolant temperature requirement 94-35), authorized the staff to extend theRegister:August 3, 1994 (59 FR 39595) in Operational Condition (OPCON) 5, DCPP operating license expiration dates.The Commission's related evaluation of Reliueling.

Because a hearing was held prior to the amendments is contained in a Safety Date of issuance:January 27, 1995 license issuance, the staff does not needEvaluation dated February 28, 1995.No Effective date:January 27, to make a final no significant hazards significant hazards consideration l995Amendment Nos. 88 and 50 consideration determination .Local Public Document Roomcomments received: No Facility Operating LicenseNos. NPF-location:California Polytechnic State Local Public Document Room 39 and NPF-85. The amendments University, Robert E. Kennedy Library, location:Osterhout Free Library, revised the Technical Specifications.

Government Documents and Maps Reference Department, 71 South Dateof initial noticein Federal Department, San Luis Obispo, California Franklin Street, Wilkes-Barre, Register:November 9, 1994 (59 FR 93407 Pennsylvania 18701. 55884) The Commission's related Pennsylvania Power and Light Pennsylvania Power and Light evaluation of the amendments is Company, Docket No. 50-387, Company, Docket Nos. 50-387 and 50- contained in a Safety Evaluation dated Susquehanna Steam Electric Station, 388 Susquehanna Steam Electric January 27, l995.No significant hazards Unit 1, Luzerne County, Pennsylvania Station, Units 1 and 2, Luzerne County, consideration comments received: No Date of application for amendment: Pennsylvania Local Public Document Room July 27, 1994, as supplemented October Date of application for amendments: location:Pottstown Public Library, 500 27, 1994 and February 3, 1995 October 28, 1994, and supplemented by High Street, Pottstown, Pennsylvania Brief description of amendment:The letter dated December 29. 1994 19464.

amendment raises the authorized Power These amendments change theBrief description of amendment is:Power Authority of The Suite of New Level from 3293 MWt to a new limit of Technical Specifications (TS) for the York, Docket No. 50-286, Indian Point 3441 MWt. Nuclear G enerating Unit No. 3, Date of issuance:February 22, 1995 1) and reference 118 (Unit 2) to Sectiontwo units by adding reference 120 (UnitWestchester County, New York and is to be implemented prior toEffective date:As of date of issuance6.9.3.2 as "PL-NF-90-001, SupplementDate of application for amendment:

startup in Cycle 9, currently scheduled 1, 'Application of Reactor Analysis November 16, 1994 to occur in May 1995. Methods for BWR Design and Analysis: Brief description of amendment t:The Amendment No.: 143 Loss of Feedwater Heating Changes and amendment revises Technical Facility Operating License No. NPF- Use of RETRAN MOD 5.l,' September Specifications Section 3.10.8 and the 14: This amendment revised the 1994." These additions reflect changesto the methodology that the licensee ismaximum allowable control rod dropassociated Bases, to reduce the Technical Specifications and license. using to perform its nuclear Hlel reload time from 2.4 to 1.8 seconds.

Dateof initial noticein Federal analysis for the two units. Date of issuance:February 21, 1995 Register:September 14, 1994 (59 FR Date of issuance:February 28, 1995 Effective date:As of the date of 47171) The Commission's related Effective date:February 28, 1995 issuance to be implemented within 30 evaluation of the amendment is Amendment Nos.: 145 and 114 days.

contained in a Safety Evaluation dated Facility Operating LicenseNos. NPF-February 22, 1995.No significant 14 and NPF-22. The amendments Amendment No.: 160 hazards consideration comments revised the Technical Specifications. Facility Operating LicenseNo. DPR-received: NoLocal Public DocurnentRoom Date of initial notice in Federal 64: Amendment revised the Technical location:Osterhout Free Library, Register:December 21, 1994 (59 FR Specifications.

Reference Department, 71 South 65819) The Commission's related Date of initial notice inFederal Franklin Street, Wilkes-Barre, evaluation of the amendments is Register:January 20, 1995 (60 FR 4203)

Pennsylvania 18701. contained in a Safety Evaluation dated The Commission's related evaluation of Pennsylvania Power and Light February 28, l995.No significant the amendment is contained in a Safety Company, Docket Nos. 50-387 and 50- hazards consideration comments Evaluation dated February 21, 1995.No 388 Susquehanna Steam Electric received: No significant hazards consideration Station, Units 1 and Z, Luzerne County, Local Public Document Room comments received: No Pennsylvania location:Osterhout Free Library, Local Public Document Room Reference Department, 71 South location:White Plains Public Library, Date of application for amendments: Franklin Street, Wilkes-Barre, 100 Martins Avenue, White Plains, New June 23, 1994 Pennsylvania 18701. York 10610.

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1 of March 1 1995 1 7

Mr. Gregory M. RuegerNuclear Power Generation, B14A Pacific Gas and Electric Company 77 Beale Street, Room 1451 p.0. Box 770000 San Francisco, California 94177

SUBJECT:

ISSUANCE OF AMENDMENTS FOR DIABLO CANYON NUCLEAR POWER PLANT,UNIT no. 1 (TAC N0. M84006) AND UNIT no. 2 (TAC no. M84007)

The Commission has issued the enclosed Amendment No. 97 to Facility Operating DPR-82 for the Diablo Canyon Power Plant (DCPP), Unit Nos. 1 and 2,License No. DPR-80 and Amendment No. 96 to Facility Operating License No.

respectively. The amendments are in response to your application dated July 9, 1992.

`\\_

These amendments extend the ocpp operating license expiration dates to September 22, 2021, for Unit 1, and to April 26, 2025, for Unit 2 or 40 yearsafter the date of the issuance of the "low-power" operating licenses.

A copy of the related Safety Evaluation is enclosed. A notice of issuance will be included in the Commissiorl's next regular biweekly Federal Register notice.

Sincerely,

Melanie A. Miller, Senior Project Manager Project Directorate IV-2 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation Docket Nos. 50-275 and 50-323 DISTRIBUT-

._ocket FE PUBLIC

Enclosures:

1.2.3.Amendment No. 97 to DPR-80 MM1 I ref DFoster-Curseen Amendment No. 96 to DPR-82 KPerkins, wcF0 GHi11 (4), T5C3 Safety Evaluation 0pA, 0265 0C/LFDCB, T9E10 JRoe PDIV-2 Reading cc w/encls: See next page TQuay OGC, 015818 CGrimes, 011E22 ACRS (4), TWFN Region IV RIV/WCFO (4)

DOCUMENT NAME: DC84006.AMD W( /45 _ _ _ l _-

'I OFC LA/DRPW 9" "

NAME DFoster-curseen M 1jler:pk o g DATE 2/ M495 .. . 2433/952/gi/95

"" 0FF1c1AL RECORD COPY MJ

1..- .m# v~. r. vi"`\\ l; r"~'*' l . ,>>, .ea 9v "ofis;l.. ., '.',>>~;;;e>> .d~. =I s>>

PDR ADDCK osooggg5 §0i'.9503080426 950301£ix,i,i>>:!a P

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. I

~>° 14)9* RE60 x I

'4 4or '9 UNITED STATES 41s 8

" Z NUCLEAR REGULATORY COMMISSION ann 5 v _ 8 WASHINGTON, D.C. 20555-0001 9*o *o March 1, 1995

Mr. Gregory M. Rueger Nuclear Power Generation, B14A 77 Beale Street, Room 1451Pacific Gas and Electric Company P.0. Box 770000 San Francisco, California 94177

Dear Mr. Rueger:

SUBJECT : ISSUANCE OF AMENDMENTS FOR DIABLO CANYON NUCLEAR POWER PLANT, UNIT no. l (TAC N0. M84006) AND UNIT N0. 2 (TAC N0. M84007)

The Commission has issued the enclosed Amendment No. 97 to Facility Operating License No. DPR-80 and Amendment No. 96 to Facility Operating License No.

DPR-82 for the Diablo Canyon Power Plant (DCPP), Unit Nos. 1 and 2, respectively.

July 9, 1992. The amendments are in response to your application dated September 22, 2021, for Unit 1, and to April 26, 2025, for Unit 2 or 40 yearsThese amendments extend the DCPP operating license expiration dates to after the date of the issuance of the "low-power" operating licenses.

A copy of the red ated Safety Evaluation is encl used. A notice of issuance will be included in the Commission's next regul at biweekly Eederal Moister notice.

Sincerely, In 4

11,4_QMQQr Melanie A. Miller, Senior Project Manager Project Directorate IV-2 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation Docket Nos. 50-275 and 50-323

Enclosures:

1.2.3.Amendment No. 97 to DPR-80 Safety EvaluationAmendment No. 96 to DPR-82

CC w/encls: See next page

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i . ,.. u

Mr. Gregory M. Rueger Pacific Gas and Electric Company Diablo Canyon

cc:

NRC Resident Inspector Mr. Steve Hsu Diablo Canyon Nuclear Power Plant Radiologic Health Branch c/o U.S. Nuclear Regulatory Commission State Department of Health Services

p. 0. Box 369 Post Office Box 942732 Avila Beach, California 93424 Sacramento, California 94234 Dr. Richard Ferguson, Energy Chair Sierra Club California Regional Administrator, Region IV 6715 Rocky Canyon U.S. Nuclear Regulator.y Conlnission Creston, California 93432 Harris Tower a Pavillion 611 Ryan Plaza Drive, Suite 400 Ms. Nancy Culver Artington, Texas 76011-8064 San Luis Obispo Mothers for Peace Mr. Peter H. Kaufman

p. 0. Box 164 Deputy Attorney General Pismo Beach, California 93448 State of California 110 West A Street, Suite 700 Ms. Jacquelyn c. Wheelerp. 0. Box 164 San Diego, California 92101 Pismo Beach, California 93448 Christopher J. Warner, Esq.

Pacific Gas & Electric Company Managing Editor Post Office Box 7442 The County ET eg.ran1_Tr_ip_ug3e San Francisco, California 94120 1321 Johnson Avenue

p. 0. Box 112 San Luis 0bisp0, California 93406 Mr. Warren H. FujimotoDiablo Canyon Nuclear Power Pl antVice President and Plant Manager Chairman p. 0. Box 56 San Luis Obispo County Board ofSupervisors Avila Beach, California 93424 Room 370 Diablo Canyon Independent Safety County Government Center Committee San Luis Obispo, California 93408 ATTN: Robert R. Wellington, Esq.Legal Counsel Mr. Truman Burns 857 Cass Street, Suite D Mr. Robert Kinosian Monterey, California 93940 California Public Utilities Commission505 Van Ness, Room 4102

San Francisco, California 94102

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Illllllilllllng L

.c 44,0G8 REGt

Lu 1.,f ,-9UNITED STATES 1- oNUCLEAR REGULATORY COMMISSION

3"' 2WASHINGTON, o.c. 20555-0001 o `"i .- 8 4% *o`

PACIFIC GAS AND EL£CTR_1C C0MP.AN1

DOCKET no. 50-275

)_CANYON NUCLEAR POWER PLANT, UNIT N0. I

AM_ENDMENT TO l:A(;IL]TY~()PERATIN(;LICENSE

Amendment No. 97 License No. DPR-80

1. The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Pacific Gas & Electric Company(the licensee) dated July 9, 1992, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's regulations set forth in 10 CFR Chapter I;

B. The facility wi11 operate in conformity with the application, theprovisions of the Act, and the rules and regulations of the Commission;

c. There is reasonable assurance (i) that the activities authorizedby this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment wi11 not be inimical to the commondefense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part51 of the [:ommission's regulations and all applicable requirements have been satisfied.

2. Accordingly, the license 'is amended as indicated in the attachment tothis license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-80 is hereby amended to read as follows:

9503080429 950301foR ADDCK 05008385

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W

2

(2) Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 97 , are hereby incorporated in the license. Pacific Gas a Electric Company shall operate thefacility in accordance with the Technical Specifications and the Environmental Protection Plan, except where otherwise stated in specific license conditions.

3. This license amendment is effective as of the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

• lu K

M ¢ MM Melanie A. Miller, Senior Project ManagerProject Directorate IV-2 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation

Attachment:

Page 9 of license*

Date of Issuance: March 1, 1995

4

• Page 9 is attached, for convenience, for the compositelicense to reflect this change.

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a 1 g

t E.

Physical Rggtectioq The licensee shall fully implement and maintain in effect all provisions of the Commission-approved physical security, guardtraining and qualification, and safeguards contingency plans

Miscellaneous Amendments and Search Requirements revisions to 10including amendments made pursuant to provisions of the CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The plans, which contain SafeguardsInformation protected under 10 CFR 73.21, are entitled: "Diablo Canyon Power Plant, Units 1 and 2 Physical Security Plan," with revisions submitted through March 4, 1988; "Diablo Canyon Power Plant, Units 1 and 2 Security Force Training and Qualification Plan," with revisions submitted through August 16, 1985; and"Diablo Canyon Power Plant, Units 1 and 2 Safeguards Contingency made in accordance with 10 CFR 73.55 shall be implemented inPlan," with revisions submitted through November 9, 1987. Changes accordance with the schedule set forth therein.

F. Antitrust Pacific Gas and Electric Company shall comply with the antitrust conditions in Appendix c to this license.

G. Reporting PG&E shall report any violations of the requirements contained in Sections 2.C(3) through 2.C(10), 2.E and 2.F, of this License within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Initial notification shall be made in accordance with the provisions of 10 CFR 50.72 with written follow-up in accordance with the procedures described in 10 CFR 50.73 (b), (c),(d) and (e).

H. Financia1_protection PG&E shall have and maintain financial protection of such type andin such amounts as the Commission shall require in accordance with Section 170 of the Atomic Energy Act of 1954, as amended, to cover public liability claims.

I. Term of License This License is effective as of the date of issuance and shall expire at midnight on September 22, 2021.

FOR THE NUCLEAR REGULATORY COMMISSION Original signed by:Eds or G. Case for

Harold R. Denton, Director Office of Nuclear Reactor Regulation

Attachments:

1. Appendix A-Technica] Specifications

2. Appendix B-Environmental Protection Plan

3. Appendix canAntitrust Conditions

Date of Issuance: November 2, 1984 Amendment No. -32>>, 97

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4

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. UNITED STATESl~<O'mNUCLEAR REGULATORY COMMISSION

z58\\.p WASHINGTON. D.C. zosss-0001 9;4% ,

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• 41

• PACIFIC GAS AND ELECTRIC COMPANY

DOCKET N0. 50-323

DIABLO CANYON NUCLEAR POWER PLAN7, UNIT N0. 2

AMENDMENT TO FACILITY OPERATING LICENSE

Amendment No. 96 License No. DPR-82

1. The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Pacific Gas & Electric Company (the licensee) dated July 9, 1992, complies with the standards and requirements of the Atomic Energy Act of1954,as amended (the Act), and the Commission's regulations set forth in 10 CFR Chapter I;

B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission;

c. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii)thatsuch activities will be conducted in compliance with the Commission's regulations;

0. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and

E. The issuance of this amendment is in accordance with 10CFR Part 51 of the Commission's regulations and all applicable requirements have beensatisfied.

2. Accordingly, the license is amended as indicated in the attachment to thislicense amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-82 is hereby amended to read as follows:

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• J

2

(2) Technical Specifications The Technical Specifications contained in Appendix A and theEnvironmental Protection Plan contained in Appendix B, as revised

through Amendment No. 96 , are hereby incorporated in the license. Pacific Gas & Electric Company shall operate thefacility in accordance with the Technical Specifications and the Environmental Protection Plan, except where otherwise stated inspecific license conditions.

3. This license amendment is effective as of the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION r \\

71@ 1.. q 69.WJQ Melanie A. Miller, Senior Project ManagerProject Directorate IV-2 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation

Attachment:

Page 7 of license*

Date of Issuance: March 1, 1995

• Page 7 is attached, for convenience, for the compositelicense to reflect this change.

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H 7

H. F_1nancia1 Protection PG&E shall have and maintain financial protection of such type and in such amounts as the Commission shall require in accordance with Section 170 of the Atomic Energy Act of 1954, as amended, to cover public liability claims.

1. Term of License This License is effective as of the date of issuance and shall expire at midnight on April 26, 2025.

FOR THE NUCLEAR REGULATORY COMMISSION Original signed by:

Harold R. Denton Harold R. Denton, Director Office of Nuclear Reactor Regulation

Attachments: Technica] Specifications (NUREG-1151)

1. Appendix A

2. Appendix BEnvironmental Protection Plan

3. Appendix CAntitrust Conditions Date of Issuance: August 26, 1985

Amendment No. 96

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ATTACHMENT TO LICENSE AMENDMENTS

AMENDMENI_N0. 97 TO FACILITY OPERATING LICENSE n0..0pl3-_80

AND AMENDMENT_N0 96_ .10 FACILITYOPERATING LICENSE NQ. DPR;§_2

DQQKEI uos,.s0-275 AND 50-323

Revise the above 'licenses by removing the pages identified be ow and inserting the enclosed pages. The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change.

0ver1 eat pages are at so included, as appropriate.

REMOVE INSERT 9, Unit 1 License 9, Unit 1 License7, Unit 2 License 7, Unit 2 License

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9 9

E. Physica] Protection

B The licensee shall fully implement and maintain in effect all provisions of the Commission-approved physical security, guardtraining and qualification, and safeguards contingency plans Miscellaneous Amendments and Search Requirements revisions to 10including amendments made pursuant to provisions of the CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR50.90 and 10 CFR 50.54(p). The plans, which contain Safeguards Canyon Power Plant, Units 1 and 2 Physical Security Plan," withInformation protected under 10 CFR 13.21, are entitled: 'Diablo revisions submitted through March 4, 1988; "Diablo Canyon Power Plant, Units 1 and 2 Security Force Training and Oualification Plan," with revisions submitted through August 16, 1985; and"Diablo Canyon Power Plant, Units 1 and 2 Safeguards Contingency Plan," with revisions submitted through November 9, 1987. Changes made in accordance with 10 CFR 73.55 shall be implemented inaccordance with the schedule set forth therein.

F. An_titrust Pacific Gas and Electric Company shall comp1uy with the antitrustconditions in Appendix C to this license.

G. Reporting PG&E shall report any violations of the requirements contained in Sections 2.C(3) through 2.C(l0), 2.E and 2.F, of this License within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Initial notification shall be made in accordance with the provisions of 10 CFR 50.72 with written follow-up in accordance with the procedures described in 10 CFR 50.73 (b), (c),(d) and (e).

H. Financial Protection PG&E shall have and maintain financial protection of such type and Section 170 of the Atomic Energy Act of 1954, as amended, to coverin such amounts as the Commission shall require in accordance with public liability claims.

I. Term of License This License is effective as of the date of issuance and shall expire at midnight on September 22, 2021.

FOR THE NUCLEAR REGULATORY COMMISSION Original signed by:

Eds or G. Case for Harold R. Denton, Director Office of Nuclear Reactor Regulation Attachments:

1. Appendix A Technical Specifications

2. Appendix B Environmental Protection Plan

3. Appendix c Antitrust Conditions Date of Issuance: November 2, 1984 Amendment No.'3'23 97

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v 7

H. Financial Protection

PG&E shall have and maintain financial protection of such type and in such amounts as the Conlnission shall require in accordance with Section 170 of the Atomic Energy Act of 1954, as amended, to cover public liability claims.

1. Term of License expire at midnight on April 26, 2025.This License is effective as of the date of issuance and shall

FOR THE NUCLEAR REGULATORY COMMISSION Harold R. DentonOriginal signed by:

Office of Nuclear Reactor RegulationHarold R. Denton, Director

Attachments:

1. Appendix A Technical Specifications (NUREG-1151)

2. Appendix BEnvironmental Protection Plan

3. Appendix C Antitrust Conditions Date of Issuance: August 26, 1985

Amendment No. 96

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H

1. . REGt44) oa UNITED STATES

v <.Lau co a 3 NUCLEAR REGULATORY COMMISSION

?>> 3 WASHINGTON, 0.0. 20555--0001 4 - 43°*o o

• * *

• 4

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REAQlDR.REGULATIQN

RELATED TO AMENDMENT N0. 97 TO FACILITY OPERATING LICENSE N0. DPR-§_0

AND AMENDMENT N0. 96 TO FACILITY OPFRATING LICENSE_l!Q_. DPR32

PACIFIC GAS AND .gLEc1l31g.c0;>>1pAn¥

DIABLO CANYON NUCLEAR POWER PLANT, UNITS I AND 2

DOCKE] nos. 50-275 AND 50-_323

1.0 INTRODUCTION

By letter of July 9, 1992, Pacific Gas and Electric Company (or the licensee) submitted a requestforchangestothe operating licenses for its Diablo Canyon Nuclear Power Plant, Units 1 and 2. The proposed amendments would change the expiration date for the Unit 1 Operating License from April 23, 2008, to September22,2021, and the expiration date for the Unit2 Operating License from December 9, 2010, to April 26,2025.

The staff issued a notice of "Proposed No Significant Hazards Consideration Determination" in the Lederal_Beqist_er(57 FR 32575) dated July 22, 1992.

This notice allowed for public comment and for a request for a hearing from "any person whose interest may be affected by this proceeding." By letter dated August 19, 1992, San Luis Obispo Mothers for Peace (SLOMP) filed a petition for leave to intervene and requested an evidentiary hearing. An Atomic Safety and Licensing Board was established to consider this matter and SLOMP was admitted into the proceeding as an intervenor pursuant to 10 CFR 2.714 after a prehearing conference held on December 10, 1992 (LBP-93-1,37 NRC 1)~

The staff issued an Environmental Assessment (EA) dated February 10, 1993 (58 FR 7899), as required by 10 CFR 51.21 and 51.22, in which it concluded that the May 1973 Final Environmental Statement for Diablo Canyon remains valid and pursuant to 10 CFR 51.31 an environmental impact statement need not be prepared for this action.

2.0 DISCUSSION

Section 103.c of the Atomic Energy Act of1954providesthata license is to be issued foraspecified period not exceeding 40years. The Code ofFederal Regulations in 10 CFR 50.51 specifies that each license will be issued fora fixed period of time not to exceed40years from dateofissuance. Also, 10 CFR50.56 and 10CFR 50.57 allow the issuance of an operating license pursuant to 10 CFR 50.51 after the construction of the facility has been substantially completed, in conformity with the construction permit and when other

gag ADOCK 050033859503080431 950301

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provisions specified in 10 CFR 50.57 are met. The currently licensed term for Diablo Canyon is 40 years, commencing with the issuance of the construction permit on April 23, 1968 for Unit 1 and December 9, 1970 for Unit 2.

Accounting for the time that was required for plant construction, this represents an effective operating license term of less than 27 years for Unit 1 and 25 years for Unit 2. Consistent with Section 103.c of the Atomic Energy Act and Sections 50.51, 50.56 and 50.57 of the Commission's regulations, the licensee, by its application of July 9, 1992, seeks extension of the operating license term from the date of operating license issuance for each unit, namely 40 years from September 22, 1981 for Unit 1 and 40 years from April 26, 1985 years provided by the Atomic Energy Act and the Code of Federal Regulations.for Unit 2. This action would extend the period of operation to the full 40

3.0 EVALUATION

The licensee's request for extension of the operating license is based on the construction of the plant. Although this does not mean that some componentsfact that a 40-year service life was considered during the design and Which maximize the inspectability of structures, systems and equipment. Thewill not wear out during the plant lifetime, design features were incorporated reactor coolant system components and support systems are analyzed for the margin) which could reasonably be expected to occur in a 40-year lifetime.integration effects of radiation damage and cyclic loadings (with added Surveillance and maintenance practices which were implemented in accordance with the American Society of Mechanical Engineers (ASME) Boiler and Pressure and the facility Technical Specifications (TS) provide assurance that anyVessel Code for If service Inspection and In service Testing of Pumps and Valves unexpected degradation in plant equipment will be identified and corrected.These TS are part of the plant's operating license and have been approved by the NRC, as are all subsequent changes to the TS. The specific provisions and requirements for ASME Code testing are set forth in 10 CFR 50.55a.

The design of the reactor vessel and its internals considered the effects of 40inyears of operation at full power and a comprehensive vessel material H that ensures the fracture toughness requirements of Appendix G are met. Assurveillance program is maintained in accordance with 10 CFR Part 50, Appendix stated in the Final Safety Analysis Report Update, reactor vessel surveillance tensile specimens; wedge opening loading fracture mechanics test specimens forcapsules are provided for post-irradiation testing of Charpy V-notch and Unit 1 and compact tension and bend bar fracture mechanics test specimens for Unit 2 are also provided.

As discussed above, the useful life of the Diablo Canyon units was intended to be 40 years. The Diablo Canyon reactor vessels were designed for transients The Pressurized Thermal Shock (PTS) rule, 10 CFR 50.61, establishes screeningconsidered to envelop design conditions over a 50-year operating period.

embrittlement beyond which plant operation cannot continue without furthercriteria that indicate a limiting level of nuclear reactor vessel plant specific evaluation. The licensee submitted a revised projection of embrittlement of its reactor vessels on December 4, 1992, which accounts for the proposed license extensions intended to recapture the construction period.

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I-.

4

Lu 3

Since the review and approval of the licensee's assessment of projectedembrittlement of the Diablo Canyon reactor vessels will be conducted for compliance with 10 CFR 50.61, a separate assessment for the purpose of this proposed amendment will not be performed.

Aging analysis has been performed for all safety-related electrical equipment in accordance with 10 CFR 50.49, "Environmental Qualification of Electrical lifetimes for this equipment. These lifetimes have been incorporated intoEquipment Important to Safety for Nuclear Power Plants," identifying qualified plant equipment maintenance and replacement practices to ensure that allsafet_y-related electrical equipment remains qualified and available to perform its safety-related function regardless of the overall age of the plant.

The staff's Safety Evaluation for environmental qualification (EQ) of safety-related electrical equipment was issued in Supplements 15 (dated September 1981) and 31 (dated April 1985) to the Diablo Canyon Safety Evaluation Report(SER). In Supplement 31 to the SER, the staff concluded that the Diablo Canyon E0 Program is acceptable and that compliance with 10 CFR 50.49 has been demonstrated.

October 16, 1974. While changes have been made to the plant design since theThe staff published its original Safety Evaluation for Diablo Canyon on original plant construction was completed, each of these changes that involved an unreviewed safety question has been reviewed and approved by the staff with Further, as required by 10 CFR 50.71(e), these changes and their effect onthe details being documented in the staff's related Safety Evaluation.

Report (FSAR) Update. Based on the ongoing review process, the staff has notaccident analyses, if any, are routinely updated in the Final Safety Analysis identified any concerns associated with approval of the proposed amendment to extend the expiration date of the license that are not already addressed by licensee commitments, operating procedures, and license requirements.

4.0 FI_NAL N0_S_IGNIEICANTHAZARDS CONSIDERATION DETERMINATION The licensee's request for amendment to the operating licenses for Diablo Canyon Units 1 and 2, including a proposed determination by the staff of nosignificant hazards consideration, was noticed in theFederal Register on July 22, 1992 (57 FR 32575). Comments were received on the proposed finding, and a hearing was held on two contentions submitted by SLOMP, one regarding the adequacy of the licensee's maintenance program and the other concerning the fire-retardant material, Thermo-Lag. The Atomic Safety and Licensing Board 180), found that Pacific Gas and Electric had justified the license extension(ASLB), in its initial decision dated November 4, 1994 (LBP-94-35, 40 NRC it sought and, subject to certain directions, as well as normal NRC staff review, should be granted the extensions (40 NRC at 185). The ASLB authorized the staff to extend the Diablo Canyon operating licenses expiration dates (40 NRC at 282) and directed the licensee to improve its Tel atemp sticker program, complete conversion of the radiation monitoring program, and undertake a study, to be submitted to the staff for review, concerning methodsfor improving communications between maintenance and other departments (40 NRC at 273). By letter dated December 9, 1994, the licensee informed the staff

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that it would complete all work and provide appropriate reports to the NRC by the end of 1995.

A final no significant hazards consideration determination need be made onlywhere a license amendment is issued prior to the conclusion of any hearing held on the application. There is no need for the staff to make a final determination in this instance where a hearing has been concluded and license amendment issuance has been authorized by an ASLB. For the same reasons, the comments received on the staff's proposed no significant hazards consideration determination need not be addressed.

5.0 stale CONSULTATION In accordance with the Commission's regulations, the California State official had no comments.was notified of the proposed issuance of the amendments. The State official

6.0 ENVIRONMENTAL CONSIDERATION

ANotice of Issuance of Environmental Assessment and Finding of No Significant Impact relating to the proposed extension of the Facility Operating License termination dates for the Diablo Canyon Nuclear Power Plant was published in the Federal Register on February 10, 1993 (58 FR 7899).

7.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that:

(a) There is reasonable assurance that the health and safety of thepublic w111 not be endangered by operation in the proposed manner,

(b) Such activities will be conducted in compl lance with theCommission's regulations, and

(C) The issuance of the amendment wil] not be inimical to the commondefense and security or to the heal th and safety of the public.

Principal Contributor: S. Peterson Date: March 1 19959

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>>~SUBJECT= EVALUATION OF DIABLO CANYON UNIT 1 SUPPLEMENTAL REACTOR VESSEL 19 RADIATION SURVEILLANCE PROGRAM (TAC no. N83285)

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I By letter dated March 31, 1992, you requested that the NRC staff review and approve a proposed supplemental reactor vessel radiation surveillance program and a revision to the reactor vessel material surveillance capsule withdrawal schedule for Diablo Canyon Unit 1 (DCPP-1). In our safety evaluation, which is enclosed, we conclude that the placement of supplemental capsules in the DCPP-1 reactor vessel is acceptable. Also, we conclude that the revised withdrawal schedule is acceptable. However, we request that you reevaluate the withdrawal schedule when test results from irradiated capsules become available or if the DCPP-1 fuel management program is significantly changed.

In addition, our safety evaluation concludes that the samples from the broken weld metal Charpy specimens may be used to determine the effect of radiation and annealing, if you demonstrate that the reconstitution process does not affect the Charpy test results. In addition, if you decide to anneal the DCPP-1 reactor vessel at some future time, all standby capsules should be annealed in order to maintain their materials in a condition equivalent to that of the reactor vessel.

This completes the staff effort on this issue and closes TAC Number M83285.

If you have any questions regarding this issue,p1ease contact me.

Sincerely, u

Original signeil' by

@Q§49 89984'ADDCK 05000275 Harry Rood,Senior Project Manager i P -u-a ,..*.=-; - --.E'DR--- Project Directorate V Division of Reactor Projects III/IV/V 990083 Office of Nuc]ear Reactor Regulation

Enclosure:

DISTRIBUTION: 3 Safety Evatuation TQuay[ll5_§,Ke.t_Ei].e _ENRC & Loca] PDRs BBoger Mvirgilio HRood CC w/enclosure: DFoster PDV Plant File CRegan See next page OGC, 15B18 ACRS(10), P315

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4% *s September 4, 1992 Docket No. 50-275 Mr. Gregory M. Rueger Nuclear Power Generation, B14APacific Gas and Electric Company 77 Beale Street, Room 1451 P.0. Box 770000 San Francisco, California 94177

1

SUBJECT:

EVALUATION OF DIABLO CANYON UNIT 1 SUPPLEMENTAL REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM (TAC N0. M83285)

By letter dated March 31, 1992, you requested that the NRC staff review and approve a proposed supplemental reactor vessel radiation surveillance program and a revision to the reactor vessel material surveillance capsule withdrawal schedule for Diablo Canyon Unit 1 (DCPP-1). In our safety evaluation, which DCPP-1 reactor vessel is acceptable. Also, we conclude that the revisedis enclosed, we conclude that the placement of supplemental capsules in the withdrawal schedule is acceptable. However, we request that you reevaluate the withdrawal schedule when test results from irradiated capsules become available or if the DCPP-1 fuel management program is significantly changed.

In addition, our safety evaluation concludes that the samples from the broken weld meta] Charpy specimens may be used to determine the effect of radiation and annealing, if you demonstrate that the reconstitution process does not DCPP-1 reactor vessel at some future time, all standby capsules should beaffect the Charpy test results. In addition, if you decide to anneal the that of the reactor vessel.annealed in order to maintain their materials in a condition equivalent to

This completes the staff effort on this issue and closes TAC Number M83285.

If you have any>>questions regarding this issue, please contact me.

Sincerely,

Harry RocR1, Senior Project Manager Project Directorate v Division of Reactor Projects III/IV/V Office of Nuclear Reactor Regulation Safety Evaluation

Enclosure:

CC w/enclosure:

See next page

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>> Mr. Gregory M. Rueger Pacific Gas and Electric Company Diablo Canyon

cc:

NRC Resident Inspector Mr. Hank Kocol Diablo Canyon Nucl ear Power PI ant Radiologic Health Branch c/oU.S.Nucl ear Regulatory Commission State Department of Health Services

p. 0. Box 369 4 Post Office Box 942732 Avil a Beach, California 93424 Sacramento, California 94234

Dr. Richard Ferguson, Energy Chair Sierra CO ub Ca1 ifornia Regional Administrator, Region V 6715 Rocky Canyon U.S. Nuclear Regulatory Commission Creston, California 93432 1450 Maria Lane, Suite 210 I

Walnut Creek, California 94596 Ms. Sandra A. Silver Mothers for Peace Mr. Peter H. Kaufman 660 Granite Creek Road Deputy Attorney Genera]

Santa Cruz, California 95065 State of California 110West A Street, Suite 700 Ms. Jacquelyn C. Wheeler San Diego, California 92101 3303 Barranca Court San Luis Obispo, California 93401 Ms. Nancy Culver 192 Luneta Street Managing Editor San Luis Obispo, California 93401 The County TeTegram Tribune 1321 Johnson Avenue Michael M. Strumwasser, Esq.

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p. 0. Box 112 Special Assistant Attorney General San Luis Obispo, California 93406 State of California Department of Justice Chairman 3580 Wilshire Boulevard, Room 800 San Luis Obispo County Board of Los Angeles, Ca1 ifornia 90010 Supervisors Room 370 County Government Center San Luis Obispo, California 93408

Christopher J. Warner, Esq.

Pacific Gas & Electric Company Post Office Box 7442 San Francisco, California 94120

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BY THF OFFICE OF NUCLEAR REACTOR REGULATION

RELATED TO SUPPLEMENTAL REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM

PACIFIC GAS AND ELECTRIC COMPANY

DIABLO CANYON PDNER PLANT. UNIT 1

DOCKET N0. 50-275

1.0 Introduction In a letter dated March 31, 1992, the Pacific Gas and Electric Company (the reactor vessel radiation surveillance program and a revision to the reactorlicensee) requested that the staff review and approve a proposed supplemental vessel material surveillance capsule withdrawal schedule for Diablo Canyon Power Plant Unit 1 (DCPP-1). The licensee stated that the current Unit 1surveillance program was established before the issuance of Appendix H to 10 CFR Part 50, and was designed in accordance with the requirements of ASTM E 185-70, the standard in effect when the reactor vessel was manufactured.

2.0 Discussion requirements of the edition of ASTM E 185 that was current on the issue dateAppendix H to 10 CFR Part 50 requires that the surveill once program meet the requires that the schedule for capsule withdrawal must be approved by the NRCof the ASME Code to which the reactor vessel was purchased. Appendix H also staff prior to implementation. ASTM E 185-70 requires that specimens be withdrawn at three or more separate times throughout the design life of the vessel and that specimens be taken from the following locations:

(1) the base meta] of the heat from the high flux location of the reactorvessel that has the highest initial ductile-brittle transition temperature, (2) weld meta] that is fully representative of fabrication practices usedfor welds in the highest flux location of the reactor vessel (weld wire and flux must come from one of the heats used in the highest flux region), and

(3) the heat affected zone of the weldment.

The current surveillance program consists of eight capsules: three "Type II" capsules containing the limiting beltline weld metal, limiting shell plate, and weld heat affected zone from an intermediate shell plate, and five

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.II 1112 I "Type I", or "standby" capsules that do not contain limiting beltline material (weld metal). The three Type II capsules will be removed to determine the effect of irradiation during the vessel's current design life of 32 effective full power years. The three Type II capsules contain base metal, weld metal, and heat affected zone samples that meet the requirements of ASTM E 185-70.

In its March 31, 1992 letter, the licensee proposed to:

(1) add four supplemental capsules (A, B, C, and D) to the surveillanceprogram,

• (2) remove and store standby capsules T and Z, (3) remove and test standby capsule Y, in accordance with the previouslyexisting withdrawal schedule,

(4) relocate capsule V, and

(5) revise the capsule withdr'awal schedule.

The four supplemental capsules (A, B, C, and D) that are to be added to the program will contain weld metal fabricated using the same heat of weld wire as the limiting beltline weld and plate material from the limiting plate at 48 EFPY. All four supplemental capsules will contain weld metal removed from a reactor vessel nozzle dropout supplied by the vessel vendor, ABB Combustion Engineering. Two of the supplemental capsules (B and D) will contain material from broken weld metal Charp\\y specimens that were previously irradiated in Capsule S. After reirradiation, the capsule B and D Charpy specimens will be reconstituted and compared to samples removed from reactor vessel nozzle dropout.

The staff believes that it is essential that the process for reconstitution not effect the Charpy test results. Hence, the staff concludes that, prior tothe removal of Capsules B and D, the licensee should develop a procedure that will ensure that the reconstitution process will not affect the Charpy test

=c results.

The withdrawal schedule for the four supplemental capsules is as follows:

(1) Capsule A will remain in the vessel as a standby capsul e.

(2) Capsul e B will be removed and tested when the accumul ated fluence isequival ent to the vessel inside surface at 48 EFPY.

(3) Capsule C will be removed when the accumulated fluence is equivalentto the vessel inside surface at 32 EFPY, the design life of DCPP-1.

Half of the specimens will be tested before annealing and half will betested after annealing to demonstrate the toughness recovery after `

thermal annealing.

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o (4) Capsule D will be removed when the accumulated fluence is equivalent to the vessel inside surface at 32 EFPY. The samples from Capsule D will be annealed, reinserted into the vessel, and subsequently removed when the accumulated fluence is equivalent to the vessel inside surface at 48 EFPY. This capsule will be used to demonstrate the degree of reembrittlement after thermal annealing.

A11 capsules, except for the standby capsules, wili be withdrawn by approximately 19.2 EFPY. These capsul es will provide data for the license t°enewa1 period and wil] determine the effect of annealing on the reactor vessel s The two capsules (T and Z) to be removed and stored contain only plate material. Since the limiting material with respect to pressurized thermal shock is the weld metal, these capsules need not be tested.

Capsule V will be moved from its present location (184 degrees) to a location Hence, it will reach its planned accumulated fluence, a value equivalent to(320 degrees) which has a greater lead factor than its present location.

the vessel inside surface at 32 EFPY, earlier in the plant's life.

3.0 Conclusions The NRC staff has reviewed the proposed supplemental reactor' vessel radiation capsule withdrawal schedule for Diablo Canyon Power Plant Unit 1. The resultssurveillance program and revision to the reactor vessel material surveillance of the review are given below.

(1) The licensee proposes to install the supplemental capsules in theDCPP-1 reactor vessel at the end of Cycle 5, during the September 1992 make room for the supplemental capsules. Capsule V will be moved to aoutage. Capsules T and Z will be removed at that time and stored to higher flux location (320 degrees). The staff finds these changesacceptable because they augment the current program, and will provide additional data On the limiting reactor vessel materials.

(2) The staff concludes that broken weld metal Charpy specimens may be reconstituted and compared to samples removed from the reactor vessel 4 dropout, provided that the licensee develops a reconstitution procedure that ensures that the reconstitution process does not affectthe Charpy test results.

(3) The staff finds the proposed withdrawal schedule to be acceptable.However, the schedule should be reviewed by the licensee when test results from irradiated capsules become available or if the licenseesignificantly changes its fuel management program.

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/ (4) If at some future time the licensee decides to anneal the DCPP-1 reactor vessel, all standby capsules should be annealed to a conditionequivalent to the reactor vessel.

Principal Contributor: B. Elliot Date: September 4, 1992

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A 1: 14 _. .27' Pacific Gas and Electric Gumpany 77 Beale StreetGregory M. Rueger San Francisco, CA 94106 Senior Vice President and 415/973-4684 General Manager Nuclear Power Generation July 9, 1992

PG&E Letter No. DcL-92-154

U.S. Nuclear Regulatory Commission ATTN: Document Control Desk

° 8 Washington, D.C. 20555

Re: Docket No. 50-275, OL-DPR-80 Docket No. 50-323, OL-DPR-82 Diablo Canyon Units 1 and 2 License Amendment Request 92-04 40-Year Operating License Application i

Gentlemen :

Pacific Gas and Electric Company applies for an amendment to Facility Operating License Nos. DPR-80 and DPR-82 to change the expiration dates of the full-power licenses for Diablo Canyon Units 1 and 2. PG&E's enclosed request would change the expiration date for the Unit 1 Operating License from April 23, 2008, to September 22, 2021, and the expiration date for the Unit 2 Operating License from December 9, 2010, to April 26, 2025. These proposed expiration dates would allow for 40 years of operation as permitted by 10 CFR 50.51.

The present operating license terms for Diablo Canyon are based on the NRC policy in effect prior to the 1982 determination by the Commission that the 40-year term of operation may begin upon issuance of the first operating license, rather than upon issuance of the construction permit. Therefore, the present operating license terms for Diablo Canyon commence with the dates of issuance of the construction permits for Units 1 and 2, April 23, 1968, and December 9, 1970, respectively.

Accordingly, the expiration date for the Unit 1 Operating License is April 23, 2008, and the expiration date for the Unit 2 Operating License is December 9, 2010.

Since 1982, the Commission has accepted and approved requests to amend existing operating licenses to change the expiration dates and recover the time between the effective dates of the construction permit and the first operating license. More than 50 such license amendments have been granted by the Commission. Based on the enclosed request, the proposed 40-year term start dates for Diablo Canyon are September 22, 1981, for Unit 1 and April 26, 1985, for Unit 2, which correspond to the effective dates of the fuel-load/low-power operating licenses for each unit.

The proposed license term changes do not affect the design, operation, or Technical Specifications of the plant. Based on a review of the Diablo Canyon Final Safety Analysis Report Update and the associated NRC Safety Evaluation Report and Supplements, PG&E concludes that the proposed changes do not involve significant

/-~ 9é57i"§o°s9"920709'hazard considerations. 4. I(

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. aIn _2 July 9, 1992 X I -DbcuMint Control Desk PG&E Letter No. DCL-92-154

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PG&E also has determined that the environment will not be adversely affected by the proposed license term changes based on a review of the Diablo Canyon Environmental

4 Report>>andSupp1ements and. the.. NRC Final Environmental Statement and Addendum.

Therefore, pursuant to 10 CFR 51.30 through 51.35, PG&E believes the proposed changes require preparation only of an environmental assessment and finding of no significant impact, and that preparation of an environmental impact statement is not required. These actions are consistent with the Commission's practice for similar amendment requests.

Based on these safety and environmental reviews, PG&E requests that the Diablo Canyon operating license expiration dates be changed from April 23, 2008, to September 22, 2021, for Unit 1 and from December 9, 2010, to April 26, 2025, for Unit 2.

Sincerely,

4

Greg ry M. Rueger

cc: Edgar Bailey, DHS Ann P. Hodgdon I John B. Martin Philip J. Morrill Harry Rood CPUC 4 Diablo Distribution

Enclosure

54088/85K/EMG/2057

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ATTACHMENT A

DIABLOCANYON POWER PLANT UNITS 1 AND 2 40-YEAR OPERATING LICENSE APPLICATION

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CONTENTS

1.

e.

1.0 DESCRIPTION

OF.AMENDMENT REQUEST 1.......................

to BACKGROUND 1...........................................

3.0 JUSTIFICATION 2......................... .................

3.13.23.33.4Baseload Generation.................................... 2333 Electric Rates........................................

Air Emissions .........................................

State and Local Economy.................................

4.0 SAFETY EVALUATION 3......................................

4.14.2 Introduction..........................................

Licensing Basis Documents/Programs 34.........................

4.2.1 FSAR and Technical Specifications.................... 4 4.2.2 Probabilistic Risk Assessment........................ 4 4.2.3 Surveillance and Maintenance Programs 5

1 4.2.3.1 ISI and IST Programs............................ 567 4.2.3.2 EQ Program ..................................

4.2.3.3 Maintenance Program............................

4.3 Plant Operating History 10.................................

4.3.1 Operating Performance >> . - ............................101113 4.3.2 Reliability/Safety-Related Plant Modifications.............

4.3.3 Regulatory Performance..........................

4.4 Assurance of Continued Functional Capability of Safety-Related Components 13...............................

4.4.1 Reactor Coolant System Pressure Boundary 13..............

4.4.1.1 General .................................... 131415 4.4.1.2 Reactor Vessel................................

4.4.1.3 Reactor Vessel Internals..........................

4.4.2 Other Mechanical Components...................... 151616 4.4.3 Electrical Components...........................

4.4.4 Structural Components...........................

4.4.4.1 Primary Containment............................

4.4.4.2 Other Structures 1617...............................

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5.0 ENVIRONMENTAL EVALUATION 18.............................

5.15.2 Introduction......................................... 18 Systems and Programs for Environmental Control and Monitoring 18.................................

I 5.2.1Waste Processing System 18.........................

5.2.2 ALARAProgram.,,,,,,_,,.,,..,__. 19............. ................

5.2.3 Process and Area Radiation Monitoring System ............ 20 5.2.4 Radiological Environmental Monitoring Program ........... 20 5.2.5 Nonradiological Surveillance Program 21.................

5.2.5.1 Environmental Protection Plan...................... 21 5.2.5.2 NPDESPermit ................................ 22

E 5_3 Environmental Impact During Normal Operation 22..................

5.3.1 Occupational Radiation Exposure..................... 22 5.3.2 Offsite Radiation Exposure........................ 22 5.3.3 Uranium Fuel _ .. _................................. 24 5.3.4 Spent Fuel Storage............ ................ 25 5..3.5 Solid Waste.................................. 26

5.3.5.1 Low Level Radioactive Waste....................... 26262727 5.3.5.2 SpentFuel. , _...................................

5.3.5.3 Waste Shipping................................

5.3.5.4 SolidWaste Conclusions..........................

5.3.6 Thermal and Ecological Effects of the Cooling Water System 5.3.7 Protection of Historic Properties ... 2728.....................

5.45.55.6Exposure from Releases During Postulated Accidents............... 293334 Environmental-Related Plant Modifications.....................

Decommissioning .....................................

6.0 NO SIGNIFICANT HAZARDS EVALUATION 35......................

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4.4.1.2 Reactor Vessel

The Diablo Canyon reactor vessels were designed and fabricated in accordance with ASME Boiler and Pressure Vessel Code, Section 111, "Nuclear Power Plant Components." They were designed for A49 t 1transients considered to envelope design conditions -over a 50-year.operating period...To ensure the continued integrity of the vessels during operation, an ISI Program (see Section 4.2.3. 1) has been in place since plant startup.' The ISI Program requires volumetric examination of each accessible pressure boundary weld at least once in each 10 year interval.

F r 1 The effects of neutron radiation embrittlement of the vessel beltline region are considered in the design and operation of the units. Compliance with all NRC regWations governing vessel integrity has been documented recently in PG&E's response to Generic Letter 92-01 (PG&E Letter No. DCL-92-150, dated

4 June 30;4992). 'Inaddition,'PG&E has instituted an Embrittlement Management Plan to manage reactor vessel embrittlement throughout the entire operating life of Diablo Canyon.

Pressurized Thermal Shock

\\ I Following Cycle 1 for each unit, the neutron fluence at the reactor vessel inner wall was reduced byinstalling 'increasingly lower neutron leakage cores, thus decreasing the reactor vessel rate of embrittlement and prolonging vessel life. The Diablo Canyon reactor pressure vessel beltline materials have been evaluated according to the NRC's Pressurized Thermal Shock (PTS) screening criterion defined in 10 CFR 50.61. The Reference Temperature for Pressurized Thermal Shock, RTws, has been calculated for each weld metal and base metal in the DCPP ~beltline regions for neutron fluences corresponding to 40 operating years. The RTf,-s for all materials will not exceed the screening limit of 270°F for base metal and longitudinal welds and 300°F for circumferential welds. Since all materials meet the screening criterion, neither additional flux reduction nor plant specific PTS analyses are required to comply with the PTS rule. Details of the PTS evaluation have been submitted to the NRC (PG&E Letter No. DCL-92-056, dated March 6, 1992).

I Based on a-conservative fluence projection for 40 operating years, DCPP will also meet the requirements of 10 CFR 50, Appendix G. Charpy Upper Shelf Energies were determined (FSAR Update Tables 5.2-19A,5.2-l9B, 5.2-21A, and 5.2-2lB) in accordance with Regulatory Guide 1.99, Revision 2. All Diablo Canyon beltline materials will remain above the 50 ft-lb Charpy Upper Shelf Energy fracture toughness requirement for more than 40 operating years. In addition, reactor vessel pressure-temperature limits will meet 10 CFR 50, Appendix G requirements for 40-year ope:ation without requiring plant modification or imposing operational restrictions.

Material Surveillance Program

I The toughness properties of the reactor vessel beltline material will be monitored throughout the proposed 40-year operating license terms with a material surveillance program that meets the requirements of 10 CFR 50, Appendix H.

The surveillance test program for DCPP Unit 1 complies with ASTM E 185-70, the standard in effect when the vessel was manufactured. Although the Unit 1 surveillance program was designed prior to the existence of 10 CFR 50, 'Appendix H, it does contain the significant -features required for later surveillance programs and will effectively monitor vessel embrittlement throughout the requested license period. The program includes a total of eight surveillance capsules. Three of the eight capsules contain the limiting weld metal and base metal, correlation monitor material, dosimeters, and thermal monitors.

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I I r The remaining five capsules contain the limiting base metal, but no weld metal. Allof the base metal charpy specimens in the capsules are longitudinally oriented.

l.,

The Unit 2 surveillance program includes six capsules and conforms to ASTM E 185-73. All capsules

.131-us.. 39 Econtain the limiting weld metal,~which>>is the limiting=beltline material..!1`he_base metalspecimens in the

l capsules are not from the limiting plate, but were machined from an adjacent plate with similar chemistry, I the same heat treatment, and similar level of embrittlement (adjusted reference temperature within 3°F) l at plate end-of-life as the limiting plate.

4 UIl To date, one surveillance capsule from Unit 1 and two capsules from=Unitl2 have been analyzed.

Analysis of these capsules confirms that the measured shifts.in nil ductility reference temperature, RTnuD for the limiting materials in the vessels are well within two standard deviations of the mean RTnDT shifts r predicted by Regulatory Guide 1.99, Rev. 2.

In addition to those required surveillance programs, a supplemental surveillance program will be I 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.

I Additional measures to monitor DCPP Units 1 and 2 vessel fluence are provided in the Reactor CavityNeutron Dosimetry Program. This voluntary program has been in effect since initial criticality and consists of irradiating and evaluating reactor cavity dosimetry, which includes multiple foil sensor sets and axial flux gradient wires attached to the metal reflective insulation surrounding the reactor vessel.

Results obtained are used to confirm and complement surveillance capsule data.

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

1 degree of embrittlement of beltline materials over the proposed 40-year operating license terms.

4.4.1.3 Reactor Vessel Internals

The design of the reactor vessel internals meets the intent of Section III of theASMEBoiler and Pressure Vessel Code. The material used for fabrication of most of the reactor vessel internals is solution heat treated, unsensitized Type 304 austenitic stainless steel conforming to ASTM specifications. Weld fabrication was done using procedures and personnel qualified in accordance with Section IX of the ASME Boiler and Pressure Code. Evaluations performed prior to initial plant startup document the ability of the reactor vessel internals to perform their intended functions when subjected to loads imposed during normal operation, abnormal operational transients, and accidents.

Periodic inspections performed under "the ISI Program ensure that any significant degradation of reactor vessel internals over the proposed 40-year operating license terms will be detected and repaired in a timely manner. -

4.4.2 Other Mechanical Components

The passive mechanical components (tanks, pump casings, andvalvebodies) associated with safety-related systems are designed to meet the intent of Regulatory Guides 1.26 and 1.29. Consideration was given to possible aging effects including corrosion, erosion, and thermal cycling fatigue. The expected service

of 54088/85K 15 ER-242

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Pacific Gas and Electric Company 77 Beale Street Gregory M. Rueger San Francisco, CA 94106 Senior Vice President and 415/973-4684 General Manager Nuclear Power Generation March 31, 1992

e I PG&E Letter No. DCL-92-072

U.S. Nucl ear Regul story Commission ATTN: Document Control Desk I

Washington, D.C. 20555 I 8 Re : Docket No. 50-275, OL-DPR-80 Diablo Canyon Unit 1 Suppl emental Reactor Vessel Radiation Surveillance Program

Gentlemen:

your review and approval is a proposed suppl emental reactor vessel radiationIn accordance with the provisions of 10 CFR 50, Appendix H, enclosed for

\\ surveill once program for Unit 1 of the Diablo Canyon Power PI ant (DCPP). .

185-82 and the draft edition of ASTM E 185-92. Additional guidance wasGuidance for the design of the suppl emental program was taken from ASTM E taken from the EPRI report, "Suppl emental Reactor Vessel Surveillance Program Guidelines," dated December 1991.

The current Unit 1 surveill once program, which was established before theissuance of Appendix H to 10 CFR 50 in 1973, was designed in accordance with was manufactured. Upon issuance of the Unit 1 fu11 power license (DPR-80)the requirements of ASTM E 185-70, the standard in effect when the vessel that the Unit 1 surveillance program complies with revised Appendix Hon November 2, 1984, the NRC Staff noted in the associated safety eval nation requirements (effective July ze, 1983) and that a previously issued exemption from certain Appendix H requirements was no 1 anger required.

The suppTemental surveilTance program consists of four new capsuTes. These capsules are scheduled to be instaTed in September 1992 during the first if service inspection of the reactor vesseT, when aT of the Unit 1 reactor vesseT internaTs wiT be removed. Therefore, PG&E requests that the NRC 1992, for use beginning in Unit 1 Cycle 6.approve the encTosed suppTementaT surveiTTance program prior to September 1,

Sincerely, x

Gregory M.§he§er cc: Ann P. Hodgdon John B. Martin phiiip J. Morritl Harry Rood Howard J. Wing CPUC Diablo Distribution

Enclosure

El 9204080 *§é033156995/85K/RLJ/1249 c'{ PDR Avé8§ 05000275 ER-243 400X(II P PDR ~~,>>>>~ ,,

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t PG Letter No. DCL-92-072wQ

ENCLOSURE

DIABLO CANYON UNIT 1 SUPPLEMENTAL REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM This enclosure describes the proposed supplemental reactor vessel radiationsurveillance program for DCPP Unit 1. The purpose of the supplemental limiting vessel beltline materials for the period beyond which the originalsurveillance program is to obtain additional embrittlement data for the program by incorporating, where possible, requirements of the latest editionsurveillance program was designed and to improve the overall surveillance of ASTM E 185 and the December 1991 Electrical Power Research Institute (EPRI)

Supplemental Reactor Vessel Surveillance Guidelines (Reference 1).

DESCRIPTION OF CURRENT APPROVED SURVEILLANCE PROGRAM The existing NRC approved surveiTTance program is described in detail inSection 5.2.4.4 of the DCPP Units 1 and 2 FSAR Update and in a Westinghouse report (Reference 2). The surveilTance program, established prior to the existence of 10 CFR 50, Appendix H, was designed in accordance with the requirements of ASTM E 185-70, the standard in effect when the vesseT was manufactured. The major elements of the program are described herein.

The surveiTTance program consists of three capsuTes, which contain the Timi ting beTtline weld metal, Timi ting shelT pTate, and weTd heat affected zone (HAZ) from an intermediate sheTT pTate. In addition, these capsuTes neutron dosimetry, and thermaT monitors. Table 1 Tists' the type and number ofcontain Heavy Section SteeT Technology (HSST) 02 correTation monitor materiaT,

specimen in the capsuTes. These are designated Type II capsuTes.

In addition, there are five capsul es that contain onTy intermediate sheTl pTate and HSST 02 correl ation monitor material. These are designated Type I capsuTes. The type and number of specimens in these capsuTes are at s0 Tisted material (weld metall), they are considered "standby" capsuTes and wilT not bein Table 1. Since the Type I capsules do not contain the Timi ting beltTine analyzed.

The originaT approved capsuTe withdrawal scheduTe included both Type I and II capsules. A revised withdrawaT scheduTe, which treats aT the Type I capsuTes as "standby", was approved by the NRC in September 1991. The revised withdrawaT scheduTe is shown in Table 2; capsul e Tocations are shown in Figure 1.

Although the surveillance program vlas designed to the 1970 edition of ASTM E 185, it incorporates most of the key elements of the later approved editions of ASTM E 185. Elements of approved editions of E 185 which are not incorporated into the current Unit 1 surveillance program include the following:

1. Five capsul es containing Timi ting weld metal, base metal, and HAZspecimens are required, based on the estimated shift in ni] ducti1it.y reference temperature, to meet the requirements of ASTM E 185-73. The Unit 1 program has three capsuTes with the Timi ting material s. The fivestandby capsules do not contain the Timi ting materials.

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2. Transverse oriented base metal Charpy and tension specimens are required.

The Unit 1 program has 1 ongitudinally oriented base meta] Charpy and tension specimens.

3. Twel ve Charpy specimens representing each material are now required.

Unit 1 program has eight Charpy specimens for each material. The

4. Capsule 1 ead factors of one to three are currently recommended.1 read factors are 1.2 and 3.6. The Unit

These differences were reviewed by the NRC Staff and they concluded that thesurveillance program complied with the intended purpose of Appendix H. An exemption to Section II.B of Appendix H was therefore granted in Supplement 9to the DCPP Safety Evaluation Report, dated June 1980. This exemption was incorporated as Section 2.D into the Unit 1 fuel load/low power license (DPR-76) issued September 22, 1981. Thereafter, upon issuance of the Unit 1 full power license (DPR-80) on November 2, 1984, the NRC Staff noted in the associated safety evaluation that the Unit 1 surveillance program complies with revised Appendix H requirements that became effective on July 26, 1983.

Hence, the previously granted exemption to Appendix H was no longer requiredand was therefore not included in DPR-80.

DESCRIPTION OF THE SUPPLEMENTAL SURVEILLANCE PROGRAM A1 though the current NRC-approved surveillance program is adequate to monitor vessel embrittl ement through 40 years of operation, its design does not accommodate operational periods significantly beyond 40 years and cannot supply at of the embrittiement data necessary to support a longer period of operation. A supplemental surveillance program, which includes additional capsules, is needed for future embrittiement management.

The purpose of the suppl ementat surveillance program is to provide sufficient embrittlement data on the limiting material s to permit effective management of vessel embrittlement during the entire operating life of the vessel.

Guidance for the design of the supplemental surveillance program was takenfrom ASTM E 185-82 and the draft edition of ASTM E 185-92, which is in the final stage of approval by the ASTM E 10 Committee. Additional guidance was taken from an EPRI report that provides guidance on supplementing an existing surveillance program (Reference 1). The ASTM E 10 Committee is studying the supplemental surveillance guidelines in Reference 1 for possible adoption as a standard.

The new surveillance program, which incorporates both the existing First, it provides embrittlement data through 48 effective fu11 power yearssurveillance capsules and the suppl emental capsules, meets three goal s:

(EFPY) or approximately 60 years of operation. Second, it provides a "standby" capsule that wi11 reside in a low 1 ead factor location and be held necessary data to demonstrate the effectiveness of thermal annealing, shouldin reserve should it be needed in the future. Third, it provides the that process be used at some future time.

The supplemental surveillance program consists of four new capsules, designated A, B, C, and D, which will be installed in September 1992 during

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because the lower vessel internals will be removed and stored in the refuelingthe first if service inspection of the reactor vessel. This time was selected cavity, making the existing capsules and holders accessible.

The design of the capsule enclosures will be identical Io the originalenclosures, except the length has been increased to accommodate more

specimens.

MATERIALS Although no archive surveillance material from the limiting weld and plate was materials, has been obtained. The existing capsules contain the base materialavailable, suitable surrogate material, representative of the limiting from lower shell plate 4106-3, which is the limiting base material at fluences through 32 EFPY. The new capsules will contain material from the intermediate shell plate 4107-1, which is the limiting base metal at 48 EFPY. The chemical compositions of both materials are shown in Table 3. Preirradiation Charpy and tension testing will be done in accordance with ASTM E 185-82 to establish the baseline with which to compare the irradiated behavior of this material.

The existing capsules contain weld metal from the limiting lower shell new capsules will contain weld metal specimens from a reactor vessel nozzlelongitudinal welds made with weld wire heat 27204 and Linde 1092 flux. The dropout supplied by ABB Combustion Engineering made with weld wire heat 27204 and Linde 1092 flux. The composition, given in Table 3, is very similar to the composition of the existing surveillance weld and will make an excellent has been completed and will be used as the baseline with which to compare thesurrogate. Preirradiation Charpy testing, in accordance with ASTM E 185-82, irradiated behavior of this material .

Capsul es B and D wi11 contain material from broken we d meta] Charpy specimens from Capsule S. These wi11 be reconstituted after reirradiation and compared with the surrogate material. Capsul e B wi11 also contain two untested wedge opening 1 oaded (HOL) specimens from Capsul e s.

Capsules A and B wilt contain HSST 02 correlation monitor material. Capsules c and D will not contain correlation monitor material because they will be used to develop plant specific thermal annealing recovery and reirradiation response and will not benefit from correlation monitor material .

Following the guidance from a draft of the next revision of ASTM E 185-92, HAZ specimens wi11 not be included in the supplemental surveillance program.

Charpy specimens supplied by EPRI. These are part of an industry researchIn addition to the Unit 1 material s, Capsutes B, C, and D wi11 at so contain program and are not part of the Unit 1 surveillance program. Testing of thesespecimens is the responsibi1it.y of EPRI.

,Pa PROGRAM DESIGN To provide room for the four new surveillance capsules, Capsules T, Y, and z will be removed and Capsule v relocated at the end of Cycle 5 (EOC 5), in September 1992. This will provide four unoccupied holders for the new capsules. In accordance with the existing withdrawal schedule, Capsule Y will be tested. Capsules T and Z will be stored. Capsule V will be moved from its

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• I. _ . .

present low lead factor location to a high lead factor location where Capsule S used to be located so it will accumulate fluence at a faster rate. The newcapsules will be installed in the vacant holders at locations 184°, 40°, 140°,

and 320°, as shown in Figure 2.

An explanation of the purpose o~f each capsule, old and new, in the proposed surveillance program is provided below. The first five capsules, S, Y, V, B, limitations of the original program and available materials. They willand A, constitute a modern ASTM E 185 surveillance program, within the provide embrittlement data through 48 EFPY. The remaining Capsules c and D will demonstrate the response of the vessel materials to thermal annealing and the rate of reembrittlement during reirradiation after annealing. Table 4 shows the proposed withdrawal schedule for all the capsules. Table 5 shows the type and number of test specimens and dosimeters in each new capsule.

Capsule s (Type II)This capsul e was tested at the end of Cycl e 1 (Reference 3).

Capsule Y (Type 11)This capsule will be tested at the end of Cycle 5 (September 1992) after it has accumulated the fluence approximately equivalent to the vessel 1/4 t location at 32 EFPY.

Capsule v (Type II)This capsule wi11 be tested at approximately 12.9 EFPY after it has accumul ated the f1 uence equival ent to the vessel inside surface at 32 EFPY.

Capsule B (New)

This capsule wi11 be inserted at EOC 5 and tested at approximately 19.2 EFPYafter it has accumul ated the f1 uence equival ent to the vessel inside surface

at 48 EFPY.

Capsule A (New)This capsule wi11 be inserted at EOC 5 and wi11 remain in the vessel throughout the vessel Lifetime. It is a standby capsul e.

Capsule c (New)

This capsule will be inserted at EOC 5 and tested at approximately 14.8 EFPYafter it has accumulated the fluence equivalent to the vessel inside surface

at 32 EFPY. One-half of the specimens will be tested before annealing and one-half will be tested after annealing. This capsule will demonstrate thetoughness recovery after thermal annealing. It will also compare the irradiation response of the surrogate weld metal to the original surveillance weld metal from Capsule v.

Capsul e D (New) .

This capsul e wi11 be inserted at EOC 5 and removed from the vessel at approximately 14.8 EFPY after it has accumul ated the fiuence equivalent to the vessel inside surface at 32 EFPY. It wi11 be anneal ed and reinserted into the vessel and removed at approximately 19.2 EFPY after it has accumul ated the demonstrate the degree of reembritti ement after thermal annealing.fluence equival ent to the vessel inside surface at 48 EFPY. This capsul e wi11

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Capsules T and z (Type 1) l, These capsules'contain only base meta] specimens and wi11 be removed at EOC 5 to make room for the new capsules and stored.

Capsules U, X, and w (Type I)

These capsules contain only base metal specimens and will remain in the vessel. These capsules will be treated as standby capsules.

References

1. Suppl emental Reactor Vessel Surveillance Program Guidelines, EPRI Project2975, December 1991.

2. Pacific Gas and Electric Company Diablo Canyon Unit 1 Reactor' Vessel Radiation Surveillance Program, Westinghouse E1 ectric Corporation, WCAP-8465, January 1975.

3. Analvsis of Capsule S from the Pacific Gas and Electric Companv Diablo Canyon Unit 1 Reactor Vessel Radiation Surveillance Program, Westinghouse Electric Corporation, WCAP-11567, December 1987.

l

I

I

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TABLE 1

CONTENTS OF EXISTING UNIT 1 SURVEILLANCE CAPSULES

TYPE I CAP$ULE$(3) TYPE II cApsuLEs(b)

Charpy Tension HOL Charpy Tension voL Weld Metal - " -8 2 2 (original 27204)

Base Meta]

(Plate 4106-3) 8 1 2 8 2 2

Base Meta] 8 1 2 (Plate 4106-2)

Base Metat 8 1 2 (Plate 4106-1)

of Heat Affected Zone M - - - - 8 .-.n..ll H "

(PI ate 4106-3)

Correlation Monitor 8 lull!- - 8 - - - - -

(Plate HSST-02)

(a) Type I capsul es contain the following dosimetry and thermal monitors:Dosimeter wires: copper, nickel, aluminum-0.15% cobal t (cadmium shielded and unshielded)

Thermal monitors: 579°F and 590°F melt points

F (b)

Type II capsul es contain the following dosimetry and thermal monitors Dosimeter wires: copper, nickel, a1uminum-0. 15% cobalt (cadmium shielded and unshielded)

Fission dosimeters: uranium 238 (cadmium shielded), neptunium 237 Thermal monitors: 579°F and 590°F melt points(cadmium shielded)

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n TABLE 2

CURRENT APPROVED UNIT 1 SURVEILLANCE CAPSULE WITHDRAWAL SCHEDULE

REM(0EVFApLy)T(1arE CAPSULE LOCATION LEAD FACTOR

S 320° 3.57 1.26 (Removed)

Y 40° 3.57 5.2

v 184° 1.17/3.57(b) 15.0

U 356° 1.17 Standby

X 176° 1.17 Standby

w 40 1.17 >> Standby

T 140° 3.57 Standby

Z 220° 3.57 Standby

i

l

(a) Approximate effective fu11 power years from plant startup, plus orminus one fuel cycle

(b) Capsule v wi11 be moved to a 3.57 lead factor 1 ocation at approximately9 EFPY

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TABLE 3

COMPOSITION OF CAPSULE TEST MATERIALS

Composition (wt%)

c Mn p CrSSiNiMo Cu

original Surveillance Base 0.20 1.33 0.011 0.012 0.25 0.46 2 0.46 0.10 Metal, Interm?d3ate Shell PI ate B4106-3 a

Supplemental Base Meta] 0.25 1.36 0.011 0.014 0.24 0.56 .- 0.48 0.13 Lower PI ateslgll 84107-1

original Surveillance Weld 0.11 1.35 0.014 0.025 0.25 1.00 0.05 0.50 0.20 Wire heat 27204 Linde 1092 f1ux(b)

Surrogate Surveillance 0.13 1.23 0.014 0.009 0.21 1.00 0.05 0.54 0.22 Neld Wire heat 27204 Lynde 1092 f1ux(c)

(a) Lukens Steel CMTR

(b) Average of four analyses

(c) Combustion Engineering analysis

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TABLE 4

PROPOSED UNIT 1 REACTOR VESSEL SURVEILLANCE PROGRAM WITHDRAWAL SCHEDULE

LEAD REMOVAL ME F ENCE CAPSULE TYPE LOCATION FACTOR (EFPY)885 (1o3n/cmZ)

S II 320° 3.57 1.26 (Tested) 0.298<b)

Y II 40° 3.57 5.8 (Tested) 1.21

1 V II 320° 3.57 12.9 1.49(°)

U I 356° 1.17 Standby lil-

X I 176° 1.17 Standby w I 40 1.17 Standby -.-.al

T I 140° 3.57 5.8 (Removed) 1.21

z I 220° 3.57 5.8 (Removed) 1.21 B(d) " 40° 3.57 19.2 2.20(&)

0(d) - - 140° 3.57 14.8 1.49(>

0(4) - - 220° 3.57 19.2(f) 2.20(e)

A(d) - -184°1.17Standby -lull-

(a) Approximate effective full power years from pl ant startup

(b) Neutron fluence measured from capsule dosimeters (C) Projected fluence at vessel inner wall at 32 EFPY

(d) Inserted at 5.8 EFPY (EOC 5)

(e) Projected fluence at vessel inner wall at 48 EFPY (f) Anneal at 14.8 EFPY and reinsert

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3- TABLE 5

CONTENTS OF SUPPLEHENTAL SURVEILLANCE CAPSULES v

CAPSULE A(8) CAPSULE Bla) CAPSULE 9(3) CAPSULE D(b)

Charnv Tension Charnv Tension voL Charuv Tension Charpv Tension

Weld Metal (Surrogate 27204) 15 - - 15 - - 30 - - 15 - -

Base Metal (Plate 4107-1) 15 3 15 3 c! In 15 3 15 3

Correlation Monitor (HSST-02 Plate) 12 Q - 12 - Z 2 2 - 2 " 2 - "

Capsule S We1 d Metal - - _ - 10(0) - - 2 _ - 10(C) - -

(Original 27204)

(a) Capsule will contain the following dosimetry and thermal monitors:

Dosimeter wires: copper, iron, nickel and aluminum-0.15% cobalt (cadmium shielded and unshielded)

Fission dosimeters: neptunium-237 (cadmium oxide shielded), and uranium 238 (cadmium oxide shielded)

Thermal monitors: 97.5% Pb, 2.5% Ag (579°F melt point), 97.5% Pb, 1.75% Ag, 0.75% Sn (590°F melt point)

(b) Capsule D will contain the following dosimeters:

Dosimeter wires: copper, iron, nickel and aluminum-0.15% cobalt (gadolinium shielded and unshielded)

Fission dosimeters: neptunium 237 (gadolinium shielded) and uranium 238 (gadolinium shielded)

Thermal monitors: will not be provided because annealing temperature will exceed the melting point of thermal monitors (c) Broken we d meta] and HAZ Charpy specimens from capsul e S, suitable for reconstitution

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.9 4 " FIGURE 1

5 ii 4 PLAN vIEw OF EXISTING SURVEILLANCE CAPSULES 1I

1 O. L

REACTCR - -THERMAL SHIELD VESSEL

SURVEILLANCE CAPSULE (TYPICAL) 'REACTOR CORE

270' 90'

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.AFTER INCORPORATION OF 4Ew~gAn§uLEs

4 O.

REACTOR -

VESSEL -THERMAL SHIELD

SURVEMLANCE

<32 CAPSULE (TYPICAL) 'REACTOR 443 CORE

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