*U.S.NRC

United States Nuclear Regulatory Commission

Protecting People and the Environment

NUREG-0847

Supplement 25

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
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*U.S.NRC

United States Nuclear Regulatory Commission

Protecting People and the Environment

NUREG-0847

Supplement 25

Safety Evaluation Report

Related to the Operation of

Watts Bar Nuclear Plant, Unit 2

Docket Number 50-391

Tennessee Valley Authority

Manuscript Completed: November 2011

Date Published: December 2011

Office of Nuclear Reactor Regulation

ABSTRACT

This report supplements the safety evaluation report (SER), NUREG-0847 (June 1982),

Supplement No. 24 (September 2011, Agencywide Documents Access and Management

System (ADAMS) Accession No. ML1 1277A148), with respect to 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).

In its SER and Supplemental SER (SSER) Nos. 1 through 20 issued by the Office of Nuclear

Reactor Regulation (NRR) of the U.S. Nuclear Regulatory Commission (NRC or the staff), the

staff documented its safety evaluation 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, the staff addressed TVA's application for a license to operate WBN Unit 2, and

provided information regarding the status of the items remaining to be resolved, which 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. Beginning with SSER 22, the staff

documented its ongoing evaluation and closure of open items in support of TVA's application for

a license to operate WBN Unit 2.

In this and future SSERs, the staff continues its documentation of its review of open items in

support of TVA's application for an operating license for WBN Unit 2.

iii

TABLE OF CONTENTS

Page

A B S T R A C T .................................................................................................... iii

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

LIS T O F TA B LE S ................................................................................................................ xi

I INTRODUCTION AND DISCUSSION .......................................................... 1-1

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

1.2 General Plant Description ......................................................................... 1-2

1.7 Summary of Outstanding Issues ..................................................... 1-4

1.8 Confirmatory Issues ..................................................................... 1-27

1.9 License Conditions ................................................................................... 1-27

1.10 Unresolved Safety Issues ......................................................................... 1-28

1.13 Implementation of Corrective Action Programs and Special Programs .... 1-28

1.13.1 Corrective Action Programs .......................................................... 1-29

1.13.2 Special Programs ......................................................................... 1-30

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

2 SITE CHARACTERISTICS ..................................................................................... 2-1

2.3 M eteorology .............................................................................................. 2-1

2.3.3 Onsite Meteorological Measurements Program ............................. 2-1

2.3.4 Short-Term (Accident) Atmospheric Dispersion Estimates ............. 2-2

2.3.5 Long-Term (Routine) Atmospheric Dispersion Estimates ............... 2-3

5 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS .......................... 5-1

5.3 Reactor Vessel ......................................................................................... 5-1

5.3.1 Reactor Vessel Materials ................................................................ 5-1

5.3.2 Pressure-Temperature Limits ......................................................... 5-2

7 INSTRUMENTATION AND CONTROLS ................................................................ 7-1

7.5 Safety-Related Display Instrumentation ................................................... 7-1

7.5.2 Post-Accident Monitoring System ................................................... 7-1

7.7 Control Systems Not Required for Safety ................................................. 7-7

7.7.1 System Description ......................................................................... 7-7

11 RADIOACTIVE WASTE SYSTEM .......................................................................... 11-1

11.3 Gaseous Waste Management .................................................................. 11-1

13 CONDUCT OF OPERATIONS ................................................................................ 13-1

13.1 Organizational Structure of Applicant ....................................................... 13-1

13.1.3 Plant Staff Organization ................................................................ 13-1

v
TABLE OF CONTENTS (Continued)

Paqe

15 ACC IDENT A NA LYSIS ........................................................................................ 15-1

15.4 Radiological Consequences of Accidents ................................................. 15-1

15.4.1 Loss-of-Coolant Accident .............................................................. 15-2

15.4.2 Main Steamline Break Outside Containment ................ 15-3

15.4.3 Steam Generator Tube Rupture (SGTR) ................... 15-4

15.4.4 Control Rod Ejection Accident ...................................................... 15-5

15.4.5 Fuel Handling Accident ................................................................. 15-5

15.4.6 Failure of Small Lines Carrying Coolant Outside Containment .... 15-8

15.4.7 Environmental Consequences of a Postulated Loss of AC

Power to the Plant Auxiliaries ....................................................... 15-8

15.4.8 Environmental Consequences of a Postulated Waste Gas

Decay Tank R upture ..................................................................... 15-10

APPENDIX A CHRONOLOGY OF RADIOLOGICAL REVIEW OF WATTS BAR

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

APPENDIX E PRINCIPAL CONTRIBUTORS TO SSER 25 ........................................... E-1

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

HH

vi
ABBREVIATIONS

ABGTS auxiliary building gas treatment system

AC or ac alternating current

ADAMS Agencywide Documents Access and Management System

ALARA as low as reasonably achievable

ANSI American National Standards Institute

ANS American Nuclear Society

ASME American Society of Mechanical Engineers

AST alternative source term

ASTM American Society for Testing and Materials

BEACON Westinghouse Best Estimate Analyzer for Core Operations-Nuclear

BL bulletin

BTP Branch Technical Position

BWR boiling-water reactor

CAP corrective action program

CCS component cooling system

CET core exit thermocouple

cfm cubic feet per minute

CFR Code of Federal Regulations

CI confirmatory issue

Ci curie

CLB current licensing basis

COMS cold overpressure mitigation system

COT channel operability test

CR control room

CSST common station service transformer

CST condensate storage tank

C, Charpy V-notch

DBA design basis accident

DC or dc direct current

DCF dose conversion factor

DCN design change notice

DCS distributed control system

DEI dose equivalent iodine-131

DF decontamination factor

D/Q deposition factor

EAB exclusion area boundary

ECCS emergency core cooling system

EDCR Engineering Document Construction Release

EDG emergency diesel generator

EGTS emergency gas treatment system

EMC electromagnetic compatibility

EMI/RFI electromagnetic/radiofrequency interference

EOF emergency operations facility

EOP emergency operating procedure

EPA Environmental Protection Agency or electrical penetration assemblies

vii
EPRI Electric Power Research Institute

EQ environmental qualification

ERCW essential raw cooling water

ERDS emergency response data system

ESF engineered safety feature

FHA fuel handling accident

FSAR final safety analysis report

FW feedwater

GDC general design criterion/criteria

GL generic letter

gpm gallons per minute

GSI generic safety issue

HEPA high efficiency particulate air

HRCAR high range containment air radiation

HVAC heating, ventilation, and air conditioning

ICC inadequate core cooling

ICRP International Commission on Radiological Protection

ICS integrated computer system

IE Office of Inspection and Enforcement

IEB Office of Inspection and Enforcement Bulletin

IEEE Institute of Electrical and Electronics Engineers

IIS in-core instrumentation system

IITA in-core instrumentation thimble assembly

IPE individual plant examination

IPEEE individual plant examination of external events

JFD joint frequency distribution

kHz kilohertz

kV kilovolt

kVA kilovolt ampere

kW kilowatt

LOCA loss-of-coolant accident

LOOP loss of offsite power

LPMS loose part monitoring system

LPZ low-population zone

LTOP low-temperature overpressure protection

LWR light-water reactor

MSIV main steam isolation valve

MSLB main steam line break

MTEB Materials Engineering Branch (of NRR)

MTP maintenance and test panel

MVA megavolt-ampere

MWt megawatts thermal

NGDC New Generation Development and Construction

NPP Nuclear Performance Plan

NP-REP Nuclear Power Radiological Emergency Plan

NRC Nuclear Regulatory Commission

viii
NRR Office of Nuclear Reactor Regulation

NSSS nuclear steam supply system

NUREG report prepared by NRC staff

OBE operating basis earthquake

ODCM Offsite Dose Calculation Manual

OL operating license

OSG original steam generator

PAD performance analysis and design

PAMS postaccident monitoring system

PDR public document room

PMF probable maximum flood

PORV power-operated relief valve

PTLR Pressure and Temperature Limits Report

PWR pressurized-water reactor

RAI request for additional information

RBPVS reactor building purge ventilation system

RCCA rod cluster control assembly

RCS reactor coolant system

RG Regulatory Guide

RPM revolutions per minute

RPV reactor pressure vessel

RV reactor vessel

SDD software design description

SE safety evaluation

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

SFP spent fuel pool

SG steam generator

SPM software program manual

SPND self-powered neutron detector

SP special program

SPS signal processing system

SRO senior reactor operator

SRP Standard Review Plan, NUREG-0800

SRS software requirements specification

SSE safe shutdown earthquake

SSER Supplemental SER

Std. Standard

SV safety valve

TID total integrated dose

TMI Three Mile Island

TPBAR tritium production burnable absorber rod

TS technical specification

TSTF Technical Specification Task Force

TVA Tennessee Valley Authority

UHS ultimate heat sink

V&V verification and validation

WBA Web-based ADAMS

WBN Watts Bar Nuclear Plant

ix
WCAP Westinghouse Commercial Atomic Power (report)

WEC Westinghouse Electric Corporation

WINCISE Westinghouse INCore Information, Surveillance, and Engineering system

x/Q atmospheric dispersion estimate

x
LIST OF TABLES

Paqe

11.1 NRC Staff's 50-Mile Collective Population Dose per Release Source ..... 11-3

15.1 Radiological Consequences of Design-Basis Accidents .......................... 15-11

15.2 Assumptions Used in the Calculation of the Radiological Consequences

Follow ing a Postulated LOCA ................................................................... 15-12

15.3 Assumptions Used to Evaluate the Radiological Consequences

Following a Postulated MSLB Accident Outside Containment ................. 15-14

15.4 Assumptions Used to Evaluate the Radiological Consequences

Following a Postulated SGTR Accident .................................................... 15-15

15.5 Assumptions Used to Evaluate the Radiological Consequences

Following a Postulated FHA Inside Closed Containment - RG 1.25

A na lys is .................................................................................................... 15 -16

15.6 Assumptions Used to Evaluate the Radiological Consequences

Following a Postulated FHA in the Auxiliary Building or in Open

Containm ent - RG 1.183 Analysis ............................................................ 15-17

15.7 Assumptions Used to Evaluate the Radiological Consequences

Following the Postulated Loss of AC Power to the Plant Auxiliaries ........ 15-18

15.8 Assumptions Used to Evaluate the Radiological Consequences

Following a Postulated Waste Gas Decay Tank Rupture ......................... 15-19

15.9 C R P aram eters ......................................................................................... 15-20

xi

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 northeast of

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

reactors (PWRs) within ice condenser containments.

In June 1982, the Nuclear Regulatory Commission staff (NRC staff or staff) issued safety

evaluation report (SER), NUREG-0847, "Safety Evaluation Report related to the operation of

Watts Bar Nuclear Plant Units 1 and 2," regarding 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 Unit 1. TVA did not complete WBN Unit 2, and the

NRC did not make conclusions regarding it.

On March 4, 2009, 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, the staff provided information regarding the status of the WBN Unit 2 items that

remain 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

SSER 22, the staff began the documentation of its evaluation and closure of open items in

support of TVA's application for a license to operate WBN Unit 2.

In this and future SSERs, the staff will continue the documentation of its evaluation and closure

of open items in support of 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 staff

added additional chapters to address the overall assessment of the facility, Nuclear

Performance Plan 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 provided in Appendix A has been discontinued, and a reference is

provided 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

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

disclosure in accordance with 10 CFR 2.390.

Appendix HH includes an Action Items Table. This table provides a status of all the open items,

confirmatory issues, and proposed license conditions that must be resolved prior to completion

of an NRC finding of reasonable assurance on the OL application for WBN Unit 2. The staff will

1-1
maintain the Action Items Table and revise Appendix HH in future SSERs, and add new

appendices, as necessary.

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 NRC, NRC contractors, or NRC

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

inspection reports, and others, are assigned accession numbers. They are searchable and

accessible from 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 documents dates.

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

NRC Public Web site at http:llwww.nrc.gov/reading-rm/adams/faq.html#1.

All WBN documents may be accessed using WBN docket numbers 05000390 and 05000391 for

Units 1 and 2, respectively.

The WBN Unit 2 Project Manager is Patrick D. Milano, who may be contacted by calling

(301) 415-1457, by e-mail to Patrick.Milano•,nrc.,ov, or by writing to the following address:

Mr. Patrick D. Milano

U.S. Nuclear Regulatory Commission

Mail Stop O-8H4

Washington, D.C. 20555

1.2 General Plant Description

A general description of WBN Unit 2 was provided by the NRC staff in Section 1.2 of the SER

(NUREG-0847, dated June 1982). TVA provided a more detailed description of WBN Unit 2 in

Section 1.2, "General Plant Description," of Amendment 104 of the Final Safety Analysis Report

(FSAR), dated June 3, 2011. An updated general description is provided below in this section

of the SSER.

WBN Unit 2 is very similar in design to WBN Unit 1 and Sequoyah Nuclear Plant Units 1 and 2.

WBN Unit 2 uses a nuclear steam supply system (NSSS) incorporating a PWR and a 4-loop

reactor coolant system (RCS). The Unit 2 reactor core design rating is 3,411 megawatts

thermal (MWt). The license application NSSS power level is 3,425 MWt, which includes 14

MWt from the reactor coolant pumps. The reactor core is composed of fuel rods made of

slightly enriched uranium dioxide pellets enclosed in ZIRLO® tubes that are grouped and

supported into assemblies. The mechanical control rods consist of rod cluster control

assemblies (RCCAs) and burnable absorber rods that are inserted into the guide thimbles of the

fuel assemblies. The absorber sections of the RCCAs are fabricated of silver-indium-cadmium

alloy slugs sealed in stainless steel tubes. The absorber material in the burnable absorber rods

is in the form of borosilicate glass sealed in stainless steel tubes. The core fuel is loaded in

three regions, each using fuel of a different enrichment of uranium-235. The new fuel is

introduced into the outer region, moved inward at successive refuelings, and removed from the

inner region to the spent fuel pool (SFP).

1-2
Water will serve as both the moderator and the coolant. It will be circulated through the reactor

vessel and core by four vertical, single-stage centrifugal pumps, one located in the cold leg of

each loop. The coolant water heated by the reactor will be circulated through the four steam

generators where heat will be transferred to the secondary system to produce saturated steam,

and then it will be returned to the pumps to repeat the cycle. An electrically heated pressurizer

connected to the hot-leg piping of one of the loops will establish and maintain the reactor

coolant pressure and provide a surge chamber and a water reserve to accommodate reactor

coolant volume changes during operation.

The steam produced in the steam generators will be used to drive a tandem arrangement of one

double-flow high-pressure turbine and three double-flow low-pressure turbines driving a directcoupled generator at 1800 revolutions per minute (RPM). The generator has a nameplate rating

of 1,411,000 kilovolt-amperes (kVA) at 0.9 power factor with 75 pounds per square inch (psi)

hydrogen pressure. The unit uses a single pass surface condenser of a horizontal, triple

pressure, single shell type. Return to the steam generator is through three stages of feedwater

pumping and seven stages of feedwater heating. Safety relief valves and power operated relief

valves, as well as a turbine bypass to the condenser are provided in the steam lines. Cooling

water from the cooling tower basin will be pumped through the tubes of the condenser to

remove the heat from, and thus condense, the steam after it has passed through the main

turbine. The condenser condensate then will be pumped back to the steam generator to be

heated for another cycle. The condenser cooling water will be passed through two natural draft

cooling towers and returned to the cooling tower basin.

The reactor will be controlled by a coordinated combination of a soluble neutron absorber (boric

acid) and mechanical control rods whose drive shafts will allow the plant to accept step load

changes of 10 percent and ramp load changes of 5 percent per minute over the range of 15 to

100 percent of full power under normal operating conditions. With steam bypass, the plant will

also have the capability to accept a 50-percent step load rejection without reactor trip.

Plant protection systems are provided that automatically initiate appropriate action whenever a

monitored condition approaches preestablished limits. These protection systems will act to shut

down the reactor, close isolation valves, and initiate operation of the engineered safety features

should any or all of these actions be required. Supervision and control of both the NSSS and

the steam and power conversion system will be accomplished from the main control room.

The emergency core cooling system for the plant consists of accumulators, upper head

injection, and both high- and low-pressure injection subsystems with provisions for recirculation

of the borated water after the end of the injection phase. Various combinations of these

features will ensure core cooling for the complete range of postulated coolant pipe break sizes.

The NSSS is housed in a separate free-standing steel containment structure within a reinforced

concrete shield building. The containment employs the ice condenser pressure-suppression

concept. A common auxiliary building adjacent to the containment structure houses the

radioactive waste treatment facilities, components of the engineered safety features, and

various related auxiliary systems for each unit. The units share a common fuel handling facility

that contains a SFP and a new fuel storage facility.

The plant is supplied with electrical power by independent transmission lines from offsite power

sources. It has independent and redundant onsite emergency power supplies capable of

supplying power to shut down the plant safely or to operate the engineered safety features in

the event of an accident. The plant electric power system consists of the main generators, the

1-3
unit station service transformers, the common station service transformers, the emergency

diesel generators (EDGs), the batteries, and the electric distribution system. Under normal

operating conditions the main generators supply electrical power through isolated-phase buses

to the main step-up transformers and through the unit station service transformers (located

adjacent to the turbine building) to the nonsafety auxiliary power system. Offsite electrical

power supplies Class 1E circuits through the 161-kilovolt (kV) system via common station

service transformers C and D. The primaries of the unit station service transformers are

connected to the isolated-phase bus at a point between the generator terminals and the lowvoltage connection of the main transformers. During normal operation, station auxiliary power is

taken from the main generator through the unit station service transformers and from the 161-kV

system through the common station service transformers. The standby onsite power is supplied

by four diesel generators.

With the exception of some shared systems, separate and similar safety-related systems and

equipment are provided for WBN Units 1 and 2. The major structures at WBN are two reactor

buildings, a turbine building, an auxiliary building, a control building, a service and office

building, diesel generator buildings, an intake pumping station, and two natural draft cooling

towers.

1.7 Summary of Outstanding Issues

The staff documented its previous review and conclusions regarding 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 staff issued an OL for WBN Unit 1 in 1996. In the SER and SSERs 1

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

conclusions were made regarding an OL for WBN Unit 2. To establish the remaining scope and

the regulatory framework for the staff's review of an OL for WBN Unit 2, the staff reviewed the

SER and SSERs 1 thorough 20. Based on this review, the staff identified "resolved" topics (i.e.,

out of scope for review) and "open" topics (i.e., in scope for staff review) for WBN Unit 2. Where

it was not clear whether the SER topic applied to Unit 2 or not, the 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 staff documented this evaluation in SSER 21 as the baseline for

resumption of the review of the OL application for Unit 2. Thus, SSER 21 reflects the status of

the staffs review of WBN Unit 2 up to 1995. The staff notes that a subsequent, more detailed

assessment may find some topics conservatively identified in the initial assessment as "open"

that should be redefined as "closed." Conversely, the NRC staff notes that there may be

circumstances that could result 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 will be identified by a footnote in future SSERs 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 FSAR until Amendment 91.

FSAR Amendment 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

the FSAR Amendment 91 into two separate FSARs for WBN Units 1 and 2. TVA has submitted

WBN Unit 2 FSAR Amendments 92 through 102 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. These FSAR amendments are currently under staff review. The staffs review of

these FSAR changes is documented in SSER 22 and subsequent supplements.

1-4
Additional general topics (e.g., financial qualifications that were not included in SSER 21, but

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

subsequent supplements.

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

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

be maintained in this and future supplements to reflect the updated status of review for each

topic.

ISSUE STATUS TABLE

Issue Status Section Note

(1) Site Envelope

(2) Geography and Demography

(3) Site Location and Description

(4) Exclusion Area Authority and

Control

(5) Population Distribution

(6) Conclusions

(7) Nearby Industrial, Transportation,

and Military Facilities

(8) Transportation Routes

(9) Nearby Facilities

(10) Conclusions

(11) Meteorology

(12) Regional Climatology

(13) Local Meteorology

(14) Onsite Meteorological

Measurements Program

(15) Short-Term (Accident) Atmospheric

Diffusion Estimates

(16) Long-Term (Routine) Diffusion

Estimates

Resolved (SSER 22)

Resolved (SER)

(SSER 22)

Resolved (SER)

(SSER 22)

Resolved (SER)

(SSER 22)

Resolved (SER)

(SSER 22)

Resolved (SSER 22)

Resolved (SER)

(SSER 22)

Resolved (SER)

(SSER 22)

Resolved (SER)

(SSER 22)

(SER)

(SSER 22)

Resolved (SER)

(SSER 22)

Resolved (SER)

(SSER 22)

Resolved (SER)

(SSER 22)

(SSER 25)

Resolved (SER)

(SSER 14)

(SSER 22)

Resolved (SER)

(SSER 14)

(SSER 22)

2

2.1

2.1.1

2.1.2

2.1.3

2.1.4

2.2

2.2.1

2.2.2

2.2.3

2.3

2.3.1

2.3.2

2.3.3

2.3.4

2.3.5

2.4

2.4.1

2.4.2

2.4.3

3

3

(17)

(18)

(19)

(20)

Hydrologic Engineering

Introduction

Hydrologic Description

Flood Potential

Resolved

Resolved

Resolved

(SER)

(SER)

(SER)

1-5
Issue Status Section Note

(21) Local Intense Precipitation in Plant

Area

(22) Roof Drainage

(23) Ultimate Heat Sink

(24) Groundwater

(25) Design Basis for Subsurface

Hydrostatic Loading

(26) Transport of Liquid Releases

(27) Flooding Protection Requirements

(28) Geological, Seismological, and

Geotechnical Engineering

(29) Geology

(30) Seismology

(31) Surface Faulting

(32) Stability of Subsurface Materials

and Foundations

(33) Stability of Slopes

(34) Embankments and Dams

(35) References

(36) Design Criteria - Structures,

Components, Equipment, and

Systems

(37) Introduction

(38) Conformance With General Design

Criteria

(39) Conformance With Industry Codes

and Standards

(40) Classification of Structures,

Systems and Components

(41) Seismic Classifications

(42) System Quality Group Classification

(43) Wind and Tornado Loadings

(44) Wind Loading

Resolved (SER)

Resolved (SER)

Resolved (SER)

Resolved (SER)

Resolved (SER)

(SSER 3)

Resolved (SER)

(SSER 22)

Open (SER)

(Inspection) (SSER 24)

Resolved (SER)

(SSER 24)

Resolved (SER)

Resolved (SER)

Resolved (SER)

Resolved (SER)

(SSER 3)

(SSER 9)

(SSER 11)

Resolved (SER)

Resolved (SER)

(SSER 22)

(SER)

(SSER 22)

Resolved (SER)

Resolved (SER)

Resolved (SSER 14)

(SSER 22)

Resolved (SER)

(SSER 3)

(SSER 5)

(SSER 6)

(SSER 8)

Open (NRR) (SER)

(SSER 3)

(SSER 6)

(SSER 7)

(SSER 9)

(SSER 22)

Resolved (SER)

2.4.4

2.4.5

2.4.6

2.4.7

2.4.8

2.4.9

2.4.10

2.5

2.5.1

2.5.2

2.5.3

2.5.4

2.5.5

2.5.6

2.6

3

3.1

3.1.1

3.1.2

3.2

3.2.1

3.2.2

3.3

3.3.1

2

1

1

2

1-6
Issue Status Section Note

(45) Tornado Loading

(46) Flood Level (Flood) Design

(47) Flood Protection

(48) Missile Protection

(49) Missile Selection and Description

(50) Structures, Systems, and

Components to be Protected from

Externally Generated Missiles

(51) Barrier Design Procedures

(52) Protection Against the Dynamic

Effects Associated with the

Postulated Rupture of Piping

(53) Plant Design for Protection Against

Postulated Piping Failures in Fluid

System Outside Containment

(54) Determination of Break Locations

and Dynamic Effects Associated

with the Postulated Rupture of

Piping

(55) Leak-Before-Break Evaluation

Procedures

(56) Seismic Design

(57) Seismic Input

(58) Seismic Analysis

(59) Seismic Subsystem Analysis

(60) Seismic Instrumentation

(61) Design of Seismic Category I

Structures

(62) Steel Containment

Resolved (SER)

Resolved (SER)

Resolved (SER)

(SSER 9)

(SSER 14)

(SSER 22)

Resolved (SER)

(SSER 2)

(SSER 22)

Resolved (SER)

Open (NRR) (SER)

(SSER 6)

(SSER 11)

Resolved (SER)

(SSER 14)

(SSER 22)

Resolved (SER)

(SSER 14)

(SSER 22)

Resolved (SSER 5)

(SSER 12)

(SSER 22)

(SSER 24)

Resolved (SER)

(SSER 6)

Resolved (SER)

(SSER 6)

(SSER 9)

(SSER 16)

Resolved (SER)

(SSER 6)

(SSER 8)

(SSER 11)

(SSER 16)

Resolved (SER)

(SSER 6)

(SSER 7)

(SSER 8)

(SSER 9)

(SSER 12)

(SSER 22)

Resolved (SER)

Resolved (SER)

(SSER 9)

Resolved (SER)

(SSER 3)

3.3.2

3.4

3.4.1

3.5

3.5.1

3.5.2

3.5.3

3.6

3.6.1

3.6.2 3

3.6.3

3.7

3.7.1

3.7.2

2

2

2

3.7.3

3.7.4

3.8

3.8.1

1

2

1-7
Issue Status Section Note

(63) Concrete and Structural Steel

Internal Structures

(64) Other Seismic Category I Structures

(65) Foundations

(66) Mechanical Systems and

Components

(67) Special Topics for Mechanical

Components

(68) Dynamic Testing and Analysis of

Systems, Components, and

Equipment

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

Components, Component

Structures, and Core Support

Structures

(70) Control Rod Drive Systems

(71) Reactor Pressure Vessel Internals

(72) Inservice Testing of Pumps and

Valves

(73) Seismic and Dynamic Qualification

of Seismic Category I Mechanical

and Electrical Equipment

(74) Environmental Qualification of

Mechanical and Electrical

Equipment

(75) Threaded Fasteners - ASME Code

Class 1, 2, and 3

(76) Reactor

Resolved (SER)

(SSER 7)

Open (NRR) (SER)

(SSER 14)

(SSER 16)

Resolved (SER)

Resolved (SER)

Resolved (SER)

(SSER 6)

(SSER 13)

(SSER 22)

Resolved (SER)

(SSER 14)

(SSER 22)

Resolved (SER)

(SSER 3)

(SSER 4)

(SSER 6)

(SSER 7)

(SSER 8)

(SSER 15)

(SSER 22)

Resolved (SER)

Open (SER)

(SSER 23)

Open (NRR) (SER)

(SSER 5)

(SSER 12)

(SSER 14)

(SSER 18)

(SSER 20)

(SSER 22)

Resolved (SER)

(SSER 1)

(SSER 3)

(SSER 4)

(SSER 5)

(SSER 6)

(SSER 8)

(SSER 9)

(SSER 23)

Open (NRR) (SSER 15)

(SSER 22)

3.8.2

3.8.3

3.8.4

3.9

3.9.1

3.9.2

3.9.3

3.9.4

3.9.5

3.9.6

3.10

3.11

Resolved (SSER 22) 3.13

4

1-8
Issue Status Section Note

(77) Introduction

(78) Fuel System Design

(79) Description

(80) Thermal Performance

(81) Mechanical Performance

(82)

(83)

Surveillance

Fuel Design Considerations

(84) Nuclear Design

(85) Design Basis

(86) Design Description

(87) Analytical Methods

(88) Summary of Evaluation Findings

(89) Thermal-Hydraulic Design

(90) Performance in Safety Criteria

(91) Design Bases

(92) Thermal-Hydraulic Design

Methodology

(93) Operating Abnormalities

(SER)

(SSER 23)

(SSER 23)

Resolved (SER)

(SSER 13)

(SSER 23)

Open (NRR) (SER)

(SSER 2)

(SSER 23)

Resolved (SER)

(SSER 2)

(SSER 10)

(SSER 13)

(SSER 23)

(SER)

Resolved (SER)

(SSER 23)

(SSER 23)

Resolved (SER)

(SSER 13)

(SSER 23)

Resolved (SER)

(SSER 13)

(SSER 15)

(SSER 23)

Resolved (SER)

(SSER 23)

Resolved (SER)

(SSER 23)

(SSER 23)

Resolved (SER)

(SSER 23)

Resolved (SER)

(SSER 12)

(SSER 23)

Resolved (SER)

(SSER 6)

(SSER 8)

(SSER 12)

(SSER 13)

(SSER 16)

SE dated

6/13/89

(SSER 23)

Resolved (SER)

(SSER 13)

(SSER 23)

4.1

4.2

4.2.1

4.2.2

4.2.3

4.2.4

4.2.5

4.3

4.3.1

4.3.2

4.3.3

4.3.4

4.4

4.4.1

4.4.2

4.4.3

4.4.4

1-9
Issue Status Section Note

(94) Loose Parts Monitoring System

(95) Thermal-Hydraulic Comparison

(96) N-1 Loop Operation

(97) Instrumentation for Inadequate Core

Cooling Detection (TMI Action Item

II.F.2)

(98) Summary and Conclusion

(99) Reactor Materials

(100) Control Rod Drive Structural

Materials

(101) Reactor Internals and Core Support

Materials

(102) Functional Design of Reactivity

Control Systems

(103) Reactor Coolant System and

Connected Systems

(104) Summary Description

(105) Integrity of Reactor Coolant

Pressure Boundary

(106) Compliance with Codes and Code

Cases

(107) Overpressurization Protection

(108) Reactor Coolant Pressure Boundary

Materials

(109) Reactor Coolant System Pressure

Boundary Inservice Inspection and

Testing

(110) Reactor Coolant Pressure Boundary

Leakage Detection

(111) Reactor Vessel and Internals

Modeling

Resolved (SER)

(SSER 3)

(SSER 5)

(SSER 16)

(SSER 23)

Resolved (SER)

(SSER 23)

Resolved (SER)

(SSER 23)

Open (NRR) (SER)

(SSER 10)

(SSER 23)

Open (NRR) (SER)

(SSER 23)

Resolved (SER)

Resolved (SER)

Resolved (SER)

(SSER 23)

Resolved (SER)

(SSER 5)

(SSER 6)

Resolved (SER)

(SSER 22)

Resolved (SER)

(SSER 2)

(SSER 15)

(SSER 24)

Resolved (SER)

(SSER 22)

Open (NRR) (SER)

(SSER 10)

(SSER 12)

(SSER 15)

(SSER 16)

(SSER 23)

Resolved (SER)

(SSER 9)

(SSER 11)

(SSER 12)

(SSER 22)

4.4.5

4.4.6

4.4.7

4.4.8

4.4.9

4.5

4.5.1

4.5.2

4.6

5

1

5.1 2

5.2

5.2.1

5.2.2

5.2.3

5.2.4

5.2.5

5.2.6

1-10
Issue Status Section Note

(112)

(113)

Reactor Vessel

Reactor Vessel Materials

(114) Pressure-Temperature Limits

(115) Reactor Vessel Integrity

(116)

(117)

Component and Subsystem Design

Reactor Coolant Pumps

(118) Steam Generators

(119) Residual Heat Removal System

(120) Pressurizer Relief Tank

(121) Reactor Coolant System Vents

(TMI Action Item ll.B.1)

(122) Engineered Safety Features

(123) Engineered Safety Feature

Materials

(124) Metallic Materials

(125) Organic Materials

(126) Postaccident Emergency Cooling

Water Chemistry

(127) Containment Systems

Resolved (SER)

(SSER 11)

(SSER 14)

(SSER 22)

(SSER 25)

Resolved (SER)

(SSER 16)

(SSER 22)

(SSER 25)

Open (NRR) (SER)

(SSER 22)

Resolved (SER)

(SSER 22)

Resolved (SER)

(SSER 1)

(SSER 4)

(SSER 22)

Resolved (SER)

(SSER 2)

(SSER 5)

(SSER 10)

(SSER 11)

(SSER 23)

Resolved (SER)

(SSER 22)

Open (SER)

(Inspection) (SSER 2)

(SSER 5)

(SSER 12)

(SSER 23)

Open (SER)

(SSER 23)

Resolved (SER)

(SSER 22)

Resolved (SER)

(SSER 22)

5.3

5.3.1

5.3.2

5.3.3

5.4

5.4.1

5.4.2

5.4.3

5.4.4

5.4.5

6

6.1

6.1.1

6.1.2

6.1.3

6.2

2

1-11
Issue Status Section Note

(128) Containment Functional Design

(129) Containment Heat Removal

Systems

(130) Secondary Containment Functional

Design

(131) Containment Isolation Systems

(132) Combustible Gas Control Systems

(133) Containment Leakage Testing

(134) Fracture Prevention of Containment

Pressure Boundary

(135) Emergency Core Cooling System

(136) System Design

Resolved (SER)

(SSER 3)

(SSER 5)

(SSER 7)

(SSER 12)

(SSER 14)

(SSER 15)

(SSER 22)

Resolved (SER)

(SSER 7)

(SSER 22)

Resolved (SER)

(SSER 18)

(SSER 22)

Resolved (SER)

(SSER 3)

(SSER 5)

(SSER 7)

(SSER 12)

(SSER 22)

Resolved (SER)

(SSER 4)

(SSER 5)

(SSER 8)

(SSER 22)

Open (NRR) (SER)

(SSER 4)

(SSER 5)

(SSER 19)

(SSER 22)

Resolved (SER)

(SSER 4)

(SSER 23)

Resolved (SER)

Open (NRR) (SER)

(SSER 6)

(SSER 7)

(SSER 11)

Resolved (SER)

(SSER 5)

Open (NRR) (SER)

(SSER 2)

(SSER 9)

Resolved (SER)

Open (NRR) (SER)

6.2.1

6.2.2

6.2.3

6.2.4

6.2.5

6.2.6

6.2.7

6.3

6.3.1

6.3.2

6.3.3

6.3.4

6.3.5

1

1

(137) Evaluation 1

(138) Testing

(139)

(140)

Performance Evaluation

Conclusions

1-12
Issue Status Section Note

(141) Control Room Habitability

(142) Engineered Safety Feature (ESF)

Filter Systems

(143) ESF Atmosphere Cleanup System

Resolved (SER)

(SSER 5)

(SSER 11)

(SSER 16)

(SSER 18

(SSER 22)

Resolved (SER)

(SSER 5)

(SSER 22)

Resolved (SER)

Open (NRR) (SER)

(SSER 22)

Resolved (SER)

(144)

(145)

Fission Product Cleanup System

Fission Product Control System

6.4

6.5

6.5.1

6.5.2

6.5.3

6.5.4

6.6

7

7.1

7.1.1

1

(146) Ice Condenser as a Fission Product I

Cleanup System

(147) Inservice Inspection of Class 2 and

3 Components

(148) Instrumentation and Controls

(149) Introduction

(150) General

(151) Comparison with Other Plants

(152) Design Criteria

Open (NRR) (SER)

(SSER 10)

(SSER 12)

(SSER 15)

(SSER 23)

(153)

(154)

Reactor Trip System

System Description

Resolved (SER)

(SSER 13)

(SSER 16)

(SSER 23)

Resolved (SER)

(SSER 23)

Resolved (SER)

(SSER 4)

(SSER 15)

(SSER 23)

Resolved (SER)

Open (NRR) (SER)

(SSER 13)

(SSER 15)

(SSER 23)

Resolved (SER)

(SSER 23)

Resolved (SER)

(SSER 23)

Resolved (SER)

(SSER 23)

Resolved (SER)

(SSER 2)

(SSER 14)

(SSER 23)

7.1.2

7.1.3

7.2

7.2.1

1

(155) Manual Trip Switches

(156) Testing of Reactor Trip Breaker

Shunt Coils

(157) Anticipatory Trips

(158) Steam Generator Water Level Trip

7.2.2

7.2.3

7.2.4

7.2.5

1

1

1-13
Issue

(159) Conclusions

(160) Engineered Safety Features System

(161) System Description

(162) Containment Sump Level

Measurement

(163) Auxiliary Feedwater Initiation and

Control

(164) Failure Modes and Effects Analysis

(165) IE Bulletin 80-06

(166) Conclusions

(167) Systems Required for Safe

Shutdown

(168) System Description

(169) Safe Shutdown from Auxiliary

Control Room

(170) Conclusions

(171) Safety-Related Display

Instrumentation

(172) Display Systems

(173) Postaccident Monitoring System

(174) IE Bulletin 79-27

(175) Conclusions

(176) All Other Systems Required for

Safety

Status

Resolved (SER)

(SSER 13)

(SSER 23)

Open (NRR) (SER)

(SSER 13)

Resolved (SER)

(SSER 13)

(SSER 14)

(SSER 23)

Resolved (SER)

(SSER 2)

(S.SER 23)

Resolved (SER)

(SSER 23)

Resolved (SER)

(SSER 23)

Resolved (SER)

(SSER 3)

(SSER 23)

Resolved (SER)

(SSER 13)

(SSER 23)

Resolved (SER)

(SSER 23)

Resolved (SER)

(SSER 7)

(SSER 23)

Resolved (SER)

(SSER 23)

Section

7.2.6

Note

7.3

7.3.1

7.3.2

7.3.3

7.3.4

7.3.5

7.3.6

7.4

7.4.1

7.4.2

7.4.3

7.5

7.5.1

7.5.2

7.5.3

7.5.4

7.6

1

Resolved

Open

(Inspection)

Open

(Inspection)

Open

(Inspection)

(SER)

(SSER 23)

(SER)

(SSER 7)

(SSER 9)

(SSER 14)

(SSER 15)

(SSER 23)

(SSER 25)

(SER)

(SSER 23)

(SER)

1-14
Issue Status Section Note

(177) Loose Part Monitoring System

(178) Residual Heat Removal System

Bypass Valves

(179) Upper Head Injection Manual

Control

(180) Protection Against Spurious

Actuation of Motor-Operated Valves

(181) Overpressure Protection during Low

Temperature Operation

(182) Valve Power Lockout

(183) Cold Leg Accumulator Valve

Interlocks and Position Indication

(184) Automatic Switchover From

Injection to Recirculation Mode

(185) Conclusions

(186) Control Systems Not Required for

Safety

(187) System Description

(188) Safety System Status Monitoring

System

(189) Volume Control Tank Level Control

System

(190) Pressurizer and Steam Generator

Overfill

(191) IE Information Notice 79-22

(192) Multiple Control System Failures

(193) Conclusions

(194) Anticipated Transient Without

Scram Mitigation System Actuation

Circuitry (AMSAC)

(195) NUREG-0737 Items

(196) Relief and Safety Valve Position

Indication (TMI Action Item ll.D.3)

Resolved (SER)

(SSER 23)

(SSER 24)

Resolved (SER)

(SSER 23)

Resolved (SER)

(SSER 23)

Resolved (SER)

(SSER 23)

Resolved (SER)

(SSER 4)

(SSER 23)

Resolved (SER)

(SSER 23)

Resolved (SER)

(SSER 23)

Resolved (SER)

(SSER 23)

Resolved (SER)

(SSER 4)

Open (NRR) (SER)

(SSER 23)

(SSER 24)

(SSER 25)

Resolved (SER)

(SSER 7)

(SSER 13)

(SSER 23)

Resolved (SER)

(SSER 23)

Resolved (SER)

(SSER 23)

Resolved (SER)

(SSER 23)

Resolved (SER)

(SSER 23)

Resolved (SER)

Resolved (SSER 9)

(SSER 14)

(SSER 23)

Resolved (SER)

(SSER 23)

Open (SER)

(Inspection) (SSER 5)

(SSER 14)

(SSER 23)

7.6.1

7.6.2

7.6.3

7.6.4

7.6.5

7.6.6

7.6.7

7.6.8

7.6.9

7.7

7.7.1

7.7.2

7.7.3

7.7.4

7.7.5

7.7.6

7.7.7

7.7.8

7.8

7.8.1

1-15
Iss u._ee Status Section Note

(197) Auxiliary Feedwater System

Initiation and Flow Indication (TMI

Action Item II.E.1.2)

(198) Proportional Integral Derivative

Control Modification (TMI Action

Item I1.K.3.9)

(199) Proposed Anticipatory Trip

Modification (TMI Action Item

I1.K.3.10)

(200) Confirm Existence of Anticipatory

Reactor Trip Upon Turbine Trip (TMI

Action Item II.K.3.12)

(201) Data Communication Systems

(202) Electric Power Systems

(203) General

(204) Offsite Power System

(205) Compliance with GDC 5

(206) Compliance with GDC 17

(207) Compliance with GDC 18

(208) Evaluation Findings

(209) Onsite Power Systems

(210) Onsite AC Power System

Compliance with GDC 17

Open

(Inspection)

Open

(Inspection)

(SER)

(SSER 23)

(SER)

(SSER 23)

Resolved (SER)

(SSER 4)

(SSER 23)

Resolved (SER)

(SSER 23)

(SSER 23)

Open (NRR) (SER)

(SSER 22)

(SSER 24)

(SER)

(SSER 22)

Open (NRR) (SER)

(SSER 13)

(SSER 22)

Open (NRR) (SER)

(SSER 2)

(SSER 3)

(SSER 13)

(SSER 14)

(SSER 15

(SSER 22)

Resolved (SER)

(SSER 22)

Open (NRR) (SER)

(SSER 22)

Resolved (SER)

(SSER 10)

(SSER 19)

(SSER 22)

Open (NRR) (SER)

(SSER 2)

(SSER 7)

(SSER 9)

(SSER 10)

(SSER 13)

(SSER 14)

(SSER 18)

(SSER 20)

(SSER 22)

7.8.2

7.8.3

7.8.4

7.8.5

7.9

8

8.1

8.2

8.2.1

8.2.2

8.2.3

8.2.4

8.3

8.3.1

1-16
Issue Status Section Note

(211) Onsite DC System Compliance with

GDC 17

(212) Common Electrical Features and

Requirements

(213) Evaluation Findings

Open (NRR) (SER)

(SSER 2)

(SSER 3)

(SSER 13)

(SSER 14)

(SSER 22)

Resolved (SER)

(SSER 2)

(SSER 3)

(SSER 7)

(SSER 13)

(SSER 14)

(SSER 15)

(SSER 16)

(SSER 22)

Open (NRR) (SER)

(SSER 2)

(SSER 3)

(SSER 7)

(SSER 13)

(SSER 14)

(SSER 15)

(SSER 16)

(SSER 22)

Open (NRR) (SSER 22)

Resolved (SER)

(SSER 10)

Resolved (SER)

Resolved (SER)

(SSER 5)

(SSER 15)

(SSER 16)

(SSER 22)

Open (NRR) (SER)

(SSER 11)

(SSER 15)

(SSER 23)

Resolved (SER)

(SSER 3)

(SSER 13)

(SSER 22)

(SSER 24)

8.3.2

8.3.3

8.3.4

8.4

9

9.1

9.1.1

9.1.2

(214)

(215)

(216)

(217)

(218)

Station Blackout

Auxiliary Systems

Fuel Storage Facility

New-Fuel Storage

Spent-Fuel Storage

1

(219) Spent Fuel Pool Cooling and

Cleanup System

(220) Fuel-Handling System

(221) Water Systems

(222) Essential Raw Cooling Water and

Raw Cooling Water System

9.1.3

9.1.4

9.2

Open (NRR) (SER) 9.2.1

(SSER 9)

(SSER 10)

(SSER 18)

(SSER 23)

1-17
Issue Status Section Note

(223) Component Cooling System

(Reactor Auxiliaries Cooling Water

System)

(224) Demineralized Water Makeup

System

(225) Potable and Sanitary Water

Systems

(226) Ultimate Heat Sink

(227) Condensate Storage Facilities

(228)

(229)

Process Auxiliaries

Compressed Air System

(230) Process Sampling System

(231) Equipment and Floor Drainage

System

(232) Chemical and Volume Control

System

(233) Heat Tracing

(234) Heating, Ventilation, and Air

Conditioning Systems

(235) Control Room Area Ventilation

System

(236) Fuel-Handling Area Ventilation

System

(237) Auxiliary Building and Radwaste

Area Ventilation System

(238) Turbine Building Area Ventilation

System

(239) Engineered Safety Features

Ventilation System

(240) Reactor Building Purge Ventilation

System

(241) Containment Air Cooling System

Open (NRR) (SER)

(SSER 5)

(SSER 23)

Resolved (SER)

(SSER 22)

Resolved (SER)

(SSER 9)

(SSER 22)

Open (NRR) (SER)

(SSER 23)

Resolved (SER)

(SSER 12)

(SSER 22)

Resolved (SER)

(SSER 22)

Resolved (SER)

(SSER 3)

(SSER 5)

(SSER 14)

(SSER 16)

(SSER 24)

Resolved (SER)

(SSER 22)

Resolved (SER)

(SSER 22)

(SSER 22)

Resolved (SER)

(SSER 9)

(SSER 22)

Resolved (SER)

(SSER 22)

Resolved (SER)

(SSER 22)

Resolved (SER)

(SSER 22)

Resolved (SER)

(SSER 9)

(SSER 10)

(SSER 11)

(SSER 14)

(SSER 16)

(SSER 19)

(SSER 22)

(SSER 22)

9.2.2

9.2.3

9.2.4

9.2.5

9.2.6

9.3

9.3.1

9.3.2

1

9.3.3

9.3.4

9.3.8

9.4

9.4.1

9.4.2

9.4.3

9.4.4

9.4.5

9.4.6

9.4.7

3

3

(SSER 22)

1-18
Issue Status Section Note

(242) Condensate Demineralizer Waste

Evaporator Building Environmental

Control System

(243) Other Auxiliary Systems

(244) Fire Protection

(245) Communications System

(246) Lighting System

(247) Emergency Diesel Engine Fuel Oil

Storage and Transfer System

(248) Emergency Diesel Engine Cooling

Water System

(249) Emergency Diesel Engine Starting

Systems

(250) Emergency Diesel Engine

Lubricating Oil System

(251) Emergency Diesel Engine

Combustion Air Intake and Exhaust

System

(252) Steam and Power Conversion

System

(253) Summary Description

(254) Turbine Generator

(255) Turbine Generator Design

(256) Turbine Disc Integrity

(257) Main Steam Supply System

(258) Main Steam Supply System (Up to

and Including the Main Steam

Isolation Valves)

(SSER 22)

Resolved (SER)

(SSER 10)

(SSER 18)

(SSER 19)

Resolved (SER)

(SSER 5)

Resolved (SER)

(SSER 22)

Resolved (SER)

(SSER 5)

(SSER 9)

(SSER 10)

(SSER 11)

(SSER 12)

(SSER 22)

Resolved (SER)

(SSER 5)

(SSER 11)

Resolved (SER)

(SSER 5)

(SSER 10)

(SSER 22)

Resolved (SER)

(SSER 3)

(SSER 5)

(SSER 10)

(SSER 22)

Resolved (SER)

(SSER 5)

(SSER 10)

(SSER 22)

Resolved (SER)

Open (NRR) (SER)

(SSER 5)

Resolved (SER)

(SSER 12)

(SSER 22)

Resolved (SER)

(SSER 23)

Resolved (SER)

Resolved (SER)

(SSER 19)

(SSER 22)

9.4.8

9.5

9.5.1

9.5.2

9.5.3

9.5.4

9.5.5

9.5.6

9.5.7

9.5.8

1

10

10.1

10.2

10.2.1

10.2.2

10.3

10.3.1

1-19
Issue Status Section Note

(259) Main Steam Supply System

(260) Steam and Feedwater System

Materials

(261) Secondary Water Chemistry

(262)

(263)

Other Features

Main Condenser

(264) Main Condenser Evacuation System

(265) Turbine Gland Sealing System

(266) Turbine Bypass System

(267) Condenser Circulating Water

System

(268) Condensate Cleanup System

(269) Condensate and Feedwater

Systems

(270) Steam Generator Blowdown System

(271) Auxiliary Feedwater System

(272) Heater Drains and Vents

(273) Steam Generator Wet Layup

Resolved (SER)

(SSER 22)

Resolved (SER)

(SSER 22)

Resolved (SER)

(SSER 5)

(SSER 22)

Resolved (SER)

(SSER 9)

(SSER 22)

Resolved (SER)

(SSER 22)

Resolved (SER)

(SSER 22)

Resolved (SER)

(SSER 5)

(SSER 22)

Resolved (SER)

(SSER 22)

Open (NRR) (SER)

(SSER 22)

Resolved (SER)

(SSER 14)

(SSER 22)

Resolved (SER)

(SSER 22)

(SSER 24)

Resolved (SER)

(SSER 14)

(SSER 23)

(SSER 24)

(SSER 22)

(SSER 22)

Resolved (SER)

(SSER 16)

(SSER 24)

Resolved (SER)

(SSER 4)

(SSER 16)

(SSER 24)

Resolved (SER)

(SSER 8)

(SSER 16)

(SSER 24)

(SSER 25)

10.3.2 2

10.3.3

10.3.4

10.4

10.4.1

10.4.2

10.4.3

10.4.4

10.4.5

10.4.6

10.4.7

10.4.8

10.4.9

10.4.10

10.4.11

(274)

(275)

System

Radioactive Waste Management

Summary Description

(276) Liquid Waste Management

(277) Gaseous Waste Management

1111.1

11.2

11.3

2

1-20
Issue Status Section Note

(278) Solid Waste Management System

(279) Process and Effluent Radiological

Monitoring and Sampling Systems

(280) Evaluation Findings

(281) NUREG-0737 Items

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

Particulate Effluent Monitors (TMI

Action Items I1.F.1(1) and I1.F.1(2))

(283) Primary Coolant Outside

Containment (TMI Action item

II1.D.1.1)

(284) Radiation Protection

(285) General

(286) Ensuring that Occupational

Radiation Doses Are As Low As

Reasonably Achievable (ALARA)

(287) Radiation Sources

(288) Radiation Protection Design

Features

Resolved (SER)

(SSER 16)

(SSER 24)

Resolved (SER)

(SSER 16)

(SSER 20)

(SSER 24)

Resolved (SER)

(SSER 8)

(SSER 16)

Open (NRR) (SER)

Open (SER)

(Inspection) (SSER 5)

(SSER 6)

Open (NRR) (SER)

(SSER 5)

(SSER 6)

(SSER 10)

(SSER 16)

Resolved (SER)

(SSER 10)

(SSER 14)

(SSER 24)

Resolved (SER)

(SSER 14)

(SSER 24)

Resolved (SER)

(SSER 14)

(SSER 24)

Open (NRR) (SER)

(SSER 10)

(SSER 14)

(SSER 18)

(SSER 24)

Open (NRR) (SER)

(SSER 14)

(SSER 24)

Open (NRR) (SER)

(SSER 10)

(SSER 14)

(SSER 24)

11.4

11.5

11.6

11.7

11.7.1

11.7.2

12

12.1

12.2 2

(289) Dose Assessment

(290) Health Physics Program

12.3

12.4

12.5

12.6

12.7

12.7.1

12.7.2

(291)

(292)

NUREG-0737 Items

Plant Shielding

(TMI Action Item Il.B.2)

Open (NRR)

Open (NRR)

(SER)

(SSER 14)

(SSER 16)

(SSER 24)

(SER)

(SSER 5)

(293) High Range In-Containment Monitor

(TMI Action Item II.F.1.(3))

1-21
Issue Status Section Note

(294) In-Plant Radioiodine Monitor (TMI

Action Item II.D.3.3)

(295) Conduct of Operations

(296) Organization Structure of the

Applicant

(297) Management and Technical

Organization

(298) Corporate Organization and

Technical Support

(299) Plant Staff Organization

Open (NRR) (SER)

(SSER 16)

(300)

(301)

Training

Licensed Operator Training

Program

(302) Training for Non-licensed Personnel

(303) Emergency Preparedness

Evaluation

(304) Introduction

(305) Evaluation of the Emergency Plan

(306) Conclusions

Resolved (SER)

(SSER 16)

(SSER 22)

Resolved (SER)

Resolved (SER)

Resolved (SER)

(SSER 8)

(SSER 22)

(SSER 25)

Resolved (SER)

(SSER 9)

(SSER 10)

(SSER 22)

Resolved (SER)

Open (NRR) (SER)

(SSER 13)

(SSER 20)

Open (NRR) (SER)

(SSER 13)

(SSER 20)

(SSER 22)

Open (NRR) (SER)

(SSER 13)

(SSER 20)

(SSER 22)

Resolved (SER)

(SSER 8)

(SSER 22)

Resolved (SER)

(SSER 22)

Resolved (SER)

(SSER 22)

Resolved (SER)

(SSER 9)

(SSER 10)

(SSER 22)

Resolved (SER)

(SSER 3)

(SSER 16)

(SSER 22)

12.7.3

13

13.1

13.1.1

13.1.2

13.1.3

13.2

13.2.1

13.2.2

13.3

13.3.1

13.3.2

13.3.3

13.4

13.5

13.5.1

13.5.2

13.5.3

(307) Review and Audit

(308) Plant Procedures

(309) Administrative Procedures

(310) Operating and Maintenance

Procedures

(311) NUREG-0737 Items

1-22
Issue Status Section Note

(312) Physical Security Plan

(313)

(314)

(315)

(316)

(317)

(317a)

(318)

Introduction

Summary of Application

Regulatory Basis

Technical Evaluation

Conclusions

Cyber Security Plan

Initial Test Program

Resolved (SER)

(SSER 1)

(SSER 10)

(SSER 15)

(SSER 20)

(SSER 22)

(SSER 22)

(SSER 22)

(SSER 22)

(SSER 22)

(SSER 22)

Resolved (SSER 24)

Resolved (SER)

(SSER 3)

(SSER 5)

(SSER 7)

(SSER 9)

(SSER 10)

(SSER 12)

(SSER 14)

(SSER 16)

(SSER 18)

(SSER 19)

(SSER 23)

13.6

13.6.1

13.6.2

13.6.3

13.6.4

13.6.5

13.6.6

14

15

15.1

15.2

15.2.1

15.2.2

15.2.3

15.2.4

15.2.5

15.3

(319) Accident Analyses

(320) General Discussion

(321) Normal Operation and Anticipated

Transients

(322) Loss-of-Cooling Transients

(323) Increased Cooling Inventory

Transients

(324) Change in Inventory Transients

(325) Reactivity and Power Distribution

Anomalies

(326) Conclusions

(327) Limiting Accidents

Resolved

Open (NRR)

(SER)

(SER)

Resolved (SER)

(SSER 13)

(SSER 14)

(SSER 24)

Resolved (SER)

(SSER 24)

Resolved (SER)

(SSER 18)

(SSER 24)

Open (NRR) (SER)

(SSER 4)

(SSER 7)

(SSER 13)

(SSER 14)

(SSER 24)

Resolved (SER)

(SSER 4)

Resolved (SER)

1-23
Issue Status Section Note

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

(329) Steamline Break

(330) Feedwater System Pipe Break

(331) Reactor Coolant Pump Rotor

Seizure

(332) Reactor Coolant Pump Shaft Break

(333) Anticipated Transients Without

Scram

Resolved (SER)

(SSER 12)

(SSER 15)

(SSER 24)

Resolved (SER)

(SSER 3)

(SSER 14)

(SSER 24)

Resolved (SER)

(SSER 14)

(SSER 24)

Resolved (SER)

(SSER 14)

(SSER 24)

Resolved (SER)

(SSER 14)

(SSER 24)

Resolved (SER)

(SSER 3)

(SSER 5)

(SSER 6)

(SSER 10)

(SSER 11)

(SSER 12)

(SSER 24)

Resolved (SER)

Resolved (SER)

(SSER 15)

(SSER 25)

Resolved (SER)

(SSER 5)

(SSER 9)

(SSER 18)

(SSER 25)

Open (NRR) (SER)

(SSER 15)

(SSER 25)

Resolved (SER)

(SSER 2)

(SSER 5)

(SSER 12)

(SSER 14)

(SSER 15)

(SSER 25)

Resolved (SER)

(SSER 15)

(SSER 25)

15.3.1

15.3.2

15.3.3

15.3.4

15.3.5

15.3.6

15.3.7

15.4

15.4.1

15.4.2

15.4.3

15.4.4

(334)

(335)

Conclusions

Radiological Consequences of

Accidents

(336) Loss-of-Coolant Accident

(337) Main Steamline Break Outside of

Containment

(338) Steam Generator Tube Rupture

(339) Control Rod Ejection Accident

1-24
Issue Status Section Note

(340) Fuel-Handling Accident

(341) Failure of Small Line Carrying

Coolant Outside Containment

(342) Postulated Radioactive Releases as

a Result of Liquid Tank Failures

(342a) Postulated Waste Gas Decay Tank

Rupture

(343) NUREG-0737 Items

(344) Thermal Mechanical Report (TMI

Action Item II.K.2.13)

(345) Voiding in the Reactor Coolant

System during Transients (TMI

Action Item II.K.2.17)

(346) Installation and Testing of Automatic

Power-Operated Relief Valve

Isolation System (TMI Action Item

I1.K.3.1) Report on Overall Safety

Effect of Power-Operated Relief

Valve Isolation System (TMI Action

Item I1.K.3.2)

(347) Automatic Trip of Reactor Coolant

Pumps (TMI Action Item I1.K.3.5)

(348) Small-Break LOCA Methods

(11.K.3.30) and Plant-Specific

Calculations (11.K.3.31)

(349) Relative Risk of Low-Power

Operation

(350) Technical Specification

(351) Quality Assurance

(352) General

(353) Organization

(354) Quality Assurance Program

Resolved (SER)

(SSER 4)

(SSER 15)

(SSER 25)

Resolved (SER)

(SSER 25)

Resolved (SER)

(SSER 25)

Resolved (SSER 25)

Resolved (SER)

(SSER 4)

(SSER 24)

Resolved (SER)

(SSER 4)

(SSER 24)

Resolved (SER)

(SSER 5)

15.4.5

15.4.6

15.4.7

15.4.8

15.5

15.5.1

15.5.2

15.5.3

15.5.4

15.5.5

15.6

16

17

17.1

17.2

17.3

17.4

17.6

18

Resolved (SER)

(SSER 4)

(SSER 16)

(SSER 24)

Open (SER)

(Inspection) (SSER 4)

(SSER 5)

(SSER 16)

Resolved (SER)

Open (NRR)

Resolved (SER)

Resolved (SER)

Resolved (SER)

(SSER 2)

(SSER 5)

(SSER 10)

(SSER 13)

(SSER 15)

(SSER 22)

(355) Resolved (SER)

(356)

(357)

Conclusions

Maintenance Rule

Control Room Design Review

1-25
Issue Status Section Note

(358) General

(359) Conclusions

(360) Report of the Advisory Committee

on Reactor Safeguards

(361) Common Defense and Security

(362) Financial Qualifications

(363) TVA Financial Qualifications for

WBN Unit 2

(364) Foreign Ownership, Control, or

Domination

(365) Financial Protection and Indemnity

Requirements

(366) General

(367) Preoperational Storage of Nuclear

Fuel

(368) Operating Licenses

(369) Quality of Construction, Operational

Readiness, and Quality Assurance

Effectiveness

(370) Program for Maintenance and

Preservation of the Licensing Basis

for Units 1 and 2

Notes:

Resolved (SER)

(SSER 5)

(SSER 6)

(SSER 15)

(SSER 16)

(SSER 22)

Resolved (SER)

(SSER 16)

(SSER 22)

(SER)

(SER)

(SER)

(SSER 22)

(SSER 23)

(SSER 22)

(SER)

(SER)

18.1

18.2

19

20

21

21.1

21.2

22

22.1

22.2

22.3

25

25.9

Open (NRR)

Open (NRR)

(SSER 22)

(SSER 22)

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 their submittals and clearly indicated them to the staff. The staff has reviewed

and confirmed that the changes made are administrative/typographical and do

not impact the staffs 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 staff characterized certain topics as "Open" pending TVA's validation of the

information contained in the section. WVA 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 their

submittals and clearly indicated the changes. The 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 staff considers this section Resolved.

1-26
3. In SSER 21, this issue was identified as "Resolved." However, TVA made

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

evaluated the changes and the results are documented in this SSER.

1.8 Confirmatory Issues

At this point in the review, there are some items that have essentially been resolved to the

staffs satisfaction, but for which certain confirmatory information has not yet been provided by

the applicant. In these instances, the applicant has committed to provide the confirmatory

information in the near future. If staff review of this information does not confirm preliminary

conclusions on an item, that item will be treated as open, and the NRC staff will report on its

resolution in a supplement to this report.

The confirmatory items, with appropriate references to subsections of this report, are noted in

Appendix HH.

1.9 License Conditions

The NRC staff proposes two license conditions discussed in Section 2.4.10 of SSER 24.

Flooding Protection Proposed License Condition No. 1:

WVA will submit to the NRC staff by August 31, 2012, for review and approval, a

summary of the results of the finite element analysis, which demonstrates that the

Cherokee and Douglas dams are fully stable under design basis probable maximum

flood loading conditions for the long-term stability analysis, including how the

preestablished acceptance criteria were met.

Flooding Protection Proposed License Condition No. 2:

WVA will submit to the NRC staff, before completion of the first operating cycle, its longterm modification plan to raise the height of the embankments associated with the

Cherokee, Fort Loudoun, Tellico, and Watts Bar dams. The submittal shall include

analyses to demonstrate that, when the modifications are complete, the embankments

will meet the applicable structural loading conditions, stability requirements, and

functionality considerations to ensure that the design basis probable maximum flood

limits are not exceeded at the Watts Bar Nuclear Plant. All modifications to raise the

height of the embankments shall be completed within 3 years from the date of issuance

of the operating license.

The NRC staff proposes two license conditions discussed in Section 13.6.6.3.22 of SSER 24.

Cyber Security Proposed License Condition 1:

The licensee shall implement the requirements of 10 CFR 73.54(a)(1)(ii) as they relate to

the security computer. Completion of these actions will occur consistent with the full

implementation date of September 30, 2014, as established in the licensee's letter dated

April 7, 2011, "Response to Request for Additional Information Regarding Watts Bar

Nuclear Plant Cyber Security Plan License Amendment Request, Cyber Security Plan

Implementation Schedule - Watts Bar Nuclear Plant Unit 1."

1-27
Cyber Security Proposed License Condition 2:

The licensee shall implement the requirements of 10 CFR 73.54(a)(1)(iii) as they relate

to the corporate based systems that support emergency preparedness. Completion of

these actions will occur consistent with the Watts Bar Nuclear Plant Unit 1

implementation schedule established in the licensee's letter dated April 7, 2011,

"Response to Request for Additional Information Regarding Watts Bar Nuclear Plant

Cyber Security Plan License Amendment Request, Cyber Security Plan Implementation

Schedule - Watts Bar Nuclear Plant Unit V."

1.10 Unresolved Safety Issues

Section 210 of the Energy Reorganization Act of 1974, as amended, states, in part,

The Commission shall develop a plan for providing for specification and analysis

of unresolved safety issues relating to nuclear reactors and shall take such action

as may be necessary to implement corrective measures with respect to such

issues.

The NRC staff continuously evaluates the safety requirements used in its review against new

information as it becomes available. In some cases, the staff takes immediate action or interim

measures to ensure safety. In most cases, however, the initial assessment indicates that

immediate licensing actions or changes in licensing criteria are not necessary. In any event,

further study may be deemed appropriate to make judgments as to whether existing

requirements should be modified. The issues being studied are sometimes called generic

safety issues because they are related to a particular class or type of nuclear facility.

The NRC staff documented its original review of Unresolved Safety Issues for WBN Units 1 and

2 in Appendix C to the safety evaluation report (SER; NUREG-0847, June 1982). A discussion

of the status of resolution of these generic issues for TVA's application for an operating license

for WBN Unit 2 is provided in Appendix C to SSER 23, dated July 2011.

1.13 Implementation of Corrective Action Programs and Special Programs

In 1985, TVA developed a corporate Nuclear Performance Plan (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 staff review of the CAPs and SPs. This SSER

and future supplements to the SER, the staff will document its evaluation and closure of open

NPP items.

1-28
1.13.1

No.

(1)

(2)

(3)

(4)

(5)

(6)

(7)

(8)

(9)

(10)

(11)

(12)

(13)

(14)

Corrective Action Programs

Title

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

Cable Tray and Tray Supports

Design Baseline and Verification Program

Electrical Conduit and Conduit Support

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

Equipment Seismic Qualification

Fire protection

Hanger and Analysis Update Program

Heat Code Traceability

Heating, Ventilation, and Air-Conditioning Duct and

Duct Supports

Instrument Lines

Prestart Test Program Plan

Quality Assurance (QA) Records

Quality-List (Q-List)

Program Review Status

Resolved

(See Appendix HH)

Resolved

Resolved

Resolved

Resolved

(See Appendix HH)

Resolved

Resolved

Resolved

Resolved

Resolved

Resolved

Resolved

Resolved

Resolved

1-29.
No.

(15)

(16)

(17)

(18)

Title Program Review Status

1.13.2

No.

(1)

(2)

(3)

(4)

(5)

(6)

(7)

(8)

(9)

(11)

Replacement Items Program (Piece Parts)

Seismic Analysis

Vendor Information Program

Welding

Special Programs

Title

Concrete Quality Program

Containment Cooling

Detailed Control Room Design Review

Environmental Qualifications Program

Master Fuse List

Mechanical Equipment Qualification

Microbiologically Induced Corrosion

Moderate Energy Line Break Flooding

Radiation Monitoring System

Use-As-Is Condition Adverse to Quality

Resolved

Resolved

Resolved

Resolved

Pro-gram Review Status

Resolved

Resolved

Resolved

Resolved

Resolved

Resolved

Resolved

Resolved

Resolved

Resolved

1-30
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."

The table below outlines the status of the resolution of the generic communications identified in

SSER 21. It should be noted that, although many of the generic communications have been

documented or otherwise resolved, the NRC staff has determined that there may be

circumstances that could result in the need to reopen a previously closed topic.

Correspondence No. Title

(1) GL 1980-14 Light-Water Reactor Primary Coolant System Pressure

Isolation Valves.

TVA Action: Submit Technical Specifications (TSs) for NRC Review.

NRC Action: To be reviewed during validation of TS 3.4.14 submitted

February 2, 2010.

(2) GL 1980-77 Refueling Water Level - Technical Specifications Changes.

TVA Action: Submit Technical Specifications for NRC Review.

NRC Action: To be reviewed during validation of TS 3.9.5 -TS 3.9.7

submitted February 2, 2010.

(3) GL 1982-28 Inadequate Core Cooling Instrumentation System.

TVA Action: Closed.

NRC Action: 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 Post-Maintenance Testing (reactor trip system components).

3.1)

Submit Technical Specifications for NRC Review.

TVA Action:

To be reviewed during validation of TS Bases 3.0.1 submitted

NRC Action: March 4, 2009.

1-31
Correspondence No.

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

TVA Action

NRC Action

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

TVA Action

NRC Action

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

TVA Action

NRC Action

Title

Post-Maintenance Testing (All Surveillance Requirement

Components).

Submit Technical Specifications and NRC Review.

To be reviewed during validation of TS Bases 3.0.1 submitted

March 4, 2009.

Reactor Trip System Reliability (Preventive Maintenance and

Surveillance Program for Reactor Trip Breakers).

Submit Technical Specifications and NRC Review.

To be reviewed during staff evaluation of Item 17 of TS

Table 3.3.1-1 submitted February 2, 2010.

Reactor Trip System Reliability (Automatic Actuation of Shunt

Trip Attachment).

Submit Technical Specifications and NRC Review.

To be reviewed during staff evaluation of Item 18 of TS

Table 3.3.1-1 submitted February 2, 2010.

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

Operation in BWRs and PWRs.

Submit Technical Specifications for NRC Review.

To be reviewed during validation of TS 3.4.4 - TS 3.4.8

submitted February 2, 2010.

Individual Plant Examination for Severe Accident Vulnerability.

Closed.

Closed. NRC letter dated August 12, 2011 (ADAMS

Accession No. ML1 11960228).

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. ML1 11960228).

(8) GL 1986-09

TVA Action

NRC Action

(9) GL 1988-20

TVA Action

NRC Action

(10) GL 1988-20sl

TVA Action

NRC Action

1-32
Correspondence No. Title

(11) GL 1988-20s2

TVA Action

NRC Action

(12) GL 1988-20s3

TVA Action

NRC Action

(13) GL 1988-20s4

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. ML1 11960228).

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).

Individual Plant Examination of External Events (IPEEE) for

Severe Accident Vulnerabilities.

Closed.

Closed. NRC letter dated September 20, 2011 (ADAMS

Accession No. ML1 11960300).

Individual Plant Examination of External Events (IPEEE) for

Severe Accident Vulnerabilities - 10 CFR 50.54(f).

(14) GL 1988-20s5

TVA Action Closed.

NRC Action

(15) GL 1989-04

TVA Action

NRC Action

Closed. NRC letter dated September 20, 2011 (ADAMS

Accession No. ML1 11960300).

Guidelines on Developing Acceptable Inservice Testing

Programs.

The proposed approach has been approved for WBN Unit 1;

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

without change.

Open.

1-33
Correspondence No. Titl_.ee

(16) GL 1989-21

TVA Action

NRC Action

(17) GL 1990-06

TVA Action

NRC Action

(18) GL 1992-08

TVA Action

NRC Action

(19) GL 1995-03

TVA Action

NRC Action

(20) GL 1995-05

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.

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.

To be reviewed during validation of TS 3.4.11 - TS 3.4.12

submitted February 2, 2010.

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.

Open. Pending NRC staff inspection verification.

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).

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).

1-34
Correspondence No. Title

(21) GL 1996-06

TVA Action

NRC Action

(22) GL 1995-07

TVA Action

NRC Action

(23) GL 1997-01

TVA Action

NRC Action

(24) GL 1997-04

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).

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).

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).

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).

1-35
Correspondence No. Title

(25) GL 1997-05

TVA Action

NRC Action

(26) GL 1997-06

TVA Action

NRC Action

(27) GL 1998-02

TVA Action

NRC Action

(28) GL 1998-04

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).

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).

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. ML1012001155).

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).

1-36
Correspondence No. Title

(29) GL 2003-01

TVA Action

NRC Action

(30) GL 2004-01

TVA Action

NRC Action

(31) GL 2004-02

TVA Action

NRC Action

(32) GL 2006-01

TVA Action

NRC Action

(33) GL 2006-02

TVA Action

NRC Action

Control Room Habitability.

No action or documentation is provided to show the 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).

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).

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.

Open.

SG Tube Integrity and Associated Technical Specifications.

No action or documentation is provided to show the 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).

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).

1-37
Correspondence No. Title

(34) GL 2006-03

TVA Action

NRC Action

(35) GL 2007-01

TVA Action

NRC Action

(36) GL 2008-01

TVA Action

NRC Action

(37) BL 1992-01 and

Supplement 1

TVA Action

NRC Action

(38) BL 1996-01

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).

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).

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).

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.

Open. Pending NRC staff inspection verification.

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. ML1 01200035) required Confirmatory Action (See

Appendix HH).

1-38
Correspondence No. Title

(39) BL 1996-02

(40) BL 2001-01

TVA Action

NRC Action

(41) BL 2002-01

TVA Action

NRC Action

(42) BL 2002-02

IVA 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).

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. See NRC Letter dated June 30, 2010 (ADAMS

Accession No. ML 100539515).

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. See NRC Letter dated June 30, 2010 (ADAMS

Accession No. ML 100539515).

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. See NRC Letter dated June 30, 2010 (ADAMS

Accession No. ML100539515).

1-39
Correspondence No. Titl.._ee

(43) BL 2003-02

TVA Action

NRC Action

(44) BL 2004-01

TVA Action

NRC Action

(45) BL 2007-01

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).

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).

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. MLI100770549).

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

to all of the following NUREG-0737 issues):

(46) 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.

Open.

1-40
Correspondence No. Titl_.ee

(47) NUREG-0737

Item I.D.1

TVA Action

NRC Action

(48) NUREG-0737

Item ll.B.3

TVA Action

NRC Action

(49) NUREG-0737

Item II.E.4.2

TVA Action

NRC Action

(50) NUREG-0737

Item II.F.2

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.

Post-accident Sampling.

No action or documentation is provided to show the 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.

Containment Isolation Dependability.

No action or documentation is provided to show the staff has

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

through submittal of a technical specification.

Open.

Instrumentation for Detection of Inadequate Core-Cooling.

TVA Action Open.

NRC Action

(51) NUREG-0737

Item II.K.3.3

TVA Action

NRC Action

Open. See SSER 23, Section 4.4.8.

Reporting SV/RV Failures/Challenges.

No action or documentation is provided to show the 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.

1-41
Correspondence No. Title

(52) NUREG-0737

Item I1.K.3.10

TVA Action

NRC Action

(53) NUREG-0737

Item III.D.1.1

TVA Action

NRC Action

(54) NUREG-0737

Item II1.D.3.4

TVA Action

NRC Action

(55) IEB 75-08

TVA Action

NRC Action

(56) IEB 77-04

Anticipatory Trip at High Power.

No action or documentation is provided to show the staff has

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

through submittal of a technical specification.

Open.

Primary Coolant Outside Containment.

No action or documentation is provided to show the staff has

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

through submittal of a technical specification.

Open.

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.

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

staff review.

Open.

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

staff review.

Open.

TVA Action

NRC Action

1-42
2 SITE CHARACTERISTICS

2.3 Meteorology

2.3.3 Onsite Meteorological Measurements Program

The U.S. Nuclear Regulatory Commission (NRC) guidance provided in NUREG-0800,

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

LWR Edition" (SRP), Section 2.3.3, "Onsite Meteorological Measurements Programs,"

Revision 2, issued July 1981, states, in part, that when the final safety analysis report (FSAR) is

docketed, an applicant should provide at least two consecutive annual cycles, including the

most recent 1-year period, of onsite meteorological measurements. The applicant should

summarize the data to describe the meteorological characteristics of the site and its vicinity in

formats suitable for making atmospheric dispersion estimates for both postulated accidental and

expected routine airborne releases of effluents. In addition, the applicant should provide

evidence of how well the data represent long-term conditions at the site.

To meet these objectives, the Tennessee Valley Authority (TVA) provided the following sets of

meteorological data for the-Watts Bar Nuclear Plant (WBN):

hourly data from 1991 through 2010 formatted to calculate short-term design-basis

accident (DBA) control room (CR) atmospheric dispersion estimates (x/Q values) using

the ARCON96 atmospheric dispersion computer code (NUREG/CR-6331, "Atmospheric

Relative Concentrations in Building Wakes," Revision 1, issued 1996), provided by letter

dated October 17, 2011 (Agencywide Documents Access and Management System

(ADAMS) Accession No. ML1 1294A461)

a joint frequency distribution (JFD) of wind speed and direction by atmospheric stability

class of the data from 1991 through 2010, provided by letter dated November 7, 2011, to

calculate the DBA exclusion area boundary (EAB) and low-population zone (LPZ) x/Q

values

a JFD of data from 1986 through 2005 provided in FSAR Amendment 105, dated

August 12, 2011 (ADAMS Accession No. ML11236A158), to calculate long-term routine

release x/Q values and deposition factors (D/Q values)

These data supplemented the JFD information from 1974 through 1993 in the WBN FSAR,

previously discussed in Supplemental Safety Evaluation Report (SSER) 15, dated

June 15, 1995 (ADAMS Accession No. ML072060488).

The NRC staff reviewed the onsite hourly meteorological data for 1991 through 2010 using the

methodology described in NUREG-0917, "NRC Staff Computer Programs for Use with

Meteorological Data," issued July 1982. Further statistical review used computer spreadsheets.

Data recovery for the 20-year period for all parameters was consistently in the upper

90 percentiles, which exceeds the recommendation of NRC Regulatory Guide (RG) 1.23,

"Onsite Meteorological Programs," issued June 1972, of at least 90-percent data recovery. With

respect to atmospheric stability measurements, stable and neutral conditions were consistently

reported to occur at night, and unstable and neutral conditions occurred during the day. The

frequency, length, and time of occurrence of stable and unstable atmospheric conditions were

congruent with expected meteorological conditions. Wind speed and direction frequency

distributions for each measurement channel were also very consistent from year to year and

2-1
when comparing measurements between the measurement heights. Meteorological data

summarized in the JFD formats were reasonably consistent with the data provided in the hourly

data summary formats. On the basis of this review, the NRC staff concluded that the data files

provided by TVA for 1991 through 2010 give an adequate representation of the site conditions

to facilitate calculation of the CR, EAB, and LPZ x/Q values for DBA and x/Q and D/Q values for

routine release dose assessments for WBN Unit 2.

The JFD 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 updated FSAR, the NRC staff will confirm that these

updates have been made by TVA. This is Open Item 136 (Appendix HH).

2.3.4 Short-Term (Accident) Atmospheric Dispersion Estimates

As discussed in its letter dated July 31, 2010 (ADAMS Accession No. ML1 02290258), WVA

initially based the WBN Unit 2 DBA x/Q values for the CR, EAB, and LPZ on meteorological

data from 1974 through 1993, which were used as the licensing basis for WBN Unit 1. WVA

subsequently compared these data with current meteorological conditions and concluded that it

was appropriate to update the WBN Unit 2 DBA x/Q values for the CR, EAB, and LPZ to

incorporate more recent meteorological data using the 20-year period from 1991 through 2010.

TVA used the ARCON96 methodology to calculate the WBN Unit 2 CR x/Q values using inputs

and assumptions initially discussed in Enclosure 1 and Attachments 11 and 12 to its letter dated

July 31, 2010. TVA stated that it based the dose analysis and associated release scenarios on

NRC guidance documentation and considered worst case single failures. The applicant

assumed loss of offsite power (LOOP) for all accidents, in addition to the worst case single

failure. It modeled all releases as ground-level point sources using the straight-line horizontal

distance as the distance of separation between each postulated source and receptor pair. No

sources were modeled as diffuse or high-energy releases. Following completion of further

analyses, TVA provided a supplemental letter dated September 15, 2011 (ADAMS Accession

No. ML1 1262A276), which summarized the revised x/Q values used in the dose assessments

for the loss-of-coolant, fuel-handling, steam generator tube rupture, main steamline break, and

loss of alternating current power accidents.

RG 1.194, "Atmospheric Relative Concentrations for Control Room Radiological Habitability

Assessments at Nuclear Power Plants," issued June 2003, states that ARCON96 is an

acceptable methodology for assessing CR x/Q values for use in DBA radiological analyses. The

NRC staff evaluated the applicability of the ARCON96 model and concluded that no unusual

siting, building arrangement, release characterization, source-receptor configuration,

meteorological regime, or terrain conditions precluded use of this model in support of the WBN

Unit 2 license application. The NRC staff qualitatively reviewed TVA's inputs to the ARCON96

computer runs and found them adequately consistent with site configuration drawings and staff

practice. The NRC staff noted that WVA used the ARCON96 default constant values for surface

roughness length and averaging sector width presented in NUREG/CR-6331, Revision 1, rather

than the default values listed in RG 1.194. The NRC staff used ARCON96 and the RG 1.194

default values to calculate x/Q values to compare with the x/Q values calculated by TVA. The

staff concluded that the CR x/Q values identified in TVA's letter dated September 15, 2011, are

acceptable for use in the DBA dose assessment associated with the WBN Unit 2 license

application, because the x/Q values calculated by TVA were either similar to, or bounding for,

the ARCON96 and the RG 1.194 default values.

2-2
By letter dated June 27, 2011 (ADAMS Accession No. ML1 1180A280), TVA provided updated

EAB and LPZ x/Q values and information related to development of the updated values that it

generated using the meteorological data from 1991 through 2010. TVA stated that it based the

updated x/Q values on guidance in RG 1.145, "Atmospheric Dispersion Models for Potential

Accident Consequence Assessments at Nuclear Power Plants," and assumed all releases were

ground level with a minimum EAB distance of 1,100 meters and LPZ distance of 4,828 meters,

which were the same bases as for WBN Unit 1.

The NRC staff qualitatively reviewed the inputs and assumptions used by TVA and found them

reasonably consistent with NRC regulatory guidance and staff practice. In addition, the NRC

staff made comparison calculations using the PAVAN computer code (NUREGICR-2858,

"PAVAN: An Atmospheric Dispersion Program for Evaluating Design Basis Accidental

Releases of Radioactive Material from Nuclear Power Stations," issued 1982) and obtained

results similar to the EAB and LPZ x/Q values generated by TVA.

On the basis of this review, the NRC staff concluded that the CR, EAB, and LPZ x/Q values are

acceptable for use in the WBN Unit 2 DBA dose assessments. Section 15.4, "Radiological

Consequences of Accidents," of this SSER discusses the updated CR, EAB, and LPZ x/Q

values with their associated DBA dose assessments.

In its letter dated September 15, 2011, TVA stated that it would revise FSAR Section 15.5 in a

future amendment to reflect the x/Q values presented in that letter. TVA reflected the EAB and

LPZ x/Q values from its letter dated June 27, 2011, in FSAR Amendment 105, dated

August 12, 2011 (ADAMS Accession No. ML1I1236A1 58). The NRC staff will confirm, upon

receipt, that WVA integrated the updated CR x/Q values from its letter dated

September 15, 2011, into a future amendment of the FSAR. This is Open Item 137

(Appendix HH).

2.3.5 Long-Term (Routine) Atmospheric Dispersion Estimates

SRP Section 2.3.5, "Long-Term Diffusion Estimates," Revision 2, issued July 1981, states that

annual average x/Q and D/Q values should be calculated for specific receptor locations

(e.g., site boundary, residence, garden, cow) and for 16 radial sectors, with each radial sector

divided into 10 sections, from the site boundary to a distance of 50 miles from the plant, using

meteorological data representative of the annual average meteorological characteristics in the

vicinity of the plant. Annual average x/Q and D/Q values are used in the dose assessment

associated with the projected routine release of effluents from the plant.

To meet this objective, TVA made the following modifications in the FSAR:

updated the annual average x/Q and D/Q values for the WBN specific receptor locations,

based upon the meteorology data for 1986 through 2005, and dose receptor and

pathway information indicated in the 2007 land use survey

provided the annual average X/OQ and D/Q values for the 16 radial sectors, based on the

data for 1986 through 2005

The updated annual average x/Q and D/Q values associated with the specific receptor locations

appear in Table 11.3-8, and the sets of values for the 16 radial sectors are in Tables 2.3-75A

and 2.3-75B, of FSAR Amendment 104, dated June 3, 2011 (ADAMS Accession

No. ML11178A155). In addition, in FSAR Section 2.3.5, TVA stated that it discussed this topic

2-3
in the WBN Offsite Dose Calculation Manual (ODCM). FSAR Section 11.3 also discusses this

topic. Upon receipt of the updated ODCM, the NRC staff will confirm that TVA made the

corresponding revisions related to the updated annual average x/Q and D/Q values. This is

Open Item 138 (Appendix HH).

The NRC staffs review and conclusions regarding the changes proposed by TVA are discussed

in Sections 2.8.4, "Atmospheric Dispersion," and 6.1, "Design Basis Accidents," of

NUREG-0498, Supplement 2, Draft Final Environmental Statement Related to the Operation of

Watts Bar Nuclear Plant, Unit 2 (ADAMS Accession No. ML1 12980199), October 2011, and in

Section 11.3, "Gaseous Waste Management," of SSER 24, September 2011 (ADAMS

Accession No. ML11277A148).

TVA stated, in the enclosure to its letter dated August 22, 2011 (ADAMS Accession

No. ML1 12350646), that, during licensing of WBN Unit 1, it developed and used site-specific

terrain adjustment factors (TAFs) to account for topography and diurnal-related factors in

determining the dose to the maximum exposed individual. TVA applied these factors to the

receptors within approximately 5 miles of the plant, as shown in FSAR Table 3.11-8. It did not

apply TAFs to the x/Q values used for determining the 50-mile total population dose for

licensing Unit 1. It developed licensing submittals for Unit 2 on the same basis as for WBN

Unit 1, and TAFs were not used in determining the 50-mile population dose. TVA provided a

summary discussion and results of comparison evaluations in the enclosure as a further basis

for determining that the 50-mile population doses presented in the FSAR are sufficiently

conservative and that no additional correction factors (i.e., near-site TAFs) need be included.

Based upon its assessment of the information provided by TVA and a site-specific review for

WBN Unit 2, the NRC staff concludes that, for WBN Unit 2 only, it is acceptable, that TVA did

not use TAFs to calculate the x/Q and D/Q values associated with the 16 radial sectors.

2-4
5 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS

5.3 Reactor Vessel

Disposition of Open Items (Appendix HH)

5.3.1 Reactor Vessel Materials

Open Item 14

In Enclosure 1 of the Tennessee Valley Authority's (TVA's) letter dated July 31, 2010

(Agencywide Documents Access and Management System (ADAMS) Accession

No. ML102290258), TVA responded to the U.S. Nuclear Regulatory Commission (NRC) staffs

Request for Additional Information (RAI) 5.3.1-1, Question 1, and proposed changes to the

pressure-temperature limits report (PTLR), including the addition of the reactor vessel (RV)

surveillance capsule withdrawal schedule and a description of the RV surveillance program. In

addition, TVA stated that the Watts Bar Nuclear Plant (WBN) Unit 2 PTLR is included in the

WBN Unit 2 "System Description for the Reactor Coolant System" (WBN2-68-4001), which

would be revised to reflect required revisions to the PTLR by September 17, 2010

(Supplemental Safety Evaluation Report (SSER) 22, Section 5.3.1, p 5-8). The staff identified

Open Item 14 to verify that TVA had revised the PTLR to include the RV surveillance capsule

withdrawal schedule and RV surveillance program description.

In a letter dated April 6, 2011 (ADAMS Accession No. ML1 10980637), TVA stated that

Revision 1 (effective August 12, 2010) of WBN2-68-4001 reflected the required revisions to the

PTLR.

The staff concludes that TVA's response to Open Item 14 is acceptable because it confirms that

TVA has revised the PTLR to include the RV surveillance program information. Therefore,

Open Item 14 is closed.

Open Item 44

In response to RAI 5.3.1-1, Question 2, contained in Enclosure 1 to WVA's letter dated

July 31, 2010 (ADAMS Accession No. ML1 02290258), WVA stated that the material test

requirements and the acceptance standard use the nil-ductility reference temperature, RTNDT,

which is determined in accordance with American Society for Testing and Materials

(ASTM) E208, "Standard Test Method for Conducting Drop-Weight Test to Determine

Nil-Ductility Transition Temperature of Ferritic Steels." Since ASTM E208 is the standard for

drop-weight testing, while NRC requirements for determination of RTNDT require a combination

of drop-weight and Charpy V-notch (Cv) tests, the NRC staff required, in Open Item 44, that TVA

provide additional information to clarify how it determined the initial and irradiated RTNDT

(SSER 22, Section 5.3.1, pp 5-9, 5-11, 5-13, 5-17, 5-28, 5-29, 5-31, 5-32).

In a letter dated April 6, 2011 (ADAMS Accession No. ML1 10980637), in response to Open

Item 44, TVA explained that the WBN Unit 2 RV fabricator used a combination of drop-weight

and C, tests to determine the initial (unirradiated) RTNDT for the RV materials. The Cv

specimens were oriented in the tangential (strong) direction rather that the axial (weak)

direction, as is required by the American Society of Mechanical Engineers (ASME) Boiler and

Pressure Vessel Code (ASME Code), Section III, Subarticle NB-2300, which is referenced in

Appendix G, "Fracture Toughness Requirements," to Title 10 of the Code of Federal

5-1
Regulations (10 CFR) Part 50, "Domestic Licensing of Production and Utilization Facilities," with

respect to the method for determining the unirradiated RTNDT. TVA explained that it designed

the WBN Unit 2 RV in accordance with the ASME Code, Section II, 1971, addenda, which

predate the current requirements for determining the unirradiated RTNDT. However, it used the

guidance in NRC Branch Technical Position (BTP) MTEB 5-2 to adjust the C, results to

determine the equivalent properties in the weak direction, which it then used to determine the

initial RTNDT. In addition, it performed unirradiated C, tests on WBN Unit 2 RV surveillance

program materials in both the weak and strong directions, and in conjunction with the dropweight tests, to determine the unirradiated RTNDT. These C, results showed that the initial RTNDT

determination based on the Cv tests in the strong direction, adjusted using BTP MTEB 5-2, are

conservative.

TVA also stated that it determined the effects of irradiation on the WBN Unit 2 RV materials

using C, specimens in the surveillance capsules, and that it only used drop-weight testing to

determine the unirradiated RTNDT.

The staff concludes that TVA's response to Open Item 44 is acceptable because it clarifies how

it determined the initial (unirradiated) and irradiated RTNDT, and because the methods TVA used

are consistent with the guidance of BTP MTEB 5-2, which the staff finds acceptable for

determining the fracture toughness properties of RVs fabricated before the current ASME Code,

Section III, requirements for determining RTNDT came into effect. Therefore, Open Item 44 is

closed.

5.3.2 Pressure-Temperature Limits

Open Item 45

TVA stated, in its response to RAI 5.3.2-2, dated July 31, 2010, that it would revise the PTLR to

incorporate the cold overpressure mitigation system (COMS) arming temperature (SSER 22,

Section 5.3.2, pp 5-28-29, 5-32). The staff identified Open Item 45 to verify incorporation of the

COMS arming temperature into the PTLR.

In a letter dated April 6, 2011 (ADAMS Accession No. ML1 10980637), in response to Open

Item 45, WVA stated that Revision 1 (effective August 12, 2010) to WBN2-68-4001 reflected the

required revision to the PTLR. Appendix B, Section 3.2, "Arming Temperature," states, "COMS

shall be armed when any RCS [reactor coolant system] cold leg temperature is < 225 OF."

The staff concludes that WVA's response to Open Item 45 is acceptable because it confirms that

WVA revised the PTLR to specify the COMS arming temperature. Therefore, Open Item 45 is

closed.

OPen Item 46

TVA did not include the low-temperature overpressure protection (LTOP) lift settings in the

PTLR but provided them in its response to RAI 5.3.2-2 in its letter dated July 31, 2010.

In its RAI response, TVA stated that it would revise the PTLR to incorporate the LTOP lift

settings (SSER 22, Section 5.3.2, pp 5-28-29, 5-32). The staff identified Open Item 46 to verify

incorporation of the LTOP settings into the PTLR.

In a letter dated April 6, 2011 (ADAMS Accession No. ML1 10980637), in response to Open

Item 46, TVA stated that Revision 1 (effective August 12, 2010) to WBN2-68-4001 reflected the

5-2
required revisions to the PTLR. Appendix B, Table 3.1-1, "Watts Bar Unit 2 PORV

[Power-Operated Relief Valve] Setpoints vs. Temperature," contains the LTOP lift settings.

The staff concludes that TVA's response to Open Item 46 is acceptable because it confirms that

TVA revised the PTLR to include the LTOP lift settings into the PTLR. Therefore, Open Item 46

is closed.

5-3

7 INSTRUMENTATION AND CONTROLS

7.5 Safety-Related Display Instrumentation

7.5.2 Postaccident Monitoring System

7.5.2.2 Common Qualified Platform-Postaccident Monitoring System

Disposition of Open Items (Appendix HH)

Open Item 72

Open Item 72 states, "The NRC staff should complete its review and evaluation of the additional

information provided by TVA regarding the ICC [inadequate core cooling] instrumentation." The

item was generated during the U.S. Nuclear Regulatory Commission (NRC) staffs review of the

Tennessee Valley Authority's (TVA's) Watts Bar Nuclear (WBN) reactor in Section 4.4.8 of

Supplemental Safety Evaluation Report (SSER) 23, issued July 2011.

The NRC staff documented its review of the Common Q postaccident monitoring system

(PAMS) and ICC in Section 7.5.2.2 of SSER 23. The staff generated Open Items 94-111

(Appendix HH) during its review. Therefore, Open Item 72 is closed, because it is subsumed

by Open Items 94-111.

Open Item 95

Open Item 95 states, "WVA 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."

In Table 7.1-1 of Final Safety Analysis Report (FSAR) Amendment 105, dated August 12, 2011,

WVA stated the following:

IEEE Std. 603-1991, "IEEE Standard Criteria for Safety Systems for Nuclear

Power Generating Stations," (See Note 13).

Note 13 Common Q Post Accident Monitoring System (PAMS) Applicability

These Rules, Regulations and standards are applicable to the design of the

Common Q PAMS system cabinets.

Based on the revised wording in FSAR Amendment 105, the NRC staff concludes that the open

item has been adequately addressed. Therefore, Open Item 95 is closed.

Open Item 96

Open Item 96 states, "WVA 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 (3) provide justification for not conforming."

By letter dated August 4, 2011 (Agencywide Documents Access and Management System

(ADAMS) Accession No. ML11222A113), TVA responded to the open item, stating that it will

add Regulatory Guide (RG) 1.100, "Seismic Qualification of Electric and Mechanical Equipment

7-1
for Nuclear Power Plants," Revision 3, issued September 2009, to Table 7.1-1 for Common Q

PAMS. In Table 7.1-1 of FSAR Amendment 105, TVA stated the following:

Regulatory Guide 1.100, Revision 3, September 2009, "Seismic Qualification of

Electric and Mechanical Equipment for Nuclear Power Plants" (See Notes 13, 14,

and 16).

Note 13 Common Q Post Accident Monitoring System (PAMS) Applicability

These Rules, Regulations and standards are applicable to the design of the

Common Q PAMS system cabinets.

Note 16 Conformance to Regulatory Guide 1.100, Revision 3 and

IEEE 344-2004

The Common Q new design modules used in the PAMS; and the RM-1000

radiation monitors comply with IEEE 344-2004 and with Regulatory Guide 1.100

Revision 3 with the exception of issues associated with testing above 33Hz.

Based on the revised words in FSAR Amendment 105, the NRC staff concludes that the open

item has been adequately addressed. Therefore, Open Item 96 is closed.

Open Item 97

Open Item 97 states, "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."

By letter dated August 4, 2011 (ADAMS Accession No. ML1 1222A1 13), TVA responded to this

open item, stating that it will add RG 1.153, "Criteria for Safety Systems," Revision 1, issued

June 1996, to FSAR Table 7.1-1 for Common Q PAMS. In Table 7.1-1 of FSAR

Amendment 105, TVA stated the following:

Regulatory Guide 1.153, Revision 1, June 1996, "Criteria for Safety Systems"

(See Note 13).

Note 13 Common Q Post Accident Monitoring System (PAMS) Applicability

These Rules, Regulations and standards are applicable to the design of the

Common Q PAMS system cabinets.

Based on the revised words in FSAR Amendment 105, the NRC staff concludes that the open

item has been adequately addressed. Therefore, Open Item 97 is closed.

Open Item 99

Open Item 99 states, "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."

By letter dated August 4, 2011 (ADAMS Accession No. ML1 1222A1 13), TVA responded to this

action item, stating that it would add Institute of Electrical and Electronics Engineers

(IEEE) Standard (Std.) 7-4.3.2, "Application Criteria for Programmable Digital Computer

Systems in Safety Systems of Nuclear Power Generating Stations," to Table 7.1-1 for Common

Q PAMS. Table 7.1-1 of FSAR Amendment 105 states the following:

7-2
IEEE Std. 7-4.3.2-2003 "Application Criteria for Programmable Digital Computer

Systems in Safety Systems of Nuclear Power Generating Stations" (See

Notes 13 and 18).

Note 13 Common Q Post Accident Monitoring System (PAMS) Applicability

These Rules, Regulations and standards are applicable to the design of the

Common Q PAMS system cabinets.

Note 18 Conformance to IEEE 7-4.3.2-2003

The Common Q PAMS meets the applicable requirements of IEEE 7-4.3.2-2003,

except as noted below:

1. The quality program is in accordance with

WCAP-16096-NP-1A "Software Program Manual" (SPM).

2. The commercial item dedication program is in accordance with

the Westinghouse 10CFR50.54 Appendix B program."

However, in its audit report dated April 27, 2011 (ADAMS Accession No. ML110980761), the

NRC staff identified several examples of practices that do not comply with the software program

manual, or where Westinghouse no longer followed the software program manual. The

software program manual also contains criteria for commercial-grade dedication. Based on the

revised wording in FSAR Amendment 105, the NRC staff concluded that TVA did not

adequately address the item.

By letter dated September 30, 2011 (ADAMS Accession No. ML1 1286A037, Enclosure 1,

Item 17), TVA stated the following:

The inclusion of Note 18 was caused by a misinterpretation of the requirements

of RG 1.152 Revision 3 as it applied to IEEE 7-4.3.2 2003 by the TVA reviewer.

Discussion with the NRC reviewer determined that the note was not required.

Note 18 was removed in FSAR Amendment 106.

Since TVA updated the FSAR to include reference IEEE Std. 7-4.3.2-2003 as applicable to the

WBN Unit 2 Common Q PAMS, Open Item 99 is closed.

Open Item 100

Open Item 100 states, "TVA should update FSAR Table 7.1-1 to reference RG 1.168,

Revision 1, IEEE Std. 1012-1998, and IEEE 1028-1997 as being applicable to the WBN Unit 2

Common Q PAMS."

By letter dated August 4, 2011 (ADAMS Accession No. ML11222A113), TVA responded to this

open item, stating that it would add RG 1.168, "Verification, Validation, Reviews, and Audits for

Digital Computer Software Used in Safety Systems of Nuclear Power Plants," Revision 1,

issued February 2004; IEEE Std. 1012-1998, "IEEE Standard for Software Verification and

Validation"; and IEEE Std. 1028-1997, "IEEE Standard for Software Reviews," to Table 7.1-1 for

Common Q PAMS. In Table 7.1-1 of FSAR Amendment 105, TVA stated the following:

Regulatory Guide 1.168, Revision 1, February 2004, "Verification, Validation,

Reviews, and Audits for Digital Computer Software Used in Safety Systems of

7-3
Nuclear Power Plants" (See Note 13)...IEEE Std. 1012-1998, "IEEE Standard for

Software Verification and Validation" (See Note 13). IEEE Std. 1028-1997,

"IEEE Standard for Software Reviews" (See Note 13).

Note 13 Common Q Post Accident Monitoring System (PAMS) Applicability

These Rules, Regulations and standards are applicable to the design of the

Common Q PAMS system cabinets.

Based on the revised words in FSAR Amendment 105, the NRC staff concludes that the open

item has been adequately addressed. Therefore, Open Item 100 is closed.

Open Item 102

Open Item 102 states, "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."

By letter dated August 4, 2011 (ADAMS Accession No. ML1 1222A1 13), TVA responded to this

open item, stating that it would add RG 1.209, "Guidelines for Environmental Qualification of

Safety-Related Computer-Based Instrumentation and Control Systems in Nuclear Power

Plants," issued March 2007, and IEEE Std. 323-2003, "IEEE Standard for Qualifying Class 1E

Equipment for Nuclear Power Generating Stations," to Table 7.1-1. In Table 7.1-1 of FSAR

Amendment 105, TVA stated the following:

Regulatory Guide 1.209, Revision 0, March 2007, "Guidelines for Environmental

Qualification of Safety-Related Computer-Based Instrumentation and Control

Systems in Nuclear Power Plants" (See Note 13)... IEEE Std. 323-2003, "IEEE

Standard for Qualifying Class 1 E Equipment for Nuclear Power Generating

Stations," (See Note 13).

Note 13 Common Q Post Accident Monitoring System (PAMS) Applicability

These Rules, Regulations and standards are applicable to the design of the

Common Q PAMS system cabinets.

Based on the revised words in FSAR Amendment 105, the NRC staff concludes that the open

item has been adequately addressed. Therefore, Open Item 102 is closed.

Open Item 103

Open Item 103 states, "WVA 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."

By letter dated August 4, 2011 (ADAMS Accession No. ML1 1222A1 13), WVA responded to this

open item; Attachment 3 to the letter included the requested evaluation.

Based on its review of the information submitted by WVA, the staff concluded that TVA provided

an adequate evaluation of conformance with RG 1.209 and IEEE Std. 323-2003. Therefore,

Open Item 103 is closed.

Open Item 104

Open Item 104 states, "The NRC staff will review the WEC [Westinghouse Electric Corporation]

self-assessment to verify that the WBN Unit 2 PAMS complies with the V&V [verification and

7-4
validation] requirements in the SPM or that deviations from the requirements are adequately

justified."

By letter dated June 10, 2011 (ADAMS Accession No. ML11167A110), TVA provided

information to address this item (see letter item 15, "NRC Common Q PAMS Audit Action Items

Response"). The information provided and the information subsequently reviewed demonstrate

that Westinghouse did not consistently follow the software program manual and that

Westinghouse did not plan and document these deviations beforehand; however, TVA did

provide sufficient information to allow the NRC staff to determine, with reasonable assurance,

that the Common Q PAMS does not pose a threat to public health and safety. Therefore, Open

Item 104 is closed.

Open Item 106

Open Item 106 states, "TVA should provide to the NRC staff documentation to confirm that the

final WBN Unit 2 Common Q PAMS SRS [system requirements specification] is independently

reviewed."

By letter dated June 10, 2011 (ADAMS Accession No. ML1 11 67A1 10), WVA responded to "NRC

Request (Item Number 372)," documenting why it concluded that the reviews were independent.

The information provided was essentially an affirmation that the documents were independently

reviewed. The Westinghouse internal guidance (reviewed during the NRC audit documented at

ADAMS Accession No. ML1 10980761) is inconsistent on how to document that a review was an

"independent review-satisfying Appendix B" versus a more informal review. However, the

NRC staff considers that the affirmation provided by IVA in its letter dated June 10, 2011, is

sufficient; therefore, Open Item 106 is closed.

Open Item 107

Open Item 107 states, "WVA should provide to the NRC staff documentation to confirm that the

final WBN Unit 2 Common Q PAMS SDDs [software design documents] are independently

reviewed."

By letter dated June 10, 2011 (ADAMS Accession No. MI111 67A1 10), TVA responded to "NRC

Request (Item Number 373)," documenting why it concluded that the reviews were independent.

The information provided was essentially an affirmation that the documents were independently

reviewed. The Westinghouse internal guidance (reviewed during the NRC audit documented at

ADAMS Accession No. ML1 10980761) is inconsistent on how to document that a review was an

"independent review-satisfying Appendix B" versus a more informal review. However, the

NRC staff considers the affirmation provided by TVA in its letter dated June 10, 2011, to be

sufficient; therefore, Open Item 107 is closed.

Open Item 109

Open Item 109 states, "TVA should demonstrate to the NRC staff acceptable data storm testing

of the Common Q PAMS."

By letter dated June 10, 2011 (ADAMS Accession No. ML1 11 67A1 10), TVA responded to

"18. NRC Verbal Request," which stated the following:

Watts Bar Unit 2 Post Accident Monitoring System went through a Data Storm

Test to verify that the safety related functions of the system driven by the Advant

7-5
Controller 160 (AC 160) and the safety related indications monitored on the

Operator Module (OM) located in the Main Control Room (MCR) are not affected

when the Ethernet network interface of the Maintenance and Test Panel (MTP) is

under data storm conditions. This test was requested by TVA.

The purpose of the data storm test was to test the ability of the MTP to handle

the possible volume of traffic generated by a broadcast storm without impacting

the safety functions. A broadcast storm occurs when a large number of

broadcast packets are received. Forwarding these packets can cause the

network to slow down or to time out.

Another objective of the data storm test was to test the ability of the MTP to

handle malformed packets possibly generated by a data storm without impacting

the safety functions.

TVA also described the pass/fail criteria and associated results. The PAMS continued to

perform its safety function during the data storm testing, as demonstrated by its meeting the

acceptance criteria.

Based on its review of the information provided by TVA, the NRC staff concludes that the open

item was acceptably addressed. Therefore, Open Item 109 is closed.

7.5.2.2.3.9.2.5 IEEE Std. 603-1991, Clause 5.6, Independence

In Section 7.5.2.2.3.9.2.5 of SSER 23, issued July 2011, the NRC staff concluded that the

Common Q PAMS meets the acceptance criteria for independence and the requirements of

Clause 5.6; therefore, the PAMS is acceptable.

Core exit thermocouples (CETs) are bundled together with the non-Class 1 E self-powered

neutron detectors in the mineral insulated cable and thus do not meet the separation

requirements between safety- and nonsafety-related equipment in IEEE Std. 384-1981, "IEEE

Standard Criteria for Independence of Class 1 E Equipment and Circuits." The NRC staff

documented its separation/isolation evaluation for the CETs in Section 7.7.1.9 of SSER 24,

issued September 2011.

7.5.2.3 High-Range Containment Area Radiation Monitors

Disposition of Open Items (Appendix HH)

Open Item 78

Open Item 78 states, "TVA intends to issue a revised calculation reflecting that the TID [total

integrated dose] in the control room is less than 1 x103 rads, which will be evaluated by the NRC

staff."

TVA's letter dated August 4, 2011 (ADAMS Accession No. ML1 1222A1 13), responded to the

open item, and Enclosure 1 to the letter provided TVA's supporting calculation (Calculation

No. WBNAPS3-127, Revision 0). TVA stated that the TID for the main control room, where the

radiation monitors are located, is 362.76 rads. This radiation dose is acceptable to the NRC

staff for mild environments according to the guidance in RG 1.209. Based on its review of the

information provided in TVA's letter dated August 4, 2011, and the calculation summary, the

staff concluded that TVA provided an adequate response. Therefore, Open Item 78 is closed.

7-6
7.7 Control Systems Not Required for Safety

7.7.1 System Description

7.7.1.9 In-Core Instrumentation System

Disposition of Open Items (Appendix HH)

Open Item 119

Open Item 119 states, "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."

In Enclosure 1 and in Attachment 2 to its letter dated September 1, 2011 (ADAMS Accession

No. ML1 1257A048), TVA responded to this open item. Attachment 2 included a copy of

Westinghouse analysis WNA-CN-00157-WBT, Revision 0, "Watts Bar 2 Incore Instrument

System (IIS) Signal Processing System (SPS) Isolation Requirements." The analysis shows the

following:

1. No fault originating either within the WINCISE (Westinghouse INCore Information,

Surveillance, and Engineering system) signal processing system (SPS) cabinet or

electrical surges on the input power lines or output communications link can result in

fault voltage affecting CET signals.

2. Inadvertent disconnection or failures of any IITA (in-core instrumentation thimble

assembly) emitter wire or wires at the WINCISE SPS cabinet, or anywhere in the cabling

between the IITA seal table connector and the WINCISE SPS cabinet, will not cause

voltage charge-up on the emitter wire to exceed 600 volts direct current during worst

case (full-power) plant operation, thus preventing the fault voltage from affecting the

associated CET.

3. For large input overvoltage conditions on the 120-volt alternating current (Vac) input

instrument bus used in both the SPS cabinet and associated PAMS train, the SPS will

not cause any failure within the PAMS train that would not otherwise occur as a direct

result of the overvoltage condition within the PAMS power supply.

The impact of a failed CET on a PAMS channel is addressed in Westinghouse analysis

WNA-AR-001 80-WBT-P, Revision 2, "Failure Modes and Effects Analysis (FMEA) for the Post

Accident Monitoring System," page 3-6, in TVA's letter dated March 2, 2011 (ADAMS Accession

No. ML110620411).

Based on the analysis results, the NRC staff concludes that TVA adequately addressed the

issue, and Open Item 119 is closed.

7-7
Open Item 120

Open Item 120 states, "TVA should 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 SPS cabinet is not routed with other cables greater than 264 VAC."

In Enclosure 1 and in Attachments 2 and 5 to its letter dated September 1, 2011 (ADAMS

Accession No. ML1 1257A048), TVA responded to this open item. The information provided by

TVA confirms that 264 Vac is the estimated maximum voltage, or surge-voltage, that could

affect WINCISE. Furthermore, the equipment qualification for the WINCISE SPS, provided in

Attachment 5, shows that the SPS cabinets can withstand a maximum surge of 4,000 volts with

no loss of function after the surge.

However, TVA's response did not confirm that the power supply cable to the WINCISE SPS

cabinet would not be routed with other cables carrying greater than 264 Vac. TVA must confirm

that the power supply cable to the SPS cabinet is not routed with other cables with voltage

greater than 264 Vac. Open Item 120 remains open.

Open Item 122

Open Item 122 states, "TVA should confirm to the NRC staff that different divisions of safety

power are supplied to the IIS SPS cabinets, with the power cables routed in separate shielded

conduits."

In Enclosure I to its letter dated September 1, 2011 (ADAMS Accession No. MLI 1257A048),

TVA confirmed that different divisions of safety power are supplied to the IIS SPS cabinets, with

the power cables routed in separate shielded conduits. Therefore, the NRC staff concludes that

TVA adequately addressed the issue, and Open Item 122 is closed.

Open Item 124

Open Item 124 states, "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."

In Enclosure 1 to its letter dated September 1, 2011 (ADAMS Accession No. ML1I1257A048),

TVA explained that the feature to automatically switch over to BEACON B is not configured for

WBN Unit 2, and the BEACON datalink on the application server can only connect to

BEACON A. TVA provided further information in Attachment 3 to its letter dated June 10, 2011

(ADAMS Accession No. ML1 1 167A1 10). The NRC staff reviewed the information provided in

TVA's letters dated June 10 and September 1, 2011, and concluded that its response was

acceptable. Therefore, Open Item 124 is closed.

Open Item 128

Open Item 128 states, "TVA should submit the seismic qualification test report procedures and

results for the SPS cabinets to the NRC staff for review."

In Enclosure 1 and in Attachment 5 to its letter dated September 1, 2011 (ADAMS Accession

No. ML11257A048), TVA responded to this open item. Attachment 5 includes a copy of the

Westinghouse report EQ-QR-39-WBT, Revision 0, "Equipment Qualification Summary Report

7-8
for WINCISE Signal Processing System." This summary report describes the seismic

qualification test performed and the results obtained. The WINCISE SPS cabinet maintained

structural integrity without any component detachment throughout the test program and thus

complied with WBN Unit 2 seismic qualification specification WB-DC-40-31.2, with testing

performed in accordance with NRC RG 1.100, IEEE Std. 344-1975, and IEEE Std. 344-1987,

"IEEE Recommended Practice for Seismic Qualification of Class 1E Equipment for Nuclear

Power Generating Stations." Based on the information in the summary report, the NRC staff

concludes that TVA adequately addressed the issue, and Open Item 128 is closed.

Open Item 130

Open Item 130 states, "TVA should provide a summary to the NRC staff of the EMC

[electromagnetic compatibility] qualification test results of the SPS cabinets."

In Enclosure 1 and in Attachment 5 to its letter dated September 1, 2011 (ADAMS Accession

No. ML11257A048), TVA responded to this open item. Attachment 5 includes a copy of the

Westinghouse report EQ-QR-39-WBT, Revision 0, "Equipment Qualification Summary Report

for WINCISE Signal Processing System." This summary report describes the EMC qualification

test performed and the results obtained. The report states that the WINCISE SPS cabinet

successfully complied with the emissions requirements of NRC RG 1.180, Revision 1,

"Guidelines for Evaluating Electromagnetic and Radio-Frequency Interference in Safety-Related

Instrumentation and Control Systems," and the supplemental-surge-withstand testing required in

WNA-CN-00157-WBT. Therefore, the NRC staff concludes that TVA adequately addressed the

issue, and Open Item 130 is closed.

7-9

11 RADIOACTIVE WASTE SYSTEM

11.3 Gaseous Waste Management

In Section 11.3 of Supplemental Safety Evaluation Report (SSER) Supplement 24 (Agencywide

Documents Access and Management System (ADAMS) Accession No. ML1 1277A148), issued

September 2011, to the safety evaluation report (SER) (NUREG-0847, "Safety Evaluation

Report Related to the Operation of Watts Bar Nuclear Plant, Units 1 and 2," issued June 1982),

the staff of the U.S. Nuclear Regulatory Commission (NRC) documented its evaluation of

Tennessee Valley Authority's (TVA's) calculated doses to members of the public as revised by

Final Safety Analysis Report (FSAR) Amendments 95 through 104. The staff concluded that the

Watts Bar Nuclear Plant (WBN) Unit 2 design meets the dose limits in Title 10 of the Code of

Federal Regulations (10 CFR) Part 20, "Standards for Protection Against Radiation," and the

design objectives in 10 CFR Part 50, "Domestic Licensing of Production and Utilization

Facilities," Appendix I, "Numerical Guides for Design Objectives and Limiting Conditions for

Operation to Meet the Criterion 'As Low as is Reasonably Achievable' for Radioactive Material

in Light-Water-Cooled Nuclear Power Reactor Effluents," Sections II.A, II.B, and II.C. However,

the revised doses did not support a conclusion that the criteria for gaseous effluents in RM 50-2

(as annexed to Appendix I to 10 CFR Part 50) are met. This was documented as Open

Item 135 (Appendix HH). In response, by letter dated July 28, 2011, TVA provided a costbenefit analysis, as required by 10 CFR Part 50, Appendix I, Section II.D, to demonstrate that a

sufficient reduction in the annual collective dose to the public within a 50-mile radius would not

be achieved by reasonable enhancements to the design of the WBN gaseous effluent

processing systems.

TVA performed the cost-benefit analysis for several potential enhancements to the gaseous

radwaste systems in accordance with the methodology presented in Regulatory Guide

(RG) 1.110, "Cost-Benefit Analysis for Radwaste Systems for Light-Water-Cooled Nuclear

Power Reactors (for Comment)," issued March 1976. In lieu of expressing the cost and benefits

in terms of dollars, TVA's analysis compares the potential annual collective dose (person-rem)

saving by the specific system enhancement to a threshold dose indicated by the annualized

cost of the enhancement (e.g., at $1,000/person-rem, an annualized cost of $2,000 would

indicate that an annual collective dose savings of at least 2 person-rem would be required for

the enhancement to have a favorable cost-benefit ratio of less than 1.0). The analysis

considered the total body and thyroid collective doses caused by the design-basis gaseous

effluent releases to the population anticipated to live within 50 miles of the plant in the

year 2040.

TVA calculated an annual collective total body dose of 6.68 person-rem resulting from gaseous

effluent releases with the current radwaste system design. Using the methods in RG 1.110,

WVA calculated the lowest threshold value for a gaseous enhancement at 6.32 person-rem.

Since this threshold dose is below the total gaseous effluent dose, the licensee was required to

consider specific dose savings associated with each system enhancement. The lowest cost

enhancement included in RG 1.110 is the addition of a steam generator blowdown flash tank

vent routed to the main condenser. The addition of this vent would only lower turbine building

vent releases. TVA calculated the total body dose from all turbine building vent releases to be

0.033 person-rem. This is much less than the threshold value, so the enhancement is not cost

beneficial. The next lowest cost enhancement is the addition of a 600-cubic foot (ft3) gas decay

tank. TVA calculated the threshold value of this enhancement to be 7.46 person-rem. This is

higher than the 6.68 person-rem total body dose from all WBN gaseous effluent releases.

11-1
Therefore, adding this enhancement was not determined to be cost beneficial. Consistent with

the RG 1.110 methodology, no additional total body dose reductions needed to be considered.

TVA calculated an annual collective thyroid dose of 13.0 person-rem from gaseous effluent

releases with the current radwaste system design. Consistent with the RG 1.110 methodology,

TVA evaluated the following seven enhancements relative to the thyroid dose in order of cost

from lowest to highest:

1. Routing the steam -generator blowdown flash tank vent to the main condenser: The

collective thyroid dose from the turbine building vent of 0.354 person-rem is significantly

below the 6.32 person-rem threshold value for adding the flash tank vent. Therefore,

this enhancement is not cost beneficial.

2. The addition of a 600-ft3 waste gas decay tank: This enhancement has already been

incorporated in the plant design. WBN already has nine 600-ft3 gas decay tanks with

sufficient capacity to store gaseous waste to a minimum of 60 days before release. The

decayed gases are discharged through a charcoal adsorber/high-efficiency particulate

air (HEPA) filter and a radiation monitor before being released through the shield

building stack. Adding a tenth gas decay tank will not reduce the thyroid dose. The

threshold value of 7.46 person-rem was found to be in excess of the potential dose

reduction.

3. The addition of a 1,000-cubic-foot-per-minute (cfm) charcoal/HEPA filter to the auxiliary

building ventilation system: TVA calculated a threshold value for this enhancement at

7.58 person-rem for a single 1,000-cfm filter unit. This is lower than the 9.15 thyroid

person-rem resulting from auxiliary building vent releases. However, the auxiliary

building vent air flow rate is 84,000 cfm. Treating only 1,000 cfm of the total release

would impact less than 2 percent of the resulting dose. Thus, in reality, the threshold

value for the auxiliary building would be many times the 7.58 value, since more than 50

1,000-cfm filter packages would be needed to effectively filter the auxiliary building

gaseous effluent. Therefore, this enhancement would not be cost beneficial.

4. A charcoal/HEPA filter to the main condenser vacuum exhaust: The threshold value for

this modification was 7.69 person-rem. This exceeds the thyroid dose release from the

turbine building. This enhancement would not be cost beneficial.

5. A 3-ton charcoal adsorber to the shield buildinq vent: RG 1.110 assumes that this

enhancement would be located in the turbine building and appended to the waste gas

decay system. The threshold dose value for this enhancement is 8.77 person-rem.

However, TVA stated that the WBN waste gas decay system is located in the auxiliary

building and is vented through the shield building stack. The total annual cost of placing

the filter in the auxiliary building would be higher than the turbine building placement

assumed in RG 1.110. Notwithstanding that issue, TVA calculated the thyroid

population dose from the reactor building (which is the location of the shield building

vent) to be 3.48 person-rem. The threshold value is greater than the potential dose

reduction. Therefore this enhancement is not cost beneficial.

6. An air eiector charcoal/HEPA filtration unit: WBN Unit 2 does not have air ejectors but

uses condenser vacuum pumps. TVA already evaluated the addition of filters on the

vacuum pump discharge and showed that it was not cost beneficial. This enhancement

has a higher cost and was eliminated from further consideration.

11-2
7. A 15,000-cfm HEPA filtration system for the turbine building or the auxiliary building:

The threshold value for a single filter unit was 16.9 person-rem. The threshold value for

this enhancement exceeds the 13.0 person-rem total thyroid population dose.

Therefore, consistent with the RG 1.110 methodology, TVA did not consider further

enhancements to the gaseous effluent processing systems.

Conclusions

The NRC staff performed an independent cost-benefit analysis to verify that these gaseous

effluent treatment enhancements are not cost effective. The annualized costs of the gaseous

effluent system enhancements identified by TVA were determined in accordance with the

guidance in RG 1.110. Using the GASPAR input parameters described in SSER 24, the staff

calculated the 50-mile radius collective doses (expected for the year 2040) listed in Table 11.1.

Table 11.1 NRC Staff's 50-Mile Collective

Population Dose per Release Source

(Person-Rem)

ORGAN AUXILIARY TURBINE CONTAINMENT TOTAL

Total Body 2.92E+00 4.19E-02 3.15E+00 6.11E+00

Thyroid 1.04E+01 4.16E-01 3.47E+00 1.42E+01

In addition, the staff assumed the minimum capital recovery factor listed in Table A-6 of

RG 1.110, which resulted in lower annualized costs of the proposed effluent treatment

enhancements by about 10 percent. Despite slightly higher collective doses and somewhat

lower annualized costs, the staffs analysis verified the TVA conclusions above, that no

enhancements evaluated result in a favorable cost-benefit ratio. Therefore, the licensee is not

required to include these enhancements in the WBN gaseous effluent systems.

Based on the NRC staffs evaluation, as documented in SSER 24, and the results of the

cost-benefit analysis, the staff concludes that the WBN effluent systems can maintain gaseous

releases within the design objectives in 10 CFR Part 50, Appendix I, Sections ILA, II.B, and II.C,

and are as low as is reasonably achievable, as required in 10 CFR Part 50, Appendix I,

Section ll.D. Therefore, Open Item 135 is closed. 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 updated FSAR, the NRC staff will confirm that the update has been

made by TVA. This is Open Item 139 (Appendix HH).

11-3

13 CONDUCT OF OPERATIONS

13.1 Or-ganizational Structure of Applicant

13.1.3 Plant Staff Organization

Disposition of Open Items (Appendix HH)

Open Item 11 (Appendix HH) states the following:

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 SRO [senior reactor operator], if one is available. The plant

administrative procedures should clearly describe this transfer of control room

duties.

In Enclosure 1 of its letter dated August 15, 2011 (Agencywide Documents Access and

Management System Accession No. ML1 1230A385), the Tennessee Valley Authority

provided its revised Watts Bar Nuclear Plant Unit 1/Unit 2 as-designed fire protection

report. The U.S. Nuclear Regulatory Commission staff reviewed Section 9.1, "Fire

Brigade Staffing," of the fire protection report and concluded that, because it ensures

that the fire brigade will not use members of the minimum shift crew, there is no need for

an administrative procedure to describe a transfer of control room duties. Therefore,

Open Item 11 is closed.

13-1

15 ACCIDENT ANALYSIS

15.4 Radiological Consequences of Accidents

To evaluate the effectiveness of the engineered safety features proposed for Watts Bar Nuclear

Plant (WBN) Unit 2 and to ensure that the radiological consequences of these accidents meet

the applicable dose criteria, the staff of the U.S. Nuclear Regulatory Commission (NRC)

reviewed the Tennessee Valley Authority's (TVA's) analyses for the loss of alternating current

(ac) power to the plant auxiliary equipment, the waste gas decay tank rupture, the

loss-of-coolant accident (LOCA), the main steamline break (MSLB), the steam generator tube

rupture (SGTR), and the fuel-handling accident (FHA). TVA did not provide an updated detailed

evaluation for the rod ejection accident because this accident is bounded by the LOCA. TVA

calculated accident doses for the exclusion area boundary (EAB), the outer boundary of the

low-population zone (LPZ), and the control room (CR). Table 15.1 of this supplemental safety

evaluation report (SSER) shows TVA's calculated doses for the analyzed accidents.

The atmospheric dispersion factors (x/Q) used in the WBN analyses are those that the NRC

staff discussed in Section 2.3.4 of the safety evaluation report (SER) (NUREG-0847, "Safety

Evaluation Report Related to the Operation of Watts Bar Nuclear Plant, Units 1 and 2," issued

June 1982) and in its SSERs. The meteorological models described in the NRC regulatory

guides (RGs) referenced in these analyses are appropriately modified by those presented in

RG 1.145, "Atmospheric Dispersion Models for Potential Accident Consequence Assessments

at Nuclear Power Plants," Revision 1, issued February 1983.

The NRC staff reviewed WBN Unit 2 Final Safety Analysis Report (FSAR) Section 15.5,

"Environmental Consequences of Accidents,", using the criteria in NUREG-75/087, "Standard

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

Edition," Revision 2, issued May 1980. Since the acceptance criteria in NUREG-0800,

"Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants"

(SRP), are not significantly different from those in NUREG-75/087, the NRC staffs conclusions

in this SSER section are also in accordance with the acceptance criteria of the SRP.

The NRC staff previously reviewed the radiological dose consequence analyses performed by

TVA for WBN Units 1 and 2 and documented the results of the review in the SER and in SSERs

through SSER 18, issued October 1995. Subsequently, the NRC issued WBN Unit 1 an

operating license, and it began commercial operation in 1996. Since the initial NRC staff

review, the licensee made changes to the radiological dose consequence analyses for Unit 1

using Title 10 of the Code of Federal Regulations (10 CFR) 50.59, "Changes, Tests, and

Experiments," and the license amendment process in 10 CFR 50.90, "Application for

Amendment of License, Construction Permit, or Early Site Permit." For this SSER, the NRC

staff used the current licensing basis (CLB) for Unit 1 as a benchmark for acceptability of the

Unit 2 analyses and addressed those revised Unit 1 parameters and assumptions that are

different for Unit 2.

TVA addresses the dose consequences of the following seven postulated design-basis

accidents in Section 15.5, "Environmental Consequences of Accidents," of the WBN Unit 2

FSAR:

(1) loss of alternating current (ac) power to the plant auxiliaries

(2) waste gas decay tank rupture

(3) LOCA

15-1
(4) MSLB

(5) SGTR

(6) FHA

(7) rod ejection accident

15.4.1 Loss-of-Coolant Accident

The NRC staff previously reviewed the LOCA radiological dose consequence analysis

performed for WBN Units 1 and 2 and documented the results of this review in the SER and in

SSERs 5, 9, 15, and 18 (issued October 1995). Subsequently, the NRC issued WBN Unit 1 an

operating license, and it began commercial operation in 1996. Since the initial NRC staff

review, TVA made changes to the LOCA radiological dose consequence analysis for Unit 1

using the 10 CFR 50.59 change process and the 10 CFR 50.90 license amendment process.

For this SSER, the NRC staff used the CLB LOCA assumptions for Unit 1 as a benchmark for

acceptability of the Unit 2 LOCA analysis and addressed the following revised Unit 1 parameters

and assumptions that are different for Unit 2.

Time-Dependent Emergency Gas Treatment System Flow Rates

The time-dependent emergency gas treatment system (EGTS) flow rates for Units 1 and 2 are

revised as a result of an alternative single failure scenario resulting in one pressure control train

in full exhaust to the shield building exhaust stack while the other train remains functional. Both

EGTS fans are in service until operator action is taken to place one fan in standby between

I and 2 hours after the accident.

The Unit 1 and Unit 2 EGTS flow rates are based on separate analyses of ventilation system

calculations reflecting the physical differences between the units.

Source Term

The source term used in the LOCA dose consequence analysis performed for Unit 2 differs from

the Unit 1 LOCA source term, because the Unit 1 source term is based on a tritium core,

whereas the Unit 2 source term is based on a standard core. The Unit 2 LOCA dose

consequence analysis did not consider tritium-producing burnable absorber rods (TPBARs),

since TVA did not request permission to produce tritium in Unit 2.

The NRC staff review concluded that the aforementioned unit-specific changes used to evaluate

the WBN Unit 2 LOCA analysis are based on physical differences in the units, unit-specific

operational differences, or updated information that has been accepted by the NRC staff for use

in dose consequence analyses. The TVA evaluation concluded that the radiological

consequences resulting from a postulated LOCA at the EAB, LPZ, and CR comply with the

reference values and the CR dose criterion provided in 10 CFR 100.11, "Determination of

Exclusion Area, Low-Population Zone, and Population Center Distance," and in 10 CFR Part 50,

"Domestic Licensing of Production and Utilization Facilities," Appendix A, "General Design

Criteria for Nuclear Power Plants," General Design Criterion (GDC) 19, "Control Room," as well

as with the accident-specific dose guidelines specified in the SRP. The NRC staff's review

concluded that TVA used analyses, assumptions, and inputs consistent with those described in

the SRP. The NRC staff concluded that the assumptions presented in Table 15.2 and TVA's

calculated dose results in Table 15.1 of this SSER were acceptable. The NRC staff concludes

that TVA's estimates of the dose consequences of a design-basis LOCA are in accordance with

the SRP acceptance criteria.

15-2
15.4.2 Main Steamline Break Outside Containment

The NRC staff previously reviewed the MSLB radiological dose consequence analysis

performed for WBN Units 1 and 2 and documented the results of this review in the SER and in

SSER 15 (issued June 1995). Subsequently, the NRC issued WBN Unit 1 an operating license,

and it began commercial operation in 1996. Since the initial NRC staff review, the licensee

made changes to the MSLB radiological dose consequence analysis for Unit 1 using the

10 CFR 50.59 change process and the 10 CFR 50.90 license amendment process. For this

SSER, the NRC staff used the CLB MSLB assumptions for Unit 1 as a benchmark for

acceptability of the Unit 2 MSLB analysis and addressed the following revised Unit 1 parameters

and assumptions that are different for Unit 2.

Steam Release Values

Unit 2 steam release values are different from Unit 1 values because the Unit 1 steam

generators have been replaced by differently designed models. The volume of steam release in

Unit 2 reflects the use of the original steam generators (OSGs).

Iodine Spike Assumptions

For Unit 2, the short-term maximum allowable dose equivalent iodine-131 (DEI) value is

14 microcuries/gram (pCi/gm). TVA has informed the NRC staff (reference Agencywide

Documents Access and Management System (ADAMS) Accession No. ML1 1252A530) that it

will submit a license amendment request to revise the Unit 1 technical specification (TS)

short-term maximum allowable DEI from 21 pCi/gm to 14 pCi/gm. This change is necessary,

since the dose conversion factors (DCFs) were changed from International Commission on

Radiological Protection (ICRP)-2, "Permissible Dose for Internal Radiation," 1959, values to the

values in RG 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor

Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I,"

Revision 1, issued October 1977. The Unit 1 TS require that RG 1.109 be used to determine

the DEL. With this change, the short-term maximum allowable DEl value of 14 pCi/gm will be

consistent for both WBN Unit 1 and Unit 2. The NRC staff has endorsed the use of the DCFs

from RG 1.109 in dose consequence analyses.

Atmospheric Dispersion Factors

The Unit 2 atmospheric dispersion coefficients are based on a release from the Unit 2 vent

valve stacks to the emergency CR intake.

The NRC staff review found that the aforementioned unit-specific changes used to evaluate the

WBN Unit 2 MSLB analysis are based on physical differences in the units, unit-specific

operational differences, or updated information that has been accepted by the NRC staff for use

in dose consequence analyses. The TVA evaluation concluded that the radiological

consequences resulting from the postulated MSLB at the EAB, LPZ, and CR comply with the

reference values and the CR dose criterion provided in 10 CFR 100.11 and 10 CFR Part 50,

Appendix A, GDC 19, as well as with the accident-specific dose guidelines specified in the SRP.

The NRC staff concluded that TVA used analyses, assumptions, and inputs consistent with

those described in the SRP. The NRC staff concluded that the assumptions presented in

Table 15.3 and TVA's calculated dose results given in Table 15.1 of this SSER were

acceptable. The NRC staff concluded that TVA's estimates of the dose consequences of a

15-3
design-basis MSLB are in accordance with the SRP acceptance criteria.

15.4.3 Steam Generator Tube Rupture

The NRC staff previously reviewed the SGTR radiological dose consequence analysis

performed for WBN Units I and 2 and documented the results of its review in the SER and in

SSERs 3, 5,12,14, and 15 (issued June 1995). Subsequently, the NRC issued WBN Unit1 an

operating license, and it began commercial operation in 1996. Since the initial NRC staff

review, TVA made changes to the SGTR radiological dose consequence analysis for Unit 1

using the 10 CFR 50.59 change process and the 10 CFR 50.90 license amendment process.

To complete this SSER, the NRC staff used the CLB SGTR assumptions for Unit 1 as a

benchmark for acceptability of the Unit 2 analysis and addressed the following revised Unit 1

parameters and assumptions that are different for Unit 2.

Steam Release Values

Unit 2 steam release values are different from Unit 1 values because the Unit 1 steam

generators have been replaced by differently designed models. The Unit 2 secondary-side

mass releases from the ruptured and intact steam generator and primary coolant are based on

a reactor coolant system (RCS) with the OSGs.

Iodine Spike Assumptions

For Unit 2, the short-term maximum allowable DEI value is 14 pCi/gm. TVA has informed the

NRC staff (reference ADAMS Accession No. ML1 1252A530) that it will submit a license

amendment request to revise the Unit 1 TS short-term maximum allowable DEl from 21 pCi/gm

to 14 pCi/gm. This change is necessary since the DCFs were changed from ICRP-2 values to

RG 1.109 values. The Unit 1 TSs require that RG 1.109 be used to determine the DEL. With

this change, the short-term maximum allowable DEI value of 14 pCi/gm will be consistent for

both WBN Unit 1 and Unit 2. The NRC staff has endorsed the use of the DCFs from RG 1.109

in dose consequence analyses.

Source Term

The source term used in the SGTR dose consequence analysis performed for Unit 2 differs from

the Unit 1 SGTR source term because the Unit 1 source term is based on the failure of two

TPBARs, whereas the Unit 2 source term is based on a standard core. The Unit 2 SGTR dose

consequence analysis did not consider TPBARs, since TVA did not request permission to

produce tritium in Unit 2.

Atmospheric Dispersion Factors

The Unit 2 atmospheric dispersion coefficients are based on the Unit 2 release location at the

Unit 2 vent valve stacks and the Unit 2 CR air intake on the east side of the control building.

The NRC staff review concluded that the aforementioned unit-specific changes used to evaluate

the WBN Unit 2 SGTR analysis are based on physical differences in the units, unit-specific

operational differences, or updated information that has been accepted by the NRC staff for use

in dose consequence analyses. The TVA evaluation concluded that the radiological

consequences resulting from the postulated SGTR at the EAB, LPZ, and CR comply with the

reference values and the CR dose criterion provided in 10 CFR 100.11 and 10 CFR Part 50,

15-4
Appendix A, GDC 19, as well as with the accident-specific dose guidelines specified in the SRP.

The NRC staffs review found that TVA used analysis, assumptions, and inputs consistent with

those described in the SRP. The NRC staff concluded that the assumptions presented in

Table 15.4 and the TVA's calculated dose results given in Table 15.1 of this SSER were

acceptable. The NRC staff concludes that TVA's estimates of the dose consequences of a

design-basis SGTR are in accordance with the SRP acceptance criteria.

15.4.4 Control Rod Ejection Accident

The NRC staff evaluated the control rod ejection accident in Section 15.4.4 of the SER

(NUREG-0847), and there have been no supplements to this section. The current WBN Unit 2

FSAR Section 15.5, "Environmental Consequences of Accidents," does not include a detailed

evaluation of this accident except to state that it is bounded by the LOCA. The LOCA dose

consequence results for WBN Unit 2 are less than 25 percent of the reference values in

10 CFR 100.11. Since the source term for a control rod ejection accident is considerably less

than for a LOCA, and the dose consequence results for the WBN Unit 2 LOCA are less than the

SRP acceptance criteria for a rod ejection accident (25 percent of the reference values in

10 CFR 100.11), the staff concludes that the dose consequence for the rod ejection accident will

be bounded by the LOCA for WBN Unit 2.

15.4.5 Fuel-Handling Accident

15.4.5.1 Fuel-Handling Accident, Regulatory Guide 1.25 Analysis

The NRC staff previously reviewed the FHA radiological dose consequence analysis performed

for WBN Units 1 and 2 in the SER and in SSERs 4 and 15 (issued June 1995). Subsequently,

the NRC issued WBN Unit 1 an operating license, and it began commercial operation in 1996.

Since the initial NRC staff review, TVA made changes to the FHA radiological dose

consequence analysis for Unit 1 using the 10 CFR 50.59 change process and the 10 CFR 50.90

license amendment process. To complete this SSER, the NRC staff used the CLB FHA

assumptions for Unit 1 as a benchmark for acceptability of the Unit 2 analysis and addressed

the revised Unit 1 parameters and assumptions that are different for Unit 2.

Source Term

In the CLB for the WBN Unit 1 FHA, two TPBARs in the dropped assembly are assumed to

break and release the entire contents of tritium. All of the tritium is conservatively assumed to

evaporate into the air. The Unit 2 analysis does not consider TPBARs since TVA did not

request permission to produce tritium in Unit 2.

In its letter dated September 23, 2011 (ADAMS Accession No. ML1 1269A064), TVA provided

the results of a revised dose consequence analysis for the FHA. The revised analysis includes

the following changes to the FHA for Unit 2:

The damper closure times for auxiliary building and main CR dampers in the normal

ventilation system are changed.

The alternative source term (AST) is used as the basis for the dose calculations for the

FHA in the auxiliary building and for the FHA in containment when the equipment hatch

is open.

15-5
The meteorology data for the 20-year period of 1991 to 2010 are incorporated, as

opposed to the 1976 to 1993 data used for licensing Unit 1.

TVA provided two separate dose consequence analyses covering three cases for the FHA. The

first case considered by TVA is for an FHA inside containment with the containment

penetrations closed to the auxiliary building and the reactor building purge ventilation system

(RBPVS) operating. This case was evaluated using the assumptions from RG 1.25,

"Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling

Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors

(Safety Guide 25)," issued March 1972, and NUREG-5009, "Assessment of the Use of

Extended Burnup Fuel in Light Water Power Reactors," issued February 1989, as reviewed by

the NRC staff in the SER and in SSERs 4 and 15. The release is from the shield building vent

with credit taken for the RBPVS filtration. A filter efficiency of 90 percent for inorganic iodine

and 30 percent for organic iodine for the purge air exhaust filters is used, since no relative

humidity control is provided.

The second case considered by the applicant is for an FHA in the spent fuel pool (SFP) area

located in the auxiliary building. This case was evaluated using the assumptions from

RG 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at

Nuclear Power Reactors," issued July 2000. In this case, no credit is taken in the analysis for

the auxiliary building gas treatment system (ABGTS) or containment purge system filters.

The third case is an open containment case for an FHA inside containment where there is open

communication between the containment and the auxiliary building. This evaluation also uses

the AST assumptions from RG 1.183 with no credit for any filtration systems.

The NRC staff reviewed the unit-specific changes used to evaluate the RG 1.25 WBN Unit 2

FHA analysis and finds that these changes are based on physical differences in the units,

unit-specific operational differences, or updated information that has been accepted by the NRC

staff for use in dose consequence analyses. The WVA evaluation concluded that the

radiological consequences at the EAB, LPZ, and CR resulting from the postulated FHA inside

containment, with the containment closed and the RBPVS operating, comply with the reference

values and the CR dose criterion provided in 10 CFR 100.11 and 10 CFR Part 50, Appendix A,

GDC 19, as well as with the accident-specific dose guidelines specified in the SRP. The NRC

staff's review concluded that TVA used analyses, assumptions, and inputs consistent with those

described in the SRP. The NRC staff concluded that the assumptions presented in Table 15.5

and TVA's calculated dose results given in Table 15.1 of this SSER were acceptable. The NRC

staff concludes that TVA's estimates of the dose consequences of a design-basis FHA inside

containment with the containment isolated and the RBPVS operating are in accordance with the

SRP acceptance criteria.

15.4.5.2 Fuel-Handling Accident, Regulatory Guide 1.183, Alternative Source Term

Analysis

This accident analysis postulates that a spent fuel assembly is dropped during fuel handling and

strikes an adjacent assembly during the fall. All of the fuel rods in the dropped assembly are

conservatively assumed to experience fuel cladding damage, releasing the radionuclides within

the fuel rod gap to the fuel pool or reactor cavity water. The affected assembly is assumed to

contain the maximum inventory of fission products. Volatile constituents of the core fission

product inventory migrate from the fuel pellets to the gap between the pellets and the fuel rod

clad during normal power operations. The fission product inventory in the fuel rod gap of the

15-6
damaged fuel rods is assumed to be instantaneously released to the surrounding water as a

result of the accident. Fission products released from the damaged fuel are decontaminated by

passage through the overlaying water in the reactor cavity or SFP, depending on their physical

and chemical form.

TVA provided its analysis of the accident to the staff in a letter dated September 23, 2011. TVA

assumed no decontamination for noble gases, a decontamination factor (DF) of 200 for

radioiodines, and retention of all particulate fission products. As described in RG 1.183, the

FHA is analyzed based on the assumption that 100 percent of the fission products released

from the reactor cavity or SFP are released to the environment in 2 hours. TVA conservatively

did not credit filtration, holdup, or dilution of the released activity. Since the assumptions and

inputs are identical for the FHA within containment and the FHA outside containment, the

results of the two events are identical.

TVA considered the analysis of the FHA both within the containment and within the auxiliary

building. The dropped fuel assembly inside the containment is assumed to occur with the

equipment maintenance hatch fully open, and the fuel assembly drop inside the auxiliary

building credits no filtration of the exhaust. A minimum water level of 23 feet above the

damaged fuel assembly is maintained for release locations both inside containment and the

auxiliary building. This minimum water covering acts as a barrier to many of the radionuclides

released from the dropped assembly. TVA assumed retention of all noniodine particulates in

the pool, while the iodine releases from the fuel gap into the pool are assumed to be

decontaminated by an overall factor of 200. This DF results in 0.5 percent (i.e., 99.5 percent of

the iodine is retained in the pool) of the radioiodine escaping the overlying water with a

composition of 70-percent elemental and 30-percent organic iodine. In accordance with

Regulatory Position 3 of RG 1.183, the applicant assumed that 100 percent of the noble gas

activity is released from the pool. All fission products released to the environment are assumed

to release over a 2-hour period. In the subject RG 1.183 FHA analysis, TVA does not credit

dilution within the surrounding structures before release to the atmosphere. These assumptions

follow the guidance of RG 1.183 and are, therefore, acceptable to the staff.

15.4.5.3 Fuel-Handling Accident, Regulatory Guide 1.183 Source Term

WVA provided its analysis of the accident to the staff in a letter dated September 23, 2011. For

the purpose of this analysis, TVA assumed a conservative estimate of 100 hours of decay time

before any movement of fuel. The core fission product inventory that constitutes the source

term for this event is the gap activity in the 264 fuel rods assumed to be damaged as a result of

the postulated design-basis FHA. Volatile constituents of the core fission product inventory

migrate from the fuel pellets to the gap between the pellets and the fuel rod cladding during

normal power operations. The fission product inventory in the fuel rod gap of the damaged fuel

rods is assumed to be instantaneously released to the surrounding water as a result of the

accident, as in Regulatory Position 1.2 of RG 1.183.

For the FHA occurring inside containment, WVA assumed that the equipment maintenance

hatch is open at the time of the accident and that the release from the containment occurs with

no credit taken for containment isolation, no credit for dilution or mixing in the containment

atmosphere, and no credit for filtration of the released effluent. For the FHA occurring in the

auxiliary building, TVA also assumed no credit for filtration of the activity released from the SFP

water before it is released to the environment.

As corrected by item 8 of Regulatory Issue Summary 2006-04, "Experience with Implementation

15-7
of Alternative Source Terms" (ADAMS Accession No. ML053460347), RG 1.183, Appendix B,

Regulatory Position 2, should read as follows:

If the depth of water above the damaged fuel is 23 feet or greater, the

decontamination factors for the elemental and organic species are 285 and 1,

respectively, giving an overall effective decontamination factor of 200 (i.e., 99.5%

of the total iodine released from the damaged rods is retained by the water).

This difference in decontamination factors for elemental (99.85%) and organic

iodine (0.15%) species results in the iodine above the water being composed of

70% elemental and 30% organic species.

As noted previously, TVA maintains a minimum water depth of 23 feet to cover the underlying

damaged fuel assembly in both the reactor cavity and SFP for the FHA analyzed. The assumed

264 damaged fuel rods in the pool release 100 percent of their gap activity within the water,

which is scrubbed by the water column as it rises. This scrubbing decontaminates the iodine

gap releases with an overall DF of 200. This DF results in 0.5 percent (i.e., 99.5 percent of the

iodine is retained in the pool) of the radioiodine escaping the overlying water with a composition

of 70-percent elemental and 30-percent organic iodine. Additionally, 100 percent of the noble

gas gap activity is assumed to exit the pool, as in Regulatory Position 3 of RG 1.183.

TVA evaluated an FHA in the SFP area and in an open containment with no credit taken for the

ABGTS or containment purge system filters and concluded that the radiological consequences

at the EAB, outer boundary of the LPZ, and CR are within the reference values and the CR

dose criterion provided in 10 CFR 50.67, "Accident Source Term," as well as the

accident-specific dose guidelines specified in SRP Section 15.0.1. The staff's review concluded

that TVA used analyses, assumptions, and inputs consistent with the regulatory guidance

specified in RG 1.183 for the evaluation of an FHA using the AST. The staff concluded that the

assumptions by TVA presented in Table 15.6 of this SSER were acceptable. Table 15.1 of this

SSER gives TVA's calculated dose results. The staff concludes that the doses estimated by

TVA for the WBN Unit 2 FHA will meet the requirements of 10 CFR 50.67 and the guidelines of

RG 1.183 and are, therefore, acceptable.

15.4.6 Failure of Small Lines Carrying Coolant outside Containment

The NRC staff previously evaluated the failure of small lines carrying coolant outside

containment in Section 15.4.6 of the SER (NUREG-0847). There have been no supplements to

this section. The current WBN Unit 2 FSAR Section 15.5, "Environmental Consequences of

Accidents," does not include an evaluation of this accident. In the analysis reviewed in the SER,

the primary coolant maximum equilibrium fission product concentration was evaluated at

1.0 pCi/gm DEL The current WBN Unit 2 TS limit for the primary coolant maximum equilibrium

fission product concentration of DEI is 0.265 pCi/gm. Therefore, the analysis reviewed in the

SER would be bounding for WBN Unit 2.

15.4.7 Environmental Consequences of a Postulated Loss of AC Power to the Plant Auxiliaries

NUREG-0847 does not include an NRC staff review for the radiological dose consequences of a

postulated loss of ac power to the plant auxiliaries. For the completion of this SSER, the NRC

staff used the CLB assumptions for Unit 1 loss of ac power as a benchmark for acceptability of

the Unit 2 analysis and only addressed the following revised Unit 1 parameters and

assumptions that are different for Unit 2.

15-8
Source Term

TVA performed a conservative analysis of the potential offsite doses resulting from this accident

with steam generator leakage as a parameter. For the Unit 2 analysis, TVA replaced the Unit 1

analysis assumptions of 1-percent defective fuel and a realistic source term with the TS

allowable secondary coolant source term of 0.1 pCi/g DEL. In addition, the source terms used in

the dose consequence analyses performed for Unit 2 differ from the Unit 1 source terms,

because the Unit 1 source terms are based on a tritium core, whereas the Unit 2 source terms

are based on a standard core. The Unit 2 dose consequence analysis did not consider TPBARs

since TVA did not request permission to produce tritium in Unit 2.

Steam Release

The volume of steam released in the Unit 1 analysis is based on the use of the replacement

steam generators, whereas the volume of steam release in the Unit 2 analysis is based on the

use of the OSGs.

Dose Conversion Factors

For the Unit 2 analysis, TVA incorporated the DCFs from RG 1.109, Appendix E. The NRC staff

has endorsed the use of the DCFs from RG 1.109 in dose consequence analyses.

Atmospheric Dispersion Coefficients

The Unit 2 atmospheric dispersion coefficients are based on a release from the Unit 2 vent

valve stacks to the emergency CR intake.

The postulated accidents involving release of steam from the secondary system will not result in

a release of radioactivity unless there is leakage from the RCS to the secondary system in the

steam generator. TVA presented a conservative analysis of the potential offsite doses resulting

from this accident with steam generator leakage as a parameter. This conservative analysis

incorporates the TS limit of 0.1 pCi/gm DEI for the secondary coolant. In addition, TVA provided

an analysis using a realistic secondary-side source term based on American National Standards

Institute/American Nuclear Society (ANSIiANS)-18.1, "Radioactive Source Term for Normal

Operation for Light Water Reactors," issued 1984. TVA used DCFs in ICRP-30, "Limits for

Intakes of Radionuclides by Workers," 1979, to determine thyroid doses in place of those found

in ICRP-2. Table 15.7 of this SSER lists the parameters used in both the realistic and

conservative analyses.

Table 15.1 of this SSER shows TVA's calculated gamma and thyroid doses for the loss of ac

power to the plant auxiliaries at the EAB and LPZ for both the realistic and conservative

analyses. The doses for this accident are well within the limits of 10 CFR 100.11. Table 15.1

presents TVA's estimated whole body and thyroid dose to CR personnel from the radiation

sources discussed above. Table 15.9 of this SSER includes parameters for the CR analysis.

The whole body dose is below the 10 CFR Part 50, Appendix A, GDC 19 limit of 5 rem for CR

personnel, and the thyroid dose is below the 30-rem acceptance criteria stated in SRP

Section 6.4, "Control Room Habitability System," Revision 3, issued March 2007; therefore, they

are acceptable.

15-9
15.4.8 Environmental Consequences of a Postulated Waste Gas Decay Tank Rupture

TVA performed two separate analyses of the postulated waste gas decay tank rupture. The first

analysis used realistic assumptions, while the second analysis is based on the assumptions

found in RG 1.24, "Assumptions Used for Evaluating the Potential Radiological Consequences

of a Pressurized Water Reactor Gas Storage Tank Failure," dated March 23, 1972. Table 15.8

of this SSER lists the parameters used for each of these analyses.

The conservative analysis assumes that the reactor has been operating at full power with

1 percent of the fuel experiencing cladding defects. The realistic analysis assumes that the

source term is consistent with ANSI/ANS-1 8.1-1984 methodology.

Both analyses assume that the tank rupture occurs immediately upon completion of the waste

gas transfer, releasing the entire contents of the tank through the auxiliary building vent to the

outside atmosphere. The assumption of the release of the noble gas inventory from only a

single tank is based on the fact that all gas decay tanks will be isolated from each other

whenever they are in use.

Both TVA analyses use conservative assumptions to evaluate the doses from the released

activity. TVA used the DCFs in ICRP-30. Table 15.1 of this SSER includes TVA's whole body

gamma and thyroid doses for the gas decay tank rupture at the EAB and LPZ and to CR

personnel for both the realistic and RG 1.24 analyses. Table 15.9 of this SSER contains the

parameters for the CR analysis.

There are no revised Unit 2 parameters for this accident analysis because the waste gas

system is common to Units 1 and 2. Therefore, the licensing basis for the proposed Unit 2

analysis is the same as the licensing basis for Unit 1 regarding the environmental

consequences of a postulated waste gas decay tank rupture. Since the NRC staff has

approved the analysis for Unit 1, it is also acceptable for Unit 2.

15-10
Table 15.1 Radiological Consequences of Design-Basis Accidents

Postulated accident

Loss of coolant

EAB rem

WB1/Thyroid

TEDE

2

2.1 E+00/4.OE+01

LPZ rem

WB/Thyroid

TEDE

2.2E+00/1.4E+01

CR rem

WB/Thyroid

TEDE

1.1 E+00/3.8E+00

Steamline break outside secondary

containment

Preaccident iodine spike

DEI-131 at 14 pCi/gm

Accident-initiated iodine spike

DEI-131 at 0.265 pCi/gm

2.7E-02/2.4E+00 1.1 E-02/1.2E+00 4.3E-03/7.4E+OO

1.OE-01/3.1E+00 1.3E-01/4.8E+00 8.OE-03/11.OE+O1

Control rod ejection Bounded by the LOCA3

FHA

In auxiliary building or open

containment-RG 1.183

In reactor building, containment

closed-RG 1.25

Small line break outside containment

SGTR

Preaccident iodine spike

DEl-131 at 14 pCi/gm

Accident-initiated iodine spike

DEI-131 at 0.265 pCi/gm

Loss of ac power

Conservative analysis

Realistic analysis

Gas decay tank rupture

RG 1.24 analysis

Realistic analysis

2.4E+00 6.7E-01 1.OE+00

4.3E-01/4.2E+01 1.2E-01/1.2E+01 2.7E-01/6.8E+00

This accident is no longer part of FSAR Section 15.5.

3.8E-01/1.4E+01

5.5E-01/7.2E+00

7.5E-04/4.6E-02

1.8E-08/1.1 E-06

6.OE-01/1.3E-02

2.9E-02/1.2E-02

1.1 E-01/3.8E+00

1.6E-0112.1E+00

4.2E-04/2.6E-02

1.OE-08/6.2E-07

1.7E-01/3.6E-03

8.1 E-03/3.4E-03

6.2E-02/1.2E+01

5.7E-02/2.OE+00

2.1 E-04/2.1 E-02

5.1 E-09/5.OE-07

8.4E-01/7.OE-03

3.8E-02/6.5E-03

1

2

3

WB is defined as whole body gamma dose.

TEDE is defined as total effective dose equivalent.

Note that the WBN Unit 2 LOCA dose is less than 25 percent of the 10 CFR 100.11 limits and, therefore,

meets the accident dose criteria for a control rod ejection accident.

15-11
Table 15.2 (Page I of 2)

Assumptions Used To Calculate the Radiological Consequences

LOCA

Power level

Primary containment free volume

Shield building annulus free volume

Primary containment deck (air return) fan flow rate

Number of containment deck air return fans operating

Fractions of core inventory available for release

Noble gases

lodines

Initial iodine composition in containment

Elemental

Organic

Particulate

Primary containment leak rates

0-24 hr

1-30 days

Percent of primary containment leakage to auxiliary building

ABGTS filter efficiencies

elemental iodine

methyl iodine

particulate iodine

Delay time of activity in auxiliary building before ABGTS operation

Delay time before filtration credit is taken for the ABGTS

Mean holdup time in auxiliary building after initial 4 minutes

ABGTS flow rate

Leakage from auxiliary building to ABGTS downstream

HVAC (bypass of filters)

Leakage from ABGTS HVAC into auxiliary building

Leakage from auxiliary building into EGTS downstream

HVAC (bypass of filters)

Leakage from auxiliary building to environment from single

failure of ABGTS (from 30 minutes to 34 minutes post-LOCA)

Percent of primary containment leakage to annulus

Percent of annulus free volume available for mixing of

recirculated activity

Number of emergency gas treatment system air-handling

units operating

15-12

Following a Postulated

3565 MWt

1.27 E6 ft3

3.75 E5 ft3

40,000 cfm

1 of 2

100%

25%

91%

4%

5%

0.25% per day

0.125% per day

25%

99%

99%

99%

None

4 minutes

0.3 hours

9000 cfm

27.88 cfm

8.87 cfm

10.7 cfm

9900 cfm (for

4 minutes)

75%

50%

1 of 2
.Table 15.2 (Page 2 of 2)

Assumptions Used To Calculate the Radiological Consequences Following a Postulated

LOCA

Emergency gas treatment system filter efficiencies

elemental iodine 99%

methyl iodine 99%

particulate iodine 99%

Shield building mixing model 50% mixing

Ice Condenser Elemental and Particulate Iodine Removal Efficiency

Time Interval Post-LOCA (Hours) Removal Efficiency

0.0 to 0.156 0.96

0.156 to 0.267 0.76

0.267 to 0.323 0.73

0.323 to 0.489 0.71

0.489 to 0.615 0.60

0.615 to 0.768 0.58

0.768 to 0.824 0.40

0.824 to 720 0.0

LOCA Atmospheric Dispersion Factors (x/Q, s/m3

)

Time Period (hr) CR EAB

0-2

2-8

8-24

24-96

96-720

West CR Intake

Used in Dose

Analysis

1.09E-03

9.44E-04

1.56E-04

1.16E-04

9.59E-05

East CR Intake

Available by

Procedure

6.38E-04

LPZ

1.784E-04

8.835E-05

6.217E-05

2.900E-05

9.811E-06

1.26E-04

9.53E-05

8.07E-05

15-13
Table 15.3

Assumptions Used To Evaluate the Radiological Consequences Following a Postulated

MSLB Accident Outside Containment

Power level

Initial maximum RCS equilibrium activity

Accident-initiated iodine spike appearance rate

Maximum preaccident spike iodine concentration

Secondary coolant iodine activity

Primary-to-secondary leak rate

Faulted steam generator

Per intact steam generator

Steam generator secondary-side iodine partition coefficients

Faulted steam generator

Intact steam generator

RCS letdown flow rate

3565 MWt

0.265 pCi/g

500 times equilibrium rate

14.0 pCi/gm

0.1 pCi/gm DEI

1.0 gpm

150 gpd

1 (none)

100

124.39 gpm

Steam releases

Faulted steam generator (0-30 minutes)

Three intact steam generators (0-2 hr)

Three intact steam generators (2-8 hr)

96,100 Ibm

433,079 Ibm

870,754 Ibm

Time period (hr)

0-2

2-8

MSLB Atmospheric Dispersion Factors (x/Q, s/m3)

CR EAB

2.59E-03 6.38E-04

2.12E-03

LPZ

1.784E-04

8.835E-05

15-14
Table 15.4

Assumptions Used To Evaluate the Radiological Consequences Following a Postulated

SGTR Accident

Power level

Initial maximum RCS equilibrium activity

Accident-initiated iodine spike appearance rate

Maximum pre-accident spike iodine concentration

Secondary coolant iodine activity

Primary-to-secondary leak rate

Faulted steam generator

Per intact steam generator

Steam generator secondary-side iodine partition coefficients

Faulted steam generator

Intact steam generator

Secondary-side mass release (ruptured steam generator)

0-2 hours

2-8 hours

Secondary-side mass release (intact steam generator)

0-2 hours

2-8 hours

3565 MWt

0.265 pCi/g

500 times equilibrium rate

14.0 pCi/gm

ANSI/ANS-1 8.1-1984

Expected levels, 150 gpd/SG

1.0 gpm

150 gpd

1 (none)

100

103,300 Ibm

32,800 Ibm

492,100 Ibm

900,200 Ibm

Primary coolant mass release

Total

Flashed

191,400 Ibm

10,077.2 Ibm

Time period (hr)

0-2

2-8

SGTR Atmospheric Dispersion Factors (x/Q, s/m3)

CR EAB

2.59E-03 6.38E-04

2.12E-03

LPZ

1.784E-04

8.835E-05

15-15
Table 15.5

Assumptions Used To Evaluate the Radiological Consequences Following a Postulated

FHA Inside Closed Containment-RG 1.25 Analysis

Core thermal power level

Radial peaking factor

Number of fuel assemblies in the core

Fuel rods per assembly

Core average assembly power

Number of fuel assemblies damaged

Minimum postshutdown fuel-handling time (decay time)

Minimum pool water depth

Fuel clad damage gap release fractions

1-131

Remainder of halogens

Kr-85

Xe-1 33

Xe-135

Remainder of noble gases

Pool DF

Noble gases and organic .iodine

Aerosols

Elemental iodine (23 ft of water cover)

Overall iodine (23 ft of water cover)

3565 MWt

1.65

193

264

18.47 MWth

1 (all rods ruptured)

100 hours

23 feet

12%

10%

14%

5%

2%

10%

1

Infinite

133

200 (effective DF)

Chemical form of iodine released

Elemental

Organic

99.75%

0.25%

RBPVS

Elemental iodine

Organic iodine

Filter efficiencies auxiliary building

90%

30%

Duration of release to the environment 2-hour release

Time period (hr)

0-2

FHA Atmospheric Dispersion Factors (x/Q, s/m3)

CR EAB

2.59E-03 6.38E-04

LPZ

1.784E-04

15-16
Table 15.6

Assumptions Used To Evaluate the Radiological Consequences Following a Postulated

FHA in the Auxiliary Building or in Open Containment-RG 1.183 Analysis

Core thermal power level

Radial peaking factor

Number of fuel assemblies in the core

Fuel rods per assembly

Core average assembly power

Number of fuel assemblies damaged

Minimum postshutdown fuel-handling time (decay time)

Minimum pool water depth

Fuel clad damage gap release fractions

1-131

Remainder of halogens

Kr-85

Remainder of noble gases

3565 MWt

1.65

193

264

18.47 MWth

1 (all rods ruptured)

100 hours

23 feet

8%

5%

10%

5%

1

Infinite

285

200 (effective DF)

Pool DF

Noble gases and organic iodine

Aerosols

Elemental iodine (23 ft of water cover)

Overall iodine (23 ft of water cover)

Chemical form of iodine released

Elemental

Organic

Filter efficiencies

Duration of release to the environment

99.85%

0.15%

None

2-hour release

FHA Atmospheric Dispersion Factors (x/Q, s/m3)

Time period (hr)

0-2

CR

2.59E-03

EAB

6.38E-04

LPZ

1.784E-04

15-17
Table 15.7

Assumptions Used To Evaluate the Radiological Consequences Following a Postulated

Loss of AC Power to the Plant Auxiliaries

Core thermal power level 3565 MWt

Steam generator tube leak rate 1 gpm

Fuel defects (clad damage)

Realistic analysis

Conservative analysis

ANSI/ANS 18.1-1984

0.1 pCi/gm DEI

Iodine partition factor 0.01

Blowdown rate 25 gpm per steam generator

Duration of plant cooldown 8 hours

Steam release (total)

0-2 hours

2-8 hours

444,875 Ibm

903,530 Ibm

Loss of AC Power Atmospheric Dispersion Factors (x/Q, s/m3)

Time period (hr)

0-2

CR

2.59E-03

EAB

6.38E-04

LPZ

1.784E-04

15-18
Table 15.8

Assumptions Used To Evaluate the Radiological Consequences Following a Postulated

Waste Gas Decay Tank Rupture

Core thermal power level

Steam generator tube leak rate

3565 MWt

1 gpm

Fuel defects (clad damage)

Realistic analysis

Conservative analysis RG 1.24

Activity released from GWPS 4

Realistic analysis

Conservative analysis RG 1.24

Xe-1 31 m

Xe-1 33

Xe-1 33m

Xe-1 35

Kr-85

Kr-85m

1-131

Time of accident

Realistic analysis

Conservative analysis RG 1.24

ANSI/ANS-18.1-1984

1%

Maximum isotopic concentrations

and actual plant flow rates

Selected isotopes (Curies)

8.9E+02

6.8E+04

1.OE+03

9.4E+02

4.2E+03

1.3E+02

4.8E-02

After tank fill

End of equilibrium core cycle

Waste Gas Decay Tank Rupture Atmospheric Dispersion Factors (x/Q, s/m3)

Time period (hr)

0-2

CR

2.56E-03

EAB

6.38E-04

LPZ

1.784E-04

GWPS is defined as gaseous waste processing system.

15-19
Table 15.9

CR Parameters

Volume

Makeup/pressurization flow

Recirculation flow

Unfiltered intake

Filter efficiency

First pass

Second pass

Isolation time

Occupancy factors

0-24 hours

1-4 days

4-30 days

257,198 cu ft

711 cfm

2889 cfm

51 cfm

95%

70%

40 seconds

100%

60%

40%

15-20
APPENDIX A

CHRONOLOGY OF RADIOLOGICAL REVIEW OF

WATTS BAR NUCLEAR PLANT, UNIT 2, OPERATING LICENSE REVIEW

Public correspondence exchanged between the NRC and TVA during the review of the

operating license 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

IVA'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

(PARS) 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

(PDF).

The NRC PDR reference staff is available to assist with ADAMS. Contact information for the

PDR staff is on the NRC Web site at httD://www.nrc.aov/readina-rm/contact-Ddr.html.

A-1

APPENDIX E

PRINCIPAL CONTRIBUTORS TO SSER 25

D. Allsopp, NRR/DIRS/IOLB

R. Alvarado, NRRPDE/EICB

L. Brown, NRRJDRNAADB

N. Carte, NRR/DE/EICB

F. Lyon, NRRPDORLILPWB

P. Milano, NRRJDORL/LPWB

J. Parillo, NRR/DRAAADB

J. Poehler, NRRJDE/EVIB

J. Poole, NRR/DORL/LPWB

L. Raghavan, NRR/DORL/LPWB

E-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, and proposed license conditions that the staff has identified. Unless otherwise noted,

the item references are to sections of this 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 83 items still open and 41 items that have been closed as of November 7, 2011.

Open Items

Item Type Action Required Lead Status

(1) Cl Review evaluations and corrective actions NRR Open

associated with a power assisted cable pull. (NRC

safety evaluation dated August 31, 2009, ADAMS

Accession No. ML092151155)

(2) CI Conduct appropriate inspection activities to verify RII Open

cable lengths used in calculations and analysis

match as-installed configuration. (NRC safety

evaluation dated August 31, 2009, ADAMS

Accession No. ML092151155)

(4) Cl Conduct appropriate inspection activities to verify RII Open

that TVA's maximum SWBP criteria for signal level

and coaxial cables do not exceed the cable

manufacturers maximum SWBP criteria. (NRC

safety evaluation dated August 31, 2009, ADAMS

Accession No. ML092151155)

(5) Cl Verify timely submittal of pre-startup core map and NRR Open

perform technical review. (TVA letter dated

September 7, 2007, ADAMS Accession No.

ML072570676)

(6) CI Verify implementation of TSTF-449. (TVA letter NRR Open

dated September 7, 2007, ADAMS Accession No.

ML072570676)

(7) Cl Verify commitment completion and review electrical RII Open

design calculations. (TVA letter dated October 9,

1990, ADAMS Accession No. ML073551056)

(8) Cl TVA should provide a pre-startup map to the NRC NRR Open

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. ML1O01200035)

(9) Cl Confirm that education and experience of RII Open

management and principal supervisory positions

down through the shift supervisory level conform to

Regulatory Guide 1.8. (SSER 22, Section 13.1.3)

HH-1
(10) Cl Confirm that TVA has an adequate number of RII Open

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)

(12) TVA's implementation of NGDC PP-20 and EDCR NRR Open

Appendix J is subject to future NRC audit and

inspection. (SSER 22, Section 25.9)

(13) TVA is expected to submit an IST program and NRR Open

specific relief requests for WBN Unit 2 nine months

before the projected date of OL issuance. (SSER

22, Section 3.9.6)

(16) Based on the uniqueness of EQ, the NRC staff must RII/NRR Open

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)

(17) The NRC staff should verify the accuracy of the RII/NRR Open

WBN Unit 2 EQ list prior to fuel load. (SSER 22,

Section 3.11.2.1)

(23) Cl Resolve whether or not TVA's reasoning for not NRR Open

upgrading the MSIV 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)

(25) Prior to the issuance of an operating license, TVA is NRR Open

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)

(26) For the scenario with an accident in one unit and NRR Open

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)

(30) TVA should confirm that all other safety-related RII/NRR Open

equipment (in addition to the Class 1 E motors) will

have adequate starting and running voltage at the

most limiting safety related components (such as

motor operated valves, contactors, solenoid valves

or relays) at the degraded voltage relay setpoint

HH-2
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)

(32) TVA should provide to the NRC staff the details of NRR Open

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)

(33) Cl TVA stated in Attachment 9 of its letter dated July RII/NRR Open

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 Vdc vital

battery system. Verification of completion of these

DCNs to the NRC staff is necessary prior to

issuance of the operating license. (SSER 22,

Section 8.3.2.3; SSER 24, Section 8.1)

(35) TVA should provide information to the NRC staff NRR Open

that the CCS will produce feedwater purity in

accordance with BTP MTEB 5-3 or, alternatively,

provide justification for producing feedwater purity to

another acceptable standard. (SSER 22, Section

10.4.6)

(37) Cl The NRC staff will review the combined WBN Unit 1 NSIR Open

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 NP-REP

since Revision 92 do not affect the bases of the

staff's findings in this SER supplement. (SSER 22,

Section 13.3.2)

(38) Cl The NRC staff will confirm the availability and RII/NSIR Open

operability of the ERDS for Unit 2 prior to issuance

of the Unit 2 OL. (SSER 22, Section 13.3.2.6)

(39) Cl The NRC staff will confirm the adequacy of the RII/NSIR Open

communications capability to support dual unit

operations prior to issuance of the Unit 2 OL. (SSER

22, Section 13.3.2.6)

(40) Cl The NRC staff will confirm the adequacy of the RII/NSIR Open

emergency facilities and equipment to support dual

unit operations prior to issuance of the Unit 2 OL.

(SSER 22, Section 13.3.2.8)

(41) Cl TVA committed to (1) update plant data displays as RII/NSIR Open

necessary to include Unit 2, and (2) to update dose

assessment models to provide capabilities for

assessing releases from both WBN units. The NRC

HH-3
staff will confirm the adequacy of these items prior

to issuance of the Unit 2 OL. (SSER 22, Section

13.3.2.9).

(42) Cl The NRC staff will confirm the adequacy of the RII/NSIR Open

accident assessment capabilities to support dual

unit operations prior to issuance of the Unit 2 OL.

(SSER 22, Section 13.3.2.9)

(43) CI Section V of Appendix E to 10 CFR Part 50 requires NSIR Open

TVA to submit its detailed implementing procedures

for its emergency plan no less than 180 days before

the scheduled issuance of an operating license.

Completion of this requirement will be confirmed by

the NRC staff prior to the issuance of an operating

license. (SSER 22, Section 13.3.2.18)

(47) The NRC staff noted that TVA's changes to NRR Open

Section 6.2.6 in FSAR Amendment 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)

(48) Cl The NRC staff should verify that its conclusions in NRR Open

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)

(49) Cl The NRC staff was unable to determine how WVA RII Open

linked the training qualification requirements of

ANSI N45.2-1971 to WVA Procedure 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)

(50) C1 TVA stated that about 5 percent of the anchor bolts RII Open

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)

HH-4
(51) Cl The implementation of TVA Procedure TI-1 19 will RII Open

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)

(52) Not used.

through

(58)

(59) The staff's evaluation of the compatibility of the ESF NRR Open

system materials with containment sprays and core

cooling water in the event of a LOCA is incomplete

pending resolution of GSI-1 91 for WBN Unit 2.

(SSER 23, Section 6.1.1.4)

(60) CI TVA should amend the FSAR description of the NRR Open

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)

(61) WVA should provide information to the NRC staff to NRR Open

demonstrate that 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)

(63) Cl TVA should confirm to the NRC staff that testing RII Open

prior to Unit 2 fuel load has demonstrated that twoway communications is impossible with the Eagle

21 communications interface. (SSER 23, Section

7.2.1.1)

(64) CI TVA stated that, "Post modification testing will be RII Open

performed to verify that the design change corrects

the Eagle 21, Rack 2 RTD 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)

(65) TVA should provide justification to the staff NRR Open

regarding why different revisions of WCAP-1 3869

are referenced in WBN Unit 1 and Unit 2. (SSER

23, Section 7.2.1.1)

(66) CI TVA should clarify FSAR Section 9.2.5 to add the NRR Open

capability of the UHS to bring the nonaccident unit

to cold shutdown within 72 hours. (SSER 23,

Section 9.2.5)

(67) Cl TVA should confirm, and the NRC staff should RII Open

verify, that the component cooling booster pumps

for Unit 2 are above PMF level. (SSER 23, Section

9.2.2)

HH-5
(68) Not used.

(69) Cl The WBN Unit 2 RCS vent system is acceptable, RII Open

pending verification that the RCS vent system is

installed. (SSER 23, Section 5.4.5)

(70) TVA should provide the revised WBN Unit 2 PSI NRR Open

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)

(71) By letter dated April 21, 2011 (ADAMS Accession NRR Open

No. ML1 11110513), TVA withdrew its commitment

to replace the Unit 2 clevis insert bolts. WVA should

provide further justification for the decision to not

replace the bolts to the NRC staff. (SSER 23,

Section 3.9.5)

(73) CI The NRC staff will inspect to confirm that TVA has RII Open

completed the WBN Unit 2 EOPs prior to fuel load.

(SSER 23, Section 7.5.3)

(74) Cl The NRC staff will verify installation of the acoustic- RII Open

monitoring system for the power-operated relief

valve (PORV) position indication in WBN Unit 2

before fuel load. (SSER 23, Section 7.8.1)

(75) CI The NRC staff will verify that the test procedures RII Open

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)

(77) It is unclear to the NRC staff which software V&V NRR Open

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)

(79) TVA should perform a radiated susceptibility survey, NRR Open

after the installation of the hardware but prior to the

RM-1 000 being placed in service, to establish the

need for exclusion distance for the 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. (SSER 23, Section 7.5.2.3)

(80) TVA should provide clarification to the staff on how NRR Open

TVA Standard Specification SS-E18-14.1 meets the

guidance of RG 1.180, and should address any

deviations from the guidance of the RG. (SSER 23,

Section 7.5.2.3)

HH-6
(81) The extent to which TVA's supplier, General NRR Open

Atomics (GA), complies with EPRI TR-106439 and

the methods that GA used for its commercial

dedication process should be provided by TVA to

the NRC staff for review. (SSER 23, Section

7.5.2.3)

(83) Cl TVA should confirm to the NRC staff the completion NRR Open

of the data storm test on the DCS. (SSER 23,

Section 7.7.1.4)

(84) Not used.

through

(89)

(90) Cl The NRC staff should verify that the ERCW dual RII/NRR Open

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 requirements

for two-unit operation. (SSER 23, Section 9.2.1)

(91) TVA should update the FSAR with information NRR Open

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)

(92) Not used.

(93) TVA should confirm to the staff that testing of the RII Open

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)

(94) TVA should provide to the staff either information NRR Open

that demonstrates that the WBN Unit 2 Common Q

PAMS meets the applicable requirements in 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)

(98) TVA should demonstrate that the WBN Unit 2 NRR Open

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)

(101) TVA should demonstrate that the WBN Unit 2 NRR Open

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)

HH-7
(105) TVA should produce an acceptable description of NRR Open

how the WBN Unit 2 Common Q PAMS SysRS and

SRS implement the design basis requirements of

IEEE Std. 603-1991 Clause 4. (SSER 23, Section

7.5.2.2.3.4.3.1)

(108) TVA should demonstrate to the NRC staff that there NRR Open

are no synergistic effects between temperature and

humidity for the Common Q PAMS equipment.

(SSER 23, Section 7.5.2.2.3.5.2)

(110) TVA should provide information to the NRC staff NRR Open

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)

(111) TVA should confirm to the staff that there are no NRR Open

changes required to the technical specifications as a

result of the modification installing the Common Q

PAMS. If any changes to the technical

specifications are required, TVA should provide the

changes to the NRC staff for review. (SSER 23,

Section 7.5.2.2.3.11)

(112) Cl TVA should provide an update to the FSAR NRR Open

reflecting the radiation protection design features

descriptive information provided in its letter dated

October 4, 2010. (SSER 24, Section 12.4)

(113) Cl TVA should provide an update to the FSAR NRR Open

reflecting the justification for the periodicity of the

COT frequency for WBN non-safety related area

radiation monitors. (SSER 24, Section 12.4)

(114) Cl TVA should update the FSAR to reflect that WBN NRR Open

meets the radiation monitoring requirements of 10

CFR 50.68. (SSER 24, Section 12.4)

(115) Cl WVA should update the FSAR to reflect the NRR Open

information regarding design changes to be

implemented to lower radiation levels as provided in

its letter the NRC dated June 3, 2010. (SSER 24,

Section 12.5)

(116) Cl TVA should update the FSAR to reflect the NRR Open

qualification standards of the RPM as provided in its

letter to the NRC dated October 4, 2010. (SSER

24, Section 12.6)

(117) Cl TVA should update the FSAR to reflect the NRR Open

calculational basis for access to vital areas as

provided in its letter dated February 25, 2011.

(SSER 24, Section 12.7.1)

(118) TVA should provide to the NRC staff a description of NRR Open

how the other vanadium detectors within the IITA

would be operable following the failure of an SPND.

(SSER 24, Section 7.7.1.9.2)

HH-8
(120) TVA must confirm to the NRC staff that the NRR Open

maximum over-voltage or surge voltage that could

affect the system is 264 VAC, assuming that the

power supply cable to the 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)

(121) TVA should submit the results to the NRC staff of a NRR Open

600 VDC dielectric strength test performed on the

IITA assembly. (SSER 24, Section 7.7.1.9.5)

(123) TVA should provide an explanation to the NRC staff NRR Open

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)

(125) TVA should provide clarification to the NRC staff of NRR Open

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)

(126) To enable the NRC staff to evaluate and review the NRR Open

IITA environmental qualification, TVA should

provide the summary report of the environmental

qualification for the IITA. (SSER 24, Section

7.7.1.9.5)

(127) TVA should provide a summary to the NRC staff of NRR Open

the electro-magnetic interference/radio-frequency

interference (EMI/RFI) testing for the MI cable

electro-magnetic compatibility (EMC) qualification

test results. (SSER 24, Section 7.7.1.9.5)

(129) TVA should verify to the NRC staff resolution of the NRR Open

open item in WNA-CN-001 57-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)

(131) TVA should review the EOP action level setpoint to NRR Open

account for the difference between core exit

temperature readings for Unit 1 and Unit 2 and

confirm the EOP action level setpoint to the NRC

staff. (SSER 24, Section 7.7.1.9.5)

(132) TVA must provide the NRC staff with analyses of NRR Open

the boron dilution event that meet the criteria of

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)

(133) In order to confirm the stability analysis of the sand NRR Open

baskets used by TVA in the WBN Unit 2 licensing

basis, TVA will perform either a hydrology analysis

HH-9
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)

(134) TVA should provide to the NRC staff supporting NRR Open

technical justification for the statements in

Amendment 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 OBE (0.09 g). Postulation of Tellico failure in

this combination has not been evaluated but is

bounded by the SSE failure of Norris, Cherokee,

Douglas and Tellico." (SSER 24, Section 2.4.10)

(136) Cl The JFD summary for the data from 1991 through NRR Open

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 updated FSAR, the NRC staff will confirm that

these updates have been made by TVA. (SSER 25,

Section 2.3.3)

(137) Cl The NRC staff will confirm, upon receipt, that TVA NRR Open

integrated the updated CR x/Q values from its letter

dated September 15, 2011, into a future

amendment of the FSAR. (SSER 25, Section 2.3.4)

(138) Cl Upon receipt of the updated ODCM, the NRC staff NRR Open

will confirm that corresponding revisions related to

the updated annual average x/Q and D/Q values

have been made to the ODCM. (SSER 25, Section

2.3.5)

(139) C1 The results of the cost-benefit analysis required by NRR Open

10 CFR Part 50, Appendix I, subsection lI.D, should

be provided in the WBN Unit 2 FSAR. Upon receipt

of the updated FSAR, the NRC staff will confirm that

the update has been made by TVA. (SSER 25,

Section 11.3)

Closed Items

(3) Cl Confirm TVA submitted update to FSAR section NRR Closed

8.3.1.4.1. (NRC safety evaluation dated August 31,

2009, ADAMS Accession No. ML092151155)

Closed in SSER 24, Section 8.1.

HH-10
(11) Cl The plant administrative procedures should clearly NRR Closed

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

SRO, 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.

(14) TVA stated that the Unit 2 PTLR is included in the NRR Closed

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.

(15) TVA should confirm to the NRC staff the completion NRR Closed

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.

(18) Based on the extensive layup period of equipment RII/NRR Closed

within WBN Unit 2, the NRC staff must review, prior

to fuel load, the assumptions used by WVA to reestablish a baseline for the qualified life of

equipment. The purpose of the staffs 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. ML1 11810890.

(19) The NRC staff should complete its review of TVA's RII/NRR Closed

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.

(20) Cl Resolve whether or not routine maintenance RII/NRR Closed

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.

(21) The NRC staff should confirm that the Electrical RII/NRR Closed

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,

HH-11
2011, ADAMS Accession No. ML1 12730197.

(22) TVA must clarify its use of the term "equivalent" NRR Closed

(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.

(24) The NRC staff requires supporting documentation NRR Closed

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.

(27) TVA should provide a summary of margin studies NRR Closed

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.

(28) TVA should provide to the NRC staff a detailed NRR Closed

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.

(29) TVA should provide information about the operating NRR Closed

characteristics of the offsite power supply at the

Watts Bar Hydro Plant (for dual-unit operation),

including the operating voltage range,

postcontingency 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.

(31) TVA should clarify the loading sequence as NRR Closed

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 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

HH-12
(34) Cl TVA stated that the method of compliance with NRR Closed

Phase I guidelines would be substantially similar to

the current Unit I 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.

(36) TVA should provide information to the NRC staff to NRR Closed

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.

(44) TVA should provide additional information to clarify NRR Closed

how the initial and irradiated RTNDT was

determined. (SSER 22, Section 5.3.1) Closed in

SSER 25, Section 5.3.1.

(45) Cl TVA stated in its response to RAI 5.3.2-2, dated NRR Closed

July 31, 2010, that the PTLR would be revised to

incorporate the COMS arming temperature. (SSER

22, Section 5.3.2) Closed in SSER 25, Section

5.3.2.

(46) Cl The LTOP lift settings were not included in the NRR Closed

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.

(62) Cl Confirm TVA's change to FSAR Section 10.4.9 to NRR Closed

reflect its intention to operate with each CST

isolated from the other. (SSER 23, Section 10.4.9)

Closed in SSER 24, Section 10.4.9.

(72) The NRC staff should complete its review and NRR Closed

evaluation of the additional information provided by

TVA regarding the ICC instrumentation. (SSER 23,

Section 4.4.8) Closed in SSER 25, Section 7.5.2.2.

(76) Cl The NRC staff will verify that the derivative time RII Closed

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.

(78) TVA intends to issue a revised calculation reflecting NRR Closed

that the TID in the control room is less than Ix103

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.

HH-13
(82) The staff concluded that the information provided by NRR Closed

TVA pertaining to the in-containment LPMS

equipment qualification for vibration was incomplete.

TVA should provide (item number 362 of ADAMS

Accession No. ML1 11050009), documentation that

demonstrates the LPMS 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.

(95) TVA should update FSAR Table 7.1-1, "Watts Bar NRR Closed

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.

(96) TVA should (1) update FSAR Table 7.1-1 to include NRR Closed

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.

(97) WVA should demonstrate that the WBN Unit 2 NRR Closed

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.

(99) TVA should update FSAR Table 7.1-1 to reference NRR Closed

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.

(100) TVA should update FSAR Table 7.1-1 to reference NRR Closed

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.

(102) TVA should update FSAR Table 7.1-1 to reference NRR Closed

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.

(103) TVA should demonstrate that the WBN Unit 2 NRR Closed

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.

(104) Cl The NRC staff will review the WEC self assessment NRR Closed

to verify that it the WBN Unit 2 PAMS is compliant to

the V&V requirements in the SPM or that deviations

HH-14
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.

(106) TVA should produce a final WBN Unit 2 Common Q NRR Closed

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.

(107) Cl TVA should provide to the NRC staff documentation NRR Closed

to confirm that the final WBN Unit 2 Common Q

PAMS SDDs that are independently reviewed.

(SSER 23, Section 7.5.2.2.3.4.3.2) Closed in SSER

25, Section 7.5.2.2.

(109) TVA should demonstrate to the NRC staff NRR Closed

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.

(119) TVA should submit WNA-CN-001 57-WBT, Revision NRR Closed

0, to the NRC by letter. The NRC staff should

confirm by review of WNA-CN-001 57-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.

(122) TVA should confirm to the NRC staff that different NRR Closed

divisions of safety power are supplied to the IIS 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.

(124) While the BEACON datalink on the Application NRR Closed

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.

(128) TVA should submit the seismic qualification test NRR Closed

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.

(130) TVA should provide a summary to the NRC staff of NRR Closed

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.

(135) TVA has not provided the analysis required by 10 NRR Closed

CFR Part 50, Appendix I, subsection lI.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,

HH-15
Section 11.3) Closed in SSER 25, Section 11.3.

Cl - Confirmatory Issue

HH-16
NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER

(12-2010) (Assigned by NRC, Add Vol.. Supp., Rev.,

NRCMD 3.7 and Addendum Numbers, if any.)

BIBLIOGRAPHIC DATA SHEET NUREG-0847

(See instructions on the reverse) Supplement No. 25

2. TITLE AND SUBTITLE 3. DATE REPORT PUBLISHED

Safety Evaluation Report MONTH YEAR

Related to the Operation of

Watts Bar Nuclear Plant Unit 2 December 2011

Docket No. 50-391 4. FIN OR GRANT NUMBER

5. AUTHOR(S) 6. TYPE OF REPORT

J. Poole, et al. Technical

7. PERIOD COVERED (Inclusive Dates)

8. PERFORMING ORGANIZATION - NAME AND ADDRESS (If NRC, provide Division, Office or Region. U.S. Nuclear Regulatory Commission, and mailing address; if contractor,

provide name and mailing address.)

U.S. Nuclear Regulatory Commission

Office of Nuclear Reactor Regulation

Division of Operating Reactor Licensing

Washington, DC 20555-0001

9. SPONSORING ORGANIZATION - NAME AND ADDRESS (If NRC, type 'Same as above,; if contractor, provide NRC Division, Office or Region, U.S. Nuclear Regulatory Commission,

and mailing address.)

Same as above.

10. SUPPLEMENTARY NOTES

Docket No. 50-391

11. ABSTRACT (200 words or less)

This report supplements the safety evaluation report (SER), NUREG-0847 (June 19 82), Supplement No. 24 (September 2011;

Agencywide Documents Access and Management System (ADAMS) Accession No. ML1 1277 A148), with respect to the

application filed by the Tennessee Valley Authority (TVA), as applicant and own er, for a license to operate Watts Bar Nuclear

Plant (WBN) Unit 2 (Docket No. 50-391).

In its SER and supplemental SER (SSER) Nos. 1 through 20 issued by the U.S Nud ear Regulatory Commission (NRC) staff,

the NRC staff documented its safety evaluation and determination that WBN Unit I met all applicable regulatory requirements.

Based on its evaluation and satisfactory inspection findings, the NRC issued a full-power operating license for WBN Unit 1 on

February 7, 1996.

In SSERs subsequent to SSER 20, the staff addressed TVA's application for a lic ense to operate WBN Unit 2, and provided

information regarding the status of items remaining to be resolved, which were open at the time that TVA deferred construction

of WBN Unit 2, and which were not evaluated and resolved as part of licensing f or WBN Unit 1. In this and future SSERs, the

staff will document its evalution and closure of open items in its review of TV A's application for an operating license for WBN

Unit 2.

12. KEY WORDS/DESCRIPTORS (List words or phrases that will assist researchers in locating the report.) 13. AVAILABILITY STATEMENT

Safety Evaluation Report (SER) unlimited

Watts Bar Nuclear Plant 14. SECURITY CLASSIFICATION

Docket No. 50-391 (This Page)

unclassified

(This Report)

unclassified

15. NUMBER OF PAGES

16. PRICE

NRC FORM 335 (12-2010)
Fe' I fmycH lnPa ge.

NUREG-0847

Supplement 25

Safety Evaluation Report Related to the Operation of

Watts Bar Nuclear Plant, Unit 2

December 2011

UNITED STATES

NUCLEAR REGULATORY COMMISSION

WASHINGTON, DC 20555-0001

OFFICIAL BUSINESS