NUREG-0847

Supplement No. 2

Safety Evaluation Report

related to the operation of

Watts Bar Nuclear Plant,

Units 1 and 2

Docket Nos. 50-390 and 50-391

Tennessee Valley Authority

U.S. Nuclear Regulatory

Commission

Office of Nuclear Reactor Regulation

January 1984
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NUREG-0847

Supplement No. 2

Safety Evaluation Report

related to the operation of

Watts Bar Nuclear Plant,

Units 1 and 2

Docket Nos. 50-390 and 50-391

Tennessee Valley Authority

U.S. Nuclear Regulatory

Commission

Office of Nuclear Reactor Regulation

January 1984

ABSTRACT

This report supplements the Safety Evaluation Report, NUREG-0847 (June 1982) and

Supplement No. 1 (September 1982), issued by the Office of Nuclear Reactor Regu- lation of the U.S. Nuclear Regulatory Commission with respect to the application filed by the Tennessee Valley Authority, as applicant and owner, for licenses

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

50-391). The facility is located in Rhea County, Tennessee, near the Watts Bar

Dam on the Tennessee River. This supplement provides recent information regarding resolution of some of the open and confirmatory items and license conditions identified in the Safety Evaluation Report.

Watts Bar SSER 2 ii i

TABLE OF CONTENTS

Page

ABSTRACT ........................................................

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

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

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

1.8 Confirmatory Issues .......................................... 1-3

1.9 License Conditions ........................................... 1-5

3 DESIGN CRITERIA - STRUCTURE, COMPONENTS, EQUIPMENT, AND SYSTEMS ... 3-1

3.5 Missile Protection ........................................... 3-1

3.5.2 Structures, Systems, and Components To Be Protected

From Externally Generated Missiles .................... 3-1

4 REACTOR ............................................................ 4-1

4.2 Fuel System Design .......................................... 4-1

4.2.2 Thermal Performance ............................... .... 4-1

4.2.3 Mechanical Performance ................................ 4-2

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

5.2 Integrity of Reactor Coolant Pressure Boundary ................ 5-1

5.2.2 Overpressure Protection ............................... 5-1

5.4 Component and Subsystem Design ............................... 5-1

5.4.3 Residual Heat Removal System .......................... 5-1

5.4.5 Reactor Coolant System Vents (II.B.1) .................. 5-2

6 ENGINEERED SAFETY FEATURES ........................................ 6-1

6.3 Emergency Core Cooling System ................................ 6-1

6.3.3 Testing ............................................. 6-1

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

7.2 Reactor Trip System ......................................... 7-1

7.2.5 Steam Generator Water Level Trip....................... 7-1

Watts Bar SSER 2 V
TABLE OF CONTENTS (Continued)

Page

7.3 Engineered Safety Features Actuation System................... 7-1

7.3.2 Containment Sump Level Measurement ..................... 7-1

8 ELECTRIC POWER SYSTEMS ............................................ 8-1

8.2 Offsite Electric Power System ................................ 8-1

8.2.2 Compliance With GDC 17 ................................ 8-1

8.3 Onsite Power Systems ........................................ 8-1

8.3.1 Onsite AC Power System Compliance With GDC 17 .......... 8-1

8.3.2 Onsite DC System Compliance With GDC 17 ............... 8-2

8.3.3 Common Electrical Features and Requirements ........... 8-2

17 QUALITY ASSURANCE .................................................. 17-1

APPENDICES

A CHRONOLOGY OF RADIOLOGICAL REVIEW OF WATTS BAR NUCLEAR PLANT,

UNITS 1 AND 2, OPERATING LICENSE REVIEW

B BIBLIOGRAPHY

E PRINCIPAL CONTRIBUTORS

G ERRATA TO WATTS BAR SAFETY EVALUATION REPORT

Watts Bar SSER 2 vi
1 INTRODUCTION AND DISCUSSION

1.1 Introduction

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

issued a Safety Evaluation Report, NUREG-0847, regarding the application by

the Tennessee Valley Authority (the applicant or TVA) for-licenses to operate

the Watts Bar Nuclear Plant, Units 1'and 2. The Safety Evaluation Report

(SER) was supplemented by Supplement No. 1 (September 1982), which discussed

the status of some outstanding issues. in-further support of the licensing

activities and addressed the recommendations of the Advisory Committee on

Reactor Safeguards (ACRS).

This supplement provides more recent information regarding resolution of some

of the open and confirmatory items and license conditions identified in the

SER and its supplement. Another supplement to the SER will be issued before

fuel loading of Unit 1 to discuss the resolution of the other open and confirmatory items and license conditions identified in-the SER.

Each of the following sections or appendices of this supplement is numbered

the same as the section or appendix of the SER that is being updated, and the

discussions are supplementary to and not in-lieu of the discussion in the SER

unless otherwise noted. Accordingly, Appendix A is a continuation of the

chronology of the safety review. Appendix B is an updated bibliography.*

Appendix E is a list of principal contributors to this supplement. Appendix G

is a list of errata for the SER. No changes in SER Appendices C, D, and F

have been made by this supplement.

The Project Manager is Thomas J. Kenyon. Mr Kenyon may be contacted by calling

(301) 492-7266, or by writing to the following address:

Mr. Thomas J. Kenyon

Division of Licensing

U.S. Nuclear Regulatory Commission

Washington, D.C. 20555

1.7- Summary of Outstanding Issues

SER Section 1.7 identified 17 outstanding issues that had not been resolved at

the time the SER was issued. This supplement updates the status of four of

those items. The current status of each of the 17 original issues is tabulated

below. For those items discussed in this supplement, the relevant section of

this document is indicated. Resolution of those issues that are, to date,

unresolved will be addressed in future supplements.

*Avaliability of all material cited is described on the inside front-cover of

this report.

Watts Bar SSER 2 1-1
Issue

(1) Potential for liquefaction beneath

ERCW pipelines and Class 1E electrical conduit

(2) Buckling loads on Class 2 and 3

supports

(3) Preservice and inservice pump and

valve test program

(4) Seismic and environmental qualification of equipment

(5) Preservice and inservice inspection

program

(6) Pressure-temperature limits for

Unit 2

(7) Model D-3 steam generator preheater

tube degradation

(8) BTP CSB 6-4 and containment isolation

dependability (II.E.4.2)

(9) H2 analysis review

(10) Safety valve sizing analysis

(WCAP-7769)

(11) Compliance of proposed design change

to the offsite power system to GDC-17

and 18

(12) Fire Protection Program

(13) Quality classification of diesel

generator auxiliary system piping

and components

(14) Diesel generator auxiliary system

design deficiencies

(15) Physical Security Plan

(16) Boron Dilution Event

Status

Awaiting information

Under review

Under review

Seismic - under review

Environmental-awaiting

information

Under review

Awaiting information

Under review

Under review

Awaiting information

Resolved

Partially resolved

Awaiting information

Under review

Under review

Resolved in SSER 1*

Under review

Section

5.2.2

8.2.2.2

(17) Q List Resolved 17

*TVA has recently submitted a revised Physical Security Plan. However, the plan

approved in SSER 1 is acceptable for use pending, approval of the new plan.

Watts Bar SSER 2 1-2
1.8 Confirmatory Issues

SER Section 1.8 identified 42 confirmatory issues for which additional information and documentation were required to confirm preliminary conclusions. This

supplement updates 15 of those items for which the confirmatory information has

subsequently been provided by the applicant and for which review has been

completed by the staff. The current status of each of the original issues is

tabulated below. For those items discussed in this supplement, the relevant

section of this supplement is noted. Resolution of issues that are outstanding,

to date, will be addressed in future supplements.

Issue

(1) Design basis ground water level for

the ERCW pipeline

(2) Material and geometric damping effect

in SSI analysis

(3) Analysis of sheetpile walls

(4) Design differential settlement of

piping and electrical components

between rock-supported structures

(5) Upgrading ERCW system to seismic

Category I

(6) Seismic classification of structures,

systems, and components important to

safety

(7) Tornado missile protection of diesel

generator exhaust

(8) Steel containment building buckling

research program

(9) Pipe support baseplate flexibility

and its effects on anchor bolt

loads (IE Bulletin 79-02)

(10) Thermal performance analysis

(11) Cladding collapse

(12) Fuel rod bowing evaluation

(13) Loose-parts monitoring system

(14) Installation of residual heat

removal flow alarm

(15) Natural circulation tests

Status

Under review

Under review

Under review.

Under review

Under review

Awaiting information

Resolved

Awaiting information

Section

3.5.2

Under review

Resolved

Resolved

Resolved

4.2.2

4.2. 2

4.2.3

Awaiting information

Awaiting verification

of installation

Awaiting information

Watts Bar SSER 2 1-3
Issue

(16) Dump Valve testing

(17) Protection against damage to containment from external pressure

(18) Designation of containment isolation

valves for main and auxiliary feedwater lines and feedwater bypass

lines

(19) Compliance with GDC-51

(20) Isolation survey (sump debris)

(21) Safety system set point methodology

(22) Steam generator water level reference

leg

(23) Containment sump level measurement

(24) IE Bulletin 80-06

(25) Overpressure protection during lowtemperature operation

(26) Availability of offsite circuits

(27) Nonsafety loads powered from the

Class 1E ac distribution system

(28) Low and/or degraded grid voltage

condition

(29) Diesel generator reliability qualification testing

(30) Diesel generator battery system

(31) Thermal overload protective bypass

(32) Sharing of dc and ac distribution

systems and power supplied between

Units 1 and 2

(33) Sharing of raceway systems between

units

(34) Testing Class 1E power systems

(35) Evaluation of penetrations capability

to withstand failure of overcurrent

protection device

Status

Resolved

Under review

Awaiting information

Section

5.4.3

Awaiting

Resolved

Awaiting

Resolved

Resolved

Awaiting

Awaiting

information

information

information

information

6.3.3

7.2.5

7.3.2

8.2.2.1.

8.3.1.1

8.3.1.2

Resolved

Resolved

Awaiting verification

of test results

Under review

Resolved

Resolved

Awaiting information

Resolved

Resolved

Awaiting information

8.3.2.4

8.3.3.1.2

8.3.3.2.3

8.3.3.5.2

Watts Bar SSER 2 1-4
Issue

(36) Missile protection for diesel

generator vent line

(37) Component booster pump relocation

Status

Awaiting verification

of modifications

Awaiting verification

of modifications

Section

(38) Electrical penetrations documentation Under review

(39) Compliance with NUREG/CR-0660

(40) No-load, low-load, and testing operations for diesel generator

(41) Initial test program

(42) Submergence of electrical equipment

as result of a LOCA

See License Condition

(22)

Awaiting verification

of procedure changes

Awaiting information

Awaiting information

1.9 License Conditions

In Section 1.9 of the SER and its-supplement, the staff identified 38 license

conditions. Since these documents were issued, the applicant has submitted

additional information on some of these items, thereby removing the necessity

to impose a condition. The license conditions are tabulated below, with the

corresponding NUREG-0737 item number given in parentheses and the relevant

section of this report noted for the updated status.

Condition Section

(1) Relief and safety valve testing (II.D.1)

(2) Preservice/inservice testing of pumps and valves --

(3) Detectors for inadequate core cooling (II.F.2) --

(4) Inservice Inspection Program

(5) Installation of reactor coolant vents (II.B.1) 5.4.5

(6) Accident monitoring instrumentation (II.F.1)

(a)

(b)

(c)

(d)

(e)

(f)

noble gas monitor

iodine particulate sampling

high range incontainment radiation monitor

containment pressure

containment water level

containment hydrogen

(7)

(8)

(9)

(10)

(11)

(12)

Modification to chemical feedlines

Containment isolation dependability (II.E.4.2)

Hydrogen control measures (II.B.7)

Status monitoring system

Installation of acoustic monitoring system (II.D.3)

Diesel generator reliability qualification testing at

normal temperatures

8.3.1.6

Watts Bar SSER 2 1-5
Condition

(13) DC monitoring and annunciation

(14) Possible sharing of dc control power to ac switchgear

(15) Testing of associated circuits

(16) Testing of non-Class 1E cables

(17) Low-temperature overpressure protection/power supplies for

pressurizer relief valves and level indicators (I.G.1)

(18) Testing of reactor coolant pump breakers

(19) Postaccident sampling system (II.B.3)

(20) Fire Protection Program

(21) Performance testing for communications systems

(22) Diesel generator reliability

(23) Secondary water chemistry monitoring and control program-

(24) Primary coolant outside containment (III.D.1.1)

(25) Independent safety engineering group (I.B.1.2)

(26) Use of experienced personnel during startup

(27) Emergency preparedness (III.A.1.1, III.A.1.2, III.A.2)

(28) Review of power ascension test procedures and emergency

operating procedures by NSSS-Vendor (I.C.7)

(29) Modifications to emergency operating instructions (I.C.8)

(30) Report on outage of emergency core cooling system (II.K.3.1

(31) Initial test program

(32) Effect of high pressure injection for small-break LOCA with

no auxiliary feedwater (II.K.2.13)

(33) Voiding in the reactor coolant system (II.K.2.17)

(34) PORV isolation system (II.K.3.1, II.K.3.2)

(35) Automatic trip of the reactor coolant pumps during a smallbreak LOCA (II.K.3.5)

(36) Revised small-break LOCA analysis (II.K.3.30, II.K.3.31)

(37) Control room design review (I.D.1)

(38) Physical Security Plan

Section

8.3.3.6

7)

Watts Bar SSER 2 1-6
3 DESIGN CRITERIA - STRUCTURE, COMPONENTS, EQUIPMENT, AND SYSTEMS

3.5 Missile Protection

3.5.2 Structures, Systems, and Components To Be Protected From Externally

Generated Missiles

As discussed in the SER, the diesel generator exhaust stacks, which protrude

approximately 2 ft above the roof grade, may be susceptible to damage or incapacitating flow blockage as a result of tornado-missile impact. By letter dated

November 24, 1982, the applicant submitted the details of proposed design modifications, which consist of installation of a reinforced concrete curb around

the diesel exhaust stacks, to prevent such damage. The staff has reviewed this

information and has concluded that the proposed design is acceptable. The

applicant has committed to completing the modifications to the exhaust stacks

before fuel loading; the staff, finds this acceptable. This item, therefore,

is closed.

Watts'Bar SSER 2 3ý-1

4 REACTOR

4.2 Fuel System Design

4.2.2 Thermal Performance

Thermal Performance Analysis

A Westinghouse fuel thermal performance code known as PAD-3.1, described in

attachments to correspondence from Westinghouse to the U.S. Atomic Energy

Commission (letters dated December 22 and 29, 1972, and January 1 and 12, 1973),

was initially used for the Watts Bar safety analysis. A more recent Westinghouse fuel thermal performance code known as PAD-3.3 (Westinghouse Topical

Report WCAP-8720) has also been approved by the NRC. The more recent code,

which contains revised models for fission gas release, helium solubility, fuel

swelling, and fuel densification, was not used in the applicant's original

submittal.

The PAD-3.3 code addresses a concern about enhanced fi'ssion gas release at'high

burnup. Because the earlier version of the code did not contain the models

necessary to analyze this effect, the staff's safety evaluation of PAD-3.3

stated that future fuel performance analyses must be done with the revised

version of the code (PAD-3.3) (letter dated February 9, 1979).

The use of the PAD-3.1 code is generally acceptable because the earlier code

produces more conservative thermal conditions than the revised code. This

margin, however, does not exist for high burnup fission gas release and fuel

rod internal pressure calculations. In a letter dated September 22, 1981, the

applicant stated that the more recent code, PAD-3;-3, is now used to analyze

the fuel thermal performance (including fission gas release and rod internal

pressure) at the Watts Bar facility. On the basis of the applicant's use of

the revised model, the staff now concludes that the Watts Bar fuel performance

analysis is acceptable.

In addition, revised internal fuel rod pressure criteria, as described in

Westinghouse Topical Report WCAP-8963-A, are also used now in the Watts Bar

safety analysis. The approved criteria are:

(1) The internal pressure is limited so that the fuel-to-cladding gap does

not increase during normal operation.

(2) Extensive departure from nucleate boiling propagation does not occur

during postulated transients-and accidents.

The Watts Bar Final Safety Analysis Report (FSAR) has been amended (Amendment 34) to incorporate these criteria. The staff finds this acceptable. This

issue, therefore, is closed.

Watts Bar SSER 2 4-1
Cladding Collapse

The staff has reviewed Westinghouse Topical Report WCAP-8377, which describes

the details of a revised cladding-flattening model that, for a given fuel region,

predicts initial flattening (collapse) time for pressurized rods containing

relatively stable fuel. This revised analysis was based on the results of TV

examinations of irradiated fuel rods, and the results indicated that the original flattening model (Westinghouse Topical Report WCAP-7982) significantly

underpredicted the time of collapse. The revised model was accepted for use in

plant safety analysis subject to provisions specified in the SER (memorandum

dated January 14, 1975) that did not permit alterations to the specified curves

used as input to the model. In a letter dated September 22, 1981, the applicant

stated that the revised cladding collapse model is used in a manner consistent

with conditions given in the staff's safety evaluation of WCAP-8377. The predicted cladding collapse time for the most limiting Watts Bar fuel has been

calculated in this manner and has been shown to be in excess of 38,000 effective

full-power hours of operation, which is greater than the expected residence

time of the fuel. The staff finds this acceptable. This issue, therefore, is

closed.

4.2.3 Mechanical Performance

Fuel Rod Bowing

The consideration of fuel rod bowing in the Watts Bar 17 x 17 fuel design was

initially analyzed by Westinghouse'in Topical Report WCAP-8346. Subsequently,

Westinghouse revised that analysis in light of new information and documented

the results in Topical Report WCAP-8692. A revision of this latter report has

been reviewed and approved by the NRC. The Watts Bar FSAR has been amended

(Amendment 31) to reflect the revised methodology. The staff concludes that

the applicant has presented an acceptable means of analyzing the effects of

fuel rod bowing and determining any residual rod bowing penalties on the departure from nucleate boiling ratio and total peaking factor for Watts Bar Units 1

and 2. This issue, therefore, is. closed.

Watts Bar SSER 2 4-2
5 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS

5.2 Integrity of Reactor Coolant Pressure Boundary

5.2.2 Overpressure Protection

The applicant referenced WCAP-7769 as part of the analytical basis for safety

valve sizing. In the SER the staff requested additional information on the

differences between the Watts Bar Nuclear Plant. and the reference plant for

WCAP-7769. The staff also requested verification that adequate protection is

provided if a reactor trip is initiated by the second safety-grade signal from

the reactor protection system as required by the Standard Review Plan (SRP,

NUREG-0800). This information has been provided by the applicant by letter

dated April 18, 1983. The only significant difference between Watts Bar and

the reference plant of WCAP-7769 is that, following a reactor trip signal, the

shutdown rods may enter the reactor core with a 2-sec delay at Watts Bar rather

than the 1-sec delay assumed for the reference plant. The design-basis overpressure transient is a combined loss of main feedwater and turbine trip. The

transient would generate reactor trip signals for each of the following conditions:

(1) turbine trip

(2) high reactor system pressure (3) high temperature difference across the core

(4) loss of feedwater

(5) low steam generator level with low steam flow

(6) low-low steam generator level

Receipt of only one of these trip signals by the reactor protection system will

trip the reactor. WCAP-7769 demonstrated that small delays in reactor trip

time would not affect the peak reactor system pressure because the reactivity

loss from core heatup would be adequate to limit the initial pressure peak from

this event. The analyses indicated that no trip was required for the first,

25 sec although reactor trip would eventually be required to compensate for the

reduction in heat sink. The staff concludes that the safety valves at Watts

Bar are adequately sized and that this issue is resolved.

5.4 Component and Subsystem Design

5.4.3 Residual Heat Removal System

The SER stated that the applicant had committed to test manually the atmospheric

steam line dump valves. In a letter dated June 16, 1983, the applicant stated

that manual actuation of the atmospheric dump valves will not be tested.

The four air-operated atmospheric dump valves at Watts Bar (one per steam generator) are seismic Category I. Electric and air power sources to the valves are

safety related. The valves can be operated from the control room. The steam

relief capacity of the valves is sufficient so that failure of any one valve to

open will not prevent the plant from reaching cold shutdown using the remaining

Watts Bar SSER 2 5-1
three valves in a reasonable period of time if required. The staff agrees with

the applicant that manual actuation testing of the atmospheric relief valves is

not necessary. The staff,*therefore, concludes that no further action need be

taken in this regard and this item is closed.

5.4.5 Reactor Coolant System Vents (II.B.1)

NUREG-0737 requires installation of reactor coolant system (RCS) and reactor

vessel head high point vents that are remotely operated from, the control room.

As stated in the SER, the applicant committed to install an acceptable RCS vent

system before fuel loading. In addition, the applicant committed to use venting

guidelines that were being developed by the Westinghouse Owners Group. In the

SER the staff concluded these commitments were acceptable but stated that modification of the venting guidelines would be required if they were indicated by

the staff's generic review. This review has been completed as documented in

NRC Generic Letter 83-22, dated June 3, 1983. The staff concluded that the

guidelines are acceptable for implementation. The staff, therefore, concludes

that a license condition regarding the modification of the venting guidelines

is not necessary and finds this item acceptable pending verification that the

RCS vent system is installed.

Watts Bar SSER 2 5-2
6 ENGINEERED SAFETY FEATURES

6.3 Emergency Core Cooling System

6.3.3 Testing

To ensure that debris following a loss-of-coolant accident will not compromise the performance of the emergency core cooling system by clogging the sump, the staff asked the applicant to perform a detailed survey of insulation materials used within the containment. The applicant provided this information in a letter dated November 23, 1982. This survey confirms the staff's initial con- clusion that the Watts Bar design to provide protection against sump debris is acceptable. The reactor system and main steam piping and components are encased in metal reflective insulation that, if dislocated by a major pipe rupture, would not form small debris particles that would clog the sump screens. Other materials (foam glass, Rubatex, fiberglass, polyurethane foam, urethane foam, and mineral wool) are either encapsulated in steel or located in areas of the containment where they would be unaffected by pipe rupture forces. The staff concludes that the Watts Bar design regarding protection against sump debris is acceptable and this issue, therefore, is closed.

Watts Bar SSER 2 6-1

7 INSTRUMENTATION AND CONTROLS

7.2 Reactor Trip System

7.2.5 Steam Generator Water Level Trip

The water level measurement channels use differential pressure transmitters.

The measurement accuracy of such a system is affected by several factors. Of

primary importance is the increase in the indicated water level caused by a

decrease of the water density in the reference leg. This can occur because of

an increase in the ambient temperature as a result of a high-energy-line break. This issue was identified in the SER as a matter to be reviewed by the staff.

In a letter dated June 21, 1982, the applicant committed to insulate the reference leg to alleviate the temperature-dependence problem. The staff finds this

acceptable. The staff will review the trip setpoint for the steam generator

level trip function during the Technical Specification review. The staff,

therefore, considers this item closed.

7.3 Engineered Safety Features Actuation System

7.3.2 Containment Sump Level Measurement

The containment sump level is monitored by four level measurement channels using

differential pressure transmitters. The staff identified a concern in the SER

that debris in the sump could block the inlets to the differential pressure

transmitters and result in a loss of the permissive signal to the initiation

logic for the automatic switchover from the injection to the recirculation mode

of the emergency core cooling system. In a letter dated September 15, 1983,

the applicant stated that the sump level sensors have been moved from inside

the sump wall to outside the sump wall. Sump water level is detected by the

sensor by means of a 1-in. sense line routed through the sump wall. The sense

line opening is protected by a 2-in.-diameter cap with ten h-in. holes to prevent large pieces of debris from blocking the opening in the sense line. On

the basis of this action, the staff considers this item closed.

Watts Bar SSER 2 7-1

8 ELECTRIC POWER SYSTEMS

8.2 Offsite Electric Power System

8.2.2 Compliance With GDC 17

8.2.2.1 Availability of Offsite Power Circuits

In the SER the *staff concluded that the offsite power system circuits at the

Watts Bar Hydro Plant Switchyard meet GDC 17 and are acceptable pending documentation in the FSAR of the additional information submitted by letter dated

October 9, 1981. The applicant documented the additional information in FSAR

Amendment 48. This item, therefore, is closed.

8.2.2.2 Minimizing the Probability of Losing all AC Power

In the SER the staff expressed the concern that the automatic transfer of loads

from the normal power source to the various other offsite sources could cause

overloading and loss of both offsite circuits. The applicant, by Amendment 48

to the FSAR, proposed a system design change. The proposed design (1) retains

provisions for automatic transfer of safety loads between the various offsite

power sources, (2) retains provisions so that a faulted or overloaded bus will

not be automatically transferred, and (3) adds two new common station transformers. The greater capacity of these new transformers (dedicated to supplying

safety-related buses and their associated loads) eliminates the original overloading concern. This item, therefore, is closed.

The staff also indicated in SER Section 8.2.3 that information had not been

provided describing the capability to test these transfers during normal power

operation. By Amendment 48 to the FSAR, the applicant indicated that these

transfers will not be tested during normal power operation because such transfers could result in transients that could cause tripping of the reactor or

turbine. Therefore, the staff concludes that the design meets GDC 18 and is

acceptable.

Testing requirements for the automatic transfers and the design, which prevents

a faulted or overloaded bus from being automatically transferred, will be reviewed with the Technical Specifications.

8.3 Onsite Power Systems

8.3.1 Onsite AC Power System Compliance With GDC 17

8.3.1.1 Nonsafety Loads Powered From the Class 1E AC Distribution System

In regard to the capability of the offsite power system, the staff concluded

in the SER that the proposed new design for the offsite power system was acceptable pending its incorporation in the FSAR. The applicant by Amendment 48

incorporated the new design in the FSAR. This item, therefore, is closed.

Watts Bar SSER 2 8-1
8.3.1.2 Low and/or Degraded Grid Voltage Condition

In the SER the staff indicated that the design for the low and/or degraded grid

voltage condition was acceptable pending verification of design implementation.

During a site visit on July 12-14, 1982, the staff reviewed Drawing 45W760-211-17,

Revision 2. On the basis of the review of this drawing, the staff confirmed

the design implementation. This item, therefore, is closed.

The staff also indicated in the SER that the voltage drop analysis and testing

would be verified. During a site visit on July 12-14, 1982, the staff verified

the analysis and found the results satisfactory. The test results that substantiate the analysis will be verified by the staff. If anyproblem areas are

identified, they will be reported in a future supplement to the SER.

8.3.1.6 Diesel Generator Reliability Qualification Testing

In the SER the staff required, as a-condition to the license, that the capability of the diesel generator to start at normal operating temperature be demonstrated before'fuel loading. In a letter dated August 31, 1983, the applicant

documented that the capability has been demonstrated by a test on a diesel generator identical to those used at Watts Bar. On the basis of this test, the staff

considers this item resolved and no longer a license condition.

The staff also indicated that the diesel generator qualification testing would

be verified. If any problem areas are identified as a result of'the staff

verification, they will be reported in a future supplement to the SER.

8.3.2 Onsite DC System Compliance With GDC 17

8.3.2.4 Diesel Generator Battery System

In the SER the staff indicated that the design analysis for demonstrating compliance of the diesel generator with regulatory requirements and guidelines was

acceptable pending the incorporation of the analysis in the FSAR. By Amendment 48, the applicant incorporated the analysis in the FSAR. This item,

therefore, is closed.

8.3.3 Common Electrical Features and Requirements

8.3.3.1 Compliance With GDC 2 and 4

8.3.3.1.2 Thermal-Overload Protection Bypass

In the SER the staff indicated that the design for bypass of thermal-overload

protective devices on safety-related motor-operated valves would be verified

during the electrical drawing review. During a site visit on July 12-14, 1982,

the staff reviewed Drawings 45W760-62-3 (Revision 7) and 45W760-270-2'(Revision 8). On the basis of the review of these drawings, the staff verified the

design. This item, therefore, is closed.

Watts Bar SSER 2 8-2
8.3.3.2 Compliance With GDC 5

8.3.3.2.3 Sharing of Raceway Systems Between Units

In the SER the staff indicated that the design for sharing of raceway systems

between units would be confirmed as part of the drawing review/site visit.

During a site visit on July 12-14, 1982, the staff traced the following cable

routings:

(1) 6,900-V 1B and 2B train cables between the diesel generator and the

switchgear

(2) 6,900- to 480-V 2A train cables between their respective switchgear

(3) cables associated with the A and B train turbine and motor-driven auxiliary

feedwater pump

On the basis of the cable tracing, the staff concludes that cables associated

with the A train ofUnits 1 and 2 are extensively routed in close proximity

and in the same raceway. Similarly, cables associated-with the B train of

Units 1 and 2 are routed together. A and B cables are routed in physically

separate raceways. This cable routing is in accordance with the accepted

Watts Bar separation criteria. This item, therefore, is closed.

8.3.3.5 Compliance With GDC 18

8.3.3.5.2 Testing of One of Two Class 1E Power Systems Versus One of Four

Systems

In the SER the staff indicated that the commitment for testing one of four

diesel generators at any one time was acceptable pending documentation of this

commitment in the FSAR. By Amendment 48 to the FSAR, the applicant provided

the required documentation. This item, therefore, is closed.

8.3.3.6 Compliance With GDC 50

In the SER the staff required, as a condition to the license, that redundant

fault current protective devices be provided in series for the reactor coolant

pump circuits in accordance with Position 1 of Regulatory Guide (RG) 1.63. By

Amendment 48 to the FSAR, the applicant documented that the design for reactor

coolant pump penetration protection would contain the required redundant circuit

breakers. The proposed design meets Position 1 of RG 1.63 and is acceptable.

This item, therefore, is resolved and is no longer a license condition.

Watts Bar SSER 2 8-3

17 QUALITY ASSURANCE

The staff review of the description of the Watts Bar Nuclear Plant quality

assurance (QA) program for the operations phase has verified that the criter3

of Appendix B to 10 CFR 50 have been adequately addressed in Section 1- 2 of

the FSAR through Amendment 48. This determination of acceptability included a

review of the list of items to which the QA program applies.

In the SER the staff stated that this list had not been approved by the staff.

Since then, the list of items has been reviewed by the staff to ensure that

safety-related items within the scope of staff review are under the QA program

controls. Differences between the staff and the applicant regarding the list

have been resolved to the staff's satisfaction. The list has been expanded to

include safety-related items reflected in NUREG-0737, "Clarification of TMI

Action Plan Requirements," November 1980. Therefore, the staff has no open

items concerning the QA program for operations or the items to which the p-ogram applies.

On the basis of its review and evaluation of the QA program description in FSAR

Section 17.2, the staff concludes:

(1) The QA organizations of TVA are provided sufficient independence from cost

and schedule (when opposed to safety considerations) and have sufficient

authority to carry out effectively the operations QA program and sufficient

access, to management at a level necessary to perform their QA functions.

(2) The QA program description contains adequate QA requirements and a comprehensive system of planned and systematic controls that address each of

the criteria of Appendix B to 10 CFR 50 in an acceptable manner. This QA

program description, therefore, can serve as an adequate basis for the

development of specific policies and procedures to implement the QA responsibilities of TVA for the operation of the Watts Bar Nuclear Plant.

Accordingly, the staff concludes that the applicant's description of the QA

program is in compliance with applicable NRC regulations and that this item is

closed.

Watts Bar SSER 2 17-1

APPENDIX A

CHRONOLOGY OF RADIOLOGICAL REVIEW

OF WATTS BAR NUCLEAR PLANT, UNITS 1 AND 2,

OPERATING LICENSE REVIEW

September 9, 1982

September 15, 1982

September 21, 1982

September

September

21,

21,

1982

1982

September 22, 1982

September 22-23,

1982

September 23, 1982

Letter, from applicant requesting extension of construction

permit for Unit 1 to March 1984 and Unit 2 to August 1985.

Letter from applicant forwarding proposed modifications to

the NRC draft version of the Technical Specifications.

Letter from applicant concerning response to.NRC

question Q22.70.

Letter from applicant concerning cement mortar lining.

Letter from applicant concerning single-failure criteria

for a boron dilution event.

Letter from applicant concerning Radiological Effluent

Technical Specifications.

Meeting with applicant to perform an audit of geotechnical

engineering documents to confirm conclusions reached during

licensing review. (Summary issued November 12, 1982).

Letter to applicant concerning seismic and dynamic qualification review of safety-related equipment.

Supplement 1 to Safety Evaluation Report (SER) issued.

Letter from applicant concerning preservice inspection

program.

Letter to applicant concerning inconsistency between requirements of 10 CFR 50.54(t) and Standard Technical Specifications-for performing audits of emergency preparedness

programs (Generic Letter 82-17).

Letter to applicant requesting additional information regarding Item II.E.4.2 of NUREG-0737.

Letter to applicant concerning Technical Specifications for

fire protection audits (Generic Letter 82-21).

Letter to applicant concerning reactor operator and senior

reactor operator requalification examinations (Generic

Letter 82-18).

September

September

29,

29,

1982

1982

October 1, 1982

October 5, 1982

October 6, 1982

October 12, 1982

Watts Bar SSER 2 A-1
October 19,

October 18,

1982

1982

October 20, 1982

October 25, 1982

October 26, 1982

October 28, 1982

October 30, 1982

November 2, 1982

November 2-3, 1982

November 4, 1982

November 9, 1982

November 10, 1982

November 10, 1982

November 16, 1982

November 18, 1982

November 19, 1982

November 23, 1982

Watts Bar SSER 2

Letter to applicant concerning review schedule.

Letter from applicant concerning program for the watchtower

power block security concept.

Letter from applicant forwarding revision to Physical

Security Plan.

Letter from applicant concerning the critical structures,

systems, and components list.

Letter to applicant concerning guidance for implementing

Standard Review Plan rule (Generic Letter 82-20).

Letter from applicant forwarding report, "Site-Specific

Top-of-Ground Motions for ERCW Pipeline."

Letter to applicant concerning inconsistency between requirements of 10 CFR 73.40(d) and Standard Technical Specifications for performing audits of safeguards contingency

plans (Security Plan)(Generic Letter 82-23).

Letter to applicant concerning generic security training

and qualification plan.

,Meeting with applicant to discuss proposed conceptual

changes to the security program.

Letter from applicant concerning cement mortar lining of

essential raw cooling water (ERCW) piping.

Letter from applicant providing schedule for responding to

NRC concerns in SER.

Letter from applicant concerning sampling program to verify

that buckling stresses in axial compression members used

in pipe support configurations do not exceed 90% of yield

strength when slenderness ratio is less than 30.

Letter from applicant concerning the loose parts monitoring

system.

Letter from applicant forwarding report, "Liquefaction

Evaluation of the ERCW Pipeline Route - Watts Bar Nuclear

Plant."

Letter from applicant concerning buckling of Class 2 and 3

supports.

Letter to applicant extending construction completion dates.

*Unit 1 completion date is now extended to March 1, 1984,

and that of Unit 2 to August 1, 1985.

Letter from applicant responding to NRC questions regarding

an insulation survey.

A-2
November 24, 1982

November 29, 1982

November 30, 1982

November 30, 1982

November 30, 1982

December 1, 1982

December 1, 1982

December 6, 1982

Letter from applicant concerning modifications to protect the diesel generator from degraded operation resulting from

a tornado-generated missile impacting the diesel generator exhaust.

Letter from applicant concerning auxiliary systems for the

diesel generator units.

Letter from applicant providing schedules for milestones

concerning the prompt notification system.

Letter from applicant concerning data availability for each

primary meterological measurements system.

Letter from applicant concerning geotechnical audit conducted September 22-24, 1982.

Letter to applicant requesting additional information

regarding Branch Technical Position CSB 6-4, "Containment

Purging During Normal Plant Operation."

Letter from applicant concerning information requested by the NRC Seismic Qualification Review Team.

Letter from applicant concerning plans for emergency opera- tions facility.

Letter from applicant forwarding copies of report, "Lique- faction Evaluation of the ERCW Pipeline Route - Watts Bar

Nuclear Plant."

Letter from applicant providing additional information concerning diesel generators.

Letter from applicant providing Revision 3 to ASME Code, Section XI, Preservice Inspection Program Technical Instruction TI-50B.

Letter to applicant forwarding Supplement 1 to NUREG-0737 - Requirements for Emergency Response Capability (Generic Letter 82-33).

Letter from applicant concerning use of the computer card

deck containing the time histories of the four artificial

earthquakes used for Watts Bar.

December 9, 1982

December 14, 1982

December 16, 1982

December 17, 1982

December 20, 1982

December 20,

December 22,

1982 Letter from applicant providing status report on proposed modifications to the Model D steam generators.

1982 Letter to applicant concerning problems with the submittals

of 10 CFR 73.21 safeguards information for licensing review

(Generic Letter 82-39).

Watts Bar' SSER 2 A-3
December 22, 1982

December 28, 1982

December 28, 1982

December 29, 1982

January 4, 1983

January 11, 1983

January

January

January

11,

17,

17,

January 21,

January 21,

January 24,

1983

1983

1983

1983

1983

1983

1983

1983

Letter to applicant concerning meeting to discuss-recent

developments for operating licensing examinations (Generic

Letter 82-38).

Letter from applicant concerning Appendix R review and

instrumentation available in the auxiliary control room.

Letter to applicant concerning filings relating to 10 CFR 50

production and utilization facilities (Generic Letter 82-30).

Letter from applicant concerning the process control program

for solidification of radwaste.

Letter from applicant submitting Amendment 47 to Final

Safety Analysis Report.

Letter to applicant concerning operator licensing examination site visit (Generic Letter 83-01).

Letter from applicant concerning compliance with Appendix R.

Letter from applicant responding to NRC staff questions.

Letter from applicant concerning compliance with 10 CFR 50,

Appendix R.

Letter from applicant concerning soil amplification studies.

Presentation by the Design Review Panel of their program

to review the Model D2/D3 steam generator modification.

Letter to applicant concerning final rulemaking concerning

reporting of changes to quality assurance programs for

nuclear power plants and fuel reprocessing plants (generic).

Letter from applicant providing comments on the SER.

Meeting with applicant to discuss Westinghouse's appeal of

the staff's position regarding reactor trip breaker testing

on Westinghouse plants.

Letter to applicant concerning certificates and revised

format for reactor operator and senior reactor operator

licenses (Generic Letter 83-06).

Letter to applicant concerning regional workshops regarding

Supplement 1 to NUREG-0737, "Requirements for Emergency

Response Capability" (Generic Letter 83-04).

Letter to applicant concerning safety evaluation of "Emergency Procedure Guidelines, Revision 2," NEDO-24934, June

1982 (Generic Letter 83-05).

January

January

25,

26,

January 31, 1983

February 1, 1983

February 1, 1983

Watts Bar SSER 2 A-4
February 1, 1983

February 3, 1983

February 17, 1983

February 22, 1983

February 22, 1983

February 24, 1983

February 24, 1983

March 1-3, 1983

March 2, 1983

March 7, 1983

March 16, 1983

March 21, 1983

March 22, 1983

March 23, 1983

March 23, 1983

Letter to applicant concerning resolution of TMI Action

Plan Item II.K.3.5, "Automatic Trip of Reactor Coolant

Pumps" (Generic Letter 83-10c).

Letter from applicant forwarding revision of Physical Security Plan.

Letter from applicant forwarding Revision 8 to ASME Code,

Section XI, Preservice Inspection Program Technical Instruction TI-50A.

Letter to applicant concerning reactor trip breaker test

appeal meeting.

Letter from applicant forwarding the process control program

for ensuring solidification and dewatering.

Letter to applicant concerning evaluation of boron dilution

event response.

Letter to applicant concerning issuance of NRC Form 398 -

Personal'Qualifications Statement - Licensee (Generic

Letter 83-12).

Caseload forecast meeting at Watts Bar site to update the

NRC's projected fuel loading date.

Letter to applicant concerning clarification of surveillance

requirements for high-efficiency particulate air filters

and charcoal adsorber units in Standard Technical Specifications on engineered safety features cleanup systems

(Generic Letter 83-13).

Letter to applicant concerning definition of "Key Maintenance Personnel" (clarification of Generic Letter 82-12)

(Generic Letter 83-14).

Letter to applicant forwarding staff evaluation of Utility

Design Review Panel Report on modification to Westinghouse

D2/D3 steam generators.

Meeting with Design Review Panel to discuss information

related to the forward flush transient on D2/D3 steam

generators.

Letter from applicant forwarding draft Technical Specifications.

Letter to applicant concerning request for withholding

information from public disclosure.

Letter to applicant concerning implementation of Regulatory

Guide 1.150, "Ultrasonic Testing of Reactor Vessel Welds

During Preservice and Inservice Examinations, Revision 1"

(Generic Letter 83-15).

Watts Bar SSER 2 A-5
March 24, 1983

March 25, 1983

March 25, 1983

March 28, 1983

March 29, 1983

April 1, 1983

April 8, 1983

April 14, 1983

April 15, 1983

April 18, 1983

April 18, 1983

April 20, 1983

April 22, 1983

April 22, 1983

April 25, 1983

April 26, 1983

Watts Bar SSER 2

Letter to applicant transmitting NUREG-0977 relative to

the anticipated transients without scram events at Salem

Generating Station, Unit No. 1 (Generic Letter 83-16).

Letter to applicant concerning alternate shutdown capabilities at the Watts Bar plant.

Letter from applicant concerning analysis of axial stresses

on buried pipe.

Letter from applicant concerning initial test program.

Letter to applicant concerning evaluation of liquefaction

potential of the soils beneath the ERCW pipeline and

electrical conduits.

Letter to applicantconcerning environmental qualification

of safety-related mechanical equipment located in harsh

environment areas.

Letter to applicant concerning integrity of the requalification examinations for renewal of reactor operator and senior

reactor operator licenses (Generic Letter 83-17).

Letter from applicant forwarding copies of TVA Design

Criteria WB-DC-10-2 pertaining to the security-power block

project.

Letter from applicant responding to Generic Letter 82-33

regarding Supplement 1 to NUREG-0737, Requirements for

Emergency Response Capability.

Letter from applicant forwarding listing of welds to date

for which relief is being requested and that have had examination reports reviewed.

Letter from applicant concerning applicability of WCAP-7769

to the issue of safety valve sizing.

Letter from applicant concerning initial test program.

Letter from applicant concerning modifications to the

chemical feed lines at the plant.

Letter from applicant responding to Generic Letter 83-10c

concerning NUREG-0737, Item II.K.3.5, "Automatic Trip of

Reactor Coolant Pumps."

Letter from applicant forwarding a list of safety-related

electrical components and their respective functions and a

summary of data used to establish set points for balanceof-plant equipment.

Letter from applicant concerning Lawrence Livermore National

Laboratory's preliminary system interaction results from

A-6
April 26, 1983

April 29, 1983

May 9, 1983

May 10, 1983

May 11, 1983

May 11,

May 17,

May 18,

May 19,

May 20,

May 24,

May 24,

May 27,

June 1,

June 3,

June 7,

1983

1983

1983

1983

1983

1983

1983

1983

1983

1983

1983

the digraph matrix analysis of the plant's safety injection

system.

Letter from applicant concerning Branch Technical Position

CSB 6-4.

Letter from applicant concerning boron dilution event

response.

Letter to applicant concerning integrated scheduling for

implementation of plant modifications (Generic Letter 83-20).

Letter from applicant concerning research and development

work performed for Sequoyah Nuclear Plant in the area of

hydrogen capabilities and its applicability to Watts Bar.

Letter to applicant concerning clarification of access

control procedures for law enforcement visits (Generic

Letter 83-21).

Letter from applicant forwarding copies of Black and

Veatch final report.

Letter from applicant concerning evaluation of potential

for liquefaction at plant.

Letter to applicant concerning environmental qualification

program.

Meeting with applicant to discuss proposed application of

the watchtower power block concept.

Meeting with applicant to discuss measures to prevent

liquefaction of the soils beneath the ERCW pipeline.

(Summary issued June 8, 1983.)

Letter to applicant providing results of NRC Caseload

Forecast Panel meeting.

Letter to applicant concerning the Power Block Physical

Security Plan.

Letter from applicant concerning schedule for steam generator modifications.

Letter to applicant requesting additional information

regarding the geotechnical concerns.

Letter to applicant forwarding safety evaluation of Emergency Response Guidelines, (Generic Letter 83-22).

Letter from applicant providing revised response to

NUREG-0737, Item II.F.1.

Watts Bar SSER 2 A-7
June 9, 1983

June 9, 1983

June 16, 1983

June 22, 1983

July 1, 1983

July 5, 1983

July 6, 1983

July 7, 1983

July 8, 1983

July 21, 1983

July 21, 1983

July 22, 1983

July 26, 1983

July 27, 1983

July 27, 1983

August 5, 1983

August 8, 1983

Letter from applicant responding to Power Systems Branch

concerns specified in the SER.

Letter from TVA forwarding generic control room design

review program plan for Sequoyah, Watts Bar, Bellefonte,

and Browns Ferry.

Letter from applicant providing updated listing showing

status of open and confirmatory items in the SER.

Letter to applicant concerning peak horizontal ground

acceleration for use in soil liquefaction analysis.

Letter from applicant concerning current status and revised

implementation schedule for prompt notification system.

Letter to applicant concerning clarification of surveillance

requirements for diesel fuel impurity level tests (Generic

Letter 83-26).

Letter to applicant concerning surveillance intervals in

Standard Technical Specifications (Generic Letter 83-27).

Letter from applicant documenting discussions between NRC

and TVA staffs regarding preoperational test programs.

Letter to applicant concerning required actions based on

generic implications of Salem ATWS events (Generic

Letter 83-28).

Letter to applicant concerning deletion of Standard Technical

Specification Surveillance Requirement 4.8.1.1.2.d.6 for

diesel generator testing (Generic Letter 83-30).

Letter from applicant concerning the analysis of sheetpile

walls and seismic analysis of buried pipes.

Letter from applicant responding to NUREG-0737, Item II.D.1,

regarding performance testing of pressurized-water reactor

relief and safety valves.

Letter from applicant responding to NUREG-0737, Item

II.F.2.

Letter from applicant forwarding proposed modifications to

the NRC draft version of the Technical Specifications.

Letter from applicant concerning requirements for performing

a preliminary control room assessment.

Letter from applicant responding to NRC question 212.35.

Letter from applicant providing revisions to ASME Code,

Section XI, Preservice Inspection Program Technical Instructions TI-50A and TI-50B.

Watts Bar SSER 2 A-8
August 11, 1983

August 18, 1983

August

August

August

22,

31,

31,

1983

1983

1983

September

September

September

September

September

September

September

September

13,

14,

15,

20,

20,

28,

29,

30,

1983

1983

1983

1983

1983

1983

1983

1983

Letter from applicant concerning liquefaction potential of

the soils beneath the ERCW pipeline and Class 1E electrical

conduit.

Letter from applicant forwarding report, "Reactor Building

Containment Integrated Leak Rate Test, Watts Bar Nuclear

Plant Unit 1."

Letter from applicant concerning the lowest service metal

temperature of +40'F for the main feedwater system.

Letter to applicant concerning fire protection issues.

Letter from applicant concerning diesel generator reliability qualification testing.

Letter from applicant forwarding generic control room design

review program plan.

Letter from applicant concerning geotechnical design

features.

Letter from applicant concerning containment sump level.

Letter from applicant concerning final response to NUREG-0737,

Item II.B.3.

Letter from applicant concerning postaccident sampling system items.

Letter from applicant concerning discontinuance of submittal

of meterological data availability reports.

Letter from applicant providing results of the preliminary

lighting survey for the main control room.

Letter to applicant concerning review of Black and Veatch's

independent review of the auxiliary feedwater system for

Unit 1.

Letter to applicant requesting additional information

regarding environmental equipment qualification.

Letter from applicant concerning safety-grade manual switches

in the main control room for closure of isolation valves in

the upper head injection system.

Letter from applicant concerning compliance with 10 CFR 50,

Appendix R.

Letter from applicant forwarding letter to Black and Veatch

concerning the handling of the independent review program.

Letter from applicant forwarding drawings related to two

sump level sensors.

A-9

October

October

4, 1983

11, 1983

October 12, 1983

October 14, 1983

October 18, 1983

Watts Bar SSER 2
October 18, 1983 Letter from applicant concerning meeting between TVA and

Black and Veatch to'discuss the independent review program.

October 19, 1983 Letter to applicant concerning NRC positions on certain

requirements of Appendix R to 10 CFR 50 (Generic.Letter 83-33).

October 24, 1983 Letter from applicant concerning NUREG-0737, Item ll.D.3,

"Direct Indication of Relief and Safety Valve Position."

October 26, 1983 Letter to applicant concerning clarification of required

actions based on generic implications of Salem ATWS events.

October 27, 1983 Letter from applicant concerning revisions to Physical

Security Plan.

October 28, 1983 Letter to applicant concerning compliance with General

Design Criterion 51.

October 28, 1983 Letter from applicant concerning their response to

NUREG-0737.

October 31, 1983 Letter to applicant transmitting NUREG-0965, "NRC Inventory

of Dams," (Generic Letter 83-38).

October 31, 1983 Letter from applicant concerning NUREG-0737, ITEM ll.D.3

(Direct Indication of Relief and Safety Valve Position).

October 31, 1983 Letter from applicant concerning his response to Generic

Letter 82-33.

November 1, 1983 Letter from applicant concerning various TVA commitments

and SER items.

November 1-2, 1983 Meeting with applicant to discuss the Technical Specifications.

November 2, 1983 Letter to applicant concerning the clarification of TMI

Action Plan Item II.K.3.31 (Generic Letter 83-35).

November 7, 1983 Letter from applicantresponding to Generic Letter 83-28

concerning the required actions based on generic implications of Salem ATWS events.

November 7, 1983 Letter from applicant concerning compliance with equipment

qualification requirements.

November 8, 1983 Letter from applicant concerning their commitments regarding control room modifications.

November 8, 1983 Letter from applicant concerning installation of high-range

noble gas monitors on the steam generator safety and PORV

release lines.

Watts Bar SSER 2 A-IO0
November 9, 1983

November 21, 1983

November 22, 1983

December 2, 1983

December 2, 1983

December 6, 1983

December 12, 1983

December 13, 1983

December 16, 1983

December 19, 1983

December 19, 1983

December 19, 1983

December 20, 1983

December 20, 1983

December 21, 1983

Meeting with applicant at plant site to review remedial

work being performed as a result of the liquefaction potential analysis for the soils beneath the ERCW pipeline.

Letter from applicant providing revisions to the ASME

Section XI, "Preservice Inspection Program Technical

Instruction."

Letter from applicant'requesting exemption for not providing high-range noble gas monitors on the auxiliary building

vent.

Generic Letter 83-32 -- NRC Staff Recommendations Regarding

Operator Action for Reactor Trip and ATWS.

Letter fromapplicant concerning program plan for evaluating

the environmental qualification of safety-related mechanical

equipment.

Letter from applicant concerning containment purge and vent

valve operability analysis report.

Letter to applicant concerning manual control of the upper

head injection system.

Meeting with applicant to discuss the Technical Specifications.

Letter to applicant concerning review of emergency action

levels.

Letter to applicant reporting requirements of 10 CFR Part 50,

Sections 50.72 and 50.73, and Standard Technical Specifications (Generic Letter 83-43).

Letter from applicant concerning postaccident sampling

capability.

Letter to applicant concerning clarification to Generic

Letter 81-07 regarding response to NUREG-0612, "Control of

Heavy Loads at Nuclear Power Plants" (Generic Letter 83-42).

Letter to applicant concerning availability of NUREG-1021,

"Operator Licensing Examiner Standards" (Generic

Letter 83-44).

Meeting with applicant to discuss requests for. exemption

from item II.F.1 of NUREG-0737.

Letter to applicant regarding operator licensing examinations (Generic Letter 83-40).

Watts Bar SSER 2 A- 11
December 22, 1983

December 23, 1983

Letter from applicant concerning Confirmatory Item 31 of

the SER.

Letter to applicant concerning comments on TVA program plan

for control room design reviews.

Watts Bar SSER 2 A-12
APPENDIX B

BIBLIOGRAPHY

Letter, Dec. 22, 1972 (NS-SL-518), from R. Salvatori, Westinghouse, to D. Knuth, AEC, Subject: PAD-3.1 - Thermal Performance Code for Westinghouse PWR Fuel.

--- , Dec. 29, 1972 (NS-SL-521), from R. Salvatori, Westinghouse, to D. Knuth, AEC, Subject: PAD-3.1 - Thermal Performance Code for Westinghouse PWR Fuel.

--- , Jan. 1, 1973 (NS-SL-527), from R. Salvatori, Westinghouse, to D. Knuth, AEC, Subject: PAD-3.1 - Thermal Performance Code for Westinghouse PWR Fuel.

--- , Jan. 12, 1973 (NS-SL-543), from R. Salvatori, Westinghouse, to D. Knuth, AEC, Subject: PAD-3.1 - Thermal Performance Code for Westinghouse PWR Fuel.

--- , Feb. 9, 1979, from J. Stolz, NRC, to T. Anderson, Westinghouse, Subject: Safety Evaluation of WCAP-8720 and WCAP-8785.

Memorandum, Jan. 14, 1975, from V. Stello, NRC, to R. DeYoung, Subject: Safety Evaluation of WCAP-7982.

Tennessee Valley Authority, "Final Safety Analysis Report for Watts Bar Nuclear

Plant, Units 1 and 2," Oct. 4, 1976.

U.S. General Services Administration, Office of the Federal Register National

Archives and Records Service, Code of Federal Regulations, Title 10, "Energy" (including General Design Criteria), U.S. Government Printing Office, Washington, DC, Jan. 1981:

U.S. Nuclear Regulatory Commission, NUREG-0737, "Clarification of TMI Action

Plan Requirements," Nov. 1980.

--- , NUREG-0800, "Standard Review Plan for Review of Safety Analysis Reports for Nuclear Power Plants---LWR Edition" (includes Branch Technical Positions), July 1981; Supplement 1, Sept. 1982.

--- , NUREG-0847, "Safety Evaluation Report Related to the Operation of Watts

Bar Nuclear Plant, Units 1 and 2," June 1982.

--- , NUREG/CR-0660, "Enhancement.of Onsite Emergency Diesel Generator Reliability," University of Dayton Research Institute, Feb. 1979.

--- , Regulatory Guide 1.63, "Electric Penetration Assemblies in Containment

Structures for Light-Water-Cooled Nuclear Power Plants."

Westinghouse Topical Report WCAP-9, "Topical Report for Overpressure Protection

for Westinghouse Pressurized Water Reactors," Rev. 1, Oct. 8, 1971.

Watts Bar SSER 2 B-1
--- , Topical Report WCAP-7982, "Final Densification Penalty Model," Oct. 1972.

--- , Topical Report WCAP-8346, "Evaluation of Fuel Rod Boring," May 1974.

--- , Topical Report WCAP-8377, "Revised Clad Flattening Model," July 1974.

--- , Topical Report WCAP-8692, "Fuel Rod Bow Evaluation," July 1979.

--- , Topical Report WCAP-8720, "Improved Analytical Models Used in Westinghouse

Fuel Rod Design Computations."

--- , Topical Report WCAP-8963-A, "Safety Analyses for Revised Fuel Rod Internal

Pressure Design Basis," Aug. 1978.(proprietary).

Watts Bar SSER 2 B-2
NRC Personnel

T.. Chan

J. Vogelwede

W. Jensen

H. Li

J. Knox

R. Giardina

J. Spraul

M. Duncan

APPENDIX E

PRINCIPAL CONTRIBUTORS

Branch

Auxiliary Systems

Core Performance

Reactor Systems

Instrumentation and Control

Power Systems

Power Systems

Quality Assurance

Licensing Branch No. 4

Watts Bar E-1 SSER2

APPENDIX G

ERRATA TO WATTS BAR SAFETY EVALUATION REPORT

Section Page Change

4.3.2.1 4-10 In the second paragraph from the bottom, change "peaking

factors less than 2.32" to "peaking factors less than 2.31".

5.4.3 5-17 In the last paragraph, the last sentence should read, "If

the alternate bypass lines are utilized, the staff requires that the malfunctioning main isolation valve be corrected

and the valve in the bypass line be closed (with power

removed) before the plant is repressurized".

9.5.1.2 9-29 Under "Elevation 692 ft", change "charging pump room" to

"charging pump rooms."

15.2 15-2 Delete Item (9) "high steam generator water level" at top

of page.

17.2 17-5 In the third paragraph, delete the words "drawings, specifications" from the sentence, "They review drawings, speci- fications, purchase requisitions, and plant instructions

and procedures covering activities such as test, calibration, special processes, maintenance, modification, and

repair for compliance with the QA program requirements."

Watts Bar SSER 2 G-1

NRC FORM 335 1. REPORT NUMBER (Assignedbv DDCj (7-77) U.S. NUCLEAR REGULATORY COMMISSION NUREG-0847

BIBLIOGRAPHIC DATA SHEET Supplement No. 2

4. TITLE AND SUBTITLE (Add Volume No., if appropriate) 2. (Leave blank)

Safety Evaluation Report related to operation of

Watts Bar Nuclear Plant, Units 1 and 2 3. RECIPIENT'S ACCESSION NO.

7. AUTHOR(S) 5. DATE REPORT COMPLETED

MONTH j YEAR

January 1984 9. PERFORMING ORGANIZATION NAME AND MAILING ADDRESS (Include Zip Code) DATE REPORT ISSUED

Division of Licensing MONTH YEAR

Office of Nuclear Reactor Regulation January 1984

U.S. Nuclear Regulatory Commission 6. (Leave blank)

Washington, D. C. 20555

8. (Leave blank)

12. SPONSORING ORGANIZATION NAME AND MAILING ADDRESS (Include Zip Code)

10. PROJECT/TASK/WORK UNIT NOSame as 9 above

11. CONTRACT NO.

13. TYPE OF REPORT . PERIOD COVERED (Inclusive dates)

Safety Evaluation Report, Supplement No. 2 1

15. SUPPLEMENTARY NOTES 14. (Leave blank)

Pertains to Docket Nos. 50-390 and 50-391

16. ABSTRACT (200 words or less) This report supplements the Safety Evaluation Report, NUREG-0847 (June 1982) and Supple- ment No. 1 (September 1982), issued by the Office of Nuclear Reactor. Regulation of the U.S. Nuclear Regulatory Commission with respect to .the application filed by the Tennessee

Valley Authority, as applicant and owner, for licenses to operate the Watts Bar Nuclear

Plant, Units 1 and 2 (Docket Nos. 50-390 and 50-391). The facility is located in Rhea

County, Tennessee, near the Watts Bar Dam on the Tennessee River. This supplement pro- vides recent information regarding resolution of some of the open and confirmatory items

and license conditions identified in the Safety Evaluation Report.

17. KEY WORDS AND DOCUMENT ANALYSIS 17a. DESCRIPTORS

17b. IDENTIFIERS/OPEN-ENDED TERMS

18. AVAILABILITY STATEMENT 19.0ECYRITYj:HkSYTh,s ncassi report) 21. NO. OF PAGES lea

Unlimited

20.6KFCLW T Y .j-SSd(Th,s page) 22. PRICE

R MS NRC FORM 335 (7 77).