Safety Evaluation Report

Related to the Operation of 

Watts Bar Nuclear Plant, Unit 2

Docket Number 50-391

Tennessee Valley Authority

Office of Nuclear Reactor Regulation

NUREG-0847

Supplement 29
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 Operation of 

Watts Bar Nuclear Plant, Unit 2

Docket Number 50-391

Tennessee Valley Authority

Manuscript Completed: October 2015

Date Published: October 2015

Office of Nuclear Reactor Regulation

NUREG-0847

Supplement 29 

iii

ABSTRACT

This report supplements the Staff’s evaluation of the application filed by the Tennessee Valley 

Authority (TVA), as applicant and owner, for a license to operate Watts Bar Nuclear Plant 

(WBN), Unit 2 (Docket No. 50-391). This supplemental evaluation documents the Staff’s 

findings with respect to the applicable elements of Title 10 of the Code of Federal Regulations

(10 CFR) 50.57, which lists the findings the Nuclear Regulatory Commission (NRC) must make 

to issue an operating license.

The application for a construction permit for WBN, Units 1 and 2 was submitted on May 14, 

1971, and constructions permits CPPR-91 and CPPR-92 were issued January 23, 1973. 

Subsequently, on October 4, 1976, an application for operating licenses for WBN, Units 1 and 2 

was submitted by TVA. In its safety evaluation report (SER) published June 1982 (Agencywide 

Documents Access and Management System (ADAMS) Accession No. ML072060491) and 

Supplemental SERs (SSERs) 1 through 20, issued by the Office of Nuclear Reactor Regulation 

(NRR) of the NRC, the NRC staff documented its safety evaluation (SE) and determination that 

WBN, Unit 1, met all applicable regulations and regulatory guidance. Based on satisfactory 

findings from all applicable inspections, on February 7, 1996, the NRC issued a full-power 

operating license (OL) to WBN, Unit 1, authorizing operation up to 100-percent power.

In SSER 21 published February 2009 (ADAMS Accession No. ML090570741), the NRC staff 

addressed TVA’s application for a license to operate WBN, Unit 2, and gave information on the 

status of the items remaining to be resolved that were outstanding at the time that TVA deferred 

construction of WBN, Unit 2, and were not evaluated and resolved as part of the licensing of 

WBN, Unit 1. SSERs 22 to 28 documented the NRC staff’s ongoing evaluation and closure of 

open items in support of TVA’s application for a license to operate WBN, Unit 2. 

This SSER documents the NRC staff’s completion of its review of open items in related to TVA’s 

application for an OL for WBN, Unit 2. In addition to closure of remaining open items, this 

SSER discusses a revision to the motor operated valve testing program, the fire protection 

program, the technical specifications, and final SER modifications made by the applicant. 

v 

TABLE OF CONTENTS

Page

ABSTRACT iii

TABLE OF CONTENTS v

ABBREVIATIONS ix

1 INTRODUCTION AND DISCUSSION 1-1

1.1 Introduction 1-1

1.7 Summary of Outstanding Issues 1-2

1.9 License Conditions 1-27

1.9.1 Flooding License Condition 1-28

1.9.2 Cyber Security License Condition 1-28

1.9.3 Core Operating Limits License Condition 1-28

1.9.4 Electrical Design License Condition 1-29

1.9.5 Fire Protection License Condition 1-29

1.9.6 Core Reload License Condition 1-29

1.13 Implementation of Corrective Action Programs and Special Programs 1-29

1.13.1 Corrective Action Programs 1-30

1.13.2 Special Programs 1-31

1.14 Implementation of Applicable Bulletin and Generic Letter Requirements 1-31

1.17 Financial Assurance for Decommissioning 1-46

1.18 NRC Staff Conclusion 1-47

3 DESIGN CRITERIA – STRUCTURE, COMPONENTS, EQUIPMENT, AND 

SYSTEMS 3-1

3.8 Design of Seismic Category I Structures 3-1

3.8.3 Other Seismic Category I Structures 3-1

3.9 Mechanical Systems and Components 3-1

3.9.3 ASME Code Class 1, 2, and 3 Components, Component Structures, 

and Core Support Structures 3-1

6 ENGINEERED SAFETY FEATURES 6-1

6.1 Engineered Safety Feature Materials 6-1

6.1.3 Postaccident Emergency Cooling Water Chemistry 6-1

6.2 Containment Systems 6-1

6.2.1 Containment Functional Design 6-1

6.2.2 Containment Heat Removal Systems 6-14

6.2.3 Secondary Containment Functional Design 6-16

6.2.4 Containment Isolation Systems 6-17

6.2.5 Combustible Gas Control Systems 6-18

6.2.6 Containment Leakage Testing 6-18

6.3 Emergency Core Cooling System 6-18

6.3.1 System Design 6-18

6.3.3 Testing 6-20

6.4 Control Room Habitability 6-20

6.5 Engineered Safety Feature (ESF) Filter Systems 6-20

6.5.1 ESF Atmosphere Cleanup System 6-20
vi

7 INSTRUMENTS AND CONTROLS 7-1

7.5 Safety-Related Display Instrumenation 7-1

7.5.2 Postaccident Monitoring System 7-1

7.7 Control Systems Not Required for Safety 7-3

7.7.1 System Description 7-3

8 ELECTRIC POWER SYSTEMS 8-1

8.3 Onsite Power Systems 8-1

 8.3.1 Onsite AC Power System Compliance with GDC 17 8-1

9 AUXILIARY SYSTEMS 9-1

9.2 Water Systems 9-1

9.2.1 Essential Raw Cooling Water and Raw Cooling Water System 9-1

9.2.2 Component Cooling System (Reactor Auxiliaries Cooling Water 

System) 9-1

9.2.6 Condensate Storage Facilites 9-1

9.5 Other Auxiliary Systems 9-2

9.5.1 Fire Protection 9-2

10 STEAM AND POWER CONVERSION SYSTEM 10-1

10.4 Other Features 10-1

10.4.9 Auxiliary Feedwater System 10-1

11 RADIOACTIVE WASTE MANAGEMENT 11-1

11.7 NUREG-0737 Items 11-1

11.7.1 Wide-Range Noble Gas, Iodine, and Particulate Effluent Monitors 

(II.F.1.1, II.F.1.2.a, and II.F.1.2.b) 11-1

11.7.2 Primary Coolant Outside Containment (III.D.1.1) 11-2

12 RADIATION PROTECTION 12-1

12.7 NUREG-0737 Items 12-1

12.7.1 Plant Shielding (II.B.2) 12-1

12.7.2 High Range In-Containment Monitor (II.F.1.2.c) 12-1

12.7.3 In-Plant Radioiodine Monitor (III.D.3.3) 12-1

15 ACCIDENT ANALYSES 15-1

15.3 Limiting Accidents 15-1

15-1

15.5 15-2

15.3.1 Loss-of-Coolant Accident (LOCA) 

NUREG-0737 Items 

15.5.5 Small-Break LOCA Methods (II.K.3.30) and Plant-Specific 

Calculations (II.K.3.31) 15-2 

16 TECHNICAL SPECIFICATIONS 16-1

16.1 Introduction 16-1

16.2 Evaluation 16-2

16.3 Conclusion 16-3

17 QUALITY ASSURANCE 17-1

17.6 Maintenance Rule 17-1
vii

22 FINANCIAL PROTECTION AND INDEMNITY REQUIREMENTS 22-1

22.3 Operating Licenses 22-1

APPENDIX A CHRONOLOGY OF RADIOLOGICAL REVIEW OF WATTS BAR 

NUCLEAR PLANT, UNIT 2, OPERATING LICENSE REVIEW A-1

APPENDIX E PRINCIPAL CONTRIBUTORS TO SSER 29 E-1 

APPENDIX FF FIRE PROTECTION PROGRAM SAFETY EVALUATION WATTS BAR 

NUCLEAR PLANT, UNITS 1 AND 2 FF-1

APPENDIX GG FINAL MEMORANDUM ON FACILITY COMPLETION IN ACCORDANCE 

WITH INSPECTION MANUAL CHAPTER 94302 GG-1

APPENDIX HH WATTS BAR, UNIT 2, ACTION ITEMS TABLE HH-1

ix

ABBREVIATIONS

3M Minnesota Mining and Manufacturing 

ABGTS auxiliary building gas treatment system

AC alternating current

ACR auxiliary control room

ACRS Advisory Committee for Reactor Safeguards

ADAMS Agencywide Documents Access and Management System

AFW auxiliary feedwater

ALARA as low as is reasonably achievable

ANSI American National Standards Institute

ANS American Nuclear Society

AOR analysis of record

APCSB Auxiliary Power Conversion Systems Branch (of NRR)

ASME American Society of Mechanical Engineers

ASTM American Society for Testing and Materials

AV analysis volume

BEACON Westinghouse Best Estimate Analyzer for Core Operations—Nuclear

BL bulletin

BOP balance of plant

BTP Branch Technical Position

BTU British thermal unit

CAP corrective action program

CCP centrifugal charging pump

CCS component cooling system

CECC Central Emergency Control Center

CFR Code of Federal Regulations

CI confirmatory issue

CO2 carbon dioxide

CP construction permit

CPU central processing unit

CRDR control room design review

CSP cyber security plan

CSST common station service transformer

CST condensate storage tank

DBA design-basis accident

DCN design change notice

DCS distributed control system 

DG diesel generator

DVR degraded voltage relay

EA Environmental Assessment

EAL emergency action level

ECCS emergency core cooling system

EDG emergency diesel generator

EMC electromagnetic compatibility

EMI electromagnetic interference

EOF emergency operations facility

EOP emergency operating procedure

EP emergency preparedness
x

EPA electrical penetration assemblies

EPIP emergency preparedness implementation procedure

EPRI Electric Power Research Institute

EPZ emergency planning zone

EQ environmental qualification

ERCW essential raw cooling water

ERDS emergency response data system

ERFBS electric raceway fire barrier system

ERO Emergency Response Organization

ESF engineered safety feature

ETE evacuation time estimate

FCV flow control valve

FEMA Federal Emergency Management Agency

FHA fuel-handling accident

FPR fire protection report

FSAR final safety analysis report

FSSD fire safe shutdown

GDC general design criterion/criteria

GL generic letter

gpm gallon per minute 

HAB hostile action based

HFT hot functional testing 

HPFP high-pressure fire protection

HRCAR high range containment air radiation

HVAC heating, ventilation, and air conditioning

ICM interim compensatory measures

IE Office of Inspection and Enforcement

IEB Office of Inspection and Enforcement Bulletin

IEEE Institute of Electrical and Electronics Engineers

IFR Interim Finding Report

IITA in-core instrumentation thimble assembly

IPEEE individual plant examination of external events

IPS intake pumping station

ISG interim staff guidance

JLD Japan Lessons-learned Directorate

kV kilovolt

kVA kilovolt ampere

LAR license amendment request

LOCA loss-of-coolant accident

LOOP loss-of-offsite power

LTOP low-temperature overpressure protection

LVR low voltage relay

M&E mass and energy

MBtu million British units

MCC motor control center

MCR main control room

MOV motor-operated valve

MSIV main steam isolation valve

MSO multiple spurious operation

NEI Nuclear Energy Institute

NFPA National Fire Protection Association
xi

NPG Nuclear Power Group

NPP nuclear performance plan

NPSH net positive suction head 

NPSHA net positive suction head available

NRC U.S. Nuclear Regulatory Commission

NRR Office of Nuclear Reactor Regulation

NSAL nuclear safety advisory letter

NTTF Near-Term Task Force

NUREG report prepared by NRC staff

OBE operating-basis earthquake

ODCM Offsite Dose Calculation Manual

ODS Operations Duty Specialist

OL operating license

OMA operator manual action

OSA on-shift staffing analysis

PAD performance analysis and design

PAMS post-accident monitoring system

PASS post-accident sampling system

PDR public document room

PORV power-operated relief valve

PPM performance prediction methodology

psi pressure per square inch 

psig pressure per square inch gauge

PRT pressurizer relief tank

PTLR Pressure and Temperature Limits Report

PWR pressurized-water reactor

QA quality assurance 

RAI request for additional information

RCP reactor coolant pump

RCS reactor coolant system

RCW raw cooling water 

REP Radiological Emergency Plan

RES radiant energy shield

RFI radio frequency interference

RG Regulatory Guide

RHR residual heat removal

RPV reactor pressure vessel

RWST refueling water storage tank

SE safety evaluation

SER safety evaluation report, NUREG-0847, dated June 1982

SFP spent fuel pool

SG steam generator

SIAS Safety Injection Actuation Signal

SPND self-powered neutron detector

SRM Staff Requirements Memorandum

SPS signal processing system

SRP Standard Review Plan, NUREG-0800

SRS software requirements specification

SSC structure, system, and component

SSE safe shutdown earthquake

SSEP safety, security, and emergency preparedness
xii

SSER Supplemental SER

Std. Standard

TCD thermal conductivity degradation

TI Technical Instruction

TMI Three Mile Island

TPS Transmission and Power Supply

TS technical specification

TSO transmission system operator

TSTF Technical Specification Task Force

TVA Tennessee Valley Authority

UFSAR Updated Final Safety Analysis Report

UL Underwriter’s Laboratories, Inc.

V volt

VAC volt alternating current

VCT volume control tank

VDC volt direct current

VPA ventilation and purge air

V&V verification and validation

WBA Web-based ADAMS

WBN Watts Bar Nuclear Plant

χ/Q atmospheric dispersion estimate
1-1

1 INTRODUCTION AND DISCUSSION

1.1 Introduction

The Watts Bar Nuclear Plant (WBN or Watts Bar) is owned by the Tennessee Valley Authority 

(TVA) and is located in southeastern Tennessee, approximately 50 miles (80 kilometers) 

northeast of Chattanooga. The facility consists of two Westinghouse-designed four-loop 

pressurized-water reactors (PWRs) within ice condenser containments.

In June 1982, the U.S. Nuclear Regulatory Commission (NRC) staff issued safety evaluation 

report (SER), NUREG-0847, “Safety Evaluation Report Related to the Operation of Watts Bar 

Nuclear Plant Units 1 and 2,” (Agencywide Documents Access and Management System 

(ADAMS) Accession No ML072060490) on TVA’s application for licenses to operate WBN,

Units 1 and 2. In SER Supplements (SSERs) 1 through 20, the NRC staff concluded that WBN,

Unit 1, met all applicable regulations and regulatory guidance, and on February 7, 1996, the 

NRC issued an operating license (OL) to WBN, Unit 1. TVA did not complete WBN, Unit 2. 

Accordingly, the NRC did not make a final approval or denial of the request for an operating 

license for WBN, Unit 2.

On March 4, 2009 (ADAMS Accession No. ML090700378), TVA submitted an updated 

application in support of its request for an OL for WBN, Unit 2, pursuant to Title 10 of the Code 

of Federal Regulations (10 CFR) Part 50, “Domestic Licensing of Production and Utilization 

Facilities.”

In SSER 21 which was published in February 2009 (ADAMS Accession No. ML090570741), the 

NRC staff reported on the WBN, Unit 2, items remaining to be resolved, which were outstanding 

at the time that TVA deferred construction of Unit 2, and which were not evaluated and resolved 

as part of the licensing of WBN, Unit 1. In SSERs 22 through 28, the NRC staff documented its 

evaluation and closure of open items in support of TVA’s application for a license to operate 

WBN, Unit 2. 

This SSER documents the NRC staff’s completion of its evaluation and closure of open items in 

response to TVA’s application.

The format of this document is consistent with the format and scope outlined in the “Standard 

Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR 

[Light-Water Reactor] Edition (NUREG-0800),” dated July 1981 (SRP, NUREG-0800). The 

NRC staff added additional chapters to address the overall assessment of the facility, nuclear 

performance plan (NPP) issues, and other generic regulatory topics.

Each of the sections and appendices of this supplement is numbered the same as the SER 

section that is being updated, and the discussions are supplementary to, and not in lieu of, the 

discussion in the SER, unless otherwise noted. For example, Appendix E continues to list the 

principal contributors to the SSER. However, the chronology of the safety review 

correspondence previously given in Appendix A has been discontinued, and a reference is 

supplied instead to the NRC’s Agencywide Documents Access and Management System 

(ADAMS) or the Public Document Room (PDR). Public correspondence exchanged between 

the NRC and TVA is available through ADAMS or the PDR. References listed as “not publicly 
1-2

available” in the SSER contain proprietary information and have been withheld from public 

disclosure in accordance with 10 CFR 2.390, “Public inspections, exemptions, requests for 

withholding.”

Appendix HH includes an action items table. This lists all the closed items, confirmatory issues, 

and proposed license conditions that have been resolved in support of an NRC finding of 

reasonable assurance on the OL application for WBN, Unit 2. 

The NRC’s ADAMS is the agency’s official recordkeeping system. ADAMS has the full text of 

regulatory and technical documents and reports written by the NRC, NRC contractors, or NRC 

licensees. Documents include NRC regulatory guides, NUREG-series reports, correspondence, 

inspection reports, and other materials. These documents are assigned accession numbers

and are searchable and accessible in ADAMS. Documents are released periodically during the 

day in the ADAMS PUBLIC/Legacy Interface Combined (ADAMS PUBLIC) and Web-based 

ADAMS (WBA) interfaces; they are released once a day in Web-based Publicly Available 

Records System (PARS). These documents in full text can be searched using ADAMS 

accession numbers or specific fields and parameters such as docket number and document 

dates. 

More information on ADAMS and help for accessing documents may be obtained on the NRC 

Public Website at http://www.nrc.gov/reading-rm/adams/faq.html#1.

All WBN documents may be accessed using WBN Docket Nos. 05000390 and 05000391 for 

Units 1 and 2, respectively.

The WBN, Unit 2, Project Manager is Justin C. Poole, who may be contacted by calling 

(301) 415-2048, by email to Justin.Poole@nrc.gov, or by writing to the following address:

Mr. Justin Poole

Mail Stop O-8G9A

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

1.7 Summary of Outstanding Issues

NUREG-0847 is the SER for the application filed by TVA, as applicant and owner, for licenses 

to operate the Watts Bar Nuclear Plant, Units 1 and 2 (Docket Nos. 50-390 and 50-391). The 

NRC staff documented its previous review and conclusions on the OL application for WBN, Unit 

1, in the SER (NUREG-0847, dated June 1982) and its Supplements 1 through 20. Based on 

these reviews, the NRC staff issued an OL for WBN, Unit 1, in 1996. In the SER and SSERs 1 

through 20, the NRC staff also reviewed and approved certain topics for WBN, Unit 2, though no 

final conclusions were made about an OL for WBN, Unit 2. To establish the remaining scope 

and the regulatory framework for the NRC staff’s review of an OL for WBN, Unit 2, the NRC staff 

reviewed the SER and SSERs 1 through 20. Based on this review, the NRC staff identified 

“resolved” topics (i.e., out of scope for review) and “open” topics (i.e., in scope for NRC staff 

review) for WBN, Unit 2. Where it was not clear whether the SER topic applied to Unit 2, or not, 

the NRC staff conservatively identified it as “open” pending further evaluation. It should be 

noted that these were not technical evaluations of each topic; rather, it was a status review to 

determine whether the topic was “open” or “resolved.” The NRC staff documented this 

evaluation in SSER 21 as the baseline for resumption of the review of the OL application for 
1-3

Unit 2. Thus, SSER 21 reflects the status of the NRC staff’s review of WBN, Unit 2, up to 1995. 

The NRC staff notes that a subsequent, more detailed assessment found some topics 

conservatively identified in the initial assessment as “open” should be redefined as “closed.” 

Conversely, the NRC staff notes that there have been circumstances that resulted in the need to 

reopen some previously closed topic areas that may have been adequately documented and 

that are considered closed in SSER 21. Such cases are identified by a footnote in the relevant

SSER to document that previous “open” topics have been recategorized as “closed” without 

requiring further review, or vice versa. 

The SER and SSERs 1 through 20 evaluated the changes to the final safety analysis report 

(FSAR) until Amendment No. 91. FSAR Amendment No. 91 was the initial licensing basis for 

WBN, Unit 1. At this time, the FSAR was applicable to both Units 1 and 2. As part of its 

updated OL application for WBN, Unit 2, TVA split FSAR Amendment No. 91 into two separate 

FSARs – one for WBN, Unit 1, and one for WBN, Unit 2. TVA has submitted WBN, Unit 2, 

FSAR Amendment Nos. 92 through 114 to address the “open” topics in support of its OL 

application for WBN, Unit 2. These FSAR amendments reflect changes that have occurred 

since 1995. Reviews of FSAR changes that have been completed by the NRC staff are 

documented in SSERs 22 through 29. 

Additional general topics (e.g., financial qualifications that were not included in SSER 21, but 

that should be resolved before issuance of an OL) are also identified in SSER 22 and 

subsequent supplements.

SSER 21 initially contained the table below documenting the status of each SER topic. The 

relevant document in which the topic was last addressed is shown in parentheses. This table 

was maintained in supplements subsequent to SSER 21 to reflect the status of review for each 

topic.

ISSUE STATUS TABLE

Issue Status Section Note

(1) Site Envelope 2

(2) Geography and Demography Resolved (SSER 22) 2.1

(3) Site Location and Description Resolved (SER)

(SSER 22)

2.1.1 3

(4) Exclusion Area Authority and 

Control

Resolved (SER)

(SSER 22)

2.1.2 3

(5) Population Distribution Resolved (SER)

(SSER 22)

2.1.3

(6) Conclusions Resolved (SER)

(SSER 22)

2.1.4

(7) Nearby Industrial, Transportation, 

and Military Facilities

Resolved (SSER 22) 2.2

(8) Transportation Routes Resolved (SER)

(SSER 22)

2.2.1

(9) Nearby Facilities Resolved (SER)

(SSER 22)

2.2.2

(10) Conclusions Resolved (SER)

(SSER 22)

2.2.3
1-4

Issue Status Section Note

(11) Meteorology Resolved (SER)

(SSER 22)

2.3

(12) Regional Climatology Resolved (SER)

(SSER 22)

2.3.1

(13) Local Meteorology Resolved (SER)

(SSER 22)

2.3.2

(14) Onsite Meteorological 

Measurements Program

Resolved (SER)

(SSER 22)

(SSER 25)

2.3.3

(15) Short-Term (Accident) Atmospheric 

Diffusion Estimates

Resolved (SER)

(SSER 14)

(SSER 22)

2.3.4

(16) Long-Term (Routine) Diffusion 

Estimates

Resolved (SER)

(SSER 14)

(SSER 22)

2.3.5

(17) Hydrologic Engineering Resolved (SSER 27)

(SSER 28)

2.4

(18) Introduction Resolved (SER) 2.4.1

(19) Hydrologic Description Resolved (SER) 2.4.2

(20) Flood Potential Resolved (SER) 2.4.3

(21) Local Intense Precipitation in Plant 

Area

Resolved (SER) 2.4.4 1

(22) Roof Drainage Resolved (SER) 2.4.5 1

(23) Ultimate Heat Sink Resolved (SER) 2.4.6

(24) Groundwater Resolved (SER) 2.4.7 1

(25) Design Basis for Subsurface 

Hydrostatic Loading

Resolved (SER)

(SSER 3)

2.4.8

(26) Transport of Liquid Releases Resolved (SER)

(SSER 22)

2.4.9 2

(27) Flooding Protection Requirements Resolved (SER)

(SSER 24)

(SSER 27)

(SSER 28)

2.4.10

(28) Geological, Seismological, and 

Geotechnical Engineering

Resolved (SER)

(SSER 24)

2.5

(29) Geology Resolved (SER) 2.5.1

(30) Seismology Resolved (SER) 2.5.2

(31) Surface Faulting Resolved (SER) 2.5.3

(32) Stability of Subsurface Materials 

and Foundations

Resolved (SER)

(SSER 3)

(SSER 9)

(SSER 11)

2.5.4

(33) Stability of Slopes Resolved (SER) 2.5.5

(34) Embankments and Dams Resolved (SER)

(SSER 22)

2.5.6

(35) References (SER)

(SSER 22)

2.6
1-5

Issue Status Section Note

(36) Design Criteria - Structures, 

Components, Equipment, and 

Systems

3

(37) Introduction 3.1

(38) Conformance With General Design 

Criteria 

Resolved (SER) 3.1.1

(39) Conformance With Industry Codes 

and Standards

Resolved (SER) 3.1.2

(40) Classification of Structures, 

Systems and Components

Resolved (SSER 14)

(SSER 22)

3.2

(41) Seismic Classifications Resolved (SER)

(SSER 3)

(SSER 5)

(SSER 6)

(SSER 8)

3.2.1

(42) System Quality Group Classification Resolved (SER)

(SSER 3)

(SSER 6)

(SSER 7)

(SSER 9)

(SSER 22)

3.2.2

(43) Wind and Tornado Loadings 3.3

(44) Wind Loading Resolved (SER) 3.3.1

(45) Tornado Loading Resolved (SER) 3.3.2

(46) Flood Level (Flood) Design 3.4

(47) Flood Protection Resolved (SER) 3.4.1

(48) Missile Protection 3.5

(49) Missile Selection and Description Resolved (SER)

(SSER 9)

(SSER 14)

(SSER 22)

3.5.1

(50) Structures, Systems, and 

Components to be Protected from 

Externally Generated Missiles

Resolved (SER)

(SSER 2)

(SSER 22)

3.5.2

(51) Barrier Design Procedures Resolved (SER) 3.5.3

(52) Protection Against the Dynamic 

Effects Associated with the 

Postulated Rupture of Piping

Resolved (SER)

(SSER 6)

(SSER 11)

3.6

(53) Plant Design for Protection Against 

Postulated Piping Failures in Fluid 

System Outside Containment

Resolved (SER)

(SSER 14)

(SSER 22)

3.6.1

(54) Determination of Break Locations 

and Dynamic Effects Associated 

with the Postulated Rupture of 

Piping

Resolved (SER)

(SSER 14)

(SSER 22)

3.6.2 3
1-6

Issue Status Section Note

(55) Leak-Before-Break Evaluation 

Procedures

Resolved (SSER 5)

(SSER 12)

(SSER 22)

(SSER 24)

3.6.3

(56) Seismic Design Resolved (SER)

(SSER 6)

3.7 2

(57) Seismic Input Resolved (SER)

(SSER 6)

(SSER 9)

(SSER 16)

3.7.1 2

(58) Seismic Analysis Resolved (SER)

(SSER 6)

(SSER 8)

(SSER 11)

(SSER 16)

3.7.2 2

(59) Seismic Subsystem Analysis Resolved (SER)

(SSER 6)

(SSER 7)

(SSER 8)

(SSER 9)

(SSER 12)

(SSER 22)

3.7.3

(60) Seismic Instrumentation Resolved (SER) 3.7.4 1

(61) Design of Seismic Category I 

Structures

Resolved (SER)

(SSER 9)

3.8 2

(62) Steel Containment Resolved (SER)

(SSER 3)

3.8.1

(63) Concrete and Structural Steel 

Internal Structures

Resolved (SER)

(SSER 7)

3.8.2

(64) Other Seismic Category I Structures Resolved (SER)

(SSER 14)

(SSER 16)

(SSER 29)

3.8.3

(65) Foundations Resolved (SER) 3.8.4

(66) Mechanical Systems and 

Components

Resolved (SER) 3.9

(67) Special Topics for Mechanical 

Components

Resolved (SER)

(SSER 6)

(SSER 13)

(SSER 22)

3.9.1

(68) Dynamic Testing and Analysis of 

Systems, Components, and 

Equipment

Resolved (SER)

(SSER 14)

(SSER 22)

3.9.2
1-7

Issue Status Section Note

(69) ASME Code Class 1, 2, and 3 

Components, Component 

Structures, and Core Support 

Structures

Resolved (SER)

(SSER 3)

(SSER 4)

(SSER 6)

(SSER 7)

(SSER 8)

(SSER 15)

(SSER 22)

(SSER 29)

3.9.3

(70) Control Rod Drive Systems Resolved (SER) 3.9.4

(71) Reactor Pressure Vessel Internals Resolved (SER)

(SSER 23)

(SSER 26)

3.9.5

(72) Inservice Testing of Pumps and 

Valves

Resolved (SER)

(SSER 5)

(SSER 12)

(SSER 14)

(SSER 18)

(SSER 20)

(SSER 22)

(SSER 27)

3.9.6

(73) Seismic and Dynamic Qualification 

of Seismic Category I Mechanical 

and Electrical Equipment

Resolved (SER)

(SSER 1)

(SSER 3)

(SSER 4)

(SSER 5)

(SSER 6)

(SSER 8)

(SSER 9)

(SSER 23)

3.10

(74) Environmental Qualification of 

Mechanical and Electrical 

Equipment

Resolved (SSER 15)

(SSER 22)

(SSER 27)

3.11

(75) Threaded Fasteners — ASME Code 

Class 1, 2, and 3

Resolved (SSER 22) 3.13

(76) Reactor 4

(77) Introduction (SER)

(SSER 23)

4.1

(78) Fuel System Design Resolved (SSER 23)

(SSER 27)

4.2

(79) Description Resolved (SER)

(SSER 13)

(SSER 23)

4.2.1

(80) Thermal Performance Resolved (SER)

(SSER 2)

(SSER 23)

(SSER 27)

4.2.2
1

-8

Issue Status Section Note

(81) Mechanical Performance Resolved (SER)

(SSER 2)

(SSER 10)

(SSER 13)

(SSER 23)

4.2.3

(82) Surveillance Resolved (SER)

(SSER 2)

(SSER 23)

4.2.4

(83) Fuel Design Considerations Resolved (SER)

(SSER 23)

4.2.5

(84) Nuclear Design Resolved (SSER 23)4.3

(85) Design Basis Resolved (SER)

(SSER 13)

(SSER 23)

4.3.1

(86) Design Description Resolved (SER)

(SSER 13)

(SSER 15)

(SSER 23)

4.3.2

(87) Analytical Methods Resolved (SER)

(SSER 23)

4.3.3

(88) Summary of Evaluation Findings Resolved (SER)

(SSER 23)

4.3.4

(89) Thermal-Hydraulic Design Resolved (SSER 23) 4.4

(90) Performance in Safety Criteria Resolved (SER)

(SSER 23)

4.4.1

(91) Design Bases Resolved (SER)

(SSER 12)

(SSER 23)

4.4.2

(92) Thermal-Hydraulic Design 

Methodology

Resolved (SER) (SSER 6)

(SSER 8)

(SSER 12)

(SSER 13)

(SSER 16)

SE dated 

6/13/89

(SSER 23)

4.4.3

(93) Operating Abnormalities Resolved (SER)

(SSER 13)

(SSER 23)

4.4.4

(94) Loose Parts Monitoring System Resolved (SER)

(SSER 3)

(SSER 5)

(SSER 16)

(SSER 23)

4.4.5

(95)Thermal-Hydraulic Comparison Resolved (SER)

(SSER 23)

4.4.6
1-9

Issue Status Section Note

(96) N-1 Loop Operation Resolved (SER)

(SSER 23)

4.4.7

(97) Instrumentation for Inadequate Core 

Cooling Detection (Three Mile Island 

((TMI) Action Item II.F.2)

Resolved (SER)

(SSER 10)

(SSER 23)

(SSER 27)

4.4.8

(98) Summary and Conclusion Resolved (SER)

(SSER 23)

(SSER 25)

4.4.9

(99) Reactor Materials 4.5

(100) Control Rod Drive Structural 

Materials

Resolved (SER) 4.5.1 1

(101) Reactor Internals and Core Support 

Materials

Resolved (SER) 4.5.2

(102) Functional Design of Reactivity 

Control Systems

Resolved (SER)

(SSER 23)

4.6

(103) Reactor Coolant System and 

Connected Systems

5

(104) Summary Description Resolved (SER)

(SSER 5)

(SSER 6)

5.1 2

(105) Integrity of Reactor Coolant 

Pressure Boundary

5.2

(106) Compliance with Codes and Code 

Cases

Resolved (SER)

(SSER 22)

5.2.1

(107) Overpressurization Protection Resolved (SER)

(SSER 2)

(SSER 15)

(SSER 24)

5.2.2

(108) Reactor Coolant Pressure Boundary 

Materials

Resolved (SER)

(SSER 22)

5.2.3

(109) Reactor Coolant System Pressure 

Boundary Inservice Inspection and 

Testing

Resolved (SER)

(SSER 10)

(SSER 12)

(SSER 15)

(SSER 16)

(SSER 23)

5.2.4

(110) Reactor Coolant Pressure Boundary 

Leakage Detection

Resolved (SER)

(SSER 9)

(SSER 11)

(SSER 12)

(SSER 22)

5.2.5

(111) Reactor Vessel 5.3

(112) Reactor Vessel Materials Resolved (SER)

(SSER 11)

(SSER 14)

(SSER 22)

(SSER 25)

5.3.1
1-10

Issue Status Section Note

(113) Pressure-Temperature Limits Resolved (SER)

(SSER 16)

(SSER 22)

(SSER 25)

5.3.2

(114) Reactor Vessel Integrity Resolved (SER)

(SSER 22)

5.3.3

(115) Component and Subsystem Design 5.4

(116) Reactor Coolant Pumps Resolved (SER)

(SSER 22)

5.4.1 2

(117) Steam Generators Resolved (SER)

(SSER 1)

(SSER 4)

(SSER 22)

5.4.2

(118) Residual Heat Removal System Resolved (SER)

(SSER 2)

(SSER 5)

(SSER 10)

(SSER 11)

(SSER 23)

5.4.3

(119) Pressurizer Relief Tank Resolved (SER)

(SSER 22)

5.4.4

(120) Reactor Coolant System Vents 

(TMI Action Item II.B.1)

Resolved (SER)

(SSER 2)

(SSER 5)

(SSER 12)

(SSER 23)

5.4.5

(121) Engineered Safety Features 6

(122) Engineered Safety Feature 

Materials

6.1

(123) Metallic Materials Resolved (SER)

(SSER 23)

(SSER 27)

6.1.1

(124) Organic Materials Resolved (SER)

(SSER 22)

6.1.2

(125) Postaccident Emergency Cooling 

Water Chemistry

Resolved (SER)

(SSER 22)

(SSER 29)

6.1.3

(126) Containment Systems 6.2

(127) Containment Functional Design Resolved (SER)

(SSER 3)

(SSER 5)

(SSER 7)

(SSER 12)

(SSER 14)

(SSER 15)

(SSER 22)

(SSER 29)

6.2.1
1-11

Issue Status Section Note

(128) Containment Heat Removal 

Systems

Resolved (SER)

(SSER 7)

(SSER 22)

(SSER 29)

6.2.2

(129) Secondary Containment Functional 

Design

Resolved (SER)

(SSER 18)

(SSER 22)

(SSER 29)

6.2.3

(130) Containment Isolation Systems Resolved (SER)

(SSER 3)

(SSER 5)

(SSER 7)

(SSER 12)

(SSER 22)

(SSER 29)

6.2.4

(131) Combustible Gas Control Systems Resolved (SER)

(SSER 4)

(SSER 5)

(SSER 8)

(SSER 22)

(SSER 29)

6.2.5

(132) Containment Leakage Testing Resolved (SER)

(SSER 4)

(SSER 5)

(SSER 19)

(SSER 22)

(SSER 26)

(SSER 29)

6.2.6

(133) Fracture Prevention of Containment 

Pressure Boundary

Resolved (SER)

(SSER 4)

(SSER 23)

6.2.7 1

(134) Emergency Core Cooling System Resolved (SER) 6.3 1

(135) System Design Resolved (SER)

(SSER 6)

(SSER 7)

(SSER 11)

(SSER 29)

6.3.1

(136) Evaluation Resolved (SER)

(SSER 5)

6.3.2 1

(137) Testing Resolved (SER)

(SSER 2)

(SSER 9)

(SSER 29)

6.3.3

(138) Performance Evaluation Resolved (SER) 6.3.4

(139) Conclusions Resolved (SER) 6.3.5
1-12

Issue Status Section Note

(140) Control Room Habitability Resolved (SER)

(SSER 5)

(SSER 11)

(SSER 16)

(SSER 18

(SSER 22)

(SSER 29)

6.4

(141) Engineered Safety Feature (ESF) 

Filter Systems

6.5

(142) ESF Atmosphere Cleanup System Resolved (SER)

(SSER 5)

(SSER 22)

(SSER 29)

6.5.1

(143) Fission Product Cleanup System Resolved (SER) 6.5.2 1

(144) Fission Product Control System Resolved (SER)

(SSER 22)

(SSER 26)

6.5.3

(145) Ice Condenser as a Fission Product 

Cleanup System

Resolved (SER) 6.5.4 1

(146) Inservice Inspection of Class 2 and 

3 Components

Resolved (SER)

(SSER 10)

(SSER 12)

(SSER 15)

(SSER 23)

6.6

(147) Instrumentation and Controls 7

(148) Introduction 7.1

(149) General Resolved (SER)

(SSER 13)

(SSER 16)

(SSER 23)

7.1.1

(150) Comparison with Other Plants Resolved (SER)

(SSER 23)

7.1.2 1

(151) Design Criteria Resolved (SER)

(SSER 4)

(SSER 15)

(SSER 23)

7.1.3

(152) Reactor Trip System Resolved (SER) 7.2

(153) System Description Resolved (SER)

(SSER 13)

(SSER 15)

(SSER 23)

(SSER 27)

7.2.1

(154) Manual Trip Switches Resolved (SER)

(SSER 23)

7.2.2 1

(155) Testing of Reactor Trip Breaker 

Shunt Coils

Resolved (SER)

(SSER 23)

7.2.3 1

(156) Anticipatory Trips Resolved (SER)

(SSER 23)

7.2.4
1-13

Issue Status Section Note

(157) Steam Generator Water Level Trip Resolved (SER)

(SSER 2)

(SSER 14)

(SSER 23)

7.2.5

(158) Conclusions Resolved (SER)

(SSER 13)

(SSER 23)

7.2.6

(159) Engineered Safety Features System Resolved (SER)

(SSER 13)

7.3

(160) System Description Resolved (SER)

(SSER 13)

(SSER 14)

(SSER 23)

7.3.1

(161) Containment Sump Level 

Measurement

Resolved (SER)

(SSER 2)

(SSER 23)

7.3.2

(162) Auxiliary Feedwater Initiation and 

Control

Resolved (SER)

(SSER 23)

7.3.3 1

(163) Failure Modes and Effects Analysis Resolved (SER)

(SSER 23)

7.3.4

(164) Office of Inspection and 

Enforcement (IE) Bulletin 80-06

Resolved (SER)

(SSER 3)

(SSER 23)

7.3.5

(165) Conclusions Resolved (SER)

(SSER 13)

(SSER 23)

7.3.6

(166) Systems Required for Safe 

Shutdown

7.4

(167) System Description Resolved (SER)

(SSER 23)

7.4.1

(168) Safe Shutdown from Auxiliary 

Control Room

Resolved (SER)

(SSER 7)

(SSER 23)

7.4.2

(169) Conclusions Resolved (SER)

(SSER 23)

7.4.3

(170) Safety-Related Display 

Instrumentation

7.5

(171) Display Systems Resolved (SER)

(SSER 23)

7.5.1

(172) Postaccident Monitoring System Resolved (SER)

(SSER 7)

(SSER 9)

(SSER 14)

(SSER 15)

(SSER 23)

(SSER 25)

(SSER 27)

(SSER 29)

7.5.2
1-14

Issue Status Section Note

(173) IE Bulletin 79-27 Resolved (SER)

(SSER 23)

(SSER 27)

7.5.3

(174) Conclusions Resolved (SER) 7.5.4

(175) All Other Systems Required for 

Safety

7.6

(176) Loose Part Monitoring System Resolved (SER)

(SSER 23)

(SSER 24)

7.6.1

(177) Residual Heat Removal System 

Bypass Valves

Resolved (SER)

(SSER 23)

7.6.2

(178) Upper Head Injection Manual 

Control

Resolved (SER)

(SSER 23)

7.6.3

(179) Protection Against Spurious 

Actuation of Motor-Operated Valves

Resolved (SER)

(SSER 23)

7.6.4

(180) Overpressure Protection during Low 

Temperature Operation

Resolved (SER)

(SSER 4)

(SSER 23)

7.6.5

(181) Valve Power Lockout Resolved (SER)

(SSER 23)

7.6.6

(182) Cold Leg Accumulator Valve 

Interlocks and Position Indication

Resolved (SER)

(SSER 23)

7.6.7

(183) Automatic Switchover From 

Injection to Recirculation Mode

Resolved (SER)

(SSER 23)

7.6.8

(184) Conclusions Resolved (SER)

(SSER 4)

7.6.9

(185) Control Systems Not Required for 

Safety

7.7

(186) System Description Resolved (SER)

(SSER 23)

(SSER 24)

(SSER 25)

(SSER 27)

(SSER 29)

7.7.1

(187) Safety System Status Monitoring 

System

Resolved (SER)

(SSER 7)

(SSER 13)

(SSER 23)

7.7.2

(188) Volume Control Tank Level Control 

System

Resolved (SER)

(SSER 23)

7.7.3

(189) Pressurizer and Steam Generator 

Overfill

Resolved (SER)

(SSER 23)

7.7.4

(190) IE Information Notice 79-22 Resolved (SER)

(SSER 23)

7.7.5

(191) Multiple Control System Failures Resolved (SER)

(SSER 23)

7.7.6

(192) Conclusions Resolved (SER) 7.7.7
1-15

Issue Status Section Note

(193) Anticipated Transient Without 

Scram Mitigation System Actuation 

Circuitry (AMSAC)

Resolved (SSER 9)

(SSER 14)

(SSER 23)

7.7.8

(194) NUREG-0737 Items Resolved (SER)

(SSER 23)

7.8

(195) Relief and Safety Valve Position 

Indication (TMI Action Item II.D.3)

Resolved (SER)

(SSER 5)

(SSER 14)

(SSER 23)

7.8.1

(196) Auxiliary Feedwater System 

Initiation and Flow Indication (TMI 

Action Item II.E.1.2)

Resolved (SER)

(SSER 23)

(SSER 29)

7.8.2

(197) Proportional Integral Derivative 

Control Modification (TMI Action 

Item II.K.3.9)

Resolved (SER)

(SSER 23)

7.8.3

(198) Proposed Anticipatory Trip 

Modification (TMI Action Item 

II.K.3.10)

Resolved (SER)

(SSER 4)

(SSER 23)

7.8.4

(199) Confirm Existence of Anticipatory 

Reactor Trip Upon Turbine Trip (TMI 

Action Item II.K.3.12)

Resolved (SER)

(SSER 23)

7.8.5

(200) Data Communication Systems (SSER 23) 7.9

(201) Electric Power Systems 8

(202) General Resolved (SER)

(SSER 22)

(SSER 24)

(SSER 27)

8.1

(203) Offsite Power System Resolved (SER)

(SSER 22)

8.2

(204) Compliance with GDC 5 Resolved (SER)

(SSER 13)

(SSER 22)

8.2.1

(205) Compliance with GDC 17 Resolved (SER)

(SSER 2)

(SSER 3)

(SSER 13)

(SSER 14)

(SSER 15

(SSER 22)

(SSER 27)

8.2.2

(206) Compliance with GDC 18 Resolved (SER)

(SSER 22)

8.2.3

(207) Evaluation Findings Resolved (SER)

(SSER 22)

8.2.4

(208) Onsite Power Systems Resolved (SER)

(SSER 10)

(SSER 19)

(SSER 22)

8.3
1

-16

Issue Status Section Note

(209) Onsite AC Power System 

Compliance with GDC 17

Resolved (SER)

(SSER 2)

(SSER 7)

(SSER 9)

(SSER 10)

(SSER 13)

(SSER 14)

(SSER 18)

(SSER 20)

(SSER 22)

(SSER 27)

(SSER 28)

(SSER 29)

8.3.1

(210) Onsite Direct Current System 

Compliance with GDC 17

Resolved (SER)

(SSER 2)

(SSER 3)

(SSER 13)

(SSER 14)

(SSER 22)

8.3.2

(211) Common Electrical Features and 

Requirements

Resolved (SER)

(SSER 2)

(SSER 3)

(SSER 7)

(SSER 13)

(SSER 14)

(SSER 15)

(SSER 16)

(SSER 22)

8.3.3

(212) Evaluation Findings Resolved (SER)

(SSER 2)

(SSER 3)

(SSER 7)

(SSER 13)

(SSER 14)

(SSER 15)

(SSER 16)

(SSER 22)

8.3.4

(213) Station Blackout Resolved (SSER 22) 8.4

(214) Auxiliary Systems Resolved (SER)

(SSER 10)

9

(215) Fuel Storage Facility 9.1

(216) New-Fuel Storage Resolved (SER) 9.1.11

(217) Spent-Fuel Storage Resolved (SER)

(SSER 5)

(SSER 15)

(SSER 16)

(SSER 22)

9.1.2
1

-17

Issue Status Section Note

(218) Spent Fuel Pool Cooling and 

Cleanup System

Resolved (SER)

(SSER 11)

(SSER 15)

(SSER 23)

(SSER 26)

9.1.3

(219) Fuel-Handling System Resolved (SER)

(SSER 3)

(SSER 13)

(SSER 22)

(SSER 24)

9.1.4

(220) Water Systems 9.2

(221) Essential Raw Cooling Water and 

Raw Cooling Water System

Resolved (SER)

(SSER 9)

(SSER 10)

(SSER 18)

(SSER 23)

(SSER 27)

(SSER 29)

9.2.1

(222) Component Cooling System 

(Reactor Auxiliaries Cooling Water 

System)

Resolved (SER)

(SSER 5)

(SSER 23)

(SSER 27)

(SSER 29)

9.2.2

(223) Demineralized Water Makeup 

System

Resolved (SER)

(SSER 22)

9.2.3

(224) Potable and Sanitary Water 

Systems

Resolved (SER)

(SSER 9)

(SSER 22)

9.2.4

(225) Ultimate Heat Sink Resolved (SER)

(SSER 23)

(SSER 27)

9.2.5

(226) Condensate Storage Facilities Resolved (SER)

(SSER 12)

(SSER 22)

(SSER 29)

9.2.6

(227) Process Auxiliaries 9.3

(228) Compressed Air System Resolved (SER)

(SSER 22)

9.3.11

(229) Process Sampling System Resolved (SER)

(SSER 3)

(SSER 5)

(SSER 14)

(SSER 16)

(SSER 24)

9.3.2

(230) Equipment and Floor Drainage 

System

Resolved (SER)

(SSER 22)

9.3.33

(231) Chemical and Volume Control 

System

Resolved (SER)

(SSER 22)

9.3.43
1-18

Issue Status Section Note

(232) Heat Tracing Resolved (SSER 22) 9.3.8

(233) Heating, Ventilation, and Air 

Conditioning Systems

9.4

(234) Control Room Area Ventilation 

System

Resolved (SER)

(SSER 9)

(SSER 22)

9.4.1

(235) Fuel-Handling Area Ventilation 

System

Resolved (SER)

(SSER 22)

9.4.2

(236) Auxiliary Building and Radwaste 

Area Ventilation System

Resolved (SER)

(SSER 22)

9.4.3

(237) Turbine Building Area Ventilation 

System

Resolved (SER)

(SSER 22)

9.4.4

(238) Engineered Safety Features 

Ventilation System

Resolved (SER)

(SSER 9)

(SSER 10)

(SSER 11)

(SSER 14)

(SSER 16)

(SSER 19)

(SSER 22)

9.4.5

(239) Reactor Building Purge Ventilation 

System

Resolved (SSER 22) 9.4.6

(240) Containment Air Cooling System Resolved (SSER 22) 9.4.7

(241) Condensate Demineralizer Waste 

Evaporator Building Environmental 

Control System

Resolved (SSER 22) 9.4.8

(242) Other Auxiliary Systems 9.5

(243) Fire Protection Resolved (SER)

(SSER 10)

(SSER 18)

(SSER 19)

(SSER 26)

(SSER 29)

9.5.1 3

(244) Communications System Resolved (SER)

(SSER 5)

9.5.2 1

(245) Lighting System Resolved (SER)

(SSER 22)

9.5.3

(246) Emergency Diesel Engine Fuel Oil 

Storage and Transfer System

Resolved (SER)

(SSER 5)

(SSER 9)

(SSER 10)

(SSER 11)

(SSER 12)

(SSER 22)

9.5.4 2

(247) Emergency Diesel Engine Cooling 

Water System

Resolved (SER)

(SSER 5)

(SSER 11)

9.5.5 1
1-19

Issue Status Section Note

(248) Emergency Diesel Engine Starting 

Systems

Resolved (SER)

(SSER 5)

(SSER 10)

(SSER 22)

9.5.6 2

(249) Emergency Diesel Engine 

Lubricating Oil System

Resolved (SER)

(SSER 3)

(SSER 5)

(SSER 10)

(SSER 22)

9.5.7 2

(250) Emergency Diesel Engine 

Combustion Air Intake and Exhaust 

System

Resolved (SER)

(SSER 5)

(SSER 10)

(SSER 22)

9.5.8 2

(251) Steam and Power Conversion 

System

10

(252) Summary Description Resolved (SER) 10.1

(253) Turbine Generator Resolved (SER)

(SSER 5)

10.2

(254) Turbine Generator Design Resolved (SER)

(SSER 12)

(SSER 22)

10.2.1

(255) Turbine Disc Integrity Resolved (SER)

(SSER 23)

10.2.2

(256) Main Steam Supply System Resolved (SER) 10.3

(257) Main Steam Supply System (Up to 

and Including the Main Steam 

Isolation Valves)

Resolved (SER)

(SSER 19)

(SSER 22)

10.3.1

(258) Main Steam Supply System Resolved (SER)

(SSER 22)

10.3.2 2

(259) Steam and Feedwater System 

Materials

Resolved (SER)

(SSER 22)

10.3.3

(260) Secondary Water Chemistry Resolved (SER)

(SSER 5)

(SSER 22)

10.3.4

(261) Other Features 10.4

(262) Main Condenser Resolved (SER)

(SSER 9)

(SSER 22)

10.4.1

(263) Main Condenser Evacuation System Resolved (SER)

(SSER 22)

10.4.2

(264) Turbine Gland Sealing System Resolved (SER)

(SSER 22)

10.4.3

(265) Turbine Bypass System Resolved (SER)

(SSER 5)

(SSER 22)

10.4.4

(266) Condenser Circulating Water 

System

Resolved (SER)

(SSER 22)

10.4.5
1-20

Issue Status Section Note

(267) Condensate Cleanup System Resolved (SER)

(SSER 22)

(SSER 27)

10.4.6

(268) Condensate and Feedwater 

Systems

Resolved (SER)

(SSER 14)

(SSER 22)

10.4.7

(269) Steam Generator Blowdown System Resolved (SER)

(SSER 22)

(SSER 24)

10.4.8

(270) Auxiliary Feedwater System Resolved (SER)

(SSER 14)

(SSER 23)

(SSER 24)

(SSER 29)

10.4.9

(271) Heater Drains and Vents Resolved (SSER 22) 10.4.10

(272) Steam Generator Wet Layup 

System

Resolved (SSER 22) 10.4.11

(273) Radioactive Waste Management 11

(274) Summary Description Resolved (SER)

(SSER 16)

(SSER 24)

11.1 2

(275) Liquid Waste Management Resolved (SER)

(SSER 4)

(SSER 16)

(SSER 24)

11.2

(276) Gaseous Waste Management Resolved (SER)

(SSER 8)

(SSER 16)

(SSER 24)

(SSER 25)

(SSER 27)

11.3

(277) Solid Waste Management System Resolved (SER)

(SSER 16)

(SSER 24)

11.4

(278) Process and Effluent Radiological 

Monitoring and Sampling Systems

Resolved (SER)

(SSER 16)

(SSER 20)

(SSER 24)

11.5

(279) Evaluation Findings Resolved (SER)

(SSER 8)

(SSER 16)

11.6

(280) NUREG-0737 Items Resolved (SER) 11.7

(281) Wide-Range Noble Gas, Iodine, and 

Particulate Effluent Monitors (TMI 

Action Item II.F.1.1, II.F.1.2.a, and 

II.F.1.2.b)

Resolved (SER)

(SSER 5)

(SSER 6)

(SSER 29)

11.7.1
1-21

Issue Status Section Note

(282) Primary Coolant Outside 

Containment (TMI Action Item 

III.D.1.1)

Resolved (SER)

(SSER 5)

(SSER 6)

(SSER 10)

(SSER 16)

(SSER 29)

11.7.2

(283) Radiation Protection 12

(284) General Resolved (SER)

(SSER 10)

(SSER 14)

(SSER 24)

12.1

(285) Ensuring that Occupational 

Radiation Doses Are As Low As 

Reasonably Achievable (ALARA)

Resolved (SER)

(SSER 14)

(SSER 24)

12.2 2

(286) Radiation Sources Resolved (SER)

(SSER 14)

(SSER 24)

12.3

(287) Radiation Protection Design 

Features

Resolved (SER)

(SSER 10)

(SSER 14)

(SSER 18)

(SSER 24)

(SSER 26)

12.4

(288) Dose Assessment Resolved (SER)

(SSER 14)

(SSER 24)

(SSER 27)

12.5

(289) Health Physics Program Resolved (SER)

(SSER 10)

(SSER 14)

(SSER 24)

(SSER 26)

12.6

(290) NUREG-0737 Items 12.7

(291) Plant Shielding

(TMI Action Item II.B.2)

Resolved (SER)

(SSER 14)

(SSER 16)

(SSER 24)

(SSER 27)

(SSER 29)

12.7.1

(292) High Range In-Containment Monitor 

(TMI Action Item II.F.1.2.c)

Resolved (SER)

(SSER 5)

(SSER 29)

12.7.2

(293) In-Plant Radioiodine Monitor (TMI 

Action Item III.D.3.3)

Resolved (SER)

(SSER 16)

(SSER 29)

12.7.3

(294) Conduct of Operations 13
1-22

Issue Status Section Note

(295) Organization Structure of the 

Applicant

Resolved (SER)

(SSER 16)

(SSER 22)

13.1

(296) Management and Technical 

Organization

Resolved (SER) 13.1.1

(297) Corporate Organization and 

Technical Support

Resolved (SER) 13.1.2

(298) Plant Staff Organization Resolved (SER)

(SSER 8)

(SSER 22)

(SSER 25)

(SSER 27)

13.1.3

(299) Training 13.2

(300) Licensed Operator Training 

Program

Resolved (SER)

(SSER 9)

(SSER 10)

(SSER 22)

13.2.1

(301) Training for Non-licensed Personnel Resolved (SER) 13.2.2

(302) Emergency Preparedness 

Evaluation

13.3

(303) Introduction Resolved (SER)

(SSER 13)

(SSER 20)

(SSER 28)

13.3.1

(304) Evaluation of the Emergency Plan Resolved (SER)

(SSER 13)

(SSER 20)

(SSER 22)

(SSER 28)

13.3.2

(305) Conclusions Resolved (SER)

(SSER 13)

(SSER 20)

(SSER 22)

(SSER 28)

13.3.3

(306) Review and Audit Resolved (SER)

(SSER 8)

(SSER 22)

13.4

(307) Plant Procedures Resolved (SER)

(SSER 22)

13.5

(308) Administrative Procedures Resolved (SER)

(SSER 22)

13.5.1

(309) Operating and Maintenance 

Procedures

Resolved (SER)

(SSER 9)

(SSER 10)

(SSER 22)

13.5.2
1-23

Issue Status Section Note

(310) NUREG-0737 Items Resolved (SER)

(SSER 3)

(SSER 16)

(SSER 22)

13.5.3

(311) Physical Security Plan Resolved (SER)

(SSER 1)

(SSER 10)

(SSER 15)

(SSER 20)

(SSER 22)

13.6

(312) Introduction Resolved (SSER 22) 13.6.1

(313) Summary of Application Resolved (SSER 22) 13.6.2

(314) Regulatory Basis Resolved (SSER 22) 13.6.3

(315) Technical Evaluation Resolved (SSER 22) 13.6.4

(316) Conclusions Resolved (SSER 22) 13.6.5

(317) Cyber security Plan Resolved (SSER 24)

(SSER 28)

13.6.6

(318) Initial Test Program Resolved (SER)

(SSER 3)

(SSER 5)

(SSER 7)

(SSER 9)

(SSER 10)

(SSER 12)

(SSER 14)

(SSER 16)

(SSER 18)

(SSER 19)

(SSER 23)

14

(319) Accident Analyses 15

(320) General Discussion Resolved (SER) 15.1

(321) Normal Operation and Anticipated 

Transients

Resolved (SER) 15.2

(322) Loss-of-Cooling Transients Resolved (SER)

(SSER 13)

(SSER 14)

(SSER 24)

15.2.1

(323) Increased Cooling Inventory 

Transients

Resolved (SER)

(SSER 24)

15.2.2

(324) Change in Inventory Transients Resolved (SER)

(SSER 18)

(SSER 24)

15.2.3
1-24

Issue Status Section Note

(325) Reactivity and Power Distribution 

Anomalies

Resolved (SER)

(SSER 4)

(SSER 7)

(SSER 13)

(SSER 14)

(SSER 24)

(SSER 26)

15.2.4

(326) Conclusions Resolved (SER)

(SSER 4)

15.2.5

(327) Limiting Accidents Resolved (SER) 15.3

(328) Loss-of-Coolant Accident (LOCA) Resolved (SER)

(SSER 12)

(SSER 15)

(SSER 24)

(SSER 29)

15.3.1

(329) Steamline Break Resolved (SER)

(SSER 3)

(SSER 14)

(SSER 24)

15.3.2

(330) Feedwater System Pipe Break Resolved (SER)

(SSER 14)

(SSER 24)

15.3.3

(331) Reactor Coolant Pump Rotor 

Seizure

Resolved (SER)

(SSER 14)

(SSER 24)

15.3.4

(332) Reactor Coolant Pump Shaft Break Resolved (SER)

(SSER 14)

(SSER 24)

15.3.5

(333) Anticipated Transients Without 

Scram

Resolved (SER)

(SSER 3)

(SSER 5)

(SSER 6)

(SSER 10)

(SSER 11)

(SSER 12)

(SSER 24)

15.3.6

(334) Conclusions Resolved (SER) 15.3.7

(335) Radiological Consequences of 

Accidents

Resolved (SER)

(SSER 15)

(SSER 25)

15.4

(336) Loss-of-Coolant Accident Resolved (SER)

(SSER 5)

(SSER 9)

(SSER 18)

(SSER 25)

15.4.1

(337) Main Steamline Break Outside of 

Containment

Resolved (SER)

(SSER 15)

(SSER 25)

15.4.2
1-25

Issue Status Section Note

(338) Steam Generator Tube Rupture Resolved (SER)

(SSER 2)

(SSER 5)

(SSER 12)

(SSER 14)

(SSER 15)

(SSER 25)

15.4.3

(339) Control Rod Ejection Accident Resolved (SER)

(SSER 15)

(SSER 25)

15.4.4

(340) Fuel-Handling Accident Resolved (SER)

(SSER 4)

(SSER 15)

(SSER 25)

15.4.5

(341) Failure of Small Line Carrying 

Coolant Outside Containment

Resolved (SER)

(SSER 25)

15.4.6

(342) Postulated Radioactive Releases as 

a Result of Liquid Tank Failures

Resolved (SER)

(SSER 25)

15.4.7

(342a) Postulated Waste Gas Decay Tank 

Rupture

Resolved (SSER 25) 15.4.8

(343) NUREG-0737 Items 15.5

(344) Thermal Mechanical Report (TMI 

Action Item II.K.2.13)

Resolved (SER)

(SSER 4)

(SSER 24)

15.5.1

(345) Voiding in the Reactor Coolant 

System during Transients (TMI 

Action Item II.K.2.17)

Resolved (SER)

(SSER 4)

(SSER 24)

15.5.2

(346) Installation and Testing of Automatic 

Power-Operated Relief Valve 

Isolation System (TMI Action Item 

II.K.3.1) Report on Overall Safety 

Effect of Power-Operated Relief 

Valve Isolation System (TMI Action 

Item II.K.3.2)

Resolved (SER)

(SSER 5)

15.5.3

(347) Automatic Trip of Reactor Coolant 

Pumps (TMI Action Item II.K.3.5)

Resolved (SER)

(SSER 4)

(SSER 16)

(SSER 24)

15.5.4

(348) Small-Break LOCA Methods 

(II.K.3.30) and Plant-Specific 

Calculations (II.K.3.31)

Resolved (SER)

(SSER 4)

(SSER 5)

(SSER 16)

(SSER 29)

15.5.5

(349) Relative Risk of Low-Power 

Operation

Resolved (SER) 15.6

(350) Technical Specification Resolved (SER)

(SSER 19)

(SSER 29)

16
1-26

Issue Status Section Note

(351) Quality Assurance 17

(352) General Resolved (SER) 17.1

(353) Organization Resolved (SER) 17.2

(354) Quality Assurance Program Resolved (SER)

(SSER 2)

(SSER 5)

(SSER 10)

(SSER 13)

(SSER 15)

(SSER 22)

17.3

(355) Conclusions Resolved (SER) 17.4

(356) Maintenance Rule Resolved (SSER 29) 17.6

(357) Control Room Design Review 18

(358) General Resolved (SER)

(SSER 5)

(SSER 6)

(SSER 15)

(SSER 16)

(SSER 22)

18.1

(359) Conclusions Resolved (SER)

(SSER 16)

(SSER 22)

18.2

(360) Report of the Advisory Committee 

on Reactor Safeguards

Resolved (SER)

(SSER 1)

(SSER 4)

(SSER 14)

(SSER 20)

(SSER 28)

19

(361) Common Defense and Security Resolved (SER) 20

(362) Financial Qualifications Resolved (SER) 21

(363) TVA Financial Qualifications for 

WBN, Unit 2

Resolved (SSER 22)

(SSER 23)

21.1

(364) Foreign Ownership, Control, or 

Domination

Resolved (SSER 22) 21.2

(365) Financial Protection and Indemnity 

Requirements

22

(366) General Resolved (SER) 22.1

(367) Preoperational Storage of Nuclear 

Fuel

Resolved (SER) 22.2

(368) Operating Licenses Resolved (SSER 22)

(SSER 29)

22.3

(369) Quality of Construction, Operational 

Readiness, and Quality Assurance 

Effectiveness

25
1-27

Issue Status Section Note

(370) Program for Maintenance and 

Preservation of the Licensing Basis 

for Units 1 and 2

Resolved (SSER 22)

(SSER 27)

25.9

Notes:

1. In the process of further validating the information in the WBN, Unit 2, FSAR, 

TVA identified minor administrative/typographical changes to sections previously 

considered resolved. TVA addressed these changes to the applicable sections 

in its submittals and clearly indicated them to the NRC staff. The NRC staff has 

reviewed and confirmed that the changes made are administrative/typographical 

and do not impact the NRC staff’s conclusions as stated in previous SSERs. 

Based on this review, no additional review is necessary and this section remains 

resolved.

2. During the assessment of the regulatory framework for completion of the project, 

the NRC staff characterized certain topics as “Open” pending TVA’s validation of 

the information contained in the section. TVA has determined that the 

information presented in the FSAR remained valid and only identified minor 

administrative or typographical changes to the section. TVA addressed the 

changes in its submittals and clearly indicated the changes. The NRC staff 

reviewed and confirmed that the changes made to the section are 

administrative/typographical and do not impact its conclusions as stated in 

previous SSERs. Therefore, no additional review is necessary and the NRC staff 

considers this section resolved.

3. In SSER 21, this issue was identified as “Resolved.” However, TVA made 

changes to the Unit 2 FSAR affecting the previous NRC staff conclusions. The 

NRC staff evaluated the changes and the results are documented in SSERs 

subsequent to SSER 21.

1.9 License Conditions

The paragraphs in 10 CFR 50.54 (with the exception of paragraphs (r) and (gg)), and the 

applicable requirements of 10 CFR 50.55a, are conditions in every nuclear power reactor 

operating license issued under 10 CFR Part 50. Per 10 CFR 50.57(b), each operating license 

will include appropriate provisions with respect to any uncompleted items of construction and 

such limitations or conditions as are required to assure that operation during the period of the 

completion of such items will not endanger public health and safety. 

During the pendency of its review of TVA’s application, the Staff has considered the following 

additional license conditions.
1-28

1.9.1 Flooding License Condition

The NRC staff had proposed two license conditions in Section 2.4 of SSER 24. TVA has 

supplemented Section 2.4 since publication of SSER 24. The NRC staff’s review of this 

information can be found in Section 2.4 of SSER 28. The NRC staff has determined that the 

previously proposed license conditions were no longer appropriate. The operating license for 

WBN, Unit 2 will include a license condition similar to the following.

Flooding Protection Proposed License Condition:

TVA shall implement permanent modifications to prevent overtopping of the 

embankments of the Fort Loudoun Dam due to the Probable Maximum Flood by 

February 1, 2017. 

1.9.2 Cyber Security License Condition

The NRC staff had proposed two license conditions discussed in Section 13.6.6.3.22 of

SSER 24. The NRC has received an updated implementation schedule from TVA. The 

previous license conditions were deleted. The operating license for WBN, Unit 2 will include a 

license condition similar to the following. 

Cyber Security Proposed License Condition:

The licensee shall fully implement and maintain in effect all provisions of the 

Commission approved cyber security plan (CSP), including changes made 

pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The licensee 

CSP was approved by NUREG-0847, Supplement 28.

1.9.3 Core Operating Limits License Condition

The NRC staff proposed a license condition discussed in Section 4.2.2 of SSER 27. The 

proposed license condition in Section 4.2.2 of SSER 27 was “PAD4TCD may be used to 

establish core operating limits prior to the initial cycle, and prior to any remaining portion of the 

initial cycle. PAD4TCD may not be used to establish core operating limits prior to any reload 

cycle, and prior to any remaining portion of any reload cycle.” Upon further consideration the 

NRC staff has decided to modify the proposed license condition. The license condition is 

modified to provide clarity and does not change the intent of the condition that was initially 

proposed in Section 4.2.2 of SSER 27. The operating license for WBN, Unit 2 will include a 

license condition similar to the following.

Performance Analysis and Design (PAD) Thermal Conductivity Degradation (TCD) 

Proposed License Condition:

PAD4TCD may be used to establish core operating limits for Cycle 1 

only. PAD4TCD may not be used to establish core operating limits for 

subsequent reload cycles.
1-29

1.9.4 Electrical Design License Condition

The NRC staff proposed a license condition discussed in Section 8.2.2 of SSER 27. The 

operating license for WBN, Unit 2 will include a license condition similar to the following.

Bulletin 2012-01, “Design Vulnerability in Electrical Power System,” Proposed License 

Condition:

By December 31, 2017, TVA will report to the NRC that the actions to resolve the 

issues identified in Bulletin 2012-01, “Design Vulnerability in Electrical Power 

System,” have been implemented.

1.9.5 Fire Protection License Condition

The NRC staff proposed a license condition discussed in Appendix FF, Section 4.3, of 

SSER 29. The operating license for WBN, Unit 2 will include a license condition similar to the 

following.

Fire Detection Monitoring Panel Proposed License Condition:

By May 31, 2018, TVA shall ensure that a listing organization acceptable to the 

NRC (as the Authority Having Jurisdiction) determines that the fire detection 

monitoring panel in the main control room either meets the appropriate 

designated standards or has been tested and found suitable for the specified 

purpose.

1.9.6 Core Reload License Condition

The NRC staff proposed a license condition discussed in Section 15.3.1, of SSER 29. The 

operating license for WBN, Unit 2 will include a license condition similar to the following.

Core Reload License Condition:

TVA will verify for each core reload that the actions taken if FQW(Z) is not within 

limits will assure that the limits on core power peaking FQ(Z) remain below the 

initial total peaking factor assumed in the accident analyses.

1.13 Implementation of Corrective Action Programs and Special Programs

In 1985, TVA developed a corporate NPP that identified and proposed corrections to problems 

concerning the overall management of its nuclear program and a site-specific plan for WBN

entitled, “Watts Bar Nuclear Performance Plan.” TVA established 18 corrective action programs 

(CAPs) and 11 special programs (SPs) to address these concerns. 

SSER 21, Table 1.13.1 documented the status of NRC staff review of the CAPs and SPs. As 

indicated all items are resolved.
1-30

1.13.1 Corrective Action Programs

No. Title Program Review Status

(1) Cable Issues

a. Silicon Rubber Insulated Cable 

b. Cable Jamming

c. Cable Support in Vertical Conduit 

d. Cable Support in Vertical Trays

e. Cable Proximity to Hot Pipes

f. Cable Pull-Bys 

g. Cable Bend Radius 

h. Cable Splices 

i. Cable Sidewall Bearing Pressure 

j. Pulling Cables Through 90° Condulet and Flexible 

Conduit 

k. Computer Cable Routing System Software and 

Database Verification and Validation

Resolved

(2) Cable Tray and Tray Supports Resolved

(3) Design Baseline and Verification Program Resolved

(4) Electrical Conduit and Conduit Support Resolved

(5) Electrical Issues

a. Flexible Conduit Installations 

b. Physical Cable Separation and Electrical Isolation

c. Contact and Coil Rating of Electrical Devices

d. Torque Switch and Overload Relay Bypass Capability 

for Active Safety-Related Valves

e. Adhesive-Backed Cable Support Mount 

Resolved

(6) Equipment Seismic Qualification Resolved

(7) Fire protection Resolved

(8) Hanger and Analysis Update Program Resolved

(9) Heat Code Traceability Resolved

(10) Heating, Ventilation, and Air-Conditioning Duct and 

Duct Supports

Resolved

(11) Instrument Lines Resolved

(12) Prestart Test Program Plan Resolved

(13) Quality Assurance Records Resolved
1-31

No. Title Program Review Status

(14) Quality-List (Q-List) Resolved

(15) Replacement Items Program (Piece Parts) Resolved

(16) Seismic Analysis Resolved

(17) Vendor Information Program Resolved

(18) Welding Resolved

1.13.2 Special Programs

No. Title Program Review Status

(1) Concrete Quality Program Resolved

(2) Containment Cooling Resolved

(3) Detailed Control Room Design Review Resolved

(4) Environmental Qualifications Program Resolved

(5) Master Fuse List Resolved

(6) Mechanical Equipment Qualification Resolved

(7) Microbiologically Induced Corrosion Resolved

(8) Moderate Energy Line Break Flooding Resolved

(9) Radiation Monitoring System Resolved

(11) Use-As-Is Condition Adverse to Quality Resolved

1.14 Implementation of Applicable Bulletin and Generic Letter Requirements

From time to time, the NRC staff issues generic requirements or recommendations in the form 

of orders, bulletins (BLs), generic letters (GLs), regulatory issue summaries, and other 

documents to address certain safety and regulatory issues. These are generally termed 

“generic communications.”
1-32

The table below outlines the status of the resolution of the generic communications. 

Correspondence No. Title

(1) GL 1980-14

TVA Action:

NRC Action:

Light-Water Reactor Primary Coolant System Pressure 

Isolation Valves.

Submit Technical Specifications (TSs) for NRC Review.

Closed based on validation of TS 3.4.14. SSER 29, Section 

16 documents the NRC staff’s review of the WBN, Unit 2 

proposed TSs.

(2) GL 1980-77

TVA Action:

NRC Action:

Refueling Water Level - Technical Specifications Changes.

Submit Technical Specifications for NRC Review.

Closed based on validation of TS 3.9 –TS 3.9.5. SSER 29, 

Section 16 documents the NRC staff’s review of the WBN, 

Unit 2 proposed TSs.

(3) GL 1982-28

TVA Action:

NRC Action:

Inadequate Core Cooling Instrumentation System.

Closed.

Closed. Subsumed as part of NRC staff review of 

Instrumentation and Controls submitted April 8, 2010.

(4) GL 1983-28 Required Actions Based on Generic Implications of Salem 

Anticipated Transient without Scram Events (Screened into 

the Items 4 through 7).

(4.a) GL 1983-28 

(item 3.1)

TVA Action:

NRC Action:

Post-Maintenance Testing (reactor trip system components).

Submit Technical Specifications for NRC Review.

Closed based on validation of TS Bases 3.0.1. SSER 29, 

Section 16 documents the NRC staff’s review of the WBN, 

Unit 2 proposed TSs.
1-33

Correspondence No. Title

(4.b) GL 1983-28 (3.2)

TVA Action

NRC Action

Post-Maintenance Testing (All Surveillance Requirement 

Components).

Submit Technical Specifications and NRC Review.

Closed based on validation of TS Bases 3.0.1. SSER 29, 

Section 16 documents the NRC staff’s review of the WBN, 

Unit 2 proposed TSs.

(4.c) GL 1983-28 (4.2)

TVA Action

NRC Action

Reactor Trip System Reliability (Preventive Maintenance and 

Surveillance Program for Reactor Trip Breakers).

Submit Technical Specifications and NRC Review.

Closed based on validation of Item 17 of TS Table 3.3.1-1. 

SSER 29, Section 16 documents the NRC staff’s review of the 

WBN, Unit 2 proposed TSs.

(4.d) GL 1983-28 (4.5)

TVA Action

NRC Action

Reactor Trip System Reliability (Automatic Actuation of Shunt 

Trip Attachment).

Submit Technical Specifications and NRC Review.

Closed based on validation of Item 18 of TS Table 3.3.1-1. 

SSER 29, Section 16 documents the NRC staff’s review of the 

WBN, Unit 2 proposed TSs.

(8) GL 1986-09

TVA Action

NRC Action

Technical Resolution of Generic Issue B-59 (N-1) Loop 

Operation in BWRs and PWRs.

Submit Technical Specifications for NRC Review.

Closed based on validation of TS 3.4.4 - TS 3.4.8. SSER 29, 

Section 16 documents the NRC staff’s review of the WBN, 

Unit 2 proposed TSs.

(9) GL 1988-20

TVA Action

NRC Action

Individual Plant Examination for Severe Accident Vulnerability.

Closed.

Closed. NRC letter dated August 12, 2011 (ADAMS 

Accession No. ML111960228).
1-34

Correspondence No. Title

(10) GL 1988-20s1

TVA Action

NRC Action

Initiation of the Individual Plant Examination for Severe 

Accident Vulnerabilities — 10 CFR 50.54.

Closed.

Closed. NRC letter dated August 12, 2011 (ADAMS 

Accession No. ML111960228).

(11) GL 1988-20s2

TVA Action

NRC Action

Individual Plant Examination for Severe Accident Vulnerability. 

Accident Management Strategies for Consideration in the 

Individual Plant Examination Process.

Closed.

Closed. NRC letter dated August 12, 2011 (ADAMS 

Accession No. ML111960228).

(12) GL 1988-20s3

TVA Action

NRC Action

Individual Plant Examination for Severe Accident Vulnerability. 

Completion of Containment Performance Improvement 

Program and Forwarding of Insights for Use in the IPE for 

Severe Accident Vulnerabilities.

Closed.

Closed. NRC letter dated August 12, 2011 (ADAMS 

Accession No. ML111960228).

(13) GL 1988-20s4

TVA Action

NRC Action

Individual Plant Examination of External Events (IPEEE) for 

Severe Accident Vulnerabilities.

Closed.

Closed. NRC letter dated September 20, 2011 (ADAMS 

Accession No. ML111960300).

(14) GL 1988-20s5

TVA Action

NRC Action

Individual Plant Examination of External Events (IPEEE) for 

Severe Accident Vulnerabilities - 10 CFR 50.54(f).

Closed.

Closed. NRC letter dated September 20, 2011 (ADAMS 

Accession No. ML111960300).
1-35

Correspondence No. Title

(15) GL 1989-04

TVA Action

NRC Action

Guidelines on Developing Acceptable Inservice Testing 

Programs.

The proposed approach has been approved for WBN, Unit 1; 

the same approach was submitted for use on WBN, Unit 2, 

without change.

Closed. NRC letter dated October 21, 2014 (ADAMS 

Accession No. ML14289A222).

(16) GL 1989-21

TVA Action

NRC Action

Request for Information Concerning Status of Implementation 

of Unresolved Safety Issue Requirements.

TVA provided an updated status of unresolved safety issues 

on September 26, 2008, as supplemented on December 2, 

2010, and January 25, 2011.

Closed. See Appendix C of SSER 23.

(17) GL 1990-06

TVA Action

NRC Action

Resolution of Generic Issues 70, “PORV [power-operated 

relief valve] and Block Valve Reliability,” and 94, “Additional 

LTOP [low-temperature overpressure] Protection for PWRs.”

Submit Technical Specifications for NRC Review.

Closed based on validation of TS 3.4.11 - TS 3.4.12. SSER 

29, Section 16 documents the NRC staff’s review of the WBN, 

Unit 2 proposed TSs.

(18) GL 1992-08

TVA Action

NRC Action

Thermo-Lag 330-1 Fire Barriers.

The proposed approach has been approved for WBN, Unit 1; 

the same approach will be proposed for use on WBN, Unit 2, 

without change.

Closed. Inspection Report 2014612 (ADAMS Accession No. 

ML15034A211).

(19) GL 1995-03

TVA Action

NRC Action

Circumferential cracking of Steam Generator (SG) Tubes.

The proposed approach has been approved for WBN, Unit 1; 

the same approach was submitted for use on WBN, Unit 2, 

without change.

Closed. NRC Letter dated January 21, 2010 (ADAMS 

Accession No. ML093631061).
1-36

Correspondence No. Title

(20) GL 1995-05 

TVA Action

NRC Action

Voltage –Based Repair Criteria for Westinghouse Steam 

Generator Tubes affected by Outside Diameter Stress 

Corrosion Cracking.

The proposed approach has been approved for WBN, Unit 1; 

the same approach was submitted for use on WBN, Unit 2, 

without change.

Closed. NRC Letter dated January 21, 2010 (ADAMS 

Accession No. ML093631061).

(21) GL 1996-06

TVA Action

NRC Action

Assurance of Equipment Operability and Containment Integrity 

During Design-Basis Accident Conditions.

The proposed approach has been approved for WBN, Unit 1; 

the same approach will be proposed for use on WBN, Unit 2, 

without change.

Closed. NRC Letter dated January 21, 2010 (ADAMS 

Accession No. ML100130227).

(22) GL 1995-07

TVA Action

NRC Action

Pressure Locking and Thermal Binding of Safety-Related 

Power-Operated Gate Valves (Not identified in SSER 21 as 

“Open”).

The proposed approach has been approved for WBN, Unit 1; 

the same approach will be proposed for use on WBN, Unit 2, 

without change.

Closed. NRC letter dated August 12, 2010 (ADAMS 

Accession No. ML100190443).

(23) GL 1997-01

TVA Action

NRC Action

Degradation of Control Rod Drive Mechanism Nozzle and 

Other Vessel Closure Head Penetrations.

The proposed approach has been approved for WBN, Unit 1; 

the same approach will be proposed for use on WBN, Unit 2, 

without change.

Closed. NRC Letter dated June 30, 2010 (ADAMS Accession 

No. ML100539515).
1-37

Correspondence No. Title

(24) GL 1997-04

TVA Action

NRC Action

Assurance of Sufficient Net Positive Suction Head for 

Emergency Core Cooling and Containment Heat Removal 

Pumps Integrity During Design-Basis Accident Conditions.

The proposed approach has been approved for WBN, Unit 1; 

the same approach was submitted for use on WBN, Unit 2, 

without change.

Closed. NRC Letter dated February 18, 2010 (ADAMS 

Accession No. ML100200375).

(25) GL 1997-05

TVA Action

NRC Action

SG Tube Inspection Techniques.

The proposed approach has been approved for WBN, Unit 1; 

the same approach was submitted for use on WBN, Unit 2, 

without change.

Closed. NRC Letter dated January 21, 2010 (ADAMS 

Accession No. ML093631061).

(26) GL 1997-06

TVA Action

NRC Action

Degradation of SG Internals.

The proposed approach has been approved for WBN, Unit 1; 

the same approach was submitted for use on WBN, Unit 2, 

without change.

Closed. NRC Letter dated January 21, 2010 (ADAMS 

Accession No. ML093631061).

(27) GL 1998-02

TVA Action

NRC Action

Loss of Reactor Coolant Inventory and Associated Potential 

for Loss of Emergency Mitigation Functions While in a 

Shutdown Condition.

The proposed approach has been approved for WBN, Unit 1; 

the same approach will be proposed for use on WBN, Unit 2, 

without change.

Closed. NRC Letter dated May 11, 2010 (ADAMS Accession 

No. ML101200155).
1-38

Correspondence No. Title

(28) GL 1998-04

TVA Action

NRC Action

Potential for Degradation of the ECCS [Emergency Core 

Cooling System] and the Containment Spray System after a 

LOCA because of Construction and Protective Coating 

Deficiencies and Foreign Material in Containment.

The proposed approach has been approved for WBN, Unit 1; 

the same approach was submitted for use on WBN, Unit 2, 

without change.

Closed. NRC Letter dated February 1, 2010 (ADAMS 

Accession No. ML100260594).

(29) GL 2003-01

TVA Action

NRC Action

Control Room Habitability.

No action or documentation is provided to show the NRC staff 

has reviewed the item for WBN, Unit 2, and the resolution is 

through submittal of a technical specification.

Closed. NRC Letter dated February 1, 2010 (ADAMS 

Accession No. ML100270076).

(30) GL 2004-01

TVA Action

NRC Action

Requirements for SG Tube Inspection.

The proposed approach has been approved for WBN, Unit 1; 

the same approach was submitted for use on WBN, Unit 2, 

without change.

Closed. NRC Letter dated January 21, 2010 (ADAMS 

Accession No. ML093631061).

(31) GL 2004-02

TVA Action

NRC Action

Potential Impact of Debris Blockage on Emergency 

Recirculation during Design-Basis Accidents at PWRs.

The proposed approach has been approved for WBN, Unit 1; 

the same approach was submitted for use on WBN, Unit 2, 

without change.

Closed. NRC Letter dated September 18, 2014 (ADAMS 

Accession No. ML14163A658).
1-39

Correspondence No. Title

(32) GL 2006-01

TVA Action

NRC Action

SG Tube Integrity and Associated Technical Specifications.

No action or documentation is provided to show the NRC staff 

has reviewed the item for WBN, Unit 2, and the resolution is 

through submittal of a technical specification.

Closed. NRC Letter dated January 21, 2010 (ADAMS 

Accession No. ML093631061) (See Appendix HH).

(33) GL 2006-02

TVA Action

NRC Action

Grid Reliability and the Impact on Plant Risk and the 

Operability of Offsite Power.

The proposed approach has been approved for WBN, Unit 1; 

the same approach was submitted for use on WBN, Unit 2, 

without change.

Closed. NRC Letter dated January 21, 2010 (ADAMS 

Accession No. ML093631061) (See Appendix HH Open 

Item 6). Staff has reviewed Revision I to the proposed 

technical specifications and found that Technical Specification 

Task Force (TSTF) 449 has been incorporated.

(34) GL 2006-03

TVA Action

NRC Action

Potentially Nonconforming Hemyc and MT Fire Barrier 

Configurations.

The proposed approach has been approved for WBN, Unit 1; 

the same approach was submitted for use on WBN, Unit 2, 

without change.

Closed. NRC Letter February 25, 2010 (ADAMS Accession 

No. ML100470398).

(35) GL 2007-01

TVA Action

NRC Action

Inaccessible or Underground Power Cable Failures that 

Disable Accident Mitigation Systems or Cause Plant 

Transients.

The proposed approach has been approved for WBN, Unit 1; 

the same approach was submitted for use on WBN, Unit 2, 

without change.

Closed. NRC Letter dated January 26, 2010 (ADAMS 

Accession No. ML100120052).
1-40

Correspondence No. Title

(36) GL 2008-01

TVA Action

NRC Action

Managing Gas Accumulation in Emergency Core Cooling, 

Decay Heat Removal, and Containment Spray Systems.

TVA submitted the information requested by the GL.

Closed. NRC letter dated August 23, 2011 (ADAMS 

Accession No. ML112232205).

(37) BL 1992-01 and 

Supplement 1

TVA Action

NRC Action

Failure of Thermo-Lag 330 Fire Barrier System to Perform its 

Specified Fire Endurance Function.

The proposed approach has been approved for WBN, Unit 1; 

the same approach will be proposed for use on WBN, Unit 2, 

without change.

Closed. Inspection Report 2014612 (ADAMS Accession No. 

ML15034A211).

(38) BL 1996-01

TVA Action

NRC Action

Control Rod Insertion Problems (PWR)

The proposed approach has been approved for WBN, Unit 1; 

the same approach was submitted for use on WBN, Unit 2, 

without change.

Closed. NRC letter dated May 3, 2010 (ADAMS Accession 

No. ML101200035) required Confirmatory Action (See 

Appendix HH Open Items 5 and 8). By letter dated July 30, 

2012, TVA provided the information in the Confirmatory Action 

and the NRC staff verified the information and has closed 

Appendix HH Open Items 5 and 8. 

(39) BL 1996-02

TVA Action

NRC Action

Movement of Heavy Loads Over Spent Fuel, Over Fuel In the 

Reactor Core, or Over Safety-Related Equipment.

The proposed approach has been approved for WBN, Unit 1; 

the same approach was submitted for use on WBN, Unit 2, 

without change.

Closed. NRC Letter dated March 4, 2010 (ADAMS Accession 

No. ML100480062).
1-41

Correspondence No. Title

(40) BL 2001-01

TVA Action

NRC Action

Circumferential Cracking of Reactor Pressure Vessel (RPV) 

Head Penetration Nozzles.

The proposed approach has been approved for WBN, Unit 1; 

the same approach was submitted for use on WBN, Unit 2, 

without change.

Closed. NRC Letter dated June 30, 2010 (ADAMS Accession 

No. ML100539515).

(41) BL 2002-01

TVA Action

NRC Action

RPV Head Degradation and Reactor Coolant Pressure 

Boundary Integrity.

The proposed approach has been approved for WBN, Unit 1; 

the same approach was submitted for use on WBN, Unit 2, 

without change.

Closed. NRC Letter dated June 30, 2010 (ADAMS Accession 

No. ML100539515).

(42) BL 2002-02

TVA Action

NRC Action

RPV Head and Vessel Head Penetration Nozzle Inspection 

Program.

The proposed approach has been approved for WBN, Unit 1; 

the same approach was submitted for use on WBN, Unit 2, 

without change.

Closed. NRC Letter dated June 30, 2010 (ADAMS Accession 

No. ML100539515).

(43) BL 2003-02 

TVA Action

NRC Action

Leakage from RPV Lower Head Penetrations and Reactor 

Coolant Pressure Boundary Integrity.

The proposed approach has been approved for WBN, Unit 1; 

the same approach was submitted for use on WBN, Unit 2, 

without change.

Closed. NRC Letter dated January 21, 2010 (ADAMS 

Accession No. ML093631061).
1-42

Correspondence No. Title

(44) BL 2004-01

TVA Action

NRC Action

Inspection of Alloy 82/182/600 Materials Used in the 

Fabrication of Pressurizer Penetrations and Steam Space 

Piping Connections at PWRs.

The proposed approach has been approved for WBN, Unit 1; 

the same approach was submitted for use on WBN, Unit 2, 

without change.

Closed. NRC letter dated August 4, 2010 (ADAMS Accession 

No. ML102080017).

(45) BL 2007-01

TVA Action

NRC Action

Security Officer Attentiveness.

The proposed approach has been approved for WBN, Unit 1; 

the same approach will be proposed for use on WBN, Unit 2, 

without change.

Closed. NRC letter dated March 25, 2010 (ADAMS Accession 

No. ML100770549).

(46) BL 2011-01

TVA Action

NRC Action

Mitigating Strategies

The proposed approach has been approved for WBN, Unit 1; 

an updated approach will be proposed for use on WBN, 

Unit 2, without change.

Closed. NRC letter dated October 14, 2015 (ADAMS 

Accession No. ML 15280A085)

(47) BL 2012-01

TVA Action

NRC Action

Design Vulnerability In Electric Power System

TVA submitted a letter on September 3, 2014 (ADAMS 

Accession No. ML14247A231) providing wording for the FSAR 

and proposed a license condition.

Resolved, see license condition described in SER 

Section 1.9.4.

NUREG-0737, TMI Action Items (TVA letter dated September 14, 1981, applies

to all of the following NUREG-0737 issues):
1-43

Correspondence No. Title

(48) NUREG-0737 

Item I.B.1.2

TVA Action

NRC Action

Independent Safety Engineering Group.

The proposed approach has been approved for WBN, Unit 1; 

the same approach will be proposed for use on WBN, Unit 2, 

without change.

Closed. TVA letter dated August 26, 1999 (ADAMS Accession 

No. ML082350970) and SSER 22, Section 17.

(49) NUREG-0737 

Item I.D.1

TVA Action

NRC Action

Control Room Design Review (CRDR).

The proposed approach has been approved for WBN, Unit 1; 

the same approach will be proposed for use on WBN, Unit 2, 

without change.

Closed in SSER 22, Section 18.2. 

(50) NUREG-0737 

Item II.B.3

TVA Action

NRC Action

Post-accident Sampling.

No action or documentation is provided to show the NRC staff 

has reviewed the item for WBN, Unit 2, and the resolution is 

through submittal of a technical specification.

Closed in SSER 24, Section 9.3.2. 

(51) NUREG-0737 

Item II.E.4.2

TVA Action

NRC Action

Containment Isolation Dependability.

No action or documentation is provided to show the NRC staff 

has reviewed the item for WBN, Unit 2, and the resolution is 

through submittal of a technical specification.

Closed. Inspection Report 05000391/2011605 (ADAMS 

Accession No. ML112201418) and SSER 29, Section 16
1-44

Correspondence No. Title

(52) NUREG-0737 

Item II.F.2

TVA Action

NRC Action

Instrumentation for Detection of Inadequate Core-Cooling.

Closed.

Closed. See SSER 25 and SSER 26, Section 7.5.2.2. Open 

Items 72, 95, 96, 97, 99, 100, 102, 103, 104, 106, 107, and 

109 were closed in SSER 25, Section 7.5.2.2. Open items 94, 

98, 101, 105, 108, 110, and 111 were closed in SSER 26, 

Section 7.5.2.2.

(53) NUREG-0737 

Item II.K.3.3

TVA Action

NRC Action

Reporting Safety Valve/Reactor Vessel Failures/Challenges. 

No action or documentation is provided to show the NRC staff 

has reviewed the item for WBN, Unit 2, and the resolution is 

through submittal of a technical specification.

Closed in SSER 22, Section 13.5.3.

(54) NUREG-0737 

Item II.K.3.10

TVA Action

NRC Action

Anticipatory Trip at High Power.

No action or documentation is provided to show the NRC staff 

has reviewed the item for WBN, Unit 2, and the resolution is 

through submittal of a technical specification.

Closed in SSER 23, Section 7.8.4.

(55) NUREG-0737 

Item III.D.1.1

TVA Action

NRC Action

Primary Coolant Outside Containment.

No action or documentation is provided to show the NRC staff 

has reviewed the item for WBN, Unit 2, and the resolution is 

through submittal of a technical specification.

Closed in SSER 29, Section 16.

(56) NUREG-0737 

Item III.D.3.4

TVA Action

NRC Action

Control-Room Habitability.

The proposed approach has been approved for WBN, Unit 1; 

the same approach will be proposed for use on WBN, Unit 2, 

without change.

Closed in SSER 22, Section 6.4.
1-45

Correspondence No. Title

(57) Office of Inspection 

and Enforcement 

Bulletin (IEB) 75-08

TVA Action

NRC Action

PWR Pressure Instrumentation.

The item has been approved either for both units at WBN or 

explicitly for WBN, Unit 2; however, a change to the original 

approval requires submittal of the technical specifications and 

NRC staff review.

Closed. The ability to continuously monitor pressure and 

temperature is noted in the UFSAR and is consistent with the 

licensing basis for WBN, Unit 1. Inspection activities verified 

that WBN has instrumentation in place to monitor pressure 

and temperature to close this Bulletin (Inspection Report 50-

391/85-08, ADAMS Accession No. ML082190701). 

(58) IEB 77-04

TVA Action

NRC Action

Calculation Error Affecting Performance of a System for 

Controlling pH of Containment Sump Water Following a 

LOCA. 

The item has been approved either for both units at WBN or 

explicitly for WBN, Unit 2; however, a change to the original 

approval requires submittal of the technical specifications and 

NRC staff review.

Closed. Item was closed in Inspection Report 10-391/78-09 

(ADAMS Accession No. ML071790565). WBN, Unit 2 has an 

ice condenser containment and utilizes sodium tetraborate in 

the ice for containment sump pH control. TVA’s response 

(ADAMS Accession No. ML073020061) to this Bulletin 

reported that limiting concentrations are included in the plant 

TSs. Surveillance Requirement 3.6.11.5 requires verification 

that the boron concentration is within a specified range and is 

consistent with the licensing basis for WBN, Unit 1.

Fukushima-Related Orders (NRC letters dated March 12, 2012):

(59) EA-12-049

TVA Action

NRC Action

Mitigating Strategies for Beyond-Design-Basis External Events

(ADAMS Accession No. ML12054A735)

Provide final compliance letter by December 17, 2014.

Closed in SSER 28, Appendix JJ.
1-46

Correspondence No. Title

(60) EA-12-051

TVA Action

NRC Action

Reliable Spent Fuel Pool Instrumentation

(ADAMS Accession No. ML12054A679)

Compliance letter sent October 29, 2014.

Closed in SSER 28, Appendix JJ.

1.17 Financial Assurance for Decommissioning

The NRC’s regulations at 10 CFR, 50.75, “Reporting and Recordkeeping for Decommissioning 

Planning, and 50.33(k), require, in part, an applicant for, or holder of, an operating license under 

Part 50 to provide a report certifying financial assurance for decommissioning in an amount that 

may be more, but not less, than the NRC minimum amount specified in 10 CFR 50.75(c)(1), 

adjusted using a rate at least equal to that stated in 10 CFR 50.75(c)(2).

By letter dated March 4, 2009 (ADAMS Accession No. ML090700378), “Watts Bar Nuclear 

Plant, Unit 2—Operating License Application Update,” Tennessee Valley Authority (TVA) 

submitted an update to its application that included certification of financial assurance for 

decommissioning of WBN, Unit 2, in an amount of $400.3 million (2008 dollars). In its submittal, 

TVA stated that it will establish a decommissioning trust fund within its Master Decommissioning 

Trust Agreement for WBN, Unit 2, before fuel load; the trustee of the fund will be New York 

Mellon Corporation. The NRC staff independently calculated the minimum funding amount 

required under 10 CFR 50.75(c)(1), as adjusted under 10 CFR 50.75(c)(2), and found the 

applicant's estimate for decommissioning funding assurance to be acceptable. 

By letter dated August 26, 2009 (ADAMS Accession No. ML092470504), “Additional Information 

Regarding Financial Information Related to Operating License Application Update,” TVA

provided its response to the staff’s request for additional information (RAI). The response 

stated TVA’s intent to use an external sinking fund as the method to provide financial assurance 

for decommissioning. TVA is an electric utility, that sets its own rates and has the ability to use 

an external sinking fund as its sole means of providing decommissioning funding assurance, 

under 10 CFR 50.75(e)(ii)(A). An external sinking fund is a fund established and maintained by 

setting funds aside periodically in an account segregated from licensee assets and outside the 

administrative control of the licensee and its subsidiaries or affiliates in which the total amount of 

funds would be sufficient to pay decommissioning costs at the time permanent termination of 

operations is expected. An external sinking fund may be in the form of a trust, escrow account, 

or Government fund, with payment by certificate of deposit, deposit of Government or other 

securities, or other method acceptable to the NRC. The NRC staff reviewed TVA's plan to use 

an external sinking fund to provide decommissioning funding assurance and found that TVA has 

complied with the requirements in 10 CFR 50.75. The NRC staff’s review was published in 

NUREG-0847, “Safety Evaluation Report Related to the Operation of Watts Bar Nuclear Plant, 

Unit 2,” Supplement 22, dated February 2011 (ADAMS Accession No. ML110390197).

By letter dated September 10, 2015, “Decommissioning Funding Update,” TVA provided an 

update to its Decommissioning Funding Plan (ADAMS Accession No. ML15253A867) for WBN, 

Unit 2, that included a revised decommissioning estimate in the amount of $517.7 million (2015 

dollars) using the adjustment factors in 10 CFR 50.75(c)(2) required for updating the NRC 

minimum formula amount. The NRC staff found the decommissioning cost estimate for WBN, 
1-47

Unit 2, to be reasonable because WBN, Unit 2, is a PWR and the decommissioning cost 

estimate was independently calculated by staff using the minimum funding amount required 

under 10 CFR 50.75(c)(1), as adjusted under 10 CFR 50.75(c)(2). In its submittal 

(Enclosure 1), TVA also supplied evidence that it has executed the Sixth Amendment to the 

Master Decommissioning Trust Agreement on June 23, 2015; this established a separate fund 

for WBN, Unit 2, including an initial deposit into the fund in the amount of $3,402,638 on 

September 8, 2015, and a schedule of the anticipated annual deposits of funding amounts into 

the trust. TVA has chosen to use an external sinking fund to provide decommissioning funding 

assurance for WBN, Unit 2, as allowed in the regulations at 10 CFR 50.75(e)(1)(ii). This 

approach is consistent with that used for TVA’s other nuclear units as discussed in TVA’s most 

recent Decommissioning Funding Status Report. The NRC staff has reviewed this information 

and finds that TVA’s submittal complies with the regulations at 10 CFR 50.75(c) and 10 CFR 

50.75(e)(1)(ii), which requires TVA to estimate needed costs for decommissioning and set aside 

funds periodically in an account segregated from TVA assets for which the total amount of funds 

will be sufficient to pay WBN, Unit 2, decommissioning costs at the time of permanent cessation 

of operations is expected.

Based on staff’s review of TVA’s submittal dated September 10, 2015, NRC staff concludes that 

TVA has complied with the decommissioning funding assurance reporting requirements in 

10 CFR 50.33(k) and the decommissioning funding assurance requirements in 10 CFR 50.75.

1.18 NRC Staff Conclusion

On the basis of the NRC staff’s review of the application, as documented in the SER and its 

supplements, as well as the NRC staff’s inspections, the NRC staff concludes that the 

requirements of 10 CFR 50.57, “Issuance of Operating License,” sections 50.57(a)(1) to (a)(6), 

have been met for WBN, Unit 2, as summarized below. 

(1) Construction of the facility has been substantially completed, in conformity with the

construction permit and the application as amended, the provisions of the Atomic Energy

Act (Act), and the rules and regulations of the Commission; 

As discussed in Appendix GG “Final Memorandum on Facility Completion In Accordance 

With Inspection Manual Chapter 94302” of this SSER, “Region II inspection activities 

support that the construction of the WBN, Unit 2 facility has been substantially 

completed in conformity with the construction permit and the application as amended. 

This conclusion is based on completion of the construction inspections necessary to 

support this finding.” 

Based on the NRC staff’s review of the application as documented in the SER and its 

supplements, and the results of the NRC staff’s inspections, the NRC staff finds that that 

construction of WBN, Unit 2 has been substantially completed in conformity with the 

construction permit and the application as amended, the provisions of the Atomic Energy 

Act, and the Commission’s regulations.

(2) The facility will operate in conformity with the application as amended, the provisions of 

the Act, and the rules and regulations of the Commission;

The NRC staff reviewed information provided by the applicant to ensure that the plant 

will operate in conformity with the application as amended, the applicable provisions of 
1-48

the Atomic Energy Act of 1954, as amended, and the applicable rules and regulations of 

the Commission.

As discussed in Appendix GG of this SSER, “Region II inspection activities support that 

WBN Unit 2 facility will operate in conformity with the application as amended. This is 

based on completion of the pre-operational testing inspections identified in IMC 2513, 

Appendix A, Light Water Reactor - Preoperational Testing Phase, that are necessary to 

support this finding.” 

Based on the NRC staff’s review of the application as documented in the SER and its 

supplements, and the results of the NRC staff’s inspections, the NRC staff finds that 

WBN, Unit 2 will operate in conformity with the application as amended, the provisions of 

the Atomic Energy Act, and the Commission’s regulations.

(3) There is reasonable assurance that the activities authorized by the operating license can 

be conducted without endangering the health and safety of the public, and that such 

activities will be conducted in compliance with the regulations in 10 CFR Ch. I; 

The NRC staff reviewed the application to assure that activities authorized by the license 

will not endanger the health and safety of the public. Specifically, the NRC staff 

evaluated the applicant’s analysis and conclusions about site-specific conditions, 

including the geography and demography of the site; nearby industrial, transportation, 

and military facilities; site meteorology; site hydrology; and site geology, seismology, and 

geotechnical engineering to ensure that issuance of the license will not endanger public 

health and safety. The review also evaluated the design of structures, components, 

equipment, and systems to ensure safe operation. 

The review confirmed that radiological releases and human doses during both normal 

operation and accident scenarios will remain within regulatory limits, which supports the 

NRC staff’s conclusion that issuance of the license will not endanger the health and 

safety of the public. 

As discussed in Appendix GG of this SSER, “Region II inspection activities support that 

there is reasonable assurance that activities authorized by the license will be conducted 

according to applicable regulations. This is based on completion of operational 

preparedness inspections items that were necessary to support this finding and 

conclusions from the Operational Readiness Assessment Team inspection (ORAT). The 

ORAT inspection (ADAMS Accession No. ML15226A212) concluded TVA adequately 

demonstrated the readiness of the facility and staff to safely begin operating the WBN 

Unit 2 facility.” 

Based on the NRC staff’s review of the application as documented in the SER and its 

supplements, as well as the NRC staff’s inspections, the NRC staff finds that the 

activities authorized by the operating license can be conducted without endangering the 

health and safety of the public, and that such activities will be conducted in compliance 

with the applicable regulations.
1-49

(4) The applicant is technically and financially qualified to engage in the activities authorized 

by the operating license in accordance with the regulations in 10 CFR Ch. I; 

The NRC staff reviewed information provided by the applicant regarding technical and 

financial qualifications. 

a. Technical Qualification. The NRC staff reviewed information provided by the applicant

regarding its technical qualifications. The review included an evaluation of the

applicant’s operating experience, organizational structure, and QA program. The review

included the fact that the applicant operates WBN, Unit 1, which is of similar design to

WBN, Unit 2. The applicant holds a 10 CFR Part 50 license for WBN, Unit 1 and has

demonstrated its ability to build and operate a nuclear power reactor. The applicant has

demonstrated the ability to choose and manage the oversight of nuclear steam supply

system vendors, architect-engineers, and constructors of nuclear-related work. The

NRC staff’s review of the applicant’s organizational structure concluded that its

management, technical support, and operating organizations are acceptable. The NRC

staff reviewed the QA program and found it acceptable. This QA program includes

requirements that will be implemented by the applicant’s engineering, procurement, and

construction contractor. The NRC staff’s evaluation of this information appears in

Chapter 13 and 17 of the SER and its supplements. Based on the NRC staff’s

evaluation of the applicant’s experience with building and operating a nuclear power

plant, its operating organization, and its QA program, the staff finds that the applicant is

technically qualified to engage in the activities authorized by the operating license in

accordance with the regulation.

b. Financial Qualifications. The NRC staff reviewed information provided by the applicant

about financial qualifications. The review included an evaluation of the financial

qualifications, decommissioning funding assurance, foreign ownership, and nuclear

insurance and indemnity. Applicable regulations and guidance considered by the NRC

staff included 10 CFR Part 140, “Financial Protection Requirements and Indemnity

Agreements”; 10 CFR 50.54(w), 10 CFR 50.33; “Contents of Applications; General

Information”; and NUREG-1577, “Standard Review Plan on Power Reactor Licensee

Financial Qualifications and Decommissioning Funding Assurance.”

The NRC staff’s evaluation of this information appears in Chapters 1, 21 and 22 of the 

SER and its supplements. As discussed in Section 21 of SSER 22, published 

February 2011 (ADAMS Accession No. ML110390197), the NRC staff concluded that 

TVA is an electric utility as defined by 10 CFR 50.2. Therefore, TVA is not subject to the 

requirements of 10 CFR 50.33(f). Based on the financial information provided by the 

applicant, the NRC staff concludes that the TVA has demonstrated that it possesses or 

has access to the financial resources necessary to meet estimated operation and 

construction costs. Therefore, the NRC staff concludes that TVA, is financially qualified 

to operate WBN, Unit 2 and to engage in the activities authorized by an operating

license in accordance with the regulations. 

(5) The applicable provisions of 10 CFR Part 140 of this chapter have been satisfied; 

As discussed in Section 22 of this SSER, the NRC staff has determined that the 

applicant has satisfied the applicable requirements in 10 CFR Part 140.
1-50

(6) The issuance of the license will not be inimical to the common defense and security or to 

the health and safety of the public.

The NRC staff reviewed the application to assure that issuance of the license will not be 

inimical to the common defense and security or to public health and safety. Specifically, 

the NRC staff evaluated the applicant’s analysis and conclusions about site-specific 

conditions, including the geography and demography of the site; nearby industrial, 

transportation, and military facilities; site meteorology; site hydrology; and site geology, 

seismology, and geotechnical engineering to ensure that issuance of the license will not 

be inimical to public health and safety. The review also evaluated the design of 

structures, components, equipment, and systems to ensure safe operation. 

The review confirmed that radiological releases and human doses during both normal 

operation and accident scenarios will remain within regulatory limits, which supports the 

NRC staff’s conclusion that issuance of the license will not be inimical to public health 

and safety. The review determined that the physical security to be implemented at the 

site is adequate to protect the facility, which supports the NRC staff’s conclusion that 

issuance of the license will not be inimical to the common defense and security.

On the basis of the NRC staff’s review of the application, as amended, as discussed in 

the SER and its supplements, operation of WBN, Unit 2 will not be inimical to the 

common defense and security or to public health and safety.
3-1 

3 DESIGN CRITERIA – STRUCTURE, COMPONENTS, EQUIPMENT, 

AND SYSTEMS

3.8 Design of Seismic Category I Structures

3.8.3 Other Seismic Category I Structures

The U.S. Nuclear Regulatory Commission (NRC) staff has reviewed the changes made in the 

Watts Bar Nuclear Plant (WBN), Unit 2 Final Safety Analysis Report related to this section up 

through Amendment 114 and found no new substantial information. The changes that were 

made were found to be editorial in nature such that the NRC staff’s evaluations in the SER, and 

SSERs 14 and 16 still remain valid. Therefore, the NRC staff considers this section to be 

closed.

3.9 Mechanical Systems and Components

3.9.3 ASME Code Class 1, 2, and 3 Components, Component Structures, and Core 

Support Structures

3.9.3.3 Design and Installation of Pressure Relief Devices

By letter dated July 29, 2015 (Agencywide Documents Access and Management System 

(ADAMS) Accession No. ML15210A140), Tennessee Valley Authority (TVA), informed the NRC 

that it has chosen to use the EPRI motor-operated valve (MOV) performance prediction 

methodology (PPM) in lieu of dynamic testing of the WBN, Unit 2 pressurizer power-operated 

relief valve (PORV) block valves.

The NRC regulations require that MOVs important to safety be treated in a manner that provide

assurance of their intended performance. Criterion 1 to Appendix A, “General Design Criteria

for Nuclear Power Plants,” to Part 50 of Title 10 of the Code of Federal Regulations (10 CFR

Part 50) states, in part, that structures, systems, and components important to safety shall be

designed, fabricated, erected, and tested to quality standards commensurate with the

importance of the safety functions to be performed. The quality assurance program to be

applied to safety-related components is described in Appendix B, “Quality Assurance Criteria for

Nuclear Power Plants and Fuel Reprocessing Plants,” to 10 CFR Part 50. In 10 CFR 50.55a,

the NRC requires licensees to establish inservice testing programs in accordance with the 

American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of

Nuclear Power Plants. 

In response to concerns regarding MOV performance, on June 28, 1989, the NRC staff issued 

Generic Letter (GL) 89-10, "Safety-Related Motor-Operated Valve Testing and Surveillance,"

(ADAMS Accession No. ML031150300) which requested that nuclear power plant licensees

and construction permit holders ensure the capability of MOVs in safety-related systems to 

perform their intended functions by reviewing MOV design bases, verifying MOV switch 

settings initially and periodically, testing MOVs under design-basis conditions where

practicable, improving evaluations of MOV failures and necessary corrective action, and 

trending MOV problems. The GL also stated that if testing at design-basis conditions is 

precluded by the existing plant configuration, an explanation should be documented including 

a description of the alternatives to the design-basis differential pressure testing or flow testing 
3-2 

that will be used. The NRC staff requested that licensees complete the GL 89-10 program

within approximately three refueling outages or 5 years from the date of issuance of the GL. 

Permit holders were requested to complete the GL 89-10 program before plant startup or in

accordance with the above schedule, whichever was later. TVA's six month response to GL 

89-10 stated that TVA planned to meet the GL 89-10 recommendations and to comply with the 

5-year implementation schedule, i.e., by June 28, 1994 (ADAMS Accession No. 

ML082320614).

On September 18, 1996, (ADAMS Accession No. ML031110010), the NRC staff issued 

GL 96-05, “Periodic Verification of Design-Basis Capability of Safety-Related Motor-Operated

Valves,” requesting each licensee to establish a program, or ensure the effectiveness of its

current program, to verify on a periodic basis that safety-related MOVs continue to be capable

of performing their safety functions within the current licensing bases of the facility.

By March 15, 1996, the NRC staff provided the results of its review of EPRI topical report TR103237, “EPRI MOV Performance Prediction Program” Revision 1, (ADAMS Accession 

No. ML15142A761). The letter stated that the staff considers the methodology to be an 

acceptable way, with conditions and limitations, to predict the thrust and torque required to 

operate gate, globe, and butterfly valves within the scope of the MOV program.

By letter dated February 24, 2009, the NRC staff provided its final safety evaluation of the EPRI 

topical report TR-103237, “EPRI MOV Performance Prediction Program” Revision 2, (ADAMS 

Accession No. ML090430444). TR-103237 Revision 2 included several improvements to the 

methodology along with software upgrades versions 3.1, 3.2, and 3.3. The NRC staff review 

concluded that the method and software improvements were acceptable for referencing in 

licensing applications as discussed in the safety evaluation enclosed with the letter and that the 

safety evaluation defines the basis for the NRC staff’s acceptance of the topical report. 

As noted in Section 1 of this Supplemental Safety Evaluation Report (SSER), TVA did not 

compete WBN, Unit 2 at the time that WBN, Unit 1 was licensed. By letter dated August 3, 

2007 (ADAMS Accession No. ML072190047), TVA notified the NRC of its intention to reactive 

the construction activities for WBN, Unit 2. Subsequently, by letter dated January 29, 2008

(ADAMS Accession No. ML080320443), TVA submitted the Regulatory Framework for the

Completion of Construction and Licensing Activities for WBN, Unit 2. The January 29, 2008 

letter contained three commitments related to the testing of MOVs: 

• Commitment 18, GL 89-10, "Safety Related Motor-Operated Valve (MOV) Testing

and Surveillance;"

• Commitment 19, GL 96-05, "Periodic Verification (PV) of Design Basis Capability of

Safety-Related MOVs;" and

• Commitment 31, [NUREG-0737, Item] II.D.1, "Relief and Safety Valve Test

Requirements."

The January 29, 2008 letter provided details regarding each of these commitments. 

Commitment 18 was to implement pressure testing and a surveillance program for safetyrelated MOVs, satisfying the intent of GL 89-10. Commitment 19 was to implement the 
3-3

Joint Owner's Group recommended GL 96-05 MOV PV program and begin testing during the

first refueling outage after startup. Commitment 31 contained four activities:

1. Testing of relief and safety valves;

2. Reanalysis of fluid transient loads for pressurizer relief and safety valve supports and

any required modifications;

3. Modifications to pressurizer safety valves, PORVs, PORV block valves and

associated piping; and

4. Change motor operated block valves.

By letter dated July 29, 2015, TVA stated that:

In consideration of the performance of Hot Functional Testing at WBN, Unit 2 in

2015, TVA reviewed the test conditions associated with the dynamic testing of the

PORV block valves. TVA determined that differential pressure testing creates an

equipment and personnel hazard. Specifically, the equipment hazards are

associated with the potential for overfilling the Pressurizer Relief Tank (PRT),

subsequently failing the PRT rupture disks, thus creating a risk for personnel

injury. Based on this review, TVA decided to use the Electric Power Research 

Institute (EPRI) MOV Performance Prediction Methodology (PPM) analysis (EPRI

TR-103237, Revision 1) (ML15142A761) in lieu of dynamic differential pressure

testing for the WBN, Unit 2 Pressurizer PORV block valves (2-FCV-068-0332 and

-0333). 

TVA used EPRI PPM software version 3.3 for this analysis.

With the limitations and conditions related to the specific application of the methodology 

described in the NRC staff’s safety evaluations in response to TR-103237, the NRC staff 

concludes that the EPRI MOV Performance Prediction Program provides an acceptable 

methodology to predict the thrust and torque required to operate gate, globe, and butterfly 

valves within the scope of the EPRI program, and to bound the effects of load sensitive behavior 

on motor-actuator thrust output in lieu of dynamic valve testing. The NRC staff reviewed WBN, 

Unit 2 pressurizer PORV block valve make, model, type, and system parameters and concluded 

that the EPRI PPM program is applicable to the WBN, Unit 2 PORV block valves and that the 

licensee should consider the limitations and conditions discussed in the NRC staff’s safety 

evaluation report for TR-103237. Therefore, the NRC staff finds the application of the EPRI 

PPM program for the pressurizer PORV block valves at WBN, Unit 2 acceptable.

6-1

6 ENGINEERED SAFETY FEATURES

6.1 Engineered Safety Feature Materials

6.1.3 Postaccident Emergency Cooling Water Chemistry 

The Tennessee Valley Authority (TVA or the applicant) provided Amendment 114 to the final 

safety analysis report (FSAR) by letter dated September 11, 2015 (Agencywide Documents 

Access and Management System (ADAMS) Accession No. ML15279A332). The 

Amendment 114 of the FSAR changed the minimum amount of ice assumed in the post–loss 

of coolant accident (LOCA) sump pH analysis from 2.26×106 lb to 2.585×106 lb. This FSAR 

revision was based on a revised LOCA containment integrity analysis of record (AOR) 

documented in FSAR Amendment 113. The U.S. Nuclear Regulatory Commission (NRC) staff’s 

evaluation of the revised LOCA containment integrity analysis is provided in Section 6.2.1 of this 

supplemental safety evaluation report (SSER).

6.2 Containment Systems

6.2.1 Containment Functional Design

The containment functional design should meet the requirements of the following NRC 

regulations: 

10 CFR 50 Appendix A-

• GDC 5 as it relates to sharing of structures, systems, and components important to 

safety shall not be shared among nuclear power units unless it can be shown that such 

sharing will not significantly impair their ability to perform their safety functions, including, 

in the event of an accident in one unit, an orderly shutdown and cooldown of the 

remaining units.

• GDC 16 as it relates to the containment and associated systems establishing a leak-tight 

barrier against the uncontrolled release of radioactivity to the environment and assuring 

that the containment design conditions important to safety are not exceeded for as long 

as the postulated accident requires.

• GDC 38 as it relates to the containment heat removal system safety function which shall 

reduce rapidly, consistent with the functioning of other associated systems, the 

containment pressure and temperature following any Loss-of-Coolant Accident (LOCA) 

and to maintain them at acceptably low levels.

• GDC 50 as it relates to the containment heat removal system which shall be designed so 

that the containment structure and its internal compartments can accommodate without 

exceeding the design leakage rate and with sufficient margin, the calculated pressure 

and temperature conditions resulting from any LOCA. 

10 CFR 50 Appendix J as it relates to the containment integrated leak rate test pressure which 

should be greater than or equal to the calculated peak containment internal pressure related to 

the design basis accident and specified in the technical specification.
6-2

The containment integrity analysis is performed to confirm that the requirements of the above 

regulations are met. The analysis consists of a LOCA mass and energy (M&E) release 

calculation which is an input to the containment pressure and temperature response calculation.

Changes in Containment Integrity Analysis in FSAR Amendment 112:

In a letter dated May 30, 2014 (ADAMS Accession No. ML14160A901), TVA submitted 

Amendment 112 that proposed changes to Watts Bar Nuclear Plant (WBN), Unit 2 FSAR. This 

letter was supplemented by letter dated December 17, 2014 (ADAMS Accession No.

ML14352A248) which provided response to NRC staff requests for additional information 

(RAIs). 

FSAR Amendment 112 revised the design basis blowdown energy of 317.3×106 British thermal 

units (Btu) and mass of 502.7×103 lb put into the containment to energy of 314.9×106 Btu and 

mass of 498.1×103 lb put into the containment in FSAR Section 6.2.1.1.1, fourth paragraph, 

item (1).

In SCVB-RAI-4 the applicant was requested to explain the reasons for the reduction in the 

LOCA blowdown M&E. In its response to SCVB-RAI-4 (ADAMS Accession No. ML14352A248), 

the applicant stated:

The subject integrated blowdown M&E data provided in the Amendment 111 contains 

data that was incorrectly applied. The correct Amendment 111 integrated blowdown 

energy and mass is 315.7×106 Btu and 499.8×103 lb, respectively.

The appropriate data for Amendment 112 is correct as presented, that is ‘The design 

basis blowdown energy of 314.9×106 Btu and mass of 498.1×103 lb put into the 

containment.’

The reason for the reduction in the integrated M&E is due to a shorter blowdown 

transient time from 27.2 seconds to 26.8 seconds for the Amendment 112 reanalysis. 

The blowdown phase transient was directly affected by the corrections relative to the 

metal heat capacity calculation with respect to the steam generator tube (primary metal) 

and subsequent secondary side metal as a result of Nuclear Safety Advisory Letter 

(NSAL) 14-2, ‘Westinghouse Loss-of-Coolant Accident Mass and Energy Release 

Calculation Issue for Steam Generator Tube Material Properties,’ March 31, 2014. In 

addition, to mitigate the impact from the work performed in response to SCVB-RAI-3 

(ADAMS Accession Number ML14352A248), a reduction in the conservative value for 

core stored energy was utilized. These changes also have an effect on downstream 

input for the blowdown calculation and transient time.

For an ice condenser containment design containment pressure analysis the blowdown 

integrated M&E release is used to initialize the LOTIC1 post-blowdown containment 

pressure response. Since the peak containment pressure occurs after the ice bed has 

melted out, the total integrated release (which is greater for the Amendment 112 

analysis) is of more importance. The total integrated energy release (calculated based 

upon the WCAP-10325-P-A, Reference 1, evaluation model) has increased from 

938.84×106 Btu (Amendment 111) to 949.05×106Btu for the Amendment 112 reanalysis.
6-3

The NRC staff finds the proposed changes acceptable because the applicant reanalyzed 

the M&E with the following changes: (a) by performing sensitivity studies that resulted in an 

increase in the condenser ice mass and a shorter blowdown transient time in the reanalysis, 

(b) reduced conservatism in the core stored energy, and (c) included corrections in M&E release 

recommended by NSAL-14-2. The M&E analysis met the acceptance criteria 1 and 2 of SRP 

6.2.1.3.

NOTE: The applicant subsequently revised the blowdown energy released into containment in 

the revision to FSAR Amendment 114. The NRC staff’s evaluation of the revision to FSAR 

Amendment 114 is discussed later in this section.

FSAR Amendment 112 changed the initial conditions in the containment from a temperature of 

100 °F in the lower and dead-ended volumes, 80 °F in the upper volume, and 15 °F in the ice 

condenser to 120 °F in the lower and dead-ended volumes, 110 °F in the upper volume, and 

15 °F in the ice condenser in Section 6.2.1.3.3, assumption (10) under the heading 

“Containment Pressure Calculation.”

In SCVB-RAI-3, the applicant was requested to refer to the initial ice condenser temperature 

assumption of 15 °F specified in assumption (10) above and the Technical Specification (TS) 

Surveillance Requirement (SR) 3.6.11.1 that states: “Verify maximum ice bed temperature is 

≤27°F.” The ice bed temperature is a key parameter for the ice condenser performance for 

pressure suppression, (i.e., assuming a lower ice bed temperature for the long-term pressure 

response is less conservative than using a higher temperature). For a conservative 

containment pressure calculation, the applicant was requested to justify using a 

nonconservative assumption of ice temperature of 15 °F instead of 27 °F.

In response to SCVB-RAI-3 (ADAMS Accession No. ML14352A248), the applicant performed a 

sensitivity (bounding) study by assuming the SR 3.6.11.1 maximum ice bed temperature of 

27 °F. From the results of the sensitivity (bounding) analysis, the applicant determined that the 

peak containment pressure increases with a higher initial ice temperature and is also influenced 

by the time taken for ice bed to melt, which in turn depends on the initial ice mass and M&E

release. A 27  °F initial ice temperature produces a higher containment peak pressure than with 

an initial ice temperature of 15  °F. The applicant performed the analysis with different initial ice 

masses until the results of the analysis closely matched the current FSAR analysis results for 

the peak containment pressure. The sensitivity studies determined that the initial ice mass 

should be increased by 70,000 lb (i.e., from 2.26 million lb to 2.33 million lb).

In its response to SCVB-RAI-3 (ADAMS Accession No. ML14352A248), the applicant provided 

the following revisions:

FSAR Section 6.2.1.3.3 assumptions (2) and (10), and the last sentence under 

heading “Containment Pressure Calculation,” reflecting ice mass changed from 

2.24×106 lbs to 2.33×106 lbs, ice bed temperature change from 15°F to 27°F, 

and maximum calculated containment pressure change from 12.86 psig to 

12.40 psig, and the approximate time at which the maximum pressure occurs 

change from: 4348 seconds to: 4346 seconds.

FSAR Section 6.7.6.1 Design Condition (2)(A)(ii) changing the minimum total weight of ice in 

columns from 2.24×106 lb to 2.33×106 lb.
6-4 

FSAR Section 6.8.3, first paragraph, maximum calculated containment pressure change from: 

12.86 psig to: 12.40 psig, and the approximate time at which the maximum pressure occurs 

change from: 4348 seconds to: 4346 seconds.

The NRC staff finds the changes in FSAR Sections 6.2.1.3.3, 6.7.6.1, and 6.8.3 acceptable 

based on the sensitivity analysis results performed with the revised initial ice mass, initial ice 

condenser temperature, and the revised initial compartment temperatures in assumption (10). 

The changes in this assumption, which revises the initial containment temperature for lower and 

dead-ended compartments from 100 °F to 120 °F and for the upper compartment from 80 °F to 

110 °F, are conservative for the long-term peak pressure analysis. 

NOTE: The applicant subsequently revised the updates to the FSAR provided in response to 

SCVB-RAI-3 in the revision to FSAR Amendment 114. The NRC staff’s review of FSAR 

Amendment 114 is discussed later in this section. 

Section 6.2.1.3.6 under heading “Long-Term LOCA Mass and Energy Releases”, refers to 

WCAP-10325-P-A, “Westinghouse LOCA Mass and Energy Release Model for Containment 

Design March 1979 Version” (ADAMS Accession Number ML080640615) for the evaluation 

model used for the long term LOCA mass and energy release calculations. 

In SCVB-RAI-5 the NRC staff identified Westinghouse issued Nuclear Safety Advisory Letters 

(NSALs)-06-6, -11-5, and -14-2 that reported errors in the WCAP-10325-P-A methodology that

requires containment analyses should be corrected. The applicant was requested by the NRC 

staff to describe changes in the following containment analyses results using the corrected 

WCAP-10325-P-A methodology that incorporates corrections listed in the these NSALs: 

(a) containment peak pressure, (b) containment peak gas temperature for environment 

equipment qualification (EEQ), (c) containment peak wall temperature, (d) containment sump 

peak water temperature, (e) pump net positive suction head available (NPSHA) for the pumps 

that draw water from the containment sump during recirculation mode of safety injection and 

containment cooling, and (f) containment minimum pressure analysis for emergency core 

cooling system (ECCS) performance capability. The applicant was also requested by the NRC 

staff to add a statement in the FSAR stating the corrected version of WCAP-10325-P-A 

methodology which removed errors reported in NSALs-06-6, -11-5, and -14-2 was used for the 

containment LOCA M&E release analysis.

In response to SCVB-RAI-5 (ADAMS Accession No. ML14352A248), the applicant stated that 

the long-term containment pressure analysis scope affected by WCAP-10325-P-A M&E release 

is in FSAR Section 6.2.1.3.3, and the resulting impact of the NSALs on the containment 

peak pressure was an increase in the calculated peak pressure from 12.6 psig to a value of 

12.86 psig. However, the revised analysis performed in response to SCVB-RAI-3 resulted in 

a lower peak containment pressure of 12.40 psig. The applicant also stated that there was no 

change to the associated peak containment temperature, and, likewise, there was no adverse 

impact on the peak sump temperature during the recirculation phase calculated for the LOCA 

analysis. In addition, neither the pump NPSHA nor the containment minimum pressure analysis 

for ECCS performance capability is impacted by the NSALs.
6-5

In response to SCVB-RAI-5, the applicant provided a markup of the first paragraph in FSAR 

Section 6.2.1.3.6, by inserting the following:

A corrected version of WCAP-10325-P-A computer codes and input, which 

removed errors reported in References 29, 30 and 31 [NSAL-06-6, -11-5, and -

14-2], was used for the containment LOCA M&E release analysis. The NSAL 

corrections are corrections to calculations in support of the approved 

methodology, and not a change in methodology.

The applicant provided revised FSAR Tables 6.2.1-3 and 6.2.1-4 on energy balances in 

Amendment 112. The applicant further revised these tables in response to SCVB-RAI-3. The 

applicant added the following references at the end of FSAR Section 6.2.1: (29) NSAL-06-6, 

“LOCA Mass and Energy Release Analysis,” June 6, 2006, (30) NSAL-11-5, “Westinghouse 

LOCA Mass and Energy Release Calculation Issues,” July 25, 2011, and (31) NSAL-14-2, 

“Westinghouse LOCA Mass and Energy Release Calculation Issue for Steam Generator Tube 

Material Properties,” March 31, 2014.

The NRC staff finds the changes in FSAR Section 6.2.1.3.6, and Table 6.2.1-3 and 6.2.1-4 

acceptable because the applicant confirmed that WCAP-10325-P-A methodology has been 

corrected from errors reported in Westinghouse NSAL-06-6, -11-5, and -14-2.

The applicant provided an updated FSAR Table 6.2.1-15 revising the Safety Injection (SI) delay 

time from 35.91 seconds to 36.13 seconds in Amendment 112. 

The applicant provided revised FSAR Tables 6.2.1-16, “Double-Ended Pump Suction Guillotine 

Break—Blowdown Mass and Energy Releases,” 6.2.1-17, “Double-Ended Pump Suction 

Guillotine Break—Reflood Mass and Energy Release—Minimum Safety Injection,” 6.2.1-18, 

“Double-Ended Pump Suction Guillotine Break—Minimum Safety Injection Principal Parameters 

During Reflood,” and 6.2.1-19, “Double-Ended Pump Suction Guillotine Break—Post-Reflood 

Mass and Energy Releases—Minimum Safety Injection” in Amendment 112.

The applicant provided markup of FSAR Table 6.2.1-25, “Double-Ended Pump Suction LOCA 

Sequence of Events,” in Amendment 112, changing the following parameters:

• Accumulator Flow Starts time from 15.7 seconds to 15.6 seconds

• Assumed Initiation of ECCS time from 35.91 seconds to 36.13 seconds

• End of Blowdown time from 27.4 seconds to 26.8 seconds

• Accumulators Empty time from 66.782 seconds to 66.700 seconds

• End of Reflood time from 239.7 seconds to 239.8 seconds

The applicant provided revised FSAR Tables 6.2.1-26a, “Mass Balance,” and 6.2.1-26b, 

“Energy Balance,” in Amendment 112.

The applicant provided revised FSAR Figures 6.2.1-1, 6.2.1-2a, 6.2.1-2b, 6.2.1-3, 6.2.1-4, 

and 6.2.1-4a in Amendment 112. The applicant further revised these figures in response to 

SCVB-RAI-3 (ADAMS Accession No. ML14352A248).
6-6

The staff finds the changes in FSAR Section 6.2.1 Tables and Figures acceptable because 

they are based on the changes made in Section 6.2.1 Amendment 112 and responses to 

SCVB-RAI-3, -4, and -5 (ADAMS Accession No. ML14352A248).

The NRC staff concludes that the changes in Amendment 112 meet the requirements of 

10 CFR Part 50 Appendix A: (1) GDC 16, because the applicant showed that the containment 

internal pressure and wall temperature during a design basis LOCA do not exceed their design 

limits, that is, internal design pressure of 15 psig or the administrative acceptance criterion of 

13.5 psig, and the wall design temperature of 250 °F respectively, (2) GDC 38, because the 

applicant showed that the containment heat removal system would reduce the containment 

pressure and temperature rapidly, following a design basis LOCA and would maintain them at 

acceptable levels and (3) GDC 50, because the applicant showed that the containment heat 

removal system is designed so that the containment structure and its internal compartments can 

accommodate without exceeding the design leakage rate and with sufficient margin, the 

calculated pressure and temperature conditions resulting from design basis LOCA. Therefore 

the NRC staff finds the changes in FSAR Amendment 112 acceptable.

Changes in Containment Integrity Analysis in FSAR Amendment 114

In a letter dated August 13, 2015 (ADAMS Accession No. ML15225A382), TVA submitted

a revised WBN, Unit 2, containment integrity analysis to address the following: (a) meeting 

10 CFR Part 50, Appendix A, Criterion 5, for the shared essential raw cooling water (ERCW) 

system and component cooling system (CCS) between WBN Units 1 and 2, and (b) an issue 

reported in Westinghouse InfoGram IG-14-1, namely the density and specific heat values used 

for the reactor coolant system (RCS) metal mass do not bound the values published by the 

American Society of Mechanical Engineers (ASME). The applicant included markup of 

Amendment 113 of the FSAR reflecting the changes. The applicant submitted a supplement 

letter dated August 28, 2015 (ADAMS Accession No. ML15243A044) that provided responses 

to NRC staff RAIs and audit questions. The applicant subsequently included the markups in 

FSAR Amendment 114 submitted September 11, 2014. The NRC staff’s evaluation of the 

proposed TSs for WBN, Unit 2, are discussed in Section 16 of this SSER.

FSAR Amendment 114 related to LOCA analysis, 10 CFR Part 50, Appendix A, Criterion 5, and 

Westinghouse InfoGram IG-14-1

FSAR Amendment 114 modified the following:

• the blowdown energy from 314.9 x 106 Btu and mass of 498.1 x 103 lb to energy of 

315.5 x 106 Btu and mass of 498.9 x 103 lb as noted in item (1) in the fourth paragraph of 

FSAR Section 6.2.1.1.1, 

• the initial ice in the ice condenser from 2.33 x 106 lb to 2.585 x 106 lb as noted in item 

(2) in the “Containment Pressure Calculation” in FSAR Section 6.2.1.3.3,

• residual heat removal (RHR) spray initiation from 4346.7 to 3600.0 seconds into the 

LOCA containment response transient as noted in item (12) in the “Containment 

Pressure Calculation” in FSAR Section 6.2.1.3.3, 
6-7

• the heat transfer coefficient (UA) for the component cooling system heat exchanger from 

5.778 x 106 Btu/hr-°F to 3.17 x 106 Btu/hr-°F as discussed in item (14) in the 

“Containment Pressure Calculation” in FSAR Section 6.2.1.3.3,

• the modeling of the ERCW flow to the component heat exchanger from 6250 to 3500 

gallons per minute as noted in item (22) in the “Containment Pressure Calculation” in 

FSAR Section 6.2.1.3.3,

• the maximum containment pressure from 12.40 pounds per square inch, gauge (psig) 

occurring at approximately 4346 seconds to 11.73 psig occurring at approximately 

3600 seconds, as noted in the last sentence in the “Containment Pressure Calculation” 

in FSAR Section 6.2.1.3.3 and FSAR Section 6.8.3,

• the analysis of the initial steam generator pressure to include an additional 13 psi 

internal steam generator pressure drop as noted in note 2 of FSAR Table 6.2.1-15, 

• the sequence of events for a double-ended pump suction LOCA. The sequence was 

revised for the end of blowdown time from 26.8 to 27 seconds, the time to empty the 

accumulators from 66.700 to 66.2 seconds, and the end of reflood time from 239.8 to 

243.6 seconds, as noted in FSAR Table 6.2.1-25,

• the minimum ice basket loading from 2.33 x 106 lb to 2.585 x 106 lb pounds of ice as 

noted in item (4) of the “Interface Requirements” in FSAR Section 6.7.4.1, 

• the minimum total weight of ice in columns from 2.33 x 106 lb to 2.585 x 106 lb as noted 

in item (2)(A)(ii) of the “Design Conditions” in FSAR Section 6.7.6.1,

• Tables 6.2.1-3, 6.2.1-4, 6.2.1-16, 6.2.1-17, 6.2.1-18, 6.2.1-19, 6.2.1-26a, and 6.2.1-26b, 

and Figures 6.2.1-1, 6.2.1-2a, 6.2.1-2b, 6.2.1-3, 6.2.1-4, and 6.2.1-4a.

The changes in the above listed FSAR sections, tables, and figures are based on revision to the 

LOCA containment integrity AOR documented in Amendment 113. 

The applicant performed the containment integrity analysis to confirm that the containment 

internal pressure, temperature, and the wall temperature do not exceed their design limits. The 

analysis should ensure that the containment heat removal systems can remove the most limiting 

LOCA or main steam line break (MSLB) M&E released into the containment without exceeding 

the WBN, Unit 2, containment internal design pressure of 15 psig or the administrative 

acceptance criterion of 13.5 psig, and the wall design temperature of 250 °F. The analysis 

should also ensure that the sump temperature change does not adversely impact the NPSHA

for the pumps that draw fluid from the sump during the LOCA recirculation phase.

The analysis consists of an M&E release calculation that is an input to the containment pressure 

and temperature response calculation. The applicant performed the M&E analysis in 

accordance with the NUREG-0800, SRP 6.2.1.3. The purpose of revising the WBN, Unit 2 

containment integrity AOR is due to the following main reasons:

(1) Correction of errors reported in Westinghouse InfoGram 14-1 would increase the stored 

energy of the RCS metal during normal plant operation, and consequently increases the 

M&E release into the containment during a design basis LOCA.
6-8

(2) Sharing of the CCS and the ERCW system flow by WBN, Units 1 and 2 during a normal 

shutdown cooling in one unit and simultaneous design basis LOCA in the second unit 

occurring prior to 48 hours from control rod insertion in the shutdown unit while it is in the 

hot shutdown mode, concurrent with a loss-of-offsite power (LOOP) in both units. 

The applicant analyzed the most limiting scenario of loss of Train A concurrent with LOOP 

during which only one CCS HX ‘C’ is available to be shared between the units. In this scenario, 

which has the most limiting heat load on the single CCS HX, the LOCA is assumed to occur in 

the accident unit at 7 hours after the initiation of shutdown of the second unit. At this instant, the 

applicant calculated the CCS HX shutdown heat load of 89.4 million British thermal units

(MBtu)/hr plus the LOCA containment heat load of 54.8 MBtu/hr. Because of sharing of the 

CCS HX between the units, this scenario results in a reduction of the following parameters of 

the CCS HX apportioned for containment heat removal in the LOCA unit: (a) the AOR value of 

ERCW flow and (b) the AOR value of UA (overall heat transfer coefficient × heat transfer area).

The applicant included the following other known changes in the containment integrity analysis: 

(a) initial RCS temperature uncertainty +7 °F, (b) 17×17 robust fuel assembly-2 (RFA-2) fuel 

(which may incorporate tritium-producing burnable absorber rods (TPBARs)) for decay heat, 

and (c) containment spray flow control valves (FCVs) increased opening stroke time to 

+13 seconds. The applicant also considered the impact of ±0.2-hertz variation in the 

emergency diesel generator (EDG) frequency and concluded its negligible impact on the 

LOCA M&E results.

In Amendment 114, the applicant revised the LOCA M&E release AOR using the NRC approved 

Westinghouse WCAP-10325-P-A (ADAMS Accession No. ML080640615) methodology and 

resolved the issues reported in Westinghouse InfoGram 14-1. The corrections consisted of 

using the revised RCS metal (stainless steel and low alloy carbon steel) density and specific

heat that bounds the values given in the current ASME Boiler and Pressure Vessel (B&PV) 

Code 2010 Edition, Section II, Part D. The applicant used the RCS metal density value of 

501 lbm/ft3 which represents the density of stainless steel 304 and 316 at 70 °F and bounds the 

density (484 lbm/ft3

) of low alloy carbon steel at 70 °F given in Table PRD of the ASME Code 

Section II, Part D. For conservative M&E release analysis, the applicant used density of 

stainless steel (501 lbm/ft3

), which bounds the density of bulk of the metal mass (carbon steel) 

in the reactor vessel and the steam generators. The applicant stated that the steam generator 

tubes are not included because they are treated separately for a conservative LOCA M&E 

analysis. The ASME Code, Section II, Part D does not directly provide the metal specific heat 

values as a function of temperature, instead its Table TCD provides thermal diffusivity which is 

related to thermal conductivity, density, and specific heat by the following equation:

Thermal Diffusivity = (Thermal Conductivity) / (Specific Heat x Density)

From the above equation, while adding a 10% uncertainty, the applicant derived a conservative 

value of the RCS metal specific heat of 0.145 Btu/lbm°F.

The NRC staff reviewed the ASME B&PV Code, Section II, Part D, and confirmed that the RCS 

metal density of 501 lbm/ft3 and specific heat of 0.145 Btu/lbm°F are conservative for the M&E 

release analysis because these are higher than the ASME values.
6-9

Resolving the issues reported in Westinghouse InfoGram IG-14-1 impacts the M&E release 

during the LOCA blowdown phase. The NRC staff finds the change in FSAR Section 6.1.3.3 

acceptable because an increase in density and specific heat of the RCS metal reported in 

InfoGram IG 14-1 increases its stored sensible heat during normal plant operation and would 

therefore release greater M&E than the AOR during the LOCA blowdown phase and is 

conservative. 

The containment peak pressure, temperature and sump fluid temperature occur during the long 

term LOCA post-reflood phase. The applicant used the revised M&E release output data for the 

analysis. The long term analysis consists of determining the containment pressure and 

temperature responses during the LOCA post-reflood phase and sump fluid temperature

response during the LOCA sump recirculation mode. 

For the M&E analysis, the applicant used the same inputs and assumptions as stated in 

Amendment 113, other than the changes noted above. The applicant replaced Tables 6.2.1-16, 

6.2.1-17, 6.2.1-18, and 6.2.1-19 of Amendment 113 with revised M&E results tables in inserts C, 

D, E, and F by letters dated August 13, 2015 (ADAMS Accession No.ML15225A382) and 

August 28, 2015 (ADAMS Accession No. ML15243A044).

For the containment response analysis, the applicant used the revised M&E analysis results and 

used the same assumptions and inputs as in Amendment 114 with the following exceptions: 

(a) ice mass in the ice condenser changed from 2.33×106 lb to 2.585×106 lb, (b) UA of the 

CCS HX apportioned (virtual) for containment cooling changed from 5.778×106 Btu/hr-°F to 

3.17×106 Btu/hr-°F, and (c) ERCW flow to the CCS HX apportioned (virtual) for containment 

cooling changed from 6250 gpm to 3500 gpm. The applicant calculated the real CCS HX 

UA = 6.82×106 Btu/hr-°F with an ERCW flow 9200 gpm and CCS flow of 10,166 gpm.

During its review, the NRC noted that the applicant split the CCS HX ‘C’ 

UA = 6.82×106 Btu/hr-°F into two virtual HXs, assigning one to the LOCA unit with a 

UA = 3.17×106 Btu/hr-°F and an ERCW flow of 3400 gpm, and the second to the shutdown unit. 

In response to the NRC staff question in a letter dated August 28, 2015 (ADAMS Accession No. 

ML15243A044) regarding the basis for determination of UA = 3.17×106 Btu/hr-°F, the applicant 

stated that the split was based on approximately an equal CCS mass flow fraction to the virtual 

HXs. The analysis results show that the peak containment pressure is 11.73 psig for the limiting 

double-ended pump suction break LOCA, assuming an ice bed mass of 2.585×106 lb. The 

peak pressure occurs at approximately 3600 seconds with ice bed melt-out at approximately 

2959 seconds and the containment spray switchover time to sump recirculation mode 

2718.7 seconds. The results are acceptable because peak calculated pressure is less than 

the design pressure of 15.0 psig as well as the administrative limit of 13.5 psig, and the margin 

between the ice bed melt-out time and spray switchover time (2959−2718.7 = 240.3 seconds) is 

greater than 150 seconds and therefore meets the GDC 16, and 38 requirements.

The NRC staff noted a significant difference in the heat loads on the two virtual CCS HXs 

(89.4 MBtu/hr on the shutdown unit versus 54.8 Btu/hr on the LOCA unit. The NRC staff 

requested the applicant reanalyze by splitting the real CCS HX (UA = 6.82×106 Btu/hr-°F) into 

two virtual HXs based on heat load fractions. With the revised split, and with the proposed 

revised total ERCW and CCS flows of 9200 gpm and 10,000 gpm, respectively, the applicant 

calculated UA = 2.64×106 Btu/hr-°F for the virtual LOCA unit and UA = 4.12×106 Btu/hr-°F for 

the virtual shutdown unit.
6-10

Since the WBN, Unit 2, containment response described above was performed with a 

virtual CCS HX UA = 3.17×106 Btu/hr-°F, while based on apportioning by heat load its 

UA = 2.64×106 Btu/hr-°F, the applicant performed a sensitivity study by changing its UA from 

3.17×106 Btu/hr-°F to 2.00×106 Btu/hr-°F (which bounds the calculated 

UA = 2.64×106 Btu/hr-°F) and used the same ERCW flow of 3504 gpm. The result showed an 

increase in peak containment pressure from 11.73 to 11.76 psig, which is considered to be 

insignificant. The sensitivity study determined that the peak containment pressure is not 

sensitive to the UA of the virtual CCS HX assigned for the LOCA unit. The NRC staff finds the 

modeling results are consistent with (i.e., proportional to) the way containment heat is removed 

during an accident. Specifically, most of the containment heat removal is performed by the 

containment spray system compared to a small fraction removed by the RHR spray cooled by 

the CCS HX ‘C’.

The NRC staff finds the containment integrity analysis acceptable because: (a) the calculated 

design basis LOCA peak containment pressure 11.73 psig is below the containment design 

pressure of 15 psig and the administrative limit of 13.5 psig, and also below the 10 CFR 50 

Appendix J Integrated Leak Test Pressure Pa = 15 psig which is the same as the containment 

design pressure as stated in TS Section 5.2.7.19, and (b) the calculated maximum containment 

vapor temperature 234.3 °F is below the containment design temperature of 250 °F. 

The applicant calculated 157.5 °F (Table 4.4.3-6 in the August 13, 2015 letter (ADAMS 

Accession No. ML15225A382), Attachment to Enclosure 1) as the maximum sump fluid 

temperature after switchover, which is a small increase from its current approximate value of 

155 °F in FSAR Amendment 113 Figure 6.2.2-3. The increase in the peak sump fluid 

temperature increases the vapor pressure by 0.262 psi (0.6 ft water) which reduces the NPSHA 

available by 0.6 ft at the inlet of the RHR and containment spray pumps during the recirculation 

mode. The decrease in the NPSHA does not impact the operation of RHR and CS pumps 

because the minimum NPSH margin (NPSHA minus NPSH required) for the most limiting pump 

given in FSAR Amendment 113 Table 6.3-12 is 6.4 ft. The most limiting NPSH margin is 

conservative because the applicant did not take credit for the sump static water level in NPSH 

available calculation as stated in FSAR Amendment 113, Section 6.3.2.14 item (1). 

Since the LOCA event requires the greatest amount of ice compared to other accidents and 

events, the analytical value of initial ice mass 2.585×106 lb based on LOCA results is acceptable 

for all other accidents and events.

The NRC staff concludes that containment integrity analysis for the sharing of CCS and ERCW 

systems between WBN, Units 1 and 2 in the event of a design basis LOCA in WBN, Unit 2 and 

an orderly shutdown of WBN, Unit 1, the design of WBN, Unit 2, as described and analyzed in 

TVA’s FSAR through Amendment 114, meet the following criteria in 10 CFR Part 50 

Appendix A:

• GDC 5, because the applicant showed that the sharing of the systems will not

significantly impair their ability to perform their safety function of mitigating the increase

in LOCA containment pressure and temperature in the LOCA unit,

• GDC 16, because the applicant showed that the WBN, Unit 2 containment design

conditions important to safety are not exceeded during a design basis LOCA in WBN,

Unit 2,
6-11

• GDC 38, because the applicant showed that the containment heat removal system 

would reduce the WBN, Unit 2 containment pressure and temperature rapidly, following 

a design basis LOCA and would maintain them at acceptable levels,

• GDC 50, because the applicant showed that the WBN, Unit 2 containment heat removal 

system is designed so that the containment structure and its internal compartments can 

accommodate without exceeding the design leakage rate and with sufficient margin, the 

calculated pressure and temperature conditions resulting from design basis LOCA. 

Additional FSAR Amendment 114 Revisions

The applicant revised the design internal pressure referenced in FSAR Section 6.2.1.2 “Primary 

Containment System Design” from 13.5 psig to 15.0 psig, which corresponds to the maximum 

internal pressure of 15 psig and is consistent with FSAR Section 6.2.2.3. FSAR 6.2.2.3 states:

The analyses were performed using the LOTIC code and show that the 

containment heat removal systems are capable of keeping the containment 

pressure below the containment maximum internal pressure of 15 psig, which 

corresponds to the code design internal pressure of 13.5 psig at 250 °F. 

The NRC staff notes that the code design pressure remains as 13.5 psig.

The applicant revised the area of ice condenser lattice frames slab 13 referenced in FSAR 

Table 6.2.1-1 “Structural Heat Sinks” from 75,860 ft2 to 75,865 ft2 to represent the as-built 

configuration.

The applicant revised the decay heat at time 15 seconds referenced in FSAR Table 6.2.1-20 

“Decay Heat Curve” from 0.477187 to 0.0477187. The NRC staff determined that the change 

did not invalidate any of the staff’s previous findings.

FSAR Amendment 114 revised FSAR Section 6.2.1.3.10, associated tables, and figures are 

based on a revision to the containment pressure and temperature response analysis for a main 

steam line break (MSLB) accident inside the containment. The change is due to an increase in 

the feedwater isolation time from 8.0 seconds in the analysis of record (AOR) to 13.5 seconds 

by the motor-operated main feedwater isolation valve (MFIV) during a MSLB and a single failure 

of a main steam isolation valve (MSIV). The following is a list of revisions in FSAR Amendment 

114 related to this reanalysis.

FSAR Section 6.2.1.3.10, page 6.2.1-34, under the heading “Pipe Break Blowdowns-Spectra 

and Assumptions”, assumption 2 is revised from:

Steam line isolation signals and feedwater line isolation signals are generated by 

either a low steam line pressure signal, high or high-high containment pressure 

signal, or high steam line pressure rate signal. An allowance of 8 seconds is 

assumed for steam line isolation including generation, processing, and delay of 

the isolation signal and valve closure. An allowance of 8 seconds is assumed for 

feedwater line isolation including generation, processing, and delay of the 

isolation signal and valve closure.
6-12

To:

Steam line isolation signals and feedwater line isolation signals are generated by 

either a low steam line pressure signal, high or high-high containment pressure 

signal, or high steam line pressure rate signal. An allowance of 8 seconds is 

assumed for steam line isolation including generation, processing, and delay of 

the isolation signal and valve closure. An allowance of 8.5 seconds is assumed 

for feedwater line isolation by the air-operated feedwater regulation valve or 13.5 

seconds by the motor-operated main feedwater isolation valve including 

generation, processing, and delay of the isolation signal and valve closure.

FSAR Section 6.2.1.3.10, page 6.2.1-34, under the heading “Pipe Break Blowdowns-Spectra 

and Assumptions”, assumption 6 is revised from:

Failure of a main steamline isolation valve (MSIV), failure of a feedwater isolation 

valve (FIV), failure of auxiliary feedwater runout control protection, and failure of 

a safety injection train are considered.

To:

Failure of a main steamline isolation valve (MSIV), failure of a feedwater 

regulation valve (FRV), failure of auxiliary feedwater runout control protection, 

and failure of a safety injection train are considered.

FSAR Section 6.2.1.3.10, page 6.2.1-35, under the heading “Single Failure Effects”, item (2) is 

revised from:

Failure of a feedwater isolation valve could only result in additional inventory in 

the feedwater line which would not be isolated from the steam generator. The 

mass in this volume can flash into the steam generator and exit through the 

break. The feedwater regulating valve closes in no more than 6.5 seconds 

precluding any additional feedwater from being pumped into the steam 

generator. The additional line volume available to flash into the steam generator 

is that between the feedwater isolation valve and the feedwater regulating valve.

To:

The additional mass inventory in the feedwater line that would not be isolated 

from the steam generator following main feedwater isolation can flash into steam 

and pass through the steam generator and exit out of the break. Failure of the 

air-operated feedwater regulation valve results in an increase of 5 seconds for 

main feedwater isolation by the motor-operated feedwater isolation valve, but 

also results in a decrease in the feedwater piping volume that is not isolated from 

the steam generator. The feedwater regulation valve closes in no more than 6.5 

seconds and the feedwater isolation valve closes in no more than 11.5 seconds 

precluding any additional feedwater from being pumped into the steam 

generator.
6-13

FSAR Section 6.2.1.3.10, page 6.2.1-36, under the heading “Worst-Case Mass and Energy 

Releases,” description of case (1) is revised from: “Full double-ended rupture at 100.6% of 

nominal full power with a failure of a FIV. This represents the limiting DER case in terms of 

calculated peak temperature.” To: “Full double-ended rupture at 100.6% of nominal full power 

with a failure of a MSIV. This represents the limiting DER case in terms of calculated peak 

temperature.”

FSAR Section 6.2.1.3.10, page 6.2.1-38, under the heading “Large Break” is revised from:

The limiting case among the double-ended ruptures, which yielded a calculated 

peak temperature of 324.3°F and a peak pressure of 10.3 psig, is the 1.4 ft2 loop 

break at 100.6% of nominal full power with a failure of a main feedwater isolation 

valve. Figure 6.2.1-69 provides the upper and lower compartment temperature 

transients, and Figure 6.2.1-70 illustrates the lower compartment pressure 

transient. Table 6.2.1-39 contains the mass and energy release rates for the 

above case.

To:

The limiting case among the double-ended ruptures, which yielded a calculated 

peak temperature of 324.4°F and a peak pressure of 10.3 psig, is the 1.4 ft2 loop 

break at 100.6% of nominal full power with a failure of a main steamline isolation 

valve. Figure 6.2.1-69 provides the upper and lower compartment temperature 

transients, and Figure 6.2.1-70 illustrates the lower compartment pressure 

transient. Table 6.2.1-39 contains the mass and energy release rates for the 

above case.

FSAR Section 6.2.1.3.10, page 6.2.1-38, under the heading “Small Break” is revised from:

The most severe transient in terms of superheat temperature duration for the 

small break spectrum is the 0.35 ft2

, 30% nominal full power, with AFW pump 

runout protection failure. The temperature transient with a peak temperature of 

325.1°F and peak pressure of 6.59 psig for the case is presented in Figure 6.2.1-

71, and the pressure transient is provided in Figure 6.2.1-72. Table 6.2.1-39 

provides the mass and energy release rates for this case.

To:

The most severe transient in terms of superheat temperature duration for the 

small break spectrum is the 0.35 ft2

, 30% nominal full power, with AFW pump 

runout protection failure. The temperature transient with a peak temperature of 

325.0°F and peak pressure of 6.59 psig for the case is presented in Figure 6.2.1-

71, and the pressure transient is provided in Figure 6.2.1-72. Table 6.2.1-39 

provides the mass and energy release rates for this case.

FSAR Tables and Figures impacted by the above changes are: Tables 6.2.1-39, 6.2.1-41, 

6.2.1-43, and 6.2.1-44, Figures 6.2.1-69, 6.2.1-70, 6.2.1-71, 6.2.1-72, 6.2.1-73, and 6.2.1-74

The most limiting case for the peak containment temperature is the full double-ended 4.6 ft2

break area occurring at the nozzle on one steam generator, however the maximum effective 
6-14

break area is limited to1.4 ft2 because of the presence of steam line flow restrictors in the steam 

generators. The feedwater isolation time increase by 5.5 seconds (from 8.0 seconds to 13.5 

seconds) which results in increased mass and energy entering into the steam generator and 

thereby through the break is partially offset by a decrease in the feedwater piping volume that is 

not isolated from the steam generator. The large break limiting temperature case resulted in a 

peak containment temperature of 324.4°F and a peak pressure of 10.3 psig.

For the most limiting small steam line break of 0.35 ft2

, at 30% nominal full power, with auxiliary 

feedwater (AFW) pump runout protection failure, the peak temperature decreased from 325.1°F 

in the AOR to 325.0°F and peak pressure of 6.59 psig.

The applicant’s revised MSLB analysis is acceptable because the increase in the peak 

containment temperature from 324.3°F (in the AOR) to 324.4°F for the large MSLB is bounded 

by the limiting small break AOR peak containment temperature of 325.1°F. The peak 

containment pressure of 10.3 psig for the limiting MSLB is acceptable because it is bounded by 

the LOCA peak containment pressure of 11.73 psig. 

6.2.2 Containment Heat Removal Systems

FSAR Amendment 112 revised Sections 6.8.1 “Design Bases,” 6.8.2 “System Description,” and 

Section 6.8.3 “Safety Evaluation,” in Section 6.8 “Air Return Fans.” The revision to Section 

6.8.1 changed the first two sentences from:

This operation takes place at the appropriate time (Section 6.7) following a 

beyond-design-basis accident. The secondary purpose of the system is to limit 

hydrogen concentration in potentially stagnant regions by ensuring a flow of air 

from these regions.

To:

This operation takes place at the appropriate time (Section 6.7) following a LOCA 

or other high energy line break within the containment. The secondary purpose 

of the system is to limit hydrogen concentration in potentially stagnant regions by 

ensuring a flow of air from these regions (Section 6.2.5).

The revision to Section 6.8.2 changed the last sentence of the first paragraph from:

Air return fan suction side is equipped with a non-return damper which prevents 

flow from the lower compartment to the upper compartment during the initial 

stages of a beyond-design-basis accident.

To:

Air return fan suction side is equipped with a non-return damper which prevents 

flow from the lower compartment to the upper compartment during the initial 

stages of a LOCA or other high energy line break.

The revision to Section 6.8.2 changed the first sentence of the fourth paragraph from: “The 

system is designed to operate continuously during degraded core conditions.” to: “The system 

is designed to operate continuously during accident conditions.”
6-15

The revision to Section 6.8.3 changed the first sentence of the first paragraph from:

The design bases of the fans are to reduce containment pressure after blowdown 

from a severe accident pipe break, prevent excessive hydrogen concentrations in 

pocketed areas, and circulate air through the ice condenser.

To:

The design bases of the fans are to reduce containment pressure after blowdown 

from a LOCA or other high energy line break, prevent excessive hydrogen 

concentrations in pocketed areas, and circulate air through the ice condenser.

The revision to Section 6.8.3 changed the last sentence of the second paragraph from: “A 

back-draft damper, normally closed, is located upstream of each deck fan to prevent reverse 

flow during the initial severe accident blowdown.” to: “A back-draft damper, normally closed, is 

located upstream of each deck fan to prevent reverse flow during the initial LOCA or other high 

energy line blowdown.”

The revision to Section 6.8.3 changed the first sentence of the fourth paragraph from: “The fans 

are designed to withstand the beyond-design-basis accident containment environment.” to: 

“The fans are designed to withstand the post DBA environment and were shown to survive the 

beyond-design-basis accident containment environment (Section 6.2.5).”

The NRC staff finds the changes in Sections 6.8.1, 6.8.2, and 6.8.3 acceptable because they 

are consistent with WBN, Unit 1. In addition, for convenience, the applicant added reference to 

FSAR Section 6.2.5 in Sections 6.8.1 and 6.8.3, which describes the requirement for the 

containment air return system during beyond-design-basis accident conditions.

FSAR Section 6.8.5 “Instrumentation Requirements” was revised to change the first sentence

Section 6.8.5, change first sentence from:

The essential instrumentation requirements are that at least one of the air return 

fans start at the appropriate time after a beyond-design-basis-accident and that 

the fan keeps running for one year. 

To:

The essential instrumentation requirements are that at least one of the air return 

fans start at the appropriate time after receipt of a Phase B isolation signal and 

that the fan is capable of running for one year. 

The NRC staff finds the corrections in Section 6.8.5 did not change any previous NRC analysis 

and acceptance of the design of WBN Unit 2.
6-16

6.2.3 Secondary Containment Functional Design

FSAR Amendment 112 revisions

FSAR Amendment 112 revised several sections to reflect that the Auxiliary Building Gas 

Treatment System is functional but not required to mitigate a fuel handling accident (FHA). 

These changes are as follows:

FSAR Amendment 112 revised Section 6.2.3.2.3 “Auxiliary Building Gas Treatment System 

(ABGTS),” to add the following statement as a third paragraph:

Although the ABGTS is available to minimize the consequences of a fuel 

handling accident, it is not required to function in order to meet control room and 

offsite dose limits based on the use of the Regulatory Guide (RG) 1.183, 

Revision 0 (Alternative Source Term) methodology.

Amendment 112 revised FSAR Section 6.2.3.3.1 “Secondary Containment Enclosures” by 

deleting “or a FHA inside containment” at two places in the first and one place in the second 

paragraph. In addition FSAR Amendment 112 added the following in the first paragraph of 

FSAR Section 6.2.3.3.1:

The original design credited the secondary containment enclosures to mitigate 

the consequences of a FHA. Although these enclosures are available to minimize 

the consequences of a FHA, based on the use of the Regulatory Guide 1.183, 

Revision 0 (Alternative Source Term) methodology for a FHA, the structures are 

no longer required for mitigation of a postulated FHA.

FSAR Section 9.4.2.1 “Design Bases” added the following at the end of the third paragraph:

Although the ABGTS is available to minimize the consequences of a fuel 

handling accident, it is not required to function in order to meet control room and 

offsite dose limits based on the use of the Regulatory Guide 1.183, Revision 0 

(Alternate Source Term) methodology.

Also, the following was added at the end of second from last paragraph of FSAR Section 

9.4.2.1:

Although the ABGTS is available to minimize the consequences of a fuel 

handling accident, it is not required to function in order to meet control room and 

offsite dose limits based on the use of the Regulatory Guide 1.183, Revision 0 

(Alternate Source Term) methodology.

FSAR Section 3.1.2.6 “Fuel and Radioactivity Control” was revised to add the following 

paragraph before the last paragraph under “Criterion 61 - Fuel Storage and Handling and 

Radioactivity Control”

The Auxiliary Building Gas Treatment System (ABGTS) includes charcoal 

filtration which can be used to minimize radioactive material releases associated 

with a postulated spent fuel handling accident. The ABGTS system is not 

required to mitigate the consequences of a spent fuel handling accident.
6-17

The NRC staff’s evaluation of a postulated FHA is provided in SSER 25, Section 15.4.5. As 

discussed in SSER 25, Section 15.4.5, the ABGTS is not credited in TVA’s FHA analysis. 

Therefore, the proposed changes are consistent with the analysis previously evaluated by the 

NRC staff and are acceptable. 

Amendment 112 revised FSAR Section 6.2.3.3.2, “Emergency Gas Treatment System (EGTS),” 

item (2), in the last paragraph under heading “Annulus Negative Pressure Control Capability” to 

change the rated flow rate for each train of the air cleanup subsystem from 4000 +10% cfm to 

4000 ±10% cfm.

The NRC staff finds the EGTS flow tolerance of -10% acceptable because it is consistent with 

WBN, Unit 1 TS SR 3.6.9.4, which states: “Verify each EGTS train produces a flow rate ≥ 3600 

cfm and ≤ 4400 cfm within 20 seconds from the initiation of a Containment Isolation Phase A 

signal. 

Amendment 112 revised FSAR Table 6.2.3-3, “Failure Modes and Effects Analysis For Passive

Failure for the ABGTS,” to change “Remarks” for item 9 on page 6 of 27, items 10 and 11, on 

page 7 of 27, and item 12 on page 8 of 27 to remove “and both dampers fail closed on loss of 

control air” in the second sentence.

The safety-related ABGTS Train A and Train B dampers should fail close with the loss of 

non-safety-related control air. This is necessary for auxiliary building isolation or in the 

presence of high radiation in refueling area to isolate the fuel handling area exhaust fans and to 

establish boundary for ABGTS

6.2.4 Containment Isolation Systems

FSAR Amendment 112 made the following revisions to FSAR Table 6.2.4-1 “Watts Bar Nuclear 

Plant Containment Penetrations and Barriers:” 

• Sheet 44 of 64; delete shield building penetration MK-71 for penetration 80.

• Sheet 57 of 64; replace shield building penetration MK-54 for penetration 99 with MK-34.

• Sheet 57 of 64; replace shield building penetration MK-54 for penetration 100 with

MK-34.

• Change “see page” to “seepage” in Note (27)

Change Note (27) in FSAR Table 6.2.4-1 (Sheet 69 of 69) 

FSAR Table 3.9-25 “Valves Required to be Active for Design Basis Events” was revised to 

change valves FCV-43-55, -58, -61, and -64 from gate to globe valves, and capitalize the first 

letter of containment in column titled “Function/Description” for valve FCV-43-55. 

The NRC staff determined that the change did not invalidate any of the staff’s previous findings.
6-18

6.2.5 Combustible Gas Control Systems

FSAR Section 3.8.2.2.3 “NRC Regulatory Guides” was revised to change the title of RG 1.7 

from “Control of Combustible Gas Concentrations in Containment Following a Loss of Coolant 

Accident” to “Control of Combustible Gas Concentrations in Containment.”

The change is consistent with the current title of RG 1.7 and does not invalidate any previous 

NRC review.

6.2.6 Containment Leakage Testing

Amendment 112 revised FSAR Table 6.2.6-3 “Valves Exempted from Type C Leak Testing” to 

change the valve type for the following valves:

• On Sheet 1 of 6, penetration X-8A, FROM: FCV 3-164A, TO: LCV-3-164A

• On Sheet 1 of 6, penetration X-8D, FROM: FCV 3-171A, TO: LCV-3-171A

• On Sheet 2 of 6, penetration X-13C, FROM: FCV 1-23, TO: PCV 1-23

• On Sheet 4 of 6, penetration X-40B, FROM: LXC 3-148A, TO: LCV 3-148A

The NRC staff finds the corrections in the valve numbers do not invalidate any previous NRC 

review. 

6.3 Emergency Core Cooling System

6.3.1 System Design

FSAR Amendment 112 revised Table 6.3-1 “Emergency Core Cooling System Component 

Parameters” (page 3 of 4) to:

• Add an asterisk to the heading labeled "Valves".

• Change the maximum stroke time for valves FCV-63-152 and FCV-63-153 from

12 seconds to 15 seconds.

• Change the maximum stroke time for valve FCV-63-1 from 20 seconds to 60 seconds.

• Change the maximum stroke time for valve FCV-74-3 from 17.1 seconds to 60 seconds.

• Change the maximum stroke time for valve FCV-74-21 from 17.1 seconds to

60 seconds.

• Change the maximum stroke time for valves FCV-63-93 and FCV-63-94 from

10 seconds to 40 seconds.

• Change the valve description for valve FCV-63-1 from “RWST to RHRPs Suction RWST

to” to “RWST to RHRPs Suction”
6-19

• Change the valve description for valve FCV-74-3 from “RHRPs Suction” to “RWST to

RHRPs Suction”

The NRC staff finds the changes do not invalidate any previous NRC review.

Amendment 112 revised FSAR Table 6.3-1 “Emergency Core Cooling System Component 

Parameters” (page 4 of 4) to:

• Add asterisks to the heading of the "Leakage Allowed" column and add a note to the

bottom of the table indicating that specific allowable leak rates are defined on valve data

sheets.

• Change “disc leakage” to “seat leakage” for valves a through f.

• Change the "Leakage Allowed" value for seat leakage per inch of nominal valve size for

items a, b and c (Conventional Globe Valves, Gate Valves and Check Valves) from "3

cc/hr" to "0-10 cc/hr."

The first two changes in page 4 of 4 of Table 6.3-1 are editorial and do not invalidate any 

previous NRC review. The general change in the gate, globe and check valve seat leakage 

from "3 cc/hr" to "0-10 cc/hr" is a change in component design requirement. For the gate, 

globe, and check valves used for containment isolation, the applicant will be required to meet 

the test requirements Type C tests in 10 CFR Part 50 Appendix J. The NRC staff finds the 

changes do not invalidate any previous NRC review.

In response to SCVB-RAI-1 (ADAMS Accession No. ML14352A248), the applicant provided a 

markup of item (1) “Residual Heat Removal Pumps” by adding the following statement: “A 

containment pressure of zero psig (building pressure of 14.3 pounds per square inch, 

atmospheric) is used in calculating the most limiting (minimum) NPSHA.”

In response to SCVB-RAI-2 (ADAMS Accession No. ML14352A248), the applicant provided a 

markup of item (1) “Residual Heat Removal Pumps” by replacing the terminology “containment 

over pressure” with “containment accident pressure.” The NRC staff finds the changes do not 

invalidate any previous NRC review.

The NRC staff has reviewed the changes made to FSAR Section 6.3.2, “System Design” from 

Amendment 92 up to Amendment 114. The NRC staff has also compared FSAR Section 6.3.2 

in Amendment 114 with the current Section 6.3.2 for WBN Unit 1. The NRC staff found most of 

the changes made were editorial in nature and, therefore, the NRC staff finds them to be 

acceptable. Those changes that were not editorial were related to updating the licensing basis 

for the modified ECCS sump strainer. By letter dated September 18, 2014 (ML14163A658), the 

NRC staff closed out its review of WBN, Unit 2’s response to Generic Letter 2004-02, “Potential 

Impact of Debris Blockage on Emergency Recirculation during Design Basis Accidents at 

Pressurized-Water Reactors.” Overall the NRC staff concluded that the information provided 

demonstrated that the debris will not inhibit the ECCS or containment spray system

performance of its intended function in accordance with 10 CFR 50.46 to assure adequate long 

term core cooling following a design-basis accident. In Section 3.2.16 of the NRC staff’s 

evaluation, the NRC staff validated that the WBN Unit 2 FSAR was updated to include the 

licensing basis for the modified ECCS sump strainer. Based on the fact the changes made in 
6-20

this section were either editorial or previously reviewed as part of the Generic Letter 2004-02 

review, the NRC staff considers the changes made to FSAR Section 6.3.2 to be acceptable.

6.3.3 Testing

The NRC staff has reviewed the changes made to FSAR Section 6.3.4, “Tests and Inspections” 

from Amendment 92 up to Amendment 114. The NRC staff also compared FSAR Section 6.3.4 

in Amendment 114 with the current Section 6.3.4 for WBN Unit 1. The NRC staff found that all 

of the changes made were editorial in nature and therefore found acceptable.

6.4 Control Room Habitability

FSAR Amendment 112 revised Section 6.4.4.2 “Toxic Gas Protection” to modify the first 

sentence of the second from last paragraph to change the nitrogen stored in the tank in the yard 

east of the control building from 286,900 standard cubic feet (scf) to 856,621 scf.

The NRC staff finds the revision acceptable because the nitrogen storage tank volume change 

is a change in the design in the holding volume of the storage tank. The NRC staff finds the 

changes do not invalidate any previous NRC review.

6.5 Engineered Safety Feature (ESF) Filter Systems

6.5.1 ESF Atmosphere Cleanup System

FSAR Amendment 112 revised Table 6.5-1 “Regulatory Guide 1.52, Rev. 2, Section 

Applicability for the Emergency Gas Treatment System” (page 1 of 2) to change “form” to “from” 

in note 1 and changed one of the referenced RGs from 1.25 to 1.183 in note 2.

The change in note 1 changing the word “form” to “from” is editorial and does not invalidate any 

previous NRC review. The NRC staff finds the change in note 2 from 1.25 to 1.183 acceptable

because the use of RG 1.183 is an option for the design of the EGTS adsorbers in RG 1.52 

Revision 4, Section C.2.c. 
7-1

7 INSTRUMENTS AND CONTROLS

7.5 Safety-Related Display Instrumentation

7.5.2 Postaccident Monitoring System

7.5.2.3 High-Range Containment Area Radiation Monitors

Disposition of Open Item (Appendix HH)

Open Item 79

Open Item 79 states: 

TVA should perform a radiated susceptibility survey, after the installation of the 

hardware but prior to the [radiation monitor (RM)]-1000 being placed in service, 

to establish the need for exclusion distance for the high range containment air 

radiation (HRCAR) monitors while using handheld portable devices (e.g., walkietalkie) in the control room, as documented in Attachment 23 to TVA’s letter, 

dated February 25, 2011, and item number 355 of TVA’s letter, dated April 15, 

2011.

By letter dated August 20, 2015 (ADAMS Accession No. ML15232A540), TVA provided the 

results of the August 3, 2015 electromagnetic interference (EMI)/radio frequency interface (RFI) 

survey that was conducted in the main control room (MCR) to address HRCAR monitors and 

other equipment in the MCR. This survey was conducted before and during the hot functional 

testing (HFT). Among other equipment, the survey included the results for the HRCAR 

monitors. In Enclosure 1, to the August 20, 2015 letter, the licensee provided the survey results 

and stated that the data were taken on two separate days. One set of data were obtained 

before HFT tests were conducted and the second set of data was obtained during the HFT 

tests.

The survey test spectrums are contained in Enclosure 1 to the August 20, 2015 letter. They 

show that the background field strength before and after HFT in the MCR are as follows:

Before HFT –

a. All but the cell phone signal is less than 80dBuV/m.

b. The cell phone signal is less than 90dBuV/m.

2. During HFT –

a. While the spectrum is slightly different the magnitudes are the same.

The above results show that the background electromagnetic compatibility (EMC) field strength 

in the MCR is very low. The field strength in the control room was observed to be less than 

100 dBuV/m. Field strength of 100 dBuV/m is equivalent to a field strength of 0.1 Volts 

(V)/meter (m) at 1 m.

Enclosure 2 to the August 20, 2015, TVA letter provides the various facets of the test including 

the survey test results. These results are applicable to the Eagle 21 Process Protection System 

and the HRCAR monitors. This enclosure describes the purpose, the regulatory commitments,
7-2

radiated survey test equipment, radiated survey methodology, acceptance criteria, analysis, 

results and conclusions, and references.

The test equipment antenna was placed in front of the HRCAR monitors in the MCR to scan 

frequencies from 10 kiloHertz to 1 gigaHertz and to monitor the radiated field strength values in 

the vicinity of the equipment including HRCAR monitors. The test results are included in the 

graphs provided in the August 20, 2015, letter.

TVA stated the acceptance criteria for the radiated EMI field strength before and during the HFT 

is 10 V/m or 140 dBuV/m. This acceptance criteria is based on TVA letter dated 

December 10, 2012 (ADAMS Accession No. ML12349A379). However, the NRC staff noted 

during teleconferences held on September 14 and 16, 2015, that the EMI/RFI tests conducted 

in that test report used the field strength of 10 V/m but these test sources were placed at a 

distance of 3 m and not next to the RM-1000 monitors as specified in the applicable acceptance 

criteria. 

During the September teleconferences, TVA was requested to clarify how the field strength of 

10V/m at a distance of 3 meters translates to the field strength at the face of or very near the 

face of the HRCAR monitors since TVA’s August 20, 2015, letter stated that there is no 

exclusion distance for the HRCAR monitors. TVA proposed to run actual tests using the walkietalkie radios and testing at all selectable frequencies at various distances to confirm whether an 

exclusion distance is needed or not. Confirmation of the successful walkie-talkie tests will also 

confirm that there is no RFI impact for the HRCAR monitors.

According to the TVA letter dated August 20, 2015, the maximum normal (with no external 

radiated emission sources e.g. no handheld portable devices) radiated field strength measured 

in the vicinity of HRCAR monitors before and during the HFT was found to be 0.1 V/m 

(100 dBuV/m). TVA also stated that extensive tests were done when TVA introduced Nextel 

walkie-talkies (radio phones) at another TVA site which produced a field strength of 

approximately 5 V/m at a distance of 1 meter. When these tests were performed on different 

equipment (Eagle 21, rod positioning equipment, process control equipment) no adverse effect 

was noticed even when the radio phones were placed next to the equipment. The Nextel 

phones were used at other TVA facilities for approximately 9 years without any observed 

performance anomalies. Based on these results, there were no required exclusion distance 

stated in the TVA letter of August 20, 2015. However, while these tests provide an indication of 

successful equipment testing, their results are not directly applicable to the WBN, Unit 2 

HRCAR monitors.

In the October 8, 2015 letter (ADAMS Accession No. ML15281A471), TVA stated that testing 

was conducted at the WBN, Unit 2 plant site using Harris Models XG-75 and XG-100 which are 

the two radio models planned for use in the WBN, Unit 2 MCR. Model XG-100 is the higher 

power of the two radios and does not use radio repeaters. Both radio models were tested on 

16 available channels. No disturbance was noticed on the HRCAR monitor at any distance 

starting from 3 feet (ft) to right next to the face of the HRCAR meter with the Harris XG-75 radio. 

No disturbance was noticed at a distance of 1 ft, 2 ft, or 3 ft when using the Harris 

XG-100 model radio. When the Harris XG-100 radio was tested immediately next to the face of 

the HRCAR monitor, some disturbance was noticed; however, no disturbance was noticed at a 

distance of 3 inches or more from the face of the HRCAR monitor.
7-3

TVA notes in its letter of October 8, 2015, that its plant training program instructs operators to 

prohibit the use of radios adjacent to the panels. TVA further stated that it is impractical to use 

the radio within 3 inches of the face of the panel where the HRCAR monitor is located. Further, 

TVA stated that HRCAR monitors do not have any actuation function and are used solely for 

indication. The NRC staff agrees that it is impractical to use the radios within 3 inches of the 

face of the panel mounted equipment. Since the only function of HRCAR monitors is indication 

even if there was some disturbance due to use of the radio at a distance of less than 3 inches it 

will not have any impact on plant safety. The NRC staff finds that based on the plant training 

program prohibiting the use of radio equipment adjacent to the MCR panels, and the 

impracticality of operating the radios within three inches of the panel where HRCAR monitors 

are mounted, it is acceptable for TVA to require no specified exclusion distance. Based on the 

satisfactory test results, in combination with no specified exclusion distance, the NRC staff finds 

the test results to be acceptable for WBN, Unit 2.

In the letter dated August 20, 2015 TVA stated that the background EMI/RFI field strength is 

assumed at 0.1 V/m whereas the measured EMI/RFI is always less than 0.1 V/m. TVA stated 

they have extensive operating experience to confirm that electronic equipment is unaffected at 

this low background EMI/RFI level. In addition, no adverse performance of the HRCAR 

equipment was observed before or during the HFT tests to confirm that background EMI/RFI 

field does not impact HRCAR monitors. The equipment was tested at 10 V/m at a distance of 3 

meters which is significantly higher than the background EMI/RFI. The field test at WBN, Unit 2 

confirmed that testing the walkie-talkies planned for use at WBN, Unit 2 at all selectable 

frequencies and at various distances had no negative impact on the HRCAR monitors in the 

control room even when the walkie-talkies were as close as 3 inches from the face of the 

HRCAR monitors. Open Item 79 has been adequately addressed therefore this item is

closed.

7.7 Control Systems Not Required for Safety

7.7.1 System Description

7.7.1.4 Distributed Control System

Disposition of Open Item (Appendix HH)

Open Item 83

Open Item 83 states: “TVA should confirm to the NRC staff the completion of the data storm 

test on the distribution control system.”

TVA provided its test plan for conducting a data storm test in a letter dated July 10, 2015 

(ADAMS Accession No. ML15196A515). This test plan describes the test requirements, test 

methodology, and the acceptance criteria for demonstrating that the nonsafety related Foxboro 

(Invensys) intelligent automation distributed control system (DCS) will continue to function with a 

failed communication network without any plant upset. 

The NRC staff evaluated the content of the TVA test plan, as well as performed an audit of the 

detailed test procedure associated with this test plan. The plan stated that two phases (loss of 

network and data storm) of network storm testing were to be administered. The TVA test plan 

provided in the July 10, 2015, letter did not describe the portion of the test plan for responding to 
7-4 

a loss of network, since this test was to be performed under a separate test. Regarding the 

data storm test, however, the plan describes the administration of a simulated data storm, to 

include the following: a broadcast storm and a multicast storm. The broadcast storm attempts 

to overwhelm the network with continuous data traffic at maximum speed, using a packet 

generator to send data onto the network with no specific address. This test is designed to 

validate the performance of the network switch data limiters. The multicast storm also attempts 

to overwhelm the network with continuous traffic at maximum speed, by using a packet 

generator to send data to specific media access control addresses. This test forces the DCS to 

attempt to process the network data and demonstrate that the addressed controller pairs 

continue to function properly independent of the network traffic. Also, the data storm testing 

consisted of one broadcast storm and one multicast storm injected at each network switch with 

all broadcast and multicast suppression capability disabled, and one broadcast and multicast 

storm injected at each network switch with the suppression capability enabled. 

TVA committed to perform the network data storm test with the DCS system installed and 

functional, prior to final commissioning. The DCS performance was to be monitored during the 

tests. System performance information and data would be accessed through monitoring of 

critical attributes of each control processor, including control processor loading, processor 

overruns, and the generation of system alarms. 

The test plan defined the acceptance criteria, which required a demonstration that the DCS is 

capable of continuing to retain its capability to perform process control functions as detailed in 

SSER-23 Section 7.7.1.4.4.1 items (1) through (8) during a simulated broadcast or multicast 

data storm. This demonstration is to be documented, in part, through monitoring of control 

system output demand signals. The test plan also required a demonstration that any failure that 

might occur within a single control processor of a pair does not propagate beyond the Foxboro 

intelligent automation mesh network from one segment to another. 

In its letter of October 9, 2015, (ADAMS Accession No. ML15282A537) TVA transmitted the 

results of this testing. The test report identified that all phases of the data storm test were 

successfully administered. During the DCS data storm test, at least one of the controllers in 

each set of DCS controller pairs serving each of the control functions continued to perform its 

required test set-up function. There were no control processor pairs where both processors 

failed during any section of data storm testing, enabling the DCS to continue to perform its 

required control functions, even with the storm suppression capability disabled. The NRC staff 

evaluated the completed test report and results. The NRC staff finds that the required test 

acceptance criteria were sufficiently met to demonstrate that the simulated multicast and 

broadcast data storms on the DCS network did not result in a complete loss of any of the DCS 

control functions, and that any test anomalies or deficiencies were appropriately addressed. 

Therefore the NRC staff concludes that the test plan and test results successfully demonstrate 

that the DCS will continue to perform its required functions in the presence of a simulated data 

storm affecting the communication network without any plant upset. Based on its review, the 

NRC staff considers Open Item 83 to be closed.
8-1 

8 ELECTRIC POWER SYSTEMS

8.3 Onsite Power Systems

8.3.1 Onsite AC Power System Compliance with GDC 17

8.3.1.1 Non-safety Loads Powered from the Class 1E AC Distribution System

During the Nuclear Regulatory Commission (NRC) staff’s review of essential raw cooling water 

(ERCW) and component cooling system (CCS), a potential vulnerability was identified by the 

NRC staff where sufficient cooling to both units during a specific postulated scenario could not 

be assured. As these systems are shared between Watts Bar Nuclear Plant (WBN), Units 1 

and 2, in order to address the NRC staff’s concern, the Tennessee Valley Authority (TVA) 

submitted a license amendment request for WBN, Unit 1 and updated the final safety analysis 

report (FSAR) for WBN, Unit 2. The NRC staff approved the WBN, Unit 1 amendment request 

by letter dated October 19, 2015 (Agencywide Documents Access and Management System 

(ADAMS) Accession No. ML15275A042). The WBN, Unit 1 amendment considered dual unit 

operation and the accompanying safety evaluation documents the NRC staff’s review. 

By letter dated September 11, 2015 (ADAMS Accession No. ML15279A332) TVA provided 

Amendment 114 to the FSAR for WBN, Unit 2, which included revisions to FSAR Section 

8.3.1.1 to address this issue. FSAR amendment 114, Section 8.3.1.1 for WBN, Unit 2 is 

consistent with the WBN, Unit 1 approved amendment, which considered dual unit operation, 

and is therefore acceptable. 

9-1 

9 AUXILIARY SYSTEMS

9.2 Water Systems

9.2.1 Essential Raw Cooling Water and Raw Cooling Water System

During the Nuclear Regulatory Commission (NRC) staff’s review of essential raw cooling water 

(ERCW) and component cooling system (CCS), a potential vulnerability was identified by the 

NRC staff where sufficient cooling to both units during a specific postulated scenario could not 

be assured. As these systems are shared between Watts Bar Nuclear Plant (WBN), Units 1 

and 2, in order to address the NRC staff’s concern, the Tennessee Valley Authority (TVA) 

submitted a license amendment request for WBN, Unit 1 and updated the final safety analysis 

report (FSAR) for WBN, Unit 2. The NRC staff approved the WBN, Unit 1 amendment request 

by letter dated October 19, 2015 (Agencywide Documents Access and Management System 

(ADAMS) Accession No. ML15275A042). The WBN, Unit 1 amendment considered dual unit 

operation and the accompanying safety evaluation documents the NRC staff’s review. 

By letter dated September 11, 2015 (ADAMS Accession No. ML15279A332) TVA provided 

Amendment 114 to the FSAR for WBN, Unit 2, which included revisions to FSAR Section 9.2.1

to address this issue. FSAR amendment 114, Section 9.2.1 for WBN, Unit 2 is consistent with 

the WBN, Unit 1 approved amendment, which considered dual unit operation, and is therefore 

acceptable. 

9.2.2 Component Cooling System (Reactor Auxiliaries Cooling Water System)

During the NRC staff’s review of ERCW and CCS, a potential vulnerability was identified by the 

NRC staff where sufficient cooling to both units during a specific postulated scenario could not 

be assured. As these systems are shared between WBN, Units 1 and 2, in order to address the 

NRC staff’s concern, TVA submitted a license amendment request for WBN, Unit 1 and updated 

the FSAR for WBN, Unit 2. The NRC staff approved the WBN, Unit 1 amendment request by 

letter dated October 19, 2015 (ADAMS Accession No. ML15275A042). The WBN, Unit 1 

amendment considered dual unit operation and the accompanying safety evaluation documents 

the NRC staff’s review. 

By letter dated September 11, 2015 (ADAMS Accession No. ML15279A332) TVA provided 

Amendment 114 to the FSAR for WBN, Unit 2, which included revisions to FSAR Section 9.2.2 

to address this issue. FSAR amendment 114, Section 9.2.2 for WBN, Unit 2 is consistent with 

the WBN, Unit 1 approved amendment, which considered dual unit operation, and is therefore 

acceptable. 

9.2.6 Condensate Storage Facilities

During the NRC staff’s review of ERCW and CCS, a potential vulnerability was identified by the 

NRC staff where sufficient cooling to both units during a specific postulated scenario could not 

be assured. As these systems are shared between WBN, Units 1 and 2, in order to address the 

NRC staff’s concern, TVA submitted a license amendment request for WBN, Unit 1 and updated 

the FSAR for WBN, Unit 2. The NRC staff approved the WBN, Unit 1 amendment request by 

letter dated October 19, 2015 (ADAMS Accession No. ML15275A042). The WBN, Unit 1 
9-2 

amendment considered dual unit operation and the accompanying safety evaluation documents 

the NRC staff’s review. 

By letter dated September 11, 2015 (ADAMS Accession No. ML15279A332) TVA provided 

Amendment 114 to the FSAR for WBN, Unit 2, which included revisions to FSAR Section 9.2.6

to address this issue. FSAR amendment 114, Section 9.2.6 for WBN, Unit 2 is consistent with 

the WBN, Unit 1 approved amendment, which considered dual unit operation, and is therefore 

acceptable. 

9.5 Other Auxiliary Systems

9.5.1 Fire Protection 

In Supplemental Safety Evaluation Report (SSER) 26 the NRC staff documented its review of 

the as-designed Fire Protection Report (FPR) submitted by TVA for (WBN), Units 1 and 2. 

SSER 26 stated:

On the basis of its review of TVA’s as-designed FPR and TVA’s supplemental 

information as referenced by this evaluation, the NRC staff concludes that the fire 

protection program for WBN, with the exception of Unit 1 specific OMAs [operator 

manual actions], meets 10 CFR 50.48(a) and GDC [General Design Criterion] 3 

of Appendix A to 10 CFR Part 50, and is consistent with Sections III.G, III.J, III.L, 

and III.O of Appendix R to 10 CFR Part 50 and Appendix A to BTP [branch 

technical position] (APCSB) 9.5-1, May 1976, with properly justified deviations 

and exceptions. Therefore, the NRC staff finds the as-designed FPR acceptable, 

contingent on the completion of the confirmatory items identified in Section 8.0 of 

this evaluation (Open items 140, 141, 142, and 143, Appendix HH). NRC 

approval of the Unit 1 OMAs is documented in SSER 18, October 1995, of 

NUREG-0847, “Safety Evaluation Report Related to the Operation of Watts Bar 

Nuclear Plant Units 1 and 2.”

Subsequent to the issuance of SSER 26, TVA submitted a revised FPR. Appendix FF of this 

SSER documents the NRC staff’s review of the revised WBN FPR. On the basis of its review of 

TVA’s as-constructed FPR and TVA’s supplemental information as referenced by this 

evaluation, the NRC staff concludes that, subject to the completion of the action in WBN, Unit 2 

license condition 2.C(9) (as described in section 4.3(b) of the staff’s evaluation in Appendix FF 

of this SSER), the fire protection program for WBN, with the exception of WBN, Unit 1 specific 

OMAs, meets Title 10 of the Code of Federal Regulations (10 CFR) 50.48(a) and GDC 3 of 

Appendix A to 10 CFR Part 50, and is consistent with Sections III.G, III.J, III.L, and III.O of 

Appendix R to 10 CFR Part 50 and Appendix A to BTP (APCSB) 9.5-1, May 1976, with properly 

justified deviations and exceptions. Therefore, the NRC staff concludes that the as-constructed 

FPR is acceptable. NRC approval of the WBN, Unit 1 OMAs is documented in SSER 18, 

October 1995, of NUREG-0847, “Safety Evaluation Report Related to the Operation of Watts 

Bar Nuclear Plant Units 1 and 2.” Therefore, the NRC staff considers Open items 140, 141, 

142, and 143 closed. 
10-1

10 STEAM AND POWER CONVERSION SYSTEM

10.4 Other Features

10.4.9 Auxiliary Feedwater System

During the Nuclear Regulatory Commission (NRC) staff’s review of essential raw cooling water 

(ERCW) and component cooling system (CCS), a potential vulnerability was identified by the 

NRC staff where sufficient cooling to both units during a specific postulated scenario could not 

be assured. As these systems are shared between Watts Bar Nuclear Plant (WBN), Units 1 

and 2, in order to address the NRC staff’s concern, the Tennessee Valley Authority (TVA) 

submitted a license amendment request for WBN, Unit 1 and updated the final safety analysis 

report (FSAR) for WBN, Unit 2. The NRC staff approved the WBN, Unit 1 amendment request 

by letter dated October 19, 2015 (Agencywide Documents Access and Management System 

(ADAMS) Accession No. ML15275A042). The WBN, Unit 1 amendment considered dual unit 

operation and the accompanying safety evaluation documents the NRC staff’s review. 

By letter dated September 11, 2015 (ADAMS Accession No. ML15279A332) TVA provided 

Amendment 114 to the FSAR for WBN, Unit 2, which included revisions to FSAR Section 10.4

to address this issue. FSAR amendment 114, Section 10.4 for WBN, Unit 2 is consistent with 

the WBN, Unit 1 approved amendment, which considered dual unit operation, and is therefore 

acceptable. 

11-1 

11 RADIOACTIVE WASTE MANAGEMENT

11.7 NUREG-0737 Items

11.7.1 Wide-Range Noble Gas, Iodine, and Particulate Effluent Monitors (II.F.1.1, 

II.F.1.2.a, and II.F.1.2.b)

Supplemental Safety Evaluation Report (SSER) 22 identified Section 11.7.1 status as “Open.” 

Section 11.7.1 relates to TMI Action Items II.F.1.1, II.F.1.2.a, and II.F.1.2.b. TMI Action Items 

Action Item II.F.1.1 requires in part, accident monitoring procedures to be available to ensure 

adequate calibration and operation of monitoring equipment for noble gas, iodine/particulate 

sampling, containment high range, containment pressure, containment water level, and 

containment hydrogen as specified in Regulatory Guide 1.97. NRC Inspection Report 

05000391/2015609, dated October 21, 2015 (Agencywide Documents Access and 

Management System (ADAMS) Accession No. ML15287A199) Section OA.1.1 states that the 

inspectors determined that accident monitoring procedures are available to ensure adequate 

calibration and operation of monitoring equipment for noble gas, iodine/particulate sampling, 

containment high range, containment pressure, containment water level, and containment 

hydrogen. Therefore, the Nuclear Regulaotry Commission (NRC) staff considers TMI Action 

Item II.F.1.1 closed.

TMI Action Item II.F.1.2.a requires that requires the installation of noble gas monitors that have 

the capability to detect and measure concentrations of noble gas fission products in plant 

gaseous effluents during and following an accident. NRC Inspection Report 05000391/2015608 

dated October 21, 2015 (ADAMS Accession No. ML15287A166) Section OA.1.4 documented 

inspections activities of TMI Action Item II.F.1 related to noble gas monitors. The NRC’s 

inspection report identified that eight radiation monitoring systems to be inspected within the 

scope of this requirement still required action to complete calibrations and verifications of 

electrical wiring and controls. In addition the NRC’s report contained non-cited violations which 

required proper installation of three misoriented main steam line monitors. NRC Inspection 

Report 05000391/2015609, dated October 21, 2015 (ADAMS Accession No. ML15287A199) 

Section OA.1.2 states that the inspectors determined that corrective actions were sufficient to 

address the non-cited violations involving improperly configured main steam line radiation 

monitors. In addition, the calibrations and verifications of radiation monitor output functions 

were sufficient to show that the outstanding actions to accomplish the installations of noble gas 

monitoring systems have been completed as required by TMI Action Item II.F.1. Therefore, the 

NRC staff considers TMI Action Item II.F.1.1.2.a related to noble gas monitoring closed. 

TMI Action Item II.F.1.2.b requires applicants to provide onsite laboratory facilities for analyses 

of radiological samples and to establish a capability to sample gaseous effluent streams to

detect post-accident releases of radioactive iodine. NRC Inspection Report 05000391/2013607, 

dated September 30, 2013 (ADAMS Accession No. ML13273A512) Section OA.1.8 states that a 

sufficient capability had been established to implement post-accident sampling of particulates as 

guided by TMI action II.F.1, Iodine Particulate Sampling. Therefore, the NRC staff considers 

TMI Action Item II.F.1.1.2.b closed. 
11-2

11.7.2 Primary Coolant Outside Containment (III.D.1.1)

SSER 22 identified Section 11.7.2 status as “Open.” Section 11.7.2 relates to TMI Action Item

III.D.1.1. TMI Action Item III.D.1.1 requires applicants to implement a program to reduce 

leakage from systems outside containment that would or could contain highly radioactive fluids 

during a serious transient or accident to as-low-as practical levels. The Watts Bar Nuclear Plant 

(WBN), Unit 2, proposed technical specification 5.7.2.4 addresses TMI action item III.D.1.1. 

The NRC staff’s review of the proposed technical specifications for WBN, Unit 2 is documented 

in SSER 29, Section 16. Therefore, the NRC staff considers TMI Action Item III.D.1.1 closed.
12-1

12 RADIATION PROTECTION

12.7 NUREG-0737 Items

12.7.1 Plant Shielding (II.B.2)

Amendment 97 revised the list of areas for which operators need access during an accident 

(vital areas) in Final Safety Analysis Report (FSAR) Section 12.3.2.2, adding three and deleting 

the post-accident sample sink. The staff requested additional information related to the dose 

consequences of these vital missions, including plant layout drawings depicting radiation zones 

during accident conditions and access/egress routes. By letters dated June 3, 2010 

(Agencywide Documents Access and Management System (ADAMS) Accession No. 

ML101600509), and December 10, 2010 (ADAMS Accession No. ML103480708), TVA provided 

dose calculations and plant layout drawings (maps) depicting the access, and egress from, 

WBN vital areas. In response to a NRC staff request for additional information, TVA 

supplemented this information in a letter dated February 25, 2011 (ADAMS Accession No. 

ML110620687). TVA committed to clarify the calculational basis, and establish corresponding 

implementing procedures, contained in the February 25, 2011, letter, which are subject to

verification by the NRC inspection program.

By letter dated June 17, 2015 (ADAMS Accession No. ML15170A473), TVA identified an 

additional vital area, (concerning starting pumps in the emergency core cooling system and 

emergency raw water system during an accident) that may be required to support dual unit 

operations. In a supplemental submittal dated July 14, 2015 (ADAMS Accession No. 

ML15197A357), TVA provided calculations, including plant layouts and access/egress paths 

estimating the mission dose for the operators completing this vital mission.

 

The NRC staff concludes that TVA has demonstrated by design calculations that the actions 

necessary to mitigate the consequences of a design-basis accident at WBN, Unit 2, will 

maintain occupational doses to plant operators within the dose criteria of General Design 

Criterion 19, per the guidance contained in NUREG-0737, Item II.B.2. Therefore, the NRC staff 

concludes that the shielding design for WBN, Unit 2, is acceptable.

12.7.2 High Range In-Containment Monitor (II.F.1.2.c)

Supplemental Safety Evaluation Report (SSER) 22 identified Section 12.7.2 status as “Open.” 

Section 12.7.2 relates to TMI Action Item II.F.1.2.c. TMI Action Item II.F.1.2.c requires 

applicants to install high range monitoring instrumentation with the capability to detect and 

measure the radiation level within containment during and following an accident. NRC 

Inspection Report 05000391/2015609, dated October 21, 2015 (ADAMS Accession No. 

ML15287A199) Section OA.1.3 states that the calibrations and verifications of radiation monitor 

output functions were sufficient to show that the outstanding actions to accomplish the 

installations of containment high range monitoring systems have been completed as required by 

TMI Action Item II.F.1.2.c. Therefore, the NRC staff considers TMI Action Item II.F.1.2.c closed.

12.7.3 In-Plant Radioiodine Monitor (III.D.3.3)

SSER 22 identified Section 12.7.3 status as “Open.” Section 12.7.3 relates to TMI Action 

Item III.D.3.3. TMI Action Item III.D.3.3 required each applicant to provide instrumentation for 

accurately determining in-plant airborne radioiodine concentrations to minimize the need for 
12-2

unnecessary use of respiratory protection equipment. NRC Inspection Report 

05000391/2012605 dated August 7, 2012 (ADAMS Accession No. ML12220A536), Section 

OA.1.14, states that based on a review of the applicant’s final completion and inspection efforts, 

the inspectors determined that the measures implemented by the applicant were sufficient to 

address the requirements of TMI Action Item III.D.3.3. Therefore, the NRC staff considersTMI 

Action Item III.D.3.3 closed.
15-1

15 ACCIDENT ANALYSES

15.3 Limiting Accidents

15.3.1 Loss-of-Coolant Accident (LOCA)

Nuclear Safety Advisory Letters Concerning Heat Flux Hot Channel Factor

As discussed in Section 16 of this SSER, the license authorizing operation of Watts Bar Nuclear 

Plant (WBN), Unit 2 includes technical specifications (TSs), which are derived from the analyses 

and evaluation included in the safety analysis report, and amendments thereto, submitted 

pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.34. Among the 

requirements in the TS 5.9.5 for WBN, Unit 2 is a requirement to establish core operating limits 

prior to the initial and each reload cycle, or prior to any remaining portion of a cycle, and to 

document those limits in the Core Operating Limits Report (COLR). The analytical methods 

used to determine the core operating limits shall be those previously reviewed and approved by 

the Nuclear Regulatory Commission (NRC), and those methods are listed in TS 5.9.5.b. As 

relevant here, one limit required to be calculated in accordance with NRC approved methods 

and documented in the COLR is the Heat Flux Hot Channel Factor, which is used in Limiting 

Condition for Operation 3.2.1. Per TS 5.9.5.b.3, the NRC will require TVA to use the following 

code: 

WCAP-10216-P-A, Revision 1A, “Relaxation of Constant Axial Offset Control 

F(Q) Surveillance Technical Specification,” February 1994 (W Proprietary). 

(Methodology for Specifications 3.2.1 - Heat Flux Hot Channel Factor (W(Z) 

Surveillance Requirements For F(Q) Methodology) and 3.2.3 - Axial Flux 

Difference (Relaxed Axial Offset Control).) 

In Nuclear Safety Advisory Letter, NSAL-09-5 Revision 1, "Relaxed Axial Offset Control FQ 

Technical Specification Actions," dated September 24, 2009, and in Nuclear Safety Advisory 

Letter, NSAL-15-1, "Heat Flux Hot Channel Factor Technical Specification Surveillance," dated 

February 3, 2015, Westinghouse alerted licensee of potential problems with the calculation of 

the Heat Flux Hot Channel Factor, and Westinghouse Electric Corporation (Westinghouse) 

suggested changes to the existing plants TSs. By letter dated September 30, 2015 

(Agencywide Documents Access and Management System (ADAMS) Accession No. 

ML15274A328) TVA provided a technical and regulatory evaluation of how the WBN, Unit 2 

proposed TSs address NSAL-09-5, Revision 1 and NSAL-15-1. TVA concluded in its evaluation

that the WBN, Unit 2 proposed TS 3.2.1 completely addresses the issues identified in both of 

these NSALs. 

On October 9, 2015, in response to the NRC staff’s questions associated with how TVA 

addressed the issues identified in the NSALs, TVA (1) provided a license condition where the 

margins provided in TS 3.2.1 will be evaluated each fuel cycle to assure that the related 

assumptions in the accident analyses are maintained, and (2) submitted a summary of a 

Westinghouse margin assessment report for Cycle 1 providing additional information related to 

the proposed TS 3.2.1.
15-2

The proposed license condition requires TVA to perform cycle-specific evaluations of the 

margins of the Required Actions in TS 3.2.1 Condition B, and states:

TVA will verify for each core reload that the actions taken if FQW(Z) is not within 

limits will assure that the limits on core power peaking FQ(Z) remain below the 

initial total peaking factor assumed in the accident analyses.

The results of this verification will be documented in the final Reload Safety Evaluation that the 

vendor provides WBN, Unit 2 for each core reload, which will be reviewed by TVA, and is a 

quality assurance record as defined in 10 CFR 50, Appendix B, and available for NRC 

inspection.

While the NRC staff has not come to a final decision on any actions required for alreadylicensed plants using different versions of TS 3.2.1 that are impacted by the NSALs, the NRC 

staff has reached a decision for WBN, Unit 2. Specifically, as detailed in Section 16 of this 

SSER, the proposed TS, which include provisions to address the NSALs, are acceptable. 

Further, the additional analyses imposed though the new license condition will provide 

additional assurances that the actions taken under TS 3.2.1 are appropriate. 

15.5 NUREG-0737 Items

15.5.5 Small-Break LOCA Methods (II.K.3.30) and Plant-Specific Calculations (II.K.3.31)

Supplemental Safety Evaluation Report (SSER) 22 identified Section 15.5.5 status as “Open.” 

Section 15.5.5 relates to TMI Action Items II.K.3.30 and II.K.3.31 which required that applicants 

verify and document that small break loss of coolant accident analysis be compliant with 

10 CFR Part 50 Appendix K and 10 CFR 50.46. NRC Inspection Report 05000391/2011603 

dated May 16, 2011 (ADAMS Accession No. ML111370702), Section OA.1.15, states that the 

inspectors reviewed various completed actions associated with TMI action items II.K.3.30 and 

II.K.3.31 to verify the adequacy of the applicant’s actions. The inspectors concluded that the 

applicant’s efforts were sufficient to satisfy the intent of the respective TMI action items. 

Therefore, the NRC staff considers TMI Action Items II.K.3.30 and II.K.3.31 closed.
16-1

16 TECHNICAL SPECIFICATIONS

16.1 Introduction

Regulatory Standard

An applicant for an operating license must propose technical specifications (TSs). Pursuant to 

Title 10 of the Code of Federal Regulations (10 CFR) 50.36(a)(1), 

Each applicant for a license authorizing operation of a production or utilization 

facility shall include in his application proposed technical specifications in 

accordance with the requirements of this section. A summary statement of the 

bases or reasons for such specifications, other than those covering 

administrative controls, shall also be included in the application, but shall not 

become part of the technical specifications.

When the NRC issues an operating license, that license will include technical specifications. As 

stated in 10 CFR 50.36(b),

Each license authorizing operation of a production or utilization facility of a type 

described in § 50.21 or § 50.22 will include technical specifications. The technical 

specifications will be derived from the analyses and evaluation included in the 

safety analysis report, and amendments thereto, submitted pursuant to § 50.34. 

The Commission may include such additional technical specifications as the 

Commission finds appropriate.

With respect to the content of the technical specifications, as stated in 10 CFR 50.36(c)(1) to 

(c)(8), TSs will include items in several categories, including: (1) safety limits, limiting safety 

system settings, and limiting control settings, (2) limiting conditions for operation, 

(3) surveillance requirements, (4) design features, and (5) administrative controls. 

Information in the Application

The Tennessee Valley Authority (TVA) stated that the proposed Watts Bar Nuclear Plant 

(WBN), Unit 2 TSs were developed from the NUREG-1431, “Standard Technical Specifications 

[STSs] – Westinghouse Plants,” Revision 0 and then incorporating the WBN, Unit 1 TS 

substantial amendments though Amendment 70. 

TVA submitted Developmental Revision I of the WBN, Unit 2 TSs by letter dated June 16, 2014 

(Agencywide Documents Access and Management System (ADAMS) Accession No. 

ML14169A525), and was made available to the Nuclear Regulatory Commission (NRC) staff for 

comment as a “proof and review” revision. 

TVA submitted Developmental Revision J by letter dated July 6, 2015 (ADAMS Accession No. 

ML15187A461), which incorporated responses to the NRC staff’s requests for additional 

information and comments on revision I and incorporated recent amendments to the WBN, 

Unit 1 TSs. The recent amendments included use of alternate qualified offsite power circuits, a 

revision to the diesel generator frequency band, and additional TSs to address shared use of 

cooling water systems between units. A supplement to Revision J, submitted by letter dated

September 4, 2015 (ADAMS Accession No. ML15247A564), added bases changes omitted 
16-2

from Revision J, incorporated late identified changes, and improved the added specifications. 

The applicable NRC staff reviewed Developmental Revision J and the supplement and identified 

final administrative issues that were resolved in the final revision to the WBN, Unit 2 TSs that

were submitted by TVA in a letter dated September 23, 2015 (ADAMS Accession No. 

ML15267A183). The final revision additionally addressed a generic potentially non-conservative 

power distribution limit. TVA certified that the final revision WBN, Unit 2 TSs were accurate.

16.2 Evaluation

The NRC staff’s review of the proposed TSs for WBN, Unit 2 was completed per the guidance in 

NUREG-0800 “Standard Review Plan for the Review of Safety Analysis Reports for Nuclear 

Power Plants, LWR Edition, “ Section 16.0 “Technical Specifications” (ADAMS Accession No. 

ML070380224) and NUREG-1431, “Standard Technical Specifications, Westinghouse Plants, 

Revision 4 (STS)” (ADAMS Accession No. ML12100A222).

The review of the proposed WBN, Unit 2 TSs was informed by the years of operational 

experience with the essentially-the-same technical specifications for WBN, Unit 1. This is in line 

with the Commission’s expectations. In a July 25, 2007, Staff Requirements Memorandum 

SECY-07-0096, “Possible Reactivation of Construction and Licensing Activities for the Watts 

Bar Nuclear Plant Unit 2,” (ADAMS Accession No. ML072060688) the Commission approved 

the NRC staff’s recommendations for the approach to licensing TVA, WBN, Unit 2. The 

Commission supported a licensing review approach that employed the current licensing basis 

for WBN, Unit 1 as the reference basis for the review and licensing of WBN, Unit 2.

Review of the WBN, Unit 2 TSs was conducted by comparing the proposed WBN, Unit 2 TSs to 

the current TSs for WBN, Unit 1. Changes were allowed where necessary to accommodate 

unique features of WBN, Unit 2 and to support dual unit operation. Additions and changes to the 

WBN, Unit 2 TSs were evaluated using the appropriate requirements and guidance.

The NRC staff evaluated the WBN, Unit 2 TSs by confirming that the WBN, Unit 2 TSs were 

substantially the same as the WBN, Unit 1 TSs, through amendment 101 and including the 

recent WBN, Unit 1 amendments submitted to support dual unit operations in accordance with 

SRM SECY-07-0096. 

In WBN, Unit 2 Developmental TSs Revisions A through J, the licensee proposed various 

changes to the WBN, Unit 2 TSs and to the Final Safety Analysis Report to bring the WBN, 

Unit 2 TSs into fidelity with current TSs of WBN, Unit 1. The NRC staff paid special attention to 

items unique to WBN, Unit 2 and any changes needed to support dual unit operation. The NRC 

staff requested additional information and commented on the various revisions. 

Where additions and changes in the TSs were necessary due to uniqueness in design, the NRC 

staff both ensured that the changes were necessary for inclusion into the TSs per the four 

criteria of 10 CFR 50.36(c)(2)(ii) and were consistent with the guidance of NUREG-0800 and 

NUREG-1431. The NRC staff evaluated the additional restrictions on plant operation to ensure 

that they enhance plant safety. 

Therefore, the NRC staff confirmed that format and content of the proposed TSs are consistent 

with the WBN, Unit 1 TSs. Differences caused by the unique design features of WBN, Unit 2 

and due to dual unit operation were evaluated for compliance with 10 CFR 50.36 and 

agreement with the guidance of NUREG-1431, Revision 4. Since the WBN, Unit 1 TSs comply 
16-3

with 10 CFR 50.36, and the proposed WBN, Unit 2 TSs are either (1) the same as WBN, 

Unit 1’s or (2) TVA adequately explained unit differences, the NRC staff concludes that the 

proposed WBN, Unit 2 TSs also will comply with 10 CFR 50.36. 

16.3 Conclusion

Overall on the basis of the staff’s review of the WBN, Unit 2 TSs, as discussed above, the NRC 

staff concludes that the WBN, Unit 2 TSs are consistent with the regulatory guidance contained 

in the Westinghouse STSs, and contain design specific parameters and additional TS 

requirements considered appropriate by the NRC staff. Therefore, the NRC staff concludes that 

the WBN, Unit 2 TSs satisfy 10 CFR 50.36 and are acceptable.

17-1

17 QUALITY ASSURANCE

17.6 MAINTENANCE RULE

Pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.65(a)(1) (the “maintenance 

rule”): 

Each holder of an operating license for a nuclear power plant ... shall monitor the 

performance or condition of structures, systems, or components, against 

licensee-established goals, in a manner sufficient to provide reasonable 

assurance that these structures, systems, and components ...are capable of 

fulfilling their intended functions.

The above regulation only applies to a holder of an operating license, whereas TVA currently 

holds a construction permit for Watts Bar Nuclear Plant (WBN), Unit 2, but does have an 

operating license for WBN, Unit 1. However, to issue an operating license for WBN, Unit 2, the 

NRC must first make the findings of 10 CFR 50.57(a)(1) to (a)(6), including the finding of 

10 CFR 50.57(a)(2) (“facility will operate in conformity with ... the rules and regulations of the 

Commission”) and 50.57(a)(3)(ii) (reasonable assurance that activities authorized by the license 

will be conducted in compliance with the Commission’s regulations). Accordingly, the Nuclear 

Regulatory Commission (NRC) staff reviewed if the Tennessee Valley Authority (TVA) would be 

meeting 10 CFR 50.65(a)(1) after an operating license was issued for WBN, Unit 2. 

In TVA’s regulatory framework, it committed to implementing the maintenance rule program for 

WBN, Unit 2 systems at least one month prior to fuel load for all systems, structures, and 

components (SSCs). By letter dated July 27, 2015, (Agencywide Documents Access and 

Management System (ADAMS) Accession No. ML15209A458) TVA stated that it considered 

this commitment met on March 3, 2014, with the implementation of Technical Instruction 0-TI119, “Maintenance Rule Performance Indicator Monitoring, Trending, and Reporting – 10 CFR 

50.65.” 

WBN, Unit 1 has been maintaining its program required by 10 CFR 50.65, and the NRC has 

been routinely inspecting the program, since the 10 CFR 50.65 took effect in 1996. By letter 

dated July 2, 2010 (ADAMS Accession No. ML101720050), the NRC staff documented the 

detailed results of its overall evaluation of TVA’s construction refurbishment plan. The

refurbishment plan was developed to ensure that the design and licensing basis, including 

original equipment design specifications, would be met. In its evaluation, the NRC staff found 

that WBN, Unit 2 SSCs that are shared between WBN, Units 1 and 2 and that are in operation 

to support WBN, Unit 1 were already covered by the existing program and met the requirements 

of 10 CFR 50.65 for WBN, Unit 1. The NRC staff concluded that TVA’s plan provided 

reasonable assurance that the structures, systems, and components are capable of fulfilling 

their intended functions relative to WBN, Unit 2.

From June 22 to June 26, 2015, the NRC staff performed an inspection at the WBN, Unit 2 site

of the readiness of TVA to operate WBN, Unit 2 and safely integrate WBN, Unit 2 into the 

current organization that is responsible for the safe operation of WBN, Unit 1. By letter dated 

August 14, 2015 (ADAMS Accession No. ML15226A212), the NRC staff published the findings 

of its operational readiness assessment team (ORAT) inspection. During the inspection, the 

NRC reviewed TVA’s actions taken under 0-TI-119. The NRC’s ORAT also looked at 
17-2

maintenance and quality assurance support activities, including the maintenance organization, 

maintenance effectiveness, and work management and prioritization. The inspection concluded 

that site maintenance and quality organizations can support startup of WBN, Unit 2, 

maintenance has been effective, and TVA has adequately demonstrated their ability to prioritize 

and complete work for a two unit site.

Based on the above, the NRC staff concludes that there is reasonable assurance that TVA’s 

maintenance rule program, as applied to WBN Unit 2, will be conducted in accordance with 

10 CFR 50.65. 
22-1

22 FINANCIAL PROTECTION AND INDEMNITY REQUIREMENTS

22.3 Operating Licenses 

Open Item 25

Open Item 25 states:

Prior to the issuance of an operating license (OL), the Tennessee Valley 

Authority (TVA) is required to provide satisfactory documentation that it has 

obtained the maximum secondary liability insurance coverage pursuant to

10 CFR 140.11(a)(4), and not less than the amount required by 10 CFR 

50.54(w), with respect to property insurance, and the NRC staff has reviewed 

and approved the documentation. 

By letter dated October 2, 2015, the Tennessee Valley Authority (TVA) submitted “Watts Bar, 

Unit 2 - Financial Protection Requirements and Indemnity Agreements - Insurance Status 

Second Supplement,” (Agencywide Documents Access and Management System (ADAMS)

Accession No. ML15272A394), to meet the requirements of Title 10 of the Code of Federal 

Regulations (10 CFR) 140 and 10 CFR 50.54(w).

The provisions of the Price-Anderson Act (PAA), Section 170 of the Atomic Energy Act, and the 

Commission’s regulations in 10 CFR 140, “Financial Protection Requirements and Indemnity 

Agreements,” require, in part, each holder of a license issued pursuant to 10 CFR 50, “Domestic 

Licensing of Production and Utilization Facilities,” to have and maintain nuclear energy liability 

insurance (also known as financial protection) to pay for claims of bodily injury and property 

damage resulting from a nuclear incident. Specifically, 10 CFR 140.11(a)(4) requires each 

licensee for a reactor with a rated capacity of 100,000 kilowatts or more to have and maintain 

financial protection in an amount equal to the sum of primary financial protection ($375 million) 

and the amount available as secondary financial protection to satisfy the PAA requirements. 

During the period of construction, and prior to the period of operation, however, an applicant for 

a license to operate a nuclear reactor who holds a construction permit for a large operating 

reactor is only required to comply with the insurance requirements in 10 CFR 140.13, “Amount 

of financial protection required of certain holders of construction permits and combined licenses 

under 10 CFR part 52.” As required by 10 CFR 140.13, an applicant shall have and maintain 

financial protection in the amount of $1 million. By letter dated April 29, 2010, “Watts Bar 

Nuclear Power Plant (WBN Unit 2 – Financial Protection Requirements and Indemnity 

Agreement – Insurance Status)” (ADAMS Accession No. ML101250300), TVA submitted a copy 

of the certificate of insurance for Watts Bar Nuclear Plan (WBN) Unit 2 to demonstrate that they 

meet the requirements under 10 CFR 140.11(a)(4) and 10 CFR 140.13. Based on its review of 

this information, the staff concludes that TVA has met the requirements for primary financial 

protection. 

By letter dated January 24, 2013, “Watts Bar Nuclear Plant Unit 2 – Documentation of Liability 

Insurance Coverage Requested for Supplemental Safety Evaluation Report (SSER) Open 

Item 25,” (ADAMS Accession No. ML13030A032), TVA submitted documentation to meet the 

secondary financial protection requirements under 10 CFR 140.11(a)(4). The documentation 

provided by TVA entitled, “Certificate of Insurance Declarations and Bond for Payment of 

Retrospective Premiums,” serves as a contract between TVA and American Nuclear Insurers, 
22-2

and requires TVA to pay into the secondary financial insurance pool should a nuclear incident 

occur. Based on its review of this information, the staff concludes that TVA has met the 

requirement under 10 CFR 140.11(a)(4). 

As required by 10 CFR 140.21, each reactor licensee should be able to demonstrate its financial 

capacity to pay into the secondary tier of financial protection for each reactor it is licensed to 

operate and insure pursuant to 10 CFR 140.11(a)(4); $121.255 million per incident and up to 

$18.963 million per year. By letter dated October 2, 2015, TVA submitted certified financial 

documents demonstrating a cash flow (i.e., cash available to a company after all operating 

expenses, taxes, interest charges, and dividends have been paid) can be obtained, and would 

be available, for payment of retrospective premiums within three (3) months after this submittal. 

TVA’s submittal also demonstrates the financial capacity to provide secondary financial 

protection for its entire nuclear fleet, which includes WBN, Unit 2 and six other plants. Based on 

its review of this information, the staff concludes that TVA has met the requirements under 

10 CFR 140.21 for proof of financial capacity to pay into the secondary tier of financial 

protection.

As required by 10 CFR 50.54(w), each reactor licensee must obtain onsite property insurance to 

cover costs associated with stabilizing the reactor and decontaminating the reactor and reactor 

site in the event of a nuclear incident. Onsite insurance must have a minimum coverage limit for 

each reactor site of either $1.06 billion or whatever amount of insurance is generally available 

from private sources, whichever is less. By letter dated September 24, 2015, “Watts Bar 

Nuclear Plant Unit 2 - Financial Protection Requirements and Indemnity Agreements -

Insurance Status Supplement (ADAMS Accession No. ML15268A528) TVA submitted a copy of 

their onsite insurance policy provided by Nuclear Electric Insurance Limited, which 

demonstrates that it possesses financial protection to cover the licensee's obligation to stabilize 

and decontaminate the reactor and the reactor site in the event of a nuclear incident. Therefore, 

the staff concludes that TVA has met the requirements under 10 CFR 50.54(w) for onsite 

insurance.

Finally, the NRC has an indemnity agreement with TVA and will amend the existing agreement 

to include WBN, Unit 2 concurrent with the issuance of the Operating License. 

In consideration of TVA’s submittals, the NRC staff concludes that TVA has adequately 

addressed the provisions of the Price-Anderson Act (Section 170 of the Atomic Energy Act) and 

the applicable Commission regulations in 10 CFR 140 and 10 CFR 50.54(w) for WBN, Unit 2. 

Open Item 25 has been adequately addressed therefore this item is closed.
A-1

APPENDIX A CHRONOLOGY OF RADIOLOGICAL REVIEW OF

WATTS BAR NUCLEAR PLANT, UNIT 2, OPERATING LICENSE 

REVIEW

Public correspondence exchanged between the U.S. Nuclear Regulatory Commission (NRC)

and the Tennessee Valley Authority (TVA) during the review of the operating license (OL) 

application for Watts Bar Nuclear Plant (WBN), Units 1 and 2, is available through the NRC’s

Agencywide Documents Access and Management System (ADAMS) or the Public Document 

Room (PDR). This correspondence includes that occurring subsequent to TVA’s letter notifying 

the NRC of its decision to reactivate construction of WBN, Unit 2, which had been in a deferred 

status under the Commission’s Policy Statement on Deferred Plants. 

Web-based ADAMS (WBA) is the latest interface to ADAMS. This search engine enables 

searching the ADAMS repository of official agency records (Publicly Available Records System 

and Public Legacy libraries) for publicly available regulatory guides, NUREG-series reports, 

inspection reports, Commission documents, correspondence, and other regulatory and technical 

documents written by NRC staff, contractors, and licensees. WBA permits full-text searching 

and enables users to view document images, download files, and print locally. New documents 

become accessible on the day they are published, and are released periodically throughout the 

day. ADAMS documents are provided in Adobe Portable Document Format. 

The NRC PDR reference staff is available to assist with ADAMS. Contact information for the 

PDR staff is on the NRC website at http://www.nrc.gov/reading-rm/contact-pdr.html.

E-1

APPENDIX E PRINCIPAL CONTRIBUTORS TO SSER 29

M. Farnan, NRR/DE/EPNB

R. Pedersen, NRR/DRA/ARCB

C. Moulton, NRR/DRA/AFPB

D. Frumkin, NRR/DRA/AFPB

E. Olvera, NRR/DIRS/IFAIB

P. Snyder, NRR/DSS/STSB

J. Poole, NRR/DORL/LPWB

G. Singh, NRR/DE/EICB

D. Rahn, NRR/DE/EICB

D. Woodyatt, NRR/DSS/SCVB

FF-1

APPENDIX FF FIRE PROTECTION PROGRAM SAFETY EVALUATION 

WATTS BAR NUCLEAR PLANT, UNITS 1 AND 2

1.0 INTRODUCTION

The Tennessee Valley Authority (TVA) is the licensee for Watts Bar Nuclear Plant (WBN), 

Unit 1, and is the applicant for an operating license for WBN, Unit 2. 

Pursuant to 10 CFR 50.48, Fire Protection:

(a)(1) Each holder of an operating license ... must have a fire protection plan that 

satisfies Criterion 3 of appendix A to [10 CFR part 50]. This fire protection plan 

must:

(i) Describe the overall fire protection program for the facility;

(ii) Identify the various positions within the licensee's organization that are 

responsible for the program;

(iii) State the authorities that are delegated to each of these positions to 

implement those responsibilities; and

(iv) Outline the plans for fire protection, fire detection and suppression 

capability, and limitation of fire damage.

(2) The FPP must also describe specific features necessary to implement the 

program described in paragraph (a)(1) of this section such as--

(i) Administrative controls and personnel requirements for fire prevention 

and manual fire suppression activities;

(ii) Automatic and manually operated fire detection and suppression 

systems; and

(iii) The means to limit fire damage to structures, systems, or components 

important to safety so that the capability to shut down the plant safely is 

ensured.

TVA submitted the As-Designed Fire Protection Report (FPR) for WBN, Units 1 and 2, to the 

U.S. Nuclear Regulatory Commission (NRC) by letter dated December 18, 2010, as revised 

and supplemented by letters dated December 20, 2010; January 14, March 16 and 31, May 6, 

18, and 26, June 7 and 17, July 1 and 22, August 5 and 15, September 30, October 28, 

November 21 and 30, 2011; March 13, April 12, 17, and 26, May 9 and 30, June 7 and 27, 

July 19, September 13, December 20, 2012; February 7 and 28, and March 13, 2013. TVA 

submitted the As-Constructed FPR for WBN, Units 1 and 2, by letter dated June 24, 2015, as 

revised and supplemented by letters dated August 27, and October 7, 2015.
FF-2

The FPR describes the measures that are established at WBN to implement a defense in depth 

fire protection program in plant areas important to safety. The objective of these measures is 

to: (1) prevent fires from starting; (2) detect rapidly, control, and extinguish promptly those fires 

that do occur; and (3) provide protection for SSCs important to safety so that a fire that is not 

promptly extinguished by the fire suppression activities will not prevent the safe shutdown of the 

plant.

In the FPR, TVA stated that “the purpose of the Fire Protection Report (FPR) is to consolidate 

a sufficiently detailed summary of the WBN regulatory required Fire Protection Program into a 

single document and to reflect the design as-constructed at the time of fuel load.” The FPR 

describes the operational phase of the fire protection program. The FPR comprises the 

following parts: 

• Part I is an introduction to the FPR and contains a summary table of fire protection 

features throughout the plant.

• Part II contains the overall fire protection plan. The fire protection plan describes (1) the 

WBN fire protection organization, (2) plant fire protection features, (3) the plant’s fire 

prevention program, (4) the plant’s emergency response organization, (5) plant 

operating requirements for fire protection features and systems, and (6) the testing and 

inspection requirements for these plant fire protection features.

• Part III contains an overview of the post-fire safe shutdown (FSSD).

• Part IV discusses alternate shutdown.

• Part V describes operator manual actions (OMAs) and repairs.

• Part VI summarizes the fire hazards analysis for each fire area by describing the 

physical characteristics of the fire area, combustible loadings and anticipated fire 

severity, and fire suppression and detection capability available in each plant area. 

Part VI also describes how the plant would achieve post-FSSD if a serious fire occurred 

in the fire area.

• Part VII documents deviations from regulatory criteria and guidance documents and 

presents engineering evaluations related to the adequacy of specific fire protection 

features, including Operator Manual Actions (OMAs) required for safe shutdown.

• Parts VIII and IX describe conformance with the guidelines in Appendix A to Branch 

Technical Position (BTP) (Auxiliary Power Conversion Systems Branch (APCSB)) 9.5-1 

and in Sections III.G, III.J, III.L, and III.O of Appendix R to 10 CFR Part 50, respectively.

• Part X contains a discussion of TVA’s compliance with National Fire Protection 

Association codes.
FF-3

TVA’s fire protection program is required to comply with the following:

• General Design Criterion (GDC) 3, “Fire Protection,” of Appendix A, “General Design 

Criteria for Nuclear Power Plants,” to Title 10 of the Code of Federal Regulations 

(10 CFR) Part 50, “Domestic Licensing of Production and Utilization Facilities”

• 10 CFR 50.48, “Fire protection,” paragraph (a)

In addition to these requirements, TVA committed in the FPR that its fire protection program has 

been developed to comply with, and is based on:

• Sections III.G, III.J, III.L, and III.O of Appendix R, “Fire Protection Program for 

Nuclear Power Facilities Operating Prior to January 1, 1979,” to 10 CFR Part 50

• Appendix A to Auxiliary Power Conversion Systems Branch’s Branch Technical 

Position 9.5-1, “Guidelines for Fire Protection for Nuclear Power Plants Docketed 

Prior to July 1, 1976.”

Because the operating licenses for WBN were not issued prior to January 1, 1979, TVA is not 

required to comply with 10 CFR Part 50 Appendix R, but has committed to do so. In the FPR, 

TVA describes inconsistancies with Appendix R as deviations. In this evaluation, these 

instances will be described as alternatives from the Appendix R information. The NRC staff 

determined that TVA’s reliance on the information in Appendix A to BTP (APCSB) 9.5-1, was 

acceptable since it was the guidance in place on April 18, 1977, when TVA first submitted a fire 

hazards analysis to the NRC for review.

In the FPR, TVA additionally stated that the applicable guidelines used as the basis for the plan 

included, in part, the following: 

• NRC letter entitled, “Nuclear Plant Fire Protection Functional Responsibilities, 

Administrative Controls and Quality Assurance,” dated June 20, 1977

• Generic Letter (GL) 81-12, “Fire Protection Rule (45 FR 76602, November 19, 1980),” 

dated February 20, 1981, and its associated clarification letter, dated March 22, 1982; 

• GL 82-21, “Technical Specifications for Fire Protection Audits,” dated October 6, 1982; 

• GL 83-33, “NRC Positions on Certain Requirements of Appendix R to 10 CFR 50,” 

dated October 19 1983; 

• GL 86-10, “Implementation of Fire Protection Requirements,” dated April 24, 1986; 

• GL 88-12, “Removal of Fire Protection Requirements from Technical Specifications,” 

dated August 2, 1988.

The following NRC guidance was used for specific topics:

• NUREG-1852, “Demonstrating the Feasibility and Reliability of Operator Manual Actions 

in Response to Fire,” issued October 2007, for WBN Unit 2 OMA evaluations
FF-4

• NRC Regulatory Guide (RG) 1.189, “Fire Protection for Operating Nuclear Power 

Plants,” Revision 0, issued April 2001, for extension of the “annual” fire protection audit 

interval

• NRC RG 1.189, “Fire Protection for Nuclear Power Plants,” Revision 2, issued 

October 2009, for OMA and multiple spurious operation (MSO) evaluations.

Accordingly, the NRC staff reviewed the entire fire protection program (except as noted 

otherwise) using the agency’s fire protection requirements and review guidance. Because 

WBN consists of two units of identical design, this evaluation applies to the fire protection 

program for both WBN, Unit 1, and WBN, Unit 2 (except as noted otherwise).

The NRC staff’s review did not include Part VII, Section 7, “Unit 1 Operator Manual Actions 

[OMAs],” of the FPR. The NRC’s approval of the WBN, Unit 1, OMAs is documented in 

Supplemental Safety Evaluation Report (SSER) 18, NUREG-0847, “Safety Evaluation Report 

Related to the Operation of Watts Bar Nuclear Plant, Units 1 and 2,” dated October 1995.

In Staff Requirements Memorandum SECY-07-0096, “Possible Reactivation of Construction 

and Licensing Activities for the Watts Bar Nuclear Plant Unit 2,” dated July 25, 2007, the 

Commission directed the NRC staff to use the existing WBN, Unit 1, licensing basis as the 

reference basis for the WBN, Unit 2, review. To that end, where applicable, the NRC staff used 

the WBN, Unit 1, approvals, as documented in SSER 18, issued October 1995, and SSER 19, 

issued November 1995, to NUREG-0847, as the basis for its approvals in this evaluation, 

instead of the agency’s current guidance. The NRC staff used the agency’s current guidance 

as the basis for approval for the WBN, Unit 2, OMAs, associated circuits, MSO, fire water 

system design demand, the auxiliary control room, and radiant energy shields.

The NRC staff met with TVA on January 19, February 3 and 15, March 29, April 22, May 12, 

June 30, July 12 and 28, August 31, November 16, and December 21, 2011, February 2, 2012, 

January 28, June 26, July 16 and 31, August 28, October 8 and 29, and November 12, 2014,

and February 11, and April 1, 2015, to discuss technical issues related to WBN’s fire protection 

program and its implementation. 

The NRC staff conducted an audit at WBN from October 25-27, 2011, which it documented in 

a report dated December 20, 2011 (Agencywide Documents Access and Management System 

(ADAMS) Accession No. ML113500239). The NRC staff conducted an additional audit at WBN 

on July 21 and 22, 2015, which it documented in a report dated August 27, 2015 (ADAMS 

Accession No. ML15239A794).

Unless otherwise noted, all information cited in this evaluation is from the WBN FPR dated 

June 24, 2015, as supplemented (ADAMS Accession Nos. ML15175A508, ML15239A794, and 

ML15280A508).

This safety evaluation is based on the as-constructed FPR, and it replaces, in its entirety, the 

dual unit fire protection safety evaluation published in SSER 26, which was based on the asdesigned FPR.
FF-5

2.0 FIRE PROTECTION PROGRAM

2.1 Purpose and Scope

The purpose of the NRC staff’s review of is to determine of TVA meets the Commission’s 

controlling regulations on fire protection in 10 CFR 50.48. The regulation at 10 CFR 50.48(a)(1) 

requires the following:

Each holder of an operating license must have a fire protection plan 

that satisfies Criterion 3 of appendix A [to 10 CFR 50.] This fire protection 

plan must:

(i) Describe the overall fire protection program for the facility;

(ii) Identify the various positions within the licensee’s 

organization that are responsible for the program;

(iii) State the authorities that are delegated to each of these 

positions to implement those responsibilities; and

(iv) Outline the plans for fire protection, fire detection and 

suppression capability, and limitation of fire damage.

The regulation at 10 CFR 50.48(a)(2) requires:

The plan must also describe specific features necessary to implement the 

program described in [10 CFR 50.48(a)(1)] such as --

(i) Administrative controls and personnel requirements for fire 

prevention and manual fire suppression activities,

(ii) Automatic and manually operated fire detection and 

suppression systems; and

(iii) The means to limit fire damage to structures, systems, or 

components [(SSCs)] important to safety so that the 

capability to shut down the plant safely is ensured.

Last, the regulation at 10 CFR 50.48(a)(3) requires: 

The licensee shall retain the fire protection plan and each change to the 

plan as a record until the Commission terminates the reactor license. The 

licensee shall retain each superseded revision of the procedures for 3 

years from the date it was superseded.
FF-6

2.1.1 Summary of TVA’s 10 CFR 50.48(a)(1) conformance

In Fire Protection Report (FPR) Part I, Section 2.0, “Purpose,” Tennessee Valley Authority 

(TVA) stated that the purpose of the FPR is to provide a detailed summary of the Watts Bar 

Nuclear Plant (WBN) fire protection program in a single document. The FPR is thus the “fire 

protection plan” document that is required by Title 10 of the Code of Federal Regulations

(10 CFR) Section 50.48(a). Section 9.5.1 of the WBN Final Safety Analysis Report (FSAR) 

incorporates the FPR by reference. In FPR Part I, TVA states that it will be updated in 

conjunction with the FSAR.

TVA’s plan provided information on 10 CFR 50.48(a)(1)(i) in FPR Part II, Section 9, “Emergency 

Response,” Section 10, “Control of Combustibles,” and Section 11, “Control of Ignition Sources.” 

TVA’s plan provided information on 10 CFR 50.48(a)(1)(ii) in FPR Part II, Section 7, “Fire 

Protection Organization/Programs,” and Section 14, “Fire Protection Systems and Features 

Operating Requirements,” and in FPR Part VI. TVA’s plan provided information on 10 CFR 

50.48(a)(1)(iii) in FPR Part II, Section 7, “Fire Protection Organization/Programs,” and in FPR 

Parts III, IV, V, and VI. TVA’s plan provided information on 10 CFR 50.48(a)(1)(iv) above in 

FPR Part II, Section 12, “Description of Fire Protection Systems and Features.” 

The NRC staff’s evaluations of compliance with 10 CFR 50.48(a)(1)(i), (a)(1)(ii), and (a)(1)(iii) 

are evaluated in Section 2 of this evaluation. The NRC staff’s evaluation of how the plan meets 

10 CFR 50.48(a)(1)(iv) is in Sections 2 through 5 of this evaluation.

2.1.2 Summary of TVA’s 10 CFR 50.48(a)(2) conformance

TVA’s plan provided information on 10 CFR 50.48(a)(2)(i) in FPR Part II, Section 9, “Emergency 

Response,” Section 10, “Control of Combustibles,” Section 11, “Control of Ignition Sources,” and 

Section 13, “Fire Protection System Impairments.” TVA’s plan provided information on 

10 CFR 50.48(a)(2)(ii) in FPR Part II, Section 12, “Description of Fire Protection Systems and 

Features.” TVA’s plan provided information on 10 CFR 50.48(a)(2)(iii) in FPR Part II, 

Section 12, “Description of Fire Protection Systems and Features,” and in FPR Parts III, IV, V, 

and VI. 

The NRC staff assesses how TVA meets 10 CFR 50.48(a)(2)(i) in Section 2 of this evaluation. 

The NRC staff considers how TVA meets 10 CFR 50.48(a)(2)(ii) in Section 4 of this evaluation. 

Finally, the NRC staff confirms how TVA meets 10 CFR 50.48(a)(2)(iii) in Sections 3 and 5 of 

this evaluation.

2.1.3 Summary of TVA’s 10 CFR 50.48(a)(3) conformance

The first part of the regulation at 10 CFR 50.48(a)(3) requires the licensee to retain the fire 

protection plan and each change to the plan as a record until the reactor license is terminated. 

In FPR Part I, Section 2 “Purpose,” TVA stated that the FPR will be updated in conjunction with 

updates to the WBN FSAR. The U.S. Nuclear Regulatory Commission (NRC) staff concludes 

that this is an acceptable method of retaining plan records, because the FSAR is maintained 

and updated in accordance with 10 CFR 50.59, “Changes, Tests, and Experiments,” and 

10 CFR 50.71(e), respectively, which have similar retention requirements and therefore meets 

the requirements of 10 CFR 50.48(a)(3).
FF-7

The second part of the regulation at 10 CFR 50.48(a)(3) requires the licensee to retain revisions 

to procedures for three years from the date they are supperceded. In FPR Part II, Section 6 

“Fire Protection Quality Assurance,” TVA described that quality assurance for the fire protection 

program provided as part of the overall WBN Quality Assurance program. In its letter dated 

October 7, 2015 (ADAMS Accession No. ML15280A508), TVA confirmed that fire protection 

program related records and procedures are retained in accordance with the requirements of 10 

CFR 50.48(a)(3). The NRC staff concludes that this meets the requirements of 10 CFR 

50.48(a)(3).

2.2 Fire Protection Organization

As described in FPR Part II, Section 7, TVA’s fire protection organization consists of corporate 

management oversight and an onsite plant implementation organization. Responsible TVA 

corporate managers include the Chief Operating Officer and Chief Nuclear Officer, the Vice 

President Nuclear Engineering, the Senior Vice President, Operations, and the Corporate 

Program Manager. The onsite implementation organization includes the Site Vice President, 

the Plant Manager, the Director, Operations, the Fire Operations Supervisor, the Fire Marshal, 

and the Site Engineering Director. 

The NRC staff reviewed the responsibilities and authorities of each position responsible for the 

fire protection program, as described in FPR Part II, Sections 7.1 through 7.2.4, and concluded 

that there is reasonable assurance that the key responsibilities for implementing the fire 

protection program at WBN have been delegated to appropriate positions within TVA’s 

organization, and that the authorities delegated to each position to implement these 

responsibilities are appropriate.

Based on its review of the FPR, the NRC staff concludes that TVA’s fire protection organization 

does not take any exceptions to Position A.1 of Appendix A to BTP (APCSB) 9.5-1, and 

therefore, is acceptable.

2.3 Fire Protection Quality Assurance Program

FPR Part II, Section 6.0, contains TVA’s description of the quality assurance (QA) program for 

fire protection at WBN. TVA stated that it used the guidance established by Appendix A to BTP 

(APCSB) 9.5-1 and the NRC’s letter dated June 20, 1977, “Nuclear Plant Fire Protection 

Functional Responsibilities, Administrative Controls, and Quality Assurance,” to develop a QA 

program for fire protection features that protects post-FSSD capability and safety-related SSCs. 

The FPR states that the WBN fire protection QA program uses the applicable parts of 

TVA-NQA-PLN89-A, “Tennessee Valley Authority Nuclear Quality Assurance Plan.”

TVA implemented a program that performs independent audits and inspections of the WBN fire 

protection program. TVA stated that its program is based on the guidance in Generic Letter 

(GL) 82-21. The FPR states that TVA’s Nuclear Assurance organization is responsible for 

conducting the fire protection-related audits.

In TVA’s letter dated May 6, 2011 (Agencywide Documents Access and Management System 

(ADAMS) Accession No. ML11129A158), in response to the NRC staff’s request for additional 

information (RAI) FPR II-26, TVA stated that the frequency of the GL 82-21 annual fire 

protection audit has been changed to 24 months. TVA stated in its letter dated August 28, 2002 

(ADAMS Accession No. ML022460173), that the plant implemented this change using a 
FF-8

performance-based schedule. In TVA’s letter dated September 30, 2011 (ADAMS Accession 

No. ML13060A225), in response to the NRC staff’s RAI FPR II-26.1, TVA stated that the change 

is being monitored on a fleetwide basis, and that deficiencies found during the biennial audits 

would result in increasing the frequency of the audits. The NRC staff concludes that this is 

consistent with Position 1.7.10.1 of Revision 0 to RG 1.189, and, therefore, is acceptable.

Based on its review of the information submitted by TVA, the NRC staff concludes that TVA’s 

fire protection QA program does not take any exceptions to Position C of Appendix A to BTP 

(APCSB) 9.5-1 and, therefore, is acceptable.

2.4 Fire Protection Administrative and Technical Controls

2.4.1 Fire Protection Program Changes, Review and Approval

TVA stated in FPR Part I, Section 2.0, that “the Fire Protection Report has been developed in 

accordance with the guidelines of NRC Generic Letter 86-10...and NRC Generic Letter 

88-12....” TVA elected to follow the guidance in GL 88-12 and incorporate the standard fire 

protection license condition as listed in GL 86-10. In addition to including, by reference, the 

NRC safety evaluations that approved the plant fire protection program, this license condition 

allows TVA to make changes to the approved program without prior approval of the NRC if 

those changes would not adversely affect the plant’s ability to achieve and maintain safe 

shutdown in the event of a fire.

Based on its review of the information submitted by TVA, the NRC staff concludes that no 

exceptions were taken to the positions in GL 88-12, and it is therefore, acceptable.

2.4.2 Fire Protection Administrative Controls

2.4.2.1 Control of Combustibles 

FPR Part II, Section 10.0, describes TVA’s program to control combustibles. The WBN 

combustible control program objectives are to (1) provide instruction and guidelines during 

general employee training on the application and use of combustible materials at WBN, 

(2) control the application and use of chemicals, (3) perform periodic plant housekeeping 

inspections and have housekeeping tours by management and the onsite fire protection 

organization, (4) control in situ combustibles through the design/modification review and 

installation process, and (5) control transient combustibles through the implementation of 

administrative controls.

TVA stated that it established a plant-wide administrative procedure to control transient 

combustibles. Implementation of this procedure will establish administrative controls for the 

handling of combustible materials such as fire-retardant wood, paper, plastic, and flammable 

and combustible gases and liquids. In addition, through its combustible control program, TVA 

established combustible control zones in the plant. TVA considers these zones to be 

subdivisions of fire areas and to limit fire spread by providing open space free of transient 

combustibles between redundant FSSD equipment or cables. Transient combustibles may not 

be stored in these zones unless an adequate fire protection engineering evaluation or 

compensatory measures, or both, are implemented.
FF-9

Based on its review of the information submitted by TVA, the NRC staff concludes that TVA’s 

program to control combustibles does not take any exceptions to Positions B.2 and B.3.c of 

Appendix A to BTP (APCSB) 9.5-1 and, therefore, is acceptable.

2.4.2.2 Control of Ignition Sources

TVA established a program for controlling ignition sources such as welding, cutting, grinding, 

and the use of open flame. TVA’s program specifies that the issuance of “hot work” permits be 

reviewed and approved based on plant conditions and a prior inspection of the proposed work 

area. The ignition source on a hot work permit is valid for only one job. Before the start of work, 

the work area is made “fire safe.” In addition, TVA’s program establishes a hot work fire watch 

for all ignition source work activities that are performed in safety-related and safe shutdown 

areas of the plant, with the exception of the specific ignition source activities of underwater 

welding, work in outside areas (on fences, light poles, etc.), and electric soldering. These fire 

watches, in addition to performing their duties during the hot work activities, will remain in the 

area for a minimum of 30 minutes after the work has been completed to ensure that potential 

residual ignition conditions do not exist.

Based on its review of the information submitted by TVA, the NRC staff concludes that TVA’s 

program to control ignition sources does not take any exceptions to Positions B.3.a and B.3.b of 

Appendix A to BTP (APCSB) 9.5-1 and, therefore, is acceptable.

2.4.3 Compensatory Measures

Compensatory measures described in FPR Part II are used to compensate for degraded or 

nonfunctional fire protection systems or features. Primarily, these compensatory measures 

take the form of both roving and continuous fire watches.

FPR Part II, Section 13.1.B states, “A roving fire watch consists of a trained individual in an 

affected location at 60 minute intervals with a 15 minute margin to accommodate and handle 

unforeseen circumstances and to report and/or resolve potential fire hazards in a location. 

Roving fire watches are required as a compensatory action in all modes of plant operation (i.e., 

Modes 1 through 6 or core empty).” The NRC staff concludes that this takes no exceptions to 

Positions B.3 and B.5.a of Appendix A to BTP (APCSB) 9.5-1 and, therefore, is acceptable.

As described in FPR Part II, Section 13.1.A, a continuous fire watch possesses the following 

attributes: (1) the trained person performing the fire watch must be in the fire area at all times; 

(2) the fire area must not contain any impediment to restrict the movements of the fire watch; 

and (3) each compartment within the fire area must be patrolled at least once every 15 minutes 

with a margin of 5 minutes. The NRC staff concludes that this takes no exceptions to 

Positions B.3 and B.5.a of Appendix A to BTP (APCSB) 9.5-1 and, therefore, is acceptable.

In the FPR, TVA identified specific exceptions to the above guidance for roving and continuous 

fire watches. In FPR Section 13.1.A, TVA identified continuous fire watch routes in more than 

one fire area that it classifies as exceptions to a continuous fire watch remaining within one fire 

area. As a basis for acceptability, TVA identified the following characteristics: (1) one or more 

rooms in different fire areas whose proximity to one another and their limited size warrant the 

combining of them into one continuous fire watch route, (2) a time study that confirms the route 

can be covered in 15 minutes without putting undue exertion on the person performing the fire 

watch, and (3) in each instance, these routes require the Fire Protection Supervisor’s approval 
FF-10

to ensure that the conditions that formed a basis for the time study have not changed in such a 

manner as to invalidate the time study. In the event that the automatic suppression or detection 

systems in the above areas cannot be restored within the time specified by FPR Part II, 

Section 14.0, TVA stated that the continuous fire watch patrols would not be allowed to include 

more than one fire area. Based on the information submitted by TVA, the NRC staff concludes 

that this takes no exceptions to Positions B.3 and B.5.a of Appendix A to BTP (APCSB) 9.5-1 

and, therefore, is acceptable.

The WBN FPR states that continuous fire watches are only required when the affected unit is in 

Mode 1 (power operation) through Mode 4 (hot shutdown). In FPR Part II, TVA stated that, 

when one unit is in Modes 5, 6, or core empty, locations where a continuous fire watch would 

be required may be combined and patrolled by a roving fire watch when approved by the Fire 

Protection Supervisor, if a fire in those locations could not affect the other unit, if it is in Modes 1 

through 4. Based on the information submitted by TVA, the NRC staff concludes that this takes 

no exceptions to Positions B.3 and B.5.a of Appendix A to BTP (APCSB) 9.5-1 and, therefore, 

is acceptable.

In FPR Part VII, Section 6.1, TVA summarized its evaluation of not performing certain 

compensatory measures for out of service fire protection features (for example, smoke 

detectors, sprinklers) for the reactor building equipment hatches when these rooms are 

inaccessible high radiation areas, that is, while the associated unit is operating. Section 6.3.1.2 

of this evaluation provides the NRC staff evaluation and approval of this configuration.

In addition, in the FPR Part II, Section 13.1, TVA identified other alternative compensatory 

measures that may be used at WBN in lieu of the above standard compensatory measures. In 

all cases, in which an alternative compensatory measure is used for a degraded or 

nonfunctional fire protection feature, TVA stated that it will perform an evaluation that 

demonstrates technical equivalency to the standard compensatory measure identified in FPR 

Part II, Section 14.0. TVA described the following alternatives that may be considered when 

supported by an appropriate technical evaluation: (1) providing additional or alternative fire 

protection equipment, (2) installing temporary or portable fire detection systems in conjunction 

with an hourly roving fire watch, (3) installing closed circuit television cameras and monitors in 

areas when special circumstances, such as personal safety or as-low-as reasonably-achievable 

(ALARA; radiological) concerns, preclude the use of a human fire watch in the area, and 

(4) taking credit in continuously manned areas for the constant manning in lieu of establishing 

either continuous or roving compensatory fire watches when the responsible individuals accept 

this responsibility. Based on its review of the information submitted by TVA, the NRC staff 

concludes that these alternatives take no exceptions to Positions B.3 and B.5.a of Appendix A 

to BTP (APCSB) 9.5-1 and, therefore, are acceptable.

2.4.4 Fire Protection Technical Controls

In FPR Part II, Section 14, TVA established functionality requirements for the following fire 

protection features: (1) fire detection instrumentation, (2) water supply, (3) water-based fire 

suppression systems, (4) carbon dioxide (CO2) suppression systems, (5) fire detection 

supervisory equipment, (6) fire hose stations and associated preaction control valves, (7) fire 

hydrants, (8) fire-rated assemblies, and (9) emergency battery lighting units.

Based on its review of the information submitted by TVA, the NRC staff concludes that TVA’s 

functionality requirement program for plant fire protection features does not take any exceptions 
FF-11

to Positions B.1, B.3, and B.5 of Appendix A to BTP (APCSB) 9.5-1, and, therefore, is 

acceptable.

GL 88-12 provides guidance for removing fire protection limiting conditions for operation and 

surveillance frequencies associated with fire detection systems, fire suppression systems, fire 

barriers, and administrative controls that address fire brigade NRC staffing from the plant’s 

technical specifications (TSs) and incorporating this information into the FSAR. In addition, 

GL 88-12 refers to GL 81-12, which requested licensees to provide TSs for equipment used for 

safe shutdown capability that is not currently covered by existing TSs. In its fire protection plan, 

TVA confirmed that the plant equipment used to achieve and maintain post-FSSD from either 

inside or outside the main control room (MCR) is included in either the plant TSs or the FPR.

Table 14.10, “Fire Safe Shutdown Equipment,” of FPR Part II, Section 14, lists the FSSD 

equipment not included in the plant’s TSs. TVA established testing and inspection requirements 

which assist in evaluating the functionality of the non-TS-related FSSD equipment and 

instrumentation. In FPR Part II, Section 14.0, TVA established the requirements with this 

equipment or instrumentation nonfunctional. TVA requires, with one or more of the required 

items of equipment listed in Table 14.10 nonfunctional (or a breaker or valve not in its safe 

shutdown position), that the plant restore the equipment to the functional status within 30 days, 

or that it either (1) place the equipment in the condition required for FSSD, (2) provide a backup 

means of instrumentation monitoring, (3) provide an alternative means of achieving post-FSSD 

(along with an evaluation justifying the alternative), or (4) be in Mode 3 within 6 hours and 

Mode 4 within the following 12 hours.

Also in FPR Part II, Section 14.0, TVA established the requirements for actions when specified 

pressurizer power-operated relief valve (PORV) block valves are in a closed position (with 

power maintained) with the plant in Modes 1, 2, or 3. Control of the pressurizer block valve is 

maintained from the MCR so that it can be opened by the MCR operator if needed. The block 

valve will be returned to the open position as soon as possible, but no later than the end of the 

next refueling outage, in accordance with WBN Technical Specification 3.4.11 for a “simmering” 

PORV. Fire watches are established in specific auxiliary building rooms where the specific 

block valve cable is in the train credited for reactor coolant system pressure control, so that a 

fire in these areas can be detected in the early stages of growth and the block valve opened if 

needed.

Based on the information provided in FPR Part II, the NRC staff concludes that TVA’s 

removal of fire protection features from the plant’s TSs and relocation to the FPR as operating 

requirements is consistent with the guidance in GL 88-12 and GL 81-12, and, therefore, is 

acceptable.

In addition, in FPR Part II, Section 14, TVA established testing and inspection requirements 

for the following fire protection features: (1) fire detection instrumentation, (2) water supply, 

(3) water-based fire suppression systems, (4) CO2 suppression systems, (5) fire detection 

supervisory equipment, (6) fire hose stations and associated preaction control valves, (7) fire 

hydrants, (8) fire-rated assemblies, (9) emergency battery lighting units, and (10) the FSSD 

equipment identified in Table 14.10.

Based on its review of the information submitted by TVA, the NRC staff concludes that TVA’s 

surveillance and test program for plant fire protection features does not take any exceptions to 

Position B.5 of Appendix A to BTP (APCSB) 9.5-1, and, therefore, is acceptable.
FF-12

2.5 Fire Brigade and Fire Response

2.5.1 Organization

FPR Part II, Section 9.1, “Fire Brigade NRC Staffing,” states that a fire brigade composed of at 

least five members will be maintained on site at all times. In the FPR, TVA stated that these 

five members will consist of the fire brigade leader and four fire brigade members. In addition, 

neither the shift operations supervisor nor the other members of the operations shift crew 

needed to perform a safe shutdown of the WBN units will be included in the fire brigade. 

TVA also stated that in addition to the five members of the fire brigade, an incident commander 

is available to direct each shift fire brigade. The incident commander has sufficient knowledge 

of plant safety systems to understand the effects of fire and fire suppression on safe shutdown 

capability. The incident commander is not the on-duty shift manager, unit supervisor, or shift 

technical advisor.

TVA stated that before initial training and annually thereafter its fire brigade program requires 

each fire brigade member to undergo a medical review and to receive medical approval to 

perform strenuous physical activities related to firefighting and to wear special respiratory 

equipment.

TVA stated that the fire brigade may comprise fewer than five members for a period of time not 

to exceed 2 hours, to accommodate unexpected conditions such as an unplanned absence or 

brigade response to a nonfire emergency.

Based on its review of the information submitted by TVA, the NRC staff concludes that TVA’s 

fire brigade staffing and organization does not take any exceptions to Positions B.4 or B.5 of 

Appendix A to BTP (APCSB) 9.5-1 and, therefore, is acceptable.

2.5.2 Training

FPR Part II, Section 9.3 “Training and Qualifications,” states that TVA’s fire brigade training 

program consists of initial training, recurrent training, and annual fire brigade training.

The initial training program includes (1) instruction and practical exercises in fire extinguishment 

and the use of firefighting equipment, (2) identification of fire hazards and types of fires that 

could occur in the plant, (3) identification of the location of firefighting equipment in each fire 

area of the plant, (4) instruction on the proper use of plant firefighting equipment, (5) instruction 

on the proper use of communications, lighting, ventilation, and emergency breathing apparatus, 

(6) instruction on the toxic characteristics of the products of combustion, and (7) instruction and 

practical exercises in fighting fires inside buildings and tunnels. In addition to initial training, the 

program instructs the fire brigade on firefighting procedures and procedure changes, the plant 

firefighting plan, with emphasis on each individual’s responsibility, and the latest plant 

modifications and changes affecting the firefighting plans.

The recurrent training consists of classroom instruction meetings held every 3 months. These 

meetings repeat the initial training subjects over a 2 year period. Each member of the fire 

brigade is required to attend this training in order to remain qualified. TVA preplans fire brigade 

drills to establish the objectives, and the fire brigade training instructor or the instructor’s 
FF-13

designee conducts these drills. The onsite fire brigade drills are conducted as follows: (1) a 

minimum of one drill per fire brigade shift will be conducted every 92 days, (2) a minimum of 

one unannounced drill will be conducted per fire brigade shift per year, and (3) at least one drill 

per fire brigade shift will be conducted on the backshift. Each fire brigade member is required 

to attend at least two drills per year.

TVA holds annual training for each fire brigade member. TVA stated that this training provides 

instruction, under actual firefighting conditions, on the proper methods for fighting various types 

of fires similar in magnitude, complexity, and difficulty to those that could be encountered in the 

plant. This training includes actual fire extinguishment and the use of firefighting equipment 

under strenuous conditions. TVA stated that if a brigade member misses or does not complete 

a training session, either annual or quarterly; the member is placed in an ineligible status, when 

their current training expires, until the training is completed.

In addition to the annual fire brigade training, TVA holds annual briefings for the local fire 

departments to ensure their continued understanding of their role in the event of a fire 

emergency at the site. TVA also holds an annual drill for the local fire department and the plant 

fire brigade. The local fire department briefings and drills are held for those departments that 

have active aid agreements with the plant.

Based on its review of the information submitted by TVA, the NRC staff concludes that TVA’s 

fire brigade training program does not take any exceptions to Positions B.5.b and B.5.c of 

Appendix A to BTP (APCSB) 9.5-1 and, therefore, is acceptable.

2.5.3 Equipment

In the FPR, TVA stated that firefighting equipment is provided throughout the plant and is 

strategically placed near the fire hazards present or anticipated. TVA stated that delays in the 

fire brigade obtaining firefighting equipment are minimized because of the distribution and 

availability of this equipment throughout the plant. TVA further stated that firefighting 

equipment may be staged adjacent to, or at the access to, areas/locations to facilitate 

equipment availability or to address equipment surveillance test concerns relative to life safety 

and ALARA practices. 

The equipment available to the fire brigade includes: (1) motorized firefighting apparatus, 

(2) portable ventilation equipment, (3) fire extinguishers, (4) self-contained breathing apparatus, 

(5) fire hose, nozzles, and fittings, (6) foam equipment, (7) personal protective equipment, 

(8) communications equipment, (9) portable lighting, and (10) ladders specifically dedicated for 

firefighting.

Based on its review of the information submitted by TVA, the NRC staff concludes that no 

exceptions were taken to Position B.5.d of Appendix A to BTP (APCSB) 9.5-1 and therefore, 

TVA’s fire brigade is acceptably equipped. 

2.5.4 Fire Emergency Procedures and Prefire Plans

As described in the FPR, TVA’s fire emergency procedures and prefire plans specify the actions 

that the individual who discovers a fire are to take and the actions that the emergency response 

organization are to consider (e.g., control room operators and the plant fire brigade). These 

procedures provide different levels of response based on whether actual fire/smoke conditions 
FF-14

are reported or whether a fire detection system annunciation occurs. (For example, a single fire 

detection system zone annunciation in a cross-zoned area will not carry the same level of 

response as a cross-zone annunciation in the same area.)

TVA stated that it implemented prefire plans to provide guidance, depending on the particular 

circumstances, to aid in firefighting efforts. TVA developed prefire plans to support the 

firefighting activities in plant areas important to safety. Specifically, these plans are developed 

for safety-related areas, FSSD areas, and areas that present a hazard to safety-related 

equipment or plant shutdown.

The prefire plans provide the following information to the fire brigade: (1) plant equipment in the 

fire area, (2) access and egress routes to the fire area, (3) firefighting strategy and tactics, 

(4) locations of fire protection features and equipment, (5) special fire, toxic, and radiological 

hazards in the area, (6) special precautions, and (7) ventilation methodology.

Based on its review of the information submitted by TVA, the NRC staff concludes that TVA’s 

proposed fire brigade preplans and fire emergency procedures conform to the guidance in the 

NRC letter dated June 20, 1977 (the contents of the June 20, 1977 NRC letter can be found 

attached to NRC letter dated August 4, 1977, ADAMS Accession No. ML031280293) and the 

requirements of 10 CFR 50.48(a)(2) and, therefore, are acceptable.
FF-15

3.0 GENERAL PLANT FIRE PROTECTION AND SAFE SHUTDOWN 

FEATURES

The NRC staff evaluated TVA’s fire protection program for compliance with 10 CFR 50.48(a), as 

described in Section 2.1 of this evaluation. TVA evaluated their program against the guidance 

in Appendix A to Branch Technical Position (BTP) (Auxiliary Power Conversion Systems Branch 

(APCSB)) 9.5-1. The NRC staff determined that TVA’s reliance on the information in Appendix 

A to BTP (APCSB) 9.5-1, was acceptable since it was the guidance in place on April 18, 1977, 

when TVA first submitted a fire hazards analysis to the NRC for review. In addition to TVA’s 

evaluation to Appendix A to BTP (APCSB) 9.5-1, since TVA used guidance published prior to 

1981, TVA also committed to meet Sections III.G, III.J, and III.O, of 10 CFR Part 50, Appendix 

R, to assure safe shutdown capability. This section of the evaluation describes the NRC staff’s 

evaluation of TVA’s compliance with portions of 10 CFR 50.48(a)(1)(iv) and 10 CFR 

50.48(a)(2)(iii). As discussed below, the NRC staff’s review found that the Watts Bar fire 

protection program meets the requirements of 10 CFR 50.48(a).

3.1 Fire Protection Design

3.1.1 Building and Compartment Fire Barriers

Tennessee Valley Authority (TVA) stated that the fire-rated assemblies at Watts Bar Nuclear 

Plant (WBN) are part of the passive fire protection features that ensure that one set of 

redundant fire safe shutdown (FSSD) components necessary to achieve and maintain FSSD 

remains free of fire damage. At WBN, fire-rated assemblies consist of fire barriers, raceway 

protection, fire doors, fire dampers, and penetration seals.

At WBN, fire areas are defined by rated wall and floor/ceiling assemblies. TVA stated that fire 

areas are separated by wall and floor/ceiling assemblies that are 2- or 3-hour equivalent fire 

barriers that are bounded by Underwriters Laboratories, Inc., (UL)-rated designs. In Fire 

Protection Report (FPR) Part II, Sections 12.10 and 12.10.1, TVA states that the walls that 

separate buildings and walls between rooms that contain safe shutdown systems are fire-rated 

assemblies. Rooms within each fire area may be separated from other rooms in the same fire 

area by regulatory or nonregulatory fire barriers. Where barriers are needed between rooms, 

TVA stated that only fire-rated barriers with a minimum 2-hour rating are relied upon, except for 

portions of the main control room (MCR) complex that have 1-hour rated barriers. 

Sections 6.2.3, 6.2.5, 6.2.6, and 6.2.7 of this evaluation provide U.S. Nuclear Regulatory 

Commission (NRC) staff evaluations of exceptions to fire barrier ratings.

In general, the fire barriers comprising compartment walls and floors/ceilings at WBN are 

constructed of reinforced concrete or concrete block. The reinforced concrete fire barriers and 

concrete block barriers are at least 8 inches thick. TVA’s evaluation of reinforced concrete 

barriers used information from Section 6, Chapter 5, of the National Fire Protection Association 

(NFPA) Fire Protection Handbook, 17th Edition (hereafter referred to as the Handbook). This 

section of the Handbook correlates fire rating and thickness of reinforced concrete. On this 

basis, Figure 6-5G in the Handbook shows that 6 inches of reinforced concrete has a fire 

resistance of approximately 4 hours. The concrete block barriers are only used when barriers 

are required to have a fire rating of 2 hours or less. TVA’s evaluation of these fire barrier 

designs concludes these are similar to UL-listed concrete block barrier designs (Design 

Nos. U904, U905, U906, and U907), which are 2- to 4-hour fire-rated.
FF-16

Based on its review of the information submitted by TVA, the NRC staff concludes that TVA’s 

proposed technical basis for the fire resistive capability of fire area boundaries offers an 

equivalent level of fire safety to that of Position D.1.j of Appendix A to BTP (APCSB) 9.5-1 and, 

therefore, is acceptable.

3.1.2 Fire Barriers Used To Separate Redundant Safe Shutdown Functions within the 

Same Fire Area 

Cable raceways that require separation due to redundant trains located in the same fire area, 

excluding primary containment and secondary containment (the annulus), are separated by 

either 1- or 3-hour fire-rated barrier systems. TVA uses a 1-hour fire-rated barrier system if 

automatic detection and automatic suppression are installed in the areas and uses a 3-hour 

fire-rated barrier system if automatic suppression is not installed in the area. Cable raceways 

that require separation due to redundant trains inside the reactor building, which includes 

primary containment (WBN Unit 1 only) and secondary containment (i.e., the annulus) (both 

units), rely on radiant energy shields (RESs) or automatic detection and suppression to provide 

separation. RESs are addressed in Section 6.1.2 of this evaluation.

In FPR Part II, Section 12.10.2, TVA stated that the 1- and 3-hour fire-rated barriers were tested

in accordance with the guidance in Supplement 1, “Fire Endurance Test Acceptance Criteria for 

Fire Barrier Systems Used to Separate Redundant Safe Shutdown Trains within the Same Fire 

Area,” to Generic Letter (GL) 86-10. This guidance includes test parameters, thermocouple 

placement, conduit and cable tray configurations, hose stream tests, and ampacity derating. 

TVA also evaluated fire barriers for seismic considerations. Configurations of raceway fire 

barriers that are not consistent with the testing have been evaluated to ensure that untested 

configurations are bounded by tested configurations. TVA has procedural controls for 

evaluating field changes to designed configurations. TVA stated that personnel who perform 

such field changes are to be cognizant of the important parameters.

Based on its review of the information submitted by TVA, the NRC staff concludes that TVA’s 

use of the guidance in Supplement 1 to GL 86-10, with the consideration of seismic events, 

bounding of untested configurations, and procedures to control field changes, offers an 

equivalent level of fire safety to that of Position D.3 of Appendix A to BTP (APCSB) 9.5-1 

and, therefore, is acceptable.

3.1.3 Equipment Hatches and Stairwells

TVA stated that at WBN equipment hatches in the floor or fire barriers in the ceiling can be 

categorized as follows:

• precast concrete plugs

• steel covers 

• open hatches and stairwells

TVA stated that the precast concrete plugs are associated with radiation shielding and, as fire 

barriers, are equivalent to the floor or ceiling fire barrier in which they are located. TVA stated 
FF-17

that the steel covers are of substantial construction and that they provide an effective barrier to 

prevent fire from propagating from one side of the barrier to the other. In addition, because the 

covers are not fire rated, they are either provided with a draft stop and water curtain around 

them or redundant safe shutdown components on either side have been separated from each 

other by a cumulative horizontal distance of 20 feet or more. In either case, automatic fire 

suppression and detection are provided on both sides of the equipment hatch cover. 

FPR Part VII, Section 2.6.4, summarizes TVA’s evaluation of the nonrated equipment hatches 

separating the control building and turbine building. Section 6.2.7.4 of this evaluation provides 

the NRC staff evaluation and approval of this configuration.

TVA stated that, in areas in which open hatches and stairwells are located, redundant shutdown 

trains are either separated by at least 20 feet horizontally, one train has been protected by a 

1-hour fire barrier, or a water curtain has been installed around the opening. In any case, fire 

detection and automatic suppression systems are located on both sides of the openings. 

Further, TVA stated that the only exceptions to this arrangement are on elevation 676 feet and 

in the refueling area of the auxiliary building.

FPR Part VII, Section 2.6.3, summarizes TVA’s evaluation of the nonrated open hatches and 

stairwells that do not fully meet the NRC staff guidance. Section 6.2.7.3 of this evaluation 

provides the NRC staff evaluation and approval of this configuration.

FPR Part VII, Section 3.1, summarizes TVA’s evaluation of the lack of area wide fire detection 

and automatic suppression in fire areas containing redundant FSSD equipment and cables. 

Section 6.1.7 of this evaluation provides the NRC staff evaluation and approval of this 

configuration.

FPR Part VII, Section 4.5, summarizes TVA’s evaluation of the lack of fire detection in the 

refueling area. Section 6.2.1 of this evaluation provides the NRC staff evaluation and approval 

of this configuration. 

Based on its review of the information submitted by TVA, the NRC staff concludes that TVA’s 

design criteria and bases related to the equipment hatches and stairwells are in accordance 

with the guidelines of Positions D.1.j and D.4.f of Appendix A to BTP (APCSB) 9.5-1 and, 

therefore, are acceptable. 

3.1.4 Fire Doors

In FPR Part II, Section 12.10.4, TVA stated that fire door assemblies (doors, frames, and 

hardware) are provided for door openings required as part of fire barriers. Fire doors have been 

evaluated in accordance with NFPA 80-1975, “Standard for Fire Doors and Fire Windows.” Fire 

doors are normally provided with closing mechanisms. In addition, TVA stated in FPR Part VII, 

Section 4.1, that some fire doors have been altered by the addition of signs and security 

hardware, or have been damaged and repaired on site. Closing mechanisms and latches 

provided on doors are inspected to ensure proper functioning. Special purpose doors 

(e.g., flood, heavy equipment) installed in fire barriers have been evaluated by a fire protection 

engineer for acceptability.

TVA installed UL-listed fire door assemblies (doors, frames, and hardware) in door openings 

that are required as part of fire barriers. These door assemblies are either A-labeled (3-hour), 
FF-18

for 3-hour fire barriers, or B-labeled (1-1/2-hour), for fire barriers having a fire rating of 2 hours 

or less. Furthermore, TVA stated that security hardware incorporated into a fire door assembly 

does not adversely impact the fire rating of the assembly in accordance with NRC staff guidance 

in Section 3.2.3 of GL 86-10.

Sliding fire doors are provided in selected locations, such as rooms protected with gaseous fire 

suppression systems. In areas not protected by gaseous fire suppression systems, these 

sliding fire doors are closed by a fusible link. TVA stated that, in areas protected by automatic 

carbon dioxide (CO2) suppression systems, fire doors close upon CO2 system actuation or by 

thermal actuation of a thermal link.

TVA stated that special purpose doors (e.g., air lock doors, equipment doors, and 

submarine-type doors) cannot be purchased as labeled fire-rated doors. FPR Part VII, 

Section 4.1, summarizes TVA’s evaluation of these types of fire door in light of the NRC staff 

guidance. Section 6.2.2 of this evaluation provides the NRC staff evaluation approval of this 

exception.

Based on its review of the information submitted by TVA, the NRC staff concludes that TVA’s 

design criteria and bases related to the installation of fire doors in fire barrier assemblies are in 

accordance with the guidelines of Position D.1.j, of Appendix A to BTP (APCSB) 9.5-1 relating 

to the fire doors and, therefore, are acceptable.

3.1.5 Fire Dampers

Fire dampers are used to maintain the required ratings of fire-rated barriers (walls, partitions, 

and floors) when they are penetrated by ductwork, with the goal of preventing the propagation 

of fire through ducts. TVA stated that fire dampers are provided in heating, ventilation, and air 

conditioning (HVAC) ducts that penetrate required fire barriers. Some duct penetrations do not 

have fire-rated dampers and are unprotected openings. Fire dampers are provided with 

appropriately rated fusible links based on the ambient temperatures in the location. Fire 

dampers in safety-related HVAC systems may have double fusible links installed if required by 

a single failure analysis. Furthermore, TVA stated that ventilation openings through fire barriers 

required to comply with NRC regulations are protected by fire dampers having a rating 

equivalent to that required of the barrier. TVA stated that fire dampers have been evaluated 

per the guidance of NFPA 90A-1975, “Standard for the Installation of Air-Conditioning and 

Ventilating Systems,” although TVA did not specifically commit to that standard.

In selected areas protected by automatic CO2 suppression systems, these dampers also close 

during the CO2 system discharge. These fire dampers that provide CO2 suppression system 

isolation capability are actuated by a release mechanism when the CO2 system activates, if not 

actuated by a thermal link prior to CO2 system discharge.

In FPR Part VII, Section 3.4, TVA stated that there are two instances of large fire dampers 

that do not meet NRC staff guidance. Section 6.2.8 of this evaluation provides the NRC staff 

evaluation and approval of these configurations.

FPR Part VII, Section 3.5, summarizes TVA’s evaluation of the fire damper in the volume control 

tank (VCT) rooms’ fire door from the NRC staff guidance. Section 6.2.10 of this evaluation 

provides the NRC staff evaluation and approval of this configuration.
FF-19

FPR Part VII, Section 6.2, summarizes TVA’s evaluation of relaxing the surveillance frequencies 

for fire dampers in high radiation or contaminated areas. Section 6.3.2 of this evaluation 

provides the NRC staff evaluation and approval of this configuration.

FPR Part VII, Section 4.7, summarizes TVA’s evaluation of 1-1/2-hour rated fire dampers in 

3-hour rated fire barriers. Section 6.2.7.5 of this evaluation provides the NRC staff evaluation 

and approval of these exceptions.

Based on its review of the information submitted by TVA, the NRC staff concludes that fire 

dampers at WBN are installed consistent with Positions D.1.j and D.4.i of Appendix A to BTP 

(APCSB) 9.5-1, and, therefore, are acceptable.

3.1.6 Fire Barrier Penetration Seals

3.1.6.1 Electrical and Mechanical Penetration Seals 

In FPR Part II, Section 12.10.6, TVA discussed seals that are installed in areas in which plant 

commodities, such as pipes, cable trays, conduits, etc., pass through fire rated barriers. TVA 

tested these seals to the time-temperature curve in American Society for Testing and Materials 

(ASTM) standard ASTM E119, “Standard Test Methods for Fire Tests of Building Construction 

and Materials,” at an independent fire testing laboratory with experience in the testing of 

penetration seals.

The testing showed that the penetration seals could withstand the fire endurance test without 

the passage of flame or gases hot enough to ignite cable or fire stop material on the unexposed 

side for a period equal to the required fire rating. In addition, for seals required to meet other 

plant design bases requirements, such as radiation shielding, HVAC pressure differential, 

and/or flood, they were tested for such capability. 

TVA stated that electrical and mechanical penetration seal configurations at WBN have 

withstood a hose stream test in accordance with Institute of Electrical and Electronics 

Engineers (IEEE) 634-1978, “Cable-Penetration Fire Stop Qualification Test,” or ASTM 

E-814-83, “Standard Test Method for Fire Tests of Penetration Firestop Systems,” without the 

hose stream causing an opening through the penetration seal that would permit a projection of 

water beyond the unexposed side.

TVA stated that the 1-, 2-, and 3-hour fire-rated mechanical penetrations were tested in 

accordance with ASTM E-814-83, for the fire rating. Penetrant service temperature and any 

thermal or mechanical movement of the penetrant were also considered in the testing of the 

mechanical penetration seals.

TVA stated that 1-, 2-, or 3-hour fire-rated electrical penetration seals were tested in accordance 

with IEEE 634-1978. Transmission of heat through the penetration seal was limited to 700 ° F 

or the lowest auto-ignition temperature of cable in the penetration, whichever is lower.

Conduit penetrations that were poured in place during plant construction have internal seals. 

TVA stated that internal seal materials, design, and locations in walls and floor/ceiling 

assemblies have been evaluated as equivalent to tested configurations. For conduits with 

external seals (e.g., the conduits passing through a sleeve larger than the conduit), the external 

seal meets the same criteria as stated for electrical penetration seals.
FF-20

Based on its review of the information submitted by TVA, the NRC staff concludes that the fire 

protection information presented in the FPR conforms to the guidelines of Positions D.1.j and 

D.3.d of Appendix A to BTP (APCSB) 9.5-1 and, therefore, is acceptable. 

3.1.6.2 Internal Conduit Fire Barrier Penetration Seals 

TVA stated that conduits that pass through fire barriers are provided with internal smoke and 

gas seals. TVA stated that these seals have a minimum of 3 inches of silicone foam and 1 inch 

of ceramic fiber damming installed at the bottom or back side of the foam seal. TVA further 

stated that conduits that terminate in closed junction boxes or other noncombustible sealed 

enclosures do not need internal smoke seals, except for conduits in the auxiliary and secondary 

containment envelope boundary. In addition, TVA stated that an electrical cubicle, such as in a 

motor control center (MCC) or in a switchgear cabinet, is considered combustible and therefore 

would have internal conduit seals at or near the fire barrier. Conduits that are routed through 

the fire area and that do not terminate in the area do not have internal seals. 

For lengths of conduit that extend less than 1 foot beyond the plane of a fire barrier, regardless 

of diameter, a fire seal is installed. For other combinations of diameters and lengths of conduit, 

TVA uses a graded approach for the installation of internal conduit seals, as provided in FPR 

Part II, Section 12.10.6. For smaller diameter conduits, a short length of conduit from the 

barrier is sufficient to restrict smoke or hot gases. For larger diameter conduits, longer lengths 

of conduit from the barriers are needed to adequately restrict the travel of smoke or hot gases.

Based on its review of the information submitted by TVA, the NRC staff concludes that TVA’s 

criteria for the installation of internal conduit fire and smoke seals are equivalent to the 

guidelines of Positions D.1.j and D.3.d of Appendix A to BTP (APCSB) 9.5-1 and, therefore, are 

acceptable. 

3.2 Safe Shutdown Capability

3.2.1 Separation of Safe Shutdown Functions 

In order to ensure that one train of equipment remains free of fire damage, where components 

of redundant trains of systems necessary to achieve and maintain hot shutdown conditions are 

located within the same fire area outside the containment, TVA stated that equipment, 

components, cables, and associated circuits of redundant, safe shutdown systems are 

separated in accordance with the following separation criteria in Section III.G.2(a) through 

Section III.G.2(c) of Appendix R to Title 10 of the Code of Federal Regulations (10 CFR) 

Part 50:

(a) Separation of cables and equipment and associated non-safety circuits of redundant 

trains by a fire barrier having a 3-hour rating. Structural steel forming a part of or 

supporting such fire barriers shall be protected to provide fire resistance equivalent to 

that required of the barrier;

(b) Separation of cables and equipment and associated non-safety circuits of redundant 

trains by a horizontal distance of more than 20 feet with no intervening combustible or 

fire hazards. In addition, fire detectors and an automatic fire suppression system shall 

be installed in the fire area; or
FF-21

(c) Enclosure of cable and equipment and associated nonsafety circuits of one redundant 

train in a fire barrier having a 1-hour rating. In addition, fire detectors and an automatic 

fire suppression system shall be installed in the fire area;

For safe shutdown components located inside the containment building, TVA used one of the 

means noted above, or one of the following means to achieve separation between trains:

• fire detectors and automatic fire suppression installed in the area; or

• separation of equipment, components, and associated circuits of redundant systems by 

a RES 

In order to conform to the fire protection and safe shutdown train separation criteria as 

described in Section III.G.2(b) of Appendix R to 10 CFR Part 50 listed above, TVA took credit 

for a safe shutdown analysis volume (AV) evaluation methodology and also took credit for 

enhanced automatic fire suppression consisting of preaction sprinklers located at the ceiling 

level and below obstructions in the large general plant areas, and areawide ionization smoke 

detection.

TVA used the AV methodology in order to subdivide a large fire area and then subject it to a 

detailed safe shutdown analysis in accordance with Appendix R to 10 CFR Part 50 and ensure 

that one train of safe shutdown capability remains free of fire damage. 

Under TVA’s AV methodology, an AV can consist of an entire fire area or a portion of a larger 

fire area. When the AV is a portion of the fire area, it can consist of multiple rooms, a single 

room, portions of a room (normally defined by column line locations), or any combination of the 

above. Each AV that involves only a portion of a room includes a 20 foot wide (minimum) 

“buffer zone” between it and the adjacent AV. The buffer zones are analyzed as part of the 

larger AV and as a separate AV. Every portion of a fire area is part of at least one AV.

In performing the safe shutdown analyses, safe shutdown components and cables are 

assigned to each AV containing the component. Additionally, components located in the buffer 

zones are assigned to an AV for the buffer zone.

TVA’s safe shutdown analysis is performed assuming that all components and cables in the AV 

are damaged by the postulated fire. A set of safe shutdown equipment is then selected and 

corrective actions designated to ensure safe shutdown functions can be maintained with the 

selected equipment.

Some AVs in the plant use the electrical raceway fire barrier system (ERFBS) for redundant 

trains located within a single AV. The ERFBS extends to the boundary of the AV to assure 

separation between redundant trains within the AV. For large AVs, this may not be a barrier; 

rather it may be the column line or other indicator of the edge of the AV. 

In order to provide reasonable assurance that WBN satisfied the technical guidance in Section 

III.G, “Fire Protection of Safe Shutdown Capability,” of Appendix R to 10 CFR Part 50, TVA 
FF-22

identified and used the following types of AV, as described with figures in FPR Part III, 

Section 10.3:

• Fire Area - The fire area is separated from other adjacent areas by rated barriers (walls, 

floors, and ceilings) that are sufficient to withstand the hazards associated with the area 

and, as necessary, to protect equipment in the area from a fire outside the area. 

• Single Room within a Fire Area - A room may be separated from other adjacent rooms 

in a fire area by regulatory fire barriers (walls, floors, and ceilings) that have a 1-hour or 

greater fire rating. 

• Combination of Rooms within a Fire Area - The combination of rooms in the AV are 

separated from other AVs within the same fire area by regulatory fire barriers that are 

rated for at least 1-hour

• Sections of Large General Areas - AVs consisting of sections of large general areas are 

separated from each other by “buffer zones” that are wider than 20 feet. In large 

general areas where buffer zones are used that include intervening combustibles, 

enhanced automatic suppression and detection systems are installed in the large 

general area. Where AVs are separated from other AVs by buffer zones, a fire in one of 

the AVs would not be expected to pass through the buffer zone and affect equipment in 

the AV on the other side of the buffer zone. TVA uses combinations of overlapping AVs 

in its analysis.

• Sections of Large Rooms - For AVs that consist of large room sections separated by an 

overlap region that is greater than 20 feet, the overlap region is considered to be part of 

both AVs. If the overlap region contains intervening combustibles, enhanced automatic 

suppression and detection systems are installed in the large room.

For large general areas and large rooms that have either buffer zones or overlap regions, refer 

to Section 6.1.4 of this evaluation for additional information regarding fire protection in those 

regions. 

Based on its review of the information submitted by TVA, the NRC staff concludes that TVA’s 

criteria for providing fire protection for safe shutdown functions provide an equivalent level of 

fire safety to Section III.G of Appendix R to 10 CFR Part 50 and therefore, is acceptable.

3.2.2 Safe Shutdown - General Plant Areas 

TVA’s methodology for assessing compliance with the separation/protection criteria described 

in Section III.G of Appendix R to 10 CFR Part 50 consisted of:

(a) determining the functions required to achieve and maintain safe shutdown

(b) producing shutdown logic diagrams that define minimum sets of systems capable of 

accomplishing each shutdown function

Each plant system or subsystem function relied on to accomplish the above safe 

shutdown functions is identified. A separate designator is assigned to each plant 

system or subsystem function to ensure consistency between analysis documents and 
FF-23

calculations. Each designator is identified as a safe shutdown “Key.” The safe 

shutdown logic diagram (FPR Figure III-5) depicts the safe shutdown system and/or 

system function, associated Key number, and logical relationships between systems 

and Keys used to demonstrate compliance with the criteria in Appendix R to 

10 CFR Part 50.

(c) grouping specific plant locations into fire areas

(d) identifying for each area, one or more paths through the shutdown logic diagrams that 

satisfy each required shutdown function

(e) Developing functional criteria that defined the required equipment for the shutdown 

paths

(f) Identifying power and control cables for shutdown-related equipment and associated 

circuits that are not isolated from shutdown cabling

For each safe shutdown key, cable block diagrams were developed for each safe 

shutdown component to identify cables required to ensure that the component can 

perform its safe shutdown function. Raceways that contain these required cables were 

then identified, and their locations documented. An interaction is defined as a place in 

the plant where redundant safe shutdown paths are not separated in accordance with 

the requirements in Section III.G.2 of Appendix R to 10 CFR Part 50. Whenever an 

interaction was identified, it was documented and evaluated for its impact on safe 

shutdown capability. An appropriate resolution was then determined and documented.

(g) Resolutions may consist of modifications, use of alternate equipment, operator manual 

actions (OMAs), fire barrier or RES installation, post-fire repairs, engineering 

evaluations prepared in accordance with the guidance in GL 86-10, or deviation 

requests.

Based on its review of the information submitted by TVA, the NRC staff concludes that TVA’s 

methodology for assessing compliance with the separation/protection criteria in Section III.G of 

Appendix R to 10 CFR Part 50, is acceptable.

3.2.3 Safe Shutdown Analysis 

TVA stated that its safe shutdown analysis demonstrated that sufficient redundancy exists for 

systems needed for hot and cold shutdown. The safe shutdown analysis included components, 

cabling, and support equipment needed to achieve hot and cold shutdown.

TVA stated that for hot shutdown, at least one train of the following safe shutdown systems 

would be available: (1) auxiliary feedwater (AFW) system, (2) steam generator (SG) 

power-operated relief valves (PORVs), (3) reactor coolant system (RCS), and (4) chemical 

and volume control system. For cold shutdown, at least one train of the residual heat removal 

(RHR) system would be available. TVA stated that the RHR system provides the capability to 

achieve cold shutdown within 72 hours after a fire, and would be used for long-term decay heat 

removal. The availability of these systems includes the components, cabling, and support 

equipment necessary to achieve cold shutdown. Support equipment includes the diesel 

generators (DGs) and associated electrical distribution system, the essential raw cooling 
FF-24

water (ERCW) system, the component cooling system (CCS), and the necessary 

ventilation systems.

TVA stated that an electrical separation study was performed to ensure that at least one train of 

such equipment is available in the event of a fire in areas that might affect these components. 

Safe shutdown equipment and cabling were identified and traced through each fire area from 

the component to the power source. Associated circuits whose fire-induced spurious operation 

could affect safe shutdown were identified by a system review to determine those components 

whose maloperation could affect safe shutdown capability. The potential for multiple spurious 

operation (MSO) was also analyzed. Further discussion of the MSO is presented below in 

Section 3.9, “Assessment of Multiple Spurious Operations.”

TVA stated that alternative shutdown measures are required only for fires in the control building. 

If a fire disables the MCR or requires the abandonment of the MCR, the auxiliary control room 

(ACR), which is located in a separate fire area in the auxiliary building, would be available to 

achieve and maintain the plant in hot standby and subsequent cold-shutdown conditions. The 

control functions and indications provided at the ACR panel are electrically isolated or otherwise 

separate and independent from the MCR. Further discussion of the alternative shutdown 

capability is presented below in Section 3.3, “Alternative Shutdown.”

Based on its review of the information submitted by TVA, the NRC staff concludes that the 

systems identified by TVA for achieving and maintaining safe shutdown in the event of a fire 

as described in Section III.G of Appendix R to 10 CFR Part 50 are acceptable.

3.2.4 Systems Required for Safe Shutdown 

TVA stated that shutdown of the reactor and reactivity control is initially performed by control 

rod insertion. Long term reactivity control is provided by adding borated water from the 

refueling water storage tank (RWST). RCS inventory is maintained by varying charging and 

letdown flow through the RCS makeup and letdown paths. Decay heat removal during hot 

shutdown is accomplished by establishing secondary-side pressure control and supplying water 

to two of the four SGs from one of the redundant motor- or turbine-driven AFW pumps. 

Long-term heat removal to establish and maintain cold-shutdown conditions is provided by the 

RHR system.

TVA stated that primary system pressure is controlled by the pressurizer heaters (if available) 

or by varying pressurizer level in combination with control of RCS temperature using SG 

PORVs. 

Based on its review of the information submitted by TVA, the NRC staff concludes that the 

systems selected by TVA are capable of satisfying the post-fire safe shutdown criteria in 

Sections III.G and III.L of Appendix R to 10 CFR Part 50, and therefore, are acceptable.

3.3 Alternative Shutdown

3.3.1 Areas in Which Alternative Shutdown Is Required 

TVA’s analysis identified that alternative shutdown capability is required for control building fires 

that also require shutdown from outside of the MCR. For these fires, cold shutdown must be 

achieved within 72 hours. TVA also indicated that it evaluates the alternative shutdown 
FF-25

capability in accordance with Sections III.G.3 and III.L of Appendix R. OMAs to support 

alternative shutdown are addressed in the same manner as OMAs discussed in Section 3.5.2, 

“OMAs for SSCs That Are Important to Safe Shutdown,” of this evaluation, below.

3.3.2 Alternative Shutdown System 

The alternative shutdown system uses existing plant systems and equipment identified in 

Section 3.2 above, and an ACR complex. TVA stated that the analysis indicates that for control 

building fires, no repairs are required to implement the alternative shutdown capability.

A loss of offsite power is required to be postulated for those locations that require alternative 

shutdown. TVA stated that the systems used during alternative shutdown can be powered by 

both onsite and offsite power.

The ACR complex is physically independent of the control building. Where required, electrical 

isolation of controls and indications provided for the ACR is achieved through the actuation of 

isolation/transfer switches. The ACR complex is divided into five independent rooms consisting 

of a Train A and Train B transfer switch room for each unit and the ACR. The ACR serves as 

the central control point during alternative shutdown from outside the MCR, and provides 

control and monitoring capability for redundant trains (Trains A and B) of equipment required to 

achieve safe shutdown. 

TVA also analyzed the potential for MSOs. Section 3.9 of this evaluation further discusses 

MSOs.

3.3.3 Alternative Shutdown Conclusion

Based on its review of the information submitted by TVA, the NRC staff concludes that the 

alternative shutdown system is consistent with Sections III.G.3 and III.L of Appendix R to 

10 CFR Part 50, and therefore, is acceptable.

3.4 Alternative Shutdown Performance Goals

TVA stated that the alternative shutdown system described in Sections 3.4.1 through 3.4.5 of 

this evaluation, was designed to enable the achievement of alternative shutdown performance 

goals outlined in Section III.L of Appendix R to 10 CFR Part 50.

3.4.1 Reactivity Control 

Initial reactivity control is provided by the control rods, which are inserted by the reactor 

protection system. Additional reactivity margin is provided by injecting borated water from the 

RWST into the RCS via the charging pumps. Source range monitoring instrumentation is 

available in the ACR to monitor reactivity and to ensure adequate reactivity margin.

3.4.2 Reactor Coolant Inventory 

Control of the RCS inventory requires maintaining the reactor coolant pump (RCP) seal integrity 

and RCS pressure boundary integrity and providing RCS makeup and letdown.
FF-26

RCP seal cooling is required to maintain seal integrity and to prevent an uncontrolled loss of 

reactor coolant inventory. Diverting a portion of the charging flow to the RCP seals achieves 

RCP seal cooling. Isolating the normal and excess letdown lines, in turn, isolates the RCS 

pressure boundary. To prevent depressurization of the RCS, the plant ensures that the 

solenoid valves in the reactor vessel head vent system remain closed.

RCS inventory is controlled by varying charging and letdown flow through RCS makeup and 

letdown paths. One of the redundant centrifugal charging pumps (CCPs) is required to provide 

makeup inventory to the RCS. The VCT is required to provide a short-term supply of water for 

makeup of RCS inventory and RCP seal cooling. A suction path from the RWST is required to 

provide a long-term source of borated water for RCS makeup. If necessary, inventory may be 

removed from the RCS by way of the pressurizer PORVs, discharging to the pressurizer relief 

tank (PRT), or discharging through the RCS head vent valves.

Reactor coolant makeup is usually available immediately following reactor trip from the 

charging system, except in a few fire locations where it is available within 75 minutes following 

reactor trip. TVA stated that an analysis was performed which demonstrates that makeup due 

to RCS leakage is not required for 75 minutes. TVA stated that for these scenarios, 

maintaining the RCS integrity is necessary to achieve adequate inventory control. The 

inadvertent opening of boundary isolation valves, such as the reactor head vent valves and 

RHR suction isolation valves, has been precluded, and adequate RCP seal integrity is 

maintained to assure safe shutdown.

3.4.3 Decay Heat Removal 

RCS temperature from power operation to hot-shutdown conditions is controlled by the rate of 

heat removal from the reactor coolant to the secondary-side coolant and from hot shutdown to 

cold shutdown via direct heat transfer by the RHR system to the ultimate heat sink. During 

RCS cooldown to RHR entry conditions, heat will be removed from the reactor and transferred 

to the SGs via natural circulation. The removal of decay heat for cooldown from reactor trip to 

hot standby conditions requires one AFW pump supplying water to two of the four SGs. The 

required makeup water supply can come from either the condensate storage tank (CST) or 

from ERCW.

The CST is normally aligned to the suction of the AFW pumps. WBN is supplied with two 

motor-driven AFW pumps per unit with only one per unit required for safe shutdown. The 

turbine-driven AFW pump (one per unit) is designed to deliver a sufficient flow to all four SGs 

and maintain SG water levels at the lower limit of the wide range level indicator, although it is 

not required for alternative shutdown from the ACR.

The RHR system is required to provide the long-term heat removal capability necessary to 

establish and maintain cold-shutdown conditions. The establishment of RHR cooling requires 

one RHR pump, a heat exchanger, and the associated flowpath to provide RCS coolant flow to 

the primary side of the RHR heat exchanger; one CCS pump and its associated flowpath to 

provide cooling to the secondary side of the RHR heat exchanger; and one ERCW pump and 

its associated flowpath to supply cooling water to the CCS heat exchanger. If the diesel 

generators (DGs) are required to supply required power, an additional ERCW pump would be 

required for cooling purposes.
FF-27

TVA’s post-fire shutdown analysis states that the pressurizer heaters are the preferred method 

of controlling RCS pressure, and will be used if available. If the pressurizer heaters are not 

available, RCS pressure can be controlled by controlling pressurizer level by using other 

systems, such as the charging system, or controlling RCS temperature using SG PORVs.

3.4.4 Process Monitoring 

Direct indication of process variables including reactor coolant hot-leg temperature, reactor 

coolant pressure, pressurizer level, SG level and pressure, source range flux, charging header 

pressure and flow, VCT level indication, and decay heat removal system flow are provided in 

the ACR.

TVA requested an alternative to Appendix R for instrumentation necessary to achieve 

alternative shutdown. Specifically, TVA has not provided wide-range SG level, tank level 

indication for the condensate and RWSTs, and RCS cold-leg temperature. Section 6.1.1 of this 

evaluation provides the NRC staff evaluation and approval of this configuration.

3.4.5 Support Functions 

The FPR and the associated shutdown logic diagram (FPR Figure III.5) identify the emergency 

power distribution system, offsite power system, ERCW system, CCS, HVAC to areas 

containing essential FSSD equipment, and control room chillers as required support functions.

TVA stated that this essential HVAC is provided for the control, auxiliary, DG, and reactor 

buildings. Portions of the systems in each building that service safe shutdown equipment 

required for compliance with Appendix R have been analyzed to ensure that at least one path 

of the required systems will be available for an Appendix R fire. These systems include the 

primary safety-related portions of the control building; the auxiliary building HVAC system for the 

480-V transformer rooms and for the general floor area on the 713.0-foot elevation; the 

turbine-driven AFW pump room; the DG HVAC systems including the DGs, associated 

batteries, and electrical boards; and the containment air cooling systems. All other areas of the 

plant that contain equipment required for safe shutdown per Appendix R have been evaluated 

and determined that acceptable temperatures will be maintained for the required equipment to 

perform its intended function if HVAC is lost.

3.4.6 Alternative Shutdown Performance Goals Conclusion

Based on its review of the information submitted by TVA, the NRC staff concludes that TVA’s 

treatment of alternative shutdown performance goals is consistent with Section III.L of 

Appendix R to 10 CFR Part 50, and therefore, is acceptable.

3.5 Operator Manual Actions

TVA’s post-FSSD analysis, and associated cable interaction studies, identified some fire areas 

where operator actions may be required to take manual control of equipment outside of the 

MCR to compensate for fire-induced equipment failures. Actions taken by operators within the 

MCR are described by TVA as Operator Actions, and are not considered OMAs (see 

Section 3.5.3 of this evaluation, below). TVA classified OMAs into two general categories: 

(1) manual actions for safe shutdown success path structures, systems, and components 
FF-28

(SSCs) and (2) manual actions for SSCs important to safe shutdown. Repairs for cold 

shutdown were also evaluated by TVA, but are not considered OMAs.

TVA referenced Revision 2 to RG 1.189 for the discussion of the distinction between safe 

shutdown success path SSCs and SSCs important to safe shutdown.

WBN, Unit 1, OMAs were approved in NRC SSER 18, prior to operation of WBN, Unit 1. TVA 

stated that OMAs for future required for safe shutdown SSCs, or such OMAs for WBN, Unit 1,

implemented since the SSER 18, will be submitted to the NRC for approval, consistent with the 

language in the FPR.

3.5.1 OMAs for Safe Shutdown Success Path SSCs 

In FPR Part V, Section 2.0, TVA stated that OMAs for SSCs in the safe shutdown success path 

require prior NRC approval. The position that OMAs for SSCs in the safe shutdown success 

path require prior NRC approval is consistent with the guidance in Revision 2 to RG 1.189, and 

is summarized in FPR Part V, Section 2.1.2.1.A. The TVA evaluations of WBN Unit 2 OMAs for 

success path SSCs are included in FPR Part VII, Section 8. Section 6.1.9 of this evaluation 

provides the NRC staff approval of these OMAs.

3.5.2 OMAs for SSCs That Are Important to Safe Shutdown 

In FPR Part V, Section 2.0, TVA stated that OMAs for SSCs that are important to safe shutdown 

do not require prior NRC review and approval. The position that OMAs for SSCs that are 

important to safe shutdown do not require prior NRC approval is consistent with the guidance in 

Revision 2 to RG 1.189, and is summarized in FPR Part V, Section 2.1.2.1.B. Area-specific 

evaluations for any area where WBN, Unit 2, OMAs involving important-to-safe-shutdown 

equipment that are needed to be performed in the area of fire origin are evaluated in 

Section 3.5.6 below.

TVA discussed the feasibility and reliability analysis criteria for evaluating OMAs. In FPR 

Part V, Section 2.1, TVA stated that these criteria are based on NUREG-1852.

For all important-to-safe-shutdown SSC manual actions, TVA considered defense-in-depth 

features, such as fire prevention (transient combustible and hot work controls), fire detection 

and suppression, and area separation. TVA also included consideration of the need for and the 

time involved in donning self-contained breathing apparatus when evaluating these OMAs. For 

any area crediting an OMA with less than 2 hours of required time, and that also lacks robust 

defense-in-depth fire protection features, additional time margin is included in addition to the 

nominal acceptance criteria.

TVA considers the following factors in its evaluation of these OMAs: (1) time, (2) environmental 

factors (smoke, lighting, noise, etc.), (3) necessary equipment, (4) procedures, and (5) staffing. 

Each of the factors included acceptance criteria. For example, all OMAs have an allowable time 

of 10 minutes or greater with 100-percent margin. Factors that could cause delays in the 

performance of the OMA have also been considered. Factors such as lighting and 

communications are supported by plant calculations.

TVA evaluated the access routes necessary to perform the OMAs. Because some areas of the 

plant are separated into separate AVs, it is possible that OMAs may occur in a portion of a fire 
FF-29

area that is remote to the portion where fire damage could affect safe shutdown equipment. In 

this event, additional access routes have been evaluated. TVA walked down these alternative 

access routes and determined that they are viable.

TVA used current NRC guidance to develop acceptance criteria for OMAs for SSCs that are 

important to safe shutdown. TVA incorporated a review of defense-in-depth, feasibility, and

reliability in its analysis.

TVA identified a few important-to-safe-shutdown OMAs that do not meet the general criteria in 

FPR Part V, Section 2.1.2.1.B.a and Section 2.1.2.1.B.b. For those cases, TVA performed a 

more detailed defense-in-depth analysis and also ensured that there is at least 50 percent time 

margin. Any new important-to-safe-shutdown OMAs that fall into this category will be analyzed 

in the same manner. This exception is rarely used and includes additional analysis of 

defense-in-depth features and, therefore, is acceptable.

Based on its review of the information submitted by TVA, the NRC staff concludes that OMAs 

for SSCs important to safe shutdown include consideration of defense-in-depth, feasibility, and 

reliability, and, therefore, this approach is acceptable.

3.5.3 Actions Related to Main Control Room Abandonment

3.5.3.1 Operator Actions Prior to Main Control Room Abandonment

TVA evaluated operator actions taken prior to MCR abandonment due to fire in FPR Part V, 

Section 2.3. TVA operators will, prior to leaving the control room, attempt seven actions for 

each reactor to prepare for staffing the ACR. For each reactor, the FPR describes that control 

room procedures direct operators to perform these seven actions prior to MCR abandonment:

(1) trip reactor,

(2) trip reactor coolant pumps (RCPs),

(3) close the block valves upstream of the pressurizer power operated relief valve (PORV) 

and close the PORVs,

(4) close the Main Steam Isolation Valves (MSIVs),

(5) close the Main Feedwater Isolation Valves,

(6) close the Steam Generator Power Operated Relief Valves (SG PORVs), and

 (7) open supply valves from the refueling water storage tank to the centrifugal 

charging pumps.

TVA relies on tripping the reactor ((1) above) and tripping the RCPs ((2) above) through 

operator action within the control room prior to abandonment. TVA indicated that there are 

two independent switches in the control room located about 20 feet apart that are available to 

trip the reactor and that the RCP breakers and RCP start bus breakers are 30 feet apart within 

the control room and will be tripped by plant operators prior to abandoning the control room. 

The distance between these switches provides assurance that they both will not be impacted by 
FF-30

a fire within the control room prior to control room abandonment. TVA stated that performing 

these actions from the control room prior to abandonment will ensure that the plant parameters 

that are impacted by tripping these systems will remain consistent with the performance goals in 

10 CFR Part 50, Appendix R, Section III.L.

Operators will attempt to close the pressurizer PORVs and the associated block valves prior to 

abandoning the control room ((3) above). If the operators are not successful in closing these 

valves and a spurious actuation occurs that opens a pressurizer PORV, the analysis considered 

cases where safety injection actuates and where safety injection fails to actuate. If safety 

injection actuates, plant operators have adequate time to staff the ACR and mitigate the effects 

of safety injection before the 10 CFR Part 50, Appendix R, Section III.L, performance goals 

have been violated. If safety injection does not actuate, TVA evaluated the plant response to a 

spuriously opened PORV. TVA’s analysis concluded that the plant response to a spuriously 

opened PORV without safety injection will result in violating the 10 CFR Part 50, Appendix R, 

Section III.L, process variable performance goals, but procedures and protected equipment are 

available to assure that the plant would not reach an unrecoverable condition. TVA stated that 

the plant would not violate the 10 CFR Part 50, Appendix R, III.L, process variable performance 

goals if either the actions performed in the control room prior to abandonment, or the automatic 

features, are successful. TVA analyzed that sufficient procedures and equipment are available 

to prevent an unrecoverable condition from occurring if the control room actions and automatic 

features fail and the 10 CFR Part 50, Appendix R, III.L, process variable performance goals are 

violated.

If the MSIVs ((4) above) or main feedwater isolation valves ((5) above) are not closed from the 

control room prior to abandonment, a safety injection signal could occur. TVA’s evaluation 

determined that a reactor trip would create conditions that would initiate signals to cause 

automatic closure of these valves. In the event that the automatic signal is challenged by the 

fire, plant procedures include steps to mitigate the safety injection from the ACR. TVA stated 

that plant parameters will remain consistent with the performance goals in 10 CFR Part 50, 

Appendix R, Section III.L if either the actions performed in the control room prior to 

abandonment, or the automatic features, are successful. TVA also stated that if the control 

room actions and automatic features fail, mitigating safety injection from the ACR will assure 

meeting the 10 CFR Part 50, Appendix R, Section III.L, process variable performance goals.

Operators will attempt to close the SG PORVs prior to abandoning the control room ((6) above). 

If the operators are not successful in closing these valves and a spurious actuation occurs that 

opens a SG PORV, a safety injection signal could occur. TVA’s evaluation determined that 

multiple spurious signals are required to open these valves, and that they fail closed. In the 

event that the spurious signals that would open one of these valves are caused by fire damage, 

plant procedures include steps to mitigate the safety injection from the ACR, and to take manual 

control of these valves, if necessary. TVA stated that plant parameters will remain consistent 

with the performance goals in 10 CFR Part 50, Appendix R, Section III.L if either the actions 

performed in the control room prior to abandonment, or the manual control of the valves, are 

successful. TVA also stated that if the control room actions and manual control fail, mitigating 

safety injection from the ACR will assure meeting the 10 CFR Part 50, Appendix R, Section III.L, 

process variable performance goals.

TVA stated that operators will also attempt to open the supply valves to the CCPs from the 

RWST prior to abandoning the control room ((7) above). Opening these valves will help to 

ensure that the CCPs have a suitable water supply. If the normal supply to the CCPs is 
FF-31

impacted by a fire, there is an automatic signal generated to open the RWST feed to the CCPs. 

In the event that both the control room operator action and the automatic actuation fail due to 

fire damage, TVA has abnormal operating procedures available to assure safe shutdown 

capability. TVA stated that performing these actions from the control room prior to 

abandonment and relying on the automatic transfer would ensure that CCPs will not be 

impacted. In the event that these features fail, TVA stated that procedures are in place 

to ensure that the plant does not reach an unrecoverable condition.

In addition, TVA assumes that a fire in the MCR would be characterized by slow growth and 

be detected in its early stages by control room operators or installed smoke detection systems. 

Fires in other areas of the control building may require MCR abandonment, such as in the cable 

spreading room or auxiliary instrument room, etc. The control building areas other than the 

MCR have installed detection and automatic suppression systems that reduce the likelihood of a 

large fire, or have an alternative documented in FPR Part VII, Section 2.3, and evaluated in 

Section 6.1.3 of this evaluation. In the event that a fire were to occur in the panel where the 

above controls are located, other controls in the MCR could serve the same function, or controls 

are available outside the control room that can be operated in sufficient time to prevent the 

adverse actions listed above.

In NRC request for additional information (RAI) FPR V-16, the NRC staff expressed a concern 

that a fire in portions of the control building that lack fire detection and automatic suppression 

could impact equipment important to safe shutdown. In TVA’s letter dated August 5, 2011 

(Agencywide Documents Access and Management System (ADAMS) Accession 

No. ML11227A257), TVA stated that, in the control building, the PORV block valve controls are 

only routed through areas that have detection and automatic suppression. Other circuits routed 

through the control building are either routed through areas with fire detection and automatic 

suppression, or areas with detection and limited combustibles and ignition sources.

Based on its review of the information submitted by TVA, the NRC staff concludes that TVA 

considered the credible fire scenarios, separation of controls or automatic features, and the 

other installed defense-in-depth features to determine that these control room actions are 

feasible and, therefore, are acceptable.

3.5.3.2 Operator Actions and OMAs Following Control Room Abandonment

TVA provided specific consideration for actions taken following a control room abandonment. 

TVA stated in FPR Part V, Section 2.1.2.1.C that actions taken at the ACR and OMAs 

performed as part of a control room abandonment have at least 10 minutes of allowed time and 

100-percent time margin.

There are a few actions that do not have this allowable time and time margin, specifically, 

OMA 1638 for Unit 1 and OMA 1639 for Unit 2. These actions are performed by the control 

room operator on the way to the ACR. So for these actions, there is no time delay for the arrival 

of an auxiliary operator. This is a very short time action—5 minutes—and TVA demonstrated 

that the action can be performed in less than 1 minute. TVA also indicated that there are 

alternative methods to achieve the same result as the preferred path that is provided by these 

OMAs. Since this action is performed by a control room operator (not a recalled auxiliary 

operator), there is over 100-percent time margin, and there are alternative (nonpreferred) 

methods to perform this function if the OMA fails, the NRC considers this action acceptable.
FF-32

Based on its review of the information submitted by TVA, the NRC staff concludes that, based 

on allowable time, time margin, and available alternative methods, actions performed by 

operators following a control room abandonment are acceptable.

3.5.4 Safe Shutdown Procedures and Manpower

TVA developed a fire response procedure, Abnormal Operating Instruction (AOI)-30.1, “Plant 

Fires,” which describes operator response and mitigating actions for plant fires. TVA also 

developed room-specific procedures as part of AOI-30.2, “Fire Safe Shutdown,” for rooms 

where OMAs may be required to mitigate damage to plant safe shutdown equipment. AOI-30.2 

is supported by controlled plant calculations. The procedures include operator-by-operator 

actions for a fire in any room of the plant that would require OMAs to shut down the plant.

TVA walked down the OMAs for both WBN, Unit 1, and WBN, Unit 2. In some cases, TVA used 

a bounding approach that involved timing more challenging or involved OMAs, and using that

timing for less challenging OMAs. Additionally, OMAs needed after 2 hours into the fire were 

not walked down, since 2 hours corresponds to the time frame for additional personnel to arrive 

at the plant in response to an event. TVA postulates that significant plant fires are interior to the 

plant; therefore, operators who are called back are not expected to have difficulty getting to the 

plant.

TVA stated that, upon confirmation of a fire, such as actuation of cross-zoned fire detection, 

high-pressure fire pump auto-start, CO2 fire extinguishing system actuation, or plant staff 

observation, steps are initiated that include recalling auxiliary unit operators needed to perform 

OMAs to their assembly location. TVA performed recall exercises and determined that auxiliary 

unit operators begin to be available within about 3 minutes. The fire detection systems are likely 

to annunciate to the control room before a fire impacts plant operation, such that for WBN, 

auxiliary unit operator recall is expected to occur prior to tripping the plant and declaring the 

Appendix R event.

TVA stated that the start of the time “clock” for the performance of OMAs is the tripping of the 

plant. Prior to tripping the reactor, the plant is considered to be in a stable operating condition. 

Once the trip is initiated, the clock starts and preventive OMAs are performed to prevent 

spurious equipment operation and to ensure safe shutdown can be accomplished. In the event 

that an Appendix R event coincides with a plant trip, without prior detection or recall of the 

auxiliary unit operators, there is still available margin for the most time-critical manual actions 

to be performed. This is based on the 100-percent available margin for every manual action, no 

manual action needed in less than 10 minutes, and the 3-minute recall time for the first auxiliary 

unit operator.

TVA analyzed the possibility that a fire could cause a plant trip prior to recalling operators to 

the control room. TVA postulated three trip scenarios and concluded that the available fire 

protection features of fire detection, automatic suppression, and physical separation will allow 

time for the MCR to recall the auxiliary equipment operators prior to an automatic trip.

TVA stated that there are some rooms that lack fire detection, but it evaluated these areas and 

determined that there were no OMAs needed for FSSD in these areas. For these areas, TVA 

will rely on normal and emergency operating instructions in the event of a fire, and OMAs and 

fire response procedures are not used.
FF-33

Most of WBN’s OMAs are preventive; however, some reactive OMAs must be taken upon fire 

damage to SSCs rather than reactor trip. TVA stated that for these reactive type actions, the 

normal plant operating procedures provide an appropriate reactive response to fire damage. 

TVA analyzed the available FSSD equipment on an area-by-area basis to assure that sufficient 

safe shutdown equipment is free of fire damage such that safe shutdown can be achieved using 

equipment that is free of fire damage.

Based on its review of the information submitted by TVA, the NRC staff concludes that TVA’s 

safe shutdown procedure structure, including both preventive and reactive OMAs, has been 

evaluated to ensure safe shutdown capability and, therefore, is acceptable.

3.5.5 Repairs 

TVA stated that repair activities (e.g., lifting/cutting leads, installing jumpers, and fuse 

replacement) are not required to achieve and maintain hot standby conditions. TVA identified 

the following three generic repairs to be performed to achieve cold shutdown: 

• RHR Room Cooler Repair, 

• RHR/RCS High-Low Pressure Boundary Valve Repair

• Long Term RCS Inventory Reduction.

Cold-shutdown repair activities include the installation of electrical jumpers, the installation of 

portable air supplies, and the installation of replacement cables and components if needed due 

to fire damage. TVA identified the specific activities to be performed and developed repair 

procedures to implement this capability. Additionally, materials necessary to accomplish the 

repairs are available on site.

Based on its review of the information submitted by TVA, the NRC staff concludes that the 

repair activities developed by TVA to achieve cold shutdown conditions are consistent with 

Appendix R to 10 CFR Part 50 and therefore, are acceptable.

3.5.6 WBN Unit 2 OMAs Involving Fire Area Re-Entry

TVA examined WBN Unit 2 OMAs that involve re-entry into plant fire areas in FPR Part VII, 

Section 8.4. This section discusses actions involving important to safe shutdown equipment, 

whereas Section 6.1.9 of this evaluation addresses OMAs involving equipment required for safe 

shutdown.

TVA indicated that all WBN Unit 2 rooms that involve re-entry to perform OMAs for important to 

safe shutdown equipment are equipped with fire detection and automatic suppression. Areas 

that do not have complete automatic suppression coverage are discussed in Section 6.1.7 of 

this evaluation. In addition, TVA stated that all such OMAs have more than 60 minutes for the 

licensee staff to extinguish the fire and to operate the equipment within the room. TVA 

determined that the equipment within the room of fire origin is unlikely to be damaged such that 

the equipment could not be operated following a postulated fire in that room. TVA performed

feasibility and reliability evaluations of the OMAs.
FF-34

Based on its review of the information submitted by TVA, the NRC staff concludes that there is 

sufficient defense-in-depth available, detection and automatic suppression is installed, and that 

the manual action provides sufficient margin to assure safe shutdown capability. Therefore, the 

NRC staff concludes that such re-entry into rooms to perform OMAs involving important to safe 

shutdown components is acceptable.

3.6 Associated Circuits

TVA examined the potential impact of fire damage on associated circuits of concern. TVA 

categorized associated circuits as follows:

• Type I – common power source, 

• Type II – spurious actuation,

• Type III – common enclosure.

TVA stated that it identified these associated circuits of concern in accordance with GL 81-12, 

the NRC staff’s clarification to GL 81-12, and GL 86-10.

3.6.1 Circuits Associated by Common Power Source 

TVA stated that, for circuits associated by a common power source, it identified all circuits 

supplied from a power source (i.e., switchgear, MCCs, and load centers) that also powers a 

circuit of equipment required for post-FSSD. For the identified circuits, TVA verified the 

coordination of electrical protection devices (e.g., fuses, circuit breakers, or relays) to ensure 

that a fire-induced fault on a branch circuit of a required supply will be cleared by at least 

one branch circuit protective device before the fault current can propagate to cause a trip of any 

feeder breaker upstream of the required supply.

In its letter dated August 5, 2011 (ADAMS Accession No. ML11227A257), in response to 

RAI FPR III-17, TVA stated that a list of the design change packages had been issued to ensure 

that the WBN, Unit 2, circuits are adequately protected with fuses/breakers to address common 

power supply and common enclosure associated circuits of concern. In its letter dated 

June 24, 2015 (ADAMS Accession No. ML15175A508), TVA stated that these actions have 

been completed.

TVA evaluated circuits associated by a common power source for multiple high impedance 

faults (MHIFs). TVA stated that MHIFs are evaluated in accordance with the base case 

conditions in Appendix B.1 to Nuclear Energy Institute (NEI) 00-01, “Guidance for Post Fire Safe 

Shutdown Circuit Analysis,” Revision 2, issued May 2009, as endorsed by Section 5.5.2 of 

Revision 2 to RG 1.189. The base case set of conditions, if met, provides reasonable 

assurance that MHIFs will not occur. The FPR, Part III, Section 7.4, analysis provided the 

NEI 00-01 base case conditions with the corresponding WBN compliance method for each base 

case condition. The FPR states: “WBN meets the NEI 00-01, Appendix B.1 base case criteria 

which establish applicability of the base case to individual plant designs.” WBN did not take any 

exceptions to the base cases. In a letter dated June 27, 2012 (ADAMS Accession 

No. ML12181A531), TVA provided a list of supporting calculations for the FPR, Part III, 

Section 7.4, MHIF analysis.
FF-35

Based on its review of the information submitted by TVA, the NRC staff concludes that TVA’s 

method of evaluating circuits associated by a common power source is consistent with the NRC 

guidance in the GLs identified in Section 3.6 of this evaluation above, and in Appendix B.1 to 

NEI-00-01, as endorsed by RG 1.189, and, therefore, is acceptable.

3.6.2 Circuits Associated by Spurious Operation 

TVA stated that cables that are not part of safe shutdown circuits may be damaged by the 

effects of postulated fires. This cable damage may consequently prevent the correct operation 

of safe shutdown components, or result in the maloperation of equipment which would directly 

prevent the proper performance of the safe shutdown systems. The effects of spurious 

operations may be conceptually divided into two subclasses as follows:

(1) maloperation of safe shutdown equipment due to control circuit electrical interlocks 

between safe shutdown circuits and other circuits (e.g., the numerous safe shutdown 

equipment automatic operation interlocks from process control and instrument circuits)

(2) maloperation of equipment that is not defined as part of the safe shutdown systems, but 

that could prevent the accomplishment of a safe shutdown function (e.g., inadvertent 

depressurization of the RCS or the main steam system by spurious opening of boundary 

valves)

TVA performed an evaluation of Appendix R to 10 CFR Part 50 events to ensure that any 

failure of associated circuits of concern by spurious operation will not prevent safe shutdown. 

Credible electrical faults considered in the analysis included open circuit, short circuit 

(conductor-to-conductor), short to ground, and cable-to-cable (hot-short) including 3-phase 

hot-shorts for high/low pressure interface valves. The analysis also considered that the 

normally ungrounded 125 VDC power distribution system may become grounded due to fire 

damage.

TVA indicated that these Type II associated circuits of concern outside of containment are 

analyzed in accordance with the criteria in Sections III.G.2.a, III.G.2.b, and III.G.2.c of 

Appendix R to 10 CFR Part 50 as required circuits. Inside containment, the Type II associated 

circuits of concern are analyzed in accordance with the criteria in Sections III.G.2.d, III.G.2.e, 

and III.G.2.f of Appendix R to 10 CFR Part 50 as required circuits.

Based on its review of the information submitted by TVA, the NRC staff concludes that TVA’s 

approach to analyze circuits associated by spurious operation, in accordance with the sections 

of Appendix R to 10 CFR Part 50 listed above, is acceptable.

3.6.3 Circuits Associated by Common Enclosure 

To address the common enclosure associated circuit concern, TVA evaluated all circuits that 

may share a common enclosure (e.g., cable tray, conduit, panel or junction box) with a circuit 

required by Appendix R to 10 CFR Part 50. On the basis of its evaluation, TVA concluded that 

the electrical protective equipment provided will ensure that electrical faults and overloads will 

not result in any more cable degradation than would be expected when operating conditions are 

below the setpoint of the electrical protective device.
FF-36

TVA stated that the plant addressed associated circuits by common enclosure by ensuring that 

all required existing (prior to 1995) circuits in buildings with safe shutdown components are 

electrically protected with a fuse or breaker that will actuate prior to the jacket of existing faulted 

cables from reaching their auto-ignition temperature. Additionally, for new circuits, associated 

circuit electrical fault protection is provided to ensure that the fuse or breaker will operate prior 

to the temperature of the insulation reaching its insulation damage temperature.

In its letter dated August 5, 2011 (ADAMS Accession No. ML11227A257), in response to 

RAI FPR III-17, TVA stated that a list of the design change packages had been issued to ensure 

that the WBN Unit 2 circuits are adequately protected with fuses/breakers to address common 

power supply and common enclosure associated circuits of concern. In its letter dated 

June 24, 2015 (ADAMS Accession No. ML15175A508), TVA stated that these actions have 

been completed.

Based on its review of the information submitted by TVA, the NRC staff concludes that TVA’s 

methodology for assessing circuits associated by common enclosure is consistent with the 

NRC guidance in the GLs identified in Section 3.6 of this evaluation, and, therefore, is 

acceptable. 

3.7 Current Transformer Secondaries

Section III.G.2 of Appendix R to 10 CFR Part 50 requires that fire induced open circuits be 

analyzed where they could prevent operation or cause maloperation of components required for 

post-FSSD. 

If a fire at a remote location causes the secondary circuit of a current transformer (CT) to open, 

the event could generate ionized gases or additional fires, or both, in other locations and could 

propagate fire to additional fire areas. 

TVA evaluated the fire hazards due to a fire-induced open circuit in the secondary circuits of 

CTs installed in high energy panels (i.e., 6.9 kV switchgear) of the required power systems. An 

evaluation of three types of CT circuits used in the auxiliary power system has been done: 

(1) ground fault, (2) differential relaying, and (3) protective relaying.

The CT circuits are contained in their respective panels for the Appendix R to 10 CFR Part 50 

required and nonrequired 480 V switchgear and the 6.9 kV switchgear. Therefore, the fire 

would have to be localized in the switchgear assembly for the CT secondary circuit to be 

opened by a fire. This would prevent the CT circuits from causing fire propagation to other fire 

areas.

The 6.9-kV CT circuit that is connected to protective relaying and a current transducer is also 

contained within the switchgear panel. The output of the current transducer is connected to a 

remote indicator, and the current transducer is an electrical isolator. Additionally, the 

output-to-input of the current transducer has been tested for 1500-VAC differential. Electrical 

isolation also exists for the Watt & volt-amperes-reactive transducers used on the 6.9-kV 

switchgear at WBN.

The board differential relaying circuits are totally internal to the switchgear panels, except for 

the following three exceptions: 
FF-37

(1) The circuits between the 6.9 kV switchgear emergency supply feeders and the DGs are 

included in the interaction analysis as required circuits. The protective relays are 

designed to operate and clear these circuits in case of fire damage.

(2) The common station service transformers transformer differential relaying circuits are 

also included in the interaction analysis as required circuits. The current imbalance 

created by an open CT circuit causes the protective differential relay to open the supply 

circuit breaker, which removes primary power to the CT, clearing the circuit, within the 

time required for protective relay and breaker operation.

(3) The circuits between the 6.9 kV start and unit boards are not required circuits. Similar 

to item (2) above, current imbalance in the protective differential relay of the 

nonrequired circuits would open the supply circuit breaker.

Based on its review of the information submitted by TVA, the NRC staff concludes that TVA’s 

approach to evaluating the fire hazards due to fire-induced open circuits in the secondary of 

CTs installed in high energy panels is in accordance with Section III.G.2 of Appendix R to 

10 CFR Part 50, and therefore, is acceptable. 

3.8 High/Low-Pressure Interfaces

TVA stated that GL 81-12, GL 86-10, and Information Notice (IN) 87-50, “Potential LOCA at 

High- and Low-Pressure Interfaces from Fire Damage,” dated October 9, 1987, describe 

special considerations for high/low pressure interfaces that are necessary to meet TVA’s 

commitment to Appendix R to 10 CFR Part 50.

In accordance with GL 81-12, the following information is necessary to ensure that high/low 

pressure boundary interfaces are adequately protected for the effects of a single fire:

(1) Identify each high/low pressure interface that uses redundant electrically controlled 

devices (such as two series motor operated valves) to isolate or preclude rupture of any 

primary coolant boundary.

(2) Identify the essential cabling for each device.

(3) Identify each location where the identified cables are separated by a barrier having less 

than a 3-hour fire rating.

(4) For the areas identified in (3) above (if any), provide the bases and justification.

GL 86-10 states that, in the case of high/low pressure interfaces, the following must be 

considered: (1) a hot short on all three phases of three phase AC circuits in the proper 

sequence to cause spurious operation of a motor, and (2) two hot shorts of proper polarity 

without grounding on ungrounded DC circuits resulting in spurious operation.

The information in IN 87-50 states that, because the high pressure from the reactor coolant 

system could result in failure of the low pressure piping, at least one isolation valve must remain 

closed despite any damage that may be caused by fire for those low pressure systems that 

connect to the reactor coolant system.
FF-38

Based on the above, TVA performed a review of the systems credited for safe shutdown to 

identify potential high/low pressure interfaces. These interfaces were evaluated to identify 

valves that, if spuriously opened, would expose low pressure piping to high pressure resulting 

in failure of the low pressure system.

The control system for RHR valves has been designed to prohibit opening unless the reactor 

coolant pressure is low enough to prevent RHR piping failure. However, if these valves opened 

spuriously, exposure of RHR piping to high pressure may cause failure of the RHR system 

piping and render the system inoperable. Therefore, the RHR/RCS isolation valves 

(1/2-FCV-74-1, -2, -8, and -9) are considered high/low pressure interface valves.

Excess letdown is not required for safe shutdown. However, the spurious opening of these 

valves could expose downstream piping to excess pressure that may cause failure resulting in 

the rupture of the primary coolant boundary. Therefore, the excess letdown isolation valves 

(1/2-FCV-62-55, and -56) are considered high/low pressure interface valves.

Normal letdown is not required for safe shutdown. However, spurious opening of these valves 

may cause failure to maintain RCS inventory control. Therefore, the normal letdown isolation 

valves (1/2-FCV-62-69A and -70A) are considered high/low pressure interface valves.

The safety injection system (SIS)/RHR interface valve with the RCS is located in piping that 

connects the SIS with the RHR system at a point between the RCS/RHR isolation valves. The 

SIS is not required for safe shutdown. However, the spurious opening of valve 1/2-FCV-63-186 

along with either 1/2-FCV-74-1-A or -9-B could expose the SIS piping to damaging pressure. 

Therefore, this valve is considered a high/low pressure interface.

The pressurizer PORV and reactor head vent isolation valves are designed to function at high 

RCS operating pressure. They provide a safe shutdown function by initially remaining closed 

for RCS inventory control purposes. The pressurizer PORVs also provide RCS pressure control 

during cooldown. The primary method of RCS inventory control during cooldown is to maintain 

pressurizer level while makeup is provided by injecting borated water through the RCP seals as 

part of seal cooling and the RCS cooldown rate is adjusted by controlling SG level and pressure 

to provide RCS volume shrinkage equal to the volume of fluid injected through the RCP seals. 

The reactor head vent isolation valves (where available) and the pressurizer PORVs can also 

be used to reduce RCS inventory. Discharge from the RCS through these valves is directed to 

the inlet of the PRT. The inlet lines are sized to accommodate vent/relief discharge flow without 

piping or component failure. Continuous blowdown to the PRT may eventually cause spillage of 

excess coolant to containment through the PRT rupture disks. Therefore, the pressurizer 

PORVs and block valve combinations, and reactor head vent isolation valves, are considered 

high/low interface valves.

To prevent fire-initiated cable faults from causing a spurious operation of the RHR isolation 

valves, all four of the motor-operated valves in the RHR suction line are kept closed (prefire 

condition) with the corresponding MCC breaker in the open position. The return lines are 

isolated by two series check valves in each line and a common motor-operated valve.

In its letter dated May 30, 2012 (ADAMS Accession No. ML12153A374), TVA stated that 

procedural controls for isolation of all potentially spurious RCS letdown paths, including 

pressurizer PORVs and reactor head vents, provide assurance (through the use of MCR actions 
FF-39

for WBN, Unit 2, and MCR actions and an OMA for WBN, Unit 1) that isolation of normal and 

excess letdown paths will be achieved.

Based on its review of the information submitted by TVA, the NRC staff concludes that TVA’s 

approach for high/low pressure interfaces meets Appendix R to 10 CFR Part 50 and follows the 

guidance in GL 81-12 and GL 86-10, and IN 87-50, and therefore, is acceptable.

3.9 Assessment of Multiple Spurious Operations

In FPR Part III, Section 11.0 “Multiple Spurious Operation (MSO) Evaluation,” TVA stated that 

Revision 2 to RG 1.189 formalized the guidance for addressing multiple fire induced circuit 

failures, or MSOs and multiple concurrent hot shorts. TVA further stated that this process was 

followed to address fire-induced spurious failures for WBN Unit 1. In a letter dated 

August 20, 2010 (ADAMS Accession No. ML102360283), TVA stated that the MSO scenarios 

requiring resolution for WBN Unit 1 would be implemented under the timing requirements 

prescribed by the NRC in Enforcement Guidance Memorandum 09-002, “Enforcement 

Discretion for Fire Induced Circuit Faults.” 

By letter dated November 5, 2010 TVA submitted TVA MSO Evaluation R-1976-20-001, “Watts 

Bar Nuclear Plant Unit 2 Multiple Spurious Operation Evaluation, Revision 1” (ADAMS 

Accession No. ML103160419). TVA stated that multiple fire-induced spurious failures were 

evaluated at WBN Unit 2 as described in Revision 2 to RG 1.189. TVA further stated that, 

based on the results of the MSO expert panel conducted at the plant for WBN Unit 1, various 

scenarios were identified and were reviewed for WBN Unit 2. Appendix B, “Unit 2 Resolutions,” 

and Appendix C, “Unit 1/Common Resolutions,” of the above report provided resolutions for 

specific unresolved MSO scenarios that affect WBN Unit 2. 

In a letter dated February 7, 2013 (ADAMS Accession No. ML130440108), TVA stated that the 

above MSO scenarios requiring resolution for WBN Unit 1 have been resolved and incorporated 

into Appendix B and Appendix C to Revision 2 to MSO Evaluation R-1976-20-001. 

TVA included Revision 3 to MSO Evaluation R-1976-20-001, as an enclosure to its letter dated 

September 18, 2014 (ADAMS Accession No. ML14265A449). In this revision, TVA indicated 

that, in a number of cases, the actions related to the WBN Unit 2 MSO scenarios requiring 

resolution in Appendix B had been completed. 

In Section 4 of the MSO Evaluation (Revisions 1, 2, and 3), TVA stated that MSO scenarios 

selected for Sequoyah Nuclear Plant, Units 1 and 2, and for WBN Unit 1 were evaluated to 

determine if the scenarios were applicable to WBN Unit 2 and how WBN Unit 2 complied with 

each scenario. Sequoyah Nuclear Plant, Units 1 and 2, and WBN Units 1 and 2, have similar 

physical and systems designs. All four units are Westinghouse four-loop pressurized water 

reactors with wet ice condenser containments and would be expected to have similar MSO 

scenarios. Additionally, the Sequoyah Nuclear Plant, Units 1 and 2, MSO scenarios were 

analyzed from a dual unit perspective. 

In a letter dated June 24, 2015 (ADAMS Accession No. ML15175A508), TVA stated that all 

MSO scenarios resolution actions for WBN Unit 2 have been completed. Finally, in FPR Part III, 

Section 11.0, TVA stated that the WBN Unit 1 MSO evaluation is documented in TVA 

calculation XDN00000020110002, “WBN Unit 1 Multiple Spurious Operation Review for RG 

1.189 Revision 2,” and that the WBN Unit 2 MSO evaluation is documented in TVA calculation 
FF-40

EDQ0009992012000085, “Appendix R - WBN Unit 2 Multiple Spurious Operation (MSO) 

Evaluations.”

Based on its review of the information submitted by TVA, the NRC staff concludes that by 

evaluating multiple fire-induced spurious failures in accordance with the guidance in Revision 2 

to RG 1.189 and by using MSO scenarios from Sequoyah Nuclear Plant, Units 1 and 2, and 

WBN Unit 1 when addressing WBN Unit 2 and dual-unit scenarios, TVA’s approach is an 

acceptable means for addressing MSO failures.

3.10 Smoke Control and Ventilation

FPR Part VIII, Section D.4, and FPR Part X, Section 3.2.9, discuss smoke control and 

ventilation. TVA stated that plant ventilation systems at WBN are not specifically designed to 

exhaust smoke or corrosive gases. TVA further stated that a combination of the normal 

ventilation exhaust systems and portable fans are used to remove smoke from specific rooms 

during and after firefighting activities. Nonrecirculating ventilation systems are provided for fire 

areas that may contain airborne radioactive materials. Smoke from fires that might occur in 

areas containing radioactive materials is monitored for radioactivity. 

Based on its review of the information submitted by TVA, the NRC staff concludes that smoke 

control and ventilation for fire protection purposes at WBN are installed consistent with 

Position D.4 of Appendix A to BTP (APCSB) 9.5-1, and, therefore, is acceptable.

3.11 Lighting and Communications

TVA stated that fixed, self-contained lighting units with individual 8-hour minimum battery power

supplies are provided in areas that must be manned for safe shutdown, and in access and 

egress routes to and from all fire areas containing controls for equipment required for safe 

shutdown. TVA stated that an assessment of the emergency lighting at OMA locations and the 

access routes to OMA locations has been performed to assure the adequacy of the lighting. 

These walkdowns were performed under local or general area blackout conditions and were 

used to document the adequacy of the lighting levels.

In FPR Part VII, Section 2.7, TVA requested an alternative to its commitment for emergency 

lighting criteria inside the reactor building, yard area, and the turbine building. Section 6.1.6 of 

this evaluation provides the NRC staff’s evaluation of this alternative.

Based on its review of the information submitted by TVA, the NRC staff concludes that the 

emergency lighting is consistent with Section lll.J of Appendix R to 10 CFR Part 50 and the 

guidelines contained in Section D.5.a of Appendix A to BTP (APCSB) 9.5-1 and, therefore, 

is acceptable.

TVA provided several means of communication to support safe shutdown operations, including 

(1) telephones, (2) a code, alarm, and paging system, (3) sound-powered phones, and 

(4) two-way radios. The two-way radio system consists of multiple radio repeaters, portable 

radios, and redundant distributed antenna systems. The radio cabinets, power cables, and 

redundant fixed repeaters are widely separated so that one fire cannot affect all the radios. The 

radios have redundant diesel and battery backed power supplies. There are antennas for the 

radios located on the auxiliary building exhaust stack, in addition to internal distributed antenna 
FF-41

systems in the control and turbine buildings. Additionally, two widely separated trunk lines feed 

the radio signal to redundant distributed antenna systems in the auxiliary building.

The two-way radio system is the primary means of communication for performing manual 

shutdown actions and for fire brigade firefighting operations. Some plant areas lack full radio 

coverage; however, coverage is available immediately outside of these rooms. Sound-powered 

phones are available in the ACR and local stations to supplement the radio system and support 

alternative shutdown.

Based on its review of the information submitted by TVA, the NRC staff concludes that TVA’s 

means of communications do not take any exceptions to Positions D.5.c and D.5.d of 

Appendix A to BTP (APCSB) 9.5-1 and, therefore, are acceptable.
FF-42

4.0 FIRE PROTECTION SYSTEMS

The NRC staff evaluated TVA’s fire protection program for compliance with 10 CFR 50.48(a), as 

described in Section 2.1 of this evaluation. TVA evaluated their program against the guidance 

in Appendix A to Branch Technical Position (BTP) (Auxiliary Power Conversion Systems Branch 

(APCSB)) 9.5-1. The NRC staff determined that TVA’s reliance on the information in Appendix 

A to BTP (APCSB) 9.5-1, was acceptable since it was the guidance in place on April 18, 1977, 

when TVA first submitted a fire hazards analysis to the NRC for review. In addition to TVA’s 

evaluation to Appendix A to BTP (APCSB) 9.5-1, since TVA used guidance published prior to 

1981, TVA also committed to meet Sections III.G, III.J, and III.O, of 10 CFR Part 50, Appendix 

R, to assure safe shutdown capability. This section of the evaluation describes the NRC staff’s 

evaluation of TVA’s compliance with 10 CFR 50.48(a)(2)(ii) and portions of 10 CFR 

50.48(a)(1)(iv). As discussed below, the NRC staff’s review found that the Watts Bar fire 

protection program meets the requirements of 10 CFR 50.48(a).

4.1 Water Supply and Distribution

Tennessee Valley Authority (TVA) described the fire water supply system at Watts Bar Nuclear 

Plant (WBN) in Fire Protection Report (FPR) Part II, Section 12.1, “Water Supply.” TVA also 

described the system in its response by letter dated August 5, 2011 (Agencywide Documents 

Access and Management System (ADAMS) Accession No. ML11227A257), to request for 

additional information (RAI) FPR II-45. TVA stated that the high-pressure fire protection (HPFP) 

water system is common to both units and that it consists of four electric motor-driven pumps 

and one diesel engine-driven pump.

TVA stated that the electrically driven pumps are seismic Category I high-pressure vertical 

turbine motor-driven pumps in accordance with Section III of the American Society of 

Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (ASME Code). Each pump 

is rated at 1,590 gallons per minute (gpm) at 130 pounds per square inch, gauge (psig). TVA 

calculated the maximum required fire water demand based on the largest automatic 

suppression system demand and hose streams, and it stated that each of these pumps can 

supply 50 percent of the required flow. The pumps are located in the seismic Category I intake 

pumping station (IPS) with a 3-hour-rated fire barrier that separates two fire pumps from the 

other two fire pumps.

TVA stated that a 100-percent capacity, Underwriters Laboratories, Inc., (UL)-listed, diesel fire 

pump is remotely located in the yard near the WBN, Unit 1, cooling tower. TVA stated that the 

diesel fire pump is capable of developing a flow of 2,500 gpm (100 percent capacity) at 125 psig 

and 3,750 gpm (150 percent capacity) at 81 psig.

TVA stated that the water supply for the electric fire pumps is taken from the Tennessee River 

and the diesel fire pump takes its water from the WBN, Unit 1, cooling tower basin. TVA stated 

that the Tennessee River is essentially unlimited and that the WBN, Unit 1, cooling tower basin 

can provide a minimum of 2 hours supply at 150 percent of the capacity of the diesel pump.

TVA stated that the electric pumps are automatically started by activation of the fire detection 

systems associated with installed automatic water-based suppression systems. Also, the 

electric pumps can be started manually from either the main control room (MCR) or the 

appropriate 480-V shutdown board. The diesel pump automatically starts on low system 
FF-43

pressure or can be manually started from the MCR or locally at the pump. The diesel pump 

can only be stopped locally at the pump.

TVA stated that each electric fire pump is powered from a separate 480-V shutdown board

and that, in the event of loss of offsite power, each 480-V shutdown board is automatically 

connected to a separate emergency diesel generator (DG). Indications of fire pump motor 

running condition and loss of line power on the line side of the switchgear are provided in the 

MCR for each electric pump. The diesel fire pump also sends annunciation signals to the MCR.

TVA stated that the electric fire pumps also serve as a backup water supply to the auxiliary 

feedwater (AFW) system in the event of a flood above plant grade (called “flood mode”). TVA 

stated that, as a result, this requires the use of pumps that meet the requirements in Section III 

of the ASME Code as opposed to traditional fire pump installations that are UL-listed or factory

mutual-approved pumps in accordance with National Fire Protection Association 

(NFPA) 20-1993, “Standard for the Installation of Centrifugal Fire Pumps,” for electric driven 

pumps. In FPR Part VII, Section 5.1, TVA stated the following:

(1) pump curve verification tests have been performed to include multiple diverse points on 

the pump curve to replicate fire pump test requirements as opposed to the single point 

verification applicable to ASME Code Section III pumps;

(2) TVA performed hydraulic calculations to demonstrate that the pumps provide adequate 

flow and pressure to the most hydraulically remote suppression systems;

(3) the electrical circuits for pump power and control meet IEEE Class 1E standards and, 

even though the pumps do not start on pressure drop in the piping system, they do start 

on activation of the fire detection systems associated with pre-action suppression 

systems; and

(4) the fire pumps can only be manually stopped from the MCR, which is continuously 

manned, or the shutdown board rooms in the auxiliary building.

Based on the above information submitted by TVA, the U.S. Nuclear Regulatory Commission 

(NRC) staff concludes that, while not designed to the guidelines in NFPA 20, the electric fire 

pump configuration will not negatively affect the performance of the fire protection system, and 

meets the purpose of the guidelines of Section E.2.c of Appendix A to Branch Technical

Position (BTP) (Auxiliary Power Conversion Systems Branch (APCSB)) 9.5-1 and, therefore, 

is acceptable.

TVA stated that the diesel fire pump installation and its associated controller are installed in 

accordance with NFPA 20-1993. Based on the information submitted by TVA that states that 

the diesel fire pump and associated controller are installed in accordance with NFPA 20-1993, 

the NRC staff concludes that the installation of the diesel fire pump is acceptable.

TVA stated that a self-cleaning strainer capable of handling 100-percent flow is provided on 

the discharge side of each pair of electric fire pumps. The strainers are located in the IPS and 

conform to the requirements in Section III of the ASME Code for seismic Category I 

components. For the diesel fire pump, TVA stated that mechanical screens are provided on the 

supply side and a strainer on the discharge.
FF-44

TVA stated that the HPFP system is interconnected to the raw cooling water (RCW) and raw 

service water (RSW) systems. Automatic isolation valves are provided to isolate the RCW 

system and selected RSW loads from the HPFP system when any fire pump is started to reduce 

the RSW load on the HPFP system to ensure adequate flow and pressure are available. During 

normal operation, HPFP system pressure is maintained by the RCW pumps.

The HPFP system mains consist of both cement-lined iron yard mains and unlined steel safety 

headers. The steel safety headers serve as a backup water supply to the AFW system in “flood 

mode,” as noted above. TVA stated that the details of the “flood mode” are documented in 

several places in the final safety analysis report (FSAR), for example, FSAR Section 2.4.14.2, 

“Plant Operation during Floods above Grade.” The piping inside buildings is unlined steel. The 

buried steel piping has an exterior coating to prevent corrosion. The electric fire pumps feed the 

steel headers and the iron yard main. The diesel pump feeds the iron yard main. The two loops 

(iron and steel) are connected at the IPS (via normally open valve 0-FCV-26-17) and at 

two remote points in the auxiliary building (via normally open valves 0-FCV-26-15 and 

0-FCV-26-16). TVA stated that pressure control is provided by a pressure control 

valve downstream of the four electric pumps and downstream of the diesel pump.

TVA stated that sectional isolation valves are provided on the iron yard main to allow 

maintenance on portions of the system while the plant maintains its firefighting capability. In 

addition, TVA stated that the sectional isolation valves in the underground and building loops 

are locked or sealed in position and that surveillance is performed to ensure proper system 

alignment. The plant has not installed any sectional valves on the underground portions of the 

steel safety headers, but there are isolation valves for each header in the intake pumping station 

(IPS) and auxiliary building. Because the two headers are redundant and because they are also 

connected to the iron yard main through valves in the auxiliary building, the plant could isolate 

either main and would still have two sources of fire water available.

TVA stated that all post-indicator-type valves are either sealed or locked open with a 

key-operated “breakaway” type lock. TVA further stated that curb box valves are not locked 

open, but TVA considers these valves to be tamper-resistant because they cannot be operated 

without a special “key” tool that is not generally available.

In the FPR, TVA stated that the WBN fire water supply system is able to provide the designed 

firefighting capacity either with one electric pump and the diesel pump unavailable or with the 

hydraulically least demanding portion of any loop main out of service. TVA further stated that 

the design flow demand consists of design flow to the largest sprinkler or water spray system 

plus design flow to nonisolated RSW loads and 500 gpm for hose streams.

TVA stated that, in most plant buildings, suppression systems and hose station standpipe 

systems are separately connected to the yard main or to headers within buildings and are fed 

from each end of the building, so that a single failure cannot impair both the primary (automatic 

suppression) and secondary (manual suppression) systems at the same time. TVA further 

stated that in the Unit 1 and Unit 2 reactor buildings and the intake pumping station, the primary 

suppression systems consist of fixed, installed systems (hose stations and some automatic 

suppression) supplied from one main supply header. In these buildings, hose stations from an 

adjoining area serve as the backup suppression systems, and are supplied from a separate, 

independent main supply header.
FF-45

As result of the concern with microbiologically induced corrosion (MIC) and other corrosion 

issues, TVA instituted a permanent monitoring program for assuring the performance of the 

standpipe and suppression systems. TVA stated that this testing is performed at the 

hydraulically most remote hose stations every 3 years. TVA uses the calculated design 

basis pressure and flow requirements for these hose stations as the basis to monitor 

system performance.

TVA’s design calculation reduces the actual unlined steel pipe inside diameter by 0.8 inches 

and uses a Hazen-Williams C factor of 55 for the sections of piping that are normally wetted. 

TVA stated that the purpose of these piping restrictions and the C factor of 55 are to account for 

the 40-year service life of the pipe. The data collected from these tests will be compared to the 

calculated values and trended to detect system failure.

TVA stated that all raw water systems, including the HPFP, are chemically treated in a manner 

that is consistent with nuclear industry practice. TVA stated that this treatment includes 

oxidizing biocide, nonoxidizing biocide, phosphate, and zinc. TVA further stated that the 

oxidizing and nonoxidizing biocides are used to control Asiatic clams, zebra mussels, slime, and 

MIC; the phosphate is used to sequester iron from existing corrosion products; and the zinc acts 

as a mild corrosion inhibitor for the carbon steel surfaces. As described in TVA’s letter dated 

August 5, 2011 (ADAMS Accession No. ML11227A257), in response to RAI FPR VII-2.1, the 

nonoxidizing biocide treatments are coordinated with periodic system flushes in order to 

distribute the biocide to normally stagnant portions of the system.

TVA stated that two programs have been implemented to combat pipe corrosion. First, TVA 

implemented the Corrosion Control Program, which primarily monitors pipe wall thickness using 

ultrasonic techniques, replacing lengths of pipe when minimum wall thickness cannot be 

maintained. Additionally, TVA stated that a WBN Buried Piping Plan has been established 

in support of Nuclear Energy Institute 09-14, “Guideline for the Management of Underground 

Piping and Tank Integrity.” TVA described this program as providing for the risk ranking of 

buried piping relative to installed conditions (e.g., design and construction practices, as well 

as soil characteristics) and consequences of a failure of the piping. TVA stated that these 

programs are intended to provide assurance in the integrity of the HPFP system boundaries.

In addition, TVA performed a code compliance review against NFPA 24-1973, “Outside 

Protection,” as documented in FPR Part X. No substantial exceptions were identified.

Based on its review of the information submitted by TVA, the NRC staff concludes that, with the 

exception of the system design demand, the fire water supply system conforms to the guidelines 

of Sections E.2 and E.3.a of Appendix A to BTP (APCSB) 9.5-1 and, therefore, is acceptable. 

Regarding the system design demand, the NRC staff concludes that it conforms to the NRC’s 

current guidance found in Position 3.2.1 of RG 1.189, Revision 2 and, therefore, is acceptable.

4.2 Active Fire Control and Suppression Features

4.2.1 Automatic Fire Suppression Systems

4.2.1.1 Sprinklers and Fixed Spray Systems with Closed Heads 

In FPR Part III, Section 10.3.1, TVA stated that all analysis volumes (AVs) that contain 

redundant safe shutdown equipment are protected to ensure that the plant maintains its safe 
FF-46

shutdown capability. In most cases, the means of protection is consistent with Section III.G.2 of 

Appendix R to Title 10 of the Code of Federal Regulations (10 CFR) Part 50. For instance, in 

areas in which cables of redundant safe shutdown equipment are located in an AV and could be 

damaged by fire, the plant ensures the function by installing a 1-hour electrical raceway fire 

barrier system (ERFBS) on one train with automatic fire detection and suppression or by 

installing a 3-hour ERFBS in areas that do not provide fire detection and suppression. TVA 

also stated that, if separation between rooms in the same fire area is less than 3-hour rated, 

the plant either provides automatic detection and suppression systems or identifies and 

justifies exceptions.

Where TVA uses an alternative to their commitment to meet Section III.G.2, with respect to 

coverage of suppression and detection systems, TVA performed evaluations to demonstrate 

that an adequate level of protection is provided. Section 6.1.7 of this evaluation provides the 

NRC staff evaluation and approval of these alternatives.

In FPR Part VII, Section 2.4, TVA described a methodology used to resolve situations where 

redundant trains are separated by more than 20 feet, but without 20 feet free of intervening 

combustibles. Section 6.1.4 of this evaluation provides the NRC staff evaluation and approval 

of this configuration.

Where provided, TVA stated that sprinkler systems and fixed water spray systems are designed 

in accordance with the applicable guidance in NFPA 13-1975, “Standard for Installation of 

Sprinkler Systems,” and NFPA 15-1973, “Standard for Water Spray Fixed Systems.” 

In addition, TVA performed a code compliance review and identified several areas in which the 

sprinkler and fixed spray systems differed from the code. The important exceptions to the 

NFPA 13-1975 code identified were as follows:

• Fire department pumper connections for the sprinkler system are only provided to 

buildings with one connection to the underground fire main. (NFPA 13, Section 2-7). 

The NRC staff concludes that this arrangement meets the intent of the provision.

 

• Strainers are provided in the supply to each pre-action sprinkler system in lieu of 

following flushing guidance. (NFPA 13, Section 3-37.3.) The NRC staff concludes that 

this arrangement meets the intent of the provision.

• Sprinklers are not provided below the open grating above the high-pressure fire pump 

flow control valve on elevation 692 feet in the WBN Unit 1 penetration room (Room 

692.0-A7). TVA stated that the combustible loading in this fire area is insignificant. This 

grating is approximately 5 feet wide by 15 feet long and is 15 feet above the room floor. 

Two sprinklers are installed approximately 3 feet above the grating. Plant procedures 

prohibit the storage of material on these grated walkways, so the gratings would be free 

of foreign obstructions. Due to the size of the grating (4 in. by 1 in.), flow from the 

sprinklers is not expected to be restricted by the grating. (NFPA 13, Section 4-4.11.) 

The NRC staff concludes the current sprinkler configuration in the Unit 1 penetration 

room is acceptable. 

The NRC staff reviewed the other code exceptions from NFPA 13 that TVA proposed in FPR 

Part X, and determined that the exceptions will not affect the performance of the systems and, 

therefore, are acceptable. 
FF-47

With respect to NFPA 15, TVA did not take any exceptions to the code for the water spray 

systems protecting outdoor transformers, the hydrogen trailer, turbine hydrogen seal oil unit, 

or the turbine lube oil reservoir. TVA used the guidance of NFPA 13 and 15 to design the 

directional fusible nozzle water spray systems used to protect certain charcoal filters and the 

reactor coolant pumps.

TVA stated that automatic preaction sprinklers are provided in areas in which it is important to 

prevent accidental discharge of water. Operation of the preaction sprinkler system is initiated by 

a signal from the fire detection system in the area. Actuation can also be initiated manually by 

mechanical operation at the deluge valve. In addition, selected preaction systems at WBN have 

manual actuation stations placed at strategic locations remote from the valve. These systems 

are provided with air supervision if the piping downstream of the system control valve supplies

more than 20 sprinkler heads.

TVA stated that, where manually activated suppression systems are installed, the piping 

network isolation valve is maintained in the closed position. Personnel are alerted to a problem 

in these areas by the fire detection system and, after confirming there is a fire, personnel open 

the appropriate isolation valve to allow water into the system. Water is then applied to the fire 

when the heat from the fire melts the fusible element in the sprinkler head. Water flow is 

subsequently stopped by manually closing the associated isolation valve.

In FPR, Part VIII, TVA stated that drainage is provided to remove the expected fire protection 

water flows or control the accumulation of water such that the water will not cause 

unacceptable damage to equipment in the area. TVA further stated that additional drainage 

can be achieved by diverting water into adjacent rooms. Finally, TVA stated that water draining 

from areas that may contain radioactivity is sampled and analyzed before being discharged into 

the environment.

TVA stated that standpipes, hose stations, and portable fire extinguishers are provided 

throughout the control building, but fixed fire suppression systems are not provided for all 

rooms. TVA justified the lack of fixed automatic suppression capability by stating that the 

control building is a single fire area with fire detection provided throughout the control building 

except in certain areas, and that there are no alternative shutdown cables or equipment located 

in the control building, thereby satisfying the design intent of maintaining safe shutdown 

capability for a postulated fire event by providing an alternate design concept. Based on TVA’s 

justification, the NRC staff concludes TVA’s approach to be acceptable. See Section 3.3 of this 

evaluation for a discussion of alternate shutdown, and Section 6.1.3 for a detailed discussion of 

the lack of areawide automatic suppression in the control building.

In all cases, TVA stated that an adequate level of protection is provided via a combination of 

limited combustible materials, administrative controls, fire-rated barriers, spatial separation, and 

active fire protection systems. Where exceptions or alternatives to NRC staff guidance, rules, or 

design standards exist, TVA stated that it has performed evaluations to ensure that an adequate 

level of protection is provided. The NRC staff reviewed TVA’s approach to the use of sprinkler 

and water spray fire suppression systems, and concludes that TVA’s design criteria and bases 

are consistent with Positions E.2 and E.3.c of Appendix A to BTP (APCSB) 9.5-1 and the 

defense-in-depth concept described in Appendix R to 10 CFR Part 50, and, therefore, are 

acceptable.
FF-48

4.2.1.2 Gas Suppression Systems 

TVA stated in FPR Part II that automatic total flooding carbon dioxide (CO2) suppression 

systems are provided for the auxiliary instrument rooms and computer room in the control 

building; and in the lube oil storage room, diesel engine rooms (4), fuel oil transfer room, and 

480-V board rooms (4), located in the DG building.

TVA stated that the CO2 systems are designed and installed in accordance with NFPA 12-1973, 

“Carbon Dioxide Extinguishment Systems,” the code of record for these systems. Further, TVA 

stated in its letter dated March 16, 2011 (ADAMS Accession No. ML13060A403), in response to 

RAI FPR II-6, that the systems installed in the computer room, DG electrical board rooms, lube 

oil storage room, and fuel oil transfer room are installed for property protection purposes only, 

and do not have soak time requirements. In addition, TVA stated that the systems are 

appropriate for the anticipated hazards and that it performed system dump tests to ensure agent 

concentration, agent reserve, and functionality of the distribution system.

TVA stated that a signal from either the fire detection system or a pushbutton station activates 

the area alarms, CO2 discharge timer (which actuates the master control valve), and the area 

selector valve permitting the CO2 to be discharged into the selected area. In addition, the 

system can be manually operated via the electromanual pilot valve for each hazard protected on 

the loss of power to the system. In designing these systems, TVA considered personnel safety 

by providing the predischarge alarm to notify anyone in the area that CO2 is going to be 

discharged, and by adding an odorant to the CO2 to warn personnel that the system has 

been discharged.

In addition, TVA stated that the actuation of these systems causes selected fire dampers and 

doors to the protected area to close and the selected heating, ventilation, and air conditioning 

(HVAC) fans to the area to shut down, ensuring that the minimum concentration of CO2 is 

maintained and preventing fire spread from the area of fire origin. TVA also stated that it 

performed full discharge tests for representative rooms in conjunction with door fan 

pressurization tests to validate CO2 concentration and soak times.

The CO2 storage tank for supplying CO2 to systems that protect the DG building is located in the 

DG building. The DGs are protected from the effects of a postulated failure of this tank by an 

18-inch-thick reinforced concrete wall. The vent path for the tank room for the storage tank 

compartment is through a set of double doors that lead into a stairwell then, if needed, through 

another set of double doors which open to the atmosphere from the stairwell.

The CO2 for the balance of the plant is supplied from a storage tank in an underground vault in 

the yard. TVA stated that the system is designed such that failure of the system cannot pose a 

threat to any safety-related areas or structures.

The NRC staff reviewed TVA’s approach to the use of automatic CO2 fire suppression systems 

and concludes that TVA’s design criteria and bases are consistent with Positions D.4.i and E.5 

of Appendix A to BTP (APCSB) 9.5-1 and, therefore, are acceptable.
FF-49

4.2.2 Manual Suppression Capability

4.2.2.1 Hose Stations 

In FPR Part II, TVA stated that hose stations for manual firefighting are located throughout the 

plant to ensure that an effective hose stream can be directed to any safety-related area in the 

plant. TVA further stated that the system is designed according to the guidance of 

NFPA 14-1974, “Standpipe and Hose System for Sizing, Spacing, and Pipe Support 

Requirements,” except for those hose stations in certain areas of the plant in which TVA 

requested an exception to exceed the 100-foot hose spacing limitation. These exceptions are 

discussed in Section 6.2.4 of this evaluation.

In addition, TVA performed a code compliance review and identified several areas in which the 

manual firefighting hose stations and standpipe system differed from the code. TVA also 

performed evaluations to justify these exceptions. The significant NFPA 14 code exceptions 

identified and associated justifications are:

• The standpipes located on elevations 676.0 feet, 692.0 feet, 713.0 feet, 729.0 feet, 

757.0 feet, 772.0 feet, and 782.0 feet of the auxiliary building are supplied by a 3-inch 

pipe rather than the 4-inch pipe, and elevation 755.0 feet of the control building has 

2-1/2-inch supply piping. TVA stated that it verified by hydraulic calculation that these 

pipe sizes are adequate. (NFPA 14, Section 212.)

• Six standpipes (0-26-677, 0-26-690, 1-26-674, 1-26-675, 2-26-674, and 2-26-675) 

are not provided with header isolation valves. TVA stated that these systems can be 

isolated using sectionalizing valves and that this would not preclude the ability to provide 

hose stream coverage in the locations normally served by these standpipes. (NFPA 14, 

Sections 413 and 622.)

• Pressure reducing devices are not installed at the hose stations. TVA justified this by 

stating that the hose stations are for fire brigade use, and the fire brigade personnel are 

trained in the use of high pressure fire hoses. TVA further stated that the hoses and 

related fittings are maintained to accommodate the expected system pressures. 

(NFPA 14, Section 442.)

• High pressure valves, pipes, and fittings are not used, even though system spikes of up 

to 190 psi occur due to pump start surges. TVA stated that the piping and fittings can 

withstand the working pressures of the system and that the system is in accordance with 

American National Standards Institute B31.1, “Code for Pressure Piping,” system 

requirements. (NFPA 14, Sections 625, 631, and 641.)

• Water flow alarms are not provided on all standpipes. TVA stated that the hose stations 

are provided for fire brigade use. Other site personnel are trained to report fires before 

using firefighting equipment (if they have been trained in its use). Therefore, TVA 

concluded that sufficient notification of standpipe use will be provided to the MCR 

without water flow alarms. (NFPA 14, Section 67.)

The NRC staff reviewed TVA’s proposed exceptions from NFPA 14 and determined that they 

will not affect the performance of the hose stations and the standpipes. Therefore, the NRC 

staff concludes that the exceptions are acceptable.
FF-50

In FPR Part VIII, TVA stated that drainage is provided to remove the expected fire protection 

water flows or control the accumulation of water such that the water will not cause unacceptable 

damage to equipment in the area. TVA further stated that additional drainage can be achieved 

by diverting water into adjacent rooms. Finally, TVA stated that water draining from areas that 

may contain radioactivity is sampled and analyzed before being discharged into the 

environment.

TVA stated in the FPR that hose station nozzles appropriate for the expected hazards 

(e.g., electrically safe) are provided for each hose station. In addition, TVA stated, in FPR 

Part VIII, and in its letter dated August 5, 2011 (ADAMS Accession No. ML11227A257), in 

response to RAI FPR II-41.1 and RAI FPR VII-17.1, that provisions have been made to supply 

water at sufficient pressure and capacity to the standpipes, hose stations, and hose connections 

for manual firefighting in areas required for safe plant shutdown in the event of a safe-shutdown 

earthquake.

Based on its review of the information submitted by TVA, the NRC staff concludes that the 

standpipe system and hose stations align with the guidance in Positions E.3.d and E.3.e of 

Appendix A to BTP (APCSB) 9.5-1 and, therefore, are acceptable.

4.2.2.2 Fire Extinguishers 

TVA stated that portable fire extinguishers of a size and type compatible with specific hazards 

are strategically located throughout the plant for use by trained personnel. TVA also stated that 

fire brigade members and fire watch personnel have been trained on the location of 

extinguishers for firefighting operations through the extinguisher inspection program. In 

addition, TVA stated that fire extinguishers are inspected on a quarterly basis. 

TVA’s proposed application and strategic distribution of portable fire extinguishers throughout 

the plant is consistent with the guidance contained in Position E.6 of Appendix A to BTP 

(APCSB) 9.5-1, and provides reasonable assurance that fire extinguishers will be readily 

available and quickly accessed in the event of a fire emergency. Therefore, the NRC staff 

concludes that TVA’s proposed application and strategic distribution of portable fire 

extinguishers is acceptable.

4.3 Fire Detection Capability

In FPR Part III, Section 10.3.1, TVA stated that all AVs containing redundant safe shutdown 

equipment are protected to ensure safe shutdown capability is maintained. In most cases, the 

means of protection is consistent with 10 CFR Part 50, Appendix R, Section III.G.2. For 

instance, where cables of redundant safe shutdown equipment are located in an AV and could 

be damaged by fire, the function is ensured either by the installation of 1-hour ERFBS on 

one train with automatic fire detection and suppression, or a 3-hour ERFBS where fire detection 

and suppression are not provided. TVA also stated that if separation between rooms in the 

same fire area is less than 3-hour rated, automatic detection and suppression systems are 

provided or exceptions are identified and justified.

In cases where coverage of suppression and detection systems does not align with the criteria 

of 10 CFR Part 50, Appendix R, Section III.G.2, TVA performed evaluations to demonstrate that 

an adequate level of protection is provided. Section 6.1.7 of this evaluation provides the NRC 

staff evaluation and approval of these configurations.
FF-51

In FPR Part VII, Section 2.4, “Intervening Combustibles,” TVA described a methodology used to 

resolve situations where redundant trains are separated by more than 20 feet, but without 

20 feet free of intervening combustibles. Section 6.1.4 of this evaluation provides the NRC staff 

evaluation and approval of this configuration.

As described by TVA in FPR Part II, the fire detection system consists of initiating devices, 

local control panels, a remote transmitter-receiver providing a remote multiples function, 

computerized multiplex central control equipment, and a power supply. A central processing

unit (CPU) of the computerized multiplex central control equipment communicates with the local 

control panels via the remote transmitter/receiver units over looped circuits. TVA stated that,

where detection is provided for the protection of safety-related or fire safe shutdown equipment, 

Class A, four-wire, supervised circuits link the fire detectors to the local control panels and 

annunciate status change to a constantly attended location. In addition, a second CPU is 

provided in a constantly attended location as an alternate for the primary processor.

TVA stated that the fire detection system uses photoelectric, ionization, and thermal detectors. 

The fire detection system also monitors duct detectors and devices for monitoring fire 

suppression system piping integrity, water or CO2 flow, and diesel fire pump status. The fire 

detection system gives an audible and visual alarm, and also annunciates in the control room. 

TVA stated that, where detection systems are provided, the detection systems are designed in 

accordance with the applicable guidance of the NFPA 72D-1975, “Installation, Maintenance and 

Use of Proprietary Signaling Systems,” and NFPA 72E-1974, “Automatic Fire Detectors.” In 

addition, TVA performed a code compliance review and identified several areas in which the 

systems differed from the code. The significant NFPA 72D and NFPA 72E code exceptions 

identified were as follows:

(a) The operation and supervision of fire alarms is not the primary function of the system 

operators (i.e., the control room operators). The operators are responsible for all control 

room alarms and controlling the plant. (NFPA 72D, Section 1223.) This is acceptable to 

the NRC staff, because, consistent with the role and training of the operators, a fire 

alarm actuation is an event that will be responded to, and will not be ignored.

(b) TVA committed to confirm that the fire detection monitoring panel in the MCR meets the 

definition of listed in proposed WBN Unit 2 license condition 2.C(9), which states: 

By May 31, 2018, TVA shall report that a listing organization acceptable 

to the NRC (as the Authority Having Jurisdiction) has determined that the 

fire detection monitoring panel in the main control room either meets the 

appropriate designated standards or has been tested and found suitable 

for the specified purpose. 

“Listing organization” is defined in NFPA 72 (2010 edition) as one that is:

...concerned with evaluation of products or services, that maintains 

periodic inspection of production of listed equipment or materials or 

periodic evaluation of services, and whose listing states that either the 

equipment, material, or service meets appropriate designated standards 

or has been tested and found suitable for a specified purpose.
FF-52

(NFPA 72D, Sections 1213 and 2022) The NRC staff concludes that this is acceptable 

because it will confirm that a listing organization acceptable to the NRC determines that 

the fire detection monitoring panel in the MCR either meets the appropriate designated 

standards or has been tested and found suitable for the specified purpose, ensuring that

the panel is in alignment with the NFPA code.

(c) Actions upon receipt of a fire alarm signal; the fire brigade is not immediately notified. 

TVA stated that, upon receipt of an alarm from a detection system, an individual 

(auxiliary or fire operator) is dispatched to the area to determine the cause of the alarm. 

If a fire exists, the individual notifies the MCR, and control room operators notify the 

plant fire brigade. If both detection zones alarm of a cross-zoned detection system, the 

fire brigade is notified immediately. (NFPA 72D, Section 1251.) The NRC staff 

concludes that this arrangement is acceptable because it allows false alarms to be 

addressed while maintaining rapid response by the site fire brigade to actual fires.

(d) The fire alarm system uses the emergency diesel generators as the automatic 

secondary power supply. The uninterruptible power supply backup and batteries inside 

the fire detection monitoring panel supply selected devices within the panel. 

(NFPA 72D, Sections 2223, 2224, and 2231). The NRC staff concludes that this 

arrangement is acceptable because it provides a reliable source of backup electrical 

power.

(e) Signal attachments and circuits (pressure switches) can be removed or tampered with 

and not cause an alarm. The site personnel access control system and the work control 

system provide assurances that work on such devices is properly controlled and 

documented. (NFPA 72D, Section 3423). The NRC staff concludes that this is 

acceptable because these devices are not installed in areas accessible to the general 

public, where tampering is a concern.

(f) Sprinkler system control valves are not electrically supervised; instead, the valves are 

locked open or sealed open and periodically inspected instead. TVA stated that 

administrative controls, including second party verification of position and strict 

site-access and work controls, will ensure that valves are in the correct position. 

(NFPA 72D, Section 3442.) The NRC staff concludes that this is acceptable because it 

provides assurance that the valves will be in the correct positions when needed. 

(g) Both visual and recorded displays meet the code, but records are not preserved for later 

inspection. Plant procedures have reporting requirements for conditions adverse to 

quality. These procedures require an adverse condition report to be completed before 

the end of the shift on which the problem was identified. Documentation from the fire 

alarm printout would be available to support the adverse condition report. (NFPA 72D, 

Section 4111.) The NRC staff concludes that this arrangement is acceptable because it 

will support the reconstruction of the sequence of events.

(h) The transmission of an alarm signal to the fire detection monitoring panel from a wire-towire short circuit cannot be recorded. TVA stated that a wire-to-wire short will generate 

a trouble signal which requires corrective action and associated compensatory 

measures as laid out in FPR Part II, Section 14. (NFPA 72D, Sections 4112 and 4311.) 

The NRC staff concludes that this is acceptable because this situation initiates 
FF-53

corrective actions and compensatory measures, which include roving or continuous fire 

watches.

(i) Some areas do not have installed detection as required by Appendix R to 

10 CFR Part 50 or Appendix A to BTP (APCSB) 9.5-1. TVA evaluated these areas in 

FPR Part VII, Section 3.1. (NFPA 72E, Section 2-6.5). The NRC staff concludes that 

this is acceptable because of the nature of the spaces as described in Section 6.1.7 of 

this evaluation.

(j) Smoke detectors in the high ceiling areas of the plant are not installed alternately on two 

levels. TVA addressed the issue of high ceilings by reducing the spacing of the 

detectors at the ceiling level. This reduced spacing is used on auxiliary building 

elevations 692 feet, 713 feet, 737 feet, 757 feet, and the waste packaging room. 

(NFPA 72E, Section 4-4.5.2.) The NRC staff concludes that this is acceptable because 

stratification is not a concern due to the ventilation system. 

(k) TVA uses duct detectors in lieu of area detectors in the reactor building upper and lower 

compartment coolers to provide protection specifically for the coolers. TVA stated that 

the regulatory guidance for detectors are met for the remainder of the reactor building. 

(NFPA-72E, Section 8-1.1.2). The NRC staff concludes that this is acceptable because 

these detectors are installed to protect these specific pieces of equipment (e.g., the 

compartment coolers) and not the general area.

(l) Duct detectors are not provided per NFPA 90A guidance, which requires that activation 

of a detector automatically stops the ventilation system. Instead, fans serving the area 

of the plant containing the fire are shut down manually to ensure that air flow will not 

prevent fire dampers from closing. (NFPA-72E, Section 8-1.2.1.) The NRC staff 

concludes that this is acceptable because it accomplishes the goal of the provision. 

Additionally, the HVAC system has been designed as described in WBN FSAR 

Chapters 3, 6, and 9, and approved by the NRC.

The NRC staff reviewed TVA’s proposed exceptions from NFPA 72D and NFPA 72E, and 

determined that they will not affect the performance of the affected systems or the ability of the 

plant to achieve and maintain safe shutdown. Therefore, the NRC considers the exceptions 

acceptable. The NRC staff concludes that TVA’s design criteria and bases for the installed 

systems are consistent with Position E.1 of Appendix A to BTP (APCSB) 9.5-1 and, therefore, 

are acceptable.

In all cases, TVA stated that an adequate level of protection is provided via a combination of 

limited combustible materials, administrative controls, fire rated barriers, spatial separation, and 

active fire protection systems. Where exceptions or alternatives to NRC staff guidance, rules, 

or design standards exist, TVA stated that it has performed evaluations to ensure that an 

adequate level of protection is provided. The NRC staff reviewed TVA’s approach to the use of 

fire detection systems and concludes that TVA’s design criteria and bases are consistent with 

Position E.1 of Appendix A to BTP (APCSB) 9.5-1 and the defense-in-depth concept described 

in 10 CFR Part 50, Appendix R, and, therefore, are acceptable.
FF-54

5.0 FIRE PROTECTION FOR SPECIFIC PLANT AREAS AND 

HAZARDS

The NRC staff evaluated TVA’s fire protection program for compliance with 10 CFR 50.48(a), as 

described in Section 2.1 of this evaluation. TVA evaluated their program against the guidance 

in Appendix A to Branch Technical Position (BTP) (Auxiliary Power Conversion Systems Branch 

(APCSB)) 9.5-1. The NRC staff determined that TVA’s reliance on the information in Appendix 

A to BTP (APCSB) 9.5-1, was acceptable since it was the guidance in place on April 18, 1977, 

when TVA first submitted a fire hazards analysis to the NRC for review. In addition to TVA’s 

evaluation to Appendix A to BTP (APCSB) 9.5-1, since TVA used guidance published prior to 

1981, TVA also committed to meet Sections III.G, III.J, and III.O, of 10 CFR Part 50, Appendix 

R, to assure safe shutdown capability. This section of the evaluation describes the NRC staff’s 

evaluation of TVA’s compliance with portions of 10 CFR 50.48(a)(1)(iv) and 10 CFR 

50.48(a)(2)(iii). As discussed below, the NRC staff’s review found that the Watts Bar fire 

protection program meets the requirements of 10 CFR 50.48(a).

5.1 Containment

Appendix A to Branch Technical Position (BTP) (Auxiliary Power Conversion Systems Branch 

(APCSB)) 9.5-1 includes guidance for fire protection in containment. In its letter dated 

May 26, 1995 (Agencywide Documents Access and Management System (ADAMS) Accession 

No. ML073230888), Tennessee Valley Authority (TVA) stated that a major fire hazard within 

containment is the lube oil in the reactor coolant pumps (RCPs). If oil leaks from the RCPs, an 

oil collection system is available to collect the oil for each RCP as described below. This 

system on each RCP is designed to collect oil from all potential leakage locations, including the 

RCP oil lift pump, system piping, overflow lines, the lube oil cooler, oil fill and drain lines, flanged 

connections on the oil lines, and the lube oil reservoirs.

The RCPs, lubricating oil systems, oil spray shields, oil collection basins, drain piping, and 

containment sump are designed to seismic Category I requirements so that they will not fail 

during a safe shutdown earthquake.

Each of the four RCPs is protected by an automatic fire suppression and detection system. A 

heat collection hood is installed directly above the RCP motors. In the event of an RCP motor 

fire, the heat collection hood acts as a ceiling, that forces the heat to stall around the detectors 

and the suppression nozzles, thus reducing the response time of these fire protection devices.

Section 6.1.8 of this evaluation provides the U.S. Nuclear Regulatory Commission (NRC) staff’s 

evaluation and approval of the RCP oil collection system configuration and associated fire 

protection features.

TVA stated that areas of divisional interaction within the annulus areas are protected by 

automatic fixed water-spray systems and photoelectric smoke detectors. Ionization duct 

detectors are provided for each lower containment cooling unit and each upper compartment 

cooling unit. In addition, ionization smoke detectors are provided for the exhaust ducts serving 

the containment purge and air exhaust systems and the emergency gas treatment system.

TVA stated that a standpipe and hose system is provided in each containment to complement 

the installed automatic suppression systems. The standpipe systems are normally dry and 

admit water when a remote control device installed at each hose station is manually operated.
FF-55

TVA stated that radiant energy shields (RESs) are relied on to separate cables and associated 

nonsafety circuits of redundant trains. TVA evaluated the combustibility of older RES materials 

in Fire Protection Report (FPR) Part VII, Section 2.2, “Non-Combustible Radiant Energy 

Shields.” Section 6.1.2 of this evaluation provides the NRC staff’s evaluation of this 

configuration. In FPR Part II, Section 12.10.2, TVA stated that the RESs installed in the Watts 

Bar Nuclear Plant (WBN), Unit 2, annulus and in the installations added to the WBN, Unit 1,

annulus during the 2012 refueling outage are constructed from a compliant material. TVA also 

stated that there are no RESs in the WBN, Unit 2, primary containment.

TVA stated that the RCP oil collection system complies with Section III.O, “Oil collection system 

for reactor coolant pump,” of Appendix R to Title 10 of the Code of Federal Regulations Part 50 

with the exception of an alternative to allow for minor amounts of oil that become entrained in 

the ventilation air to escape the oil collection system. TVA evaluated the RCP oil collection 

system in FRP Part VII, Section 2.8, “Reactor Coolant Pump Oil Collection System.” See 

Section 6.1.8 of this evaluation for a detailed evaluation and approval of the configuration.

Based on its review of the information provided by TVA the NRC staff concludes that the fire 

protection features for containment conform to the guidance in Position F.1 of Appendix A to 

BTP (APCSB) 9.5-1 and, therefore, are acceptable.

5.2 Control Room Complex

5.2.1 Control Room 

Appendix A to BTP (APCSB) 9.5-1 includes guidance for fire protection in the main control room 

(MCR). The MCR is common to both units and contains circuits for safe shutdown for fires 

outside of the control building. TVA designated the control building, which contains the MCR, 

an alternative shutdown area. As a result, independent alternative shutdown capability has 

been provided for this area. Discussion of alternative shutdown is located in FPR Part IV and 

Section 3.3 of this evaluation. The entire control building is considered a single fire area and is 

separated from other fire areas (e.g., the auxiliary building, turbine building) by 3-hour fire 

barriers, as documented in FPR Part VI.

In FPR Part VII, Section 2.6.4, TVA evaluated the effect of nonrated metal hatch covers 

between the mechanical equipment rooms and the turbine building. Section 6.2.7.4 of this 

evaluation provides the NRC staff’s evaluation and approval of this configuration.

FPR Part VIII summarizes the fire barriers that separate the MCR from the balance of the 

control building. The MCR is separated from adjacent rooms on the same elevation in the 

control building by 1-hour rated fire barriers. These barriers contain two hollow metal security 

doors that are similar in construction to 1-1/2-hour rated fire doors, although they are not 

specifically fire-rated. The doors between the control room and the turbine building and the 

control room and auxiliary building are 3-hour fire-rated doors. The MCR and the cable 

spreading room are not separated by a rated fire barrier.

FPR Part VIII describes the use of cables in the MCR. TVA stated that (1) wiring for lighting 

terminates in the lighting fixtures, (2) instrumentation and control wiring that enters the control 

room terminates inside panels or control boards, and (3) cables are not routed through the 

control room from one area to another area.
FF-56

In FPR Part VIII, TVA described manual firefighting operations. TVA stated that fire 

extinguishers are provided in the MCR. Standpipe and hose stations are located in the 

stairwells at each end of the MCR. TVA also stated that the hose stations have electrically 

qualified nozzles in alignment with the expected hazards.

TVA stated that ionization smoke detectors are provided in selected cabinets and smoke 

detection is installed in the MCR ventilation system intakes. TVA further stated that fire alarms 

in other parts of the plant, as well as the MCR, alarm and annunciate in a constantly attended 

location in the MCR.

FPR Part VIII also summarizes smoke control features for the MCR. The MCR ventilation air 

intakes are provided with remotely controlled dampers to prevent smoke from entering the 

control room. Manual venting of the control room can be achieved by using portable smoke 

ejectors available on site and by opening the doors of the MCR. TVA also stated that breathing 

apparatuses are available for the control room NRC staff.

TVA evaluated the impact of not providing an automatic suppression system (as required for 

alternative shutdown locations) in the MCR and corridor in FPR Part VII, Section 2.3. 

Section 6.1.3 of this evaluation provides the NRC staff’s evaluation and approval of this 

configuration.

In FPR Part VII, Section 4.1, TVA evaluated MCR doors C49 and C50 for altering the doors by 

adding signs and security hardware or by repairing onsite damage. Section 6.2.2 of this 

evaluation provides the NRC staff’s evaluation and approval of this configuration.

Based its review of the information submitted by TVA, the NRC staff concludes that an 

equivalent level of safety to the separation guidance in Position F.2 of Appendix A to BTP 

(APCSB) 9.5-1 has been achieved by TVA because of (1) the installed detection and 

suppression in the cable spreading room, (2) the low combustible loading and installed 

automatic suppression and detection in adjacent non-MCR control building areas, (3) the 

provision for alternative shutdown for control building fires through use of the independent 

auxiliary control room (ACR) complex, and (4) the fire safe shutdown (FSSD) evaluation that 

demonstrates the use of the ACR to achieve post-FSSD, and therefore, is acceptable.

5.2.2 Auxiliary Control Room 

TVA designated the control building as an alternative shutdown area. FSSD activities take 

place outside of the control building for large or damaging fires in the control building. The ACR 

at WBN provides independent alternative shutdown capability for control building fires. 

Discussion of alternative shutdown is located in FPR Part IV and Section 3.3 of this evaluation.

TVA stated that the ACR is independent from the control building, which includes the cable 

spreading room, MCR, and auxiliary instrument room. The ACR is located in the auxiliary 

building, and is divided into five independent, dedicated rooms. Each room is separated from 

the others and from the rest of the auxiliary building by at least 2-hour rated fire barriers and 

from the control building by 3-hour rated fire barriers. The five independent rooms consist of a 

Train A and a Train B transfer switch room for each unit and a common ACR containing multiple 

instrumentation and control panels for both units. Ionization smoke detectors and preaction 

sprinkler systems are provided in each of the five rooms. Standpipe and hose stations are 
FF-57

provided for manual firefighting activities in the ACR complex from adjacent Rooms 757.0-A2 

and -A24.

In FPR Part IV, TVA described the ACR as designed to control the FSSD activities after control 

has been established at the ACR following MCR abandonment. Systems requiring operator 

manipulations have the controls located in the ACR along with their associated transfer switches 

located in the adjacent transfer switch rooms. TVA stated that operators are periodically trained 

in shutdown procedures from the ACR. TVA further stated that the instruments and controls 

located in the ACR are separated from, or can be electrically isolated from, the corresponding 

instrumentation and controls located in the control building.

In FPR Part VII, Section 2.4, TVA evaluated the effect for intervening combustibles, such as 

insulation on cables in trays and Thermo-Lag® in the ACR. Section 6.1.4 of this evaluation 

provides the NRC staff’s evaluation and approval of this configuration.

Based on its review of the information submitted by TVA, the NRC staff concludes that the 

installed fire protection features are consistent with the NRC’s current guidance in 

Position 6.1.6, “Alternative and Dedicated Shutdown Panels,” in Revision 2 to RG 1.189 and,

therefore, are acceptable.

5.3 Cable Spreading Room

The cable spreading room is common to both units and contains circuits for redundant safe 

shutdown features. TVA designated the control building, which contains the cable spreading 

room, an alternative shutdown area. As a result, independent alternative shutdown capability 

has been provided for this area. Discussion of alternative shutdown is located in Part IV of the 

FPR and Section 3.3 of this evaluation. 

TVA stated that the cable spreading room is separated from the adjacent buildings by 3-hour 

rated barriers. TVA also stated that fire brigade access to the cable spreading room is provided 

by doors from the turbine building and from enclosed stairways within the control building. TVA 

stated that portable extinguishers that are located inside and immediately outside the cable 

spreading room are available. Additionally, standpipe and hose stations are provided from the 

two stairwells and from the turbine building. 

In the FPR Part VIII, TVA summarized the fire protection features for the cable spreading room 

and stated that these features provide full coverage detection and automatic suppression. The 

automatic preaction sprinkler system has a ceiling layer and an intermediate layer of sprinklers 

under the grating and staggered between the upper level heads. TVA further stated that the 

installed cables are designed to allow wetting without faulting.

Based on its review of the information submitted by TVA, the NRC staff concludes that the fire 

protection features for the cable spreading room do not take any exceptions to Position F.3 of 

Appendix A to BTP (APCSB) 9.5-1 and, therefore, are acceptable.

5.4 Switchgear Rooms

TVA stated that the Trains A and B 6.9 kV and 480 V switchgear rooms are located within the 

auxiliary building, but separated from each other and from other rooms within the auxiliary 

building by 2-hour fire rated barriers and from the control building by 3-hour fire rated barriers. 
FF-58

Except as noted below, each room is provided with a full area coverage automatic preaction 

sprinkler system that is actuated by a cross-zoned areawide ionization smoke detection system. 

Water spray shields have been installed to protect safety related electrical equipment against 

the effects of inadvertent or advertent actuation of the automatic suppression system. 

Additionally, standpipe and hose stations are provided for each of the switchgear rooms.

TVA evaluated the impact of the lack of total area suppression in the 480 V Board Rooms 1B 

(Room 772.0 A2; Fire Area 33) and 2B (Room 772.0 A15; Fire Area 45) in FPR Part VII, 

Section 3.1. Section 6.1.7 of this evaluation provides the NRC staff evaluation and approval of 

this configuration.

Based on its review of the information submitted by TVA, the NRC staff concludes that the fire 

protection features for the essential switchgear rooms provide an equivalent level of fire safety 

to Position F.5 of Appendix A to BTP (APCSB) 9.5-1 and, therefore, are acceptable.

5.5 Battery Rooms

TVA stated in FPR Part VIII, that the required Vital Battery Rooms I through IV are separated 

from all other plant areas by 3-hour fire-rated barriers. The Fifth Vital Battery Room is a spare 

that can be used for any of the other four vital batteries. TVA further stated that the Fifth Vital 

Battery Room is separated from other plant areas by 2-hour fire-rated barriers that exceed the 

hazards to which they could be exposed.

TVA also stated that ceiling vents are provided for each battery room with a direct exhaust to 

outside the building to maintain the concentration of hydrogen below 2 percent by volume within 

the battery rooms. Additional details of these exhaust systems are available in WBN, Unit 2,

Final Safety Analysis Report Section 9.4.3.2.5, “Auxiliary Board Rooms Air-Conditioning 

Systems.”

TVA provided a summary of the fire protection features for the battery rooms in FPR Part VIII. 

TVA stated that full coverage automatic smoke detection and manually actuated sprinkler 

system are provided for Vital Battery Rooms I to IV. Smoke detection and an automatic 

preaction sprinkler system are provided for the Fifth Vital Battery Room. With regard to manual 

firefighting, TVA stated that hose stations and portable fire extinguishers are available for fire 

brigade use.

Based on its review of the information submitted by TVA, the NRC staff concludes that the fire 

protection features for the battery rooms do not take exceptions to Position F.7 of Appendix A to 

BTP (APCSB) 9.5-1 and, therefore, are acceptable.

5.6 Turbine Lubrication and Control Oil Storage and Use Areas

TVA stated in FPR Part VI that a fire in the turbine building would not impact equipment 

required to achieve safe shutdown, and that Trains A and B systems and components would be 

utilized without mitigating actions. TVA further stated that cable tray and door penetrations 

through the 3-hour fire rated fire barrier separating the turbine building from the control building 

are sealed with 3-hour fire-rated penetration seals and are provided with automatic water 

curtain protection on the turbine building side. TVA stated in FPR Part VIII that the penetration 

seals are located approximately 5 feet above the elevation of the turbine building floor, limiting 

the access of any turbine oil to the penetration seals. 
FF-59

TVA stated in FPR Part VIII that turbine building oil reservoir hazards are protected by fixed 

water spray systems. Additionally, standpipe and hose stations are provided on each elevation 

of the turbine building.

Based on its review of the information submitted by TVA, the NRC staff concludes that the fire 

protection features for the turbine building provide an equivalent level of safety as the 

guidelines in Position F.8 of Appendix A to BTP (APCSB) 9.5-1 and, therefore, are acceptable.

5.7 Diesel Generator Areas

5.7.1 Diesel Generator Building

In the FPR Part VIII, TVA stated that the diesel generator (DG) building is remotely located and 

is not adjacent to any other safety-related building or structure, and that each DG and its 

associated equipment are separated from each other by 3-hour fire barriers.

TVA described the automatic fire suppression systems installed in these areas as follows. Each 

DG area is provided with full coverage detection that alarms and annunciates in the control 

room and alarms locally. Automatic, total flooding CO2 suppression systems protect each DG, 

the associated day tanks, and the electrical board room. TVA also stated that the DG building 

pipe gallery and corridor are protected by a preaction sprinkler system. For manual 

suppression, TVA stated that standpipes and hose stations are available on both elevations of 

the DG building, with backup from hydrants in the yard.

TVA stated that the two 550-gallon day tanks are located in the same room as the associated 

tandem DG.

TVA evaluated the impact of a 1-1/2-hour rated fire damper in the 3-hour fire-rated floor 

separating the Unit 1 A-A DG exhaust room from the DG building corridor in FPR Part VII, 

Section 4.7. Section 6.2.7.5.1 of this evaluation provides the NRC staff evaluation and approval 

of this configuration.

Based on its review of the information submitted by TVA, the NRC staff concludes that the fire 

protection features for the DG building are consistent with Position F.9 of Appendix A to BTP 

(APCSB) 9.5-1 and, therefore, are acceptable.

5.7.2 FLEX Diesel Generator Rooms

In the FPR Part VIII, TVA stated that two additional DGs are located in separate rooms on the 

roof of the auxiliary building. These two rooms are separated from all other plant areas, and 

each other, by 3-hour fire barriers. These DGs have been installed to mitigate Beyond Design 

Basis Accidents, and are neither safety-related, nor required for safe shutdown.

TVA stated that the FLEX DG rooms are protected by a thermal fire detection system and a 

preaction sprinkler system. For manual suppression, TVA stated that standpipes and hose 

stations are available outside the rooms on the auxiliary building roof.

TVA stated that each of the FLEX DGs has a double-walled 185-gallon fuel tank mounted on 

the same skid as the DG.
FF-60

Based on its review of the information submitted by TVA, the NRC staff concludes that the 

fire protection features for the FLEX DG rooms do not take any exceptions to Position F.9 of 

Appendix A to BTP (APCSB) 9.5-1 and, therefore, are acceptable.

5.8 Diesel Generator Fuel Oil Storage Areas

In FPR Part VIII, TVA stated that the aboveground diesel fuel oil storage tanks are located more 

than 50 feet from any safety-related building or structure, and that they are located within a 

diked area sized to contain leaks or spills of fuel oil.

TVA further stated that the 7-day fuel oil storage tanks for each DG are buried under the floor 

of the DG building. The only portions of the tanks that are not buried are the manway access 

openings to each tank within the diesel rooms and in the common corridor outside the diesel 

rooms. TVA evaluated the impact of these nonrated manway access openings in the FPR 

Part VII, Section 4.4, “Fire Barriers between DG [Diesel Generator] Storage Tank and DG 

Corridor.” Section 6.2.9 of this evaluation provides the NRC staff evaluation and approval of 

this configuration.

TVA evaluated the impact of an untested penetration assembly in the fire barrier between the 

fuel oil transfer pump room and the DG corridor in the FPR Part VII, Section 4.6, “Fire Barriers 

between Fuel Oil Transfer Pump Room and Diesel Generator Building Corridor.” Section 6.2.5 

of this evaluation provides the NRC staff’s evaluation and approval of this configuration.

Based on its review of the information submitted by TVA, the NRC staff concludes that the fire 

protection features for the diesel fuel oil storage areas are consistent with Position F.10 of 

Appendix A to BTP (APCSB) 9.5-1 and, therefore, are acceptable.

5.9 Safety-Related Pump Areas

5.9.1 CCS Pump Area 

As described in TVA’s response dated May 26, 2011 (ADAMS Accession No. ML111520119), to 

request for additional information (RAI) FPR VII-3, the component cooling system (CCS) pumps 

are located in the same fire area in the auxiliary building on elevation 713.0 feet. The 

two Train A CCS pumps are separated from the two Train B pumps, and the spare, by a 

partial-height fire barrier.

TVA evaluated the partial-height fire barrier between the CCS pumps and the ensuing 

redundant train separation issues in FPR Part VII, Section 2.5. Section 6.1.5 of this evaluation 

provides the NRC staff’s evaluation and approval of this configuration.

TVA stated in FPR Part VII, that the area containing the CCS pumps is provided with automatic 

preaction sprinkler system protection at the ceiling and under the grated mezzanine over the 

CCS pumps as well as full coverage automatic smoke detection. Further, in FPR Part VI, TVA 

stated that hose stations are available to support manual firefighting.

Based on its review of the information submitted by TVA, the NRC staff concludes that the fire 

protection features for the CCS pumps are consistent with Position F.11 of Appendix A to BTP 

(APCSB) 9.5-1 and, therefore, are acceptable.
FF-61

5.9.2 Charging Pumps 

As described in the FPR Part VI, each charging pump is located in its own 2-hour fire-rated 

compartment. TVA stated that the pump rooms and the corridor outside these rooms are 

protected by full coverage detection and an automatic preaction sprinkler system. However, 

detection and suppression is not extended into the entrance labyrinth of the charging pump 

rooms. Further, TVA stated that hose stations are located in the corridor leading to these rooms 

and are available to support manual firefighting inside the pump rooms.

TVA evaluated the impact of the lack of total area suppression and detection in the FPR 

Part VII, Section 3.1, “Lack of Total Area Suppression and Detection.” Section 6.1.7 of this 

evaluation provides the NRC staff evaluation and approval of this configuration.

Based on its review of the information submitted by TVA, the NRC staff concludes that the fire 

protection features for the charging pumps provide an equivalent level of safety to Position F.11 

of Appendix A to BTP (APCSB) 9.5-1 and, therefore, are acceptable.

5.9.3 AFW Pumps 

As described in the FPR Part VI, the two turbine-driven auxiliary feedwater (AFW) pumps (one 

for each Unit) are located in the auxiliary building on elevation 692.0 feet. Each pump is located 

in its own 2-hour fire-rated fire compartment. TVA stated that each pump room is provided with 

full coverage automatic detection and an automatic preaction sprinkler system. Further, TVA 

stated that hose stations are located in the corridor leading to these rooms and are available to 

support manual firefighting inside the pump rooms.

As described in TVA’s response dated May 26, 2011 (ADAMS Accession No. ML111520119), to 

RAI FPR VII-3, the motor-driven AFW pumps (two per Unit) are located on opposite ends of the 

auxiliary building on elevation 713.0 feet. TVA stated that there is approximately 126 feet 

separating the Unit 1 and Unit 2 AFW pumps. TVA further stated that the area in which these 

pumps are located is protected by an automatic preaction sprinkler system and that automatic 

fire detection is provided throughout the area. TVA stated in FPR Part VI that hose stations are 

available in the area to support manual firefighting operations.

Based on its review of the information submitted by TVA, the NRC staff concludes that the 

fire protection features provided for the motor- and turbine-driven AFW pumps provide an 

equivalent level of fire safety to Position F.11 of Appendix A to BTP (APCSB) 9.5-1 and, 

therefore, are acceptable.

5.9.4 RHR Pumps 

As described in the FPR Part VI, each residual heat removal (RHR) pump is located in its own 

2-hour fire-rated fire compartment. Each RHR pump room is a separate fire area and none of 

the rooms contain redundant trains of equipment or cables. TVA stated that the corridor outside 

these rooms has full coverage fire detection installed. In each pump room, fire detection is 

installed, except in the entrance labyrinths. TVA stated that the combustible loading in these 

rooms is insignificant, consisting mainly of the lube oil associated with the pump and valve. 

TVA stated that for each fire area, the capability to achieve safe shutdown has been 

demonstrated through analysis. Therefore, a fire in any of these fire areas will not endanger 
FF-62

other safety-related equipment required for safe plant shutdown. Further, TVA stated that hose 

stations are located in the corridor leading to these rooms and are available to support manual 

firefighting inside the individual RHR pump rooms.

Based on its review of the information submitted by TVA, the NRC staff concludes that the fire 

protection features for the RHR pumps provide an equivalent level of fire safety to Position F.11 

of Appendix A to BTP (APCSB) 9.5-1 and, therefore, are acceptable.

5.9.5 ERCW Pumps 

As described in FPR Part VI, the redundant essential raw cooling water (ERCW) pumps are 

separated by 3-hour fire-rated barriers. These pumps are also separated from the traveling 

screen pumps by 3-hour barriers. However, these barriers have open scuppers at the base of 

the wall of the ERCW pump rooms.

TVA stated in FPR Part VI that heat detectors are installed over the ERCW pumps and that no 

redundant FSSD cables or equipment are installed in these areas. Further, TVA stated that 

manual fire suppression capability is available through use of hose stations installed in the 

ERCW strainer room and the screen wash pump room.

TVA evaluated the impact of the open scuppers in the fire barriers that separate the pumps from 

the traveling screens in the FPR Part VII, Section 2.6.2, “Justification for Scupper Openings.” 

Section 6.2.7.2 of this evaluation provides the NRC staff evaluation and approval of this 

configuration.

Based on its review of the information submitted by TVA, the NRC staff concludes that the fire 

protection features for the ERCW pumps provide an equivalent level of fire safety to 

Position F.11 of Appendix A to BTP (APCSB) 9.5-1 and, therefore, are acceptable.

5.10 Other Plant Areas

5.10.1 Areas without Exceptions or Evaluations

The NRC staff reviewed TVA’s compliance with the following positions of Appendix A to 

BTP (APCSB) 9.5-1, as documented in FPR Part VIII:

• Position F.4 – “Plant Computer Room”

• Position F.6 – “Remote Safety-Related Panels”

• Position F.15 – “Decontamination Areas”

• Position F.16 – “Safety-Related Water Tanks”

• Position F.17 – “Cooling Towers”

• Position F.18 – “Miscellaneous Areas”

Based on its review of the information submitted by TVA, the NRC staff concludes that the fire 

protection features provided in these areas provide an equivalent level of fire safety as the 
FF-63

guidance in these sections of Appendix A to BTP (APCSB) 9.5-1, and, therefore, are 

acceptable.

5.10.2 Areas with Exceptions or Evaluations

The NRC staff reviewed TVA’s conformance with the following positions of Appendix A to 

BTP (APCSB) 9.5-1, as documented in FPR Part VIII:

• Position F.12 – “New Fuel Area”

• Position F.13 – “Spent Fuel Pool Area”

• Position F.14 – “Radwaste Building”

TVA evaluated the impact of the lack of installed fire detection for areas related to 

Positions F.12 and F.13 in FPR Part VII, Section 4.5. Section 6.2.1 of this evaluation 

provides the NRC staff’s evaluation and approval of this configuration.

TVA evaluated the impact of 1-1/2-hour rated fire dampers in 3-hour-rated fire barriers for 

areas related to Position F.14 in FPR Part VII, Section 4.7. Section 6.2.7.5.2 of this evaluation 

provides the NRC staff evaluation and approval of this configuration.

Based on its review of the information submitted by TVA, the NRC staff concludes that the fire 

protection features for these areas provide an equivalent level of fire safety to Positions F.12, 

F.13, and F.14 of Appendix A to BTP (APCSB) 9.5-1 and, therefore, are acceptable.

5.11 Specific Hazards

5.11.1 Hydrogen Piping

TVA stated in the FPR that a 1-inch seismically designed hydrogen line is routed through the 

auxiliary building on elevation 713.0 feet to each Unit’s volume control tank. Two isolation 

valves are installed in the hydrogen supply line outside the auxiliary building. These valves 

close automatically when the downstream flow rate reaches 50 standard cubic feet per minute 

(scfm). TVA stated that any hydrogen leakage less than 50 scfm will be diffused and carried 

away by the auxiliary building ventilation system, keeping the hydrogen concentration in any 

given area below the lower explosive limit.

Based on its review of the information submitted by TVA, the NRC staff concludes that the 

hydrogen supply piping in the auxiliary building does not take any exceptions to Position D.2.b 

of Appendix A to BTP (APCSB) 9.5-1 and, therefore, is acceptable.

5.11.2 Transformers Installed Inside Buildings 

TVA stated that transformers located inside of buildings are either dry type or medium voltage 

transformers that contain “high fire point” transformer liquid. The use of dry type transformers is 

consistent with the NRC guidance in Appendix A of BTP 9.5-1, Element D.1.g, but the use of 

transformers with “high fire point” silicone fluid is not included as part of the guidance.
FF-64

In TVA’s response dated August 5, 2011 (ADAMS Accession No. ML11227A257), to 

RAI VIII-21, TVA provided its justification for the use of the “high fire point” silicone fluid in lieu of 

the noncombustible liquid described in Appendix A to BTP 9.5-1. TVA stated that the 

noncombustible transformer liquids contained polychlorinated biphenyl (PCB) fluids. PCB fluids 

are considered noncombustible, but constitute an occupational health and safety, as well as 

environmental, concern if leaked or spilled. Therefore, TVA decided to remove PCB fluids from 

the plant. Although the “high fire point” liquid is considered combustible, it is not considered 

flammable in accordance with the definition of flammable and combustible provided by 

NFPA 30-1973, “Flammable and Combustible Liquids Code.”

TVA stated that all areas where these transformers are located have sprinkler protection. 

Based on the vendor information provided by TVA in Attachment 4 of its letter dated 

September 30, 2011 (ADAMS Accession No. ML13060A225), sprinkler systems are effective 

at extinguishing silicone fluid fires. TVA also considered dikes to contain the volume of the 

silicone fluid if it were to leak from the transformers.

In its response dated September 30, 2011, to RAI VIII-21.1, TVA provided additional information 

regarding the installation of transformers containing “high fire point” silicone fluid. The NRC 

staff questioned the location of these transformers in plant areas that constitute buffer zones 

between analysis volumes, since the transformers were not described as being located in the 

buffer zones. TVA confirmed, in its RAI response, that the transformers are not located in buffer 

zones for large fire areas except for in the electrical equipment room in the intake pumping 

station (IPS).

The transformers in the electrical equipment room in the IPS have dikes, are protected with 

automatic fire suppression systems, and there is 20 feet of separation between the transformers 

and the redundant FSSD train. TVA stated that the 20 feet of separation has intervening 

combustibles, but the combustibles are not continuous. Therefore, in the event that a 

transformer fire were to occur in this area, automatic suppression and spatial separation 

is available to assure that safe shutdown capability is assured.

Based on its review of the information submitted by TVA, the NRC staff concludes that TVA’s 

use of dry type transformers in plant areas is consistent with Appendix A of BTP 9.5-1, 

Element D.1.g, and, therefore, is acceptable. The use of “high fire point” silicone fluid in 

transformers in plant areas is acceptable where the transformers are installed in areas with 

automatic sprinkler systems and spatial separation, either buffer zones or 20 feet without

continuous intervening combustibles, and where transformers have dikes large enough to 

contain the volume of the transformer fluid.
FF-65

6.0 ALTERNATIVES, EXCEPTIONS, AND EVALUATIONS 

The NRC staff evaluated TVA’s fire protection program for compliance with 10 CFR 50.48(a), as 

described in Section 2.1 of this evaluation. TVA evaluated their program against the guidance 

in Appendix A to Branch Technical Position (BTP) (Auxiliary Power Conversion Systems Branch 

(APCSB)) 9.5-1. The NRC staff determined that TVA’s reliance on the information in Appendix 

A to BTP (APCSB) 9.5-1, was acceptable since it was the guidance in place on April 18, 1977, 

when TVA first submitted a fire hazards analysis to the NRC for review. In addition to TVA’s 

evaluation to Appendix A to BTP (APCSB) 9.5-1, since TVA used guidance published prior to 

1981, TVA also committed to meet Sections III.G, III.J, and III.O, of 10 CFR Part 50, Appendix 

R, to assure safe shutdown capability. This section of the evaluation describes the NRC staff’s 

evaluation of TVA’s proposed exceptions and alternatives to the guidance in the BTP and to 

their commitment to the information in 10 CFR Part 50, Appendix R. As discussed below, the 

NRC staff’s review found that the exceptions and alternatives were acceptable.

6.1 Alternatives and Evaluations Related to Criteria in Appendix R to 10 CFR 

Part 50

6.1.1 Alternative – Required Instrumentation for Alternative Shutdown 

TVA committed to maintain safe shutdown capability during and after a fire in accordance with 

Section III.L of Appendix R to Title 10 of the Code of Federal Regulations (10 CFR) Part 50. 

Section III.L.2.d of Appendix R to 10 CFR Part 50 states that the process monitoring function for 

alternative shutdown be capable of providing direct readings of the process variables necessary 

to perform and control a plant cooldown.

TVA has not provided instrumentation in the auxiliary control room (ACR) for (1) tank level 

indication for the condensate storage tank (CST) or the refueling water storage tank (RWST), 

(2) wide-range steam generator (SG) level indication, and (3) cold-leg temperature (TC) 

indication. TVA evaluated these alternatives in FPR Part VII, Section 2.1. TVA’s justification for 

omitting this instrumentation is given below.

The CST level indication is not considered essential in the ACR because automatic switchover 

of the auxiliary feedwater (AFW) pump suction from the CST to the essential raw cooling water 

(ERCW) header is independent of the control building and therefore would be available when 

control is established in the ACR.

The RWST level indication is not considered essential in the ACR because the RWST contains 

almost 20 times the inventory required for cold shutdown. Because the RWST is primarily used 

as makeup water for contraction resulting from cooldown over a period of hours, the excess 

inventory in the RWST is considered sufficient without level indication in the ACR.

Wide-range SG level indication is not provided in the ACR. Instead, the narrow-range SG level 

and AFW flow indication to each SG are provided in the ACR and are sufficient for use in safe 

shutdown procedures whenever the ACR is utilized. This instrumentation also provides input to 

the automatic control utilized to maintain SG level during plant shutdown from the ACR. 

Although wide-range instrumentation is available in the main control room (MCR), no automatic 

control or safety system inputs are derived from this instrumentation. Therefore, the AFW flow 

indication is sufficient for the operator to confirm that adequate post-trip SG inventory is 
FF-66

available in the event that SG level falls below the range of the narrow range indicators that 

are located in the ACR.

Cold-leg temperature indication is not provided in the ACR. Cold-leg temperature is used 

for monitoring natural circulation. Rather than using TC, TVA monitors natural circulation by 

inferring TSAT, the saturation temperature corresponding to the secondary-side SG pressure. In 

the natural circulation mode of operation, the difference between the hot-leg and cold-leg 

temperature (TH−TC) provides an effective indication of when natural circulation is established 

and whether it is being maintained. TSAT will be used to monitor natural circulation in the 

reactor coolant loop in the operating range from full power to the hot standby condition. To 

demonstrate that TSAT will accurately monitor natural circulation in the operating range from hot 

standby to cold shutdown, TVA analyzed the correlation between TSAT and TC while a reactor 

was brought to cold-shutdown condition.

TVA stated that the Westinghouse Owners Group document “Emergency Response Guidelines, 

Generic Issue on Natural Circulation,” Revision 1, provides specific guidelines on how an 

operator can verify that natural circulation has been established without TC being available. The 

Westinghouse Owners Group recommends the use of the following criteria for verifying natural 

circulation: (1) reactor coolant system is subcooling (conversion of pressurizer pressure to TSAT

and subtracting TH), (2) TH is stable or decreasing, and (3) SG pressure is stable or decreasing. 

The instrumentation needed to use these methods of verifying natural circulation is available to 

the operator in the ACR. Therefore, the installed indication is sufficient to compensate for the 

lack of TC indication in the ACR.

10 CFR Part 50, Appendix R, Section III.L.2.d states that the process monitoring function shall 

be capable of providing direct readings of the process variables necessary to perform and 

control the reactivity control function, the reactor coolant makeup function, and the reactor heat 

removal function. The direct reading of process monitoring functions are required to ensure that 

during postfire shutdown, the reactor coolant system process variables are maintained within 

those predicted for a loss of normal AC power and that there is no effect on fission product 

boundary integrity (i.e., no fuel clad damage, no rupture of any primary coolant boundary, and 

no rupture of the containment boundary). Although TVA has not provided instrumentation in the 

ACR for (1) tank level indication for the CST or RWST, (2) wide-range SG level indication, and 

(3) cold-leg temperature (TC) indication, the NRC staff concludes that these alternatives are 

acceptable because (1) the CST level indication is not considered essential in the ACR because 

of the automatic switchover of the AFW pump suction from the CST to the ERCW, and this has 

no impact on the performance and control of the reactivity control function, the reactor coolant 

makeup function, and the reactor heat removal function; the RWST level indication is not 

considered essential in the ACR because the RWST contains almost 20 times the inventory 

required for cold shutdown and this has no impact on the performance and control of the 

reactivity control function, the reactor coolant makeup function, and the reactor heat removal 

function, (2) the narrow-range SG level and AFW flow indication to each SG are provided in the 

ACR and are sufficient for use in safe shutdown procedures whenever the ACR is utilized and

this results in no impact on the performance and control of the reactivity control function, the 

reactor coolant makeup function, and the reactor heat removal function, and (3) natural 

circulation is monitored by inferring TSAT, the saturation temperature corresponding to the 

secondary-side SG pressure and in the natural circulation mode of operation, the difference 

between the hot-leg and cold-leg temperature (TH−TC) provides an effective indication of when 

natural circulation is established and whether it is being maintained, and this has no impact on 

the performance and control of the reactivity control function, the reactor coolant makeup 
FF-67

function, and the reactor heat removal function. Therefore, the NRC staff concludes that there 

is reasonable assurance that the reactor coolant system process variables will be maintained 

within those predicted for a loss of normal AC power and that there would be no effect on fission 

product boundary integrity (i.e., no fuel clad damage, no rupture of any primary coolant 

boundary, and no rupture of the containment boundary).

6.1.2 Alternative – Noncombustible Radiant Energy Heat Shields 

TVA committed to maintain safe shutdown capability during and after a fire in accordance with 

Section III.G of Appendix R to 10 CFR Part 50. Section III.G.2.f of Appendix R to 

10 CFR Part 50 states that inside non-inerted containments, separation of cables and 

equipment and associated non-safety circuits of redundant trains by a noncombustible RES is 

an acceptable method of ensuring that a redundant train of equipment and circuits are 

protected from a fire.

The acceptance criteria included in previous revisions to NUREG-0800, “Standard Review Plan 

for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water 

Reactor] Edition,” Chapter 9, “Auxiliary Systems,” Section 9.5.1.1, “Fire Protection Program,” 

have been removed and have been incorporated in Revision 2 of RG 1.189. Section 6.1.1.1, 

“Containment Electrical Separation,” of RG 1.189 states the following:

Inside noninerted containments, one of the fire protection means specified in 

Regulatory Position 5.3.1.1, or one of the following, should be provided:

a. separation of cables and equipment and associated nonsafety circuits of 

redundant trains by a horizontal distance of more than 6.1 m [meters] (20 

feet) with no intervening combustibles or fire hazards,

b. installation of fire detectors and an automatic fire suppression system in 

the fire area, or

c. separation of cables and equipment and associated nonsafety circuits of 

redundant trains by a noncombustible radiant energy shield [RES] having 

a minimum fire rating of 30 minutes, as demonstrated by testing or 

analysis.

The version of the Standard Review Plan at the time of publication of Generic Letter (GL) 86-10

was Branch Technical Position (BTP) Chemical Engineering Branch (CMEB) 9.5-1.

Section 3.7.1, to GL 86-10 states the following:

The guidelines in BTP CMEB 9.5-1, Section C.7.a.(1)b. indicate that these 

shields should have a fire rating of 1⁄2 hour. In our opinion any material with a 

1⁄2 hour fire rating should be capable of performing the required function.

TVA evaluated this alternative in FPR Part VII, Section 2.2. The RESs installed inside the 

WBN, Unit 1, reactor building are constructed of Minnesota Mining and Manufacturing 

(3M) M20C in the WBN, Unit 1, primary containment, and primarily of 3M M20A in the WBN, 

Unit 1, annulus (there are also compliant RESs constructed of 3M E54C in the Unit 1 

containment). TVA stated that site calculations EPM-BFS-041895 and EPM-BFS-053195 
FF-68

provide the design basis for the number of layers of M20A and M20C required to provide 

approximately 1⁄2 hour RESs for electrical raceways containing circuits required for fire safe 

shutdown (FSSD). These calculations were based on fire tests performed by 3M to 

Underwriters Laboratories, Inc. (UL), Subject 1724, “Fire Tests for Electrical Circuit Protective 

Systems.” The fire exposure used in the tests is the standard time-temperature curve from 

American Society for Testing and Materials (ASTM) E119.

TVA had a series of fire resistance tests performed on the material at Omega Point Laboratories 

for combustibility of the installed materials. The 3M M20A and M20C did not meet the criteria 

for noncombustibility per ASTM E136, “Standard Test Method for Behavior of Materials in a 

Vertical Tube Furnace at 750 ° C.” Additional fire tests to the criteria in ASTM E1354, “Standard 

Test Method for Heat and Visible Smoke Release Rates for Materials and Products Using an 

Oxygen Consumption Calorimeter,” were performed with various RES materials. The results 

indicated that the peak heat release rate (HRR) and the total heat release rate (THR) for the 

3M M20A and M20C was lower than that of Marinite board. Since Marinite board is accepted in 

GL 86-10 as an acceptable RES material, and the 3M materials used at WBN have lower HRR 

and THR than Marinite board, the 3M materials are also considered sufficiently noncombustible 

for use as RES.

10 CFR Part 50, Appendix R, Section III.G.2.f states that inside non-inerted containments, 

separation of cables and equipment and associated non-safety circuits of redundant trains by a 

noncombustible RES is an acceptable method of ensuring that a redundant train of equipment 

and circuits are protected from a fire. This separation of cables and equipment is required in 

order to limit fire damage to one train of systems necessary to achieve and maintain hot 

shutdown conditions and to ensure that systems necessary to achieve and maintain cold 

shutdown can be repaired within 72 hours. Although the 3M M20A and M20C RES used by 

TVA was tested and determined to not meet the criteria for non-combustibility, the NRC staff 

concludes that the RES constructed of 3M M20A used in WBN, Unit 1 annulus, and 3M M20C 

used in WBN, Unit 1 Containment are acceptable because the licensee demonstrated that 

these RESs have a minimum approximate fire rating of 30 minutes in accordance with RG 

1.189, Position 5.3.1.1, and because testing indicated that these RESs have a lower heat 

release rate and total heat release than materials that are considered acceptable RES per GL 

86-10. Therefore, the NRC staff concludes that there is reasonable assurance that fire damage 

would be limited to one train of systems necessary to achieve and maintain hot shutdown 

conditions and that systems necessary to achieve and maintain cold shutdown would be able to 

be repaired within 72 hours.

6.1.3 Alternative – Lack of Automatic Fire Suppression in Alternative Shutdown 

Locations 

TVA committed to maintain safe shutdown capability during and after a fire in accordance with 

Section III.G of Appendix R to 10 CFR Part 50. Section III.G.3 of Appendix R to 

10 CFR Part 50 states that fire detection and a fixed fire suppression system shall be installed 

in the areas, rooms, or zones requiring alternative or dedicated shutdown capability.

TVA requested an alternative to Appendix R for a number of control building rooms that lack 

fixed fire suppression, and some rooms that also lack fire detection. 

The control building is separated from the ACR and adjacent plant areas by equivalent 3-hour 

fire rated barriers except for the equipment hatch in the ceiling separating the control building 
FF-69

from the turbine building. The justification for the hatch opening through the ceilings of 

Rooms 692.0-C1 and 692.0-C10 to the turbine building is evaluated in Section 6.2.7.4 of this 

evaluation. The turbine building is separated from the ACR and adjacent plant areas by 

equivalent 3-hour fire rated barriers. This separation provides assurance that safe shutdown 

capability is assured for a fire in the control building.

All the control building rooms that lack fixed fire suppression have limited ignition sources and 

low or insignificant combustible loading. In addition, all of the rooms have standpipes and hose 

stations available for manual firefighting. Only a few rooms lack full area detection. These 

rooms are stairwells, shower rooms, the telephone room, and the space above the living area 

on the 755.0 foot elevation. Frequent use of the stairwells would lead to discovery of a fire in its 

early stages and would also reduce the likelihood that combustibles could accumulate there. 

The other rooms all are described as having negligible combustible loading.

10 CFR Part 50, Appendix R, Section III.G states that fire detection and a fixed fire suppression 

system shall be installed in the areas, rooms, or zones requiring alternative or dedicated 

shutdown capability. These systems are required to limit fire damage to one train of systems 

necessary to achieve and maintain hot shutdown conditions and to ensure that systems 

necessary to achieve and maintain cold shutdown can be repaired within 72 hours. Although 

TVA is proposing that certain control building rooms continue to not have fire detection and/or 

fixed fire suppression systems, the NRC staff concludes that this is acceptable because these 

rooms have low or insignificant combustible loading, have limited ignition sources, and have 

standpipes and hose stations for manual firefighting capability. Therefore, the NRC staff 

concludes that there is reasonable assurance that fire damage would be limited to one train of 

systems necessary to achieve and maintain hot shutdown conditions and that systems 

necessary to achieve and maintain cold shutdown would be able to be repaired within 72 hours.

6.1.4 Alternative – Intervening Combustibles 

TVA committed to maintain safe shutdown capability during and after a fire in accordance with 

Section III.G of Appendix R to 10 CFR Part 50. Section III.G.2.b of Appendix R to 

10 CFR Part 50 states that separation of redundant trains of safe-shutdown cables and 

equipment by a horizontal distance of more than 20 feet with no intervening combustibles. In 

addition, fire detection and an automatic fire suppression system shall be installed in the area.

In FPR Part VII, Section 2.4, TVA requested an alternative to Section III.G of Appendix R to 

10 CFR Part 50 for 20 feet horizontal distance with no intervening combustibles for safe

shutdown components and cables in the auxiliary building and the intake pumping station (IPS)

electrical equipment room. WBN stated that safe-shutdown components in the auxiliary building 

and IPS electric equipment room are in compliance with Section III.G.2.b of Appendix R to 

10 CFR Part 50, except that intervening combustibles are located between the redundant 

components.

The intervening combustibles in the auxiliary building are primarily in the form of insulation on 

cables in open ladder type cable trays and Thermo-Lag fire barrier material. The remaining 

in situ combustible loading consists of plastics in electrical panels, junction boxes, etc., and 

lubricating oil in pumps, motors, and valves. The intervening combustibles in the IPS electric 

equipment room are mainly in the form of insulation on cables in open ladder type cable trays 

and transformer silicone liquid. The remaining in situ combustible loading consists of 

lubricating oil in small pumps and plastics associated with electrical panels, junction boxes, etc. 
FF-70

Discussion of the nature of the transformer silicon liquid can be found in Section 5.11.2 of this 

evaluation. 

The presence of these intervening combustibles is a concern because they add to a fire’s 

intensity at the ceiling and they could serve as a path for fire propagation between the 

redundant safe-shutdown trains.

For intervening combustibles in the auxiliary building, TVA stated that existing sprinkler heads, 

which are capable of fully developing spray patterns at the ceiling, provide acceptable floor 

coverage if there are no intermediate obstructions in their patterns, which are greater than 

48 inches wide. Additional intermediate sprinklers are provided for 48 inch wide obstructions 

and for combinations of obstructions that, when overlapped, constitute a 48 inch wide 

obstruction, that overlap or combinations of obstructions have less than a 4 inch flue space 

between them when viewed from immediately below. No combination of obstructions may 

traverse the 4 inch flue space and block more than 2 feet of any 8 feet of flue space. To 

mitigate the effects of an exposure fire from transient combustibles at the floor level, TVA stated 

that floor level sprinkler coverage is provided under intermediate obstructions for up to a 30 foot 

wide path where spatially separated redundant FSSD components exist.

TVA stated that for intervening combustibles in the IPS electrical equipment room, sprinkler 

protection has been provided at the ceiling level. Due to the presence of obstructions such as 

heating, ventilation, and air conditioning (HVAC) ducts, cable trays, pipes, and supports, these 

systems have been upgraded. Sprinkler heads were added to provide full coverage at the 

ceiling level and to compensate for large intermediate-level obstructions. To mitigate the effects 

of an exposure fire from transient combustibles at the floor level, TVA provided floor-level 

sprinkler coverage under intermediate obstructions for up to a 30-foot-wide path for spatially 

separated redundant FSSD components.

TVA concluded that, if a fire were to occur, these sprinkler systems would develop effective 

spray patterns at the ceiling, and the water would cascade down through the cable trays in the 

intervening spaces. The cooling effect of these sprinklers, once actuated, would help cool the 

layer of hot gas at the ceiling, prevent the formation of a high temperature plume, and cool the 

room. The sprinklers under the intermediate level obstructions would actuate to ensure that 

floor level coverage is provided under the obstructions. In addition, the coverage provided by 

the ceiling sprinklers would produce sufficient cooling to reduce the likelihood that fire will 

propagate across the intervening space between the redundant trains.

10 CFR Part 50, Appendix R, Section III.G.2.b states that separation of cables and equipment 

and associated non-safety circuits of redundant trains by a horizontal distance of more than 20 

feet with no intervening combustible or fire hazards and that fire detectors and an automatic fire 

suppression system shall be installed in the fire area. This separation of cables and equipment 

and the installation of fire detectors and an automatic fire suppression system is required in 

order to limit fire damage to one train of systems necessary to achieve and maintain hot 

shutdown conditions and to ensure that systems necessary to achieve and maintain cold 

shutdown can be repaired within 72 hours. Although TVA is proposing to have less than 20 feet 

horizontal distance with no intervening combustibles for safe-shutdown components and cables 

in the auxiliary building and IPS electrical equipment room, the NRC concludes that this is 

acceptable because the intervening combustibles are limited to cable insulation on open ladder 

type cable trays and thermo-lag fire barrier material, and because the coverage provided by the 

ceiling sprinklers would produce sufficient cooling to reduce the likelihood that fire would 
FF-71

propagate across the intervening space between the redundant trains. Therefore, the NRC staff 

concludes that there is reasonable assurance that fire damage would be limited to one train of 

systems necessary to achieve and maintain hot shutdown conditions and that systems 

necessary to achieve and maintain cold shutdown would be able to be repaired within 72 hours. 

6.1.5 Alternative – Partial Fire Wall between CCS Pumps 

TVA committed to maintain safe shutdown capability during and after a fire in accordance with 

Section III.G of Appendix R to 10 CFR Part 50. Section III.G.2.b of Appendix R to 

10 CFR Part 50 states that one compliance strategy is the separation of cables and equipment 

and associated nonsafety circuits of redundant trains by a horizontal distance of more than 

20 feet with no intervening combustibles or fire hazards. In addition, fire detectors and an 

automatic fire suppression system shall be installed in the fire area. Section III.G.2.c states that 

another compliance strategy is the enclosure of cables and equipment and associated 

nonsafety circuits of one redundant train in a fire barrier having a 1-hour rating. In addition, fire 

detectors and an automatic fire suppression system shall be installed in the fire area.

In FPR Part VII, Section 2.5, TVA requested an alternative to Appendix R for redundant 

component cooling system (CCS) pumps that are protected by fire detectors and an automatic 

fire suppression system, but are separated by a partial-height and -width noncombustible wall.

The five CCS pumps are located in Fire Area 8, Room 713.0-A1, in 

subsections 713.0-A1A1, -A1A2 and -A1A3, on elevation 713.0 feet of the auxiliary building. 

The two Train B pumps are separated from both Train A pumps and the spare pump by a 

noncombustible wall which extends 3 feet above the highest point of the pumps. A ceiling-level 

preaction sprinkler system is provided for cable tray and general area coverage. Automatic 

sprinkler coverage has also been provided under the pipe-break barrier for the Unit 1 

motor-driven AFW pumps and under the mezzanine for all five CCS pumps. Cross-zoned 

ionization smoke detectors are provided to actuate the preaction suppression systems and 

give early warning of a fire.

The combustibles in Room 713.0-A1 consist of lube oil in the pumps, motors, and valves; 

plastics associated with the electrical panels, boxes and lights, insulation on cables routed in 

cable trays; and anticipated amounts of radwaste trash and laundry. The fire severity for this 

room is classified as moderately severe. However, TVA stated that approximately 95 percent 

of the in situ combustible loading in this area is due to the insulation on cables routed in cable 

trays and the Thermo-Lag fire barrier material. The majority of the remaining combustible 

loading in the immediate area of the CCS pumps is due to the approximately 6 gallons of lube 

oil associated with each CCS pump and approximately 45 gallons of lube oil associated with 

each of the two Unit 1 AFW pumps. The cables are protected electrically with appropriately 

sized circuit protective devices (breakers and fuses) that will actuate on electrical faults prior 

to the jacket material of faulted cables reaching their auto-ignition temperature. A fire due to 

transient combustibles located near the edge of the partial-height fire barriers would not pose 

a threat to more than one CCS pump due to the lack of combustibles. Additionally, raceways 

containing the redundant cables for the CCS pumps are separated by 20 feet or more or by 

noncombustible barriers.

10 CFR Part 50, Appendix R, Section III.G.2.b states that separation of cables and equipment 

and associated non-safety circuits of redundant trains by a horizontal distance of more than 20 

feet with no intervening combustible or fire hazards and that fire detectors and an automatic fire 
FF-72

suppression system shall be installed in the fire area. 10 CFR Part 50, Appendix R, 

Section III.G.2.c states that enclosure of cables and equipment and associated non-safety 

circuits of one redundant train in a fire barrier having a 1-hour rating and that fire detectors and 

an automatic fire suppression system shall be installed in the fire area. This separation of 

cables and equipment and the installation of fire detectors and an automatic fire suppression 

system is required in order to limit fire damage to one train of systems necessary to achieve and 

maintain hot shutdown conditions and to ensure that systems necessary to achieve and 

maintain cold shutdown can be repaired within 72 hours. Although TVA is proposing the use of 

a partial-height and -width noncombustible wall as the fire barrier for redundant CCS pumps, the 

NRC staff concludes that its use is acceptable because a fire would be detected and 

suppression would begin prior to it becoming a threat to the redundant pumps on the other side 

of the noncombustible barrier, and until the fire is suppressed, the noncombustible barrier would 

shield the pumps from radiant heat on one side and from fire on the other side. Therefore, the 

NRC staff concludes that there is reasonable assurance that fire damage would be limited to 

one train of systems necessary to achieve and maintain hot shutdown conditions and that 

systems necessary to achieve and maintain cold shutdown would be able to be repaired within 

72 hours.

6.1.6 Alternative – Emergency Lighting 

TVA committed to provide emergency lighting to assure safe shutdown capability is maintained 

during and after a fire in accordance with Section III.J of Appendix R to 10 CFR Part 50. 

Section III.J of Appendix R to 10 CFR Part 50 states that emergency lighting units with at least 

an 8-hour battery power supply be provided in all areas needed for operation of safe-shutdown 

equipment and for necessary access and egress routes. In FPR Part VII, Section 2.7, TVA 

requested an alternative for emergency lighting in each containment, the turbine building, the 

yard, and areas impacted by fire where operator manual actions (OMAs) are performed in the 

impacted area (“post-fire” areas).

Dedicated and maintained portable battery powered lights are used in lieu of installed battery 

pack lighting units in both containments. Emergency diesel generator-backed standby lighting 

is installed and maintained for the turbine building. Security diesel generator (DG)-backed 

standby lighting is installed and maintained for the yard. To perform actions the nuclear 

assistant unit operators (NAUOs) carry portable lighting to supplement yard and turbine 

building diesel-backed lighting systems to provide additional task lighting capability.

Based on its review of the information submitted by TVA, the NRC staff concludes that the use 

of installed standby lighting and hand-held portable lighting units for the yard and turbine 

building is an acceptable alternative to the lighting criteria required by Section III.J, of 

Appendix R to 10 CFR Part 50, and, therefore, is acceptable.

OMAs requiring entry into primary containment would only result from fire damage to the 

residual heat removal (RHR) isolation valves or cables near the valves which are located inside 

lower containment. The OMAs to align the RHR isolation valves may be performed anytime 

within 2 hours after reactor trip. This allows ample time to extinguish the fire, obtain the portable 

lanterns, and operate the valves. As described above, WBN has dedicated hand-held portable 

lighting units for use in supporting manual firefighting and safe shutdown OMAs for fires in the 

lower containment, in addition to the portable lighting normally used by NAUOs.
FF-73

A fire affecting the RHR isolation valves could damage lighting circuits in the immediate vicinity, 

but would not be expected to disable all lower containment lighting, since different circuits are 

used at each elevation. Additionally, two standby lighting circuits, with fixtures strategically 

located throughout lower containment, provide lighting in case of fire damage to the normal 

lighting cabinet.

TVA’s concerns regarding the installation of 8-hour emergency lighting units inside containment 

include the reduced life of the batteries in the high temperature and humidity environment 

experienced inside the primary containment. Also, ALARA concerns would limit testing and 

maintenance to reactor outages, since access into the primary containment during plant 

operations is restricted.

Based on its review of the information submitted by TVA, the NRC staff concludes that, based 

on the complications of testing and maintaining 8-hour fixed emergency lighting units, and 

TVA’s design description of the installed lighting in the lower containment complemented by the 

dedicated hand-held portable lighting units, the installation of 8-hour emergency lighting units is 

unnecessary to provide access and egress to the manual action sites and perform safe 

shutdown actions in primary containment. Therefore, the use of installed lighting and hand-held 

portable lighting units for this area is an acceptable alternative to the lighting criteria required by 

Section III.J of Appendix R to 10 CFR Part 50.

For OMAs for post-fire areas there is emergency lighting installed on the route to the area. The 

operators have access to dedicated, hand-held portable lanterns that are maintained for OMAs

in these areas. In addition, OMAs are not needed for 60 minutes following the fire, which will 

allow for time for the operators to arrive and prepare to perform the OMA (see Sections 3.5.6 

and 6.1.9 of this evaluation). Portable lanterns stored outside of the post-fire area would not be 

impacted by fire, whereas installed emergency lighting units within the area may not be 

available due to fire damage. Therefore, the use of dedicated, hand-held portable lanterns for 

post-fire areas is an acceptable alternative to the lighting criteria required by Section III.J of 

Appendix R to 10 CFR Part 50.

10 CFR Part 50, Appendix R, Section III.J states that emergency lighting units with at least an 

8-hour battery power supply be provided in all areas needed for operation of safe-shutdown 

equipment and for necessary access and egress routes. Emergency lighting is required to 

ensure that safe shutdown capability is maintained during and after a fire. Although TVA is 

proposing portable battery powered lights in the containment, emergency diesel powered 

standby lighting in the turbine building, and hand-held portable lights carried by the NAUOs, the 

NRC staff concludes that these alternatives acceptable because they will provide adequate 

lighting to allow OMAs to be performed. Therefore, the NRC staff concludes that there is 

reasonable assurance that the emergency lighting provided will maintain safe shutdown 

capability during and after a fire. 

6.1.7 Evaluation – Lack of Total Area Suppression and Detection

TVA committed to meet Section III.G.2 of Appendix R to 10 CFR Part 50 for hot shutdown 

capability, which states that when redundant trains of cables or equipment necessary for 

post-FSSD are installed in the same fire area, fire detectors and automatic fire suppression 

must be installed, unless one train is protected by a 3-hour-rated fire barrier. Position 5 of the 

Attachment to GL 86-10 states that, to meet Section III.G.2 of Appendix R to 10 CFR Part 50, 
FF-74

less than full area coverage may be adequate to comply with the regulation if the suppression 

and detection installed is sufficient to protect against the hazards of the fire area.

In FPR Part VII, Section 3.1, TVA evaluated portions of fire areas that contain both trains of safe 

shutdown success paths, but do not have full coverage fire detection and suppression installed. 

The WBN plant has some fire areas that include multiple subdivisions, called rooms. These 

rooms may not be separated from the other rooms within the fire area by rated fire barriers.

The NRC staff found that for fire areas composed of multiple rooms, the rooms which contain 

redundant safe shutdown equipment have either 3-hour-rated barriers to protect one train of the 

safe shutdown equipment, or the rooms are equipped with fire detection and automatic 

suppression, and have some spatial separation between trains (see Section 3.2.1 of this 

evaluation). Therefore, these rooms are not considered to be credible exposure hazards to the 

other rooms in the fire area that have redundant safe shutdown equipment.

Some of the rooms contain safe shutdown equipment, but there is not redundant safe shutdown 

equipment required for hot shutdown in the room. In other cases, the safe shutdown equipment 

is needed for cold shutdown or for alternative shutdown. In still other cases, the safe shutdown 

equipment is not used to provide for plant safe shutdown for a fire in the room; that is, it is relied 

upon for a fire elsewhere in the plant. In any of these cases, safe shutdown equipment is 

available outside of the room if there is a fire in the room, and any exposure hazard in the 

room to another room would be mitigated by the protection in the other room.

Based on the information provided by TVA, there are rooms that lack full area fire detection and 

suppression that do not contain redundant safe shutdown equipment needed for hot shutdown 

and do not constitute exposure hazards to other rooms within the fire area. The NRC staff 

reviewed this information and concludes that this is acceptable.

The descriptions in the evaluations state that the plant provided only one train of FSSD 

equipment and cables in Centrifugal Charging Pump (CCP) Rooms 1B-B (Room 692.0-A10; 

Fire Area 6), 2A-A (Room 692.0-A23; Fire Area 68), and 2B-B (Room 692.0-A22; Fire Area 67). 

However, Fire Areas 6, 67, and 68 consist solely of the single CCP room. Because these 

rooms do not contain redundant trains of equipment or cables, the NRC staff did not review 

these evaluations.

Rooms that contain redundant cables or equipment necessary for post-FSSD

480-V Board Rooms 1B (Room 772.0-A2; Fire Area 33) and 2B (Room 772.0-A15; Fire 

Area 45)

In FPR Part VII Section 3.1.8, TVA stated that in 480-V Board Rooms 1B (Room 772.0-A2; Fire 

Area 33) and 2B (Room 772.0-A15; Fire Area 45), preaction sprinkler systems are provided 

throughout both rooms except for the portion of each room that contains one set of vital battery 

inverters and chargers. Additionally, ionization detection is installed throughout both rooms. 

TVA further stated that the redundant inverters and chargers and associated cables are 

separated by a minimum of 42 feet and are located at opposite ends of each room. Additionally, 

TVA stated that other redundant components in the rooms are located within the suppressed 

area of each room and are separated in accordance with Section III.G.2 of Appendix R to 

10 CFR Part 50. A fire in the unsprinklered locations in these rooms would be detected by the 

installed fire detection systems before propagating significantly. If the fire propagated rapidly 
FF-75

before the fire brigade arrived, individual sprinklers in the protected portions of the rooms would 

operate to limit the spread of fire and to protect the redundant systems until the fire was 

controlled and suppressed by the plant fire brigade.

10 CFR Part 50, Appendix R, III.G.2 states that when redundant trains of cables or equipment 

necessary for post FSSD are installed in the same fire area, fire detectors and automatic fire 

suppression must be installed, unless one train is protected by a 3 hour rated fire barrier. This 

separation of cables and equipment and the installation of fire detectors and an automatic fire 

suppression system is required in order to limit fire damage to one train of systems necessary to 

achieve and maintain hot shutdown conditions and to ensure that systems necessary to achieve 

and maintain cold shutdown can be repaired within 72 hours. Although TVA is proposing the 

use of partial preaction sprinkler protection in the 480-V Board Rooms 1B and 2B, the NRC staff 

concludes that this is acceptable because a fire in the unsprinklered locations in these rooms 

would be detected by the fire detection systems before propagating significantly, individual 

sprinklers in the protected portions of the rooms would operate to limit the spread of fire and to 

protect the redundant systems until the fire was controlled and suppressed by the plant fire 

brigade. Therefore, the NRC staff concludes that there is reasonable assurance that fire 

damage would be limited to one train of systems necessary to achieve and maintain hot 

shutdown conditions and that systems necessary to achieve and maintain cold shutdown would 

be able to be repaired within 72 hours.

Unit 1 Penetration Room (Room 692.0-A7; Fire Area 5) and Unit 2 Penetration Room

(Room 692.0-A25; Fire Area 70)

In FPR Part VII Sections 3.1.1.a and 3.1.9.a, TVA stated that in Unit 1 Penetration Room

(Room 692.0-A7; Fire Area 5) and Unit 2 Penetration Room (Room 692.0-A25; Fire Area 70), 

preaction sprinkler systems and ionization smoke detection systems are provided throughout 

both rooms except for the portion of each room that consists of the containment spray and RHR 

valve vault. TVA further stated that the walls, floors and ceilings, of each vault are reinforced 

concrete at least 6 inches thick with a steel liner. These portions of the rooms are not normally 

accessible due to the access opening being covered by a heavy steel hatch. TVA stated that 

the rooms and the vaults have insignificant fire loadings and no credible ignition sources. 

Finally, TVA stated that the vaults have been included in the safe shutdown analysis, and the 

valves would not be needed to achieve and maintain hot shutdown conditions. The Unit 1 area 

needs an 1140 minute required time OMA to achieve cold shutdown, whereas Unit 2 does not 

need an OMA for a fire in this area to achieve cold shutdown.

10 CFR Part 50, Appendix R, III.G.2 states that when redundant trains of cables or equipment 

necessary for post FSSD are installed in the same fire area, fire detectors and automatic fire 

suppression must be installed, unless one train is protected by a 3 hour rated fire barrier. This 

separation of cables and equipment and the installation of fire detectors and an automatic fire 

suppression system is required in order to limit fire damage to one train of systems necessary to 

achieve and maintain hot shutdown conditions and to ensure that systems necessary to achieve 

and maintain cold shutdown can be repaired within 72 hours. Although TVA is proposing the 

use of partial fire detection and suppression protection in the Unit 1 and Unit 2 Penetration 

rooms, the NRC staff concludes that this is acceptable because of the minimal fire loading in the 

rooms, the inaccessibility of the vaults, the lack of exposed combustibles and ignition sources in 

the vaults, the separation of the unprotected locations from the balance of the rooms which 

would prevent fire spread into or out of these areas, and inclusion of these configurations in the 

WBN safe shutdown analysis. Therefore, the NRC staff concludes that there is reasonable 
FF-76

assurance that fire damage would be limited to one train of systems necessary to achieve and 

maintain hot shutdown conditions and that systems necessary to achieve and maintain cold 

shutdown would be able to be repaired within 72 hours.

6.1.8 Evaluation – Reactor Coolant Pump Oil Collection System 

TVA committed to meet Section III.O of Appendix R to 10 CFR Part 50. This section states, 

in part, that reactor coolant pumps (RCPs) be equipped with an oil collection system if the 

containment is not inerted during normal operation and that the system be capable of collecting 

lube oil from all potential pressurized and unpressurized leakage sites in the RCP lube oil 

system.

In FPR Part VII, Section 2.8, TVA stated that the RCP oil collection system must function in an 

area with significant ventilation airflows from both the control rod drive mechanism cooling units 

and the RCP motor itself. A minor leak in the lubrication system that causes oil to drip in an 

area where the ventilation airflow is strong can result in the oil’s becoming entrained in 

ventilation air, which in turn could prevent the leak from ever entering the collection system. 

The need for ventilation around the RCP dictates that some ventilation flow areas must be 

present in areas around the lube oil system and the oil collection system. In designing the 

oil collection system, it is not feasible in all instances to prevent minor amounts of oil from 

becoming entrained in the ventilation air and escaping the collection system. This oil may 

become a thin film on the piping mirror insulation and supports in the vicinity of the RCPs.

TVA described the RCP oil collection systems in a letter dated May 26, 1995 (Agencywide 

Documents Access and Management System (ADAMS) Accession No. ML073230888). TVA 

used the following design criteria as the basis for the oil collection systems.

The oil collection system on each RCP collects oil from all potential leakage locations, including 

the RCP oil lift pump, system piping, overflow lines, the lube oil cooler, oil fill and drain lines, 

flanged connections on the oil lines, and the lube oil reservoirs. Each RCP oil collection system 

consists of spray shields/deflectors, a collection basin, a lift pump collection tray, a lower 

bearing collection tray and drain, drain piping, and a closed, vented container (reactor building 

floor and equipment drain sump).

The drain piping from each RCP’s oil collection basin is directed to a drain header. The drain 

header runs through the shield wall and into the raceway area inside primary containment and 

runs through the floor into the 1600-gallon capacity sump. As required by Appendix R, the 

sump is a closed container and is equipped with a flame arrester on the vent line. Each unit’s 

sump has sufficient capacity to hold the entire RCP oil inventory of all four RCPs.

TVA stated that up to 17 gallons of oil could collect in the lower motor support housing before 

beginning to drain to the collection system. The RCPs are equipped with control loop level 

indication that would initiate an alarm in the MCR if 2 or more gallons of lube oil are lost from the 

RCP. Collection of oil within the lower motor support housing is acceptable since the oil, and 

possible fire, would be contained within the RCP and would not impact surrounding equipment 

such that safe shutdown could be affected. In addition, the RCP is equipped with a water-based 

fire suppression system such that a fire at the RCP would have automatic suppression 

available.
FF-77

The RCP pumps, lubricating oil systems, oil spray shields, oil collection basins, drain piping, and 

containment sumps are designed to seismic Category I requirements so as not to fail during a 

safe-shutdown earthquake.

Each of the four RCPs is protected by a fixed fire suppression and detection system. A heat 

collection hood is installed directly above the RCP motors. Each of the RCPs is protected by a 

separate closed-head preaction automatic water spray system that is installed under this hood. 

Each system has a ring header containing eight nozzles. The header is located approximately 

4 feet above the top of the RCP motor and the nozzles, which actuate at 500 ° F, are oriented 

so as to provide optimum coverage of the RCP motor from above. In addition, there are 

four rate-compensating/fixed-temperature spot-type thermal detectors located above the RCP 

motors on the bottom side of the heat-collection hood. These detectors are Class A supervised, 

have a thermal rating of between 200 ° F and 225 ° F and are alarmed and annunciated in the 

MCR. In the event of a fire, this hood acts as a ceiling, forcing the heat to stall around the 

detectors and the suppression nozzles, thus reducing the response time of these fire 

protection devices.

10 CFR Part 50, Appendix R, Section III.O states, in part, that reactor coolant pumps (RCPs) be 

equipped with an oil collection system (if the containment is not inerted during normal operation)

and that the system be capable of collecting lube oil from all potential pressurized and 

unpressurized leakage sites in the RCP lube oil system. This oil collection system is required to 

minimize the possibility of a severe lube oil fire from occurring during normal or design basis 

accident conditions. Although TVA is proposing to allow collection of oil in the lower motor 

support housing and to allow minor amounts of oil to escape the oil collection system and 

become a thin film on piping mirror insulation and supports in the vicinity of the RCPs, the NRC 

staff concludes that these deviations are acceptable because large leakages would be alarmed 

to the control room and the RCP cubicles are equipped with detection and fixed fire 

suppression. Therefore, the NRC staff concludes that there is reasonable assurance that the oil 

collection system will minimize the possibility of a lube oil fire from occurring during normal or 

design basis accident conditions.

6.1.9 Evaluation – WBN Unit 2 Operator Manual Action Feasibility and Reliability

TVA committed to meet Section III.G of Appendix R to 10 CFR Part 50. Section III.G of 

Appendix R to 10 CFR Part 50 provides a number of acceptable methods of providing 

reasonable assurance that one of the safe-shutdown trains is free of fire damage using a 

combination of physical separation, fire wraps, fire detection, and fire suppression. Section III.G 

of Appendix R to 10 CFR Part 50, does not permit the use of OMAs as a means of assuring that 

a safe-shutdown train is free of fire damage, although NRC approval may be granted for the use 

of OMAs. Discussion of OMAs needed for equipment important for safe shutdown is included in 

Section 3.5.2 of this evaluation. Discussion of the treatment of recall time for auxiliary unit 

operators to perform these actions is discussed in Section 3.5.4 of this evaluation. TVA stated 

that some OMAs in this evaluation serve two roles. These OMAs may address 

important-to-safe-shutdown and required-for-safe-shutdown situations. This section reviews 

only the required-for-safe-shutdown aspects. See Section 3.5 of this evaluation for a discussion 

of the important-to-safe-shutdown acceptance criteria.

TVA developed evaluations based on the information in NUREG-1852, “Demonstrating the 

Feasibility and Reliability of Operator Manual Actions in Response to Fire,” to demonstrate that 

OMAs are capable of accomplishing various safe shutdown functions and terminating spurious 
FF-78

equipment operations that have the potential to interfere with safe shutdown. TVA also 

described the fire protection defense-in-depth features, which include fire prevention, fire 

detection, and fire suppression, within each room that reduce the likelihood that an OMA would 

be needed. The NRC staff reviewed the OMAs for the rooms and zones listed in the table 

below.

The following table describes, by room or zone, the margin available for the OMA and the 

defense-in-depth features. TVA used the information from NUREG-1852 in determining the 

feasibility and reliability of manual actions. The following criteria were used to consider 

feasibility and reliability: (1) Adequate Time Available to Perform Actions and (2) Adequate 

Time Available to Ensure Reliability. Manual actions that had at least 40 minutes of margin 

were considered feasible and reliable. If a postulated fire has been evaluated to have the 

potential to cause a plant trip, 5 minutes has been added to the performance time in the margin 

calculation. Margin in this context is:

Time when − Demonstrated time = Margin

OMA is needed to perform OMA

TVA also considered: (1) Environmental Factors, (2) Equipment Functionality and Accessibility, 

(3) Available Indications, (4) Communications, (5) Portable Equipment, (6) Personnel Protection 

Equipment, and (7) Procedures and Training. TVA stated that the above criteria do not 

adversely affect the performance of the action for the OMAs evaluated in this section. The 

diagnosis time for OMAs is discussed in Section 3.5.4 of this evaluation.

OMAs that have less than 40 minutes of margin have been identified in the table below. Any 

OMA with less than 40 minutes of margin references a note that includes an evaluation of the 

time margin and defense-in-depth features.

Fire protection defense-in-depth features, such as fire prevention, fire detection, and fire 

suppression, apply to each of these rooms. Most rooms have full area fire detection and 

automatic fire suppression. For those rooms that lack full fire detection and automatic fire 

suppression, the NRC staff evaluated these areas specifically below. This evaluation uses the 

available defense-in-depth information and information about the OMA to determine if having 

less than full area suppression and detection is acceptable. Using full area suppression and 

automatic detection as criteria for OMAs is not intended to imply that they are required; rather, 

the NRC staff deemed full detection and automatic suppression as a robust level of protection. 

Less than full detection and automatic suppression received a more detailed review by the 

NRC staff.
FF-79

The table below is an analysis of OMAs based on the room or zone of fire origin. OMAs that 

have been evaluated by TVA to meet NUREG-1852 and have at least 40 minutes of margin are 

considered to be acceptable without additional evaluation, if the postulated fire occurs in a room 

equipped with fire detection and automatic suppression. Exceptions to these factors are 

discussed in notes that refer to information explained below the table.

Part VII 

Evaluation

Room/ 

Zone

Exceptions to Full 

Area Detection and 

Automatic 

Suppression 

OMA # Exceptions to 40 

Minutes of 

Margin

Notes

8.3.1 692.0-A1B None 1159 & 

1160

OMA involves 

re-entry 

Note 1

8.3.2 692.0-A22 No detection or 

suppression in the 

entrance labyrinth

1159 & 

1160

None Note 2

8.3.3 692.0-A25 No detection or 

suppression in the 

valve vault

1065 None Note 2

1066 OMA involves reentryNote 1

8.3.4 713.0-A1A No suppression over 

the boric acid tanks

1022 None Note 2

1023 None Note 2

8.3.5 713.0-A1B No suppression over 

the boric acid tanks

1016 None Note 2

1023 None Note 2

1615 & 

1275

Less than 40 

minutes of margin

Note 3

8.3.6 713.0-A27 None 1022 None

1023 None

8.3.7 2ANN None 1023 None

8.3.8 737.0-A1A None 1022 None

1023 None

8.3.9 737.0-A1B None 1016 None

1024 None

1540 &

1542

Less than 40 

minutes of margin

Note 4

8.3.10 737.0-A1C None 1023 None

8.3.11 737.0-A1N No total area 

suppression and 

detection in the air 

locks

1016 None Note 5

1024 None Note 5

1535 & 

1536

Less than 40 

minutes of margin

Note 6

8.3.12 737.0-A5M None 1023 None

8.3.13 737.0-A5N None 1023 None

8.3.14 737.0-A5S None 1022 None

1023 None

8.3.15 737.0-A9M None 1022 None
FF-80

Part VII 

Evaluation

Room/ 

Zone

Exceptions to Full 

Area Detection and 

Automatic 

Suppression 

OMA # Exceptions to 40 

Minutes of 

Margin

Notes

1023 None

8.3.16 737.0-A9N None 1022 None

1023 None

8.3.17 737.0-A9S None 1022 None

1023 None

8.3.18 737.0-A12 None 1024 None

8.3.19 757.0-A1 None 1016 None

1024 None

8.3.20 757.0-A2 None 1185 None Note 7

1184 None Note 7

1712 None

1713 None

1577 None

8.3.21 757.0-A3 Fire sprinkler system 

is manual, not 

automatic

1024 None Note 8

8.3.22 757.0-A4 Fire sprinkler system 

is manual, not 

automatic

1022 None Note 8

1023 None Note 8

8.3.23 757.0-A5 None 1016 None

1024 None

8.3.24 757.0-A9 None 1185 None Note 7

1184 None Note 7

1712 None

1713 None

1577 None

8.3.25 757.0-A10 None 1016 None

1024 None

8.3.26 757.0-A16 None 1022 None

1023 None

8.3.27 757.0-A17 None 1037 None Note 9

1038 None Note 9

8.3.28 757.0-A21 None 1023 None

1516 Less than 40 

minutes of margin

Note 10

8.3.29 757.0-A22 Fire sprinkler system 

is manual not 

automatic

1016 None Note 8

1024 None Note 8

8.3.30 757.0-A23 Fire sprinkler system 

is manual not 

automatic

1022 None Note 8

1023 None Note 8

8.3.31 757.0-A24 None 1037 None Note 9

1038 None Note 9

8.3.32 757.0-A26 None 1024 None

8.3.33 757.0-A27 None 1022 None
FF-81

Part VII 

Evaluation

Room/ 

Zone

Exceptions to Full 

Area Detection and 

Automatic 

Suppression 

OMA # Exceptions to 40 

Minutes of 

Margin

Notes

1023 None

8.3.34 757.0-A28 None 1016 None

1024 None

8.3.35 772.0-A1 None 1022 None

1023 None

8.3.36 772.0-A2 

East

No suppression 

between column lines 

A6-A8/Q-R 

1016 None Note 11

1024 None Note 11

8.3.37 772.0-A2 

West

No suppression 

between column lines 

A6-A8/Q-R 

1022 None Note 11

1023 None Note 11

8.3.38 772.0-A4 Fire sprinkler system 

is manual not 

automatic

1022 None Note 8

1023 None Note 8

8.3.39 772.0-A5 None 1016 None

8.3.40 772.0-A6 None 1023 None

8.3.41 772.0-A8 None 1016 None

1022 None

1023 None

1024 None

8.3.42 772.0-A9 None 1022 None

1023 None

8.3.43 772.0-A10 None 1022 None

1023 None

8.3.44 772.0-A11 None 1024 None

8.3.45 772.0-A12 None 1023 None

8.3.46 772.0-A13 Fire sprinkler system 

is manual not 

automatic

1016 None Note 8

1024 None Note 8

8.3.47 772.0-A14 Fire sprinkler system 

is manual not 

automatic

1022 None Note 8

1023 None Note 8

1148 None Note 8

8.3.48 772.0-A15 

East

No suppression 

between column lines 

A8-A10/Q-R

1016 None Note 11

1024 None Note 11

8.3.49 772.0-A15 

West

No suppression 

between column lines 

A8-A10/Q-R

1016 None Note 11

1024 None Note 11

1495 Less than 40 

minutes of margin

Note 12

558 & 

622

None Note 11

1665 & 

1666

Less than 40 

minutes of margin

Note 13

1664 None Note 11
FF-82

Part VII 

Evaluation

Room/ 

Zone

Exceptions to Full 

Area Detection and 

Automatic 

Suppression 

OMA # Exceptions to 40 

Minutes of 

Margin

Notes

1669 & 

1670

Less than 40 

minutes of margin

Note 14

1667 & 

1668

Less than 40 

minutes of margin

Note 15

8.3.50 772.0-A16 None 1022 None

1023 None

8.3.51 782.0-A1 None 1016 None

1024 None

8.3.52 782.0-A2 None 1016 None

1024 None

8.3.53 782.0-A3 None 1022 None

1023 None

8.3.54 782.0-A4 None 1022 None

1023 None

8.3.55 DBIPS-A No detection or 

suppression in duct 

bank

1023 Less than 40 

minutes margin

Note 16

8.3.56 DBIPS-B No detection or 

suppression in duct 

bank

1024 Less than 40 

minutes of margin

Note 16

8.3.57 IPS-A Partial detection, no 

suppression

1023 None Note 17

8.3.58 IPS-B Partial detection, no 

suppression

1024 None Note 17

8.3.59 IPS-C 

(East)

None 1024 None

8.3.60 IPS-C 

(Middle)

None 1024 None

8.3.61 IPS-C 

(West)

None 1023 None

Note 1 - Applies where re-entry of the fire area was considered after 60 minutes. Areas 

involving re-entry have automatic fire suppression. Areas involving re-entry have more than 

40 minutes of margin, from the plant trip, but less than 40 minutes if 60 minutes is needed to 

suppress the fire before re-entering. In other cases, the OMA may be needed in a room, but 

not the same room, on the same elevation as the postulated fire. In this event, TVA considered 

the possible environmental effects of the fire and concluded that those factors would not 

prevent the performance of the OMA. These areas have full suppression and detection, or 

have been evaluated and approved in Section 6.1.7 of this evaluation, therefore this is 

considered acceptable by the NRC staff.

Note 2 – Note 2 applies to multiple areas, as described in FPR Part VII, Section 3.1. TVA 

performed an analysis of the fire hazards in the area and determined that they do not warrant 

the installation of fire detection and automatic suppression. These areas are typically pipe 

chases, tunnels, tank rooms, entrance labyrinths, corridors, or portions of larger rooms where 
FF-83

the majority of the room is protected. The evaluation and approval of these areas’ lack of full 

area fire detection and automatic suppression is provided in Section 6.1.7 of this evaluation. 

Based on the available defense-in-depth features as evaluated in Section 6.1.7 of this 

evaluation, and the minimum 40 minutes of margin, the staff concludes that this reduced 

protection is acceptable.

Note 3 – OMAs 1275 and 1615, in zone 713.0-A1B, lack 40 minutes of margin. Zone 713.0-

A1B is part of a room that lacks area wide detection and automatic suppression. The room has 

a floor area of over 17,000 square feet and has a ceiling height of 23 feet. The primary 

combustible in this room is insulation on cables. The suppression system does not extend over 

the boric acid tank area. TVA designated the area around the tank a combustible control zone. 

The actions must be completed in 20 minutes. The demonstrated time was less than 7 minutes, 

resulting in over 13 minutes of margin. Based on the installed defense-in-depth features, the 

size of the fire area, and over 13 minutes of time margin, the NRC staff concludes that this 

OMA is acceptable for this specific room.

Note 4 - OMAs 1540 and 1542, for a fire in zone 737.0-A1B, must be completed within 20 

minutes. Demonstration of each pair of actions indicated a travel and performance time of less 

than 2 minutes, resulting in over 18 minutes of margin in each case. Zone 737.0-A1B is a 

portion of a room which has a primary combustible of cable insulation, a floor area of over 

23,000 square feet, and a nominal ceiling height of 19 feet. TVA determined that a plant trip 

could be caused by a fire in this zone, which may have the potential of not having the 

necessary operators in the control room at the time of the plant trip. TVA included a 5 minute 

reduction in margin time to account for this possibility. The room is equipped with fire detection 

and an automatic sprinkler system. Based on the installed defense-in-depth features and over 

13 minutes of time margin, the NRC staff concludes that these OMAs are acceptable for this 

specific room.

Note 5 – The room 737.0-A1 has area wide detection and suppression. However, the fire area 

lacks total area suppression and detection because there is neither detection nor suppression 

in the air locks (Rooms 737.0-A4 and 737.0-A11). FPR Section 3.1.7 states that the air locks do 

not contain fire safe shutdown components. Therefore, the lack of total area suppression and 

detection in the air locks is acceptable.

Note 6 - OMAs 1535 and 1536, for a fire in zone 737.0-A1N, must be completed within 

20 minutes. Demonstration of each pair of actions indicates a travel and performance time of 

less than 3 minutes, resulting in over 17 minutes of margin in each case. Zone 737.0-A1N is a 

portion of a room which has a primary combustible of cable insulation, a floor area of over 

23,000 square feet, and a nominal ceiling height of 19 feet. The fire area is equipped with fire 

detection and an automatic sprinkler system, except in the air locks which lack fire safe 

shutdown components, as discussed in Note 5. Based on the installed defense-in-depth 

features and over 17 minutes of time margin, the NRC staff concludes that these OMAs are 

acceptable for this specific room.

Note 7 – A fire in room 757.0-A2 or 757.0-A9 has the potential to challenge the ability to 

operate steam generator relief valves to control steam generator pressure. TVA implemented 

OMAs 1184 and 1185 to manipulate nitrogen stations to manually control pressure in the steam 

generators in the event that the fire detection and automatic fire suppression systems installed 

in rooms 757.0-A2 and 757.0-A9 do not prevent fire damage. The fire barrier(s) between the 

fire rooms and the OMA location (757.0-A24) have installed backdraft dampers that close when 
FF-84

the ventilation system is shut down. Proceduralized manual shutdown of the ventilation system 

will allow the backdraft dampers to close. Despite this, TVA considered the possibility that 

smoke from the fire affected room could enter the OMA location, creating a challenge to the 

operator performing the OMA. The operator sent to perform the OMA will have SCBA to allow 

performance of the OMA even if a limited amount of smoke enters the OMA location. TVA 

assumes that plant fires will be suppressed in 60 minutes. Thus, TVA assumed that 

performance of the OMA begins at 60 minutes to account for fire suppression activities to have 

completely suppressed the fire. There is 75 minutes for the operator to perform the OMA. The 

OMA has been demonstrated to take less than 8 minutes. Based on the installed defense-indepth features, the operator actions to shut down ventilation to close the backdraft dampers, 

and the available time to perform the OMA, the NRC staff concludes that these OMAs are 

acceptable for these rooms.

Note 8 - Rooms 757.0-A3, 757.0-A4, 757.0-A22, 757.0-A23, 772.0-A4, 772.0-A13, and 

772.0-A14, are equipped with ionization fire detection systems and manually actuated fire 

sprinkler systems, rather than the automatic fire sprinkler system. These vital battery board 

rooms are described as having battery and instrument boards, transformers, control panels, 

and junction boxes. Electrical components have circuit protection. Transient combustibles are 

controlled by plant procedures. OMAs in these rooms and zones have at least 40 minutes of 

time margin. Based on the installed ionization fire detection systems, the manually actuated 

fire suppression systems, the limited combustibles, and the available time margin, the NRC 

staff concludes that this is acceptable.

Note 9 – A fire in room 757.0-A17 or 757.0-A24 has the potential to challenge the ability to 

operate steam generator relief valves to control steam generator pressure. TVA implemented 

OMAs 1037 and 1038 to manipulate nitrogen stations to manually control pressure in the steam 

generators in the event that the fire detection and automatic fire suppression systems installed 

in rooms 757.0-A17 and 757.0-A24 do not prevent fire damage. The fire barrier(s) between the 

fire rooms and the OMA location (757.0-A21) have installed backdraft dampers that close when 

the ventilation system is shut down. Proceduralized manual shutdown of the ventilation system 

will allow the backdraft dampers to close. Despite this, TVA considered the possibility that 

smoke from the fire affected room could enter the OMA location, creating a challenge to the 

operator performing the OMA. The operator sent to perform the OMA will have SCBA to allow 

performance of the OMA even if a limited amount of smoke enters the OMA location. TVA 

assumes that plant fires will be suppressed in 60 minutes. Thus, TVA assumed that 

performance of the OMA begins at 60 minutes to account for fire suppression activities to have 

completely suppressed the fire. There is 75 minutes for the operator to perform the OMA. The 

OMA has been demonstrated to take less than 8 minutes. Based on the installed defense-indepth features, the operator actions to shut down ventilation to close the backdraft dampers, 

and the available time to perform the OMA, the NRC staff concludes that these OMAs are 

acceptable for these rooms.

Note 10 - OMA 1516 for a fire in room 757.0-A21, must be completed within 20 minutes. 

Demonstration of the action shows a travel and performance time of less than 3 minutes. This 

provides over 17 minutes of margin for this action. The room is equipped with fire detection 

and an automatic sprinkler system. The room has a floor area of 2244 square feet with a 

nominal ceiling height of 14 feet. TVA determined that a plant trip could be caused by a fire in 

this zone, which may have the potential of not having the necessary operators in the control 

room at the time of the plant trip. TVA included a 5 minute reduction in margin time to account 
FF-85

for this possibility. Based on the installed defense-in-depth features and over 12 minutes of 

time margin, the NRC staff concludes that this OMA is acceptable for this specific room.

Note 11 - Zones 772.0-A2 East and West, are in room 772.0-A2, the 480V Board Room 1B, 

and zones 772.0-A15 East and West, are in room 772.0-A15, the 480V Board Room 2B. A 

315 square foot portion in each of the rooms that contains the vital battery inverters and 

charger does not have automatic suppression. The combustible in that portion of each room is 

insulation on cables in one cable tray. In addition, there is one conduit wrapped with 

Thermo-Lag 770 in room 772.0-A2, which is considered a combustible. Section 6.1.7 of this 

evaluation includes a discussion of the partial suppression in these zones. Based on the small 

size of the unprotected area, the detection and automatic suppression throughout the other 

portions of the area, and at least 40 minutes of available margin, the NRC staff concludes that 

this is acceptable.

Note 12 – OMA 1495 for a fire in room 772.0-A15 West, lacks 40 minutes of time margin, and 

the room lacks full automatic suppression. The lack of automatic suppression is discussed in 

Note 11, which describes that detection is installed throughout the area, the area without 

suppression is small, and if a fire were to occur it would be controlled by suppression outside of 

the 315 square foot unprotected portion. Section 6.1.7 of this evaluation includes a discussion 

of the partial suppression in these zones. This OMA must be performed within 20 minutes, and 

the demonstrated performance time for the OMA is less than 2 minutes. TVA evaluated the 

OMA and concluded that the OMA would not be impacted if a fire were to occur in the 

unprotected area. Based on the small size of the unprotected area, TVA’s evaluation that an 

unsuppressed fire would not impede the manual action, and at least 18 minutes of time margin, 

the NRC staff concludes that this is acceptable.

Note 13 – OMAs 1665 and 1666 for a fire in room 772.0-A15 West, lack 40 minutes of time 

margin, and the room lacks full automatic suppression. The lack of automatic suppression is 

discussed in Note 11, which describes that detection is installed throughout the area, the area 

without suppression is small, and if a fire were to occur it would be controlled by suppression 

outside of the 315 square foot unprotected portion. Section 6.1.7 of this evaluation includes a 

discussion of the partial suppression in these zones. These OMAs must be performed within 

45 minutes, and the demonstrated performance time for the OMA is less than 8 minutes. TVA 

evaluated the OMA and concluded that the OMA would not be impacted if a fire were to occur 

in the unprotected area. Based on the small size of the unprotected area, TVA’s evaluation that 

an unsuppressed fire would not impede the manual action, and at least 37 minutes of time 

margin, the NRC staff concludes that this is acceptable.

Note 14 – OMAs 1669 and 1670 for a fire in room 772.0-A15 West, lack 40 minutes of time 

margin, and the room lacks full automatic suppression. The lack of automatic suppression is 

discussed in Note 11, which describes that detection is installed throughout the area, the area 

without suppression is small, and if a fire were to occur it would be controlled by suppression 

outside of the 315 square foot unprotected portion. Section 6.1.7 of this evaluation includes a 

discussion of the partial suppression in these zones. These OMAs must be performed within 

20 minutes, and the demonstrated performance time for the OMA is less than 5 minutes. TVA 

evaluated the OMA and concluded that the OMA would not be impacted if a fire were to occur 

in the unprotected area. Based on the small size of the unprotected area, TVA’s evaluation that 

an unsuppressed fire would not impede the manual action, and at least 15 minutes of time 

margin, the NRC staff concludes that this is acceptable.
FF-86

Note 15 – OMAs 1667 and 1668 for a fire in room 772.0-A15 West, lack 40 minutes of time 

margin, and the room lacks full automatic suppression. The lack of automatic suppression is 

discussed in Note 11, which describes that detection is installed throughout the area, the area 

without suppression is small, and if a fire were to occur it would be controlled by suppression 

outside of the 315 square foot unprotected portion. Section 6.1.7 of this evaluation includes a 

discussion of the partial suppression in these zones. TVA evaluated that the OMAs are needed 

within 22 minutes, but more specifically TVA’s analysis indicates that there is 30 minutes to 

perform the operator manual actions for WBN Unit 2 to prevent steam generator overfill. In 

addition, TVA indicated that some controls, such as tripping pumps that contribute to steam 

generator overfill have their cables traced for a fire in this area. The analysis assumes that the 

pumps are not tripped, so tripping the pumps would allow more time to successfully complete 

the OMA before overfilling the steam generators. Based on the small size of the unprotected 

area, greater than 13 minutes available to perform the actions, and actions available in the 

control room to slow the overfilling of the steam generators, the NRC staff concludes that this is 

acceptable.

Note 16 - DBIPS-A and DBIPS-B, the IPS Duct Banks, have no detection or suppression and 

lack 40 minutes of time margin. These areas have no credible ignition sources for the installed 

cables in the area. Since these are underground electrical conduit banks, no transient 

combustibles are expected. The manual actions that may be needed for fires in these duct 

banks are OMAs 1023 (DBIPS-A) and 1024 (DBIPS-B). OMA 1023 (DBIPS-A) has 46 minutes 

of margin, and OMA 1024 (DBIPS-B) has 45 minutes of margin. Since these areas do not have 

fire detection, TVA allows for 10 minutes for visual fire validation, which results in over 

35 minutes of margin when visual fire validation time is included. Based on the limited ignition 

sources for these underground duct banks, and the available time margin, the NRC staff 

concludes that this is acceptable.

Note 17 – Fire areas 58 and 59 (IPS-A and IPS-B), the Intake Pumping Station areas A and B, 

have fire detection over the ERCW pumps and in each of the ERCW strainer rooms. Each has 

a floor area in excess of 3500 square feet and a ceiling height of at least 13 feet. The 

combustibles in the areas consist of the insulation on cables in trays, oil associated with the 

pumps, transformers, and electrical panels. The areas do not contain redundant safe shutdown 

equipment. OMAs 1023 (IPS-A), and 1024 (IPS-B) have at least 40 minutes of time margin. 

Based on the partial detection, the size of the rooms, and the available time margin, the NRC 

staff concludes that this is acceptable. 

Conclusion –Manual Actions 

The NRC staff reviewed the information submitted by TVA regarding these specific OMAs and 

the fire scenarios that would cause them to be performed. The NRC staff concludes that, 

based on the fire protection defense-in-depth features and the feasibility and reliability of the 

OMAs, performance of these manual actions provides reasonable assurance that the capability 

to safely shut down will be available, and is, therefore, acceptable.

6.1.10 Evaluation – Fire Hazards Analysis in Lieu of 10 CFR Part 50, Appendix R, 

Section III.G.2 Separation 

In FPR Part VII, Section 2.9, TVA stated that there are rooms at WBN that lack the separation 

described by Section III.G.2 of Appendix R to 10 CFR Part 50. For these rooms, TVA relied 
FF-87

upon a fire hazards analysis (FHA) and an analysis of the safe shutdown capability rather than 

OMAs. In many cases these rooms are part of larger fire areas.

For all the rooms included as part of this evaluation, transient combustibles and ignition sources 

have been reported by TVA to be controlled by plant procedures. TVA provided a justification 

why certain ignition sources were not considered credible ignition sources. In addition, 

separation between adjacent rooms has been evaluated, and TVA concluded that no credible 

fire could spread either from or to adjacent rooms. TVA reported that room fires affecting the 

FSSD equipment would neither initiate nor require a plant trip.

10 CFR Part 50, Appendix R, Section III.G.2 states, requires, in certain circumstances, 

separation of cables and equipment and associated non-safety circuits of redundant trains by a 

fire barrier having a 3-hour rating, or by a horizontal distance of more than 20 feet with no 

intervening combustible or fire hazards, along with fire detectors and an automatic fire 

suppression system. This separation of cables and equipment and the installation of fire 

detectors and an automatic fire suppression system is required in order to limit fire damage to 

one train of systems necessary to achieve and maintain hot shutdown conditions and to ensure 

that systems necessary to achieve and maintain cold shutdown can be repaired within 72 hours. 

Although TVA has some areas that lack the required separation, the NRC staff concludes that 

these areas are acceptable.

6.1.10.1 Rooms without Credible Ignition Sources and Redundant Trains

• Rooms 692.0-A29 and 692.0-A30 – Boric Acid Evaporator Package Rooms A and B

• Rooms 729.0-A1 and 737.0-A6 – WBN Unit 1 South Main Steam Valve Room and Air 

Lock

• Room 729.0-A2 – WBN Unit 1 North Main Steam Valve Room

• Room 729.0-A6 – Nitrogen Storage Area

• Room 729.0-A10 – WBN Unit 2 North Main Steam Valve Room

• Rooms 729.0-A11 and 737.0-A10 – WBN Unit 2 South Main Steam Valve Room and Air 

Lock

• Room 729.0-A12 – WBN Unit 1 Steam Valve Instrument Room A

• Room 729.0-A13 – WBN Unit 2 Steam Valve Instrument Room B

• Rooms 729.0-A15 and 763.5-A2 WBN Unit 2 Additional Equipment Building

TVA evaluated fire protection defense-in-depth for these rooms. These rooms have been 

reported to have minimal combustible loading consisting of plastics associated with small 

components or grease and oil associated with valves. Cables related to FSSD are installed 

within these rooms within conduit. Air lines that have a related FSSD function may be installed 
FF-88

within these areas and are of welded steel construction. TVA evaluated air operated 

components, such as solenoid valves, based on the fire hazards in the areas and determined 

that there is no credible fire that would impact the valves. Also, for areas where such air 

operated components are installed, TVA determined that the system capacity is sufficient to 

allow for the loss of a device without impacting system functionality. Other than cables within 

conduit, welded steel air piping, and air operated components, no other FSSD equipment is 

installed in these rooms. TVA evaluated the installed equipment in these rooms and concluded 

that there are no credible in situ ignition sources. Other ignition sources and transient 

combustibles are controlled in accordance with plant procedures. TVA determined that, even 

without any installed fire detection or suppression, no fire scenarios could credibly affect the 

cables or air lines that are involved in plant safe shutdown. TVA determined that for each of 

these areas, if fire damage were to occur to the installed equipment a plant trip would not be 

initiated or required. 

The NRC staff concludes that the lack of separation in these rooms is acceptable because of 

the fire protection defense-in-depth features, the limited combustibles, ignition sources and 

combustible controls, and because of no credible fire scenarios affecting FSSD equipment that 

would either initiate or require a plant trip. 

6.1.10.2 Room 757.0-A13 – Refueling Floor and New Fuel Storage Vault

The refueling floor has two fixed ignition sources installed, specifically two auxiliary air 

compressor units and equipment associated with hydraulic cranes and hoists. The air 

compressors, although credible ignition sources, are more than 20 feet separated from each 

other with no intervening combustibles. Therefore, a fire affecting one compressor would not 

be expected to affect the other compressor. A failure of one of the compressors could cause 

the air supply system to lose supply pressure. The other train would be available. In addition, 

a low pressure alarm on the affected system would be annunciated in the MCR.

The crane and hoist are in operation only when plant personnel are operating them. Therefore, 

any fire would be quickly identified by personnel in the immediate vicinity, and this would 

provide assurance that other FSSD equipment would not be damaged.

The new fuel storage vault has negligible combustibles and no credible ignition sources. 

The NRC staff concludes that the lack of separation in this room is acceptable because of the 

fire protection defense-in-depth features, the limited combustibles and combustible controls, the 

separation between redundant trains within the room with no intervening combustibles, the 

continuous staffing when cranes and hoists are used, and because there are no credible fire 

scenarios affecting FSSD equipment that would either initiate or require a plant trip. 

6.1.10.3 Room 757.0-A14 – Unit 2 Reactor Building Access Room and Room 757.0-A15 –

WBN Unit 2 Reactor Building Equipment Hatch

In contrast to the other rooms evaluated, these rooms have more than minimal combustible 

loading. The combustible loading is composed primarily of thermoset cable. The electrical 

circuits in the cables have circuit protection that reduces the likelihood of a self-ignited cable 

fire. TVA reported that there are no credible ignition sources in these rooms. In addition, each 

of these rooms is equipped with fire detection and automatic fire suppression systems.
FF-89

For each of these rooms TVA identified five sets of redundant components. Each set of 

components is discussed below.

SGs 2 and 3 Main Steam Isolation Valves – The main steam isolation valves (MSIVs) are 

normally energized and fire damage that deenergizes the train will cause the MSIVs to close. 

Closed is the normal safe shutdown configuration. The fire damage failure mode of concern is 

a sustained hot short that keeps the MSIVs open. 

In the unlikely event that damage causes a sustained hot short, given the limited ignition 

sources, full area detection and automatic suppression, the main steam system can be isolated 

from the MCR using the steam load valves.

RCP Seal Injection – An instrument cable for control circuits for the valve that controls the 

charging flow is located in these rooms near the ceiling. Based on the limited ignition sources 

and installation of an automatic fire suppression system, fire damage at the ceiling of these 

rooms is unlikely. In the unlikely event that the control circuits are damaged and the control 

valve spuriously operates, the indication is available and MCR operators could operate the 

valve using a different pressurizer level input or manually.

Control Cable for SG 3 PORV – A control cable for SG 3 power-operated relief valve (PORV) is

routed through these rooms. A hot short to the control cable would cause the PORV to close, 

and not to be used for safe shutdown. In the unlikely event that a fire were to start, given the 

limited ignition sources, and the fire was not extinguished by the installed fire suppression 

system, the location of the cable in conduit over 20 feet above the floor provides assurance that 

cable damage would not occur.

Main Feedwater Isolation for SGs 2 and 3 – Main feedwater isolation valve control cables are 

installed in conduit in these rooms. Fire damage to these cables could interfere with the 

isolation of main feed water. In the unlikely event that a fire were to start, given the limited 

ignition sources, and the fire was not extinguished by the installed fire suppression system, 

operators in the MCR would still have available indication and controls over other valves that 

would be available to isolate the main feedwater flow.

Main Feedwater Bypass Line Isolation Valve Circuits for SGs 2 and 3 – Main feedwater bypass 

lines could remain open upon concurrent hot shorts of the control cables. In the unlikely event 

that a fire were to start, given the limited ignition sources, and fire was not extinguished by the 

installed fire suppression system, the control valves could still be closed by operator actions 

from the MCR.

The NRC staff concludes that the lack of separation in these rooms is acceptable because of 

the fire protection defense-in-depth features, the limited ignition sources, the available detection 

and suppression systems, and the height of the cables above the floor, or the alternative ways 

of meeting the safe shutdown goals using MCR actions. 

6.1.10.4 WBN Unit 2 Containment Rooms

• Room 2RIR – WBN Unit 2 Reactor Instrument Room

• Rooms 2RA1, 2RA2, 2RA3, and 2RA4 – WBN Unit 2 Accumulator Rooms 1, 2, 3, and 4
FF-90

• Rooms 2RF1 and 2RF2 – WBN Unit 2 Reactor Building Fan Rooms 1 and 2

• Rooms 2RI-1, 2RI-2, 2RI-3, and 2RI-4 – WBN Unit 2 Reactor Building Inside Crane Wall 

Rooms

• Rooms 2RO-1, 2RO-2, 2RO-3, and 2RO-4 – WBN Unit 2 Reactor Building Outside 

Crane Wall Rooms

TVA stated that these rooms have stronger combustible controls than other plant areas, since 

these areas are considered combustible control zones. In addition, many of these areas are 

inaccessible during power operations and involve the climbing of ladders for entry, which will 

reduce the likelihood of transient combustibles and ignition sources. TVA stated that none of 

these rooms have credible in situ ignition sources. TVA provided a discussion that concluded 

fires in adjoining rooms would not affect the FSSD equipment in these rooms, due to either the 

types and quantities of combustibles in the adjoining rooms, or installed automatic suppression 

and detection in the adjoining rooms. TVA stated that a fire in one of these rooms affecting 

FSSD equipment would neither initiate nor require a plant trip.

In addition to defense-in-depth features described above, the FSSD capability has one or more 

of the additional features that provide(s) additional assurance that a fire in one of these rooms 

will not challenge plant safe shutdown:

• redundant cables are separated by at least 2 feet horizontally

• cables are installed in conduit,

• alternate systems are available in the control room to shut down the plant,

• spurious actuations are avoided by the use of dedicated conduit with no other energized 

conductors, 

• spurious actuations are avoided since they would only occur if there were a proper 

polarity two or three phase hot short,

• targets are high above the floor, at least 10 feet, and/or

• redundant trains may be located in the analysis volume (AV), but not in the room being 

evaluated.

The NRC staff concludes that the lack of separation in these rooms is acceptable because of 

the limited combustibles and ignition sources, the failure of the FSSD equipment or cables 

would not initiate or require a plant trip, and because the redundant safe shutdown circuits have 

one or more of the additional criteria described above. 

6.1.10.4 Conclusion for section 6.1.10

Therefore, the NRC staff concludes that, for the areas described above, there is reasonable 

assurance that fire damage would be limited to one train of systems necessary to achieve and 

maintain hot shutdown conditions and that systems necessary to achieve and maintain cold 

shutdown would be able to be repaired within 72 hours.
FF-91

6.2 Exceptions and Evaluations Related to BTP (APSCB) 9.5-1, 

Appendix A Guidance

6.2.1 Exception – Fire Detection in Refueling Room and New Fuel Storage Vault 

TVA committed to follow the guidance in Positions F.12 and F.13 of Appendix A to BTP 

(Auxiliary Power Conversion Systems Branch (APCSB)) 9.5-1, which states that fire detectors 

should be installed in new fuel and spent fuel pool areas. Contrary to the guidance, the 

refueling room (Room 757.0-A13), which includes the New Fuel Storage Vault (elevation 

741.5 feet), is not provided with a detection system.

TVA states that the refueling room is constructed of reinforced concrete. This room has a large 

open area with a floor area of approximately 16,000 square feet and a nominal ceiling height of 

56 feet. The walls, floor, and penetration seals have a fire resistance rating of 2 hours or 

greater. The doors are not UL-listed doors, but have been evaluated as equivalent to fire-rated 

doors as listed in the FPR Part II, Table 14.8.1 (Fire Doors). The dampers have a rating of 

1.5 or 3 hours, based on the rating of the barrier they are installed in, in accordance with NFPA 

90A “Standard for the Installation of Air Conditioning and Ventilation systems.”

During normal operations, the in situ combustible loading in the refueling room and the new 

fuel storage vault is insignificant, resulting in an equivalent fire severity of less than 5 minutes. 

There are no ignition sources in the new fuel storage vault. The combustible materials in the 

refueling room are widely dispersed, which further diminishes the magnitude of a postulated fire. 

The combustibles consist of InstaCote (a plastic type fuel transfer canal coating); lube oil in air 

compressors; hoists and cranes; plastics associated with the electrical equipment, panels, fuel 

pool boundary, lighting and boxes; rubber fire hose; and anticipated amounts of radwaste trash 

and laundry. TVA further stated that transient combustibles in the room are controlled by WBN 

procedure NPG-SPP-18.4.7, “Control of Transient Combustibles.” The potential ignition 

sources in the room are panels, air compressors, transformers, and lighting cabinets. The only 

ignition sources that could impact an FSSD component or cable are the Trains A and B auxiliary 

air compressors.

The room is manned during an outage, which can assist in the early detection of a fire. The 

new fuel storage vault is only accessible from the refueling room, and that access is normally 

closed with a steel hatch cover. The cover is removed when new fuel is received and stored 

until needed for a refueling outage. Due to the high ceiling and limited amount of combustibles, 

a fire in this area may not have sufficient energy to create the necessary air currents to carry the 

smoke to the ceiling. In this situation, the smoke detectors at the ceiling level may not be able 

to provide early detection in the event of a fire.

Standpipe and hose stations are provided in the refueling room and in adjacent rooms.

The Trains A and B auxiliary air compressors supply backup air to the Trains A and B air 

header if the normal air supply from the station air compressors is unable to maintain minimum 

pressure on the air header. A fire involving either of the auxiliary air compressors would not 

impact the normal air supply or the other auxiliary air compressor. The worst case fire scenario 

would be a loss of one train of auxiliary control air, which would not require either unit to shut

down. The other FSSD circuits are routed in conduits in the refueling floor area and are outside 
FF-92

the fire zone of influence of the compressors. Therefore, a fire in the refueling room will not 

impact FSSD capability.

Appendix A to BTP (APCSB) 9.5-1, Positions F.12 and F.13 state that fire detectors should be 

installed in new fuel and spent fuel pool areas. Fire detectors are required in these areas to 

ensure a fire would be promptly detected so that manual suppression activities could be initiated 

thereby limiting fire damage. Although TVA is proposing that the refueling room (including the 

new fuel storage vault) not be equipped with a detection system, the NRC staff concludes that 

this is acceptable because of the size of the refueling room, the limited amounts of in situ and 

transient combustibles, the separation of the room from other plant areas by fire-rated barriers, 

and the routing of FSSD circuits in conduits away from credible ignition sources. Therefore, the 

NRC staff concludes that there is reasonable assurance that fire damage would be limited were 

a fire to occur in these areas.

6.2.2 Exception – Fire Doors 

TVA committed to the guidance in Position D.1.j in Appendix A to BTP (APCSB) 9.5-1, which 

states that door openings should be protected with equivalently rated fire doors, frames, and 

hardware that have been tested and approved by a nationally recognized laboratory.

In FPR Part VII, Section 4.1, TVA stated that, contrary to the guidance, a number of fire doors 

have been altered by the addition of signs and security hardware, or have been damaged and 

repaired on site. Additionally, special-purpose doors, such as flood doors and pressure doors, 

are not UL labeled.

The fire doors that are not listed or labeled as fire-rated assemblies have been evaluated to the 

guidance of National Fire Protection Association (NFPA) 80-1975, “Fire Doors and Windows,” 

by TVA or nationally recognized laboratories for fire door assemblies. The evaluation criteria for 

fire door assemblies are documented and controlled by WBN General Engineering 

Specification-73, “Installation, Modification and Maintenance of Fire Protection Systems 

and Features.”

FPR Part II, Table 14.8.1, lists the plant fire doors and the doors’ fire rating in hours. The table 

identifies doors that are not UL listed as having been evaluated and identified as equivalent to 

fire-rated doors or they have been evaluated as being acceptable. A number of the fire doors at 

WBN have been altered by the addition of signs and security hardware or have been damaged 

and repaired. Examples of other fire doors that are not UL rated are special purpose doors 

such as flood doors and pressure doors, security doors in the MCR that are constructed of 

heavy welded steel construction, and hollow core metal swinging doors.

Appendix A to BTP (APCSB) 9.5-1, Position D.1.j states that door openings should be protected 

with equivalently rated fire doors, frames, and hardware that have been tested and approved by 

a nationally recognized laboratory. The protection of door openings in this manner is required to 

limit fire damage to the area where the fire occurred. Although TVA is proposing the use of 

doors that may not have been tested and approved by a nationally recognized testing 

laboratory, the NRC staff concludes that these doors are acceptable because they were 

evaluated to the guidance of NFPA 80-1975, and WBN General Engineering Specification-73 

controls and documents the installation, modifications, and maintenance of all fire doors. 

Therefore, the NRC staff concludes that there is reasonable assurance that these doors would 

limit fire damage to the area where the fire occurred.
FF-93

6.2.3 Exception – Openings in Fire Walls 

TVA committed to the guidance in Section D.1.j of Appendix A to BTP (APCSB) 9.5-1, which 

states that fire barriers should be capable of withstanding the fire hazards to which they could 

be exposed. NRC GLs and guidance documents state that penetrations in walls, floors, and 

ceilings forming part of a fire barrier should be protected with seals or closure devices having 

a fire resistive rating equivalent to that required of the barrier.

In FPR Part VII, Section 4.2, TVA stated that there is a 6-inch-wide by 3-inch-deep gutter that 

penetrates each stairwell enclosure (Stairwells C1 and C2) from the corridor (Room 692.0-C11) 

in the control building.

These two stairwells are located at the opposite ends of the corridor (approximately 70 feet 

apart). The gutter penetrates the walls separating the stairwells from the corridor. Located 

in the gutter, there is one floor drain in each stairwell and two floor drains in the corridor.

The in situ combustible loading for the corridor is low and results in an equivalent fire severity of 

less than 20 minutes. The corridor is provided with a preaction sprinkler system that is actuated 

by an ionization type smoke detection system. Standpipe and hose stations are in the 

two stairwells, and portable extinguishers are provided in the corridor.

The in situ combustible liquids on elevation 692.0 feet of the control building are 35 gallons 

of lube oil associated with each of the two electrical board room chiller packages. The chiller 

packages are located in the Unit 2 mechanical equipment room, which is not part of 

Stairwells C1 or C2 or the corridor. However, the room is separated from Stairwell C2 by a 

2-hour-rated reinforced concrete wall. The combustibles in the Unit 2 mechanical equipment 

room consist of lube oil in the chillers, plastics associated with the electrical panels, boxes, 

lights, and insulation on piping. The in situ combustible loading in this room is low resulting 

in an equivalent fire severity of less than 5 minutes. This room also has full detection and 

suppression installed.

Appendix A to BTP (APCSB) 9.5-1, Position D.1.j states that fire barriers should be capable of 

withstanding the fire hazards to which they could be exposed. This requirement for fire barriers 

ensures that the fire barrier will limit fire damage to the area where the fire occurred. Although 

TVA is proposing to allow a 6-inch-wide by 3-inch-deep gutter that penetrates each stairwell 

enclosure (Stairwells C1 and C2) from the corridor (Room 692.0-C11) in the control building.

use of doors that may not have been tested and approved by a nationally recognized testing 

laboratory, the NRC staff concludes that these doors are acceptable because they were 

evaluated to the guidance of NFPA 80-1975, and WBN General Engineering Specification-73 

controls and documents the installation, modifications, and maintenance of all fire doors. 

Therefore, the NRC staff concludes that there is reasonable assurance that these doors would 

limit fire damage to the area where the fire occurred.

6.2.4 Exception – Manual Hose Stations 

TVA committed to the guidance in Section E.3.d of Appendix A to BTP (APCSB) 9.5-1, which 

states that interior manual hose installations should be able to reach any location with at least 

one effective hose stream. To accomplish this, standpipes with hose connections equipped with 
FF-94

a maximum of 75 feet of 1-1/2-inch woven jacket lined fire hose and suitable nozzles should be 

provided.

In FPR Part VII, Section 4.3, TVA stated that there are manual hose stations with more than 

75 feet of 1-1/2-inch UL listed or factory mutual-approved fire hose located throughout the plant. 

The pressure loss in fire hoses due to conditions such as friction with the inner wall of the hose 

and turbulent water flow is directly proportional to the length of the hose. If the pressure loss is 

excessive, the hose stream may not be effective.

To justify the use of hoses of greater than 75 feet in length up to 100 feet in length, TVA stated 

that these installations are consistent with the guidelines of NFPA 14-1974, “Standard for the 

Installation of Standpipe and Hose Systems,” which allow up to 100 feet of hose connected to 

the standpipe.

For hose stations with more than 100 feet of hose, TVA stated that although those specific 

hose stations may not have been tested, hose stations at a higher elevation in the respective 

buildings were tested at a minimum of 65 psig at 500 gpm at a 2.5-inch hose connection. Also,

TVA calculated that there is 6 psi additional pressure loss for each additional 25-foot section of 

hose. TVA stated, in its letter dated May 30, 2012 (ADAMS Accession No. ML12153A374), that 

the tested hose stations are 31.5 feet higher in elevation than the hose stations with the 

additional hose. TVA calculated that 31.5 feet of elevation equates to approximately 13.5 psig 

of additional pressure at the lower elevation. This additional pressure on lower elevations would 

provide sufficient additional pressure to compensate for the approximately 6 psi of pressure loss 

for each of the two additional hose sections and, therefore, would provide sufficient pressure 

and flow to meet the guidance of NFPA 14-1974.

Appendix A to BTP (APCSB) 9.5-1, Position E.3.d, states that interior manual hose installations 

should be able to reach any location with at least one effective hose stream, and that to 

accomplish this, standpipes with hose connections equipped with a maximum of 75 feet of 

1-1/2-inch woven jacket lined fire hose and suitable nozzles should be provided. Manual hose 

installations to reach any location are required to ensure that there is a means to limit fire 

damage to SSCs important to safety so that the capability to safely shutdown the plant is 

maintained. Although the licensee is providing more than 75 feet of 1-1/2 inch fire hose on 

some hose stations, the NRC staff concludes that this is acceptable because NFPA 14-1974 

allows up to 100 feet of hose, and because the licensee has demonstrated that for hose stations 

with 125 feet of hose installed, there is sufficient pressure and flow to reach any location with 

one effective hose stream which also meets the requirements of NFPA 14-1974. Therefore, the 

NRC staff concludes that there is reasonable assurance that the manual hose installation would 

limit fire damage to SSCs important to safety in order to maintain safe shutdown capability.

6.2.5 Exception – Fire Barrier Penetration between Fuel Oil Transfer Pump Room and 

the Diesel Generator Building Corridor 

TVA committed to the guidance in Position D.1.j of Appendix A to BTP (APCSB) 9.5-1, which 

states that penetrations in fire barriers, including conduits and piping, should be sealed or 

closed to provide a fire resistance rating at least equal to that of the fire barrier itself. The fire 

hazard in each area should be evaluated to determine barrier ratings.

In FPR Part VII, Section 4.6, TVA stated that the fire barrier separating the fuel oil transfer pump 

room (Room 742.0-D8) from the DG building corridor (Room 742.0-D9) is a 2-hour-rated fire 
FF-95

barrier and has a penetration containing a steel box. This penetration is not a tested fire-rated 

penetration assembly.

The fire barrier separating the fuel oil transfer pump room and the corridor is constructed of 

8-inch-thick reinforced concrete block and is fire-rated for 2 hours. The annular gap between 

the block wall and the box is filled with concrete grout, but no sealant material is installed within 

the box. The box back (inside the fuel oil transfer pump room) is a steel plate. The front of the 

panel is a steel plate with cutouts for three metal junction boxes.

The in situ combustible loading of the fuel oil transfer pump room is approximately 3,730 Btu/ft2

and is due to insulation on cables associated with panel 0-L-162, hand switches, an emergency 

lighting unit, and foam plastic insulation. The in situ combustible loading of the corridor is 

approximately 77,700 Btu/ft2 of which approximately 96 percent is due to insulation on cables in 

cable trays. The other in situ combustibles are dispersed throughout the corridor and do not 

present a direct exposure hazard to the box. The corridor width at the panel is approximately 

6 feet. The door into the 2B-B DG is across from the box and the door to the fuel oil transfer 

pump room is next to the box. The end of the corridor is less than 6 feet from this door. TVA 

stated that this arrangement minimizes the probability of transient combustibles being stored 

near the box.

The fuel oil transfer pump room is provided with a fire detection system and a total flooding 

automatic CO2 suppression system. The detection system alarms in the MCR and actuates the 

suppression system. The corridor is provided with a fire detection system and an automatic 

sprinkler system. The detection system alarms in the MCR and actuates the suppression 

system.

The top of the box is located approximately 13 feet below the ceiling. TVA stated that, in light of 

this distance and the location of the box at the end of the corridor, the detection system should 

alarm the MCR and actuate the suppression system before a hot gas layer could challenge 

the box.

TVA stated that the fuel oil transfer pump room and the corridor (analysis volume AV-081B) do 

not contain components required for safe shutdown in the event of a fire in these rooms. The 

small amount of in situ combustibles and the lack of free floor space limit the quantity of 

transient combustibles, thereby limiting the severity of a postulated fire in the room. The failure 

of a fuel oil line or pump that resulted in a fire is addressed by the total flooding, automatic CO2

suppression system that will also control a postulated transient fire until the fire brigade 

responds.

Appendix A to BTP (APCSB) 9.5-1, Position D.1.j states that penetrations in fire barriers, 

including conduits and piping, should be sealed or closed to provide a fire resistance rating at 

least equal to that of the fire barrier itself and that the fire hazard in each area should be 

evaluated to determine barrier requirements. Penetrations in fire barriers are required to be 

rated as they are a means to limit fire damage so that the capability to shutdown the plant safety 

is maintained. Although TVA is proposing the use of an unrated penetration in a 2 hour rated 

barrier between the fuel oil transfer pump room and the DG building corridor, the NRC staff 

concludes that this configuration is acceptable because the unrated penetration is configured in 

such a way as to prevent the passage of flames, hot gases, or water from the corridor to the fuel 

oil transfer pump room or vice versa; because neither area contains components requirement 

for safe shutdown; because fire severity in either room is limited; and, because automatic 
FF-96

suppression is provided. Therefore, the NRC staff concludes that there is reasonable 

assurance that the unrated penetration would not impact fire damage limits or the capability to 

safely shutdown the plant.

6.2.6 Exception – Undampered Penetrations between the Unit 1 Penetration Room and 

the Unit 1 Annulus and the Unit 2 Penetration Room and the Unit 2 Annulus

In FPR Part VII, Section 3.2, TVA stated that the walls separating the Unit 1 penetration room 

(Room 713.0-A6) from the Unit 1 Annulus and the Unit 2 penetration room (Room 713.0-A19) 

from the Unit 2 Annulus are 3-hour rated fire barriers. The containment purge air system return 

and exhaust ducts penetrate these walls in three places. The penetrations are not provided 

with fire dampers.

TVA provided the following details regarding these configurations:

• The ducts are constructed of 0.25 inch thick steel plates and welded schedule 10 pipe.

• The connection between the duct and the purge air system is protected by 3M M20A 

wrap in Room 713.0-A6 and 3M E54C wrap in Room 713.0-A19. Both wraps provide 

3-hour fire resistance.

• The ducts are rigidly attached to the concrete wall.

• The penetrations are not straight-through, instead the openings in the concrete wall are 

offset to provide radiation protection.

• The ducts have no openings in the pipe chase.

• There is automatic detection and suppression installed in the annuluses and pipe 

chases.

• The two annuluses and the areas under the ducts in the pipe chases are combustible 

control zones.

Appendix A to BTP (APCSB) 9.5-1, Position D.1.j states that penetrations for ventilation

systems should be protected by a standard "fire door damper" where required. Penetrations in 

fire barriers are required to be rated as they are a means to limit fire damage so that the 

capability to safely shutdown the plant is maintained. Although TVA is proposing that no fire 

dampers be installed in the containment purge air system return and exhaust ducts in the fire 

barriers between the WBN Unit 1 penetration room (Room 713.0-A6) and the WBN Unit 1 

Annulus, and the WBN Unit 2 penetration room (Room 713.0-A19) from the WBN Unit 2 

Annulus, the NRC staff concludes that this is acceptable because of the physical configuration

as described above, the installed fire protection systems, and the administrative controls. 

Therefore, the NRC staff concludes that there is reasonable assurance that the lack of fire 

dampers would not impact fire damage limits or the capability to safely shutdown the plant.
FF-97

6.2.7 Exception – Openings in Fire Barriers

Section D.1.j of Appendix A to BTP (APCSB) 9.5-1, “Guidelines for Fire Protection for Nuclear 

Plants Docketed Prior to July 1, 1976,” specifies that penetrations in walls, floors, and ceiling 

forming part of a fire barrier be protected with self-closure devices having a fire-resistive rating 

equivalent to that of the barrier.

6.2.7.1 Ventilation and Purge Air Room Ventilation Penetrations

In FPR Part VII, Section 2.6.1, TVA stated that the ventilation and purge air (VPA) rooms 

(Rooms 737.0-A5 and 737.0-A9), the post-accident sampling system (PASS) rooms 

(Rooms 729.0-A8 and 729.0-A9), and the nitrogen storage room (Room 729.0-A6) are 

separated by 2-hour fire-rated barriers. The walls and floor of the VPA rooms are penetrated 

by HVAC ducts that pass from the PASS rooms, enter the VPA rooms, and then exit into the 

PASS and nitrogen storage rooms. TVA stated that the ducts have no fire dampers, but they 

also have no openings into the VPA rooms. Additionally, one duct enters each VPA room from 

the nitrogen storage room and terminates at a normally closed isolation damper. The ducts are 

constructed from Schedule 40 carbon steel pipe. Pipe sleeves are provided where the ducts 

penetrate the barriers between the VPA rooms and the PASS rooms and nitrogen storage 

rooms. Further, the annular space between the sleeves and the ducts is sealed with a 

fire-rated seal.

TVA stated that each of these rooms contains safe shutdown equipment. TVA further stated 

that the VPA and PASS rooms have fire detection and automatic fire suppression systems 

installed, and the nitrogen storage room has ionization smoke detection. Standpipe and hose 

systems are available in adjacent rooms and portable extinguishers are also available for 

manual firefighting in these rooms.

TVA stated that the significant fire exposure to the ducts from the VPA rooms consists 

of charcoal filter units in each VPA room. TVA also stated that closed-head water-spray 

suppression systems are provided for the charcoal filters and are actuated by duct-mounted 

ionization smoke detectors.

TVA stated that the effect of a fire in the PASS rooms or the nitrogen storage room would be 

experienced in the VPA rooms in the form of radiant heat from hot gases passing through the 

ducts. In the VPA rooms, TVA stated that no fixed combustibles are located in the immediate 

vicinity of these ducts, and the ducts are separated from the nearest safe shutdown circuit by 

more than 20 feet.

Appendix A to BTP (APCSB) 9.5-1, Position D.1.j states that penetrations for ventilation

systems should be protected by a standard "fire door damper" where required. Penetrations in 

fire barriers are required to be rated as they are a means to limit fire damage so that the 

capability to safely shutdown the plant is maintained. Although TVA is proposing that no fire 

dampers be installed in the HVAC ducts that pass from the PASS rooms, enter the VPA rooms, 

and then exit into the PASS and nitrogen rooms, the NRC staff concludes that this is acceptable 

because of the installed fire detection and automatic fire suppression systems, the special 

hazard protection for the charcoal filters, and the construction of the ducts. Therefore, the NRC 

staff concludes that there is reasonable assurance that the lack of fire dampers would not 

impact fire damage limits or the capability to safely shutdown the plant.
FF-98

6.2.7.2 Scuppers

6.2.7.2.1 ERCW Pump Room 

TVA committed to the guidance in Position D.1.j of Appendix A to BTP (APCSB) 9.5-1, which 

states that penetrations in fire barriers, including conduits and piping, be sealed or closed to 

provide a fire resistance rating at least equal to that of the fire barrier itself.

In FPR Part VII, Section 2.6.2.1, TVA stated that, contrary to Position D.1.j, on elevation 

741.0 feet of the IPS, there are four scupper openings penetrating the fire wall between the 

ERCW pump rooms and traveling screen room.

The wall separating the redundant ERCW pumps and the wall separating the ERCW pumps 

from the traveling screen pumps are 3-hour fire-rated barriers with the exception of the

four scupper openings. These scupper openings are located at the floor and provide drainage 

of rainwater from the ERCW pump rooms to the traveling screen wells. The floor slopes away 

from the ERCW pumps toward the scuppers so that a fire in one ERCW pump room will not 

propagate through the scuppers and jeopardize a redundant train of ERCW pumps.

The wall separating the ERCW pump rooms and traveling screen room is intended to protect 

the rooms from the radiant heat of an exposure fire. The roof is designed as a missile shield 

and has beams that will allow free air flow from a fire to dissipate heat to the outside 

environment.

ERCW Pump Rooms A and B have heat detectors installed over the ERCW pumps, and 

standpipe and hose stations are accessible for manual firefighting activities. TVA stated that 

even though these rooms are not provided with suppression and full area detection, the fire area 

barrier ratings are sufficient given the combustible loadings in the area.

Appendix A to BTP (APCSB) 9.5-1, Position D.1.j states that penetrations in fire barriers, 

including conduits and piping, should be sealed or closed to provide a fire resistance rating at 

least equal to that of the fire barrier itself and that the fire hazard in each area should be 

evaluated to determine barrier requirements. Penetrations in fire barriers are required to be 

rated as they are a means to limit fire damage so that the capability to shutdown the plant safety 

is maintained. Although TVA is proposing four unrated scupper openings penetrating the fire 

wall between the ERCW pump rooms and traveling screen room, the NRC staff concludes that 

this is acceptable because of the location of the scupper openings, the floor slope away from 

the ERCW pumps, the installed heat detectors over the ERCW pumps, and the availability of 

standpipe and hose systems for manual firefighting activities. Therefore, the NRC staff 

concludes that there is reasonable assurance that the unrated penetrations would not impact 

fire damage limits or the capability to safely shutdown the plant.

6.2.7.2.2 Yard Duct Bank 

TVA committed to the guidance in Position D.1.j of Appendix A to BTP (APCSB) 9.5-1, which 

states that penetrations in fire barriers, including conduits and piping, be sealed or closed to 

provide a fire resistance rating at least equal to that of the fire barrier itself.
FF-99

In FPR Part VII, Section 2.6.2.1, TVA stated that contrary to Position D.1.j, there are scupper 

openings in the Train A and Train B yard duct banks that run from the auxiliary building to the 

IPS where they share a common wall in three manholes.

Manholes 1A and 1B, 2A and 2B, and 3A and 3B are used to access the Train A and Train B 

duct banks that connect the auxiliary building to the IPS. The Train A and Train B duct banks 

are separated by a 12-inch thick reinforced concrete wall at each pair of manholes. One 

manhole in each pair contains a sump pump and is connected to the other manhole by a 2-inch 

diameter scupper opening. There are no other openings in the common wall separating the 

Train A and Train B manholes.

Cable insulation is the only combustible material in the yard duct banks where they share a 

common wall. The sump pumps are the only equipment in the yard duct banks where they 

share a common wall.

TVA stated that a postulated fire in the cable insulation of one duct bank or in the sump pump 

will not propagate through the scupper openings to the adjacent duct bank due to the lack of 

continuity of combustible materials between duct banks.

Appendix A to BTP (APCSB) 9.5-1, Position D.1.j states that penetrations in fire barriers, 

including conduits and piping, should be sealed or closed to provide a fire resistance rating at 

least equal to that of the fire barrier itself and that the fire hazard in each area should be 

evaluated to determine barrier requirements. Penetrations in fire barriers are required to be 

rated as they are a means to limit fire damage so that the capability to shutdown the plant 

safety is maintained. Although TVA is proposing four unrated scupper openings in the Train A 

and Train B yard duct banks that run from the auxiliary building to the IPS where they share a 

common wall in three manholes, the NRC staff concludes that this is acceptable because of the 

lack of combustibles, the lack of ignition sources, and the inability of a fire to propagate from 

one duct bank to the other. Therefore, the NRC staff concludes that there is reasonable 

assurance that the unrated penetrations would not impact fire damage limits or the capability to 

safely shutdown the plant.

6.2.7.3 Auxiliary Building Penetrations

In FPR Part VII, Section 2.6.3, TVA described the following unprotected openings in the 

auxiliary building:

• Open Stairs and Hatches. TVA stated that water curtains designed in accordance with

NFPA 13-1974, Section 4-4.8.2, have been installed to protect the openings listed in

FPR Part VII, Section 2.6.3.1.

• Sheet Metal Ducts That Are Not Provided with Fire Dampers. TVA stated that these

ducts are constructed of minimum 22 gauge sheet metal, are securely fastened to the

fire barrier with angle steel, and that automatic suppression and detection is provided on

at least one side of the opening. Finally, TVA stated that the safe shutdown analysis

considered these openings as unprotected and ensured that a fire on either side of the

opening would not impact both paths of redundant safe shutdown components, cables,

or equipment.
FF-100

• Round HVAC Ducts Constructed of Spiral Welded Pipe or Schedule 10 Piping. TVA

stated that these ducts are treated as normal mechanical penetrations with appropriate

fire rated mechanical penetration seals.

• Spare Conduit Sleeves. As described in FPR Part VII, Section 2.6.3.3, TVA stated that

spare conduit sleeves which penetrate fire barriers are provided with approved sealant

material, capped on each end with metal caps or plugs, or a combination of the two.

• Unrated Steel Hatches into Monolithic Concrete Enclosures. As described in FPR

Part VII, Section 2.6.3.4, TVA stated that the monolithic enclosures in which the steel

hatches are located are not open to other rooms on other elevations. Further, TVA

stated that there are no safe shutdown cables or components within the monolithic

enclosures.

Appendix A to BTP (APCSB) 9.5-1, Position D.1.j states that penetrations in fire barriers, 

including conduits and piping, should be sealed or closed to provide a fire resistance rating at 

least equal to that of the fire barrier itself and that the fire hazard in each area should be 

evaluated to determine barrier requirements. Appendix A to BTP (APCSB) 9.5-1, Position D.1.j 

states that penetrations for ventilation systems should be protected by a standard "fire door 

damper" where required. Penetrations in fire barriers are required to be rated as they are a 

means to limit fire damage so that the capability to shutdown the plant safety is maintained. 

Although TVA is proposing open stairs and hatches, sheet metal ducts without fire dampers, 

round HVAC ducts constructed of spiral welded pipe or schedule 10 piping, spare conduit 

sleeves, and unrated steel hatches into monolithic concrete enclosures, in the Auxiliary 

building, the NRC staff concludes that these configurations are acceptable because of either 

the installed water curtains, the physical configuration of the ducts, the installed automatic 

detection and suppression systems, the installed sealant material and/or caps and plugs, or the 

lack of safe shutdown cables or components. Therefore, the NRC staff concludes that there is 

reasonable assurance that these configurations would not impact fire damage limits or the 

capability to safely shutdown the plant.

6.2.7.4 Control Building Equipment Hatches to the Turbine Building

In FPR Part VII, Section 2.6.4, TVA stated that the mechanical equipment rooms in the control 

building (Rooms 692.0-C1 and 692.0-C10) are provided with equipment hatches in the ceiling 

separating them from the turbine building. The equipment hatches have flush fitting steel

covers which are not fire rated. TVA stated that the covers are vital area boundaries with 

access control and security features attached to the undersides, to prevent inadvertent 

removal.

TVA stated that the covers do not form a watertight seal, but will limit any flammable and 

combustible liquid spills through the hatch openings into the control building mechanical 

equipment rooms. Seepage could occur around the perimeter where the covers are mounted 

to the floor and through the small diameter holes in the covers that are provided to facilitate 

their removal.

TVA stated that there are no safe shutdown components in the turbine building within 20 feet of 

the equipment hatches, so that a fire that spreads up into the turbine building will not impact 

FSSD capability. Further, TVA stated that the mechanical equipment rooms are provided with 

automatic detection and preaction sprinkler systems, including sidewall heads in the vicinity of 
FF-101

the hatches. TVA stated that the installed detection and suppression systems would control or 

extinguish postulated fires passing through the hatch covers prior to arrival of the fire brigade.

Appendix A to BTP (APCSB) 9.5 1, Position D.1.j states that penetrations in fire barriers, 

including conduits and piping, should be sealed or closed to provide a fire resistance rating at 

least equal to that of the fire barrier itself and that the fire hazard in each area should be 

evaluated to determine barrier requirements. Penetrations in fire barriers are required to be 

rated as they are a means to limit fire damage so that the capability to shutdown the plant 

safety is maintained. Although TVA is proposing unrated equipment hatches between the 

mechanical equipment rooms of the control building and the turbine building, the NRC staff 

concludes that these hatches are acceptable because of the hatch configuration, the separation 

between FSSD equipment, and the installed fire protection systems. Therefore, the NRC staff 

concludes that there is reasonable assurance that these configurations would not impact fire 

damage limits or the capability to safely shutdown the plant.

6.2.7.5 1-1⁄2 Hour Rated Fire Dampers

6.2.7.5.1 Fire Damper 0-ISD-30-620

In FPR Part VII, Section 4.7, TVA stated that fire damper 0-ISD-30-620 located in the 3-hour 

fire-rated floor separating the DG building corridor (Room 742.0-D9A) from the Unit 1 A-A DG 

exhaust room (Room 760.5-D3) is a 1-1/2-hour fire resistance rated damper. TVA also stated 

that there is detection installed in both rooms and an automatic preaction sprinkler system in the 

corridor and that manual suppression capability is available from hose stations and hydrants at 

the DG building.

TVA stated that the DG building corridor has a low in situ combustible loading (less than 1 hour) 

and the DG exhaust room has a moderate in situ combustible loading (less than 2 hours). In 

accordance with NFPA Standard 90A, “Standard for the Installation of Air-Conditioning and 

Ventilating Systems,” a fire damper with a 1-1/2-hour fire resistance rating is appropriate for 

installation in barriers with a fire resistance rating of 2 hours.

TVA determined that a fire in the DG building will not result in a loss of offsite power. TVA also 

stated that there is no equipment in the DG building required to achieve post-FSSD for a fire in 

the DG building. The only time these rooms would be needed for post-FSSD is for fires in other 

buildings and areas of the plant.

Appendix A to BTP (APCSB) 9.5-1, Position D.1.j states that penetrations for ventilation

systems should be protected by a standard "fire door damper" where required. Penetrations in 

fire barriers are required to be rated as they are a means to limit fire damage so that the 

capability to safely shutdown the plant is maintained. Although TVA is proposing the use of 1-

1/2 hour fire resistance rated dampers in the 3-hour rated barrier between the DG building 

corridor and the Unit 1 A-A DG exhaust room, the NRC staff concludes that this is acceptable 

because of the installed fire protection features, the hazards present in the rooms, and the lack 

of equipment needed for post-FSSD. Therefore, the NRC staff concludes that there is 

reasonable assurance that the installed 1-1/2 hour fire resistance rated dampers would not 

impact fire damage limits or the capability to safely shutdown the plant.

6.2.7.5.2 Fire Dampers 0-ISD-31-2427 and 0-ISD-31-2429
FF-102

In FPR Part VII, Section 4.7, TVA stated that fire dampers 0-ISD-31-2427 and 0-ISD-31-2429 

located in the 3-hour fire-rated wall separating the waste packaging room (Room 729.0-A4) from 

the condensate demineralizer waste evaporator (CDWE) building are 1-1/2-hour fire resistance 

rated dampers. TVA also stated that there is detection and an automatic preaction sprinkler 

system installed in the waste packaging room, and that manual suppression capability is 

available from hose stations in the CDWE and the auxiliary building railroad bay (an adjoining 

room to the waste packaging room).

TVA stated that the waste packaging room has a moderate (less than 2 hours) in situ 

combustible loading and the CDWE has a low (less than 1 hour) in situ combustible loading. 

In accordance with NFPA Standard 90A, “Standard for the Installation of Air-Conditioning and 

Ventilating Systems,” a fire damper with a 1-1/2-hour fire resistance rating is appropriate for 

installation in barriers with a fire resistance rating of 2 hours.

TVA stated that a fire in either room does not adversely impact post-FSSD of the plant since 

there is no equipment in either the CDWE or the waste packaging room that is needed for 

post-FSSD for a fire in any plant area.

Appendix A to BTP (APCSB) 9.5-1, Position D.1.j states that penetrations for ventilation

systems should be protected by a standard "fire door damper" where required. Penetrations in 

fire barriers are required to be rated as they are a means to limit fire damage so that the 

capability to safely shutdown the plant is maintained. Although TVA is proposing the use of 1-

1/2 hour fire resistance rated dampers in the 3-hour rated barrier between the waste packaging 

room and the CDWE building, the NRC staff concludes that this is acceptable because of the 

installed fire protection features, the hazards present in the rooms, and the lack of equipment 

needed for post-FSSD. Therefore, the NRC staff concludes that there is reasonable assurance 

that the installed 1-1/2 hour fire resistance rated dampers would not impact fire damage limits or 

the capability to safely shutdown the plant.

6.2.8 Evaluation – Large Fire Dampers 

TVA committed to the guidance in Position D.1.j of Appendix A to BTP (APCSB) 9.5-1, which 

states that fire dampers should be tested and approved by a nationally recognized laboratory 

and the tests shall bound the installed configurations. In FPR Part VII, Section 3.4, TVA stated 

that in a December 12, 1984, report, UL stated that the maximum sizes of dampers covered by 

its classification and followup service program are 90 inches wide by 72 inches high in multiple 

assemblies (maximum sections being 30 inches wide by 36 inches high) and that dampers 

exceeding this are not eligible to be labeled. Contrary to this, fire dampers 1-ISD-31-3807 and 

2-ISD-31-3882 consist of four 24-inch-wide and 24-1/2-inch-high damper sections resulting in 

an opening 98-5/8 inches wide by 24-1/2 inches high. This exceeds the UL-rated damper width 

by 8-5/8 inches.

TVA further stated that fire tests reports dated June 15 and July 19, 1984, document the 

results of tests conducted by UL for Ruskin (the damper manufacturer) on large-size damper 

installations. The large damper configurations in the two tests (100 inches by 91 inches and 

100 inches by 72 inches) both passed the 3-hour fire endurance acceptance criteria by 

remaining in place and not having an opening in the damper configuration. Both configurations, 

however, failed the hose stream test at the end of the 3-hour fire exposure. The report dated 

December 12, 1984, documented UL’s evaluation of WBN’s installation of the large dampers.
FF-103

The large fire damper installations at WBN are constructed from individual damper sections 

which are smaller than the maximum allowed by UL. The UL-listed assembly is three sections 

wide by two sections high, but the WBN configuration is one section high and four sections 

wide, thus making the assembly more rigid and less susceptible to buckling and twisting under 

actual fire conditions. Also, the individual damper sections are 24 inches wide by 24-1/2 inches 

high, which are less than the UL-allowable 30 inches wide by 36 inches high. The overall 

damper height is 24-1/2 inches high, and the UL-allowable height is 72 inches, when 

two 36-inch dampers are stacked.

In the December 12, 1984, report, UL indicated that the WBN dampers (98-5/8 inches wide by 

24-1/2 inches high) should have significantly less buckling and twisting of the vertical mullions 

than the tested damper (91 inches wide and 72 inches high) noted in the June 15, 1984, report. 

UL also concluded that the large damper installations at WBN provide adequate protection for 

their HVAC penetration.

Appendix A to BTP (APCSB) 9.5-1, Position D.1.j, states that fire dampers should be tested and 

approved by a nationally recognized laboratory and the tests shall bound the installed 

configurations. Penetrations in fire barriers are required to be rated as they are a means to limit 

fire damage so that the capability to safely shutdown the plant is maintained. Although TVA is 

proposing the use of large fire dampers that are not tested and approved by a nationally 

recognized laboratory, the NRC staff concludes that this is acceptable because of the physical 

configuration of the dampers, and because of the UL conclusion that the large damper 

installations provide adequate protection for the HVAC penetrations. Therefore, the NRC staff 

concludes that there is reasonable assurance that the large fire dampers would not impact fire 

damage limits or the capability to safely shutdown the plant.

6.2.9 Evaluation – Emergency Diesel Generators 7 Day Storage Tanks 

TVA committed to the guidance in Position F.10 of Appendix A to BTP (APCSB) 9.5-1, which 

states that diesel fuel oil tanks with a capacity of over 1,100 gallons should not be located inside 

buildings containing safety-related equipment. If located inside such buildings, the tanks should 

be separated by 3-hour fire barriers. Buried tanks are considered to meet the 3-hour fire 

resistance rating.

In FPR Part VII, Section 4.4, TVA stated that there are four 7-day (70,248-gallon) storage tank 

assemblies, one per DG, that are almost entirely buried below the floor of the DG building. The 

fuel oil storage assembly for each DG consists of four interconnected tanks, each with its own 

manway access openings, one at either end of the tank. There are a total of 16 manway access 

openings to the tanks from the corridor and four in each DG room. The manway access 

openings are the only portion of the tanks that are not buried underneath the floor of the 

DG building.

Each manway access opening is in a pit covered by a removable plate cover sitting over the top 

of the pit flush with the floor. The cover is 1/4-inch-thick steel plate, secured to the top of the 

tank by eighteen (18) 1/2-inch bolts. There are three normally closed openings in the cover 

plates in the corridor. Two of the openings are provided for fuel oil circulation, and the other

is for taking fuel oil samples.

The Pipe Gallery and Corridor (Room 742.0-D9) and DG Units 1A-A, 2A-A, 1B-B, and 2B-B 

(Rooms 742.0-D4, D5, D6 and D7) are provided with full area detection and automatic 
FF-104

suppression systems. The DG units each have heat detectors and a total flooding CO2

suppression system. Standpipe and hose stations are provided within the DG building on both 

elevations, and there are also fire hydrants available in the yard. The Pipe Gallery and Corridor 

has smoke detectors and an automatic preaction sprinkler system. A standpipe and hose 

station is provided in the Pipe Gallery and Corridor.

Fire effects on the emergency diesel generators and associated cables in the DG building will 

not have an adverse effect on safe shutdown. The DGs are not credited for any fire in the DG

building. The DG building is located remotely from other buildings containing equipment or 

cables needed for safe shutdown. This is because offsite power capabilities have been 

evaluated and determined not to be affected or required for a fire in the DG building, including 

the corridor.

Appendix A to BTP (APCSB) 9.5-1, Position F.10, states that diesel fuel oil tanks with a capacity 

of over 1,100 gallons should not be located inside buildings containing safety-related 

equipment, that if located inside such buildings, the tanks should be separated by 3-hour fire 

barriers, and that buried tanks are considered to meet the 3-hour fire resistance requirements. 

The proper placement of diesel fuel oil tanks is a means to limit fire damage so that the 

capability to safely shutdown the plant is maintained. Although TVA is proposing the use of four 

partially buried tanks in the DG building, the NRC staff concludes that this configuration is 

acceptable because of the physical construction of the manway access openings, the manway 

access openings being the only portion of the tanks that are not buried, the installed detection 

and suppression systems, the DGs not being required for any fire in the DG building, location of 

the DG building, and the offsite power capabilities not being affected by a DG building fire. 

Therefore, the NRC staff concludes that there is reasonable assurance that the partially buried 

tanks in the DG building would not impact fire damage limits or the capability to safely shutdown 

the plant.

6.2.10 Evaluation – Fire Dampers in the VCT Room Doors

In FPR Part VII, Section 3.5, TVA stated that a fire damper in the door connecting each of the 

two volume control tank (VCT) rooms with the associated pipe gallery has been changed from a 

blade-type to a curtain-type configuration. The new dampers are damper/sleeve assemblies, 

installed with the damper inside the doors. The sleeve extends a short distance on each side of 

the opening. The door was tested with the original damper, but not with the new damper.

TVA provided the following details regarding these configurations:

• The combustible loading in the immediate vicinity of the doors is insignificant.

• The new dampers are listed dampers.

• The rooms on both sides of the doors are provided with automatic fire detection and 

suppression, with the exception of the VCT rooms’ entrance labyrinths.

Appendix A to BTP (APCSB) 9.5-1, Position D.1.j, states that door openings should be 

protected with equivalent rated doors, frames, and hardware that have been tested and 

approved by a nationally recognized laboratory and that such doors should be normally closed 

and locked or alarmed with alarm and annunciation in the control room. Penetrations in fire 
FF-105

barriers are required to be rated as they are a means to limit fire damage so that the capability 

to safely shutdown the plant is maintained. Although TVA is proposing the use of untested 

doors connecting each of the two VCT rooms with the associated pipe gallery, the NRC staff 

concludes that these doors are acceptable because of the physical configuration, the installed 

fire protection systems, and the limited combustibles in the area. Therefore, the NRC staff 

concludes that there is reasonable assurance that use of these doors would not impact fire 

damage limits or the capability to safely shutdown the plant.

6.2.11 Evaluation – Plexiglas Windows in the Radiation Protection Control Point 

Building on the Refueling Floor 

TVA committed to the guidance in Position D.1.d in Appendix A to BTP (APCSB) 9.5-1, which 

states, in part, that interior finishes should be noncombustible or have a flame spread rating of 

25 or less. 

In FPR Part VIII, TVA stated that, contrary to the guidance, the windows in the radiation 

protection control point building (on the 757.0 foot elevation on the Refueling Floor) was built 

with Plexiglas windows, which do not meet the flame spread criteria. TVA stated the following 

concerning the Plexiglas windows:

• Based on operating experience at Sequoyah Nuclear Plant (i.e., a near-miss 

incident), glass windows pose a safety concern.

• Available alternatives either do not meet the flame spread criteria, or are not 

sufficiently transparent.

• The Plexiglas windows add an insignificant amount of combustibles to a large room. 

• The Plexiglas windows have no effect on the safe shutdown analysis.

• The building is not used for safe shutdown.

Appendix A to BTP (APCSB) 9.5-1, Position D.1.d states, in part, that interior finishes should be 

noncombustible or have a flame spread rating of 25 or less. Use of non-combustible interior 

finishes or interior finishes with a flame spread rating of 25 or less is required in order to limit fire 

damage so that the capability to safely shutdown the plant is maintained. Although TVA is 

proposing the use of Plexiglas windows that do not meet the flame spread criteria in the 

radiation protection control point building, the NRC staff concludes that this is acceptable 

because the use of Plexiglas windows only adds an insignificant amount of combustibles to a 

large room, the use of Plexiglas windows has no effect on the safe shutdown analysis, and 

because the building is not used for safe shutdown. Therefore, the NRC staff concludes that 

there is reasonable assurance that use of Plexiglas would not impact fire damage limits or the 

capability to safely shutdown the plant.
FF-106

6.3 Additional Engineering Evaluations

6.3.1 Evaluations Related to the Reactor Buildings’ Equipment Hatches

6.3.1.1 Relaxation of FPR Surveillance Frequencies for the Reactor Buildings’ Equipment 

Hatches

FPR Part VII, Section 6.1, summarizes TVA’s evaluation of relaxing the surveillance frequencies 

for fire protection features (smoke detectors, sprinklers, Thermo-Lag, penetration seals) from 

their regular schedules elsewhere in the plant for the equipment hatches (Rooms 757.0-A11 

and -A15). TVA stated that these actions will be performed during outages, because these 

areas are inaccessible high-radiation areas while the associated unit is operating.

These rooms connect the refueling floor and the reactor buildings and provide equipment 

access. TVA stated that the rooms are constructed of reinforced concrete and are provided with 

smoke detectors and automatic preaction sprinkler systems. The rooms are inaccessible during 

plant operations by the closure of the equipment hatch doors and placement of concrete shield 

blocks. FPR Part VI states that the rooms’ barriers are 3-hour fire-rated, with the exception of 

the blast door into the reactor building. TVA stated that these doors are of heavy metal 

construction that would prevent a fire from propagating from either the reactor building into the 

room or from the room into the reactor building. TVA further stated that combustible loading in 

the rooms is comprised of cable insulation, light covers, and Thermo-Lag (Room 757.0-A11 

only), and that there are no ignition sources in the rooms during power operation.

Periodic surveillances of fire protection system and equipment are required in order to ensure 

that the installed fire protection features are available to limit fire damage so that the capability 

to safely shutdown the plant is maintained. Although TVA is proposing to only conduct fire 

protection systems and equipment surveillances for the equipment hatches (Rooms 757.0-A11 

and -A15) during outages, the NRC staff concludes that this is acceptable because the rooms 

are not accessible during normal plant operations, the rooms are equipped with 3-hour fire rated 

barriers, the limited amount of combustible loading in the rooms, and the lack of ignition 

sources. Therefore, the NRC staff concludes that there is reasonable assurance that the 

reduced fire protection surveillances would not impact fire damage limits or the capability to 

safely shutdown the plant.

6.3.1.2 Elimination of Certain Compensatory Measures for the Reactor Buildings’ Equipment 

Hatches

FPR Part VII, Section 6.1, also summarizes TVA’s evaluation of not performing certain 

compensatory measures for out of service fire protection features (smoke detectors, sprinklers) 

for the equipment hatches (Rooms 757.0-A11 and -A15). TVA stated that compensatory 

measures performed inside these rooms are impractical when these rooms are inaccessible 

high radiation areas, that is, while the associated unit is operating.

These rooms connect the refueling floor and the reactor buildings, and provide equipment 

access. TVA stated that the rooms are constructed of reinforced concrete and are provided 

with smoke detectors and automatic preaction sprinkler systems. The rooms are inaccessible 

during plant operations by the closure of the equipment hatch doors and placement of concrete 

shield blocks. FPR Part VI states that the rooms’ barriers are 3-hour fire-rated, with the 

exception of the blast door into the reactor building. TVA stated that these doors are of heavy 
FF-107

metal construction that would prevent a fire from propagating from either the reactor building 

into the room or from the room into the reactor building. TVA further stated that combustible 

loading in the rooms is comprised of cable insulation, light covers, and Thermo-Lag (Room 

757.0-A11 only), and that there are no ignition sources in the rooms during power operation.

The performance of compensatory measures for out of service fire protection features are 

required in order to maintain some level of fire protection to limit fire damage so that the 

capability to safely shutdown the plant is maintained. Although TVA is proposing to not perform 

compensatory measures for out of service fire protection features (smoke detectors, sprinklers) 

for the equipment hatches (Rooms 757.0-A11 and -A15) during plant operation, the NRC staff 

concludes that this is acceptable because the rooms are not accessible during normal plant 

operations, the rooms are equipped with 3-hour fire rated barriers, the limited amount of 

combustible loading in the rooms, and the lack of ignition sources. Therefore, the NRC staff 

concludes that there is reasonable assurance that not performing certain compensatory 

measures during plant operations would not impact fire damage limits or the capability to safely 

shutdown the plant.

6.3.2 Relaxation of FPR Surveillance Period for Fire Dampers in High Radiation and 

Contaminated Areas

In FPR Part VII, Section 6.2, TVA evaluated the need to perform surveillance for fire dampers in 

high radiation or contaminated areas. TVA evaluated the consequences of the failure of the 

following fire dampers to close during a fire event: 0-ISD-31-3846, 0-ISD-31-3847, and 

0-ISD-31-3848. TVA stated that these fire dampers are located in contaminated areas and are 

considered to be inaccessible.

6.3.2.1 Fire Damper 0-ISD-31-3846

TVA stated that fire damper 0-ISD-31-3846 is located in a 24-inch-diameter embedded duct that 

starts at an embedded collector box located in the Fuel Transfer Canal wall and runs for 40 feet 

where it exits the concrete wall of the VPA room (Room 737.0-A5) and then enters a large 

(64-inch by 54-inch) duct.

TVA also stated that there is no combustible hazard in the fuel transfer canal, and negligible 

quantities of combustibles in the vicinity of the duct in the VPA room. TVA further stated that 

the room is provided with smoke detection and automatic suppression. Finally, TVA stated that 

should a fire breach the walls of the duct in the VPA room, the fire would have to travel a 

distance of 40 feet to reach the fuel transfer canal.

Periodic surveillances of fire protection system and equipment are required in order to ensure 

that the installed fire protection features are available to limit fire damage so that the capability 

to safely shutdown the plant is maintained. Although TVA is not proposing to conduct periodic 

surveillance of fire dampers in high radiation or contaminated areas, the NRC staff concludes 

that this is acceptable because of the limited combustibles in each room, the distance the fire 

would have to travel to reach the other room, the automatic suppression installed in the VPA 

room, and the ALARA concern identified by TVA. The NRC staff also concludes that not 

performing surveillance of this fire damper is consistent with Interpretation 4, “Fire Area 

Boundaries,” of GL 86-10 and, therefore, is acceptable. Therefore, the NRC staff concludes 

that there is reasonable assurance that not conducting these fire protection surveillances would 

not impact fire damage limits or the capability to safely shutdown the plant.
FF-108

6.3.2.2 Fire Dampers 0-ISD-31-3847 and 0-ISD-31-3848

TVA stated that one of the fire dampers is located in a 24-inch-diameter embedded duct that 

starts at an embedded collector box located in the spent fuel pit wall, runs for approximately 

5 feet where it exits the concrete wall, traverses a corridor, and penetrates the concrete wall of 

the VPA room, and then enters a large (58-inch by 54-inch) duct. TVA stated that the other fire 

damper is located in a 30-inch-diameter embedded duct that starts at an embedded collector 

box located in the opposite wall of the spent fuel pit, runs for approximately 80 feet where it exits 

the spent fuel pit wall (near the 24-inch duct), traverses the corridor, and penetrates the wall of 

the VPA room and enters the large duct.

TVA stated that both ducts are coated with 2 inches of fire protective material (Pyrocrete) where 

they traverse the corridor. Further, TVA stated that there are no combustible hazards in the 

spent fuel pit and negligible quantities of combustibles in the vicinity of the ducts in the corridor 

and near the ducts in the VPA room. In addition, the corridor and the VPA room are provided 

with smoke detection, and the VPA room is also provided with automatic suppression. Finally, 

TVA indicated that, should a fire breach the walls of the ducts in the VPA room, the fire would 

have to travel the width of the corridor plus a distance of 5 feet (24-inch duct) or 80 feet (30-inch 

duct) to reach the spent fuel pit, which is filled with water.

Periodic surveillances of fire protection system and equipment are required in order to ensure 

that the installed fire protection features are available to limit fire damage so that the capability 

to safely shutdown the plant is maintained. Although TVA is not proposing to conduct periodic 

surveillance of fire dampers in high radiation or contaminated areas, the NRC staff concludes 

that this is acceptable because of the limited combustibles in each room, the distance the fire 

would have to travel to reach the other room, the automatic suppression installed in the VPA 

room, and the ALARA concern identified by TVA. The NRC staff also concludes that not 

performing surveillance of this fire damper is consistent with Interpretation 4, “Fire Area 

Boundaries,” of GL 86-10 and, therefore, is acceptable. Therefore, the NRC staff concludes 

that there is reasonable assurance that not conducting these fire protection surveillances would 

not impact fire damage limits or the capability to safely shutdown the plant.

6.3.3 Gap between Door and Frame for Fire Door W9

In FPR Part VII, Section 6.3, TVA stated that a portion of the gap between the door and frame 

of fire door W9 exceeds the maximum 3/16-inch clearance. TVA further stated that the fire 

door is located in the wall that separates the raw cooling water (RCW) pump deck from the 

Train A ERCW pump room. TVA stated the following concerning the environment of door W9:

• The RCW pump deck is open to the atmosphere on three sides and does not have a 

roof. 

• The ERCW pump room does not have a roof.

• The nearest RCW pump is located 17 feet horizontally from the door and the bottom of 

the door is 13.5 feet above the RCW pump deck.

 

• The in situ combustible load of the RCW pump deck consists primarily of lube oil 

associated with the RCW pumps. 
FF-109

• There are no in situ combustibles located directly under the door and the stairs and 

landings prevent any appreciable quantities of transient combustibles from being stored 

under the door. 

• The door opens into a labyrinth that does not contain any in situ combustibles, nor are 

transient combustibles stored in the labyrinth.

Gaps between fire doors and frames are required to be limited in size as a means to limit fire 

damage so that the capability to safely shutdown the plant is maintained. Although TVA is 

proposing a larger than required door gap between the door and frame for fire door W9, the 

NRC staff concludes that this is acceptable because of the physical configuration that would 

prevent the formation of a hot gas layer, the distance to the nearest source of combustibles, and 

the limited amount of combustibles in this area. The NRC staff also concludes that this is 

consistent with Interpretation 4, “Fire Area Boundaries,” of GL 86-10. Therefore, the NRC staff 

concludes that there is reasonable assurance that the excessive door gap would not impact fire 

damage limits or the capability to safely shutdown the plant.

6.3.4 Relaxation of FPR Surveillance Period for Penetration Seals in High Radiation and 

Contaminated Areas

In FPR Part VII, Section 6.4, TVA evaluated the need to perform surveillance for penetration 

seals in high radiation areas by evaluating the consequences of the failure of the penetration 

seals for each of the rooms. TVA stated that its evaluations considered the locations not 

inspected, the proximity of combustibles, and the construction features of the rooms on either 

side of the seals.

6.3.4.1 Spent Resin Tank Room (Room 692.0-A15)

TVA stated that the penetration seals of interest in Room 692.0-A15 are installed in the wall 

separating it from the pipe gallery and chase room (Room 692.0-A24), which is a 2-hour rated 

fire barrier of reinforced concrete construction. TVA stated that the penetration seals are 

accessible for surveillance inspection from Room 692.0-A24, however, they are not accessible 

for inspection from the spent resin tank room due to the radiation posting of the room.

TVA stated that there is no safe shutdown equipment in the spent resin tank room. FPR Part VI 

stated that the combustible loading in both rooms is insignificant. TVA also stated that there is 

smoke detection installed in Room 692.0-A24.

Periodic surveillances of fire protection system and equipment are required in order to ensure 

that the installed fire protection features are available to limit fire damage so that the capability 

to safely shutdown the plant is maintained. Although TVA is not proposing to conduct periodic 

surveillance of penetration seals in high radiation areas, the NRC staff concludes that this is 

acceptable because of the minimal amount of combustibles in each room, the lack of safe 

shutdown equipment or cables in the spent resin tank room, the automatic smoke detection 

installed in the pipe gallery and chase room, and the ALARA concern identified by TVA. The 

NRC staff also concludes that not performing surveillance of this penetration seal is consistent 

with Interpretation 4, “Fire Area Boundaries,” of GL 86-10. Therefore, the NRC staff concludes 

that there is reasonable assurance that not conducting these fire protection surveillances would 

not impact fire damage limits or the capability to safely shutdown the plant.
FF-110

6.3.4.2 Tritiated Drain Collector Tank Room (Room 674.0-A1)

TVA stated that the penetration seals of interest in Room 674.0-A1 are installed in the wall 

separating it from the RHR Pump Room 1A-A (Room 676.0-A11) which is a 2-hour fire rated 

barrier of reinforced concrete construction. TVA stated that the penetration seals are 

accessible for surveillance inspection from Room 676.0-A11; however, they are not accessible 

for inspection from Room 674.0-A1 due to the radiation posting of the room. 

TVA stated that there is no safe shutdown equipment installed in the tritiated drain collector 

tank room which is required for a fire in the auxiliary building. FPR Part VI stated that the 

combustible loading in both rooms is insignificant. TVA also stated that there is smoke 

detection installed in Room 676.0-A11.

Periodic surveillances of fire protection system and equipment are required in order to ensure 

that the installed fire protection features are available to limit fire damage so that the capability 

to safely shutdown the plant is maintained. Although TVA is not proposing to conduct periodic 

surveillance of penetration seals in high radiation areas, the NRC staff concludes that this is 

acceptable because of the limited amount of combustibles in each room, the lack of safe 

shutdown equipment or cables in the tritiated drain collector tank room, the automatic smoke 

detection installed in the RHR Pump Room 1A-A, and the ALARA concern identified by TVA. 

The NRC staff also concludes that not performing surveillance of these penetration seals is 

consistent with Interpretation 4, “Fire Area Boundaries,” of GL 86-10. Therefore, the NRC staff 

concludes that there is reasonable assurance that not conducting these fire protection 

surveillances would not impact fire damage limits or the capability to safely shutdown the plant.

6.3.4.3 Hold Up Tank Rooms A and B (Rooms 676.0-A2 and 676.0-A3)

TVA stated that Rooms 676.0-A2 and 676.0-A3 are separated from adjacent non-high-radiation 

area rooms by 2- and 3-hour fire-rated barriers of reinforced concrete construction. TVA stated 

that the penetration seals are accessible for surveillance inspection from these adjacent rooms. 

The penetrations are not accessible from inside the holdup tank rooms for surveillance 

inspection due to the radiation posting of the rooms.

TVA stated that there is no safe shutdown equipment installed in the holdup tank rooms, nor 

any equipment that could initiate a plant trip. FPR Part VI stated that the combustible loading in 

both rooms is insignificant. Additionally, TVA stated that all the adjacent rooms which contain 

cables or equipment needed for FSSD have installed smoke detection, with the exception of 

Room 692.0, which contains one cable related to FSSD. TVA further stated that fire damage to 

this cable will not cause spurious operation or otherwise prevent safe shutdown.

Periodic surveillances of fire protection system and equipment are required in order to ensure 

that the installed fire protection features are available to limit fire damage so that the capability 

to safely shutdown the plant is maintained. Although TVA is not proposing to conduct periodic 

surveillance of penetration seals in high radiation areas, the NRC staff concludes that this is 

acceptable because of the limited amount of combustibles in these rooms, the lack of safe 

shutdown equipment or cables in the holdup tank room, the automatic smoke detection installed 

in the adjacent rooms which contain FSSD equipment or cables, and the ALARA concern 

identified by TVA. The NRC staff also concludes that not performing surveillance of these

penetration seals is consistent with Interpretation 4, “Fire Area Boundaries,” of GL 86-10. 
FF-111

Therefore, the NRC staff concludes that there is reasonable assurance that not conducting 

these fire protection surveillances would not impact fire damage limits or the capability to safely 

shutdown the plant.

6.3.4.4 Gas Decay Tank Rooms (Rooms 692.0-A3 and 692.0-A5)

TVA stated that Rooms 692.0-A3 and 692.0-A5 are separated from adjacent non-high radiation 

area rooms by 2- and 3-hour fire rated barriers of reinforced concrete construction. TVA stated 

that the penetration seals are accessible for surveillance inspection from these adjacent rooms. 

The penetration seals are not accessible for inspection from the gas decay tank rooms due to 

the radiation posting of the rooms.

TVA stated that there is no safe shutdown equipment installed in the gas decay tank rooms, nor 

any equipment that could initiate a plant trip. TVA also stated that Room 692.0 contains one 

cable related to FSSD, but that fire damage to this cable will not cause spurious operation or 

otherwise prevent safe shutdown. FPR Part VI stated that the combustible loading in both 

rooms is insignificant. Additionally, TVA stated that all the adjacent rooms that contain cables 

or equipment needed for FSSD have installed automatic smoke detection.

Periodic surveillances of fire protection system and equipment are required in order to ensure 

that the installed fire protection features are available to limit fire damage so that the capability 

to safely shutdown the plant is maintained. Although TVA is not proposing to conduct periodic 

surveillance of penetration seals in high radiation areas, the NRC staff concludes that this is 

acceptable because of the limited amount of combustibles in these rooms, the lack of safe 

shutdown equipment or cables in the gas decay tank rooms, the automatic smoke detection 

installed in the adjacent rooms that contain FSSD equipment or cables, and the ALARA 

concern identified by TVA. The NRC staff also concludes that not performing surveillance of 

these penetration seals is consistent with Interpretation 4, “Fire Area Boundaries,” of GL 86-10. 

Therefore, the NRC staff concludes that there is reasonable assurance that not conducting 

these fire protection surveillances would not impact fire damage limits or the capability to safely 

shutdown the plant.

6.3.4.5 Barriers between High Radiation Area Rooms (Rooms 676.0-A2, 676.0-A3, 692.0-A3 

and 692.0-A5)

TVA stated that the barriers between Rooms 676.0-A2 and 676.0-A3, Rooms 676.0-A2 and 

692.0-A3, and Rooms 692.0-A3 and 692.0-A5 are not accessible because of the high levels of 

radiation present in these rooms.

TVA stated that there is no safe shutdown equipment installed in any of these rooms, nor any 

equipment that could initiate a plant trip. TVA also stated that Room 692.0 contains one cable 

related to FSSD, but that fire damage to this cable will not cause spurious operation or 

otherwise prevent safe shutdown. FPR Part VI stated that the combustible loading in all the 

rooms is insignificant. Additionally, TVA stated that all the adjacent rooms that contain cables 

or equipment needed for FSSD have installed automatic smoke detection.

Periodic surveillances of fire protection system and equipment are required in order to ensure 

that the installed fire protection features are available to limit fire damage so that the capability 

to safely shutdown the plant is maintained. Although TVA is not proposing to conduct periodic 

surveillance of penetration seals in high radiation areas, the NRC staff concludes that this is 
FF-112

acceptable because of the limited amount of combustibles in these rooms, the lack of safe 

shutdown equipment or cables in Rooms 676.0-A2, 676.0-A3, 692.0-A3 and 692.0-A5, the 

automatic smoke detection installed in the adjacent rooms that contain FSSD equipment or 

cables, and the ALARA concern identified by TVA. The NRC staff also concludes that not 

performing surveillance of these penetration seals is consistent with Interpretation 4, “Fire Area 

Boundaries,” of GL 86-10. Therefore, the NRC staff concludes that there is reasonable 

assurance that not conducting these fire protection surveillances would not impact fire damage 

limits or the capability to safely shutdown the plant.

6.3.5 Diesel Generator Building Lube Oil Storage Room Fire Door

In FPR Part VII, Section 5.2, TVA stated that the lube oil storage room (Room 742.0-D2) has a 

3-hour fire-rated barrier separating it from the adjacent Rooms 742.0-D4 (DG 1A-A) and 

742.0-D9 (pipe gallery/corridor). There is an opening between Rooms 742.0-D9 and 742.0-D2 

that is protected with a swinging hollow metal door in the opening and a sliding fire door. There 

is a preaction sprinkler system installed in Room 742.0-D9 and a total flooding CO2 suppression 

system installed in the lube oil storage room.

The 3-hour fire-rated self-closing sliding door is held in the open position and closes only when 

a thermal link above the door melts or the CO2 suppression system for the lube oil storage room 

discharges. In addition to the sliding door, TVA installed a hollow metal side-hinged door in the 

opening, which is normally closed. TVA stated that this door is similar to rated fire doors and is 

expected to prevent smoke and hot gases from a fire from passing through the opening until the 

fusible links melt or the fire suppression system in the lube oil storage room actuates, and the 

sliding door closes.

NFPA 80 requires fusible links on both sides of a sliding fire door, but the installed configuration 

only has these links on the lube oil storage room side. The NRC staff concludes that the 

installed configuration provides an equivalent level of safety because of the presence of 

automatic suppression systems on both sides of the opening, the addition of a CO2 system 

actuation release for the sliding door, the presence of the normally closed swinging door, and 

the relatively higher fire loading in the lube oil storage room.

Based on its review of the information submitted by TVA, the NRC staff concludes that the fire 

door configuration in the lube oil storage room complies with Position D.1.j of Appendix A to 

BTP (APCSB) 9.5-1 and, therefore, is acceptable.
FF-113

7.0 CONCLUSION

The U.S. Nuclear Regulatory Commission (NRC) staff reviewed Tennessee Valley Authority’s 

(TVA’s) as-constructed Fire Protection Report (FPR) and TVA’s supplemental information as 

referenced by this evaluation. Based on its review, the NRC staff concludes, with the exception 

of WBN, Unit 1, specific operator manual actions (OMAs) which the NRC staff documented its 

approval of in SSER 18, that the fire protection program for WBN:

• meets Title 10 of the Code of Federal Regulations (10 CFR) Section 50.48(a),

• meets General Design Criterion 3 of Appendix A to 10 CFR Part 50,

• is consistent with Sections III.G, III.J, III.L, and III.O of Appendix R to 10 CFR Part 50, and 

• is consistent with Appendix A to Branch Technical Position (Auxiliary Power Conversion 

Systems Branch) 9.5-1, May 1976.

All alternatives and exceptions have been properly justified. Therefore, the NRC staff concludes 

that the as-constructed FPR is acceptable. 

GG-1 

APPENDIX GG FINAL MEMORANDUM ON FACILITY COMPLETION IN 

ACCORDANCE WITH INSPECTION MANUAL CHAPTER 94302
UNITED STATES

NUCLEAR REGULATORY COMMISSION

REGION II

245 PEACHTREE CENTER AVENUE NE, SUITE 1200

ATLANTA, GEORGIA 30303-1257

October 15, 2015

MEMORANDUM TO: William M. Dean, Director

 Office of Nuclear Reactor Regulation

FROM: Leonard D. Wert, Jr. /RA/

Acting Regional Administrator

SUBJECT: READINESS OF WATTS BAR UNIT 2 TO RECEIVE AN 

OPERATING LICENSE

Region II has completed the inspections needed to support issuing an operating license for 

Watts Bar Unit 2. NRC Inspection Procedure 94302, “Status of Watts Bar Unit 2 Readiness for 

an Operating License,” lists the NRC findings necessary before an operating license can be 

issued. Those findings as described in Title 10 of the Code of Federal Regulations (10 CFR) 

Part 50.57 are: “(a)(1) Construction of the facility has been substantially completed, in 

conformity with the construction permit and the application as amended...;” “(a)(2) the facility 

will operate in conformity with the application as amended...;” and (a)(3)(ii) there is reasonable 

assurance that facility will be operated in accordance with the regulations....” 

The NRC staff developed a plant-specific construction inspection program for Watts Bar 

Nuclear Plant (WBN) Unit 2, and specific inspection items were listed in the Inspection 

Planning and Scheduling (IP&S) data base. Items in IP&S include construction inspection 

procedures, corrective action programs and special programs Tennessee Valley Authority 

(TVA) developed to address historical quality assurance problems, generic communications, 

inspection open items, allegations and refurbishment activities. There were 560 items 

identified for WBN Unit 2. The NRC staff verified that the items needed for this readiness 

determination have been sufficiently addressed. Review and closure of the items are 

documented in 64 inspection reports issued during the past eight years. Those inspection 

reports are available on the WBN Unit 2 Reactivation website (http://www.nrc.gov/infofinder/reactor/wb/watts-bar.html). The twenty-one (21) IP&S items that have not been fully 

closed are listed in Enclosure 2, Open Items List, along with a description of why the items do 

not impact the readiness decision.

An overview of the WBN Unit 2 construction inspection program is provided in Enclosure 1. In 

addition to the IP&S items, the NRC staff has completed all pre-operational test procedure 

reviews and witnessed mandatory and safety-related system tests needed for the 10 CFR 

50.57 findings. The NRC staff will perform additional inspections as TVA’s pre-operational 

testing program is completed. Those testing activities are included in Enclosure 2.

CONTACT: William B. Jones, RII/DCP

 404-997-4200

GG-2
W. Dean 2 

Region II operational preparedness inspections also verified that TVA has adequate 

management controls and procedures, including quality assurance programs, security, 

operations, and radiological controls necessary for an operating reactor. 

In 2009, the NRC established the WBN Unit 2 Reactivation Assessment Group (WRAG), led by 

division level management from the Office of Nuclear Reactor Regulation (NRR) and Region II. 

The WRAG responsibilities include informing the Region II Regional Administrator when the 

activities in Inspection Manual Chapter (IMC) 2517, Watts Bar 2 Inspection Program, needed to 

support an operating license, have been completed. 

On August 12, 2015, TVA submitted their substantially complete letter (ADAMS Accession No.

ML15224B482) and requested that the NRC issue an operating license for WBN Unit 2. The 

staff reviewed that letter and briefed the WRAG on the results of the reviews. The staff’s review 

of the letter was also factored into the three 10 CFR 50.57 findings.

Region II inspection activities support that the construction of the WBN Unit 2 facility has been 

substantially completed in conformity with the construction permit and the application as 

amended. This conclusion is based on completion of the construction inspections necessary to 

support this finding. 

Region II inspection activities support that WBN Unit 2 facility will operate in conformity with the 

application as amended. This is based on completion of the pre-operational testing inspections 

identified in IMC 2513, Appendix A, Light Water Reactor - Preoperational Testing Phase, that 

are necessary to support this finding. The NRC staff will continue to implement IMC 2517,

through WBN Unit 2 commercial operations. This will include completion of IMC 2513, Light 

Water Reactor Inspection Program - Preoperational Testing and Operational Preparedness 

Phase, IMC 2514, Light Water Reactor Inspection Program -- Startup Testing Phase, and the 

implementation of the reactor oversight program including IMC 2515, Light-Water Reactor 

Inspection Program-Operations Phase, at the issuance of the operating license.

Finally, Region II inspection activities support that there is reasonable assurance that activities 

authorized by the license will be conducted according to applicable regulations. This is based 

on completion of operational preparedness inspections items that were necessary to support 

this finding and conclusions from the Operational Readiness Assessment Team inspection 

(ORAT). The ORAT inspection (ADAMS Accession No. ML 15226A212) concluded TVA 

adequately demonstrated the readiness of the facility and staff to safely begin operating the

WBN Unit 2 facility. 

Inspections that have not been completed are listed in Enclosure 2 along with justification that 

the open items do not impact the determination that construction of the facility has been 

substantially completed, that the facility will operate in conformity with the application and that 

activities authorized by the license will be conducted according to the applicable regulations. 

For those items listed in Enclosure 2, a sufficient amount of inspection has been completed to 

provide reasonable assurance that associated systems, structures, components, or operational 

programs will perform as described in the application or applicable regulations. 

In addition, Region II found no pending or open enforcement issues or open allegations that 

would affect a determination of reasonable assurance of the three 10 CFR 50.57 inspection 

findings. Integrated inspection report (IIR) 05000391/2015607 (ADAMS Accession No. 

ML15273A452) that was issued on September 29, 2015, contained a Severity Level IV Notice of 

Violation (NOV), involving deliberate misconduct by a contract employee. The staff conducted 

GG-3
W. Dean 3 

an initial assessment of TVA’s corrective actions while determining the appropriate enforcement 

actions for this matter. Based on this review, Region II determined this issue does not impact 

the 94302 recommendation. The NRC staff will continue review of the enforcement action in 

accordance with NRC procedures. Region II has reviewed the one currently open allegation for 

WBN Unit 2, including technical aspects and corrective actions. Region II determined the 

allegation does not impact the 94302 recommendation. The NRC staff will continue actions to 

address the allegation in accordance with our established process.

The NRC staff has assessed the safety culture for WBN Unit 2 with a variety of activities. On a 

routine bases, the resident inspectors monitored TVA’s employee concerns program to review 

concerns and trends. During the annual problem identification and resolution inspections, the 

safety conscious work environment was assessed while interviewing the construction staff and 

reviewing related documents. Allegations for WBN Unit 2 also provided insight on safety 

culture. Overall, the NRC staff determined that TVA had established an adequate safety culture 

for WBN Unit 2. 

The WRAG has reviewed the inspections completed, the items listed in Enclosure 2, and 

supports the Region II assessment that the necessary inspections and reviews have been 

completed to support the 10 CFR 50.57 (a)(1), (a)(2), and (a)(3)(ii) findings. 

Enclosures: 

As Stated

GG-4
October 15, 2015

MEMORANDUM TO: William M. Dean, Director

 Office of Nuclear Reactor Regulation

FROM: Leonard D. Wert, Jr. /RA/

Acting Regional Administrator

SUBJECT: READINESS OF WATTS BAR UNIT 2 TO RECEIVE AN

OPERATING LICENSE

Region II has completed the inspections needed to support issuing an operating license for 

Watts Bar Unit 2. NRC Inspection Procedure 94302, “Status of Watts Bar Unit 2 Readiness for 

an Operating License,” lists the NRC findings necessary before an operating license can be 

issued. Those findings as described in Title 10 of the Code of Federal Regulations (10 CFR) 

Part 50.57 are: “(a)(1) Construction of the facility has been substantially completed, in 

conformity with the construction permit and the application as amended...;” “(a)(2) the facility 

will operate in conformity with the application as amended...;” and (a)(3)(ii) there is reasonable 

assurance that facility will be operated in accordance with the regulations....” 

The NRC staff developed a plant-specific construction inspection program for Watts Bar Nuclear 

Plant (WBN) Unit 2, and specific inspection items were listed in the Inspection Planning and 

Scheduling (IP&S) data base. Items in IP&S include construction inspection procedures, 

corrective action programs and special programs Tennessee Valley Authority (TVA) developed 

to address historical quality assurance problems, generic communications, inspection open 

items, allegations and refurbishment activities. There were 560 items identified for WBN Unit 2. 

The NRC staff verified that the items needed for this readiness determination have been 

sufficiently addressed. Review and closure of the items are documented in 64 inspection 

reports issued during the past eight years. Those inspection reports are available on the WBN 

Unit 2 Reactivation website (http://www.nrc.gov/info-finder/reactor/wb/watts-bar.html). The 

twenty-one (21) IP&S items that have not been fully closed are listed in Enclosure 2, Open 

Items List, along with a description of why the items do not impact the readiness decision.

An overview of the WBN Unit 2 construction inspection program is provided in Enclosure 1. In 

addition to the IP&S items, the NRC staff has completed all pre-operational test procedure 

reviews and witnessed mandatory and safety-related system tests needed for the 10 CFR 50.57 

findings. The NRC staff will perform additional inspections as TVA’s pre-operational testing 

program is completed. Those testing activities are included in Enclosure 2.

CONTACT: William B. Jones, RII/DCP

 404-997-4200

PUBLICLY AVAILABLE NON-PUBLICLY AVAILABLE SENSITIVE NON-SENSITIVE

ADAMS: Yes ACCESSION NUMBER:ML15288A305________________ SUNSI REVIEW COMPLETE FORM 665 ATTACHED

OFFICE RII:DCP RII:DCP RII:DCP NRR NRR NRR RII:ORA RII:ORA

SIGNATURE \RA via Email RCH WJ \RA via Email \RA via Email \RA via Email LAD LW

NAME E. Patterson R. Haag W. Jones J. Poole J. Quichocho J. Trapp L. Dudes L. Wert

DATE 10/09/15 10/08/15 10/09/15 10/08/15 10/08/;15 10/08/15 10/09/15 10/15/15

E-MAIL COPY? YES NO YES NO YES NO YES NO YES NO YES NO YES NO YES NO 

OFFICIAL RECORD COPY

GG-5
Enclosure 1

Overview of the Watts Bar Unit 2 Construction and Testing Inspection Programs

Inspection Manual Chapter (IMC) 2517 was issued to establish policy for implementing the

WBN Unit 2 construction, preoperational testing, and startup testing inspection programs that 

are covered under IMCs 2512, 2513, and 2514. The construction inspection program was 

expanded to include items specific to WBN Unit 2 such as the refurbishment program and TVAspecific Corrective Action Programs (CAPs) and Special Programs (SPs). Since many of the 

WBN Unit 2 components had been installed during initial construction, the refurbishment 

program was implemented to assure the components meet or exceed their original 

specifications. The CAPs and SPs were implemented to correct historical problems 

encountered during initial construction. The construction inspection program also factored in 

both historical and current industry initiatives. Examples include motor operated valve testing 

and surveillance, potential impact of pressurized water reactor containment sump blockage, 

Fukushima response actions, and Design Vulnerability in Electrical Power System (Open Phase 

Condition).

A. Construction Phase Inspections

IMC 2512

IMC 2512 prescribes all the required NRC construction inspection procedures (IPs). Many 

of these inspections were completed or partially completed prior to suspension of WBN Unit

2 construction in the mid-1980s. To understand the amount of previously completed WBN 

Unit 2 construction inspections, the staff performed a reconstitution of the 2512 inspection

program. The reconstitution effort involved a computer assisted search of historical 

inspection reports to identify portions/ samples of 2512 IPs that had already been 

completed. Unfinished items in the IPs were noted, with the understanding that these items 

would be inspected following resumption of construction, such that all aspects of the 2512 

IPs would be accomplished at the completion of WBN Unit 2 construction. New work 

(activities performed following resumption of construction in 2008) would also be inspected 

under the 2512 IPs, even if reconstitution had determined that the applicable section(s) of 

an IP had been completed.

Additional items were considered when developing a comprehensive construction inspection 

program for WBN Unit 2. The following items were added to the WBN Unit 2 construction 

inspection program:

• Old allegations – closed allegations were screened to identify possible WBN Unit 2 

unresolved items.

• Generic Issues – The staff reviewed TVA’s responses for all applicable issued generic 

communications including generic letters, bulletins, and Three Mile Island action items. 

Generic communications that were issued after construction on WBN Unit 2 resumed in 

2008 were reviewed on a routine basis and added to the inspection program as 

applicable. Inspection was performed to address the resolution of each of the issues. 

Additionally, applicable Temporary Instructions were re-issued and utilized to perform 

the required inspections of outstanding design, licensing, and regulatory issues for WBN 

Unit 2. 

• Construction Deficiency Reports (CDRs) – CDRs issued during initial construction that

were not previously closed were reviewed to determine the current applicability on WBN 

Unit 2 construction and inspected as required.

GG-6
2 

• Inspection items identified during the licensing review process – Items identified for 

inspection follow-up in Appendix HH to NUREG-0847 "Safety Evaluation Report [SER] 

Related to the Operation of Watts Bar Nuclear Plant Units 1 and 2," and its supplements 

(SSER). 

• Inspection Open Items – Unresolved items and violations identified in inspection reports 

were tracked and inspected for closure.

Construction Refurbishment Program 

Because plant construction was inactive for a long time and the scope of equipment layup 

activities was limited, TVA developed and submitted its Construction Refurbishment 

Program to ensure that the design and licensing basis, including original equipment design 

specifications, would be met. The Construction Refurbishment Program was intended to 

refurbish or replace most active components and instruments. For other equipment, the 

program determines the potential degradation mechanism for each category of components, 

taking into account the environmental conditions, the acceptance criteria, and the 

refurbishment or inspection activities necessary to demonstrate compliance with applicable 

vendor and design specifications or requirements. The NRC staff reviewed TVA’s program 

and on July 2, 2010, issued its evaluation, which concluded that, upon proper 

implementation, the Construction Refurbishment Program would provide reasonable 

assurance that the equipment would meet its design criteria and perform its intended 

functions. A new IP 37002, “Construction Refurbishment Process—Watts Bar Unit 2,” was 

developed to provide guidance for inspection of the refurbishment program. RII conducted 

several refurbishment inspections throughout the project. The IP was closed in inspection 

report 05000391/2014605 (ADAMS Accession No. ML14226A049). 

Inspection of Nuclear Performance Plan (NPP) Implementation

To address historical construction quality issues that were identified in 1985, TVA 

implemented its NPP which included Corrective Action Programs (CAPs) and Special 

Programs (SPs) to correct the deficient areas. At that time, TVA’s NPP efforts were focused 

on WBN Unit 1. The NRC staff reviewed components of the NPP for WBN Unit 1 and 

recognized the general approaches of various corrective actions. From 1985 to 1993, the 

NRC focused on inspecting the CAPs and SPs as they applied to WBN Unit 1 and no 

conclusions were stated for WBN Unit 2. 

To address the quality issue for WBN Unit 2, TVA developed CAPs and SPs that contained 

many of the same corrective actions performed at WBN Unit 1. TVA submitted its plans in 

2008 detailing specific actions for the CAPs and SPs to resolve historical quality issues for 

WBN Unit 2. The staff reviewed these plans and determined that when implemented 

thoroughly, the proposed corrective actions should address the identified deficiencies for 

WBN Unit 2. Region II has inspected all of the CAPs and SPs and has concluded that the 

programs were adequate. Additional inspections are necessary for one CAP and two SPs 

to verify the implementation of the programs. These inspection items are listed in 

Enclosure 2.

TVA Implementation of Fukushima Actions at WBN, Unit 2

Of the three Orders issued by the NRC related to the implementation of Fukushima actions, 

two were applicable to WBN based on plant design (spent fuel pool instrumentation and 

mitigating strategies).

GG-7
3 

The Orders state that prior to issuance of an operating license, the holder of a construction 

permit must state that they fully comply with the Order via letter to the NRC. TVA submitted 

a final Compliance Letter for the Spent Fuel Pool Order on December 19, 2014 (ADAMS 

Accession No. ML 15002A202). A Full Compliance Letter for the Mitigating Strategies Order 

was submitted on March 12, 2015 (ADAMS Accession No. ML15072A116). Since that time, 

TVA has implemented several of the actions, made modifications to the plant, and acquired

new equipment. The NRC inspected a sample of these modifications focused on critical 

design and structural attributes using Temporary Instruction 2515/191. Additionally, 

inspections were performed of as-built and completed modifications to verify that 

specifications, drawings, requirements, and standards were met. The results of this 

inspection were issued on June 22, 2015, in inspection report 05000391/2015616 (ADAMS 

Accession No. ML15173A317).

Inspection of Design Vulnerability in Electrical Power System (Open Phase Condition)

An inspection was performed to verify actions to address the concerns identified in NRC 

Bulletin 2012-01, Design Vulnerability in Electrical Power System. The inspectors confirmed 

the interim actions taken by WBN Unit 2 in response to Bulletin 2012-001 were consistent 

with SSER 27, their September 3, 2014 and February 3, 2014, RAI responses and the TVA 

Response to Bulletin 2012-001, dated October 25, 2012.

B. Preoperational Testing Inspections

IMC 2513 Appendix A

Region II implemented the preoperational testing inspections specified in Appendix A of IMC 

2513. Appendix A contains the procedures applicable to verifying that systems and 

components important to safety of the plant are fully tested to demonstrate that they satisfy 

their design requirements. Region II has completed all of the selected IMC 2513 

Appendix A procedure review inspections and the majority of the test witnessing 

inspections. The groupings of preoperational testing inspections are listed as follows:

Mandatory Tests 

- Reactor Coolant System Hydrostatic Test

- Reactor Protection System Test

- Engineered Safety Features Test

- Loss of Offsite Power Test

- Containment Integrated Leak Rate Test

- Integrated Hot Functional Test

Primal System Testing

A minimum selection of five safety-related systems was required to complete the primal 

system testing inspection requirement. Region II developed an inspection sample based 

primarily on the risk-significance of systems at WBN Unit 2. The following systems were 

selected for review: auxiliary feedwater (AFW), chemical and volume control (CVCS), safety 

injection (SI), essential raw cooling water (ERCW), component cooling water (CCS), 

containment spray (CS), residual heat removal (RHR), main steam isolation valve (MSIV), 

main feedwater (MFW), ice condenser, and containment hydrogen mitigation.

GG-8
4 

Other Inspections Areas Covered Under IMC 2513, Appendix A

IMC 2513 lists several other IPs that involve inspection of the testing program infrastructure 

and inspections of related areas that assess the quality of the plant as construction is near 

completion. Example IPs include “Quality Assurance for Preoperational Testing”, 

“Comparison of As-Built Plant to FSAR Description”, and “Testing of Pipe Support and 

Restraint Systems”.

C. Operational Preparedness Inspections 

IMC 2513 Appendix B

Operational preparedness inspections are specified in Appendix B of IMC 2513 and address 

functional areas such as operations, maintenance, radiological controls, security, etc. Many 

of these programs and procedures for WBN Unit 2 will be the same or nearly identical to 

those already established and in use for WBN Unit 1. The NRC staff reviewed each of the 

IPs in Appendix B and developed revised requirements for performing these IPs, focusing 

on aspects of these programs and procedures that are unique to WBN Unit 2 or required 

substantive changes to address WBN Unit 2. The inspection plan for the Appendix B 

inspections was issued to TVA by letter dated November 8, 2013 (ADAMS Accession No.

ML13312A082. 

Operational Readiness Assessment Team (ORAT) Inspection 

The ORAT inspection was completed at WBN Unit 2 on June 26, 2015. The ORAT team 

performed an independent assessment of TVA’s readiness to operate and integrate WBN 

Unit 2 into a currently operating facility, WBN Unit 1. The ORAT inspection team looked at 

five broad areas: management oversight, control of safety-significant activities, operations 

training and experience, corrective action program, and maintenance support activities. The 

ORAT determined that TVA has programs and processes to effectively turnover systems 

from construction, to testing, and then to operations. Based on this determination and a 

review of other dual unit transition activities, the ORAT concluded that TVA has 

demonstrated their readiness to safety startup and conduct power operations on WBN Unit 

2. The final inspection report was issued on August 14, 2015 (ADAMS Accession No. 

ML15226A212).

GG-9
OPEN INSPECTION ITEMS 

Inspection Item

Basis to Remain Open After IP 94302 Issuance

IMC 2512 - Construction Inspection (IP&S Items)

Fire Protection Corrective 

Action Program (CAP) 

NRC verified TVA’s CAP has adequate corrective actions for the fire protection deficienci

sample of fire protection deficiencies were adequately resolved. Additional NRC inspecti

implementation of the fire protection program corrective actions will be conducted to clos

Mechanical Equipment 

Qualification Special Program 

(SP) 

NRC inspection activities verified that applicable equipment is included in the mechanica

SP. This was documented in inspection report 05000391/2012603. NRC also inspected 

program for the containment spray system, including specific preventive maintenance pla

was documented in inspection report 05000391/2015605. The NRC will inspect addition

this item. 

Radiation Monitoring System 

SP

NRC inspection verified that formal design criteria had been adequately established for th

monitoring system. Previous inspections of this SP were documented in inspection repor

05000391/2010605, and 05000391/2011607. Further inspection will evaluate the implem

verifying the installed equipment and its design. 

Inspection Procedure (IP) 

63050, Containment Structural 

Integrity Test

This IP consists of the procedure review, test witness, and results review for the containm

test. NRC reviewed the test procedure and determined that it was adequate and that the 

the design requirements. NRC also witnessed the successful completion of the test. NR

of the test to complete the inspection procedure.

Bulletin 80-06, Emergency 

safety feature reset controls

This item has been inspected as part of the emergency safety features (ESF) preoperatio

inspection that the reset feature meets its design requirements. The ESF test consists of 

instructions (PTIs), two of which have been witnessed by NRC inspectors and were deter

NRC will inspect the remaining portion of the ESF test to complete this item. 

GG-10
2 

Inspection Item Basis to Remain Open After IP 94302 Issuance

Three Mile Island (TMI) II.D.1, 

relief and safety valve test 

requirements

During hot functional testing (HFT), NRC witnessed the power operated relief valve (POR

is addressing PORV test deficiencies that were identified during the HFT. This includes t

testing facility at normal operating conditions. NRC determined that the test plan is adeq

PORV stroke testing following re-installation. 

Generic Letter (GL) 88-14, 

Instrument air supply system 

problems affecting safetyrelated equipment

Previous NRC inspection of this item verified that implementation of design changes and 

were consistent with TVA’s approach documented in the NRC regulatory framework lette

ML 080320443). This was documented in inspection report 05000391/2013605. Addition

performed to verify that the test results meet the design requirements of the system.

GL 89-04, Guidelines on 

developing acceptable inservice testing (IST) programs

NRC verified through inspection that WBN Unit 2’s IST program is the same as WBN Un

program is in compliance with regulatory requirements. This initial IST program inspectio

completed and documented in inspection report 05000391/2014614. NRC will review ad

testing to complete the inspection of GL 89-04.

GL 04-02, Potential impact of 

debris blockage on 

emergency recirculation 

during design basis accidents 

at pressurized water reactors 

(PWRs) 

Previous NRC inspection of the emergency recirculation system verified the adequacy of 

containment sump and coatings inside containment that could impact containment sump 

inspection was documented in inspection report 05000391/2015604. NRC will review the 

containment latent debris walkdown to complete inspection of this generic letter.

Temporary Instruction (TI) 

2500/19, USI A-26: Licensee's 

actions taken to implement 

USI A-26: Transient protection

NRC inspected WBN Unit 2’s mitigation system for low-temperature overpressure transie

concluded that it was adequate. Inspections of the system implementation verified the de

controls, procedures, modifications, training, and surveillances. The inspection results w

inspection reports 05000391/2013607 and 2015605. NRC will review the remaining POR

complete this item. 

TI 2500/20, ATWS Rule, GL 83-

28 & 85-06

Previous NRC inspections verified that the current design and sampled installation were 

approved design. These inspections were documented in inspection reports 05000391/2

2014615. Additional inspections will review system testing to complete this item.

GG-11
3 

Inspection Item Basis to Remain Open After IP 94302 Issuance

TI 2515/110, Performance of 

Safety-Related Check Valves, 

NRC has verified through inspection that WBN Unit 2’s IST program which covers check 

WBN Unit 1’s program. Unit 1’s program is in compliance with regulatory requirements. 

inspection for Unit 2 was completed and documented in inspection report 05000391/2014

additional IST baseline testing to complete the inspection of this item.

TI 2515/114, In-Service 

Testing, GL 89-04

NRC has verified through inspection that WBN Unit 2’s IST program is the same as Unit 

program is in compliance with regulatory requirements. An initial IST program inspection 

and documented in inspection report 05000391/2014614. NRC will review additional IST 

complete the inspection requirements of this item.

TI 2515/166, Containment 

sump blockage

A previous NRC inspection of the emergency recirculation system verified the adequacy 

the containment sump and coatings inside containment that could impact containment su

inspection was documented in inspection report 05000391/2015604. NRC will inspect th

containment latent debris walkdown to complete inspection of this item.

Inspector Follow-up Item 86-

10-03, U2 Instrument Air 

Preoperational Test

Previous NRC inspection of this item verified that implementation of design changes and 

were consistent with TVA’s approach documented in the NRC regulatory framework lette

ML 080320443), and was documented in inspection report 05000391/2013605. Addition

that the system test results meet the design requirements.

Construction Deficiency 

Report (CDR) 83-61, Failure to 

provide self-contained lights 

as committed to NRC

NRC verified through inspection that TVA’s emergency lighting system modification meth

The NRC will review additional samples when they are available to verify that the implem

meets applicable regulatory requirements.

CDR 86-11, Lack of thermal 

qualification for Systems 43 & 

90 piping

This item required inspection of two systems, the radiation sampling system (system 43) 

monitoring system (system 90). Previous NRC inspections of the radiation sampling syst

design, analysis, and implementation were adequate. This was documented in inspectio

05000391/2015607. Additional inspections of the radiation monitoring system implement

complete this item. 

GG-12
4 

Inspection Item Basis to Remain Open After IP 94302 Issuance

CDR 89-09, Significant trend 

associated with damaged, 

loose, or missing hardware

NRC verified through inspection that the damaged, loose, and missing hardware program 

procedures were adequate. This was documented in inspection report 05000391/201360

several subsequent inspections of safety systems and determined that the program is be

appropriately. These inspections included hand-over-hand walkdowns of the systems as 

NRC will walkdown additional systems and areas of the plant to complete this item.

CDR 93-02, Loose flexible 

conduit fittings

NRC has reviewed TVA’s corrective actions and concluded that the program was adequa

documented in inspection report 05000391/2014604. NRC has performed several subse

safety systems and determined that the program is being implemented appropriately. Th

system and area walkdowns. NRC will walkdown additional systems and areas of the pla

Final CAP/SP Inspection - This

item is to satisfy Section 03.02 

of CAP/SP related TI's

The TI for each CAP/SP requires a final inspection be performed after TVA certifies that t

completed. All CAP/SPs have been inspected for technical adequacy to verify that the co

developed to address the issues were acceptable. A sample of 20 certified CAP/SPs we

documented in inspection report 05000391/2015607. NRC will perform additional inspec

certified the remaining CAPs/SPs are completed. 

Enforcement Action, EA-15-

112; inspection report 

05000391/2015607

This severity level (SL) IV notice of violation (NOV) was the result of the enforcement pro

Investigations Report 2-2013-017. The staff reviewed the initial corrective actions as par

process and verified that the anchor bolt in question was acceptable and will not affect th

coolant drain tank pump 6 to perform its function. Further inspection of this issue will be 

TVA’s written response to the NOV.

GG-13
5 

Inspection Item Basis to Remain Open After IP 94302 Issuance

IMC 2513 Appendix A – Pre-Operational Testing Inspection

IP 70329, 

Preoperational Test 

Results Evaluation

Previous NRC inspections of preoperational test results have included the inspection req

of IP 70329 into the scope of those reviews. These inspections concluded that the test re

Additional inspections of final test result evaluations will be performed to complete this in

IP 71302, 

Preoperational Test 

Program

Implementation 

Verification

Previous NRC inspections concluded that TVA’s implementation of the preoperational tes

These inspections were documented in integrated inspection report 05000391/2014607. 

the preoperational test program implementation will be conducted to complete this inspec

IP 70370, Testing 

Piping Support and 

Restraint Systems

NRC has completed all aspects of the inspection procedure that were required during pre

2513). Additional inspections during plant operations will be conducted to complete this i

ESF

IP 70315, Engineering 

Safety Features Test 

Witnessing

This mandatory test consists of three PTIs. NRC has determined the PTIs were adequat

acceptance criteria met the design requirements. Also, NRC witnessed the acceptability 

stated in their substantially complete letter that this mandatory test would be completed p

witness the remaining PTI to complete the inspection procedure. 

ESF

IP 70322, Engineering 

Safety Features Test 

Results Evaluation

Previous NRC inspections verified that the test procedures’ acceptance criteria met the d

during the test the acceptance criteria was met, and that test deficiencies were properly i

accordance with TVA procedures. NRC will review the final test results to complete the in

RPS

IP 70317, Reactor 

Protection System

(RPS) Test 

Witnessing

This mandatory test consists of seven PTIs. NRC has determined the PTIs were adequa

acceptance criteria met the design requirements. Also, NRC witnessed the acceptability 

stated in their substantially complete letter that this mandatory test would be completed p

witness the remaining two PTIs to complete the inspection procedure. 

RPS

IP 70325, Reactor 

Protection System 

Test Results 

Evaluation

Previous NRC inspections verified that the test procedures’ acceptance criteria met the d

during the test the acceptance criteria was met, and that test deficiencies were properly i

accordance with TVA procedures. NRC will review the final test results to complete the in

GG-14
6 

Inspection Item Basis to Remain Open After IP 94302 Issuance

IMC 2513 Appendix A – Pre-Operational Testing Inspection

LOOP 

IP 70326, Loss of 

Offsite Power Test 

Results Evaluation

Previous NRC inspections for the Loss of Offsite Power (LOOP) pre-operational tests ver

procedures’ acceptance criteria met the design requirements, that during the test the acc

and that test deficiencies were properly identified for correction in accordance with TVA p

review the results of the LOOP tests to complete the inspection procedure.

CILRT

IP 70323, 

Containment System: 

Integrated Leak Rate 

Test Results 

Evaluation

Previous inspections of this mandatory test verified that the acceptance criteria in the tes

requirements, that during the test the acceptance criteria was met, and that test deficienc

identified for correction in accordance with TVA procedures. NRC will review the final re

complete the inspection procedure. 

MFW

IP 70400, 

Preoperational Test 

Results Evaluation

NRC inspections for the primal preoperational test of the main feedwater (MFW) system 

acceptance criteria in the test procedure met design requirements, that during the test the 

met, and that test deficiencies were properly identified for correction in accordance with T

will review the final results of the test to complete the inspection procedure.

AFW

IP 70438, Auxiliary 

Feedwater System

(AFW) Test 

Witnessing

This primal test consists of three PTIs. NRC has verified through inspection that the acce

procedures met the design requirements, and witnessed the acceptability of the test for tw

will witness the remaining PTI to complete the inspection procedure.

AFW

IP 70400, 

Preoperational Test 

Results Evaluation

NRC conducted previous inspections of the AFW preoperational tests by verifying that th

the test procedures met design requirements, that during the test the acceptance criteria 

deficiencies were properly identified for correction in accordance with TVA procedures. N

results of the tests to complete the inspection procedure.

CVCS

IP 70400, 

Preoperational Test 

Results Evaluation

This primal test consisted of three PTIs. NRC conducted previous inspections of the che

system (CVCS) preoperational tests, including verification that the acceptance criteria in 

design requirements, that during the test the acceptance criteria was met, and that test d

identified for correction in accordance with TVA procedures. NRC will review the final res

complete the inspection procedure.

GG-15
7 

Inspection Item Basis to Remain Open After IP 94302 Issuance

ERCW IP 70400, 

Preoperational Test 

Results Evaluation

This primal test consisted of four PTIs. NRC conducted previous inspections for the esse

(ERCW) preoperational tests, including verification that the acceptance criteria in the test 

requirements, that during the test the acceptance criteria was met, that test deficiencies w

for correction in accordance with TVA procedures, and that the final results of three of the 

NRC will review the final results for the remaining PTI to complete the inspection procedu

RHR

IP 70400 

Preoperational Test 

Results Evaluation

This primal test consisted of three PTIs. NRC conducted previous inspections for the res

system preoperational tests, including verification that the acceptance criteria in the test p

requirements, that during the test the acceptance criteria was met, that test deficiencies w

for correction in accordance with TVA procedures, and that the final results of two of the 

NRC will review the final results for the remaining PTI to complete the inspection procedu

H2

IP 70442, 

Containment 

Combustible Gas (H2)

Control System Test 

Witnessing

This primal test consists of one PTI. NRC has reviewed the acceptance criteria in the tes

that they met design requirements, and documented this in inspection report 05000391/2

inspections have determined the design and constructability of the system were adequate

were documented in inspection reports 05000391/2011602, 2011604, 2011605, 2012603

2015602. NRC will witness the test to complete this inspection procedure.

H2

IP 70400, 

Preoperational Test 

Results Evaluation

NRC has reviewed the acceptance criteria in the hydrogen mitigation test procedure to en

design requirements, and documented this in inspection report 05000391/2015605. Othe

determined the design and constructability of the system were adequate. These inspecti

inspection reports 05000391/2011602, 2011604, 2011605, 2012603, 2013605, 2014604

review the final test results to complete this inspection procedure.

GG-16
8 

Inspection Item Basis to Remain Open After IP 94302 Issuance

IMC 2513 Appendix B – Operational Preparedness Inspections

IP 35744, Quality Assurance 

(QA) Program - Design 

Changes and Modifications

An inspection of the design change and modification program has been performed and d

report 05000391/2013607. This inspection concluded that the program was in complianc

regulatory requirements. Further NRC inspection is planned if temporary modifications a

six months of operation.

IP 35749, QA Program – Tests 

and Experiments

An inspection of the 10 CFR 50.59 program procedures was performed and documented 

05000391/2014607. This inspection concluded that the program was in compliance with 

requirements. The 50.59 program would not apply to Unit 2 until an operating license is 

TVA’s implementation of this program to complete the inspection procedure. 

IP 64704, Fire Protection 

Program

NRC has verified through inspection that the TVA’s new dual unit fire protection program 

as-constructed fire protection report. SSER 29 will document the final fire protection repo

NRC will review additional samples of TVA’s ongoing implementation of in-plant modifica

the fire protection program.

IP 83526, Control of 

Radioactive Materials and 

Contamination, Surveys, and 

Monitoring

A previous inspection of the radiation protection survey and monitoring program conclude

capabilities currently in place for portable surveys, protective clothing, contamination con

monitoring are adequate to support dual-unit operation. For the in-place radiation monito

to WBN Unit 2, the components have been installed. Additional inspections will review th

preoperational testing for these components.

IP 84523, Liquids and Liquid 

Wastes

A previous inspection reviewed the liquid radwaste system and liquid effluent monitors an

processing, sampling, and discharge protocols and methodologies used for the liquid rad

were the same as those used in WBN Unit 1. The inspection also determined that the ins

the liquid effluent monitor was adequate. NRC will review additional testing results of the 

to complete this inspection procedure.

IP 84524, Gaseous Waste 

System

A previous inspection of the gaseous radwaste system and gaseous effluent monitors de

components unique to WBN Unit 2 had been adequately installed. NRC will review addit

gaseous radwaste system and gaseous effluent monitors to complete this inspection proc

GG-17

HH-1 

APPENDIX HH WATTS BAR, UNIT 2, ACTION ITEMS TABLE

This table provides a status of required action items associated of all open items, confirmatory 

issues (Cis), and proposed license conditions that the U.S. Nuclear Regulatory Commission 

(NRC) staff has identified. Unless otherwise noted, the item references are to sections of this

Supplemental SER (SSER). Items that are still open are listed first, and items that have been 

closed are listed second. Some numbers were not used in the sequential list. There are no

open and 139 items that have been closed.

Closed Items

Item Type Action Required Lead Status

(1) Confirmatory 

Issue (CI)

Review evaluations and corrective actions 

associated with a power assisted cable pull.

(NRC safety evaluation, dated August 31, 

2009, Agencywide Documents Access and 

Management System (ADAMS) Accession 

No. ML092151155). Closed. Inspection 

Report 0500391/2015607 dated 

September 29, 2015 (ADAMS Accession No. 

ML15273A452) stated that based on the 

results of this inspection and past inspections 

on this activity, adequate controls were in 

place during the reviewed cable pulling 

activities. The NRC staff’s evaluation of the 

corrective actions program (CAP) plans for 

cable and electrical issues was documented in 

a letter dated August 31, 2009 (ADAMS 

Accession No. ML092151154) which stated 

that the NRC has completed its review of the 

CAP plans for the cable and electrical issues.

Based on its review, the NRC staff concluded

that these CAP plans provide acceptable 

guidelines for identification, resolution, 

implementation, and inspection of the various 

issues addressed in the program plans. 

Based on the inspection and NRC staff 

review, the staff considers this item closed.

RII Closed

(2) CI Conduct appropriate inspection activities to 

verify cable lengths used in calculations and 

analysis match as-installed configuration. 

(NRC safety evaluation, dated August 31, 

2009, ADAMS Accession No. ML092151155). 

Closed in Inspection Report 

05000391/2013604, dated June 27, 2013, 

ADAMS Accession No. ML13179A079.

RII Closed
HH-2 

(3) CI Confirm TVA submitted update to FSAR 

section 8.3.1.4.1. (NRC safety evaluation (SE) 

dated August 31, 2009, ADAMS Accession 

No. ML092151155) Closed in SSER 24, 

Section 8.1.

NRR Closed

(4) CI Conduct appropriate inspection activities to 

verify that TVA’s maximum SWBP criteria for 

signal level and coaxial cables do not exceed 

the cable manufacturer’s maximum SWBP 

criteria. (NRC safety evaluation, dated 

August 31, 2009, ADAMS Accession No. 

ML092151155) Closed in Inspection Report 

0500391/2012602, dated March 27, 2012, 

ADAMS Accession No. ML12087A324.

RII Closed

(5) CI Verify timely submittal of pre-startup core map 

and perform technical review. (TVA letter, 

dated September 7, 2007, ADAMS Accession 

No. ML072570676). By letter dated July 30, 

2012, TVA provided the pre-startup core map. 

The staff has verified the information and has 

closed Appendix HH Open Items 5 and 8, 

which came from the review of BL 1996-01. 

NRR Closed

(6) CI Verify implementation of TSTF-449. (TVA 

letter dated September 7, 2007, ADAMS 

Accession No. ML072570676). Staff has 

reviewed Revision I to the proposed TSs and 

found that TSTF 449 has been incorporated.

NRR Closed

(7) CI Verify commitment completion and review 

electrical design calculations. (TVA letter 

dated October 9, 1990, ADAMS Accession 

No. ML073551056). Closed in Inspection 

Report 05000391/2013610, dated 

February 14, 2014, ADAMS Accession 

No. ML14049A158.

RII Closed

(8) CI TVA should provide a pre-startup map to the 

NRC staff indicating the rodded fuel 

assemblies and a projected end of cycle 

burnup of each rodded assembly for the initial 

fuel cycle 6-months prior to fuel load. (NRC 

safety evaluation, dated May 3, 2010, ADAMS 

Accession No. ML101200035). By letter 

dated July 30, 2012, TVA provided the 

pre-startup core map. The staff has verified 

the information and has closed Appendix HH 

Open Items 5 and 8, which came from the 

review of BL 1996-01.

NRR Closed
HH-3 

(9) CI Confirm that education and experience of 

management and principal supervisory 

positions down through the shift supervisory 

level conform to Regulatory Guide 1.8. 

(SSER 22, Section 13.1.3). Closed in 

Inspection Report 0500391/2014603, dated 

May 9, 2014, ADAMS Accession 

No. ML14129A381.

RII Closed

(10) CI Confirm that TVA has an adequate number of 

licensed and non-licensed operators in the 

training pipeline to support the preoperational 

test program, fuel loading, and dual unit 

operation. (SSER 22, Section 13.1.3). Closed 

in Inspection Report 0500391/2014603, dated 

May 9, 2014, ADAMS Accession No. 

ML14129A381.

RII Closed

(11) CI The plant administrative procedures should 

clearly state that, when the Assistant Shift 

Engineer assumes his duties as Fire Brigade 

Leader, his control room duties are 

temporarily assumed by the Shift Supervisor 

(Shift Engineer), or by another senior reactor 

operator, if one is available. The plant 

administrative procedures should clearly 

describe this transfer of control room duties. 

(SSER 22, Section 13.1.3) Closed in 

SSER 25, Section 13.1.3.

NRR Closed

(12) TVA’s implementation of New Generation 

Development and Construction PP-20 and 

Engineering Document Construction Release 

Appendix J is subject to future NRC audit and 

inspection. (SSER 22, Section 25.9). Closed 

in SSER 27, Section 25.9.

NRR Closed

(13) TVA is expected to submit an inservice testing

program and specific relief requests for WBN, 

Unit 2, 9 months before the projected date of 

OL issuance. (SSER 22, Section 3.9.6). 

Closed in SSER 27, Section 3.9.6.

NRR Closed

(14) TVA stated that the Unit 2, Pressure and 

Temperature Limits Report (PTLR) is included 

in the Unit 2, System Description for the 

Reactor Coolant System (WBN2-68-4001), 

which will be revised to reflect required 

revisions to the PTLR by September 17, 2010. 

(SSER 22, Section 5.3.1) Closed in SSER 25, 

Section 5.3.1.

NRR Closed
HH-4 

(15) TVA should confirm to the NRC staff the 

completion of Primary Stress Corrosion 

Cracking (PWSCC) mitigation activities on the 

Alloy 600 dissimilar metal butt welds 

(DMBWs) in the primary loop piping. (SSER 

22, Section 3.6.3) Closed in SSER 24, 

Section 3.6.3. 

NRR Closed

(16) Based on the uniqueness of environmental 

qualification (EQ), the NRC staff must perform 

a detailed inspection and evaluation prior to 

fuel load to determine how the WBN, Unit 2, 

EQ program complies with the requirements of 

10 CFR 50.49. (SSER 22, Section 3.11.2) 

Closed in Inspection Report 

05000391/2015605, dated August 14, 2015, 

ADAMS Accession No. ML15226A345.

RII/NRR Closed

(17) The NRC staff should verify the accuracy of 

the WBN, Unit 2, EQ list prior to fuel load. 

(SSER 22, Section 3.11.2.1) Closed in 

Inspection Report 05000391/2014615, dated 

February 13, 2014, ADAMS Accession No. 

ML15044A424.

RII/NRR Closed

(18) Based on the extensive layup period of 

equipment within WBN, Unit 2, the NRC staff 

must review, prior to fuel load, the 

assumptions used by TVA to re-establish a 

baseline for the qualified life of equipment. 

The purpose of the staff’s review is to ensure 

that TVA has addressed the effects of 

environmental conditions on equipment during 

the layup period. (SSER 22, Section 3.11.2.2) 

Closed in Inspection Report 

0500391/2011604, dated June 29, 2011, 

ADAMS Accession No. ML111810890.

RII/NRR Closed

(19) The NRC staff should complete its review of 

TVA’s EQ Program procedures for WBN, 

Unit 2, prior to fuel load. (SSER 22, Section 

3.11.2.2.1) Closed in Inspection Report 

0500391/2011604, dated June 29, 2011, 

ADAMS Accession No. ML111810890.

RII/NRR Closed

(20) CI Resolve whether or not routine maintenance 

activities should result in increasing the EQ of 

the 6.9 kV motors to Category I status in 

accordance with 10 CFR 50.49. (SSER 22, 

Section 3.11.2.2.1; SSER 24, Section 8.1) 

Closed in Inspection Report 

0500391/2011605, dated August 5, 2011, 

ADAMS Accession No. ML112201418.

RII/NRR Closed
HH-5 

(21) The NRC staff should confirm that the 

electrical penetration assemblies (EPAs) are 

installed in the tested configuration, and that 

the feedthrough module is manufactured by 

the same company and is consistent with the 

EQ test report for the EPA. (SSER 22, 

Section 3.11.2.2.1) Closed in Inspection 

Report 05000391/2011607, dated 

September 30, 2011, ADAMS Accession No. 

ML112730197.

RII/NRR Closed

(22) TVA must clarify its use of the term 

“equivalent” (e.g., identical, similar) regarding 

the replacement terminal blocks to the NRC 

staff. If the blocks are similar, then a similarity 

analysis should be completed and presented 

to the NRC for review. (SSER 22, Section 

3.11.2.2.1) Closed in SSER 24, Section 8.1.

NRR Closed

(23) Resolve whether or not TVA’s reasoning for 

not upgrading the main steam isolation valve

solenoid valves to Category I is a sound 

reason to the contrary, as specified in 10 CFR 

50.49(l). (SSER 22, Section 3.11.2.2.1; 

SSER 24, Section 8.1). Closed in SSER 27, 

Section 3.11.2.2.1.

NRR Closed

(24) The NRC staff requires supporting 

documentation from TVA to justify its 

establishment of a mild environment threshold 

for total integrated dose of less than 1x103

rads for electronic components such as 

semiconductors or electronic components 

containing organic material. (SSER 22, 

Section 3.11.2.2.1) Closed in SSER 24, 

Section 8.1.

NRR Closed

(25) Prior to the issuance of an OL, TVA is 

required to provide satisfactory documentation 

that it has obtained the maximum secondary 

liability insurance coverage pursuant to

10 CFR 140.11(a)(4), and not less than the 

amount required by 10 CFR 50.54(w), with 

respect to property insurance, and the NRC 

staff has reviewed and approved the 

documentation. (SSER 22, Section 22.3)

Closed in SSER 29, Section 22.3.

NRR Closed
HH-6 

(26) For the scenario with an accident in one unit 

and concurrent shutdown of the second unit 

without offsite power, TVA stated that Unit 2, 

pre-operational testing will validate the diesel 

response to sequencing of loads on the Unit 2, 

emergency diesel generators (EDGs). The 

NRC staff will evaluate the status of this issue 

and will update the status of the EDG load 

response in a future SSER. (SSER 22, 

Section 8.1). Closed in SSER 27, Section 8.1.

NRR Closed

(27) TVA should provide a summary of margin 

studies based on scenarios described in 

Section 8.1 for CSSTs A, B, C, and D. (SSER 

22, Section 8.2.2) Closed in SSER 24, 

Section 8.1. 

NRR Closed

(28) TVA should provide to the NRC staff a 

detailed discussion showing that the load tap 

changer is able to maintain the 6.9 kV bus 

voltage control band, given the normal and 

post-contingency transmission operating 

voltage band, bounding voltage drop on the 

grid, and plant conditions. (SSER 22, 

Section 8.2.2) Closed in SSER 24, 

Section 8.1. 

NRR Closed

(29) TVA should provide information about the 

operating characteristics of the offsite power 

supply at the Watts Bar Hydro Plant (for dualunit operation), including the operating voltage 

range, post contingency voltage drops 

(including bounding values and post-unit trip 

values), and operating frequency range. 

(SSER 22, Section 8.2.2) (corrected version of 

Open Item 29 from SSER 22 Appendix HH) 

Closed in SSER 24, Section 8.1.

NRR Closed

(30) TVA should confirm that all other safetyrelated equipment (in addition to the Class 1E 

motors) will have adequate starting and 

running voltage at the most limiting safetyrelated components (such as motor-operated 

valves (MOVs), contactors, solenoid valves or 

relays) at the degraded voltage relay setpoint 

dropout setting. TVA should also confirm that 

the final Technical Specifications are properly 

derived from these analytical values for the 

degraded voltage settings. (SSER 22, Section 

8.3.1.2) Closed in SSER 28, Section 8.3.1.2.

RII/NRR Closed
HH-7 

(31) TVA should clarify the loading sequence as 

explained in its letter dated December 6, 

2010, to the staff. TVA should clarify whether 

the existing statements in FSAR regarding 

automatic sequencing logic are correct. If the 

FSAR description is correct, TVA should 

explain how the EDG and logic sequencing 

circuitry will respond to a LOCA followed by a 

loss-of-offsite power (LOOP) scenario. (SSER 

22, Section 8.3.1.11) (corrected version of 

Open Item 31 from SSER 22 Appendix HH) 

Closed in SSER 24, Section 8.1 

NRR Closed

(32) TVA should provide to the NRC staff the 

details of the administrative limits of EDG 

voltage and speed range, and the basis for its 

conclusion that the impact is negligible, and 

describe how it accounts for the administrative 

limits in the Technical Specification 

surveillance requirements for EDG voltage 

and frequency. (SSER 22, Section 8.3.1.14). 

Closed in SSER 27, Section 8.3.

NRR Closed

(33) CI TVA stated in Attachment 9 of its letter dated 

July 31, 2010, that certain design change 

notices (DCNs) are required or anticipated for 

completion of WBN, Unit 2, and that these 

DCNs were unverified assumptions used in its 

analysis of the 125 volt direct current (VDC)

vital battery system. Verification of completion 

of these DCNs to the NRC staff is necessary 

prior to issuance of the OL. (SSER 22, 

Section 8.3.2.3; SSER 24, Section 8.1) 

Closed in Inspection Report 

05000391/2015602, dated March 24, 2015, 

ADAMS Accession No. ML15083A276.

RII/NRR Closed

(34) CI TVA stated that the method of compliance 

with Phase I guidelines would be substantially 

similar to the current Unit 1,program and that 

a new Section 3.12 will be added to the Unit 2, 

FSAR that will be materially equivalent to 

Section 3.12 of the current Unit 1, FSAR. 

(SSER 22, Section 9.1.4) Closed in SSER 24, 

Section 9.1.4.

NRR Closed

(35) TVA should provide information to the NRC 

staff that the CCS will produce feedwater 

purity in accordance with BTP Materials 

Engineering Branch 5-3 or, alternatively, 

provide justification for producing feedwater 

purity to another acceptable standard. 

(SSER 22, Section 10.4.6). Closed in 

SSER 27, Section 10.4.6.

NRR Closed
HH-8 

(36) TVA should provide information to the NRC 

staff to enable verification that the SGBS 

meets the requirements and guidance 

specified in the SER or provide justification 

that the SGBS meets other standards that 

demonstrate conformance to GDC 1 and 

GDC 14. (SSER 22, Section 10.4.8) Closed in 

SSER 24, Section 10.4.8. 

NRR Closed

(37) CI The NRC staff will review the combined WBN, 

Unit 1, and 2 Appendix C prior to issuance of 

the Unit 2, OL to confirm (1) that the proposed 

Unit 2, changes were incorporated into 

Appendix C, and (2) that changes made to 

Appendix C for Unit 1, since Revision 92 and 

the changes made to the Nuclear Power 

Radiological Emergency Plan (REP) since 

Revision 92 do not affect the bases of the 

staff’s findings in this SER supplement. (SSER 22, Section 13.3.2) Closed in 

SSER 28, Section 13.3.2.

NSIR Closed

(38) CI The NRC staff will confirm the availability and 

operability of the ERDS for Unit 2, prior to 

issuance of the Unit 2 OL. (SSER 22, Section 

13.3.2.6) Closed in Inspection Report 

05000391/2014614, dated December 29, 

2014, ADAMS Accession No. ML14363A315.

RII/NSIR Closed

(39) CI The NRC staff will confirm the adequacy of the 

communications capability to support dual unit 

operations prior to issuance of the Unit 2 OL. 

(SSER 22, Section 13.3.2.6) Closed in 

Inspection Report 0500391/2011609, dated 

December 16, 2011, ADAMS Accession No. 

ML11350A229. 

RII/NSIR Closed

(40) CIThe NRC staff will confirm the adequacy of the 

emergency facilities and equipment to support 

dual unit operations prior to issuance of the 

Unit 2, OL. (SSER 22, Section 13.3.2.8) 

Closed in Inspection Report 

05000391/2014614, dated December 29, 

2014, ADAMS Accession No. ML14363A315.

RII/NSIR Closed

(41) CI TVA committed to (1) update plant data 

displays as necessary to include Unit 2, and 

(2) to update dose assessment models to 

provide capabilities for assessing releases 

from both WBN Units. The NRC staff will

confirm the adequacy of these items prior to 

issuance of the Unit 2 OL. (SSER 22, Section 

13.3.2.9) Closed in Inspection Report 

05000391/2015603, dated May 1, 2015, 

ADAMS Accession No. ML15124A921.

RII/NSIR Closed
HH-9 

(42) CI The NRC staff will confirm the adequacy of the 

accident assessment capabilities to support 

dual unit operations prior to issuance of the 

Unit 2 OL. (SSER 22, Section 13.3.2.9) 

Closed in Inspection Report 

0500391/2011609, dated December 16, 2011, 

ADAMS Accession No. ML11350A229.

RII/NSIR Closed

(43) CI Section V of Appendix E to 10 CFR Part 50 

requires TVA to submit its detailed 

implementing procedures for its emergency 

plan no less than 180 days before the 

scheduled issuance of an OL. Completion of 

this requirement will be confirmed by the NRC 

staff prior to the issuance of an OL. 

(SSER 22, Section 13.3.2.18) Closed in 

SSER 28, Section 13.3.2.18

NSIR Closed

(44) TVA should provide additional information to 

clarify how the initial and irradiated RTNDT 

was determined. (SSER 22, Section 5.3.1) 

Closed in SSER 25, Section 5.3.1.

NRR Closed

(45) CI TVA stated in its response to RAI 5.3.2-2, 

dated July 31, 2010, that the PTLR would be 

revised to incorporate the cold overpressure 

mitigation system arming temperature. 

(SSER 22, Section 5.3.2) Closed in SSER 25, 

Section 5.3.2.

NRR Closed

(46) CI The LTOP lift settings were not included in the 

PTLR, but were provided in TVA’s response to 

RAI 5.3.2-2 in its letter dated July 31, 2010. 

TVA stated in its RAI response that the PTLR 

would be revised to incorporate the LTOP lift 

settings into the PTLR. (SSER 22, 

Section 5.3.2) Closed in SSER 25, Section 

5.3.2.

NRR Closed
HH

-10

(47) CI The NRC staff noted that TVA’s changes to 

Section 6.2.6 in FSAR Amendment No. 97, 

regarding the implementation of Option B of 

Appendix J, were incomplete, because several 

statements remained regarding performing 

water-sealed valve leakage tests “as specified 

in 10 CFR [Part] 50, Appendix J.” With the 

adoption of Option B, the specified testing 

requirements are no longer applicable; 

Option A to Appendix J retains these 

requirements. The NRC discussed this 

discrepancy with TVA in a telephone 

conference on September 28, 2010. TVA 

stated that it would remove the inaccurate 

reference to Appendix J for specific water 

testing requirements in a future FSAR 

amendment. (SSER 22, Section 6.2.6) 

Closed in SSER 26, Section 6.2.6.

NRR Closed

(48) CI The NRC staff should verify that its 

conclusions in the review of FSAR 

Section 15.4.1 do not affect the conclusions of 

the staff regarding the acceptability of 

Section 6.5.3. (SSER 22, Section 6.5.3) 

Closed in SSER 26, Section 6.5.3.

NRR Closed

(49) CI The NRC staff was unable to determine how 

TVA linked the training qualification 

requirements of ANSI N45.2-1971 to TVA 

Procedure Technical Instruction (TI)-119. 

Therefore, the implementation of training and 

qualification for inspectors will be the subject 

of future NRC staff inspections. (NRC letter 

dated July 2, 2010, ADAMS Accession No. 

ML101720050). Closed in Inspection Report 

0500391/2014602, dated March 27, 2014, 

ADAMS Accession No. ML14086A063.

RII Closed

(50) CI TVA stated that about 5 percent of the anchor 

bolts for safety-related pipe supports do not 

have quality control documentation, because 

the pull tests have not yet been performed. 

Since the documentation is still under 

development, the NRC staff will conduct 

inspections to follow-up on the adequate 

implementation of this construction 

refurbishment program requirement. (NRC 

letter dated July 2, 2010, ADAMS Accession 

No. ML101720050) Closed in Inspection 

Report 0500391/2013612, dated March 28, 

2013, ADAMS Accession No. ML13088A066.

RII Closed
HH-11

(51) CI The implementation of TVA Procedure TI-119 

will be the subject of NRC follow-up inspection 

to determine if the construction refurbishment 

program requirements are being adequately 

implemented. (NRC letter dated July 2, 2010, 

ADAMS Accession No. ML101720050). 

Closed in Inspection Report 

0500391/2014602, dated March 27, 2014, 

ADAMS Accession No. ML14086A063.

RII Closed

(52) 

through 

(58)

Not used.

(59) The staff’s evaluation of the compatibility of 

the ESF system materials with containment 

sprays and core cooling water in the event of 

a LOCA is incomplete pending resolution of 

Generic Safety Issue-191 for WBN, Unit 2. 

(SSER 23, Section 6.1.1.4). Closed in 

SSER 27, Section 6.1.

NRR Closed

(60) CI TVA should amend the FSAR description of 

the design and operation of the spent fuel pool 

cooling and cleanup system in FSAR Section 

9.1.3 as proposed in its December 21, 2010, 

letter to the NRC. (SSER 23, Section 9.1.3) 

Closed in SSER 26, Section 9.1.3.

NRR Closed

(61) TVA should provide information to the NRC 

staff to demonstrate that Performance 

Analysis and Design (PAD) 4.0 can 

conservatively calculate the fuel temperature 

and other impacted variables, such as stored 

energy, given the lack of a fuel thermal 

conductivity degradation model. (SSER 23, 

Section 4.2.2). Closed in SSER 27, 

Section 4.2.

NRR Closed

(62) CI Confirm TVA’s change to FSAR Section 

10.4.9 to reflect its intention to operate with 

each condensate storage tank isolated from 

the other. (SSER 23, Section 10.4.9) Closed 

in SSER 24, Section 10.4.9.

NRR Closed

(63) CI TVA should confirm to the NRC staff that 

testing prior to Unit 2 fuel load has 

demonstrated that two-way communications is 

impossible with the Eagle 21 communications 

interface. (SSER 23, Section 7.2.1.1). Closed 

in SSER 27, Section 7.2.

RII Closed
HH-12

(64) CI TVA stated that, “Post modification testing will 

be performed to verify that the design change

corrects the Eagle 21, Rack 2 resistance 

temperature detector accuracy issue prior to 

WBN, Unit 2 fuel load.” This issue is open 

pending NRC staff review of the testing 

results. (SSER 23, Section 7.2.1.1). Closed 

in Inspection Report 05000391/2014602, 

dated March 27, 2014, ADAMS Accession No. 

ML14086A063.

RII Closed

(65) TVA should provide justification to the staff 

regarding why different revisions of 

Westinghouse Commercial Atomic Power13869 are referenced in WBN, Units 1 and 2. 

(SSER 23, Section 7.2.1.1) Closed in SSER 

26, Section 7.2.1.1.

NRR Closed

(66) CI TVA should clarify FSAR Section 9.2.5 to add 

the capability of the ultimate heat sink to bring 

the nonaccident unit to cold shutdown within 

72 hours. (SSER 23, Section 9.2.5). Closed 

in SSER 27, Section 9.2.5.

NRR Closed

(67) CI TVA should confirm, and the NRC staff should 

verify, that the component cooling booster 

pumps for Unit 2 are above probable 

maximum flood level. (SSER 23, Section 

9.2.2). Closed in SSER 27, Section 9.2.2 and 

Inspection Report 05000391/2014615, dated 

February 13, 2014, ADAMS Accession 

No. ML15044A424.

NRR Closed

(68) Not used.

(69) CI The WBN, Unit 2, reactor coolant system 

(RCS) vent system is acceptable, pending

verification that the RCS vent system is 

installed. (SSER 23, Section 5.4.5) Closed in 

Inspection Report 05000391/2014614, dated 

December 29, 2014, ADAMS Accession No. 

ML14363A315.

RII Closed

(70) TVA should provide the revised WBN, Unit 2, 

preservice inspection program ASME Class 1, 

2, and 3 Supports “Summary Tables,” to 

include numbers of components so that the 

NRC staff can verify that the numbers meet 

the reference ASME Code. (Section 3.2.3 of 

Appendix Z of SSER 23). Closed in SSER 27, 

Section 3.2.3 of Appendix Z.

NRR Closed
HH

-13

(71) By letter dated April 21, 2011 (ADAMS 

Accession No. ML111110513), TVA withdrew 

its commitment to replace the Unit 2 clevis 

insert bolts. TVA should provide further 

justification for the decision to not replace the 

bolts to the NRC staff. (SSER 23, Section 

3.9.5) Closed in SSER 26, Section 3.9.5.

NRR Closed

(72) The NRC staff should complete its review and 

evaluation of the additional information

provided by TVA regarding the inadequate 

core cooling instrumentation. (SSER 23, 

Section 4.4.8) Closed in SSER 25, 

Section 7.5.2.2.

NRR Closed

(73) CI The NRC staff will inspect to confirm that TVA 

has completed the WBN, Unit 2, emergency 

operating procedures (EOPs) prior to fuel 

load. (SSER 23, Section 7.5.3). Closed in 

Inspection Report 05000391/2014604, dated 

June 25, 2014, ADAMS Accession No. 

ML14177A214.

RII Closed

(74) CI The NRC staff will verify installation of the 

acoustic-monitoring system for the power-

operated relief valve (PORV) position 

indication in WBN, Unit 2, before fuel load. (SSER 23, Section 7.8.1) Closed in 

Inspection Report 05000391/2015604, dated 

June 29, 2015, ADAMS Accession No. 

ML15181A446.

RII Closed

(75) CI The NRC staff will verify that the test 

procedures and qualification testing for 

auxiliary feedwater initiation and control and 

flow indication are completed in WBN, Unit 2, 

before fuel load. (SSER 23, Section 7.8.2) 

Closed in Inspection Report 

05000391/2015608, dated October 21, 2015, 

ADAMS Accession No. ML15287A166.

RII Closed

(76) CI The NRC staff will verify that the derivative 

time constant is set to zero in WBN, Unit 2, 

before fuel load. (SSER 23, Section 7.8.3) 

Closed in Inspection Report 

05000391/2011607, dated September 30, 

2011, ADAMS Accession No. ML112730197.

RII Closed

(77) It is unclear to the NRC staff which software

verification and validation (V&V) documents 

are applicable to the HRCAR monitors. TVA 

should clarify which software V&V documents 

are applicable, in order for the staff to 

complete its evaluation. (SSER 23, 

Section 7.5.2.3) Closed in SSER 26, Section 

7.5.2.3.4

NRR Closed
HH

-14

(78) TVA intends to issue a revised calculation 

reflecting that the total integrated dose in the 

control room is less than 1×10

3 rads, which 

will be evaluated by the NRC staff. (SSER 23, 

Section 7.5.2.3) Closed in SSER 25, 

Section 7.5.2.3.

NRR Closed

(79) TVA should perform a radiated susceptibility 

survey, after the installation of the hardware 

but prior to the RM-1000 being placed in 

service, to establish the need for exclusion 

distance for the high range containment air 

radiation (HRCAR) monitors while using 

handheld portable devices (e.g., walkie-talkie) 

in the control room, as documented in 

Attachment 23 to TVA’s letter, dated February 

25, 2011, and item number 355 of TVA’s 

letter, dated April 15, 2011. (SSER 23, 

Section 7.5.2.3) Closed in SSER 29, 

Section 7.5.2.3.

NRR Closed

(80) TVA should provide clarification to the staff on 

how TVA Standard Specification SS-E18-14.1 

meets the guidance of Regulatory Guide (RG) 1.180, and should address any 

deviations from the guidance of the RG. 

(SSER 23, Section 7.5.2.3). Closed in 

SSER 27, Section 7.5.2.3.

NRR Closed

(81) The extent to which TVA’s supplier, General 

Atomics (GA), complies with Electric Power 

Research Institute TR-106439 and the 

methods that General Atomics used for its 

commercial dedication process should be 

provided by TVA to the NRC staff for review. (SSER 23, Section 7.5.2.3) Closed in 

SSER 26, Section 7.5.2.3.4.

NRR Closed

(82) The staff concluded that the information 

provided by TVA pertaining to the in-

containment loose part monitoring system 

equipment qualification for vibration was 

incomplete. TVA should provide (item number 

362 of ADAMS Accession No. ML111050009), 

documentation that demonstrates the loose 

part monitoring system in-containment 

equipment has been qualified to remain 

functional in its normal operating vibration 

environment, per RG 1.133, Revision 1. 

(SSER 23, Section 7.6.1) Closed in SSER 24, 

Section 7.6.1.4.5.

NRR Closed
HH-15

(83) CI TVA should confirm to the NRC staff the 

completion of the data storm test on the 

distribution control system. (SSER 23, 

Section 7.7.1.4) Closed in SSER 29, 

Section 7.7.1.4.

NRR Closed

(84) 

through 

(89)

Not used.

(90) CI The NRC staff should verify that the essential 

raw cooling water (ERCW) dual unit flow 

balance confirms that the ERCW pumps meet 

all specified performance requirements and 

have sufficient capability to supply all required 

ERCW normal and accident flows for dual-unit 

operation and accident response, in order to 

verify that the ERCW pumps meet GDC 5 for 

two-unit operation. (SSER 23, Section 9.2.1) 

Closed in Inspection Report 

05000391/2015608, dated October 21, 2015, 

ADAMS Accession No. ML15287A166.

RII/NRR Closed

(91) TVA should update the FSAR with information 

describing how WBN, Unit 2, meets GDC 5, 

assuming the worst case single failure and a 

LOOP, as provided in TVA’s letter dated 

April 13, 2011. (SSER 23, Section 9.2.1). 

Closed in SSER 27, Section 9.2.1.

NRR Closed

(92) Not used.

(93) TVA should confirm to the staff that testing of 

the Eagle 21 system has sufficiently 

demonstrated that two-way communication to 

the ICS is precluded with the described 

configurations. (SSER 23, Section 7.9.3.2). 

Closed in SSER 27, Section 7.9.

RII Closed

(94) TVA should provide to the staff either 

information that demonstrates that the WBN, 

Unit 2, Common Q post-accident monitoring 

system (PAMS) meets the applicable 

requirements in Institute of Electrical and 

Electronics Engineers (IEEE) Std. 603-1991, 

or justification for why the Common Q PAMS 

should not meet those requirements. 

(SSER 23, Section 7.5.2.2.3) Closed in 

SSER 26, Section 7.5.2.2.

NRR Closed

(95) TVA should update FSAR Table 7.1-1, “Watts 

Bar Nuclear Plant NRC Regulatory Guide 

Conformance,” to reference IEEE Std. 603-

1991 for the WBN, Unit 2, Common Q PAMS. 

(SSER 23, Section 7.5.2.2.3) Closed in 

SSER 25, Section 7.5.2.2.

NRR Closed
HH

-16

(96) TVA should (1) update FSAR Table 7.1-1 to 

include RG 1.100, Revision 3, for the 

Common Q PAMS, or (2) demonstrate that the 

Common Q PAMS is in conformance with RG 

1.100, Revision 1, or provide justification for 

not conforming. (SSER 23, Section 7.5.2.2.3) 

Closed in SSER 25, Section 7.5.2.2.

NRR Closed

(97) TVA should demonstrate that the WBN, 

Unit 2, Common Q PAMS is in conformance 

with RG 1.153, Revision 1, or provide 

justification for not conforming. (SSER 23, 

Section 7.5.2.2.3) Closed in SSER 25, 

Section 7.5.2.2.

NRR Closed

(98) TVA should demonstrate that the WBN, Unit 

2, Common Q PAMS is in conformance with 

RG 1.152, Revision 2, or provide justification 

for not conforming. (SSER 23, 

Section 7.5.2.2.3) Closed in SSER 26, 

Section 7.5.2.2.3.

NRR Closed

(99) TVA should update FSAR Table 7.1-1 to 

reference IEEE 7-4.3.2-2003 as being 

applicable to the WBN, Unit 2, Common Q 

PAMS. (SSER 23, Section 7.5.2.2.3; 

SSER 25, Section 7.5.2.2) Closed in SSER 

25, Section 7.5.2.2.

NRR Closed

(100) TVA should update FSAR Table 7.1-1 to 

reference RG 1.168, Revision 1; IEEE 1012-

1998; and IEEE 1028-1997 as being 

applicable to the WBN, Unit 2, Common Q 

PAMS. (SSER 23, Section 7.5.2.2.3) 

Closed in SSER 25, Section 7.5.2.2.

NRR Closed

(101) TVA should demonstrate that the WBN, Unit 

2, Common Q PAMS application software is in 

conformance with RG 1.168, Revision 1, or 

provide justification for not conforming. (SSER 23, Section 7.5.2.2.3) Closed in 

SSER 26, Section 7.5.2.2.

NRR Closed

(102) TVA should update FSAR Table 7.1-1 to 

reference RG 1.209 and IEEE Std. 323-2003 

as being applicable to the WBN, Unit 2, 

Common Q PAMS. (SSER 23, Section 

7.5.2.2.3) Closed in SSER 25, 

Section 7.5.2.2.

NRR Closed

(103) TVA should demonstrate that the WBN, 

Unit 2, Common Q PAMS conforms to RG 

1.209 and IEEE Std. 323-2003, or provide 

justification for not conforming. (SSER 23, 

Section 7.5.2.2.3) Closed in SSER 25, 

Section 7.5.2.2.

NRR Closed
HH

-17

(104) CI The NRC staff will review the Westinghouse 

Electric Corporation self-assessment to verify 

that it the WBN, Unit 2, PAMS is compliant to 

the V&V requirements in the software program 

manual or that deviations from the 

requirements are adequately justified. 

(SSER 23, Section 7.5.2.2.3.4.2) Closed in 

SSER 25, Section 7.5.2.2.

NRR Closed

(105) TVA should produce an acceptable 

description of how the WBN, Unit 2, Common 

Q PAMS System Requirements Specification 

and software requirements specification (SRS) 

implement the design-basis requirements of 

IEEE Std. 603-1991 Clause 4. (SSER 23, 

Section 7.5.2.2.3.4.3.1) Closed in SSER 26, 

Section 7.5.2.2.

NRR Closed

(106) TVA should produce a final WBN, Unit 2, 

Common Q PAMS SRS that is independently 

reviewed. (SSER 23, Section 7.5.2.2.3.4.3.1) 

Closed in SSER 25, Section 7.5.2.2.

NRR Closed

(107) CI TVA should provide to the NRC staff 

documentation to confirm that the final WBN, 

Unit 2, Common Q PAMS software design 

descriptions that are independently reviewed. 

(SSER 23, Section 7.5.2.2.3.4.3.2) Closed in 

SSER 25, Section 7.5.2.2.

NRR Closed

(108) TVA should demonstrate to the NRC staff that 

there are no synergistic effects between 

temperature and humidity for the Common Q 

PAMS equipment. (SSER 23, Section 

7.5.2.2.3.5.2) Closed in SSER 26, 

Section 7.5.2.2.

NRR Closed

(109) TVA should demonstrate to the NRC staff 

acceptable data storm testing of the Common 

Q PAMS. (SSER 23, Section 7.5.2.2.3.7.1.8) 

Closed in SSER 25, Section 7.5.2.2.

NRR Closed

(110) TVA should provide information to the NRC 

staff describing how the WBN, Unit 2, 

Common Q PAMS design supports periodic 

testing of the RVLIS function. (SSER 23, 

Section 7.5.2.2.3.9.2.6) Closed in SSER 26, 

Section 7.5.2.2.

NRR Closed

(111) TVA should confirm to the staff that there are 

no changes required to the TSs as a result of 

the modification installing the Common Q 

PAMS. If any changes to the TSs are 

required, TVA should provide the changes to 

the NRC staff for review. (SSER 23, 

Section 7.5.2.2.3.11) Closed in SSER 26, 

Section 7.5.2.2.

NRR Closed
HH

-18

(112) CI TVA should provide an update to the FSAR 

reflecting the radiation protection design 

features descriptive information provided in its 

letter dated October 4, 2010. (SSER 24, 

Section 12.4) Closed in SSER 26, 

Section 12.4.

NRR Closed

(113) CI TVA should provide an update to the FSAR 

reflecting the justification for the periodicity of 

the channel operability test frequency for WBN 

non-safety-related area radiation monitors. 

(SSER 24, Section 12.4) Closed in SSER 26, 

Section 12.4.

NRR Closed

(114) CI TVA should update the FSAR to reflect that 

WBN meets the radiation monitoring 

requirements of 10 CFR 50.68. (SSER 24, 

Section 12.4) Closed in SSER 26, 

Section 12.4.

NRR Closed

(115) CI TVA should update the FSAR to reflect the 

information regarding design changes to be 

implemented to lower radiation levels as 

provided in its letter to the NRC dated June 3, 

2010. (SSER 24, Section 12.5). Closed in 

SSER 27, Section 12.5

NRR Closed

(116) CI TVA should update the FSAR to reflect the 

qualification standards of the radiation 

protection manager as provided in its letter to 

the NRC dated October 4, 2010. (SSER 24, 

Section 12.6) Closed in SSER 26, 

Section 12.6.

NRR Closed

(117) CI TVA should update the FSAR to reflect the 

calculational basis for access to vital areas as 

provided in its letter dated February 25, 2011. 

(SSER 24, Section 12.7.1). Closed in 

SSER 27, Section 12.7.1

NRR Closed

(118) TVA should provide to the NRC staff a 

description of how the other vanadium 

detectors within the in-core instrumentation 

thimble assembly (IITA) would be operable 

following the failure of a self-powered neutron 

detector (SPND). (SSER 24, Section 

7.7.1.9.2) Closed in SSER 26, 

Section 7.7.1.9.

NRR Closed

(119) TVA should submit WNA-CN-00157-WBT, 

Revision 0, to the NRC by letter. The NRC 

staff should confirm by review of WNA-CN-

00157-WBT, Revision 0, that no credible 

source of faulting can negatively impact the 

CETs or PAMS train. (SSER 24, Section 

7.7.1.9.5) Closed in SSER 25, 

Section 7.7.1.9.

NRR Closed
HH-19

(120) TVA must confirm to the NRC staff that the 

maximum over-voltage or surge voltage that 

could affect the system is 264 VAC, assuming 

that the power supply cable to the signal 

processing system (SPS) cabinet is not routed 

with other cables greater than 264 VAC. 

(SSER 24, Section 7.7.1.9.5; SSER 25, 

Section 7.7.1.9) Closed in SSER 26, 

Section 7.7.1.9.

NRR Closed

(121) TVA should submit the results to the NRC 

staff of a 600 VDC dielectric strength test 

performed on the IITA assembly. (SSER 24, 

Section 7.7.1.9.5) Closed in SSER 26, 

Section 7.7.1.9.

NRR Closed

(122) TVA should confirm to the NRC staff that 

different divisions of safety power are supplied 

to the in-core instrumentation system SPS 

cabinets, with the power cables routed in 

separate shielded conduits. (SSER 24, 

Section 7.7.1.9.5) Closed in SSER 25, 

Section 7.7.1.9.

NRR Closed

(123) TVA should provide an explanation to the 

NRC staff of how the system will assign a data 

quality value to notify the power distribution 

calculation software to disregard data from a 

failed SPND. (SSER 24, Section 7.7.1.9.5) 

Closed in SSER 26, Section 7.7.1.9.

NRR Closed

(124) While the BEACON datalink on the Application 

server can connect to either BEACON 

machine, only BEACON A is used for 

communication. TVA should clarify to the 

NRC staff whether automatic switchover to the 

other server is not permitted. (SSER 24, 

Section 7.7.1.9.5) Closed in SSER 25, 

Section 7.7.1.9.

NRR Closed

(125) TVA should provide clarification to the NRC 

staff of the type of connector used with the MI 

cable in Unit 2, and which EQ test is 

applicable. (SSER 24, Section 7.7.1.9.5) 

Closed in SSER 26, Section 7.7.1.9.

NRR Closed

(126) To enable the NRC staff to evaluate and 

review the IITA EQ, TVA should provide the 

summary report of the EQ for the IITA. 

(SSER 24, Section 7.7.1.9.5) Closed in 

SSER 26, Section 7.7.1.9.

NRR Closed
HH-20

(127) TVA should provide a summary to the NRC 

staff of the electro-magnetic 

interference/radio-frequency interference 

testing for the MI cable electro-magnetic 

compatibility (EMC) qualification test results. 

(SSER 24, Section 7.7.1.9.5) Closed in 

SSER 26, Section 7.7.1.9.

NRR Closed

(128) TVA should submit the seismic qualification 

test report procedures and results for the SPS 

cabinets to the NRC staff for review. 

(SSER 24, Section 7.7.1.9.5) Closed in SSER 

25, Section 7.7.1.9.

NRR Closed

(129) TVA should verify to the NRC staff resolution 

of the open item in WNA-CN-00157-WBT for 

the Quint power supply (to be installed in the 

SPS cabinet) to undergo EMC testing of 4 kV 

to validate the assumptions made in the 

Westinghouse analysis. (SSER 24, 

Section 7.7.1.9.5) Closed in SSER 26, 

Section 7.7.1.9.

NRR Closed

(130) TVA should provide a summary to the NRC 

staff of the EMC qualification test results of the 

SPS cabinets. (SSER 24, Section 7.7.1.9.5) 

Closed in SSER 25, Section 7.7.1.9.

NRR Closed

(131) TVA should review the EOP action level 

setpoint to account for the difference between 

core exit temperature readings for Units 1 and

2, and confirm the EOP action level setpoint to 

the NRC staff. (SSER 24, Section 7.7.1.9.5). 

Closed in SSER 27, Section 7.7.

NRR Closed

(132) TVA must provide the NRC staff with analyses 

of the boron dilution event that meet the 

criteria of Standard Review Plan (SRP) 

Section 15.4.6, including a description of the 

methods and procedures used by the 

operators to identify the dilution path(s) and 

terminate the dilution, in order for the staff to 

determine that the analyses comply with 

GDC 10. (SSER 24, Section 15.2.4.4) 

Closed in SSER 26, Section 15.2.4.4.

NRR Closed
HH-21

(133) In order to confirm the stability analysis of the 

sand baskets used by TVA in the WBN, 

Unit 2, licensing basis, TVA will perform either 

a hydrology analysis without crediting the use 

of the sand baskets at the Fort Loudoun Dam 

for the seismic dam failure and flood 

combination, or TVA will perform a seismic 

test of the sand baskets, as stated in TVA’s 

letter dated April 20, 2011. TVA will report the 

results of this analysis or test to the NRC by 

October 31, 2011. (SSER 24, Section 2.4.10). 

Closed in SSER 27, Section 2.4.10.

NRR Closed

(134) TVA should provide to the NRC staff 

supporting technical justification for the 

statements in Amendment No. 104 of FSAR 

Section 2.4.4.1, “Dam Failure Permutations,” 

page 2.4-32 (in the section “Multiple Failures”) 

that, “Fort Loudoun, Tellico, and Watts Bar 

have previously been judged not to fail for the 

operating basis earthquake (OBE) (0.09 g). 

Postulation of Tellico failure in this 

combination has not been evaluated but is 

bounded by the SSE [safe shutdown 

earthquake] failure of Norris, Cherokee, 

Douglas and Tellico.” (SSER 24, 

Section 2.4.10) Closed in SSER 28, 

Section 2.4.10

NRR Closed

(135) TVA has not provided the analysis required by 

10 CFR Part 50, Appendix I, subsection II.D. 

TVA must demonstrate with a cost-benefit 

analysis that a sufficient reduction in the 

collective dose to the public within a 50-mile 

radius would not be achieved by reasonable 

changes to the design of the WBN gaseous 

effluent processing systems. (SSER 24, 

Section 11.3) Closed in SSER 25, 

Section 11.3.

NRR Closed

(136) CI The joint frequency distribution summary for 

the data from 1991 through 2010 provided by 

letter dated November 7, 2011, and a 

discussion of the long-term 

representativeness of these data should be 

provided in the WBN, Unit 2, FSAR. Upon 

receipt of the UFSAR, the NRC staff will 

confirm that these updates have been made 

by TVA. (SSER 25, Section 2.3.3) Closed in 

SSER 26, Section 2.3.3.

NRR Closed
HH-22

(137) CI The NRC staff will confirm, upon receipt, that 

TVA integrated the updated control room

atmospheric diversion estimate (χ/Q) values 

from its letter dated September 15, 2011, into 

a future amendment of the FSAR. (SSER 25, 

Section 2.3.4) Closed in SSER 26, 

Section 2.3.4.

NRR Closed

(138) CI Upon receipt of the updated Offsite Dose 

Calculation Manual (ODCM), the NRC staff 

will confirm that corresponding revisions 

related to the updated annual average χ/Q 

and deposition factor values have been made 

to the ODCM. (SSER 25, Section 2.3.5) 

Closed in SSER 26, Section 2.3.5.

NRR Closed

(139) CI The results of the cost-benefit analysis 

required by 10 CFR Part 50, Appendix I, 

subsection II.D, should be provided in the 

WBN, Unit 2, FSAR. Upon receipt of the 

UFSAR, the NRC staff will confirm that the 

update has been made by TVA. (SSER 25, 

Section 11.3). Closed in SSER 27, 

Section 11.3

NRR Closed

(140) CI TVA to confirm to the staff the completion of 

the Unit 2, OMA feasibility walkdowns. (SSER 

26, Appendix FF, Section 8.0). Closed in 

SSER 29, Section 9.5

NRR Closed

(141) CI TVA to confirm to the staff the completion of 

the multiple spurious operation scenario 

resolution actions for scenarios that only affect 

Unit 2. (SSER 26, Appendix FF, Section 8.0). 

Closed in SSER 29, Section 9.5

NRR Closed

(142) CI TVA to confirm to the staff the completion of 

the electrical coordination modifications. 

(SSER 26, Appendix FF, Section 8.0). Closed 

in SSER 29, Section 9.5)

NRR Closed

(143) CI TVA to confirm the as-built fire protection 

report aligns with as-designed fire protection 

report. Gaps to be submitted to the NRC for 

approval. (SSER 26, Appendix FF, Section 

8.0). Closed in SSER 29, Section 9.5

NRR Closed






NUREG

-0847

Supplement 29

Safety Evaluation Report Related to the Operation of 

Watts Bar Nuclear Plant, Unit 2