Safety Evaluation Report

Related to the License Renewal 

of Sequoyah Nuclear Plant

Units 1 and 2

Docket Numbers 50-327 and 50-328

Tennessee Valley Authority

Office of Nuclear Reactor Regulation

NUREG-2181

Supplement 1

 
AVAILABILITY OF REFERENCE MATERIALS

IN NRC PUBLICATIONS

NRC Reference Material

As of November 1999, you may electronically access 

NUREG-series publications and other NRC records at 

NRC’s Library at www.nrc.gov/reading-rm.html. Publicly 

released records include, to name a few, NUREG-series

publications; Federal Register notices; applicant, 

licensee, and vendor documents and correspondence; 

NRC correspondence and internal memoranda; bulletins 

and information notices; inspection and investigative 

reports; licensee event reports; and Commission papers 

and their attachments.

NRC publications in the NUREG series, NRC regulations, 

and Title 10, “Energy,” in the Code of Federal Regulations

may also be purchased from one of these two sources.

1. The Superintendent of Documents

U.S. Government Publishing Office

Mail Stop IDCC

Washington, DC 20402-0001

Internet: bookstore.gpo.gov

Telephone: (202) 512-1800

Fax: (202) 512-2104

2. The National Technical Information Service

5301 Shawnee Rd., Alexandria, VA 22312-0002

www.ntis.gov

1-800-553-6847 or, locally, (703) 605-6000

A single copy of each NRC draft report for comment is

available free, to the extent of supply, upon written

request as follows:

Address: U.S. Nuclear Regulatory Commission

Office of Administration

Publications Branch

Washington, DC 20555-0001

E-mail: distribution.resource@nrc.gov

Facsimile: (301) 415-2289

Some publications in the NUREG series that are posted 

at NRC’s Web site address www.nrc.gov/reading-rm/

doc-collections/nuregs are updated periodically and may 

differ from the last printed version. Although references to 

material found on a Web site bear the date the material 

was accessed, the material available on the date cited 

may subsequently be removed from the site.

Non-NRC Reference Material

Documents available from public and special technical 

libraries include all open literature items, such as books, 

journal articles, transactions, Federal Register notices, 

Federal and State legislation, and congressional reports. 

Such documents as theses, dissertations, foreign reports 

and translations, and non-NRC conference proceedings 

may be purchased from their sponsoring organization.

Copies of industry codes and standards used in a

substantive manner in the NRC regulatory process are 

maintained at—

The NRC Technical Library

Two White Flint North

11545 Rockville Pike

Rockville, MD 20852-2738

These standards are available in the library for reference 

use by the public. Codes and standards are usually 

copyrighted and may be purchased from the originating 

organization or, if they are American National Standards, 

from—

American National Standards Institute

11 West 42nd Street

New York, NY 10036-8002

www.ansi.org

(212) 642-4900

Legally binding regulatory requirements are stated only in 

laws; NRC regulations; licenses, including technical specifications; or orders, not in NUREG-series publications. The 

views expressed in contractorprepared publications in this 

series are not necessarily those of the NRC.

The NUREG series comprises (1) technical and administrative reports and books prepared by the staff (NUREG–

XXXX) or agency contractors (NUREG/CR–XXXX), (2) 

proceedings of conferences (NUREG/CP–XXXX), (3) reports 

resulting from international agreements (NUREG/IA–XXXX), 

(4) brochures (NUREG/BR–XXXX), and (5) compilations of 

legal decisions and orders of the Commission and Atomic 

and Safety Licensing Boards and of Directors’ decisions 

under Section 2.206 of NRC’s regulations (NUREG–0750).

DISCLAIMER: This report was prepared as an account 

of work sponsored by an agency of the U.S. Government. 

Neither the U.S. Government nor any agency thereof, nor 

any employee, makes any warranty, expressed or implied, 

or assumes any legal liability or responsibility for any third 

party’s use, or the results of such use, of any information, 

apparatus, product, or process disclosed in this publication, 

or represents that its use by such third party would not 

infringe privately owned rights.
Safety Evaluation Report

Related to the License Renewal 

of Sequoyah Nuclear Plant

Units 1 and 2

Docket Numbers 50-327 and 50-328

Tennessee Valley Authority

Manuscript Completed: September 2015

Date Published: September 2015

Office of Nuclear Reactor Regulation

NUREG-2181

Supplement 1

 

 iii 

ABSTRACT 

This document is a supplement to NUREG-2181, “Safety Evaluation Report Related to the 

License Renewal of the Sequoyah Nuclear Plant, Units 1 and 2,” for the license renewal 

application (LRA) for Sequoyah Nuclear Plant (Sequoyah or SQN), Units 1 and 2. By letter 

dated January 7, 2013, Tennessee Valley Authority (TVA or the applicant) submitted the LRA in 

accordance with Title 10, Part 54, “Requirements for Renewal of Operating Licenses for Nuclear 

Power Plants,” of the Code of Federal Regulations. TVA requests renewal of the operating 

licenses (Facility Operating License Nos. DPR 77 and DPR 79) for a period of 20 years beyond 

the current expiration at midnight September 17, 2020, for Unit 1, and at midnight September 

15, 2021, for Unit 2. 

This supplement to NUREG 2181 (SSER) documents the staff’s review of supplemental 

information provided by the applicant since the issuance of the Safety Evaluation Report (SER). 

This evaluation considered TVA’s LRA annual updates dated April 22, 2014, August 17, 2015, 

and August 28, 2015, supplemental information regarding recent industry operating experience 

related to SQN Unit 1 reactor vessel internals, Commitment No. 28 revisions (pertaining to 

reactor vessel internals and surveillance capsule monitoring), information required by 10 CFR 

54.21(b), as well as information provided in response to staff requests for additional information. 

This document discusses only the changes to the SER.

v 

TABLE OF CONTENTS 

ABSTRACT ................................................................................................................................ iii 

ABBREVIATIONS ...................................................................................................................... vii 

SECTION 1 INTRODUCTION AND GENERAL DISCUSSION ................................................ 1-1 

1.1 Introduction ............................................................................................................. 1-1

SECTION 2 STRUCTURES AND COMPONENTS SUBJECT TO AGING 

MANAGEMENT REVIEW .................................................................................... 2-1

2.5 Scoping and Screening Results: Electrical and Instrumentation and 

Control Systems ...................................................................................................... 2-1

2.5.1 Electrical, and Instrumentation, and Controls Commodity Groups ........... 2-1

2.5.1.1 Summary of Technical Information in the Application ................ 2-1

2.5.1.2 Staff Evaluation.......................................................................... 2-1

2.5.1.3 Conclusion ................................................................................. 2-2

SECTION 3 AGING MANAGEMENT REVIEW RESULTS ...................................................... 3-1

3.0 Applicant’s Use of the Generic Aging Lessons Learned Report ............................. 3-1

3.0.3 Aging Management Programs .................................................................. 3-1

3.0.3.2 AMPs Consistent with the GALL Report with Exceptions 

or Enhancements ..................................................................... 3-1

SECTION 5 REVIEW BY THE ADVISORY COMMITTEE ON REACTOR 

SAFEGUARDS ..................................................................................................... 5-1

SECTION 6 CONCLUSION ....................................................................................6-1 

APPENDIX A: SEQUOYAH NUCLEAR PLANT, UNITS 1 AND 2, LICENSE 

RENEWAL COMMITMENTS ........................................................................... A-1

APPENDIX B: CHRONOLOGY ............................................................................................... B-1

APPENDIX C: PRINCIPAL CONTRIBUTORS ....................................................................... C-1

APPENDIX D: REFERENCES ................................................................................................ D-1

vii 

ABBREVIATIONS 

ACRS Advisory Committee on Reactor Safeguards 

ADAMS Agencywide Documents Access and Management System 

AMG Aging Management Guide 

AMP aging management program 

AMR aging management review 

ANS American Nuclear Society 

ANSI American National Standards Institute 

APCSB Auxiliary Power Conversion System Branch 

ASCE American Society of Civil Engineers 

ASM American Society for Metals 

ASME American Society of Mechanical Engineers 

CFR Code of Federal Regulations

CLB current licensing basis/bases 

DOE Department of Energy 

EOC End(s) of Cycle 

EPRI Electric Power Research Institute 

ER Environmental Report (Applicant’s Environmental Report Operating License 

Renewal Stage) 

FOSAR foreign object search and retrieval 

FR Federal Register

GALL Generic Aging Lessons Learned [Report] 

GEIS Generic Environmental Impact Statement 

GL generic letter 

HTP High Thermal Performance 

I&C instrumentation and control(s) 

IEEE Institute of Electrical and Electronics Engineers 

ksi kilogram(s) per square inch 

MeV mega electron-volts 

MEB metal-enclosed bus 

MRP Materials Reliability Program 

N/cm2 newton(s) per square centimeter 

NEI Nuclear Energy Institute 

NRC U.S. Nuclear Regulatory Commission 

NUMARC Nuclear Management and Resources Council 

OE operating experience 

PTLR pressure-temperature limits report 
viii 

PWR pressurized-water reactor 

RAI request for additional information 

RCP reactor coolant pump 

RCPB reactor coolant pressure boundary 

RFO refueling outage 

RG regulatory guide 

RVI reactor vessel internals 

SCC stress-corrosion cracking 

SE safety evaluation 

SER safety evaluation report 

SQN Sequoyah Nuclear Plant 

SSER supplemental safety evaluation report 

TAC Technical Assignment Control 

TR Technical Report 

TS Technical Specification 

TVA Tennessee Valley Authority 

UFSAR updated final safety analysis report 

WCAP Westinghouse Commercial Atomic Power 
1-1 

1SECTION 1 

INTRODUCTION AND GENERAL DISCUSSION 

1.1 Introduction 

This document is the first supplement to NUREG 2181, “Safety Evaluation Report Related to the 

License Renewal of Sequoyah Nuclear Plant (Sequoyah or SQN), Units 1 and 2, for the license 

renewal application (LRA) for Sequoyah Nuclear Plant (Sequoyah or SQN), Units 1 and 2,” as 

filed by Tennessee Valley Authority (TVA or the applicant). By letter dated January 7, 2013, 

TVA submitted its application to the U.S. Nuclear Regulatory Commission (NRC) for renewal of 

the Sequoyah, Units 1 and 2, operating licenses for an additional 20 years. The NRC staff (the 

staff) prepared this report to summarize the results of its safety review of the LRA for 

compliance with Title 10, Part 54, “Requirements for renewal of operating licenses for nuclear 

power plants,” of the Code of Federal Regulations (10 CFR Part 54). This information includes 

information committed to by TVA as documented in Commitment No. 28 (pertaining to reactor 

vessel internals), information required by 10 CFR 54.21(b), updated information and 

commitments in response to the recent industry operating experience, as well as information 

provided in response to staff requests for additional information. This document discusses only 

the changes to the safety evaluation report (SER). Unless otherwise noted, the staff reviewed 

and considered information submitted through August 28, 2015. This supplemental safety 

evaluation report (SSER) supplements portions of SER Sections 2, 3, and Appendices A 

through D. 

In accordance with 10 CFR 54.21(b), the NRC requires that, each year following submission of 

the LRA and at least 3 months before the scheduled completion of the staff’s review, the 

applicant submit an LRA amendment identifying any current licensing basis (CLB) changes to 

the facility that materially affect the contents of the LRA, including the updated final safety 

analysis report supplement. With annual updates dated April 22, 2014, August 17, 2015, and 

August 28, 2015, the applicant submitted an LRA update which summarizes the CLB changes 

that have occurred during the staff’s review of the LRA. This submission satisfies 10 CFR 

54.21(b) requirements up to the publication of this document. For this supplement, the staff 

reviewed the information in the latest LRA annual update and found all proposed CLB changes 

to be acceptable or have no effect on the staff’s findings. 

The staff does not have any further changes or updates to this section of the SER. 

2-1 

2SECTION 2

STRUCTURES AND COMPONENTS SUBJECT TO AGING 

MANAGEMENT REVIEW 

2.5 Scoping and Screening Results: Electrical and Instrumentation and 

Control Systems 

2.5.1 Electrical and Instrumentation and Controls Commodity Groups 

2.5.1.1 Summary of Technical Information in the Application 

The staff’s summary of technical information in the application evaluation is as documented in 

subsection in the “Summary of Technical Information in the Application” subsection 2.5.1.1 of 

NUREG-2181. 

2.5.1.2 Staff Evaluation 

The staff’s evaluation of the applicant’s electrical and instrumentation and control (I&C) systems 

is as documented in the “Staff Evaluation” subsection of Section 2.5.1.2 of NUREG 2181. This 

evaluation remains the same with the exception of the supplemental information that follows. 

The staff reviewed LRA Section 2.5, the second annual update to license renewal application 

(LRA) Section 2.5, and updated final safety analysis report (UFSAR) Chapters 7 and 8 using the 

evaluation methodology described in SER section 2.5 and the guidance in SRP-LR Section 2.5, 

“Scoping and Screening Results: Electrical and Instrumentation and Controls Systems.” 

By letter dated August 14, 2015 (ADAMS Accession No. ML15226A592), the applicant provided 

a second annual update to the SQN LRA Section 2.5 regarding the removal of penetrations 

SQN-1-PENE-302-0026-S and SQN-1-PENE-302-0040-S from the EQ program. The staff held 

a conference call with the applicant on August 20, 2015, to clarify the license renewal scoping of 

electrical components associated with the referenced non-EQ penetrations. Based on the 

staff’s discussions with the applicant, the staff understands that: 1) SQN electrical and I&C 

penetration assemblies in the EQ program (10 CFR 50.49) are not subject to aging 

management review (AMR), 2) electrical components associated with non-EQ penetrations 

(such as SQN-1-PENE-302-0026-S and SQN-1-PENE-302-0040-S) are within the scope of 

license renewal, 3) all non-EQ cables and connections associated with both EQ and non-EQ 

electrical and I&C penetration assemblies are evaluated in the insulated cables and connections 

commodity group that is subject to AMR as identified in Table 2.5-1, and 4) the pressure 

boundary function of non-EQ penetration assemblies are evaluated in Section 2.4.1. By letter 

dated August 28, 2015, the applicant supplemented the second annual update to the SQN LRA 

Section 2.5 to reflect these clarifications. Therefore, since cables and connections associated 

with the referenced non-EQ penetrations are within the scope of license renewal and subject to 

AMR, the staff’s concern is resolved. 
2-2 

2.5.1.3 Conclusion 

The staff’s conclusion for the scoping and screening results for electrical and I&C systems is 

documented in the “Scoping and Screening Results: Electrical and Instrumentation and Control 

Systems” subsection 2.5.1.3 of NUREG 2181. The staff does not have any changes or updates 

to its conclusion.
3-1 

3SECTION 3

AGING MANAGEMENT REVIEW RESULTS 

This section of the supplemental safety evaluation report (SSER) evaluates aging management 

programs (AMPs) and aging management reviews (AMRs) for Sequoyah Nuclear Plant (SQN), 

by the staff of the U.S. Nuclear Regulatory Commission (NRC) (the staff). 

The staff’s supplemental evaluation for the Reactor Vessel Internals Program and the Reactor 

Vessel Surveillance Program are detailed below. There were no updates to AMRs. 

3.0 Applicant’s Use of the Generic Aging Lessons Learned Report 

3.0.3 Aging Management Programs 

3.0.3.2 AMPs Consistent with the GALL Report with Exceptions or Enhancements 

3.0.3.2.17 Reactor Vessel Internals Program 

The summary of the Reactor Vessel Internals Program (License Renewal Application (LRA) 

Section B.1.34) and the staff’s evaluation of the Reactor Vessel Internals Program are as 

documented in Section 3.0.3.2.17 of NUREG 2181, “Safety Evaluation Report Related to the 

License Renewal of Sequoyah Nuclear Plant Units 1 and 2,” as documented in Agencywide 

Documents Access and Management System (ADAMS) Accession No. ML15187A206). This 

evaluation supplements the evaluation in the final SER Section 3.0.3.2.17 and provides an 

updated operating experience (OE) evaluation of the recent reactor vessel capsule event that 

occurred in Unit 1 during the spring 2015 refueling outage (RFO) for the unit. 

Summary of Technical Information in the Application. The summary of the technical information 

for the applicant’s Reactor Vessel Internals Program is as documented in the “Summary of 

Technical Information” subsection in Section 3.0.3.2.17 of NUREG-2181. 

Staff Evaluation. The staff’s evaluation of the applicant’s Reactor Vessel Internals Program is 

as documented in the “Staff Evaluation” subsection of Section 3.0.3.2.17 of NUREG-2181, with 

the exception of the supplemental OE that follows. 

Operating Experience. As discussed in the applicant’s letter dated July 10, 2015, SQN, Unit 1, 

inservice inspections of the reactor vessel internals (RVI) components during the End of Cycle 

(EOC) 20 outage revealed that two reactor vessel surveillance capsules had become dislodged 

from their basket holders. The inspections also revealed that capsule pieces or specimens from 

at least one of these capsules had become loose inside the Unit 1 reactor vessel. Tennessee 

Valley Authority (TVA or the applicant) noted damage in some cases to some RVI component. 

The staff noted that any damage to RVI components would need to be assessed in order to 

demonstrate that stress profiles for the damaged components would remain bounded by those 

assumed for the as built component configurations in the Electric Power Research Institute 

(EPRI) Materials Reliability Program (MRP) Technical Report (TR) MRP 227 A or the EPRI 

MRP reports used to develop the inspection and evaluation guidelines in TR MRP 227 A. The 

staff also noted that any damage that occurred from the impacts of the loose parts (e.g., causing 
3-2 

cold work to the affected components) could potentially create a preferential site for the initiation 

of further degradation effects, such as stress corrosion cracking (SCC), during the period of 

extended operation. Therefore, the staff concluded that the impact of this OE event would need 

to be evaluated further by the applicant for any impacts on the assumptions used to develop the 

inspection and evaluation criteria in MRP 227 A. 

By letter dated June 22, 2015, the staff issued Request for Additional Information (RAI) 

3.0.3.2.17 1a, Parts 1 and 2, requesting that the applicant address these issues. In RAI 

3.0.3.2.17 1a, Part 1, the staff asked the applicant to provide an adequate technical justification 

for not considering the potential effects of loose part induced cold work on the likelihood of 

degradation of the damaged RVI, in particular the long term prospects for initiation of SCC in the 

damaged internals. In RAI 3.0.3.2.17 1a, Part 2, the staff asked the applicant to provide a basis 

for not performing subsequent reinspections of the damaged locations of the Unit 1 RVI 

components using inspection methods that will effectively identify impact related damage. 

The applicant responded to RAI 3.0.3.2.17 1a, Parts 1 and 2, in a letter dated August 3, 2015. 

In its response to RAI 3.0.3.2.17 1a, Part 1, the applicant stated that any light scratches or rub 

marks in RVI components that were impacted by loose parts during plant operations after the 

capsule dislodged event would not be cold worked sufficiently to change the stress profiles of 

the components during plant operations. The applicant also indicated that, for components with 

deep rubs, the applicant’s analysis sufficiently demonstrated that either the components were in 

areas of low stress due to their design locations and configurations, or the stress levels of the 

components were less than 5 ksi. The applicant further stated that, based on these 

observations and evaluations, it concluded that (a) potential loose part induced surface cold 

work is not expected to have a significant impact on the affected RVI components, and (b) the 

assumptions used in the screening and expert panel review of MRP 191 and the inspection and 

evaluation guidance of MRP 227 A still remain valid for the SQN Unit 1 RVI components. 

In its review, the staff noted that the applicant’s OE review provides sufficient demonstration that 

any loose part impacts on the RVI components have not appreciably changed the component 

conditions from those evaluated in the design basis for those RVI components. The staff finds 

that the applicant has provided an acceptable basis to conclude that the inspection and 

evaluation criteria in MRP 227 A remain valid because the applicant has demonstrated that 

there has not been sufficient damage to the RVI components such that operational stress 

profiles for the components would no longer be bounded by those assumed in MRP 227 A, or in 

the background MRP reports used to develop MRP 227 A. 

In its response to RAI 3.0.3.2.17 1a, Part 2, the applicant stated that it would continue to 

implement foreign object search and retrieval (FOSAR) inspections for further loose parts that 

may have resulted from the surveillance capsule failure event. The applicant also stated that it 

retrieved any loose parts, which were detected as a result of the past FOSAR inspections 

subsequent to the capsule failure event, using gripping tools or by vacuum into a debris basket. 

The applicant further stated that, since the startup of Unit 1 following the capsule dislodged 

event, the loose part monitoring activities have not identified any indication of loose parts in the 

Unit 1 reactor vessel. 

In addition, the applicant stated that, if Foreign Material Exclusion inspections in subsequent 

Unit 1 RFOs find loose parts, the findings will be entered into the corrective action program for 

further evaluation of potential impacts to the Reactor Vessel Internals Program. The staff finds 

this to be an acceptable basis for resolving any questions regarding potential undetected loose 

parts from this event because the staff has confirmed that the applicant’s basis is consistent 
3-3 

with implemented Operating Experience Commitment 37 and with LRA Section A.1, Appendix 

A. RAI 3.0.3.2.17 1a, Parts 1 and 2, are resolved. 

Based on this review, the staff concludes that the applicant has adequately evaluated and 

resolved the impact of the reactor vessel surveillance capsule dislodged event on the program 

element criteria for the Reactor Vessel Internals Program and that the applicant has provided 

sufficient demonstration that the assumptions and criteria in MRP 227 A remain valid with 

respect to this OE event. 

UFSAR Supplement. The staff’s evaluation of the applicant’s updated final safety analysis 

report (UFSAR) supplement of Reactor Vessel Internals Program (LRA Appendix A, 

Section A.1.34) is as documented in “UFSAR Supplement” subsection in Section 3.0.3.2.17 of 

NUREG-2181. 

Conclusion. On the basis of its audit and review of the applicant’s Reactor Vessel Internals 

Program, the staff determined that the program elements for which the applicant claimed 

consistency with the Generic Aging Lessons Learned (GALL) Report are consistent. Also, the 

staff reviewed the enhancements and confirmed that their implementation prior to the period of 

extended operation will make the AMP adequate to manage the applicable aging effects. The 

staff concludes that the applicant has demonstrated that the effects of aging will be adequately 

managed so that the intended function(s) will be maintained in a way consistent with the current 

licensing basis (CLB) for the period of extended operation, as required by Title 10 of the Code of 

Federal Regulations (10 CFR) section 54.21(a)(3). The staff also reviewed the updated 

UFSAR supplement for this AMP and concludes that it provides an adequate summary 

description of the program, as required by 10 CFR 54.21(d). 

The staff also concludes that the applicant has properly evaluated the impacts of the reactor 

pressure vessel surveillance capsule dislodge event and the assumptions used to develop the 

criteria in the MRP-227-A report, as applicable to the design of RVI components at Units 1 and 

2. The applicant has demonstrated that the inspection and evaluation activities will remain valid 

for Units 1 and 2 with respect to the OE event during the period of extended operation. 

3.0.3.2.18 Reactor Vessel Surveillance Program 

Supplement to SER Section 3.0.3.2.18, “Reactor Vessel Surveillance Program.” 

The staff’s evaluation of the applicant’s Reactor Vessel Surveillance Program (LRA Section 

B.1.35) is documented in Section 3.0.3.2.18, “Reactor Vessel Surveillance Program,” of 

NUREG-2181, “Safety Evaluation Report Related to the License Renewal of Sequoyah Nuclear 

Plant Units 1 and 2,” July 31, 2015 (ADAMS Accession No. ML15187A206). Following the Unit 

1 event in which two reactor vessel surveillance capsules were dislodged from capsule holders 

and were damaged during the Cycle 20 operation (November 2013 – April 2015), the applicant 

revised the Reactor Vessel Surveillance Program (LRA Section B.1.35), as described below. 

The following safety evaluation supplements the staff’s evaluation regarding the Reactor Vessel 

Surveillance Program that is documented in Section 3.0.3.2.18 of NUREG-2181. 

Summary of Technical Information in the Application. The summary of the technical information 

for the applicant’s program is as documented in the “Summary of Technical Information” 

subsection in Section 3.0.3.2.18 of NUREG-2181, with the exception that follows. 
3-4 

In its letter dated July 10, 2015, the applicant revised Enhancement 2 of the applicant’s program 

to include an updated reference that addresses changes to the Unit 1 reactor vessel 

surveillance capsule withdrawal schedule (ADAMS Accession No. ML15197A176). The 

applicant’s letter indicates that the revision to Enhancement 2 results from the Unit 1 

surveillance capsules event. 

Staff Evaluation. The staff’s evaluation of the applicant’s program is as documented in the 

“Staff Evaluation” subsection in Section 3.0.3.2.18 of NUREG-2181, with the exception of 

Enhancement 2 as described below. 

Enhancement 1. The summary of Reactor Vessel Surveillance Program, Enhancement 1, for 

the applicant’s program is as documented in the “Enhancement 1” subsection in 

Section 3.0.3.2.18 of NUREG-2181. 

Enhancement 2. In its letter dated July 10, 2015, the applicant indicated that Enhancement 2 of 

the program is revised as a result of the following Unit 1 event regarding dislodged surveillance 

capsules. 

During the Unit 1 End of Cycle (EOC) 19 outage, the applicant relocated Capsule S to a higher 

fluence location (40-degree azimuthal location) in accordance with the staff-approved capsule 

withdrawal schedule documented in the safety evaluation dated September 27, 2013 (ADAMS 

Accession No. ML13240A320). During the EOC 19 outage, Capsule W (standby capsule) was 

also relocated to another higher fluence location (220-degree azimuthal location) for potential 

future use. During the EOC 20 outage, applicant’s inspections revealed that Capsules S and W 

of Unit 1 had been dislodged from their designated capsule basket holders and the materials 

test specimens contained in the capsules had been damaged. 

The applicant also indicated that, since these surveillance specimens were no longer available 

to provide neutron embrittlement data, a request for revision to the Unit 1 capsule withdrawal 

schedule was submitted to the NRC by letter dated May 14, 2015 (ADAMS Accession No. 

ML15134A377). In its proposed revision to the Unit 1 surveillance capsule withdrawal schedule, 

the applicant’s plan is to relocate Unit 1 Capsule V to a higher fluence location (140-degree 

azimuthal location) during the EOC 21, 22, or 23 outage. In its plan, the applicant will also 

withdraw the relocated capsule during the EOC 31 outage. The applicant further indicated that 

the proposed revision to the capsule withdrawal schedule is consistent with GALL Report AMP 

XI.M31 because the fast neutron fluence for capsule withdrawal is between one and two times 

the peak reactor vessel wall neutron fluence of Unit 1 (2.66x1019 n/cm2; E > 1 MeV) projected 

at the end of the period of extended operation. For example, the applicant indicated that, if Unit 

1 Capsule V is relocated during the EOC 23 outage and withdrawn during the EOC 31 outage, 

the fast neutron fluence for the capsule withdrawal will be 2.97x1019 n/cm2 (E > 1 MeV), 

consistent with the guidance in the GALL Report. In addition, the applicant indicated that Unit 1 

Capsule Z will not be relocated and will remain as a standby capsule in the reactor vessel. 

In its letter dated July 10, 2015, the applicant revised Enhancement 2 to include an updated 

reference to the May 14, 2015, submittal which addresses the proposed revision to the Unit 1 

capsule withdrawal schedule. The applicant also revised Commitment No. 28, Item B 

associated with Enhancement 2 and the UFSAR supplement (LRA Section A.1.35), consistent 

with the revision to Enhancement 2. 

In addition, the applicant’s July 10, 2015, letter states that the direct cause of this event is 

installation errors and installation procedural inadequacies. By letter dated August 3, 2015, the 
3-5 

applicant supplemented the July 10, 2015, letter indicating that it planned to perform visual 

inspections on the two Unit 2 surveillance capsules that were relocated during the Unit 2 EOC 

20 outage in Fall 2015 (ADAMS Accession No. ML15215A656). The applicant also indicated 

that these inspections will be performed to confirm that the relocated surveillance capsules are 

adequately recessed and seated into the groove in the top region of the capsule holder. The 

applicant further stated that the root cause analysis for the Unit 1 event did not identify any 

potential aging effect as a direct or contributing cause. The applicant also provided the 

following inspection results to support the conclusion of the root cause analysis. 

 The two dislodged Unit 1 capsules (Capsules S and W) are those that had been 

relocated in the Unit 1 reactor vessel during the prior outage (EOC 19 outage). The 

two remaining Unit 1 capsules (Capsules V and Z), which had not been relocated in 

the reactor vessel during the EOC 19 outage, remain intact and secured. 

 A visual inspection was conducted on the accessible areas of the two remaining Unit 

1 capsules and their respective capsule holders. The inspections revealed no signs 

of degradation. 

 A visual inspection was also conducted on the external surfaces of the capsule 

holders from which the two Unit 1 capsules had been dislodged. The inspections 

revealed no signs of degradation. 

In its review, the staff finds that Enhancement 2 as revised by letter dated July 10, 2015, is 

acceptable because (1) the revised enhancement adequately includes an updated reference to 

the May 14, 2015, submittal that requests NRC review of revision to the Unit 1 capsule 

withdrawal schedule in accordance with 10 CFR Part 50, Appendix H; and (2) the applicant 

confirmed that the proposed revision to the Unit 1 capsule withdrawal schedule is consistent 

with the recommendation in GALL Report AMP XI.M31 regarding the capsule fluence range for 

the period of extended operation. In addition, the staff approved the applicant’s request for 

revision to the Unit 1 capsule withdrawal schedule in accordance with 10 CFR Part 50, 

Appendix H, as documented in the NRC letter dated September 4, 2015 (ADAMS Accession 

No. ML15244B222). 

Enhancement 3. The summary of Reactor Vessel Surveillance Program, Enhancement 1, for 

the applicant’s program is as documented in the “Enhancement 3” subsection in 

Section 3.0.3.2.18 of NUREG-2181. 

Operating Experience. The staff’s evaluation of the operating experience regarding the 

applicant’s program is as documented in the “Operation Experience” subsection in Section 

3.0.3.2.18 of NUREG-2181. 

UFSAR Supplement. The staff’s evaluation of the applicant’s UFSAR supplement of Reactor 

Vessel Surveillance Program (LRA Section A.1.35) is as documented in the “UFSAR 

Supplement” subsection in in Section 3.0.3.2.18 of NUREG-2181, with the exception that 

follows. 

As previously discussed in the evaluation section for Enhancement 2 above, the applicant 

revised the UFSAR supplement by letter dated July 10, 2015. The staff finds the revised 

UFSAR supplement acceptable because it adequately includes an updated reference to the 
3-6 

May 14, 2015, submittal which addresses the proposed revision to the Unit 1 capsule 

withdrawal schedule 

Conclusion. On the basis of its audit and review of the applicant’s Reactor Vessel Surveillance 

Program, the staff determined that the program elements for which the applicant claimed 

consistency with the GALL Report are consistent. The staff also reviewed the enhancements 

and confirmed that their implementation prior to the period of extended operation will make the 

AMP adequate to manage the applicable aging effects. The staff concludes that the applicant 

has demonstrated that the effects of aging will be adequately managed so that the intended 

function(s) will be maintained in a way consistent with the CLB for the period of extended 

operation, as required by 10 CFR 54.21(a)(3). In addition, the staff reviewed the updated 

UFSAR supplement for this AMP and concludes that it provides an adequate summary 

description of the program, as required by 10 CFR 54.21(d). 
 5-1 

4SECTION 5 

REVIEW BY THE ADVISORY COMMITTEE ON 

REACTOR SAFEGUARDS

In accordance with Title 10 of the Code of Federal Regulations Part 54, “Requirements for 

renewal of operating licenses for nuclear power plants,” the Advisory Committee on Reactor 

Safeguards (ACRS) performed a detailed review of the license renewal application for 

Sequoyah Nuclear Plant, Units 1 and 2. The ACRS issued a report discussing the results of the 

review which was enclosed in Section 5 of NUREG-2181. This supplemental safety evaluation 

report was submitted to the ACRS for review, and the ACRS had no further comments 

regarding the supplemental information and evaluation performed by the staff. 

6-1 

5SECTION 6 

CONCLUSION 

The staff concludes that the additional information provided by Tennessee Valley Authority does 

not alter the conclusion proffered in the safety evaluation report issued in January 2015 and that 

the requirements of Title 10 of the Code of Federal Regulations section 54.29(a) have been 

met.

A-1 

6APPENDIX A

SEQUOYAH NUCLEAR PLANT, UNITS 1 AND 2, LICENSE 

RENEWAL COMMITMENTS 

During the review of the Sequoyah Nuclear Plant (Sequoyah, SQN), Units 1 and 2, license 

renewal application (LRA) by the staff of the U.S. Nuclear Regulatory Commission, Tennessee 

Valley Authority (TVA, the applicant) made commitments related to aging management 

programs to manage the aging effects of structures and components. During the period of 

review for this supplemental safety evaluation report, one commitment was revised. The 

following table lists the updated commitment along with the implementation schedules and 

sources for the commitment. 
A-2 

Table A-1 Sequoyah License Renewal Commitments 

Item 

Number 

Commitment UFSAR

Supplement 

Section/LRA 

Section 

Enhancement or

Implementation 

Schedule 

Source

28 A. Revise Reactor Vessel Surveillance Program procedures to 

consider the area outside the beltline, such as nozzles, penetrations and discontinuities, to determine if more restrictive P-T limits are required than would be determined by just considering the reactor vessel beltline materials. B1. Revise Unit 2 Reactor Vessel Surveillance Program procedures to incorporate an NRC-approved schedule for capsule withdrawals to meet ASTM E185-82 requirements, including the possibility of operation beyond 60 years (refer to the TVA Letter to NRC, “Sequoyah Reactor Pressure Vessel Surveillance Capsule Withdrawal Schedule Revision Due to License Renewal Amendment,” dated 01/10/13, ADAMS Accession No. ML13032A251; NRC final safety evaluation report approved on 09/27/13, ADAMS Accession No. ML13240A320). B2. Revise Unit 1 Reactor Vessel Surveillance Program procedures to incorporate an NRC-approved schedule for capsule withdrawals to meet ASTM E185-82 requirements, including the possibility of operation beyond 60 years (refer to the TVA Letter to NRC, “Sequoyah Nuclear Plant, Revision to Reactor Pressure Vessel Surveillance Capsule Withdrawal Schedule for License Renewal,” dated May 14, 2015). C. Revise Reactor Vessel Surveillance Program procedures to withdraw and test a standby capsule to cover the peak fluence expected at the end of the period of extended operation. 

B.1.35 SQN1: Prior to 

03/17/2020 SQN2: Prior to 03/15/2021 

Letter 

ML13190A276 

(7/1/13) 

Letter 

ML15197A176 

(7/10/15) 

This Commitment Revision supersedes all previous versions. 
B-1 

7APPENDIX B 

CHRONOLOGY

This appendix lists chronologically the routine licensing correspondence between the staff of the 

U.S. Nuclear Regulatory Commission (NRC) (the staff) and the Tennessee Valley Authority 

(TVA, the applicant) for Sequoyah Nuclear Plant (SQN), Units 1 and 2, during the period of 

review for this supplemental safety evaluation report. This appendix also lists other 

correspondence regarding the staff’s review of the license renewal application (LRA) 

(under Docket Nos. 50-327 and 50-328). 

Table B-1 Chronology 

Date Accession No.a or 

Federal Register No.

Subject 

9/27/2013 ML13240A320 Letter to Mr. Joseph W. Shea, TVA, “Sequoyah Nuclear Plant, Units 1 and 

2 - Revise the Reactor Pressure Vessel Material Surveillance Capsule 

Withdrawal Schedule Due to License Renewal Amendment (TAC 

Nos. MF0631 and MF0632).” 

5/14/2015 ML15134A377 Letter from Mr. Joseph W. Shea, TVA, “Sequoyah, Units 1 and 2 - Revision 

to the Reactor Pressure Vessel Surveillance Capsule Withdrawal Schedule 

for License Renewal.” 

6/22/2015 ML15159B164 Letter to Mr. Joseph W. Shea, TVA, “Requests for Additional Information 

for the Review of the Sequoyah Nuclear Plant, Units 1 and 2, License 

Renewal Application – Set 25 (TAC Nos. MF0481 and MF0482).” 

6/23/2015 ML15152A031 Summary of Telephone Conference Call Held on May 13, 2015, between 

the U.S. NRC and TVA, concerning RAI, Set 24 Pertaining to the 

Sequoyah Nuclear Plant License Renewal Application. (TAC Nos. MF0481 

and MF0482) 

7/10/2015 ML15197A176 Letter from Mr. Joseph W. Shea, TVA, “Sequoyah, Units 1 and 2 - Revision 

to Commitment No. 28 and Review of Impacts to the SQN Reactor Vessel 

Internals Aging Management Program Due to Dislodged Reactor Vessel 

Surveillance Capsules in Unit 1 Reactor.” 

8/3/2015 ML15215A647 Letter from Mr. Joseph W. Shea, TVA, “Sequoyah Nuclear Plant, Units 1 

and 2 – Response to NRC Request for Information regarding the Review 

of the License Renewal Application, Set 25 (TAC Nos. MF0481 and 

MF0482).” 

8/3/2015 ML15215A656 Letter from Mr. Joseph W. Shea, TVA, "Supplement to TVA Letter, 

“Sequoyah Nuclear Plant - Revision to Commitment No. 28 and Review of 

Impacts to the SQN Reactor Vessel Internals Aging Management Program 

Due to Dislodged Reactor Vessel Surveillance Capsules in Unit 1 Reactor.” 

8/17/2015 ML15226A592 Letter from Mr. Joseph W. Shea, TVA, “Second Annual Update to the 

Sequoyah Nuclear Plant, Units 1 and 2, License Renewal Application (TAC 

Nos. MF0481 and MF0482).” 

8/28/2015 ML15240A145 Letter from Mr. Joseph W. Shea, TVA, “Supplement to Second Annual 

Update to the Sequoyah Nuclear Plant, Units 1 and 2, License Renewal 

Application (TAC Nos. MF0481 and MF0482).” 
B-2 

Date Accession No.a or 

Federal Register No.

Subject 

8/28/2015 ML15236A067 Summary of Telephone Conference Call Held on August 20, 2015, 

between the U.S. NRC and TVA, concerning the August 17, 2015 Annual 

Update to the Sequoyah Nuclear Plant License Renewal Application. (TAC 

Nos. MF0481 and MF0482) 

9/1/2015 ML15233A035 Summary of Telephone Conference Call Held on July 20, 2015, between 

the U.S. NRC and TVA, concerning RAI, Set 25 Pertaining to the 

Sequoyah Nuclear Plant License Renewal Application. (TAC Nos. MF0481 

and MF0482) 

9/4/2015 ML15238A342 Summary of Telephone Conference Call Held on July 22, 2015, between 

the U.S. NRC and TVA, concerning RAI, Set 25 Pertaining to the 

Sequoyah Nuclear Plant License Renewal Application. (TAC Nos. MF0481 

and MF0482) 

9/24/2015 ML15244B222 Letter to Mr. Joseph W. Shea, TVA, “Sequoyah Nuclear Plant, Units 1 and 

2 – Revision To The Unit 1 Reactor Vessel Surveillance Capsule Schedule 

For License Renewal (TAC NOS. MF6620 AND MF6621)” 

a

 Accession numbers can be used to find documents in the NRC’s Agencywide Documents Access and Management System. 
C-1 

8APPENDIX C 

PRINCIPAL CONTRIBUTORS 

This appendix lists the principal contributors for the development of this supplemental safety 

evaluation report and their areas of responsibility. 

Table C-1 Principal Contributors 

Name Responsibility

Buford, Angela Reviewer – Structural 

Diaz-Sanabria, Yoira Management Oversight 

Doutt, Cliff Reviewer – Electrical 

Foli, Adakou Reviewer – Electrical 

Lubinski, John Management Oversight 

Marshall, Michael Management Oversight 

Medoff, Jim Reviewer – Reactor Systems 

Miller, Chris Management Oversight 

Min, Seung Reviewer – Reactor Systems 

Morey, Dennis Management Oversight 

Purtcher, Patrick Reviewer – Reactor Vessel Internals 

Sayoc, Emmanuel Project Management 

Wittick, Brian Management Oversight 

Zimmerman, Jacob Management Oversight 

D-1 

9APPENDIX D 

REFERENCES 

This appendix lists the references used throughout this supplemental safety evaluation report 

for review of the license renewal application (LRA) for Sequoyah Nuclear Plant (SQN), Units 1 

and 2. 

U.S. Nuclear Regulatory Commission (NRC) and Tennessee Valley Authority (TVA) Documents for SQN 

NRC Letter – “Request for Publication in Biweekly FR Notice - Notice of Issuance of Amendments to Facility 

Operating Licenses (TAC Nos. ME6538 and ME6539),” September 26, 2012 (ADAMS Accession 

No. ML12249A426) 

Report – “Sequoyah Nuclear Plant (SQN) Unit 2 Reactor Vessel Surveillance Capsule Y Analysis Summary 

Report,” March 8, 2000 (ADAMS Accession No. ML003691809) 

TVA Letter – “Sequoyah License Amendment Request,” December 15, 2011 

TVA Letter – “Sequoyah Nuclear Plant, Units 1 and 2 Issuance of Amendments To Revise the Technical 

Specification To Allow Use of AREVA Advanced W17 High Thermal Performance Fuel (TS-SQN-2011-07) (TAC 

Nos. ME6538 and ME6539),” September 26, 2012 (ADAMS Accession No. ML12249A394) 

TVA Letter – “Sequoyah Reactor Pressure Vessel Surveillance Capsule Withdrawal Schedule Revision Due to 

License Renewal Amendment,” January 10, 2013 (ADAMS Accession No. ML13032A251) 

UFSAR Update – “Sequoyah, Units 1 and 2, Updated Final Safety Analysis Report Amendment 23,” 

December 14, 2011 (ADAMS Accession No. ML11364A031) 
D-2 

Other NRC Documents by Type

GL 92-01, Revision 1, “Reactor Vessel Structural Integrity,” May 19, 1995 

NUREG-0011, “Safety Evaluation Report to Sequoyah Nuclear Plant, Units 1 and 2,” Supplements 1, 2, and 5 

NUREG-0588, “Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment,” 

July 1981 

NUREG-1437, “Generic Environmental Impact Statement for License Renewal of Nuclear Plants (GEIS),” 

including draft Supplement 53, “Generic Environmental Impact Statement for License Renewal of Nuclear 

Plants regarding Sequoyah Nuclear Plant Units 1 and 2,” July 31, 2014 

NUREG-1800, Revision 2, “Standard Review Plan for Review of License Renewal Applications for Nuclear Power 

Plants,” December 2010 

NUREG-1801, Revision 2, “Generic Aging Lessons Learned (GALL) Report,” December 2010 

NUREG-2181, "Safety Evaluation Report Related to the License Renewal of Sequoyah Nuclear Plant, Units 1 and 

2," July 31, 2015 (ADAMS Accession No. ML15187A206) 

RG 1.89, “Environmental Qualification of Certain Electric Equipment Important to Safety for Nuclear Power Plants,” 

June 1984 

RG 1.163, “Performance-Based Containment Leak-Test Program,” March 2013 

U.S. NRC RIS 2012-02, “Insights into Recent LRA Consistency with the Generic Aging Lessons Learned Report,” 

January 24, 2012 

SE, “Westinghouse Owners Group,” October 15, 2001 (ADAMS Accession No. ML012890230) 

SE, “Sequoyah Nuclear Plant, Units 1 and 2 – Issuance of amendment regarding technical specification change 

No. 00-14, pressure temperature limits report (TAC Nos. MB6436 and MB6437),” September 15, 2004 (ADAMS 

Accession No. ML042600465) 

SE, “Final Safety Evaluation of Electric Power Research Institute (EPRI) Report, Materials Reliability Program 

(MRP) Report 1016596 (MRP-227), Revision 0, ‘Pressurized Water Reactor (PWR) Internals Inspection and 

Evaluation Guidelines’ (TAC No. ME0680),” Revision 1, December 16, 2011 (ADAMS Accession 

No. ML11308A770) 

United States Legislation

National Environmental Policy Act of 1969 

Regulations

U.S. Code of Federal Regulations (CFR), “Agency Rules of Practice and Procedure,” Part 2, Title 10, “Energy” 

(10 CFR Part 2) 

CFR, “Requirements for Renewal of Operating Licenses for Nuclear Power Plants,” Part 54, Title 10, “Energy” 

(10 CFR Part 54) 

CFR, “Operators’ Licenses,” Part 55, Title 10, “Energy” (10 CFR Part 55) 

CFR, “Reactor Site Criteria,” Part 100, Title 10, “Energy” (10 CFR Part 100) 
D-3 

Industry Codes, Standards, and Other Documents, by Source

American Nuclear Society (ANS) 

ANS N18.2, “American National Standard Revision and Addendum to Nuclear Safety Criteria for the Design of 

Stationary Pressurized Water Reactor Plants,” 1973 

American National Standards Institute (ANSI) 

ANSI/ANS 56.8, “Containment System Leakage Testing Requirements,” 2002 

ANSI N18.2-1973, “Revision and Addendum to Nuclear Safety Criteria for the Design of Stationary Pressurized 

Water Reactor Plants,” 1973 

ANSI and American Society of Civil Engineers (ASCE) 

ANSI/ASCE 11-99, “Guideline for Structural Condition Assessment of Existing Buildings” 

AREVA 

AREVA ANP-2986, “Sequoyah HTP Fuel Transition Report,” June 2011 

American Society of Mechanical Engineers (ASME) 

ASME, Boiler and Pressure Vessel Code, Section III, “Rules for Construction of Nuclear Power Plant 

Components,” American Society of Mechanical Engineers, New York, NY 

ASME Code, Section VIII, “Rules for Construction of Pressure Vessels” 

ASME Code, Section XI, “Rules for Inservice Inspection of Nuclear Power Plant Components,” including 

2001 Edition with 2003 Addenda 

American Society for Metals (ASM) 

ASM Handbook, Volume 19, “Fatigue and Fracture,” 1996 

ASTM International 

ASTM E185-82, “Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power 

Reactor Vessels” 
D-4 

Electric Power Research Institute (EPRI) 

EPRI Report 1003057, “License Renewal Handbook” 

EPRI Report 1012987, Revision 2, “Steam Generator Integrity Assessment Guidelines,” 

EPRI Report 1013234, “Materials Reliability Program: Screening, Categorization, and Ranking of Reactor Internals 

Components for Westinghouse and Combustion Engineering PWR Design (MRP-191)” 

EPRI Report 1013706, “Pressurized Water Reactor Steam Generator Examination Guidelines” 

EPRI Report 1016596, “PWR Internals Inspection and Evaluation Guidelines (MRP-227, Revision 0)” 

EPRI Report 1016609, “Materials Reliability Program: Inspection Standard for PWR Internals – 2012 Update 

(MRP-228)” 

EPRI Report 1022863, “Materials Reliability Program: Pressurized Water Reactor Internals Inspection and 

Evaluation Guidelines (MRP-227-A)” 

Institute of Electrical and Electronics Engineers (IEEE) 

IEEE Std. 1406-1998, “IEEE Guide to the Use of Gas-In-Fluid Analysis for Electric Power Cable Systems” 

Nuclear Energy Institute (NEI) 

NEI 95-10, Revision 6, “Industry Guideline for Implementing the Requirements of 10 CFR Part 54 – The License 

Renewal Rule,” June 2005 

NEI 03-08, “Guideline for the Management of Materials Issues” 

Sandia National Laboratory 

Department of Energy (DOE) Aging Management Guide (AMG) contractor report SAND96-0344, “Aging 

Management Guideline for Commercial Nuclear Power Plants – Electrical Cable and Terminations” 

Tennessee Valley Authority (TVA) 

TVA Letter Response to Generic Letter 92-01, “Reactor Vessel Structural Integrity,” NUMARC recommended 

response guidelines utilized to extent practical,” July 7, 1992 (ADAMS Accession No. 9207090177) 

TVA response letter of November 18, 1999, to Revision 2 of the Reactor Vessel Integrity Database 

TVA Report Nos. PTLR-1, Revision 4, and PTLR-2, Revision 5 
D-5 

Westinghouse Electric Co. 

WCAP-10456, “Effects of Thermal Aging on the Structural Integrity of Cast Stainless Steel Piping for 

Westinghouse Nuclear Steam Supply Systems,” Westinghouse Proprietary Class 2, November 1983 

WCAP-13333, “Analysis of Capsule X from the Tennessee Valley Authority Sequoyah Unit No. 1 Reactor Vessel 

Radiation Surveillance Program,” June 30, 1992 

WCAP-13545, “Analysis Of Capsule X From The Tennessee Valley Authority Sequoyah Unit 2 Reactor Vessel 

Radiation Surveillance Program,” March 12, 1993 

WCAP-15224, “Analysis of Capsule Y from the Tennessee Valley Authority Sequoyah Unit 1 Reactor Vessel 

Radiation Surveillance Program,” June 1999 

WCAP-15320, “Analysis of Capsule Y from the Tennessee Valley Authority Sequoyah Unit 2 Reactor Vessel 

Radiation Surveillance Program,” March 2000 

WCAP-17539-(Non-Proprietary), Revision 0, “Sequoyah Units 1 and 2 Time-Limited Aging Analysis on Reactor 

Vessel Integrity” 





NUREG

-2181

Supplement 1

Safety Evaluation Report Related to the License Renewal of 

Sequoyah Nuclear Plant Units 1 and 2