ML24089A011

03.25.24 Corrected Petitioners Opening Brief

ML24089A011
Person / Time
Issue date:	03/25/2024
From:	Eric Michel NRC/OGC
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Case 23-3884
Download: ML24089A011 (1)

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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 OPENING BRIEF

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

March 20, 2024 Corrected March 25, 2024

CORPORATE DISCLOSURE STATEMENT In accordance with Federal Rule of Appellate Procedure 26.1, Petitioners certify that they are nonprofit organizations that have no parent or subsidiary entities. No Petitioners have stock, and therefore, no publicly held company owns 10 percent or more of its stock.

Table of Contents Table of Authorities.iv Glossaryviii JURISDICTIONAL STATEMENT...1 A. Hobbs Act Jurisdiction1

B. Standing of Petitioners1

STATEMENT OF ISSUES3 STATUTORY ADDENDUM.3 STATUTORY AND REGULATORY BACKGROUND..3 A. Atomic Energy Act.3

B. Implementing Regulations and Guidance for Reactor Vessel Surveillance Programs..4

STATEMENT OF THE CASE..6 I. INTRODUCTION6

II. STATEMENT OF THE FACTS9

A. Construction Permit and Operating License for Diablo Canyon Unit 1...9

B. Initial Program for Monitoring Unit 1 Reactor Pressure Vessel..9

C. NRC promulgation of license renewal rule11

D. PG&Es Supplemental Surveillance Program11

1. PG&E application and NRC approval11

2. Withdrawal and testing of Capsules S, Y, and V13

E. License Amendments for Recovery of Time for Construction and Low-Power Testing.14

1. 1995 license amendment to recover thirteen-year construction period.....14

2. 2006 license amendment to allow recovery of low-power testing interval14

a. NRC policy for recovery of low-power testing time...14

b. 2006 license amendment for Diablo Canyon Unit 116

F. NRC Decisions Extending Schedule for Withdrawal of Capsule B From Unit 1 Pressure Vessel20

1. 2008 Extension Decision: from 2007 to 2009..20

2. 2010 Extension Decision: from 2010 to 2012..21

3. 2012 Extension Decision: from 2012 to 2022..23

4. 2023 Extension Decision: from 2022 to 2023 or 2025 or indefinitely delayed..25

G. Subsequent Treatment of PG&Es 2009 License Application:

Submission, Withdrawal, and Re-Submission..26

H. Petitioners Hearing Request and Request for Emergency Action by the Commissioners..27

I. Petition for Review30

STANDARD OF REVIEW.30

SUMMARY

OF THE ARGUMENT..32

ARGUMENT..36

ii

I. THE NRC VIOLATED THE ATOMIC ENERGY ACT WHEN IT AMENDED THE OPERATING LICENSE FOR DIABLO CANYON UNIT 1 WITHOUT PROVIDING PUBLIC NOTICE OR THE OPPORTUNITY TO REQUEST A HEARING AND WITHOUT FINDING THAT THE LICENSE AMENDMENTS WERE ADEQUATE TO PROTECT PUBLIC HEALTH AND SAFETY...36 A. The NRCs 2006 License Amendment Decision Conditioned the Three-Year Extension of the Unit 1 Operating License on Specific Requirements for PG&Es Reactor Vessel Surveillance Program..36

B. The NRC Has Repeatedly Amended the License Condition Imposed on PG&E by the 2006 License Amendment37

C. The NRC Violated the Atomic Energy Act by Failing to Provide Public Notice or a Hearing Opportunity Each Time It Extended the Schedule for Withdrawal of Capsule B from the Unit 1 Pressure Vessel38

D. The NRC Violated the Atomic Energy Act by Failing to Evaluate Whether Changes to PG&Es License Condition Would Provide Adequate Protection to Public Health and Safety...39

II. THE NRC VIOLATED THE ATOMIC ENERGY ACT AND THE ADMINISTRATIVE PROCEDURE ACT WHEN IT DENIED PETITIONERS A HEARING ON THE 2023 EXTENSION OF THE SCHEDULE FOR WITHDRAWING CAPSULE B40 III. THE NRCS ABANDONMENT, WITHOUT A REASONED EXPLANATION, OF THE SURVEILLANCE PROGRAM IT IMPOSED ON PG&E AS A CONDITION OF EXTENDING THE TERM OF PG&ES LICENSE WAS UNREASONABLE AND ARBITRARY AND CAPRICIOUS..41

A. The Extension Decisions Were Unreasonable.41

B. The Extension Decisions Were Arbitrary and Capricious...42

IV. CONCLUSION AND REQUEST FOR RELIEF43

iii

Table of Authorities Judicial Decisions Alaska Ctr. for the Envt v. United States Forest Serv.,

189 F.3d 851 (9th Cir. 1999)...31, 41

Alaska Wilderness Recreation & Tourism v. Morrison, 67 F.3d 723 (9th Cir. 1995)...30, 40-41

Cal. Ex. rel. Lockyer v. USDA, 575 F.3d 999, 1011 (9th Cir. 2009)..41

Citizens Awareness Network v. NRC, 59 F.3d 284 (1st Cir. 1995)3-4, 33, 37, 38

Deukmejian v. NRC, 751 F.2d 1287, 1314 (D.C. Cir. 1984).37

Encino Motorcars, LLC v. Navarro, 579 U.S. 211(2016).42

FCC v. Fox Television Stations, Inc., 556 U.S. 502 (2009)..42

Guard v. U.S. Nuclear Reg. Commn, 753 F.2d 1144 (D.C. Cir. 1985)...31, 42

Hatch v. FERC, 654 F.2d 825 (D.C. Cir. 1981).42

Heckler v. Chaney, 470 U.S. 821 (1985)...30

Honeywell Intl v. NRC, 628 F.3d 568 (D.C. Cir. 2010)42

Hunt v. Wash. State Apple Adver. Commn, 432 U.S. 333 (1977)...2

Lujan v. Defenders of Wildlife, 504 U.S. 555 (1992)..2

Marsh v. Oregon Natural Resources Council, 490 U.S. 360 (1989).31

Motor Vehicle Manufacturers Association v. State Farm Mutual Automobile Insurance Co., 463 U.S. 29, 42 (1983).42

iv

Northcoast Envtl. Ctr. v. Glickman, 136 F.3d 660 (9th Cir. 1998).31, 41

Price Rd. Neighbor. Assn v. U.S. Dept. of Transp,

113 F.3d 1505 (9th Cir. 1997)31

Public Citizen v. NRC, 573 F.3d 916 (9th Cir. 2009).30

Safe Energy Coalition v. U.S. Nuclear Reg. Commn, 866 F.2d 1473 (D.C. Cir. 1989)..30

San Luis Obispo Mothers for Peace v. NRC,

449 F.3d 1016 (9th Cir. 2006).31

San Luis Obispo Mothers for Peace, et al. v. U.S. Nuclear Reg. Commn, No.23-852 (pending in Ninth Circuit)26

Students for Fair Admissions, Inc. v. President and Fellows of Harvard University, 143 S.Ct. 2141 (2023)..2

Union of Concerned Scientists v. NRC, 711 F.2d 370 (D.C. Cir. 1983)37

Statutes Atomic Energy Act

42 U.S.C. § 2133...34, 39

42 U.S.C. §§ 2131-21334

42 U.S.C. § 2232...3

42 U.S.C. § 2237...4

42 U.S.C. § 2239(a)(1)(A)..1, 4, 38

42 U.S.C. § 2239(b)..1

v

Hobbs Act

28 U.S.C. § 2342(4)1

28 U.S.C. § 2344.1

Administrative Procedure Act

5 U.S.C. § 702...1

Administrative Decisions Calvert Cliffs 3 Nuclear Project, L.L.C. and Unistar Nuclear Operating Services, L.L.C., 70 N.R.C. 911 (2009).2

Cleveland Electric Illuminating Co.,

44 N.R.C. 315, 317 (1996)5, 6, 10

Yankee Atomic Electric Co., 34 N.R.C. 3 (1991)6

Regulations 10 C.F.R. § 2.109(b)..19 10 C.F.R. § 2.206...29 10 C.F.R. § 50.61.....4 10 C.F.R. § 50.92(a)....34, 39 10 C.F.R. Part 50, Appendix H4 Guidance documents ASTM E 185...5, 6, 10, 12, 18, 23 ASTM E 185-70..10, 17, 19, 21, 23, 29, 32, 33, 36, 27 ASTM E 185-8218, 19, 20, 22, 23, 32, 36, 38, 43 NUREG-1801, Generic Aging Lessons Learned (GALL) Report20

vi

Federal Register Notices Notice of License Renewal Application, 75 Fed. Reg. 3,493 (Jan. 21, 2010)..21, 26

Final Rule, Fracture Toughness Requirements for Light Water Reactor Pressure Vessels, 60 Fed. Reg. 65,456, 65,457 (Dec. 19, 1995)6, 7, 32

Final Rule, Reactor License Renewal, 56 Fed. Reg. 64,943 (Dec. 13, 1991).11

Notice of Withdrawal of License Renewal Application, 83 Fed. Reg. 17,688 (Apr. 23, 2018)26

vii

GLOSSARY

APA Administrative Procedure Act

ASTM American Society for Testing of Materials

Commission U.S. Nuclear Regulatory Commission

EFPY Effective Full Power Years

EOL End of Life

NRC U.S. Nuclear Regulatory Commission

PG&E Pacific Gas & Electric Co.

viii

JURISDICTIONAL STATEMENT

A. Hobbs Act Jurisdiction

This case involves an appeal of a final Order entered on October 2, 2023

(Denial Order), by the United States Nuclear Regulatory Commission (the

NRC or Commission) regarding the operating license held by Pacific Gas and

Electric Co. (PG&E) for Unit 1 of the Diablo Canyon nuclear power plant. 1-ER-003. 1 The Commissions Order is reviewable by this Court under the Atomic

Energy Act (AEA), 42 U.S.C. § 2239(b); the Hobbs Act, 28 U.S.C. § 2342(4);

and the Administrative Procedure Act (APA), 5 U.S.C. § 702. The appeal was

timely filed pursuant to 28 U.S.C. § 2344, because it was docketed on December 1,

2023, within 60 days of the date of the Commissions Order.

B. Standing of Petitioners

Section 189a of the Atomic Energy Act, 42 U.S.C. § 2239(a), requires the NRC

to grant a hearing upon the request of any person whose interest may be affected

by the proceeding. Petitioners, with the support and authorization of their

members, seek a hearing before the NRC to address their concerns that PG&Es

ongoing lack of knowledge regarding the condition of the Unit 1 pressure vessel

poses an unacceptable risk to their health and safety. See 2-ER-043 ER-045, 2-

1 This appeal concerns only the Unit 1 pressure vessel and does not include Unit 2.

ER-119 ER-121. And these concerns are germane to the purposes of the

organizations. Id.

The NRC has found that petitioners who live within approximately fifty miles

of a nuclear reactor meet the judicial test for an affected person, i.e., injury in

fact, causal connection, and redress by a favorable decision. Calvert Cliffs 3

Nuclear Project, L.L.C. and Unistar Nuclear Operating Services, L.L.C., 70

N.R.C. 911, 917 (2009) ( citing Lujan v. Defenders of Wildlife, 504 U.S. 555, 572

n.7 (1992)). Each of the Petitioners meets this test through members who live,

work, and own property within 50 miles of the Diablo Canyon reactors. 2-ER-119 ER-121.

Petitioners also meet the judicial test for organizational standing because:

[their] members would otherwise have standing to sue in their own right; (b) the interests [they] seek[] to protect are germane to the [organizations]

purpose; and (c) neither the claim asserted nor the relief requested requires the participation of individual members in the lawsuit.

Students for Fair Admissions, Inc. v. President and Fellows of Harvard University,

143 S.Ct. 2141, 2157 (2023) (quoting Hunt v. Wash. State Apple Adver. Commn,

432 U.S. 333, 343 (1977)).

2

STATEMENT OF ISSUES

1. Did the NRC violate the Atomic Energy Act and the APA by denying Petitioners request for a hearing on the NRCs decision to extend the surveillance schedule for the Unit 1 pressure vessel and by failing to give any reason for its decision?

2. In extending the scheduled date for PG&E to remove Capsule B from the Unit 1 pressure vessel for embrittlement testing - from 2009 until 2024 and possibly beyond -- did the NRC amend a condition in PG&Es operating license, thereby requiring compliance with the procedural requirements of the Atomic Energy Act for license amendments?

3. In extending the schedule for removing Capsule B from the Diablo Canyon Unit 1 pressure vessel without acknowledging that it had altered the terms of PG&Es amended operating license or explaining the reasons for those changes, did the NRC violate the Atomic Energy Act and the APA?

STATUTORY ADDENDUM

In accordance with Ninth Circuit Rule 28-2.7, pertinent statutes and

regulations are included in the Addendum to this Brief, beginning on Page A-1.

STATUTORY AND REGULATORY BACKGROUND

A. Atomic Energy Act

The NRC is responsible for ensuring that operation of nuclear reactors

provides adequate protection to the health and safety of the public. 42 U.S.C. §

2232. Operation of a nuclear reactor must be carried out under a license that the

NRC has determined will meet this statutory standard. If the licensee wishes to

modify the facility or take actions not specifically authorized by the license, the

licensee must first seek an amendment to its license from the Commission. Citizens 3

Awareness Network v. NRC, 59 F.3d 284, 287 (1st Cir. 1995) (citing 42 U.S.C. §§

2131-2133, 2237 (1988)).

Section 189a of the Atomic Energy Act requires the NRC to provide

interested members of the public with a prior opportunity for a hearing on any

proposed decision to amend, grant, or revoke an operating license for a nuclear

facility. 42 U.S.C. § 2239(a)(1)(A). The NRC must also provide public notice in

the Federal Register of its proposed licensing decisions. Id.

B. Implementing Regulations and Guidance for Reactor Vessel Surveillance Programs

NRC regulation 10 C.F.R. § 50.61 establishes requirements that nuclear

reactor licensees must satisfy in order to demonstrate that reactor vessels in U.S.

pressurized light-water reactor facilities will have adequate protection against the

consequences of pressurized thermal shock events throughout their service lives.

Requirements for reactor vessel surveillance programs are found in § (b)(2) and 10

C.F.R. Part 50, Appendix H. As summarized by the Commission, Appendix H:

sets forth a surveillance program to monitor the fracture toughness of beltline materials in light-water reactor vessels. Appendix H directs licensees to attach a particular number of surveillance capsules to specified areas within the reactor vessel, typically near the inside vessel wall at the beltline. Each capsule contains a number of material specimens that remain exposed to radiation during plant operation. Under the Appendix H surveillance program, licensees must periodically withdraw capsules from the reactor vessel. Capsule removal permits the material specimens to be tested for changes in ductility and fracture toughness - effects of the neutron irradiation and elevated temperatures in a given reactor pressure vessel.

4

Cleveland Electric Illuminating Co., 44 N.R.C. 315, 317 (1996) ( Cleveland

Electric).

In Section III.B, Appendix H also requires that a reactors surveillance

program must satisfy ASTM E 185, the established industry guidance of the

American Society for Testing of Materials. ASTM E 185 provides licensees with

the criteria for determining both the minimum number of surveillance capsules that

need to be installed within the reactor vessel at the start of the plants life, and

when in the plants life -- measured in effective full-power years [EFPY] -- a

capsule should be withdrawn for evaluation. Cleveland Electric, 44 N.R.C. at 317.2

ASTM E 185 E has been revised multiple times since the first edition in

1970. With respect to the edition of ASTM E 185 that is applicable to a particular

reactor,Section III.B of Appendix H provides:

The design of the surveillance program and the withdrawal schedule must meet the requirements of the edition of the ASTM E 185 that is current on

2 EFPY is an irradiation or fluence-based measure of the life of a reactor. At the end of a reactors design life of forty calendar years (or end or life [EOL]), a reactor typically has accumulated a fluence of 32 EFPY. At the end of a renewed operating license totaling 60 years of operation, a reactor typically will have accumulated a fluence of 48 EFPY. 2-ER-245 ER-247.

Because capsules are located closer to the reactor core than the wall of the reactor vessel itself, their fluence will be higher than the fluence of the reactor vessel itself. And some capsules are deliberately located closer to the core than others such that their fluence may be significantly higher. See, e.g., 2-ER-247 (proposing relocation of Capsule V in order to accumulate fluence at a faster rate.).

5

the issue date of the ASME code to which the reactor vessel was purchased; for reactor vessels purchased after 1982, the design of the surveillance program and the withdrawal schedule must meet the requirements of ASTM E 185-82. For reactor vessels purchased in or before 1982, later editions of ASTM E 185 may be used, but including only those editions through 1982.

The Commission has held that if an operating license provides for

compliance with the ASTM standard, changes to the surveillance schedule may be

made without amending the reactors operating license. Cleveland Electric, 44

N.R.C. at 328. However, the licensee must seek NRC review before changing the

schedule, to allow the NRC to verify whether the changed schedule continues to

conform to the applicable edition of ASTM E 185. Id., 44 N.R.C. at 328.

STATEMENT OF THE CASE

I. INTRODUCTION

Petitioners seek their rightful opportunity to hear from NRC, and be heard, on a

series of decisions that carry inordinate safety risks for communities and the

environment surrounding the Diablo Canyon Nuclear Power Plant regarding a

schedule for monitoring the integrity of the Unit 1 reactor vessel. As the receptacle

that holds the highly radioactive core of a nuclear reactor, the pressure vessel is

perhaps the most important single component in the reactor coolant system.

Final Rule, Fracture Toughness Requirements for Light Water Reactor Pressure

Vessels, 60 Fed. Reg. 65,456, 65,457 (Dec. 19, 1995). See also Yankee Atomic

Electric Co., 34 N.R.C. 3, 12 (1991) ( Yankee Rowe ) (pressure vessel is one of

6

the key components of a reactor.) Because it has no backup, the pressure vessel

must be protected continuously against the risk of fracture and failure, which could

lead to core melt and catastrophic consequences to public health and safety and the

environment. 2-ER-079. See also 60 Fed. Reg. at 65,456 ([m]aintaining the

structural integrity of the reactor pressure vessel... is a critical concern related to

the safe operation of nuclear power plants.).

Petitioners seek review of four consecutive unlawful decisions by the NRC,

starting in 2008 and culminating in the 2023 decision upheld by the NRC in the

Denial Order that is the subject of this Petition for Review. These decisions

cumulatively extended, by a period of more than fourteen years and perhaps

indefinitely, the schedule for withdrawing Capsule B from the Unit 1 pressure vessel and testing it for embrittlement. 3 In each of these decisions, and without

explanation or rationale, the NRC abandoned a 2006 license amendment that had

imposed a specific reactor vessel surveillance schedule on PG&Es Unit 1

3 See the 2008 Extension Decision, extending the withdrawal of Capsule B from 2007 to 2009, 1-ER-029 (discussed below in Section F.1); the 2010 Extension Decision, extending the withdrawal of Capsule B from 2010 to 2012, 1-ER-023 (discussed below in Section F.2); the 2012 Extension Decision, extending the withdrawal of Capsule B from 2012 to 2022, 1-ER-017 (discussed below in Section F.3); and the 2023 Extension Decision, extending the withdrawal of Capsule B from 2022 to 2024, 2025, or indefinitely, 1-ER-009 (discussed below in Section F.4).

Petitioners refer to these four decisions collectively as Extension Decisions.

7

operating license. 2-ER-155. (2006 License Amendment). The NRC had

imposed that specific surveillance schedule in exchange for granting PG&Es

request to extend its operating license term by three years, changing the license

expiration date from September 22, 2021 to November 2, 2024. 2-ER-161.

Despite the fact that these Extension Decisions qualified as license

amendments, the NRC did not publish notice of any of the Decisions in the Federal

Register. Nor did the NRC explain why -- or even acknowledge -- that it was

abandoning the license condition that the agency had imposed in 2006 in exchange

for extending Unit 1s operating license by three years. As a result, since 2021, the

NRC has allowed PG&E to operate Diablo Canyon Unit 1 in violation of the 2006

license condition on which the extended operating license term for Unit 1 is

founded. And despite the importance of the Unit 1 pressure vessel to the safety of

the reactors operation, the required 2009 inspection of the pressure vessel has

been delayed by more than fourteen years. All of this has been done without a

meaningful opportunity for public input.

In this appeal, Petitioner asks the Court to reverse and vacate the NRCs Denial

Order and cumulative Extension Decisions based on multiple violations of federal

law. First, the NRC contravened the Atomic Energy Act by failing to proactively

give public notice and offer the opportunity for a public hearing before amending

or revoking PG&Es license condition. Second, the NRC violated both the Atomic

8

Energy Act and the Administrative Procedure Act by summarily denying

Petitioners public hearing request. Finally, Petitioners seek review of NRCs

related violation of the Atomic Energy Act by completely failing to justify or even

acknowledge its abandonment of the terms of the 2006 License Amendment,

including the lack of any safety rationale for abandoning the license condition

imposed by the 2006 License Amendment.

II. STATEMENT OF THE FACTS

A. Construction Permit and Operating License for Diablo Canyon Unit 1

In 1968, the NRC issued PG&E a construction permit for Diablo Canyon Unit

1. 2-ER-226. In 1981, following completion of Unit 1 construction, the NRC

issued a low-power license for the sole purpose of testing the reactor. After three

years of low-power testing, the NRC issued PG&E a full-power operating license

for Unit 1 on November 2, 1984. The license allowed PG&E to operate Unit 1 for

forty years from the date of issuance of the construction permit, or until April 23,

2008. 2-ER-236.

B. Initial Program for Monitoring Unit 1 Reactor Pressure Vessel

In the 1970s, while construction was underway, PG&E established reactor

vessel surveillance programs for the Diablo Canyon reactor pressure vessels. 2-ER-

258. The purpose of these programs was to monitor and ensure the structural

9

integrity of reactor pressure vessels. Cleveland Electric, 44 N.R.C. at 317. This

threat to reactor vessel integrity arises from [l]ong-term exposure to neutron

radiation and elevated temperatures, causing the ductility of the reactor vessel

materials to decrease and thereby decreasing the vessel materials fracture

toughness, or resistance to fracture. Id. A significant decrease in ductility renders

the reactor vessel vulnerable to rupture if cold water were to be injected into the

reactor vessel during a loss of coolant accident. 2-ER-085. Because the reactor

vessel was purchased in the 1970s, the applicable ASTM E standard was ASTM E

185-70. 2-ER-244.

The Unit 1 surveillance program provided for placement inside the pressure

vessel of capsules containing specimens or coupons of representative metal

samples. Capsules S, Y, and V consisted of three Type II capsules containing limiting weld metal and base metal content. 4 In compliance with ASTM E 185-

70, the surveillance program scheduled capsules S, Y, and V for removal at

specific intervals over the forty-year operating life of the reactor and tested for

embrittlement characteristics. 2-ER-247. The schedule was based on projections of

when the capsules would reach certain levels of fluence -- exposure to neutron

4 Limiting materials are envisioned to be the weakest components when embrittled and hence are those that will likely fail first. 2-ER-083.

10

irradiation -- based on their location in relation to the reactor core. Five other

capsules that did not contain the limiting material were designated standby,

without a schedule for removal and testing. Id.

C. NRC promulgation of license renewal rule

In 1991, for the first time, the NRC promulgated safety regulations for the

renewal of nuclear reactor licenses to allow operation for an additional twenty

years after expiration of their initial forty-year licenses. 56 Fed. Reg. 64,943 (Dec.

13, 1991). The new regulations established standards for the management of aging

safety equipment, including reactor pressure vessels, during a twenty-year renewal

term.

D. PG&Es Supplemental Surveillance Program

1. PG&E application and NRC approval

In March 1992, PG&E applied to supplement the original Unit 1 surveillance

program by adding Capsules A, B, C, and D. 2-ER-243. The supplemental

surveillance program had two purposes: to provide embrittlement data for a

possible additional twenty-year license renewal term and to improve the overall

surveillance program by incorporating, where possible, more recent industry

and government guidance. 2-ER-244.

The proposed supplemental surveillance program incorporate[d] both the

existing surveillance capsules and the supplemental capsules, i.e., Capsules S, Y,

11

V, B, and A. 2-ER-245. PG&E described these first five capsules as a modern

ASTM E 185 surveillance program, within the limitations of the original program

and available materials. 2-ER-247. But only four of these capsules -- Capsules S,

Y, V, and B -- were given scheduled dates for withdrawal during the forty-year

operating license term. Capsule A was designated Standby, i.e., reserved for future use with no specified withdrawal date. 2-ER-247. 5

PG&Es schedule for removing and testing capsules showed that Capsules S

and Y had already been removed and tested. 2-ER-247, 2-ER-252. Capsule V was

scheduled for removal at 12.9 EFPY or approximately 2002. Id. PG&E estimated

that at 12.9 EFPY, Capsule V would provide the fluence equivalent to the vessel

surface at 32 EFPY or approximately forty years of operation. 2-ER-247. And

Capsule B was scheduled for withdrawal at EFPY 19.2, or approximately 2007. Id.

At that point, due to its location relatively close to the core, PG&E estimated that

Capsule B would provide embrittlement data through 48 effective full power

years (EFPY) or approximately 60 years of operation. 2-ER-245. See also note 2,

supra.

5 Capsules C and D were not listed as part of the modern ASTM E 185 surveillance program because they had the separate purpose of demonstrating the response of the vessel material to thermal annealing and the rate of reembrittlement during reirradition after annealing. 2-ER-247.

12

The NRC Staff approved PG&Es supplemental surveillance program,

concluding that PG&Es proposed changes to the program were acceptable

because they augment the current program, and will provide additional data on the

limiting reactor vessel materials. 2-ER-234.

2. Withdrawal and testing of Capsules S, Y, and V

In 2002, PG&E withdrew and tested Capsule V from Diablo Canyon Unit 1.

When the test showed that Unit 1 would be approaching a regulatory threshold for

concern about embrittlement at the end of its operating life in 2021, PG&E discounted the data as not... credible. 2-ER-188. 6 Instead, PG&E substituted

generic data and data from other reactors. 2-ER-081. But PG&E stated that it did

not intend to rely on generic data and data from other reactors indefinitely. Instead,

PG&E asserted that Capsule V is not the last planned capsule to be evaluated in

the [Diablo Canyon Unit 1] surveillance program. 2-ER-189.

6 Petitioners dispute whether PG&E complied with NRC guidance in rejecting this data and instead relying on generic data and data from other reactors. This dispute is one of the bases for Petitioners hearing request and their concern that the NRC lacks sufficient information about the condition of the Unit 1 pressure vessel to support a conclusion that it is operating safely. See 2-ER-081, 2-ER-100 ER-101 and discussion below in Section II.H.

13

E. License Amendments for Recovery of Time for Construction and Low-Power Testing

1. 1995 license amendment to recover thirteen-year construction period

In 1995, the NRC Staff approved PG&Es application for a license

amendment to recover or recapture the thirteen-year construction period for

Unit 1 by changing the Unit 1 operating license expiration date from April 23,

2008 to September 22, 2021. 2-ER-213, 2-ER-211. In support of its decision, the

Staff generally cited, inter alia, PG&Es comprehensive vessel material

surveillance program [that] is maintained in accordance with 10 CFR Part 50,

Appendix H that ensures the fracture toughness requirements of Appendix G are

met. 2-ER-226.

2. 2006 license amendment to allow recovery of low-power testing interval

a. NRC policy for recovery of low-power testing time

In 1999, the NRC Commissioners established a new policy of allowing reactor

licensees to recover the initial time of low-power testing of a newly-constructed

reactor under a low-power license, by adding the same amount of time to the term of a full-power license. 2-ER-191. 7 While the vast majority of reactors needed only

7 In other words, the NRC would change the commencement date of a forty-year full-power operating license from the date the low-power license was issued to the later date when the full-power operating license was issued.

14

a few months for low-power testing, for some reactors -- like Diablo Canyon Unit

1 -- it took years to complete. Id., 2-ER-195. For those cases, the NRC Staff

devised an approach to consider the aging effects on the pressure vessel caused by

allowing it to be irradiated for several more years beyond its forty-year design life.

Id., 2-ER-194.

In Commission-approved Policy Memorandum SECY-98-296, addressing a

request from the Grand Gulf Nuclear Station to recover low-power testing time, the

Staff set out its approach to the safety review as follows:

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.

2-ER-194. Consistent with this approach, the Staff reviewed the Grand Gulf

reactor vessel surveillance program and found that it will aid in adjusting the

operational conditions in order to maintain sufficient safety margin for the

prevention of brittle failure of the reactor vessel. 2-ER-204.

The Staffs review also covered specific details of the reactor vessel

surveillance program:

15

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.

2-ER-205. Based on the above, the Staff concluded that there is reasonable

assurance that the [reactor pressure vessel] 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 [Grand Gulf

Nuclear Station] license. 2-ER-205.

b. 2006 license amendment for Diablo Canyon Unit 1

In 2005, citing the new NRC policy for recovery of time spent on low-

power testing of nuclear reactors, PG&E applied to extend the Unit 1 initial

operating license term by three years - the time that had been consumed by low-power testing of Unit 1. 2-ER-170. 8 In support of its application, PG&E cited both

the original surveillance program and the supplemental surveillance program it

had proposed and agreed to when it sought in 1995 to recapture the 13-year

construction period. 2-ER-176 ER-177.

8 PG&E clarified that the proposed extension does not constitute license renewal.

2-ER-189.

16

In 2006, the NRC Staff granted the three-year license amendment. 2-ER-

155, 2-ER-152. Consistent with the NRC policy set forth in Policy Memorandum

SECY-98-269, the Staff performed a Safety Evaluation for the proposed license

amendment that assessed whether or how the addition of three years to the license

term for Unit 1 would affect the safety of the pressure vessel, including the

[i]mpact on the RVMSP [reactor vessel material surveillance program]. 2-ER-

165 ER-166.

In reviewing PG&Es license amendment application, the Safety Evaluation

cited the Staffs previous determination that:

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.

2-ER-165. The Staff also clarified that this supplemental program consisted of

four capsules, Capsule S, Y, V, and B, [that] were designated for removal from

the [Diablo Canyon Unit 1 reactor vessel]. Id.

Further, the Staff noted that Capsules S, Y, and V have been removed and

tested in accordance with the licensees program. Id. This left only the fourth

capsule, Capsule B.

As per Policy Memorandum SECY-98-296, the Safety Evaluation then

assessed whether adding three years to the operating life of Unit 1 would affect the

adequacy of the reactor vessel surveillance program. 2-ER-163. The Staff

17

concluded 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 because:

The request to recover the testing time for DCPP-1 amends the projected withdrawal for Capsule B to approximately 20.7 EFPY [i.e., around 2009],

when the capsule is projected to achieve a neutron fluence of 2.9 x 10 19 n/cm2 (E > 1.0 MeV). Therefore, the capsule will achieve a neutron fluence approximately equal to twice the projected limiting inside RV fluence for DCPP-1 at the EOL (i.e., approximately 2

• 1.43 x 10 19 n/cm2 [E > 1.0 MeV]).

2-ER-165 (emphasis added). Further, the Staff found that this amended withdrawal

schedule complies with the criterion in ASTM E185-82 for withdrawal of the

final capsule of a four capsule withdrawal program. Id.

Based on these findings, the Safety Evaluation stated:

The NRC staff has reviewed PG&Es license amendment request to recover the low-power testing time that was performed during the initial startup of the

[Diablo Canyon Units 1 and 2] reactors. The NRC has determined that authorization of the requested license may be granted based on the following conclusions:

(3) The [reactor vessel] surveillance capsule withdrawal schedules for

[Diablo Canyon Units 1 and 2] remain in compliance with the requirements of 10 CFR Part 50, Appendix H, and the ASTM E 185 versions of record for the units.

Id. (emphasis added). Accordingly, the Safety Evaluation concluded as a

general matter that:

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 18

operation in the proposed manner, (2) such activities will be conducted in compliance with the Commissions regulations, and (3) issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

2-ER-167. The license amendment therefore changed the Unit 1 operating license

expiration date from to September 22, 2021 to November 2, 2024. 2-ER-161. This three-year extension was equivalent to a new end of life EFPY of 35.2. 2-ER. 9

Thus, in approving the license amendment to add three more years to the

operating license of Unit 1, the NRC Staff upgraded the Unit 1 surveillance

program from a three-capsule program compliant with ASTM E 185-70 to a four-

capsule program compliant with ASTM E 185-82. Further, there was no question

that the Staff intended withdrawal of Capsule B to be carried out during the initial

operating license term for Unit 1, because -- by PG&Es own assertion -- the license amendment proceeding had nothing to do with license renewal. 10

9 In 2023, by operation of an NRC-issued exemption to the agencys timely renewal rule, 10 C.F.R. § 2.109(b), this termination date was changed to the date when the NRC makes a decision on PG&Es pending license renewal application.

Petitioners appeal of the exemption is pending before this Court in San Luis Obispo Mothers for Peace, et al. v. U.S. Nuclear Reg. Commn, No.23-852.

10 As PG&E stated in its license amendment application, [t]he proposed amendments do not constitute license renewal. 2-ER-174.

19

F. NRC Decisions Extending Schedule for Withdrawal of Capsule B From Unit 1 Pressure Vessel

1. 2008 Extension Decision: from 2007 to 2009

In 2008, PG&E asked the NRC to extend the schedule for removing Capsule B

from the Unit 1 pressure vessel, from 20.7 EFPY (2009) to 21.9 EFPY (2010). 2-

ER-147. PG&E did not ask for an amendment to its operating license to change the

Capsule B withdrawal schedule that the NRC Staff had incorporated into its 2006

license amendment. Instead, it simply stated that the current withdrawal schedule

did not meet the requirement of NRC license renewal guidance document

NUREG-1801, which requires that a reactor vessel surveillance program must have

a vessel material coupon that has a fluence exposure equivalent to 60 years of

operation. 2-ER-148. According to PG&E, [a] removal time of approximately

21.9 EFPY for Capsule B satisfies NUREG-1801. 2-ER-150.

In 2008, the Staff granted PG&Es request to extend the schedule for

withdrawing Capsule B. 1-ER-029. But the Staff neither acknowledged that its

decision constituted a license amendment, nor gave public notice of a hearing

opportunity as required by § 189a of the Atomic Energy Act, 42 U.S.C. §

2239(a)(1)(A). To the contrary, the Staffs decision did not even mention that its

2006 license amendment had been based upon PG&E carrying out an ASTM E

185-82-compliant four-capsule reactor surveillance program that included Capsule

B. Instead, the Staff erroneously stated that the withdrawal and testing of Capsule 20

V [in 2003] fulfilled the third and final recommendation of ASTM E 185-70 for the current... Unit 1 operating license. 1-ER-033. 11

2. 2010 Extension Decision: from 2010 to 2012

That same year, in 2010, PG&E applied to the NRC for a second extension of

the schedule for removing Capsule B from the Unit 1 pressure vessel, now seeking

to change the date from 2010 to 2012. 2-ER-140. This time, PG&E asserted the

extension was necessary because it had not been able to remove Capsule B in

2010. Id.

Once again, PG&E did not seek a license amendment, nor did it mention the

NRCs 2006 license amendment decision. Instead, it asserted that it had

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. 2-ER-141. Further, PG&E repeated the Staffs misrepresentation of its

1995-approved surveillance program as a three-capsule program asserting that the

withdrawal of Capsule V in 2003 had fulfilled the third and final recommendation

of ASTM E 185-70 for the current DCPP operating license. Id. Instead, PG&E

proposed to withdraw Capsule B at a fluence of approximately 60 EFPY for the

11 In 2009, following the First Extension Decision, PG&E applied for renewal of its operating license. 75 Fed. Reg. 3,493 (Jan. 21, 2010). See also Section G below.

21

reactor pressure vessel, to provide fluence data for the period of extended

operation for license renewal. Id.

The NRC approved the requested schedule change, again without providing

public notice or an opportunity to request a hearing, or providing a reason for why

it was abandoning the alteration of the vessel surveillance schedule on which it had

conditioned the 2006 license amendment. 1-ER-023 - 028. And instead of

acknowledging that in 2006 it had approved a four-capsule ASTM E 185 compliant surveillance program culminating with the withdrawal of Capsule B in

2009, the Staff again mischaracterized the Unit 1 surveillance program as a three-

capsule program for which the third and already-removed Capsule V - not the

fourth and still-remaining Capsule B - was the final capsule. 1-ER-026. The

Staff also incorrectly asserted that Capsule B did not currently form a part of the

licensees surveillance program. Id. Instead, according to the Staff, Capsule B

would become part of a future surveillance schedule for the license renewal period. Id.12

12 The excision of Capsule B from the current surveillance program was further emphasized in the following paragraph:

The surveillance capsule withdrawal plan spanning the initial license period has already been completed and, as such, forms no part of this evaluation. The DCPP LRA and the associated withdrawal schedule have not yet been approved; however, the NRC staff believes that the proactive consideration of Surveillance Capsule B for the period of extended operation adds to the consideration of this request and addresses it below.

22

Finally, in an apparent effort to obfuscate the obvious violation of ASTM E

185-82, the Staff strung together a partially false statement with a vague and

misleading characterization of applicable ASTM standards:

[T]he evaluation criteria of ASTM E-70 do not apply to Surveillance Capsule B, since it does not currently form a part of the licensees surveillance program.

The licensees [reactor vessel] material surveillance program conforms to ASTM E 185.

1-ER-026. The Staff was correct in stating that the evaluation criteria of ASTM E-

70 do not apply to Surveillance Capsule B - but incorrect about the reason. The

reason for not applying ASTM E-70 to Capsule B was that the 2006 License

Amendment upgraded the applicable ASTM E standard to ASTM E-82. And the

statement that PG&Es surveillance program conforms to ASTM E 185 is vague

and misleading, because it is not specific about what edition of ASTM E 185 is

applicable to PG&Es surveillance program. The Staff failed to acknowledge that

in granting the 2006 License Amendment, it explicitly decided that ASTM E-82

constituted the applicable industry standard for PG&Es surveillance program.

3. 2012 Extension Decision: from 2012 to 2022

In 2011, PG&E asked the NRC for a third extension of the schedule for

removing Capsule B from the Unit 1 pressure vessel. 2-ER-134. This time, PG&E

sought a ten-year extension, from 2012 to 2022. The new date would bring

1-ER-026. As discussed in Sections F.3 and F.4 below, this same paragraph would appear as boilerplate language in the two additional extension decisions to follow.

23

withdrawal of Capsule B to within two years of the expiration of the Unit 1

operating license in 2024.

The requested extension did not relate at all to the NRC-approved surveillance

program for Diablo Canyon Unit 1. Rather, it was proposed to support data

acquisition for an industry-wide research program related to reactor license

renewal. Id. For these research purposes, PG&E sought to delay removing Capsule

B until it had accumulated approximately twice the 60-year fluence that would

have been achieved by removing the capsule in 2012 - and which would take

another ten years. Id.

PG&Es application did not request a license amendment or even mention the

2006 license amendment decision. Instead, PG&E again asserted that the

withdrawal of Capsule V in 2003 had fulfilled the third and final recommendation

of ASTM E 185-70 for the current Unit 1 operating license. Id.

In 2012, the NRC Staff approved the schedule change. 1-ER-017. The Staff did

not provide public notice or offer a hearing on this additional amendment to

PG&Es amended operating license, nor did it mention the four-capsule

surveillance program upon which the Staff had conditioned the 2006 license

amendment. Instead, using the now-boilerplate language it had employed in the

previous extension decision, the Staff asserted that the surveillance program for the

initial license term had already been fulfilled and therefore formed no part of

24

this evaluation. 1-ER-020. See also note 12 above.

4. 2023 Extension Decision: from 2022 to 2023 or 2025 or indefinitely delayed

In 2023, PG&E requested a fourth extension for withdrawal of Capsule B, from

2022 until either late 2023 or sometime in 2025 - either the very eve of the

extended license expiration or after expiration. 2-ER-125. Again, PG&E did not

seek a license amendment or even mention the 2006 license amendment decision.

Instead, PG&E characterized Capsule B as a standby capsule - a capsule with no

scheduled withdrawal date at all. 2-ER-132.

The NRC Staff approved the extension. 1-ER-009. Once again, the Staff

offered neither a rationale for abandoning its 2006 license amendment decision,

nor public notice of its decision or an opportunity to request a hearing. The Staff

agreed with PG&E that Capsule B was a standby capsule for which no

withdrawal schedule existed and approved a new scheduled withdrawal date of

2023 or 2025. 1-ER-012. The Staffs approval letter also included the same

boilerplate language that had appeared in the previous two extension decisions,

stating that the surveillance program for the initial license period had been

completed and thus formed no part of the Staffs evaluation. 1-ER-020. See

also note 12 above.

The NRC Staff also left open the possibility that the schedule for removing

Capsule B will be extended yet again, stating that the Staff does not make any 25

conclusion regarding the future use of the subject capsule in any potential future

licensing applications or license periods. 1-ER-009.

G. Subsequent Treatment of PG&Es 2009 License Application:

Submission, Withdrawal, and Re-Submission

PG&E applied for renewal of its operating license in 2009, 75 Fed. Reg. 3,493,

but withdrew the application in 2018. 83 Fed. Reg. 17,688 (Apr. 23, 2018). After

that, until 2022, PG&E made plans to close the reactors on their operating license

expiration dates of 2024 and 2025. Perhaps for this reason, PG&E never sought an

extension of the 2022 deadline for removing Capsule B; nor, to Petitioners

knowledge, did it attempt to remove Capsule B.

In 2022, following passage of state legislation offering a substantial public

subsidy to encourage PG&E to seek license renewal once again, PG&E reversed its

decision to retire the reactors. The company then obtained an exemption from the

NRCs Timely Renewal Rule, allowing it to continue operating the Diablo Canyon

reactors indefinitely pending NRC action on a forthcoming license renewal application. 13

13 As discussed above in note 9, Petitioners appeal of the exemption is pending before this Court in San Luis Obispo Mothers for Peace, et al. v. U.S. Nuclear Reg.

Commn, No.23-852.

26

H. Petitioners Hearing Request and Request for Emergency Action by the Commissioners

On September 14, 2023, after learning of the NRCs decision to extend the

schedule for withdrawal of Capsule B to 2024, 2025, or later, Petitioners submitted

a hearing request to the NRC Commissioners. 2-ER-036. NRCs response to

Petitioners hearing request gives rise to the issues currently before the Court.

Petitioners hearing request included the following contention:

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

2-ER-061. In the same pleading, Petitioners asked the Commissioners to order the

immediate shutdown of Unit 1 for failure to obtain needed embrittlement data for

over twenty years due to the repeated extensions of time for withdrawing and

testing Capsule B. Petitioners asked the Commissioners to keep the reactor in a

shutdown condition pending the removal and testing of Capsule B and a thorough

assessment of the state of embrittlement of the Unit 1 pressure vessel. 2-ER-065 -

2-ER-068.

Both the hearing request and request for emergency enforcement action were

supported by the declaration of Dr. Digby Macdonald, Professor in Residence at

27

the University of California at Berkeley and a highly qualified expert on materials

embrittlement in nuclear reactor pressure vessels. 2-ER-072. Dr. Macdonalds

lengthy and detailed declaration explained the basis for his expert opinion that the

current operation of Diablo Canyon Unit 1 poses an unreasonable risk to public

health and safety due to serious indications of an unacceptable degree of

embrittlement, coupled with a lack of information to establish otherwise. 2-ER-

076. In Dr. Macdonalds professional judgment, PG&E had inappropriately

rejected data from Capsule V indicating that the Unit 1 pressure vessel could

approach an unacceptable state of embrittlement by 2021; and furthermore, that

PG&E had failed to withdraw and test Capsule B and therefore had no additional

reactor-specific data on which it could rely. 2-ER-079 - ER-082. Therefore, Dr.

Macdonald recommended that the reactor should be closed until PG&E obtains

and analyzes additional data regarding its condition. 2-ER-076.

Despite the gravity of the concern and detailed support provided by Petitioners

and Dr. Macdonald for their charges, the Commissioners themselves did not

respond to the hearing request or the request for emergency action. Instead, on

their behalf, the Secretary of the Commission issued a brief three-page decision

denying both requests.

28

The Secretarys grounds for denying the hearing request consisted of two

paragraphs that merely restated the mischaracterizations of the record that had been

repeated time and again in the Staffs four extension decisions:

The Petitioners argue that they are entitled to a hearing because the Extension Approval constitutes a license amendment. But the Extension Approval, by its own terms, does not amend or otherwise affect Diablo Canyons current license.

The Extension Approval does not grant the licensee any greater operating authority, or otherwise alter the original terms of the license, the relevant factors in determining whether a Staff action constitutes a license amendment.

In its evaluation of the schedule revision, the Staff specifically notes that

additional capsules are not needed to satisfy the requirements of Appendix H to 10 CFR Part 50 and ASTM E 185-70 for the current operating license period the licensees compliance with Appendix H to 10 CFR Part 50 and ASTM E 185-70 with respect to the current operating license period for Diablo Canyon, Unit 1 forms no part of the NRC staffs evaluation of the licensees proposed revision to the withdrawal schedule for supplemental surveillance.

The Staff further observes that it does not make any conclusion regarding the future use of the subject capsule in any potential future licensing applications or license periods.

Therefore, the Secretary concluded that [b]ecause the current license for Diablo

Canyon, Unit 1, has not been amended, the Extension Approval does not trigger an

opportunity to request a hearing. 1-ER-005.

With respect to Petitioners request for emergency action, the Secretarys only

response was to state: I refer Petitioners underlying concerns to the Executive

Director for Operations for consideration under 10 C.F.R. § 2.206. 1-ER-005. Thus,

the Secretary referred Petitioners request for emergency action to the very same

29

agency officials who had unlawfully extended the schedule for withdrawing Capsule

B, under the regulatory framework of a petition for discretionary enforcement action

whose outcome would be unreviewable by this court or any other. See, e.g., Safe

Energy Coalition v. U.S. Nuclear Reg. Commn, 866 F.2d 1473, 1479 (D.C. Cir.

1989) (citing Heckler v. Chaney, 470 U.S. 821 (1985) (holding that NRCs denial of

an enforcement petition for revocation or modification of an existing license constitutes an unreviewable exercise of agency discretion). 14

I. Petition for Review

On December 1, 2023, Petitioners submitted a petition for review of the NRCs

decision denying their hearing request without a meaningful explanation.

STANDARD OF REVIEW

Under the APA, agency decisions will be set aside if arbitrary, capricious,

an abuse of discretion, or otherwise not in accordance with law. Public Citizen v.

NRC, 573 F.3d 916, 923 (9th Cir. 2009) (citing 5 U.S.C. § 706(2)(A)). In

reviewing predominantly legal questions rather than factual ones, this Court

applies a standard of reasonableness. Alaska Wilderness Recreation & Tourism

v. Morrison, 67 F.3d 723, 727 (9 th Cir. 1995) ([I]t makes sense to distinguish the

14 The Staff denied the petition for emergency enforcement action on March 8, 2024. The decision post-dates the decisions on review and therefore is not part of the record.

30

strong level of deference we accord an agency in deciding factual or technical

matters from that to be accorded in disputes involving predominantly legal questions.). See also Northcoast Envtl. Ctr. v. Glickman, 136 F.3d 660, 667 (9th

Cir. 1998); Price Rd. Neighbor. Assn v. U.S. Dept. of Transp, 113 F.3d 1505 (9th

Cir. 1997)); San Luis Obispo Mothers for Peace v. NRC, 449 F.3d 1016, 1028 (9th

Cir. 2006).

When reviewing an agencys application of its own regulation, the agency's

interpretation of its regulation must be given controlling weight unless it is plainly

erroneous or inconsistent with the regulation. Alaska Ctr. for the Envt v. United States Forest Serv., 189 F.3d 851, 857 (9th Cir. 1999). But an agencys

interpretation of its own regulation will not be upheld if it lacks the quality

necessary to attract judicial deference. Guard v. United States Nuclear Reg.

Commn, 753 F.2d 1144, 1148-49 (D.C. Cir. (1985). To determine whether agency

action is arbitrary or capricious, a court must consider whether the decision was

based on a consideration of the relevant factors and whether there has been clear

error of judgment. Id. at 859 (citing Marsh v. Oregon Natural Resources Council,

490 U.S. 360, 378 (1989)). Precedent behests this Court to reverse the NRC under

the arbitrary and capricious standard if:

[T]he agency has relied on factors that Congress has not intended it to consider, has entirely failed to consider an important aspect of the problem, or has offered an explanation for that decision that runs counter to the evidence before the agency or is so implausible that it 31

could not be ascribed to a difference in view or the product of agency expertise.

Public Citizen, 573 F.3d at 923.

SUMMARY

OF THE ARGUMENT

Petitioners challenge the NRCs wholesale abandonment of a condition in

PG&Es amended operating license, without providing public notice, explanation,

or any opportunity to challenge the NRCs abdication in a hearing. The license

condition that the NRC abandoned was designed by the NRC to ensure that

extending the operation of Diablo Canyon Unit 1 by three years past its 2021

expiration date would not pose an undue accident risk to the pressure vessel, which

is perhaps the most important single component in the reactor coolant system. 60

Fed. Reg. at 65,457. In that license condition, imposed via a 2006 License

Amendment for Unit 1, the NRC upgraded the industry standard applicable to the

Unit 1 reactor vessel surveillance program from ASTM E 185-70 to ASTM E 185-

82, requiring a four-capsule surveillance program instead of a three-capsule

program. And it required that Capsule B -- the fourth capsule -- must be withdrawn

at 20.7 EPFY or approximately in 2009.

Over the following fifteen-year period, from 2008 to 2023, the NRC issued a

series of Extension Decisions that not only postponed the schedule for removing

Capsule B from the Unit 1 reactor vessel by fifteen years or perhaps indefinitely,

but that attempted -- without explanation or rationale -- to erase the condition 32

imposed by the NRC in the 2006 License Amendment as a predicate for adding

three more years to Unit 1s operating license term, changing the expiration date

from 2021 to 2023.

Instead of acknowledging the 2006 License Amendment or the condition it had

imposed on the Unit 1 operating license, the NRC asserted that Unit 1 was

governed by the outdated ASTM E 185-70 standard and PG&Es previous three-

capsule surveillance program. And the Staff maintained that this three-capsule

program had been fulfilled by the removal of Capsule V in 2003, thus making it

unnecessary to remove Capsule B in the current license term.

The agencys abandonment of this important license condition in its four

Extension Decisions and its refusal to grant Petitioners a hearing on those

Decisions violated the Atomic Energy Act and the APA in three significant ways.

First, the NRC violated the Atomic Energy Act and the APA by refusing to

grant Petitioners a hearing on the 2023 Extension Decision and its predecessor

decisions. These decisions collectively amended conditions in the Unit 1 operating

license to authorize PG&E to operate Unit 1 in a manner that exceeded the limits

imposed by the 2006 License Amendment, thereby triggering the procedural

obligations of the Atomic Energy Act to provide a hearing opportunity. Citizens

Awareness Network, 59 F.3d at 295.

33

Second, the NRC violated the Atomic Energy Acts requirement that changes to

operating licenses must be supported by findings that those changes will not pose

an unreasonable risk to public health and safety. As required by 10 C.F.R. §

50.92(a), the NRCs review of license amendment applications must be guided by

the considerations which govern the issuance of initial licenses. These

considerations include whether the license amendment will protect the health and

safety of the public. 42 U.S.C. § 2133(b).

Finally, the NRCs failure to support or even acknowledge its abandonment of

the 2006 License Amendment violates the APAs requirement for reasonable

decision-making on the primarily legal question under the Atomic Energy Act of

whether it was required to justify a change to a previous safety determination. The

NRCs unannounced and unexplained abandonment of the condition it imposed in

2006 was also arbitrary and capricious because the agency failed to provide any

basis, let alone a reasoned basis, for its change of position. To the extent the Denial

Order did attempt to explain the basis for the NRCs decision, its explanation

[ran] counter to the evidence before the agency and was so implausible that it

could not be ascribed to a difference in view or the product of agency expertise.

Public Citizen, 573 F.3d at 923.

The NRCs violations of the Atomic Energy and the APA have practical,

significant ramifications for public health and safety as well as the credibility of

34

the agency. With respect to public health and safety, PG&E has now operated Unit

1 for more than twenty years without withdrawing any capsule from the Unit 1

pressure vessel. And as discussed in Section D.2 above, PG&E has no data from

the most recently withdrawn capsule - Capsule V in 2003 - that it considers

credible. Further, given that the NRC has now dubbed Capsule B a standby capsule, it appears unlikely that Capsule B will be withdrawn any time soon. 15 In

the meantime, the NRC has prevented Petitioners, if not the general public, from

holding the agency accountable for this regulatory manipulation, by denying

Petitioners hearing request without a word of explanation.

15 As stated in the NRC decision rejecting Petitioners hearing request: The Staff further clarifies that it does not make any conclusion regarding the future use of the subject capsule in any potential future licensing applications or license periods. 1-ER-005.

35

ARGUMENT

I. THE NRC VIOLATED THE ATOMIC ENERGY ACT WHEN IT AMENDED THE OPERATING LICENSE FOR DIABLO CANYON UNIT 1 WITHOUT PROVIDING PUBLIC NOTICE OR THE OPPORTUNITY TO REQUEST A HEARING AND WITHOUT FINDING THAT THE LICENSE AMENDMENTS WERE ADEQUATE TO PROTECT PUBLIC HEALTH AND SAFETY.

A. The NRCs 2006 License Amendment Decision Conditioned the Three-Year Extension of the Unit 1 Operating License on Specific Requirements for PG&Es Reactor Vessel Surveillance Program.

In 2006, following the analytical method set forth in SECY-98-296, the NRC

Staff assessed the impact of the requested license extension on PG&Es

surveillance program for the Unit 1 pressure vessel. 2-ER-163, 2-ER-165. As a

result of that evaluation, in order to provide[] adequate protection to the health

and safety of the public for the enlarged term, the Staff required that: (1) PG&Es

program be upgraded from ASTM E185-70 to ASTM E185-82; (2) consistent

with that upgrade, PG&Es program would be increased from three to four

capsules, of which Capsule B was the last; and (3) also consistent with that

upgrade, the projected withdrawal for Capsule B was amended to 20.7 EFPY

(approximately 2009). See Section E.2 above.

By citing these explicit elements of the reactor surveillance program to

justify the three-year extension of the Unit 1 operating license term, the NRC met

the two-pronged test that established them as conditions of the Unit 1 operating

license. First, the NRC relied on the elements of PG&Es reactor vessel 36

surveillance program to support a license amendment that would grant greater

operating authority to PG&E, i.e., the authority to operate Unit 1 beyond 2021 to

2024. In re Three Mile Island Alert, 771 F.2d 720, 729 (3d Cir. 1985). See also

Citizens Awareness Network, 59 F.32d at 295 (emphasis in original) (license

amendment undeniably supplemented [PG&Es] operating authority.).

Second, by establishing specific new surveillance requirements that must be

carried out as a condition precedent to the extended operation permitted by the

2006 license amendment, the NRC altered the original terms of the operating

license. Deukmejian v. NRC, 751 F.2d 1287, 1314 (D.C. Cir. 1984). See also

Union of Concerned Scientists v. NRC, 711 F.2d 370, 382 (D.C. Cir. 1983)

(holding that NRC amended reactor licenses by changing the binding substantive

norms.).

B. The NRC Has Repeatedly Amended the License Condition Imposed on PG&E by the 2006 License Amendment.

In four separate Exemption Decisions issued since 2006 -- in 2008, 2010, 2012,

and 2023 -- the NRC Staff has amended the license condition imposed on PG&E

by the 2006 License Amendment as a safety-based predicate for extending Unit 1s

operating license term by three years past its 2021 expiration date. Without even

acknowledging the license condition imposed in 2006, these Decisions have

effectively discarded it by (1) extending the scheduled date for removal of Capsule

B; (2) declaring that the applicable ASTM E standard was ASTM E 185-70 rather 37

than the updated ASTM E 185-82; (3) declaring that PG&Es surveillance program

was a three-capsule program instead of a four-capsule program, and (4) asserting

that PG&Es surveillance program was completed with the removal of Capsule V in 2002. 16

Fifteen years after the 2006 License Amendment decision, nothing remains of

the license condition. The only part of the License Amendment that has any

recognized effect is that PG&E has continued to operate the Unit 1 reactor for

years past the pre-2006 expiration date of 2021, now unencumbered by the safety

requirements on which that extension was based. Thus, the Exemption Decisions

has supplemented PG&Es operating authority. Citizens Awareness Network,

59 F.3d at 295.

C. The NRC Violated the Atomic Energy Act by Failing to Provide Public Notice or a Hearing Opportunity Each Time It Extended the Schedule for Withdrawal of Capsule B from the Unit 1 Pressure Vessel.

Before amending an operating license, the NRC must comply with the

requirements of Section 189a of the Atomic Energy Act to provide public

notice and the opportunity to request a hearing. 42 U.S.C. § 2239(a). Citizens

Awareness Network, 59 F.32d at 295. The NRC violated this statutory mandate

16 See also discussion above in Sections F.1 through F.4.

38

by failing to provide any public notice of the four Extension Decisions or to

offer the public an opportunity to be heard.

D. The NRC Violated the Atomic Energy Act by Failing to Evaluate Whether Changes to PG&Es License Condition Would Provide Adequate Protection to Public Health and Safety.

By approving changes to PG&Es license as amended by the 2006 License

Amendment without evaluating how those changes would affect public health and

safety, the NRC violated the Atomic Energy Act. As required by 10 C.F.R. §

50.92(a), the NRCs review of license amendment applications must be guided by

the considerations which govern the issuance of initial licenses. These

considerations include whether the license amendment will protect the health and

safety of the public. 42 U.S.C. § 2133(b). The NRC failed even to acknowledge

the existence of the license condition, let alone address how changing it would

affect public health and safety. Therefore, the Decisions are unlawful under the

Act.

39

II. THE NRC VIOLATED THE ATOMIC ENERGY ACT AND THE ADMINISTRATIVE PROCEDURE ACT WHEN IT DENIED PETITIONERS A HEARING ON THE 2023 EXTENSION OF THE SCHEDULE FOR WITHDRAWING CAPSULE B.

Section 189a of the Atomic Energy Act, 42 U.S.C. § 2239(a), requires the

NRC to grant a hearing upon the request of any person whose interest may be

affected by the proceeding. As discussed above in Section II.H, Petitioners

demonstrated their interest in the proceeding by submitting standing declarations

and by setting forth, in a specific and well-supported contention, the facts

demonstrating that the NRC had amended PG&Es operating license by granting

the 2023 extension and the three extensions preceding it. Nevertheless, the

Secretary summarily denied Petitioners hearing request. 1-ER-005.

The Secretarys Denial Order violated Section 189a of the Act by utterly failing

to engage on, or even entertain Petitioners claims that after granting the 2006

License Amendment, the NRCs decisions granting multiple extensions of the time

for removing Capsule B pivot[ed] sharply away from the rationale for the 2006

License Amendment, to the point where the Staff now considers withdrawal of

Capsule B a discretionary task that PG&E may undertake on its own schedule. 2-

ER-056 ER-059. The Secretarys decision is unreasonable because it addresses

the legal question of whether the Staff impermissibly changed or discarded a

license condition by simply parroting the demonstrably unacceptable language on

which Petitioners seek a hearing. Alaska Wilderness Recreation & Tourism, 67

40

F.3d at 727. It is also arbitrary and capricious because it fails to consider whether

the decision was based on a consideration of the relevant factors and whether there

has been clear error of judgment. Alaska Ctr. for the Envt v. United States Forest

Serv., 189 F.3d at 859.

III. THE NRCS ABANDONMENT, WITHOUT A REASONED EXPLANATION, OF THE SURVEILLANCE PROGRAM IT IMPOSED ON PG&E AS A CONDITION OF EXTENDING THE TERM OF PG&ES LICENSE WAS UNREASONABLE AND ARBITRARY AND CAPRICIOUS.

A. The Extension Decisions Were Unreasonable.

In all four Extension Decisions at issue, the NRC abandoned a duly-

established license condition that it had imposed in 2006 in compliance with the

substantive and procedural requirements of the Atomic Energy Act. By failing

to even acknowledge the existence or applicability of the 2006 License

Amendment, the NRC unlawfully abdicated its statutory duties under the

Atomic Energy Act, which require support of decisions with a safety analysis

and to provide public notice and a hearing opportunity.

Under Ninth Circuit jurisprudence, agency decisions that are primarily

legal in nature are entitled less deference than those that are factual in

nature. Cal. Ex. rel. Lockyer v. USDA, 575 F.3d 999, 1011 (9th Cir. 2009)

(citing Northcoast Environmental Center v. Glickman, 136 F.3d at 667 and

Alaska Wilderness Recreation & Tourism, 67 F.3d at). The NRCs unexplained

41

abdication of the agencys statutory duty must be rejected because it is

unreasonable.

B. The Extension Decisions Were Arbitrary and Capricious.

The Administrative Procedure Act invalidates agency actions that are

arbitrary and capricious. 5 U.S.C. § 706(2)(A). A long line of Supreme Court

and lower federal court cases have concluded that, while an agency can change

its legal position, an agency must provide a reasoned explanation for any

failure to adhere to its own precedents." Hatch v. FERC, 654 F.2d 825, 834

(D.C. Cir. 1981). See also Encino Motorcars, LLC v. Navarro, 579 U.S. 211, 22

(2017) (citing FCC v. Fox Television Stations, Inc., 556 U.S. 502, (2009)).

(agency must present a reasoned explanation for disregarding facts and

circumstances that were engendered by [a] prior policy.); Motor Vehicle

Manufacturers Association v. State Farm Mutual Automobile Insurance Co.,

463 U.S. 29, 42 (1983) (Accordingly, an agency changing its course by

rescinding a rule is obligated to supply a reasoned analysis for the change

beyond that which may be required when an agency does not act in the first

instance.)). See also Honeywell Intl v. U.S. Nuclear Reg. Commn, 628 F.23d

568, 578 (D.C. Cir. 2010) and Guard v. United States Nuclear Regulatory

Com., 753 F.2d 1144 (D.C. Cir. 1985) (NRC decisions held arbitrary and

capricious for ignoring previous NRC positions without reasoned explanation).

42

Here, the NRC's action falls plainly afoul of this long line of cases. Without

so much as acknowledgement, let alone a reasoned explanation, the agency

arbitrarily and abruptly reversed position - on multiple occasions. In its 2006

license amendment, the agency conditioned a three-year extension of Unit 1s

license on PG&E implementing the ASTM E 185-82 four-capsule surveillance

program, including removal of Capsule B in approximately 2009; then the

agency granted four separate extensions of the deadline for removing Capsule B

over a period of 15 years, each one based on the NRCs assertion that the three-

capsule surveillance program that had been supplanted in 2006 was the

applicable program, and that it had been completed. The NRC offered no reason

for its change of position: indeed, it did not even acknowledge that it had made

any change.

CONCLUSION AND REQUEST FOR RELIEF

For the foregoing reasons, Petitioners respectfully request the Court to

declare that the 2023 Extension Decision and the three preceding Extension

Decisions constituted unlawfully issued amendments or revocations of the license

condition imposed by the NRC in 2006 and reverse and vacate them. In addition,

Petitioners request the Court to order the Commission to grant a hearing on

whether it should have issued the 2023 Extension Decision or any of the previous

Extension Decisions leading up to it. Finally, because these license amendments

43

have cumulatively allowed PG&E to operate Unit 1 in violation of the license

condition on which extended operation past 2021 is predicated, the Court should

order the Commission to expedite the hearing and any other response to the

Courts decision that may be required.

Respectfully submitted,

/s/Diane Curran 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

/s/Richard E. Aryes 2923 Foxhall Road, N.W.

Washington, D.C. 20016 202-744-6930 ayres@ayreslawgroup.com Counsel to Petitioner Friends of the Earth

March 20, 2024 Corrected March 25, 2024

44

PETITIONERS CERTIFICATE OF COMPLIANCE Pursuant to Fed. R. App. P. 32(a)(7)(C), the undersigned hereby certifies that this brief complies with the type-volume limitations of Fed. R. App. P.

32(a)(7)(B)(i) and Rule 29(a)(f).

1. Exclusive of the exempted portion of the brief provided in Fed. R. App. P.

32(a)(7)(B), the brief contains 10,041 words.

2. The brief has been prepared in proportionally spaced typeface using Microsoft Word in 14-point Times New Roman font. As permitted by Fed. R. App.

P. 32(a)(7)(B), the undersigned has relied upon the word count feature of this word processing system in preparing this certificate.

Respectfully submitted,

/s/ Diane Curran Diane Curran

March 25, 2024 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 ADDENDUM OF PERTINENT STATUTES AND REGULATIONS

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

March 20, 2024

Table of Contents Statutes Atomic Energy Act 42 U.S.C. § 2133A-001 42 U.S.C. § 2131A-003 42 U.S.C. § 2132.A-006 42 U.S.C. § 2232.A-009 42 U.S.C. § 2237.A-011 42 U.S.C. § 2239.A-012 Hobbs Act 28 U.S.C. § 2342..A-014 28 U.S.C. § 2344..A-015 Administrative Procedure Act 5 U.S.C. § 702..A-016 Regulations 10 C.F.R. § 2.206..A-017 10 C.F.R. § 50.61..A-018 10 C.F.R. Part 50, Appendix H..A-026 10 C.F.R. § 50.92...A-029 42 USCS § 2133

Current through Public Law 118-40, approved March 1, 2024.

United States Code Service > TITLE 42. THE PUBLIC HEALTH AND WELFARE (Chs. 1 164) >

CHAPTER 23. DEVELOPMENT AND CONTROL OF ATOMIC ENERGY (§§ 2011 2297h-13) >

ATOMIC ENERGY (§§ 2011 2296b-7) > ATOMIC ENERGY LICENSES (§§ 2131 2142)

§ 2133. Commercial licenses

(a) Conditions. The Commission is authorized to issue licenses to persons applying therefor to transfer or receive in interstate commerce, manufacture, produce, transfer, acquire, possess, use, import, or export under the terms of an agreement for cooperation arranged pursuant to section 123 [42 USCS § 2153], utilization or production facilities for industrial or commercial purposes. Such licenses shall be issued in accordance with the provisions of chapter 16 [42 USCS §§ 2231 et seq.] and subject to such conditions as the Commission may by rule or regulation establish to effectuate the purposes and provisions of this Act [42 USCS §§ 2011 et seq.].

(b)Nonexclusive basis. The Commission shall issue such licenses on a nonexclusive basis to persons applying therefor (1) whose proposed activities will serve a useful purpose proportionate to the quantities of special nuclear material or source material to be utilized; (2) who are equipped to observe and who agree to observe such safety standards to protect health and to minimize danger to life or property as the Commission may by rule establish; and (3) who agree to make available to the Commission such technical information and data concerning activities under such licenses as the Commission may determine necessary to promote the common defense and security and to protect the health and safety of the public. All such information may be used by the Commission only for the purposes of the common defense and security and to protect the health and safety of the public.

(c) License period. Each such license shall be issued for a specified period, as determined by the Commission, depending on the type of activity to be licensed, but not exceeding forty years from the authorization to commence operations, and may be renewed upon the expiration of such period.

(d)Limitations. No license under this section may be given to any person for activities which are not under or within the jurisdiction of the United States, except for the export of production or utilization facilities under terms of an agreement for cooperation arranged pursuant to section 123 [ 42 USCS § 2153], or except under the provisions of section 109 [42 USCS § 2139]. No license may be issued to an alien or any [any] corporation or other entity if the Commission knows or has reason to believe it is owned, controlled, or dominated by an alien, a foreign corporation, or a foreign government. In any event, no license may be issued to any person within the United States if, in the opinion of the Commission, the issuance of a license to such person would be inimical to the common defense and security or to the health and safety of the public.

(e) [Not enacted]

(f) Accident notification condition; license revocation; license amendment to include condition.

Each license issued for a utilization facility under this section or section 104(b) [42 USCS § 2134(b)] shall require as a condition thereof that in case of any accident which could result in an unplanned release of quantities of fission products in excess of allowable limits for normal operation established by the Commission, the licensee shall immediately so notify the Commission. Violation of the condition prescribed by this subsection may, in the Commissions discretion, constitute grounds for license revocation. In accordance with section 187 of this Act [42 USCS § 2237], the Commission shall promptly amend each license for a utilization facility issued under this section or section 104(b) [42 USCS § 2134(b)] which is in Page 4 of 20

§ 2133. Commercial licenses

effect on the date of enactment of this subsection [enacted June 30, 1980] to include the provisions required under this subsection.

History

HISTORY:

Aug. 1, 1946, ch 724, Title I, Ch. 10, § 103, as added Aug. 30, 1954, ch 1073, § 1, 68 Stat. 936; Aug. 6, 1956, ch 1015, §§ 12, 13, 70 Stat. 1071; Dec. 19, 1970, P. L.91-560, § 4, 84 Stat. 1472; June 30, 1980, P. L.96-295, Title II, § 201, 94 Stat. 786; Oct. 24, 1992, P. L. 102-486, Title IX, § 902(a)(8), 106 Stat. 2944; Aug. 8, 2005, P. L. 109-58, Title VI, Subtitle B, § 621, 119 Stat. 782.

Annotations

Notes

HISTORY; ANCILLARY LAWS AND DIRECTIVES

References in text:

Explanatory notes:

Amendment Notes

1956.

1970.

1980.

2005.

Other provisions:

References in text:

The Commission, referred to in this section, was the Atomic Energy Commission, which was abolished by Act Oct.

11, 1974, P.L.93-438, Title I, § 104(a), 88 Stat. 1237, and its functions and personnel transferred (see 42 USCS § 2014 note).

Explanatory notes:

The word any has been enclosed in brackets in subsec. (d) to indicate the probable intent of Congress to delete it.

Act Oct. 24, 1992, P. L. 102-486, Title IX, § 902(a)(8), 106 Stat. 2944, amended the Atomic Energy Act of 1954, which appears generally as 42 USCS §§ 2011 et seq., by inserting TITLE I-ATOMIC ENERGY before the Chapter 1 heading.

Amendment Notes 1956.

42 USCS § 2131

Current through Public Law 118-40, approved March 1, 2024.

United States Code Service > TITLE 42. THE PUBLIC HEALTH AND WELFARE (Chs. 1 164) >

CHAPTER 23. DEVELOPMENT AND CONTROL OF ATOMIC ENERGY (§§ 2011 2297h-13) >

ATOMIC ENERGY (§§ 2011 2296b-7) > ATOMIC ENERGY LICENSES (§§ 2131 2142)

§ 2131. License required

It shall be unlawful, except as provided in section 91 [42 USCS § 2121], for any person within the United States to transfer or receive in interstate commerce, manufacture, produce, transfer, acquire, possess, use, import, or export any utilization or production facility except under and in accordance with a license issued by the Commission pursuant to section 103 or 104 [42 USCS § 2133 or 2134].

History

HISTORY:

Aug. 1, 1946, ch 724, Title I, Ch. 10, § 101, as added Aug. 30, 1954, ch 1073, § 1, 68 Stat. 936; Aug. 6, 1956, ch 1015, § 11, 70 Stat. 1071; Oct. 24, 1992, P. L. 102-486, Title IX, § 902(a)(8), 106 Stat. 2944.

Annotations

Notes

HISTORY; ANCILLARY LAWS AND DIRECTIVES

References in text:

Explanatory notes:

Amendment Notes

1956.

References in text:

The Commission, referred to in this section, was the Atomic Energy Commission, which was abolished by Act Oct.

11, 1974, P.L.93-438, Title I, § 104(a), 88 Stat. 1237, and its functions and personnel transferred (see 42 USCS § 2014 note).

Explanatory notes:

Page 13 of 20

§ 2131. License required

Act Oct. 24, 1992, P. L. 102-486, Title IX, § 902(a)(8), 106 Stat. 2944, amended the Atomic Energy Act of 1954, which appears generally as 42 USCS §§ 2011 et seq., by inserting TITLE I-ATOMIC ENERGY before the Chapter 1 heading.

Amendment Notes

1956.

Act Aug. 6, 1956, inserted use,.

NOTES TO DECISIONS

1.Generally

2.Purpose

3.Construction

1.Generally

NRC is empowered by 42 USCS § 2131 to regulate nuclear power plants off-site transmission lines; in exercise of that power, it must pursue objectives of 42 USCS §§ 2011 et seq. and NEPA (42 USCS §§ 4321 et seq.)

simultaneously. Detroit Edison Co. v. United States Nuclear Regulatory Com., 630 F.2d 450, 14 Env't Rep. Cas. (BNA) 2090, 10 Envtl. L. Rep. 20879, 1980 U.S. App. LEXIS 14312 (6th Cir. 1980).

2.Purpose

Purpose of license requirement under 42 USCS § 2131 is to provide adequate examination by Nuclear Regulatory Commission of such factors as safety features, design requirements, and competent supervision and operation of nuclear facilities. Drake v. Detroit Edison Co., 443 F. Supp. 833, 1978 U.S. Dist. LEXIS 20040 (W.D. Mich. 1978).

3.Construction

Failure to submit plans and schedules for making plant modifications required by NRC regulations relating to fire safety, and failure to implement modifications in accordance with prescribed timetables did not constitute continuing offenses for limitation purposes in action charging violations of criminal statutes; statutes themselves contained no explicit language compelling conclusion that crimes described or referenced therein were continuing offenses for limitation purposes; while Atomic Energy Act (42 USCS §§ 2011 et seq.) reflected congressional concern about safety of nuclear power facilities, that fact alone was insufficient to justify construing violations of Act as continuing offenses; regulations which specified date for performance provided substantive basis for charged offenses and defined nature of offenses and there was nothing inherent about failing to submit accurate reports or make plant modifications which demanded that actions or omissions be construed as continuing crimes. United States v. Del Percio, 870 F.2d 1090, 1989 U.S. App. LEXIS 3409 (6th Cir. 1989), reh'g denied, 1989 U.S. App. LEXIS 12378 (6th Cir. July 21, 1989).

Holder of radioactive material storage license was not deprived of value of license without due process when DOE commenced providing free storage of unenriched uranium, in connection with feed usage agreements, for enrichment customers; provision of storage services was not inconsistent with statutory criteria (42 USCS § 2201)

Page 14 of 20

§ 2131. License required

governing uranium enrichment contract. Nuclear Transp. & Storage v. United States, 890 F.2d 1348, 1989 U.S. App.

LEXIS 15832 (6th Cir. 1989), cert. denied, 494 U.S. 1079, 110 S. Ct. 1807, 108 L. Ed. 2d 938, 1990 U.S. LEXIS 1936 (1990).

Neither a purchaser nor the Tennessee Valley Authority (TVA) was entitled to summary judgment in purchasers action for specific performance of parties contract regarding sale and purchase of nuclear power plant site because transfer of site without approval of the Nuclear Regulatory Commission would have been unlawful under Section 101 of the Atomic Energy Act of 1954, and there were genuine disputes of material fact as to whether TVA fulfilled its duties under the parties agreement, and as to whether the purchaser was ready, willing, and able to close on the closing date. Nuclear Dev., LLC v. TVA, 532 F. Supp. 3d 1154, 2021 U.S. Dist. LEXIS 68114 (N.D. Ala. 2021).

42 USCS § 2132

Current through Public Law 118-40, approved March 1, 2024.

United States Code Service > TITLE 42. THE PUBLIC HEALTH AND WELFARE (Chs. 1 164) >

CHAPTER 23. DEVELOPMENT AND CONTROL OF ATOMIC ENERGY (§§ 2011 2297h-13) >

ATOMIC ENERGY (§§ 2011 2296b-7) > ATOMIC ENERGY LICENSES (§§ 2131 2142)

§ 2132. Utilization and production facilities for industrial or commercial purposes

(a) Issuance of licenses. Except as provided in subsections (b) and (c), or otherwise specifically authorized by law, any license hereafter issued for a utilization or production facility for industrial or commercial purposes shall be issued pursuant to section 103 [42 USCS § 2133].

(c) Facilities constructed or operated under 42 USCS § 2134(b). Any license hereafter issued for a utilization or production facility for industrial or commercial purposes, the construction or operation of which was licensed pursuant to subsection 104(b) [ 42 USCS § 2134(b)] prior to enactment into law of this subsection, shall be issued under subsection 104(b) [42 USCS § 2134(b)].

(a) Cooperative Power Reactor Demonstration facilities. Any license for a utilization or production facility for industrial or commercial purposes constructed or operated under an arrangement with the Commission entered into under the Cooperative Power Reactor Demonstration Program shall, except as otherwise specifically required by applicable law, be issued under subsection 104(b) [42 USCS § 2134(b)].

History

HISTORY:

Aug. 1, 1946, ch 724, Title I, Ch. 10, § 102, as added Aug. 30, 1954, ch 1073, § 1, 68 Stat. 936; Dec. 19, 1970, P. L.91-560, § 3, 84 Stat. 1472; Oct. 24, 1992, P. L. 102-486, Title IX, § 902(a)(8), 106 Stat. 2944.

Annotations

Notes

HISTORY; ANCILLARY LAWS AND DIRECTIVES

References in text:

Explanatory notes:

Amendment Notes

1970.

Page 18 of 20

§ 2132. Utilization and production facilities for industrial or commercial purposes

The Commission, referred to in this section, was the Atomic Energy Commission, which was abolished by Act Oct.

11, 1974, P.L.93-438, Title I, § 104(a), 88 Stat. 1237, and its functions and personnel transferred (see 42 USCS § 2014 note).

Explanatory notes:

Act Oct. 24, 1992, P. L. 102-486, Title IX, § 902(a)(8), 106 Stat. 2944, amended the Atomic Energy Act of 1954, which appears generally as 42 USCS §§ 2011 et seq., by inserting TITLE I-ATOMIC ENERGY before the Chapter 1 heading.

Amendment Notes

1970.

Act Dec. 19, 1970, substituted the text of this section for text which read: Whenever the Commission has made a finding in writing that any type of utilization or production facility has been sufficiently developed to be of practical value for industrial or commercial purposes, the Commission may thereafter issue licenses for such type of facility pursuant to section 103..

NOTES TO DECISIONS

1.Generally

2.Emergency preparedness plan

1.Generally

NRC regulations interpreting antitrust review (42 USCS § 2135(c)), which provide that such review is to be conducted only when new licenses are issued, not when licenses are renewed (42 USCS §§ 2133, 2134(b)), are valid interpretation of statute. American Pub. Power Ass'n v. United States Nuclear Regulatory Comm'n, 990 F.2d 1309, 301 U.S. App. D.C. 39, 23 Envtl. L. Rep. 20824, 1993-1 Trade Cas. (CCH) ¶ 70193, 1993 U.S. App. LEXIS 7605 (D.C.

Cir. 1993).

2.Emergency preparedness plan

Scheduled exercise of utilitys emergency preparedness plan for nuclear power station is necessary to fulfill Nuclear Regulatory Commissions (NRC) responsibility under 42 USCS §§ 2132, 2201, and thus will not be canceled, since it will (1) provide information as to whether state and local governments lack of emergency planning co-operation results in significant defects under NRCs emergency planning standards, 10 CFR § 50.47, and (2) test utilitys ability to accommodate ad hoc governmental participation in actual emergency. In re Long Island Lighting Co., 24 N.R.C.

36, 1986 NRC LEXIS 131 (N.R.C. Jan. 30, 1986).

42 USCS, Ch. 23, Atomic Energy, Atomic Energy Licenses

Current through Public Law 118-40, approved March 1, 2024.

United States Code Service > TITLE 42. THE PUBLIC HEALTH AND WELFARE (Chs. 1 164) >

CHAPTER 23. DEVELOPMENT AND CONTROL OF ATOMIC ENERGY (§§ 2011 2297h-13) >

ATOMIC ENERGY (§§ 2011 2296b-7) > ATOMIC ENERGY LICENSES (§§ 2131 2142)

ATOMIC ENERGY LICENSES

Annotations

Notes

HISTORY; ANCILLARY LAWS AND DIRECTIVES

Explanatory notes:

Provisions similar to those comprising 42 USCS §§ 2131-2140 were contained in Act Aug. 1, 1946, ch 724, § 7, 60 Stat. 764 (Formerly appearing as 42 USCS § 1807), prior to the complete amendment and renumbering of Act Aug. 1, 1946 by Act Aug. 30, 1954, ch 107 3, 68 Stat. 921.

Research References & Practice Aids

Cross

References:

This subchapter is referred to in 42 USCS §§ 2014, 5842.

Hierarchy Notes:

42 USCS, Ch. 23

United States Code Service Copyright

© 2024 All rights reserved.

End of Document 42 USCS § 2232

Current through Public Law 118-40, approved March 1, 2024.

United States Code Service > TITLE 42. THE PUBLIC HEALTH AND WELFARE (Chs. 1 164) >

CHAPTER 23. DEVELOPMENT AND CONTROL OF ATOMIC ENERGY (§§ 2011 2297h-13) >

ATOMIC ENERGY (§§ 2011 2296b-7) > JUDICIAL REVIEW AND ADMINISTRATIVE PROCEDURE (§§ 2231 2243)

§ 2232. License applications

(a) Contents and form. Each application for a license hereunder shall be in writing and shall specifically state such information as the Commission, by rule or regulation, may determine to be necessary to decide such of the technical and financial qualifications of the applicant, the character of the applicant, the citizenship of the applicant, or any other qualifications of the applicant as the Commission may deem appropriate for the license. In connection with applications for licenses to operate production or utilization facilities, the applicant shall state such technical specifications, including information of the amount, kind, and source of special nuclear material required, the place of the use, the specific characteristics of the facility, and such other information as the Commission may, by rule or regulation, deem necessary in order to enable it to find that the utilization or production of special nuclear material will be in accord with the common defense and security and will provide adequate protection to the health and safety of the public.

Such technical specifications shall be a part of any license issued. The Commission may at any time after the filing of the original application, and before the expiration of the license, require further written statements in order to enable the Commission to determine whether the application should be granted or denied or whether a license should be modified or revoked. All applications and statements shall be signed by the applicant or licensee. Applications for, and statements made in connection with, licenses under sections 103 and 104 [42 USCS §§ 2133 and 2134] shall be made under oath or affirmation. The Commission may require any other applications or statements to be made under oath or affirmation.

(b) Review of applications by Advisory Committees on Reactor Safeguards; report. The Advisory Committee on Reactor Safeguards shall review each application under section 103 or section 104(b) [42 USCS § 2133 or 2134(b)] for a construction permit or an operating license for a facility, any application under section 104(c) [42 USCS § 2134(c)] for a construction permit or an operating license for a testing facility, any application under section 104(a) or (c) [42 USCS § 2134(a) or (c)] specifically referred to it by the Commission, and any application for an amendment to a construction permit or an amendment to an operating license under section 103 or 104(a), (b), or (c) [42 USCS § 2133 or 2134(a), (b), or (c)]

specifically referred to it by the Commission, and shall submit a report thereon which shall be made part of the record of the application and available to the public except to the extent that security classification prevents disclosure.

(c) Commercial power; publication. The Commission shall not issue any license under section 103 [42 USCS § 2133] for a utilization or production facility for the generation of commercial power until it has given notice in writing to such regulatory agency as may have jurisdiction over the rates and services incident to the proposed activity; until it has published notice of the application in such trade or news publications as the Commission deems appropriate to give reasonable notice to municipalities, private utilities, public bodies, and cooperatives which might have a potential interest in such utilization or production facility; and until it has published notice of such application once each week for four consecutive weeks in the Federal Register, and until four weeks after the last notice.

(d) Preferred consideration. The Commission, in issuing any license for a utilization or production facility for the generation of commercial power under section 103 [42 USCS § 2133], shall give preferred Page 2 of 8

§ 2232. License applications

consideration to applications for such facilities which will be located in high cost power areas in the United States if there are conflicting applications for a limited opportunity for such license. Where such conflicting applications resulting from limited opportunity for such license include those submitted by public or cooperative bodies such applications shall be given preferred consideration.

History

HISTORY:

Aug. 1, 1946, ch 724, Title I, Ch. 16, § 182, as added Aug. 30, 1954, ch 1073, § 1, 68 Stat. 953; Aug. 6, 1956, ch 1015, § 5, 70 Stat. 1069; Sept. 2, 1957, P. L.85-256, § 6, 71 Stat. 579 ; Aug. 29, 1962, P. L.87-615, § 3, 76 Stat.

409; Dec. 19, 1970, P. L.91-560, § 9, 84 Stat. 1474; Oct. 24, 1992, P. L. 102-486, Title IX, § 902(a)(8), 106 Stat.

2944.

Annotations

Notes

HISTORY; ANCILLARY LAWS AND DIRECTIVES

References in text:

Explanatory notes:

Amendment Notes

1956.

1957.

1962.

1970.

Other provisions:

References in text:

The Commission, referred to in this section, was the Atomic Energy Commission, which was abolished by Act Oct.

11, 1974, P.L.93-438, Title I, § 104(a), 88 Stat. 1237, and its functions and personnel transferred (see 42 USCS § 2014 note).

Explanatory notes:

Act Oct. 24, 1992, P. L. 102-486, Title IX, § 902(a)(8), 106 Stat. 2944, amended the Atomic Energy Act of 1954, which appears generally as 42 USCS §§ 2011 et seq., by inserting TITLE I-ATOMIC ENERGY before the Chapter 1 heading.

Amendment Notes 42 USCS § 2237

Current through Public Law 118-40, approved March 1, 2024.

United States Code Service > TITLE 42. THE PUBLIC HEALTH AND WELFARE (Chs. 1 164) >

CHAPTER 23. DEVELOPMENT AND CONTROL OF ATOMIC ENERGY (§§ 2011 2297h-13) >

ATOMIC ENERGY (§§ 2011 2296b-7) > JUDICIAL REVIEW AND ADMINISTRATIVE PROCEDURE (§§ 2231 2243)

§ 2237. Modification of license

The terms and conditions of all licenses shall be subject to amendment, revision, or modification, by reason of amendments of this Act [42 USCS §§ 2011 et seq.] or by reason of rules and regulations issued in accordance with the terms of this Act [42 USCS §§ 2011 et seq.].

History

HISTORY:

Aug. 1, 1946, ch 724, Title I, Ch. 16, § 187, as added Aug. 30, 1954, ch 1073, § 1, 68 Stat. 955; Oct. 24, 1992, P. L.

102-486, Title IX, § 902(a)(8), 106 Stat. 2944.

Annotations

Notes

HISTORY; ANCILLARY LAWS AND DIRECTIVES

Explanatory notes:

Act Oct. 24, 1992, P. L. 102-486, Title IX, § 902(a)(8), 106 Stat. 2944, amended the Atomic Energy Act of 1954, which appears generally as 42 USCS §§ 2011 et seq., by inserting TITLE I-ATOMIC ENERGY before the Chapter 1 heading.

Research References & Practice Aids

Cross

References:

This section is referred to in 42 USCS § 2133.

Code of Federal Regulations:

Nuclear Regulatory CommissionDomestic licensing of source material, 10 CFR 40.1 et seq.

Nuclear Regulatory CommissionDomestic licensing of production and utilization facilities, 10 CFR 50.1 et seq.

42 USCS § 2239

Current through Public Law 118-40, approved March 1, 2024.

United States Code Service > TITLE 42. THE PUBLIC HEALTH AND WELFARE (Chs. 1 164) >

CHAPTER 23. DEVELOPMENT AND CONTROL OF ATOMIC ENERGY (§§ 2011 2297h-13) >

ATOMIC ENERGY (§§ 2011 2296b-7) > JUDICIAL REVIEW AND ADMINISTRATIVE PROCEDURE (§§ 2231 2243)

§ 2239. Hearings and judicial review

(a)

(1)

(A) In any proceeding under this Act [42 USCS §§ 2011 et seq.], for the granting, suspending, revoking, or amending of any license or construction permit, or application to transfer control, and in any proceeding for the issuance or modification of rules and regulations dealing with the activities of licensees, and in any proceeding for the payment of compensation, an award or royalties under sections [section] 153, 157, 186(c), or 188 [42 USCS § 2183, 2187, 2236(c), or 2238], the Commission shall grant a hearing upon the request of any person whose interest may be affected by the proceeding, and shall admit any such person as a party to such proceeding. The Commission shall hold a hearing after thirty days notice and publication once in the Federal Register, on each application under section 103 or 104(b) [42 USCS § 2133 or 2134(b)] for a construction permit for a facility, and on any application under section 104(c) [42 USCS § 2134(c)]

for a construction permit for a testing facility. In cases where such a construction permit has been issued following the holding of such a hearing, the Commission may, in the absence of a request therefor by any person whose interest may be affected, issue an operating license or an amendment to a construction permit or an amendment to an operating license without a hearing, but upon thirty days notice and publication once in the Federal Register of its intent to do so. The Commission may dispense with such thirty days notice and publication with respect to any application for an amendment to a construction permit or an amendment to an operating license upon a determination by the Commission that the amendment involves no significant hazards consideration.

(B)

(i) Not less than 180 days before the date scheduled for initial loading of fuel into a plant by a licensee that has been issued a combined construction permit and operating license under section 185(b) [42 USCS § 2235(b)], the Commission shall publish in the Federal Register notice of intended operation. That notice shall provide that any person whose interest may be affected by operation of the plant, may within 60 days request the Commission to hold a hearing on whether the facility as constructed complies, or on completion will comply, with the acceptance criteria of the license.

(ii)A request for hearing under clause (i) shall show, prima facie, that one or more of the acceptance criteria in the combined license have not been, or will not be met, and the specific operational consequences of nonconformance that would be contrary to providing reasonable assurance of adequate protection of the public health and safety.

(iii)After receiving a request for a hearing under clause (i), the Commission expeditiously shall either deny or grant the request. If the request is granted, the Commission shall determine, after considering petitioners prima facie showing and any answers thereto, whether during a Page 2 of 19

§ 2239. Hearings and judicial review

period of interim operation, there will be reasonable assurance of adequate protection of the public health and safety. If the Commission determines that there is such reasonable assurance, it shall allow operation during an interim period under the combined license.

(iv) The Commission, in its discretion, shall determine appropriate hearing procedures, whether informal or formal adjudicatory, for any hearing under clause (i), and shall state its reasons therefor.

(v) The Commission shall, to the maximum possible extent, render a decision on issues raised by the hearing request within 180 days of the publication of the notice provided by clause (i) or the anticipated date for initial loading of fuel into the reactor, whichever is later.

Commencement of operation under a combined license is not subject to subparagraph (A).

(2)

(A) The Commission may issue and make immediately effective any amendment to an operating license or any amendment to a combined construction and operating license, upon a determination by the Commission that such amendment involves no significant hazards consideration, notwithstanding the pendency before the Commission of a request for a hearing from any person.

Such amendment may be issued and made immediately effective in advance of the holding and completion of any required hearing. In determining under this section whether such amendment involves no significant hazards consideration, the Commission shall consult with the State in which the facility involved is located. In all other respects such amendment shall meet the requirements of this Act [42 USCS §§ 2011 et seq.].

(B) The Commission shall periodically (but not less frequently than once every thirty days) publish notice of any amendments issued, or proposed to be issued, as provided in subparagraph (A).

Each such notice shall include all amendments issued, or proposed to be issued, since the date of publication of the last such periodic notice. Such notice shall, with respect to each amendment or proposed amendment (i) identify the facility involved; and (ii) provide a brief description of such amendment. Nothing in this subsection shall be construed to delay the effective date of any amendment.

(C) The Commission shall, during the ninety-day period following the effective date of this paragraph, promulgate regulations establishing (i) standards for determining whether any amendment to an operating license or any amendment to a combined construction and operating license involves no significant hazards consideration; (ii) criteria for providing or, in emergency situations, dispensing with prior notice and reasonable opportunity for public comment on any such determination, which criteria shall take into account the exigency of the need for the amendment involved; and (iii) procedures for consultation on any such determination with the State in which the facility involved is located.

(b) The following Commission actions shall be subject to judicial review in the manner prescribed in chapter 158 of title 28, United States Code [28 USCS §§ 2341 et seq.], and chapter 7 of title 5, United States Code [5 USCS §§ 701 et seq.]:

(1) Any final order entered in any proceeding of the kind specified in subsection (a).

(2) Any final order allowing or prohibiting a facility to begin operating under a combined construction and operating license.

(3) Any final order establishing by regulation standards to govern the Department of Energys gaseous diffusion uranium enrichment plants, including any such facilities leased to a corporation established under the USEC Privatization Act.

(4) Any final determination under section 1701(c) [42 USCS § 2297f(c)] relating to whether the gaseous diffusion plants, including any such facilities leased to a corporation established under the USEC Privatization Act, are in compliance with the Commissions standards governing the gaseous diffusion plants and all applicable laws.

28 USCS § 2342

Current through Public Law 118-40, approved March 1, 2024.

United States Code Service > TITLE 28. JUDICIARY AND JUDICIAL PROCEDURE (§§ 1 5001)

> Part VI. Particular Proceedings (Chs. 151 190) > CHAPTER 158. Orders of Federal Agencies; Review (§§ 2341 2353)

§ 2342. Jurisdiction of court of appeals

The court of appeals (other than the United States Court of Appeals for the Federal Circuit) has exclusive jurisdiction to enjoin, set aside, suspend (in whole or in part), or to determine the validity of (1) all final orders of the Federal Communications Commission made reviewable by section 402(a) of title 47; (2) all final orders of the Secretary of Agriculture made under chapters 9 and 20A of title 7 [7 USCS §§ 181 et seq. and §§ 501 et seq.], except orders issued under sections 210(e), 217a, and 499g(a) of title 7 [7 USCS §§ 210(e), 217a, and 499g(a)];

(3) all rules, regulations, or final orders of (A) the Secretary of Transportation issued pursuant to section 50501, 50502, 56101-56404, or 57109 of title 46 [46 USCS §§ 50501, 50502, 56101-56404, or 57109] or pursuant to part B or C of subtitle IV, subchapter III of chapter 311, chapter 313, or chapter 315 of title 49 [49 USCS §§ 13101 et seq., 15101 et seq., 31131 et seq., 31301 et seq., or 31501 et seq.]; and (B) the Federal Maritime Commission issued pursuant to section 305, 41304, 41308, or 41309 or chapter 421 or 441 of title 46 [46 USCS § 305, 41304, 41308, or 41309 or §§ 42101 et seq. or 44101 et seq.];

(4) all final orders of the Atomic Energy Commission made reviewable by section 2239 of title 42; (5) all rules, regulations, or final orders of the Surface Transportation Board made reviewable by section 2321 of this title [28 USCS § 2321];

(6) all final orders under section 812 of the Fair Housing Act [42 USCS § 3612]; and (7) all final agency actions described in section 20114(c) of title 49.

Jurisdiction is invoked by filing a petition as provided by section 2344 of this title [28 USCS § 2344].

History

HISTORY:

Added Sept. 6, 1966, P. L.89-554, § 4(e), 80 Stat. 622; Jan. 2, 1975, P. L.93-584, § 4, 88 Stat. 1917; Oct. 13, 1978, P. L.95-454, Title II, § 206, 92 Stat. 1144; Oct. 15, 1980, P. L.96-454, § 8(b)(2), 94 Stat. 2021; April 2, 1982, P. L.97-164, P. L.97-164, Title I, Part A, § 137, 96 Stat. 41; Oct. 30, 1984, P. L.98-554, Title II, § 227(a)(4), 98 Stat, 2852; June 19, 1986, P. L.99-336, § 5(a), 100 Stat. 638; Sept. 13, 1988, P. L. 100-430, § 11(a), 102 Stat.

1635; Sept. 3, 1992, P. L. 102-365, § 5(c)(2), 106 Stat. 975; July 5, 1994, P. L. 103-272, § 5(h), 108 Stat. 1375; Dec. 29, 1995, P. L. 104-88, Title III, Subtitle A, § 305(d)(5)-(7), 109 Stat. 945; Oct. 11, 1996, P. L. 104-287, § 6(f)(2), 110 Stat. 3399; Aug. 10, 2005, P. L. 109-59, Title IV, Subtitle A, § 4125(a), 119 Stat. 1738 ; Oct. 6, 2006, P.

L. 109-304, § 17(f)(3), 120 Stat. 1708.

28 USCS § 2344

Current through Public Law 118-40, approved March 1, 2024.

United States Code Service > TITLE 28. JUDICIARY AND JUDICIAL PROCEDURE (§§ 1 5001)

> Part VI. Particular Proceedings (Chs. 151 190) > CHAPTER 158. Orders of Federal Agencies; Review (§§ 2341 2353)

§ 2344. Review of orders; time; notice; contents of petition; service

On the entry of a final order reviewable under this chapter [28 USCS §§ 2341 et seq.], the agency shall promptly give notice thereof by service or publication in accordance with its rules. Any party aggrieved by the final order may, within 60 days after its entry, file a petition to review the order in the court of appeals wherein venue lies. The action shall be against the United States. The petition shall contain a concise statement of (1) the nature of the proceedings as to which review is sought; (2) the facts on which venue is based; (3) the grounds on which relief is sought; and (4) the relief prayed.

The petitioner shall attach to the petition, as exhibits, copies of the order, report, or decision of the agency.

The clerk shall serve a true copy of the petition on the agency and on the Attorney General by registered mail, with request for a return receipt.

History

HISTORY:

Added Sept. 6, 1966, P. L.89-554, § 4(e), 80 Stat. 622.

Annotations

Notes

HISTORY; ANCILLARY LAWS AND DIRECTIVES

Prior law and revision:

5 USCS § 702

Current through Public Law 118-40, approved March 1, 2024.

United States Code Service > TITLE 5. GOVERNMENT ORGANIZATION AND EMPLOYEES (§§ 101 13146) > Part I. The Agencies Generally (Chs. 1 10) > CHAPTER 7. Judicial Review (§§ 701 706)

§ 702. Right of review

A person suffering legal wrong because of agency action, or adversely affected or aggrieved by agency action within the meaning of a relevant statute, is entitled to judicial review thereof. An action in a court of the United States seeking relief other than money damages and stating a claim that an agency or an officer or employee thereof acted or failed to act in an official capacity or under color of legal authority shall not be dismissed nor relief therein be denied on the ground that it is against the United States or that the United States is an indispensable party. The United States may be named as a defendant in any such action, and a judgment or decree may be entered against the United States: Provided, That any mandatory or injunctive decree shall specify the Federal officer or officers (by name or by title), and their successors in office, personally responsible for compliance. Nothing herein (1) affects other limitations on judicial review or the power or duty of the court to dismiss any action or deny relief on any other appropriate legal or equitable ground; or (2) confers authority to grant relief if any other statute that grants consent to suit expressly or impliedly forbids the relief which is sought.

History

HISTORY:

Sept. 6, 1966, P. L.89-554, § 1, 80 Stat. 392; Oct. 21, 1976, P. L.94-574, § 1, 90 Stat. 2721.

Annotations

Notes

HISTORY; ANCILLARY LAWS AND DIRECTIVES

Prior law and revision:

Amendment Notes

1976.

Prior law and revision:

10 CFR 2.206

This document is current through the Mar. 13, 2024 issue of the Federal Register, with the exception of the amendments appearing at 89 FR 16820, 89 FR 17728, 89 FR 17693, and 89 FR 17902.

LEXISNEXIS CODE OF FEDERAL REGULATIONS > Title 10 Energy > Chapter I Nuclear Regulatory Commission > Part 2 Agency Rules of Practice and Procedure > Subpart B Procedure for Imposing Requirements by Order, or for Modification, Suspension, or Revocation of a License, or for Imposing Civil Penalties

§ 2.206 Requests for action under this subpart.

(a) Any person may file a request to institute a proceeding pursuant to § 2.202 to modify, suspend, or revoke a license, or for any other action as may be proper. Requests must be addressed to the Executive Director for Operations and must be filed either by hand delivery to the NRCs Offices at 11555 Rockville Pike, Rockville, Maryland; by mail or telegram addressed to the Executive Director for Operations, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; or by electronic submissions, for example, via facsimile, Electronic Information Exchange, e-mail, or CD-ROM. Electronic submissions must be made in a manner that enables the NRC to receive, read, authenticate, distribute, and archive the submission, and process and retrieve it a single page at a time. Detailed guidance on making electronic submissions can be obtained by visiting the NRCs Web site at http://www.nrc.gov/site-help/e-submittals.html; by e-mail to

MSHD.Resource@nrc.gov

or by writing the Office of the Chief Information Officer, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001. The request must specify the action requested and set forth the facts that constitute the basis for the request. The Executive Director for Operations will refer the request to the Director of the NRC office with responsibility for the subject matter of the request for appropriate action in accordance with paragraph (b) of this section.

(b) Within a reasonable time after a request pursuant to paragraph (a) of this section has been received, the Director of the NRC office with responsibility for the subject matter of the request shall either institute the requested proceeding in accordance with this subpart or shall advise the person who made the request in writing that no proceeding will be instituted in whole or in part, with respect to the request, and the reasons for the decision.

(c)

(1) Directors decisions under this section will be filed with the Office of the Secretary. Within twenty-five (25) days after the date of the Directors decision under this section that no proceeding will be instituted or other action taken in whole or in part, the Commission may on its own motion review that decision, in whole or in part, to determine if the Director has abused their discretion. This review power does not limit in any way either the Commissions supervisory power over delegated staff actions or the Commissions power to consult with the staff on a formal or informal basis regarding institution of proceedings under this section.

(2) No petition or other request for Commission review of a Directors decision under this section will be entertained by the Commission.

(3) The Secretary is authorized to extend the time for Commission review on its own motion of a Directors denial under paragraph (c) of this section.

Statutory Authority 10 CFR 50.61

This document is current through the Mar. 13, 2024 issue of the Federal Register, with the exception of the amendments appearing at 89 FR 16820, 89 FR 17728, 89 FR 17693, and 89 FR 17902.

LEXISNEXIS CODE OF FEDERAL REGULATIONS > Title 10 Energy > Chapter I Nuclear Regulatory Commission > Part 50 Domestic Licensing of Production and Utilization Facilities

> Issuance, Limitations, and Conditions of Licenses and Construction Permits

§ 50.61 Fracture toughness requirements for protection against pressurized thermal shock events.

(a) Definitions. For the purposes of this section:

(1) ASME Code means the American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section III, Division I, Rules for the Construction of Nuclear Power Plant Components, edition and addenda and any limitations and modifications thereof as specified in § 50.55a.

(2) Pressurized Thermal Shock Event means an event or transient in pressurized water reactors (PWRs) causing severe overcooling (thermal shock) concurrent with or followed by significant pressure in the reactor vessel.

(3) Reactor Vessel Beltline means the region of the reactor vessel (shell material including welds, heat affected zones and plates or forgings) that directly surrounds the effective height of the active core and adjacent regions of the reactor vessel that are predicted to experience sufficient neutron radiation damage to be considered in the selection of the most limiting material with regard to radiation damage.

(4) RT[NDT] means the reference temperature for a reactor vessel material, under any conditions. For the reactor vessel beltline materials, RT[NDT] must account for the effects of neutron radiation.

(5) RT [NDT(U)] means the reference temperature for a reactor vessel material in the pre-service or unirradiated condition, evaluated according to the procedures in the ASME Code, Paragraph NB-2331 or other methods approved by the Director, Office of Nuclear Reactor Regulation.

(6) EOL Fluence means the best-estimate neutron fluence projected for a specific vessel beltline material at the clad-base-metal interface on the inside surface of the vessel at the location where the material receives the highest fluence on the expiration date of the operating license.

(7) RT[PTS] means the reference temperature, RT[NDT], evaluated for the EOL Fluence for each of the vessel beltline materials, using the procedures of paragraph (c) of this section.

(8) PTS Screening Criterion means the value of RT[PTS] for the vessel beltline material above which the plant cannot continue to operate without justification.

(b) Requirements. (1) For each pressurized water nuclear power reactor for which an operating license has been issued under this part or a combined license issued under Part 52 of this chapter, other than a nuclear power reactor facility for which the certification required under § 50.82(a)(1) has been submitted, the licensee shall have projected values of RT[PTS] or RT[MAX-X], accepted by the NRC, for each reactor vessel beltline material. For pressurized water nuclear power reactors for which a construction permit was issued under this part before February 3, 2010 and whose reactor vessel was designed and fabricated to the 1998 Edition or earlier of the ASME Code, the projected values must be in accordance with this section or § 50.61a. For pressurized water nuclear power reactors for which a construction permit is issued under this part after February 3, 2010 and whose reactor vessel is designed and fabricated to an ASME Code after the 1998 Edition, or for which a combined license is issued under Part 52, the projected values must Page 2 of 9 10 CFR 50.61

be in accordance with this section. When determining compliance with this section, the assessment of RT[PTS] must use the calculation procedures described in paragraph (c)(1) and perform the evaluations described in paragraphs (c)(2) and (c)(3) of this section. The assessment must specify the bases for the projected value of RT[PTS] for each vessel beltline material, including the assumptions regarding core loading patterns, and must specify the copper and nickel contents and the fluence value used in the calculation for each beltline material. This assessment must be updated whenever there is a significant 2change in projected values of RT[PTS], or upon request for a change in the expiration date for operation of the facility.

(2) The pressurized thermal shock (PTS) screening criterion is 270° F for plates, forgings, and axial weld materials, and 300° F for circumferential weld materials. For the purpose of comparison with this criterion, the value of RT[PTS] for the reactor vessel must be evaluated according to the procedures of paragraph (c) of this section, for each weld and plate, or forging, in the reactor vessel beltline. RT[PTS]

must be determined for each vessel beltline material using the EOL fluence for that material.

(3) For each pressurized water nuclear power reactor for which the value of RT[PTS] for any material in the beltline is projected to exceed the PTS screening criterion using the EOL fluence, the licensee shall implement those flux reduction programs that are reasonably practicable to avoid exceeding the PTS screening criterion set forth in paragraph (b)(2) of this section. The schedule for implementation of flux reduction measures may take into account the schedule for submittal and anticipated approval by the Director, Office of Nuclear Reactor Regulation, of detailed plant-specific analyses, submitted to demonstrate acceptable risk with RT[PTS] above the screening limit due to plant modifications, new information or new analysis techniques.

(4) For each pressurized water nuclear power reactor for which the analysis required by paragraph (b)(3) of this section indicates that no reasonably practicable flux reduction program will prevent RT[PTS] from exceeding the PTS screening criterion using the EOL fluence, the licensee shall submit a safety analysis to determine what, if any, modifications to equipment, systems, and operation are necessary to prevent potential failure of the reactor vessel as a result of postulated PTS events if continued operation beyond the screening criterion is allowed. In the analysis, the licensee may determine the properties of the reactor vessel materials based on available information, research results, and plant surveillance data, and may use probabilistic fracture mechanics techniques. This analysis must be submitted at least three years before RT[PTS] is projected to exceed the PTS screening criterion.

(5) After consideration of the licensees analyses, including effects of proposed corrective actions, if any, submitted in accordance with paragraphs (b)(3) and (b)(4) of this section, the Director, Office of Nuclear Reactor Regulation, may, on a case-by-case basis, approve operation of the facility with RT[Sub]PTS in excess of the PTS screening criterion. The Director, Office of Nuclear Reactor Regulation, will consider factors significantly affecting the potential for failure of the reactor vessel in reaching a decision.

(6) If the Director, Office of Nuclear Reactor Regulation, concludes, pursuant to paragraph (b)(5) of this section, that operation of the facility with RT[PTS] in excess of the PTS screening criterion cannot be approved on the basis of the licensees analyses submitted in accordance with paragraphs (b)(3) and (b)(4) of this section, the licensee shall request and receive approval by the Director, Office of Nuclear Reactor Regulation, prior to any operation beyond the criterion. The request must be based upon modifications to equipment, systems, and operation of the facility in addition to those previously proposed in the submitted analyses that would reduce the potential for failure of the reactor vessel due to PTS events, or upon further analyses based upon new information or improved methodology.

2 Changes to RT[PTS] values are considered significant if either the previous value or the current value, or both values, exceed the screening criterion before the expiration of the operating license or the combined license under part 52 of this chapter, including any renewed term, if applicable for the plant.

Page 3 of 9 10 CFR 50.61

(7) If the limiting RT[PTS] value of the plant is projected to exceed the screening criteria in paragraph (b)(2), or the criteria in paragraphs (b)(3) through (b)(6) of this section cannot be satisfied, the reactor vessel beltline may be given a thermal annealing treatment to recover the fracture toughness of the material, subject to the requirements of § 50.66. The reactor vessel may continue to be operated only for that service period within which the predicted fracture toughness of the vessel beltline materials satisfy the requirements of paragraphs (b)(2) through (b)(6) of this section, with RT[PTS] accounting for the effects of annealing and subsequent irradiation.

(c) Calculation of RT[PTS]. RT[PTS] must be calculated for each vessel beltline material using a fluence value, f, which is the EOL fluence for the material. RT[PTS] must be evaluated using the same procedures used to calculate RT[NDT], as indicated in paragraph (c)(1) of this section, and as provided in paragraphs (c)(2) and (c)(3) of this section.

(1) Equation 1 must be used to calculate values of RT[NDT] for each weld and plate, or forging, in the reactor vessel beltline.

Equation 1: RT[NDT] = RT[NDT](U) + M + [DELTA]RT[NDT]

(i) If a measured value of RT[NDT](U) is not available, a generic mean value for the class 3of material may be used if there are sufficient test results to establish a mean and a standard deviation for the class.

(ii) For generic values of weld metal, the following generic mean values must be used unless justification for different values is provided: 0° F for welds made with Linde 80 flux, and -56° F for welds made with Linde 0091, 1092 and 124 and ARCOS B-5 weld fluxes.

(iii) M means the margin to be added to account for uncertainties in the values of RT[NDT](U),

copper and nickel contents, fluence and the calculational procedures. M is evaluated from Equation 2.

Equation 2 Click here to view this image.

(A) In Equation 2, [sigma][U] is the standard deviation for RT[NDT](U). If a measured value of RT[NDT](U) is used, then s[U] is determined from the precision of the test method. If a measured value of RT[NDT](U) is not available and a generic mean value for that class of materials is used, then [sigma][U] is the standard deviation obtained from the set of data used to establish the mean. If a generic mean value given in paragraph (c)(1)(i)(B) of this section for welds is used, then [sigma][U] is 17° F.

(B) In Equation 2, [sigma]D is the standard deviation for [DELTA]RT[NDT]. The value of

[sigma]D to be used is 28° F for welds and 17° F for base metal; the value of [sigma][DELTA]

need not exceed one-half of [DELTA]RT[NDT].

(iv) [DELTA]RT[NDT] is the mean value of the transition temperature shift, or change in RT[NDT],

due to irradiation, and must be calculated using Equation 3.

Equation 3: [DELTA]RT[NDT]=(CF)f [Sup] (0.28-0.10 log f)

(A) CF (° F) is the chemistry factor, which is a function of copper and nickel content. CF is given in Table 1 for welds and in Table 2 for base metal (plates and forgings). Linear interpolation is permitted. In Tables 1 and 2, Wt-% copper and Wt-% nickel are the best-estimate values for the material, which will normally be the mean of the measured values for a plate or forging. For a weld, the best estimate values will normally be the mean of the measured values for a weld deposit made using the same weld wire heat number as the critical vessel weld. If these values are not available, the upper limiting values given in the material specifications to which the vessel material was fabricated may be used. If not available,

3 The class of material for estimating RT[NDT](U) is generally determined for welds by the type of welding flux (Linde 80, or other), and for base metal by the material specification.

Page 4 of 9 10 CFR 50.61

conservative estimates (mean plus one standard deviation) based on generic data 4may be used if justification is provided. If none of these alternatives are available, 0.35% copper and 1.0% nickel must be assumed.

(B) f is the best estimate neutron fluence, in units of 10<19> n <2> (E greater than 1 MeV), at the clad-base-metal interface on the inside surface of the vessel at the location where the material in question receives the highest fluence for the period of service in question. As specified in this paragraph, the EOL fluence for the vessel beltline material is used in calculating KRT[PTS].

(v) Equation 4 must be used for determining RT[PTS] using equation 3 with EOL fluence values for determining DRT[PTS].

Equation 4: RT[Sub]PTS=RT[NDT](U)+M+[DELTA]RT[PTS]

(2) To verify that RT[NDT] for each vessel beltline material is a bounding value for the specific reactor vessel, licensees shall consider plant-specific information that could affect the level of embrittlement.

This information includes but is not limited to the reactor vessel operating temperature and any related surveillance program 5results.

(i) Results from the plant-specific surveillance program must be integrated into the RT[NDT]

estimate if the plant-specific surveillance data has been deemed credible as judged by the following criteria:

(A) The materials in the surveillance capsules must be those which are the controlling materials with regard to radiation embrittlement.

(B) Scatter in the plots of Charpy energy versus temperature for the irradiated and unirradiated conditions must be small enough to permit the determination of the 30-foot-pound temperature unambiguously.

(C) Where there are two or more sets of surveillance data from one reactor, the scatter of

[DELTA]RT[NDT] values must be less than 28° F for welds and 17° F for base metal. Even if the range in the capsule fluences is large (two or more orders of magnitude), the scatter may not exceed twice those values.

(D) The irradiation temperature of the Charpy specimens in the capsule must equal the vessel wall temperature at the cladding/base metal interface within [+/-] 25° F.

(E) The surveillance data for the correlation monitor material in the capsule, if present, must fall within the scatter band of the data base for the material.

(ii)

(A) Surveillance data deemed credible according to the criteria of paragraph (c)(2)(i) of this section must be used to determine a material-specific value of CF for use in Equation 3. A material-specific value of CF is determined from Equation 5.

Equation 5 Click here to view this image.

(B) In Equation 5, n is the number of surveillance data points, A[i] is the measured value of

[DELTA]RT[NDT] and f[i] is the fluence for each surveillance data point. If there is clear evidence that the copper and nickel content of the surveillance weld differs from the vessel weld, i.e. differs from the average for the weld wire heat number associated with the vessel

4 Data from reactor vessels fabricated to the same material specification in the same shop as the vessel in question and in the same time period is an example of generic data.

5 Surveillance program results means any data that demonstrates the embrittlement trends for the limiting beltline material, including but not limited to data from test reactors or from surveillance programs at other plants with or without surveillance program integrated per 10 CFR Part 50, Appendix H.

Page 5 of 9 10 CFR 50.61

weld and the surveillance weld, the measured values of [DELTA]RT[Sub]NDT must be adjusted for differences in copper and nickel content by multiplying them by the ratio of the chemistry factor for the vessel material to that for the surveillance weld.

(iii) For cases in which the results from a credible plant-specific surveillance program are used, the value of [sigma]D to be used is 14° F for welds and 8.5° F for base metal; the value of [sigma]D need not exceed one-half of [DELTA]RT[NDT].

(iv) The use of results from the plant-specific surveillance program may result in an RT[NDT] that is higher or lower than those determined in paragraph (c)(1).

(3) Any information that is believed to improve the accuracy of the RT[PTS] value significantly must be reported to the Director, Office of Nuclear Reactor Regulation Any value of RT[PTS] that has been modified using the procedures of paragraph (c)(2) of this section is subject to the approval of the Director, Office of Nuclear Reactor Regulation when used as provided in this section.

Table 1.

Chemistry Factor for Weld Metals, degrees F

Nickel, wt-%

Copper, wt- 0 0.20 0.40 0.60 0.80 1.00 1.20

0 20 20 20 20 20 20 20

0.01 20 20 20 20 20 20 20

0.02 21 26 27 27 27 27 27

0.03 22 35 41 41 41 41 41

0.04 24 43 54 54 54 54 54

0.05 26 49 67 68 68 68 68

0.06 29 52 77 82 82 82 82

0.07 32 55 85 95 95 95 95

0.08 36 58 90 106 108 108 108

0.09 40 61 94 115 122 122 122

0.10 44 65 97 122 133 135 135

0.11 49 68 101 130 144 148 148

0.12 52 72 103 135 153 161 161

0.13 58 76 106 139 162 172 176

0.14 61 79 109 142 168 182 188

0.15 66 84 112 146 175 191 200

0.16 70 88 115 149 178 199 211

0.17 75 92 119 151 184 207 221

0.18 79 95 122 154 187 214 230

0.19 83 100 126 157 191 220 238 Page 6 of 9 10 CFR 50.61

Table 1.

Chemistry Factor for Weld

Metals, degrees F

Nickel, wt-%

Copper, wt- 0 0.20 0.40 0.60 0.80 1.00 1.20

0.20 88 104 129 160 194 223 245

0.21 92 108 133 164 197 229 252

0.22 97 112 137 167 200 232 257

0.23 101 117 140 169 203 236 263

0.24 105 121 144 173 206 239 268

0.25 110 126 148 176 209 243 272

0.26 113 130 151 180 212 246 276

0.27 119 134 155 184 216 249 280

0.28 122 138 160 187 218 251 284

0.29 128 142 164 191 222 254 287

0.30 131 146 167 194 225 257 290

0.31 136 151 172 198 228 260 293

0.32 140 155 175 202 231 263 296

0.33 144 160 180 205 234 266 299

0.34 149 164 184 209 238 269 302

0.35 153 168 187 212 241 272 305

0.36 158 172 191 216 245 275 308

0.37 162 177 196 220 248 278 311

0.38 166 182 200 223 250 281 314

0.39 171 185 203 227 254 285 317

0.40 175 189 207 231 257 288 320

Table 2.

Chemistry Factor for Base

Metals, degrees F

Nickel, wt-%

Copper, wt- 0 0.20 0.40 0.60 0.80 1.00 1.20 Page 7 of 9 10 CFR 50.61

Table 2.

Chemistry Factor for Base

Metals, degrees F

Nickel, wt-%

Copper, wt- 0 0.20 0.40 0.60 0.80 1.00 1.20

0 20 20 20 20 20 20 20

0.01 20 20 20 20 20 20 20

0.02 20 20 20 20 20 20 20

0.03 20 20 20 20 20 20 20

0.04 22 26 26 26 26 26 26

0.05 25 31 31 31 31 31 31

0.06 28 37 37 37 37 37 37

0.07 31 43 44 44 44 44 44

0.08 34 48 51 51 51 51 51

0.09 37 53 58 58 58 58 58

0.10 41 58 65 65 67 67 67

0.11 45 62 72 74 77 77 77

0.12 49 67 79 83 86 86 86

0.13 53 71 85 91 96 96 96

0.14 57 75 91 100 105 106 106

0.15 61 80 99 110 115 117 117

0.16 65 84 104 118 123 125 125

0.17 69 88 110 127 132 135 135

0.18 73 92 115 134 141 144 144

0.19 78 97 120 142 150 154 154

0.20 82 102 125 149 159 164 165

0.21 86 107 129 155 167 172 174

0.22 91 112 134 161 176 181 184

0.23 95 117 138 167 184 190 194

0.24 100 121 143 172 191 199 204

0.25 104 126 148 176 199 208 214

0.26 109 130 151 180 205 216 221

0.27 114 134 155 184 211 225 230

0.28 119 138 160 187 216 233 239 Page 8 of 9 10 CFR 50.61

Table 2.

Chemistry Factor for Base Metals, degrees F

Nickel, wt-%

Copper, wt- 0 0.20 0.40 0.60 0.80 1.00 1.20

0.29 124 142 164 191 221 241 248

0.30 129 146 167 194 225 249 257

0.31 134 151 172 198 228 255 266

0.32 139 155 175 202 231 260 274

0.33 144 160 180 205 234 264 282

0.34 149 164 184 209 238 268 290

0.35 153 168 187 212 241 272 298

0.36 158 173 191 216 245 275 303

0.37 162 177 196 220 248 278 308

0.38 166 182 200 223 250 281 313

0.39 171 185 203 227 254 285 317

0.40 175 189 207 231 257 288 320

Statutory Authority

Authority Note Applicable to 10 CFR Ch. I, Pt. 50

History

[56 FR 22304, May 15, 1991; 60 FR 65456, 65468, Dec. 19, 1995; 61 FR 39278, 39300, July 29, 1996; 72 FR 49352, 49500, Aug. 28, 2007; 73 FR 5709, 5722, Jan. 31, 2008; 75 FR 13, 23, Jan. 4, 2010; 84 FR 65639, 65644, Nov. 29, 2019, as corrected at 84 FR 66561, Dec. 5, 2019]

Annotations

Notes

[EFFECTIVE DATE NOTE:

72 FR 49352, 49500, Aug. 28, 2007, revised paragraph (b)(1), effective Sept. 27, 2007; 73 FR 5709, 5722, Jan.

31, 2008, revised paragraphs (a)(5) and (c)(3), effective Jan. 31, 2008; 75 FR 13, 23, Jan. 4, 2010, revised paragraph (b)(1), effective Feb. 3, 2010; 84 FR 65639, 65644, Nov. 29, 2019, amended this section, effective Dec.

30, 2019.]

10 CFR PART 50 APPENDIX H

This document is current through the Mar. 13, 2024 issue of the Federal Register, with the exception of the amendments appearing at 89 FR 16820, 89 FR 17728, 89 FR 17693, and 89 FR 17902.

LEXISNEXIS CODE OF FEDERAL REGULATIONS > Title 10 Energy > Chapter I Nuclear Regulatory Commission > Part 50 Domestic Licensing of Production and Utilization Facilities

APPENDIX H TO PART 50 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM REQUIREMENTS

I.Introduction II.Definitions III. Surveillance Program Criteria IV. Report of Test Results I.Introduction The purpose of the material surveillance program required by this appendix is to monitor changes in the fracture toughness properties of ferritic materials in the reactor vessel beltline region of light water nuclear power reactors which result from exposure of these materials to neutron irradiation and the thermal environment. Under the program, fracture toughness test data are obtained from material specimens exposed in surveillance capsules, which are withdrawn periodically from the reactor vessel.

These data will be used as described in Section IV of Appendix G to Part 50.

ASTM E 185-73, Standard Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels; ASTM E 185-79, Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels; and ASTM E 185-82, Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels; which are referenced in the following paragraphs, have been approved for incorporation by reference by the Director of the Federal Register. Copies of ASTM E 185-73, -79, and -82, may be purchased from the American Society for Testing and Materials, 1916 Race Street, Philadelphia, PA 19103 and are available for inspection at the NRC Library, 11545 Rockville Pike, Two White Flint North, Rockville, MD 20852-2738.

II.Definitions All terms used in this Appendix have the same meaning as in Appendix G.

III. Surveillance Program Criteria A. No material surveillance program is required for reactor vessels for which it can be conservatively demonstrated by analytical methods applied to experimental data and tests performed on comparable vessels, making appropriate allowances for all uncertainties in the measurements, that the peak neutron fluence at the end of the design life of the vessel will not exceed 10<17> n <2> (E > 1 MeV).

B. Reactor vessels that do not meet the conditions of paragraph III.A of this appendix must have their beltline materials monitored by a surveillance program complying with ASTM E 185, as modified by this appendix.

1. The design of the surveillance program and the withdrawal schedule must meet the requirements of the edition of the ASTM E 185 that is current on the issue date of the ASME Code to which the reactor vessel was purchased; for reactor vessels purchased after 1982, the design of Page 2 of 4 10 CFR PART 50 APPENDIX H

the surveillance program and the withdrawal schedule must meet the requirements of ASTM E 185-82. For reactor vessels purchased in or before 1982, later editions of ASTM E 185 may be used, but including only those editions through 1982. For each capsule withdrawal, the test procedures and reporting requirements must meet the requirements of ASTM E 185-82 to the extent practicable for the configuration of the specimens in the capsule. If any of the optional provisions in paragraphs III.B.4(a) through (d) of this section are implemented in lieu of ASTM E 185, the number of specimens included or tested in the surveillance program shall be adjusted as specified in paragraphs III.B.4(a) through (d) of this section.

2. Surveillance specimen capsules must be located near the inside vessel wall in the beltline region so that the specimen irradiation history duplicates, to the extent practicable within the physical constraints of the system, the neutron spectrum, temperature history, and maximum neutron fluence experienced by the reactor vessel inner surface. If the capsule holders are attached to the vessel wall or to the vessel cladding, construction and inservice inspection of the attachments and attachment welds must be done according to the requirements for permanent structural attachments to reactor vessels given in Sections III and XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code). The design and location of the capsule holders must permit insertion of replacement capsules. Accelerated irradiation capsules may be used in addition to the required number of surveillance capsules.

3. A proposed withdrawal schedule must be submitted with a technical justification as specified in § 50.4. The proposed schedule must be approved prior to implementation.

4. Optional provisions. As used in this section, references to ASTM E 185 include the edition of ASTM E 185 that is current on the issue date of the ASME Code to which the reactor vessel was purchased through the 1982 edition.

a. First Provision: Heat-Affected Zone SpecimensThe inclusion or testing of weld heat-affected zone Charpy impact specimens within the surveillance program as specified in ASTM E 185 is optional.

b. Second Provision: Tension SpecimensIf this provision is implemented, the minimum number of tension specimens to be included and tested in the surveillance program shall be as specified in paragraphs III.B.4(b)(i) and (ii) of this section.

i. Unirradiated Tension SpecimensTwo tension specimens from each base and weld material required by ASTM E 185 shall be tested, with one specimen tested at room temperature and the other specimen tested at the service temperature; and ii. Irradiated Tension SpecimensTwo tension specimens from each base and weld material required by ASTM E 185 shall be included in each surveillance capsule and tested, with one specimen tested at room temperature and the other specimen tested at the service temperature.

c. Third Provision: Correlation Monitor MaterialsThe testing of correlation monitor material specimens within the surveillance program as specified in ASTM E 185 is optional.

d. Fourth Provision: Thermal Monitor The inclusion or examination of thermal monitors within the surveillance program as specified in ASTM E 185 is optional.

C. Requirements for an Integrated Surveillance Program.

1. In an integrated surveillance program, the representative materials chosen for surveillance for a reactor are irradiated in one or more other reactors that have similar design and operating features.

Integrated surveillance programs must be approved by the Director, Office of Nuclear Reactor Regulation on a case-by-case basis. Criteria for approval include the following:

Page 3 of 4 10 CFR PART 50 APPENDIX H

a. The reactor in which the materials will be irradiated and the reactor for which the materials are being irradiated must have sufficiently similar design and operating features to permit accurate comparisons of the predicted amount of radiation damage.

b. Each reactor must have an adequate dosimetry program.

c. There must be adequate arrangement for data sharing between plants.

d. There must be a contingency plan to assure that the surveillance program for each reactor will not be jeopardized by operation at reduced power level or by an extended outage of another reactor from which data are expected.

e. There must be substantial advantages to be gained, such as reduced power outages or reduced personnel exposure to radiation, as a direct result of not requiring surveillance capsules in all reactors in the set.

2. No reduction in the requirements for number of materials to be irradiated, specimen types, or number of specimens per reactor is permitted.

3. After (the effective date of this section), no reduction in the amount of testing is permitted unless previously authorized by the Director, Office of Nuclear Reactor Regulation or the Director, Office of New Reactors, as appropriate.

IV. Report of Test Results A. Each capsule withdrawal and the test results must be the subject of a summary technical report to be submitted, as specified in § 50.4, within eighteen months of the date of capsule withdrawal, unless an extension is granted by the Director, Office of Nuclear Reactor Regulation.

B. The report must include the data required by ASTM E 185, as specified in paragraph III.B.1 of this appendix, and the results of all fracture toughness tests conducted on the beltline materials in the irradiated and unirradiated conditions.

C. If a change in the Technical Specifications is required, either in the pressure-temperature limits or in the operating procedures required to meet the limits, the expected date for submittal of the revised Technical Specifications must be provided with the report.

Statutory Authority

Authority Note Applicable to 10 CFR Ch. I, Pt. 50

History

[48 FR 24011, May 27, 1983, as amended at 51 FR 40311, Nov. 6, 1986; 53 FR 43420, Oct. 27, 1988; 57 FR 61786, Dec. 29, 1992; 59 FR 50689, Oct. 5, 1994; 60 FR 65456, 65476, Dec. 19, 1995; 68 FR 75388, 75390, Dec.

31, 2003; 73 FR 5709, 5723, Jan. 31, 2008; 84 FR 65639, 65644, Nov. 29, 2019, as corrected at 84 FR 66561,

Dec. 5, 2019; 85 FR 62199, 62207, Oct. 2, 2020, as confirmed at 85 FR 85503, Dec. 29, 2020; 88 FR 57873, 57878, Aug. 24, 2023]

Annotations

Notes

[EFFECTIVE DATE NOTE:

10 CFR 50.92

This document is current through the Mar. 13, 2024 issue of the Federal Register, with the exception of the amendments appearing at 89 FR 16820, 89 FR 17728, 89 FR 17693, and 89 FR 17902.

LEXISNEXIS CODE OF FEDERAL REGULATIONS > Title 10 Energy > Chapter I Nuclear Regulatory Commission > Part 50 Domestic Licensing of Production and Utilization Facilities

> Amendment of License or Construction Permit at Request of Holder

§ 50.92 Issuance of amendment.

(a) In determining whether an amendment to a license, construction permit, or early site permit will be issued to the applicant, the Commission will be guided by the considerations which govern the issuance of initial licenses, construction permits, or early site permits to the extent applicable and appropriate. If the application involves the material alteration of a licensed facility, a construction permit will be issued before the issuance of the amendment to the license, provided however, that if the application involves a material alteration to a nuclear power reactor manufactured under part 52 of this chapter before its installation at a site, or a combined license before the date that the Commission makes the finding under § 52.103(g) of this chapter, no application for a construction permit is required. If the amendment involves a significant hazards consideration, the Commission will give notice of its proposed action:

(1) Under § 2.105 of this chapter before acting thereon; and (2) As soon as practicable after the application has been docketed.

(b) The Commission will be particularly sensitive to a license amendment request that involves irreversible consequences (such as one that permits a significant increase in the amount of effluents or radiation emitted by a nuclear power plant).

(c) The Commission may make a final determination, under the procedures in § 50.91, that a proposed amendment to an operating license or a combined license for a facility or reactor licensed under §§ 50.21(b) or 50.22, or for a testing facility involves no significant hazards consideration, if operation of the facility in accordance with the proposed amendment would not:

(1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in a margin of safety.

Statutory Authority

Authority Note Applicable to 10 CFR Ch. I, Pt. 50

History

[51 FR 7767, Mar. 6, 1986; 72 FR 49352, 49504, Aug. 28, 2007 ]

Page 2 of 2 10 CFR 50.92

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