NUREG-0847

Supplement No. 3

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 1985

o REG

- -ý
NOTICE

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NUREG-0847

Supplement No. 3

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 1985

ABSTRACT

This report supplements the Safety Evaluation Report, NUREG-0847 (June 1982), Supplement No. 1 (September 1982), and Supplement No. 2 (January 1984) issued

by the Office of Nuclear Reactor Regulation of the U.S. Nuclear Regulatory Commission wi.th 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 3 iii

TABLE OF CONTENTS

Page

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

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

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

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

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

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

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

2.4 Hydrologic Engineering ...................................... 2-1

2.4.8 Design Basis for Subsurface Hydrostatic Loading ...... 2-1

2.5 Geology and Seismology ...................................... 2-1

2.5.4 Stability of Subsurface Materials and Foundations .... 2-1

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

3.2 Classification of Structures, Systems, and Components ....... 3-1

3.2.1 Seismic Classification. .............................. 3-1

3.2.2 System Quality Group Classification .................. 3-1

3.8 Design of Seismic Category I Structures ...................... 3-2

3.8.1 Steel Containment .................................... 3-2

3.9 Mechanical Systems and Components ........................... 3-3

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

Component Supports, and Core Support Structures ...... 3-3

3.10 Seismic and Dynamic Qualification of Safety-Related

Electrical and Mechanical Equipment ....... I ........ ......... 3-3

3.10.1 Generic Concerns .................................... 3-4

3.10.2 Specific Concerns ................................... 3-6

3.10.3 Summary .............................................. 3-13

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

•4.4 Thermal-Hydraulic Design ....... I ......................... 4-1

Watts Bar SSER 3 V
TABLE OF CONTENTS (Continued)

Page

4.4.5 Loose Parts Monitoring System ........................ 4-1

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

6.2 Containment Systems ......................................... 6-1

6.2.1 Containment FunctionalDesign ........................ 6-1

6.2.4 Containment Isolation System ......................... 6-2

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

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

7.3.5 IE Bulletin 80-06 .................................... 7-1

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

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

8.3 Onsite Power Systems........................................ 8-2

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

8.3.3 Common Electrical Features and Requirements .......... 8-3

9 AUXILIARY SYSTEMS ................................................. 9-1

9.1 Fuel Storage Facility ....................................... 9-1

9.1.4 Fuel-Handling System ................................. 9-1

9.3 Process Auxiliaries ......................................... 9-1

9.3.2 Process Sampling System ............................... 9-1

9.5 Other Auxiliary Systems ..................................... 9-5

9.5.7 Emergency Diesel Engines Lubricating Oil System ...... 9-5

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

13.5 Plant Procedures ............................................ 13-1

13.5.3 NUREG-0737 Items ................................... 13-1

14 INITIAL TEST PROGRAM ............................................. 14-1

15 ACCIDENT ANALYSIS ............................................... 15-1

15.3 Limiting Accidents ......................................... 15-1

Watts Bar SSER 3 vi
TABLE OF CONTENTS (Continued)

Page

15.3.2 Steamline Break........................ ... 15-1

15.3.6 Anticipated Transients Without Scram ................. 15-2

15.4 Radiological Consequences of Accidents ...................... 15-6

15.4.3 Steam Generator Tube Rupture ........................ 15-6

APPENDICES

A CHRONOLOGY OF RADIOLOGICAL REVIEW OF WATTS BAR NUCLEAR PLANT, UNITS 1

AND 2, OPERATING LICENSE REVIEW

B BIBLIOGRAPHY

C NUCLEAR REGULATORY COMMISSION UNRESOLVED SAFETY ISSUES

E PRINCIPAL CONTRIBUTORS

LIST OF TABLES

2.1 Summary of Groundwater Level Estimates for Essential Raw Cooling

Water Pipeline.................................................2-5

2.2 Foundation Damping for Lumped Mass Modal Analysis .................. 2-6

Watts Bar SSER 3 vii

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 (TVA or the applicant) for licenses to operate the Watts Bar Nuclear Plant, Units 1 and 2. The Safety Evaluation Report (SER)

was supplemented by Supplement No. 1 (SSER 1, September 1982) and Supplement

No. 2 (SSER 2, January 1984), 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 the resolution or

status of some of the open and confirmatory items and license conditions identified in the SER and its supplements. 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.* Appen- dix C is an update to the status of the unresolved safety issues that were discussed in the SER. Appendix E is a list of principal contributors to this sup- plement. No changes in SER Appendices D, F, and G 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 five 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.

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

this report.

Watts Bar SSER 3 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

Status

Resolved (SSER 3)

Under review

Under review

Seismic - partially

resolved (SSER 3)

Environmental - under

review

Under review

Under review

Under review

Resolved (SSER 3);

see License

Condition (8)

Under review

Resolved (SSER 2)

Partially resolved

(SSER 2, SSER 3)

Under review

Under review

Partially resolved

(SSER 3)

Resolved (SSER 1)*

Section

2.5.4.4

3.10

(8) BTP CSB 6-4 6.2.4

(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

8.2

9.5.7

*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 3 1-2
Issue ,Status Section

(16) Boron dilution event

(17) Q List

Under review

Resolved (SSER 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 14 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 Status

(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

Resolved (SSER 3)

Resolved (SSER 3)

Resolved (SSER 3)

Resolved (SSER 3)

Awaiting verification

of installation

Awaiting verification

of installation

Resolved (SSER 2)

Resolved (SSER 3)

Under review

Resolved (SSER 2)

Section

2.4.8

2.5.4.2

2.5.4.2

2.5.4,.3

3.2.1,

3.2.2

3.2.1

3.8.1

*TVA has recently submitted a revised quality assurance program. However, the

program approved in SSER 2 is acceptable for use pending approval of the new

program.

Watts Bar SSER 3 1-3
Issue

(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

(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) Insulation 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

Status

Resolved (SSER 2)

Resolved (SSER 2)

Resolved (SSER 3); see

License Condition (42)

Awaiting verification

of installation

Awaiting information

Resolved (SSER 2)

Resolved (SSER 3)

Under review

Under review

Resolved (SSER 2)

Under review

Resolved (SSER 2)

Secti on

4.4.5

6.2.1.1

Resolved (SSER

Resolved (SSER

Under review

Resolved (SSER

Resolved (SSER

2)

3)

2)

2)

7.3.5

Awaiting verification

of test results

Awaiting verification

of acceptability of

test results

Resolved (SSER 2)

Watts Bar SSER 3 1-4
Issue Status

(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

(36) Missile protection for diesel

generator vent line

(37) Component booster pump relocation

(38) Electrical penetrations documentation

(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

Resolved (SSER 2)

Section

Under review 8.3.3.2.2

Resolved (SSER 2)

Resolved (SSER 2)

Under review

Awaiting verification

of modifications

Awaiting verification

o~f modifications

Under review

See License Condition (22)

Awaiting verification

of procedure changes

Resolved (SSER 3)

8.3.3.6

14

Under review 8.3.3.1.1

1.9 License Conditions

In Section 1.9 of the SER and Supplement No. 1 to the SER, 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 Status Section

3.9.3.3 (1) Relief and safety valve testing

(II.D.1)

(2) Preservice/inservice testing of

pumps and valves

Resolved (SSER 3)

Under review

Watts Bar SSER 3 1-5
Condition

(3), Detectors for inadequate core

cooling (II.F.2)

(4) Inservice Inspection Program

(5) Installation of reactor coolant

vents (ll.B.1)

(6) Accident monitoring

instrumentation (II.F.1)

(a) noble gas monitor

(b) iodine particulate sampling

(c) high range incontainment

radiation monitor

(d) containment pressure

(e) containment water level

(f) containment hydrogen

(7) Modification to chemical

feedlines

(8) Containment isolation

dependability (II.E.4.2)

(9) Hydrogen control measures

(II.B.7)

(10) Status monitoring system

(11) Installation of acoustic

monitoring system (II.D.3)

(12) Diesel generator reliability

qualification testing at

normal operating temperature

(13) DC monitoring and annunciation

(i4) Possible sharing of dc control

power to ac switchgear

(15) Testing of associated circuits

Status

Awaiting information

Under review

Awaiting verification of

installation

Awaiting information

Awaitinq information

Awaiting verification of

installation

Awaiting verification of

installation

Awaiting verification of

installation

Awaiting verification of

installation

Under review

Section

Under review 6.2.4

Under review

Unchanged (SER)

Awaiting verification of

installation

Resolved (SSER 2)

Partially resolved

(SSER 3)

Resolved (SSER 3)

Resolved (SSER 3)

8.3.2.2

8.3.3.2.4

8.3.3.3

Watts Bar SSER 3 1-6
Condition

(16) Testing of non-Class 1E cables

(17) Low-temperature overpressure

protection/power supplies for

pressurizer relief valves and

level indicators (II.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

(NUREG/CR-0660)

(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.3.17)

(31) Initial test program

(32) Effect of high pressure injection

for small-break LOCA with no

auxiliary feedwater (II.K.2.13)

Watts Bar SSER 3 1-7

Status

Resolved (SSER 3)

Under review

Resolved (SSER 2)

Partially resolved

(SSER 3)

Awaiting information

Under review

Awaiting verification of

modifications

Unchanged (SER)

Awaiting information

Awaiting information

Unchanged (SER)

Under review

Under review

Awaiting verification of

modi fications

Resolved (SSER 3)

Partially resolved

(SSER 3)

Under review

Section

8.-3.3.3

8.3.3.4

9.3.2

13.5.3

14
Condition Status Section

(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

Under review

Under review

Under review

Under review

Awaiting verification

of modifications

Unchanged (SER)

There are additional issues for which license conditions have been determined

to be desirable to ensure that staff requirements are met during plant operation.

These license conditions are tabulated below with the relevant section of this

report.

(39) Control of heavy loads

(NUREG-0612)

(40) Anticipated transients without

scram (Generic Letter 83-28)

(41) Steam Generator Tube Rupture

(42) Loose parts monitoring system

Added in SSER 3

Added in SSER 3

Added in SSER 3

Added in SSER 3

9.1.4

15.3.6

15.4.3

4.4.5

Watts Bar SSER 3 1-8
2 SITE CHARACTERISTICS

2.4 Hydrologic Engineering

2.4.8 Design Basis for Subsurface Hydrostatic Loading

The SER lists the design-basis groundwater level for the essential raw cooling

water (ERCW) pipeline as a confirmatory issue. The applicant had not provided

any groundwater information for the ERCW pipeline and in lieu of factual data,

the staff conservatively assumed ground elevation as the design groundwater

level for the ERCW pipeline.

The applicant subsequently installed eight piezometers along the ERCW pipeline

to establish groundwater levels. These data were also used in conjunction with other long-term groundwater and rainfall records to project an approximate

25-year (probability of 0.04 per year) groundwater level for use with combined

seismic events for design purposes. The 25-year groundwater level was chosen

on the basis of the combined events criteria in American Nuclear Society (ANS) Standard 2.8,,"Standards for Determining Design Basis Flooding at Power Reactor

Sites." Although that standard does not specifically address the combination

of a seismic event and a high groundwater level, it does consider a safe shutdown earthquake (SSE) combined with a 25-year flood to be an acceptable combination when considering dam failure floods.

Amendment 50 to the Watts Bar Final Safety Analysis Report (FSAR), Section

2.5.4.6, contains a complete description of the analyses used to determine the

projected 25-year groundwater level for the ERCW pipeline. FSAR Table 2.5-73

shows the estimated 25-year groundwater elevation for each of the eight ERCW

piezometers. The staff has reviewed,the applicant's analyses and estimated

25-year groundwater elevations (listed in Table 2.1) and finds them to be

acceptable and to meet the requirements of General Design Criterion (GDC) 2.

Therefore, Confirmatory Item (1) is resolved.

2.5 Geology and Seismology

2.5.4 Stability of Subsurface Materials and Foundations

2.5.4.2 Subsurface Conditions and Geotechnical Properties

In the SER, the staff stated-that it would perform an audit of Tennessee

Valley Authority's (TVA's) calculations to confirm the adequacy of the proce- dures used for accounting for the material and geometric damping in the soilstructure interaction (SSI) analysis of *the soil- and pile-supported structures.

The staff performed this audit on September 24, 1982, and, on the basis of its

review of the information during the audit, the staff finds that the applicant

used the following analysis techniques to analyze the soil-supported and pilesupported Category I structures.

Watts Bar SSER 3 2-1
Structure Analysis technique

Diesel generator Lumped mass modal analysis

building with equivalent soil springs

Waste-packaging Lumped mass modal analysis

area with equivalent soil springs

Condensate Lumped mass modal analysis

demineralizer waste with equivalent soil springs

evaporator (CDWE) and finite element analysis

building (pile supported) (LUSH (Lysmer et al., 1974))

Refueling water Lumped mass modal analysis

storage tank with equivalent soil springs

and finite elements analysis

(FLUSH (Lysmer et al., 1975))

The calculated soil geometrical and material damping and the design damping

values used in the analyses of various structures are shown in Table 2.2. In

the SSI analyses, the applicant varied the shear wave velocity of the soil by

a minimum of ±30% to account for uncertainties in the dynamic properties of

the soil. On the basis of its review of the damping values used in the analyses (Table 2.2) and in view of the applicant's use of ±30% variation in the

values of the soil shear wave velocities, the staff finds that the applicant

has used reasonable assumptions for dynamic modulus and damping values in his

soil-structure interaction analyses and, therefore, considers Confirmatory

Item (2) closed.

2.5.4.3 Foundation Construction and Evaluation

Foundations on Rock

In March 1982, the applicant informed the staff that t~he intended design differential settlement of 1 in. between rock-supported structures was not actually

used in the design of piping and electrical components passing between adjacent

rock-supported structures. The affected components include heating, ventilation,

and air conditioning (HVAC) duct, cable trays, Category I piping, instrument

lines, and conduits that pass between the reactor building of Unit 1 and the

auxiliary building and between the reactor building of Unit 2 and the auxiliary

building.

By letter dated September 14, 1983, the applicant notified the staff that the

capability of individual safety-relatedcomponents located at the interfaces

of rock-supported structures has been evaluated. On the basis of this evaluation, the applicant concluded that all affected safety-related HVAC duct, cable

trays, and their supports can safely withstand a 1-in. differential settlement.

In addition, the applicant had previously indicated in the Watts Bar FSAR that

the expected differential settlement of adjacent rock-supported structures,

based on settlement computations, is less than 0.3 in. On the basis of this

information and the applicant's analysis presented to the staff on September 16,

1983, the staff agrees with the applicant that he has adequately addressed

Watts Bar SSER 3 2-2
settlement effects on the Category I piping, conduit, and instrumentation lines.

The staff finds the applicant's analysis and the results acceptable and considers Confirmatory Item (4) closed.

Concrete Wingwalls and Sheetpile Retaining Walls

As discussed in the SER, the applicant informed the staff by letter dated

April 27, 1982, that he had discovered errors in the final design calculations

for the two soil-anchored sheetpile walls that extend from each end of the intake pumping station toward the main plant and retain backfill surrounding

Category I pipes. On the basis of revised calculations, the applicant concluded

that the sheet-pile walls will be overstressed in bending during an SSE. The

applicant committed to institute corrective field measures to reinforce these

walls to eliminatesuch overstressing. The staff found this approach reasonable

and acceptable.

During an audit of the applicant's calculations on September 24, 1982, the

staff reviewed the design procedures, assumptions, and input soil parameters

used by the applicant for the tied sheetpile wall design. At this meeting,

the applicant informed the staff that he has investigated the stability of

sheetpile walls using saturated soil to an elevation of 700 ft on one side,

with no water on the opposite side. The design is based on Coulomb's equation

for calculating earth pressures using angle of internal friction, d, for the

soil of 320. The staff informed the applicant that the design procedures,

assumptions, and input soil parameters used in the analyses are generally

acceptable. However, the staff requested the applicant to investigate further

the effects of the following conditions on the stability of sheetpile walls:

(1) reduction in the available soil passive pressure when the water level

rises above the normal water level as a result of flooding

(2) blocked weep holes

The applicant provided responses to this staff request by letters dated November 30, 1982, and July 21, 1983. On the basis of its review of this response

and previously submitted information, the staff finds that the applicant has

adequately analyzed the effect of potential reduction in the available soil

passive pressures when the water level rises above the normal water level as a

result of flooding. The applicant has determined that under these conditions,

the sheetpile wall has a factor of safety of 1.00, which the staff considers

adequate for these loading conditions. In the letter dated July 21, 1983, the

applicant also stated that he has investigated the stability of sheetpile

walls for the case where the weep holes were not properly functioning and has

determined that the sheetpile walls would have a factor of safety of 1.68 for

this case. The staff finds these design loading assumptions andanalyses

results reasonable and acceptable. On the basis of the results of its FSAR

review, the confirmatory audit of the applicant's calculations on September 24,

1982, and a review of the information submitted by the applicant since the

audit meeting, the staff concurs with the applicant's conclusion that the

sheetpile walls have an adequate margin of safety under the design static and

dynamic loading conditions and, therefore, considers Confirmatory Item (3)

closed.

Watts Bar SSER 3 2-3
2.5.4.4 Liquefaction Potential

ERCW Pipeline and Class 1E Electrical Conduit Support

During a meeting on May 20, 1983, the applicant informed the staff that he had

reassessed the liquefaction potential of soils below the water table in all the

borings along the ERCW pipeline and electrical conduit duct banks. For these

analyses, the applicant used the Seed and Idriss (1971) procedure and assumed

a maximum horizontal peak acceleration of O.4g at the ground surface. The

applicant found, in using this analysis criterion, that soils may liquefy in

several areas along the ERCW pipelines. One of the susceptible areas was found

to be continuous beneath slopi.ng ground close to the ERCW pumping station. In

other areas, the liquefaction-susceptible soils are either isolated or under

very flat ground. The applicant indicated that he would provide field remedial

measures (an underground barrier) to stabilize the area between the pumping

station and boring #141. The applicant also agreed to address the effects of

postearthquake settlement of liquefied soils.

The staff completed its review of the applicant's-use of a maximum ground acceleration of O.4g at the soil surface for evaluating liquefaction potential. By

letter dated June 22, 1983, the staff informed the applicant that his use of a

peak horizontal acceleration of O.4g for liquefaction analysis is appropriate

and acceptable.

The staff has reviewed the applicant's use of the Seed and Idriss (1971) method

for liquefaction analysis and concluded that the use of this method for evaluating liquefaction potential at the Watts Bar facility and the conclusions based

on this analysis are acceptable.

The field remedial measures described in Amendment 50 of the Watts Bar FSAR

show the construction of underground barriers along the ERCW pipelines. The

purpose of these measures is to contain the soil beneath the pipelines in case

of any liquefaction in this area. The applicant designed the barriers to withstand a maximum ground acceleration of 0.4g. The compacted soil embankment is

about 100 ft wide at the base and about 25 ft deep from the ground surface to

underlying shale deposit. The distance of the barriers from the ERCW pipelines

varies from 50 ft to 500 ft. The compacted barriers are formed by compacting

native borrow material from the location of the barriers. The compaction criterion is 100% standard Proctor density. The applicant has filled ground surface depressions in the area between the location of the pipelines and the top

of the barriers to preclude potential ground movement in this area. The applicant also has provided appropriate modification to accommodate anticipated

settlement effects.

The staff and its consultant, the U.S. Army Corps of Engineers, Tulsa District,

have reviewed the applicant's as-built drawings submitted with Amendment 50 of

the FSAR. On the basis of these reviews, the staff agrees with the applicant

that the underground barriers will provide sufficient confinement to any

liquefied soil in the affected area. Therefore, the staff concludes that Open

Item (1) is closed.

Watts Bar SSER 3 2-4
Table 2.1 Summary of groundwater level

estimates for essential raw

cooling water (ERCW) pipeline

ERCW

piezometer

number

25-year

groundwater

estimate

P1 702.9

P2 717.6

P3 716.8

P4 714.4

P5 712.5

P6 710.2

P7 718.4

P8 723.4

Watts Bar SSER 3 2-5
Table 2.2 Foundation damping for lumped mass modal analysis

Damping (% of critical)

Mode Predominant

Structure no. motion Geometrical' Material Design

Diesel generator 1 Horizontal translation 60 10 10

building of foundation

2 Rocking and horizontal 30 10 10

translation of

foundations

3 Primary building 5 10

deformation

Waste packaging 1 Rocking and horizontal 20 10 10

building translation of

foundation

2 Horizontal 20 10 10

translation of

foundation

3 Primary building 5 10

deformation

Condensate 1 Rocking and hori- 5 5

demineralizer zontal translation

waste evaporator of foundation

building

2 Rocking and horizontal 55

translation of

foundation

3 Vertical -7 5 5

Refueling water 1 Horizontal 10 10

storage tank translation of

foundation

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

3.2 Classification of Structures, Systems, and Components

3.2.1 Seismic Classification

As was noted in the SER, staff acccptance of the emergency raw cooling water

system (ERCWS) was contingent on a satisfactory upgrading or replacement of

portions of the ERCWS that service areas containing equipment required for plant safety. These portions of the ERCWS were incorrectly classified and installed

to nonseismic Category I design requirements. To meet the guidance of Position C.1.g of Regulatory Guide (RG) 1.29, the portions of the ERCWS required for

plant safety should be designed to withstand the effect of a safe shutdown earthquake (SSE) and remain functional, that is, seismic Category I. These portions of the ERCWS include heating, ventilation, and air conditioning (HVAC) air cooling units and chiller packages that provide cooling to areas that contain safety- related equipment such as safety injection pumps, containment spray pumps, residual heat removal pumps, and auxiliary feedwater pumps.

The applicant has reviewed the documentation for HVAC equipment that cools areas

where equipment required for plant safety is located and has verified that this

HVAC equipment meets the seismic Category I requirements of the applicable design criteria and is in accordance with a quality assurance program that is

in compliance with.10 CFR 50, Appendix B.

The applicant has also reviewed the documentation for HVAC equipment that cools

areas where equipment not required for plant safety is located and has verified

that although this HVAC equipment is not designed to seismic Category I require- ments, it is seismically designed to maintain the pressure boundary integrity of

the ERCWS.

In Amendment 49 to the FSAR, the applicant revised Tables 3.2-2 and 3.2-2a to

reflect the current status of the ERCWS. The staff has reviewed FSAR Section 3.2, flow diagram Figures 9.2-1 through 9.2-4a, and the series of progress

reports on the improper classification of the ERCWS that were issued by the

applicant between April 24, 1981, and September 28, 1983 (letters dated Apr. 24, June 8, July 14, Sept. 2, and Dec. 9, 1981, and Feb. 17, Aug. 11, Oct. 19, Apr. 27, and Sept. 28, 1983), and finds the seismic classification of the ERCWS

acceptable. The staff will also verify that these actions have been implemented for the ERCWS. Pending verification of these actions, the staff concludes that

those portions of the ERCWS that are required to withstand the effects of an

SSE and remain functional are properly classified as seismic Category I in accordance with RG 1.29. This constitutes an acceptable basis for satisfying, in part, the requirements of GDC 2 and is, therefore, acceptable.

3.2.2 System Quality Group Classification

As was noted in the SER, staff acceptance of the ERCWS was contingent on a

satisfactory upgrading or replacement of portions of the ERCWS that service

Watts Bar SSER 3 3-1
areas containing equipment required for plant safety. These portions of the

ERCWS were incorrectly classified and installed to Quality Group D standards.

To meet the guidance of Position C.2.1 of RG 1.26, the portions of the ERCWS

required for plant safety should be constructed to Quality Group C standards,

that is, American Society of Mechanical Engineers, "Boiler and Pressure Vessel

Code" (ASME Code), Section III, Class 3.

The portion of the ERCWS that was installed incorrectly was piping between the

first two isolation valves of the chillers/coolers. Approximately 1,200 ft of

ERCWS piping was installed incorrectly as TVA Class M (Quality Group D) instead

of to Quality Group C standards, and about 400 of 900 welds were made during

this installation without welder identification. The applicant has replaced

all of the affected piping with new piping that is constructed to Quality

Group C standards. As was noted in Section 3.2.1, the staff has reviewed FSAR

Section 3.2, the appropriate flow diagrams, and the series of progress reports

on the improper classification of the ERCWS and finds the Quality Group C

classification of the replacement piping acceptable. The staff will also

verify that these actions have been implemented for the ERCWS. Pending verification of these actions, the staff concludes that those portions of the ERCWS

that are required to be constructed to Quality Group C standards are properly

classified in accordance with RG 1.26. This constitutes an acceptable basis

for satisfying the requirements of GDC 1 and is, therefore, acceptable.

3.8 Design of Seismic Category I Structures

3.8.1 Steel Containment

In the SER, the staff requested the applicant to provide an additional verification of his containment buckling analytical methodology. This request was made

because new information, discussed below, had been obtained through NRC-sponsored

research programs (NUREG/CR-2165). By letter dated May 16, 1984, the applicant

stated that he has reviewed the NRC research results and concluded that the new

information from the research had been already adequately accounted for in his

calculation.

The most important new information resulting from the research was that reinforcement of an opening according to the area replacement method of ASME Code,

Section III, might not restore original buckling strength. However, the applicant stated that the design by the Watts Bar steel containment contractor,

Chicago Bridge and Iron, compensated for the weakness of the area replacement

method by providing extra reinforcing members around the opening. This extra

reinforcing consisted of stiffening above and around the opening so as to avoid

stress concentration. Also, the vertical stiffening was placed in two spans

between ring stiffeners. In view of the fact that the applicant provided-an

additional safety margin for loading to account for potential shortcomings in

the calculational methodology for asymmetrical shell buckling, the staff has

concluded that the applicant's evaluation is acceptable and, therefore, considers

Confirmatory Item (8) to be closed.

Watts Bar SSER 3 3-2
3.9 Mechanical Systems and Components

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

Support Structures

3.9.3.3 Design and Installation of Pressure Relief Devices

Relief and Safety Valve Test Requirements

As required by NUREG-0737, "Clarification of TMI Action Plan Requirements,"

Item II.D.1, all pressurized-water-reactor (PWR) plant licensees and applicants

are required to demonstrate that their pressurizer safety valves, power-operated

relief valves (PORVs), PORV block valves, and all associated discharge piping

will function adequately under conditions predicted for design-basis transients

and accidents. In response to this requirement, the Electric Power Research

Institute (EPRI), on behalf of the PWR Owners Group, has completed a full-scale

valve testing program, and the PWR Owners Group submitted these test results to

the NRC by letter dated April 1, 1982. Additionally, each PWR plant applicant

for an operating license (OL) was required to submit a report by fuel loading

that would demonstrate the operability of these valves and the associated piping.

On July 22, 1983, the applicant responded to this requirement with a submittal

that contains information from the EPRI valve test program results that applies

to Watts Bar Units 1 and 2. The submittal also states that the safety and relief

valve discharge piping and supports are being modified to ensure functionability.

The staff has not completed a detailed review of the applicant's submittal;

however, on the basis of a preliminary review, the staff finds that the general

approach of using the EPRI test results to demonstrate operability of the safety

valves, PORVs, and PORV block valves is acceptable. In the submittal the applicant notes that at Watts Bar Units 1 and 2, safety valves, PORVs, and PORV block

valves are used that are the same size and model as those that performed satisfactorily for test sequences considered representative or that bound conditions

to which the valves at Watts Bar Units 1 and 2 could be exposed.

In summary, on the basis of a preliminary review, the staff has concluded that

the applicant's general approach to responding to this TMI item is acceptable

and provides adequate assurance that the Watts Bar reactor coolant system overpressure protection systems can adequately perform their intended functions

until the staff completes its detailed review. If this detailed review reveals

that modification or adjustments to safety valves, PORVs, PORV block valves, or

associated piping are needed to ensure that the overpressure protection systems

can perform their intended functions, the staff will require that the applicant

make appropriate modifications. Therefore, the staff considers License Condition (1) to be resolved.

3.10 Seismic and Dynamic Qualification of Safety-Related Electrical and

Mechanical Equipment

In Section 3.10 of Supplement No. 1 and during the site audit of safety-related

electrical and mechanical equipment, several generic and specific concerns

regarding seismic and dynamic qualification were identified. These generic and

Watts Bar SSER 3 3-3
specific concerns are discussed in Supplement No. 1 and a letter to the applicant dated September 23, 1982. In an effort to resolve the generic and specific

concerns, the applicant provided post-audit information to the NRC in submittals

dated December 1, 1982, June 10, 1983, and February 9, May 17, May 25, and

June 19, 1984. The current status of each concern, along with the disposition,

when applicable, is discussed in the following sections.

3.10.1 Generic Concerns

The status of the generic concerns is:

(1) Single-frequency and single-axis testing was frequently used to qualify

electrical equipment. The applicant was requested to provide justification for using this form of testing for all equipment qualified in this

manner. The applicant responded by referring to qualification reports

(Westinghouse Topical Reports WCAP-8673, -8694, and -8624) that were previously submitted to the NRC in regard to this issue. The applicant also

referred to an NRC memorandum from J. P. Knight to R. C. DeYoung, dated

August 26, 1976, in support of a supplemental seismic demonstration program

to ensure that adequate margin exists for electrical equipment tested by

Westinghouse before May 1974. The staff received the necessary documents

and further discussed the issue during a meeting with Westinghouse representatives in Bethesda, Maryland, on April 28, 1983. During the meeting,

the applicant was requested to provide verification of analysis, including

the following:

(a) The effect of directional coupling should be considered if applicable.

(b) Where applicable, verification should be provided that acceleration

at each device location is less than 0.95g because relay chatter at

higher acceleration levels is expected.

(c) The test response spectrum (TRS) envelopes the required response

spectrum (RRS) for all directions because the RRS for different directions can be different for buildings or cabinets.

The applicant was requested to verify this in writing for equipment at

Watts Bar procured from Westinghouse. By letter dated May 17, 1984, the

applicant responded to this request. The nature of the response was acceptable, but the scope was not. The response to the concern about directional

coupling was limited to audit items. The response to the concern about

acceleration level was limited to devices mounted in the solid-state protection system. The response to the envelopment concern did not indicate

that Seismic Qualification Review Team forms for all safety-related equipment supplied by Westinghouse were provided. To be acceptable, the

responses must address the concerns for all safety-related equipment

supplied by Westinghouse. This issue remains open.

(2) In numerous cases the RRS were not broadened at the peaks to account for

uncertainty in the prediction of natural frequencies of the supporting

structures. In addition, the TRS should include sufficient margin to

Watts Bar SSER 3 3-4
account for uncertainty in the manufacturing process and testing apparatus.

The applicant's December 1, 1982, submittal stated that requirements were

imposed to ensure +10% broadening for balance-of-plant (BOP) equipment,

and it also addressed RRS broadening for certain specific nuclear steam

supply system (NSSS) items as discussed in Section 3.10.2. In that submittal the applicant did not adequately demonstrate that the broadening

requirement had actually been applied or considered in qualifying all

safety-related equipment. To better ensure that the broadening require- ment would pose no difficulties for any equipment, the applicant was

requested to evaluate the effects of peak broadening for a sampling of

equipment where RRS were not broadened. The applicant stated in the

June 10, 1983, submittal that the qualifications'for all Watts Bar safetyrelated equipment (BOP and NSSS) included consideration of spectra peak

broadening. In all cases the spectra were broadened by +10% throughout

the critical frequency range. Several examples of qualifications where

such broadening was applied, which were evidently representative of all

safety-related equipment, were provided with the submittal. On the basis

of the applicant's assurance that this policy was used throughout the

qualification program, this issue is closed.

(3) In numerous cases, particularly for electrical cabinets, equipment is field

mounted by welding but test mounted by.bolting. The applicant was asked to

identify and justify all such cases. In the December 1, 1982, submittal,

the applicant maintained that the welding used in all configurations is

stronger than the corresponding bolting used in the tests. The applicant

stated that more metal area is used in the welded connections and that,

if anything, the welded connections will increase rigidity of the cabinet

response. The staff requested that the applicant further evaluate whether

an increase in cabinet rigidity may shift the cabinet frequency into a more

critical range of the response spectrum curve, and also whether using welding could adversely affect damping in the system. Additionally, instead

of relying on the results of purely analytical predictions, the staff

required that the applicant select a cabinet and verify its natural frequencies by in situ testing. In the June 10, 1983, submittal, the applicant asserted that the base steel assemblies used to mount the cabinets are

rigid and that the welded-versus-bolted configuration would result in a

very small difference in equipment seismic response. The applicant

indicated that results of Westinghouse tests show that welding can be

substituted for bolting in these base steel assemblies with a 6% or smaller

change in equipment resonance frequencies. The applicant was requested,

however, to demonstrate by in situ tests on a Watts Bar cabinet that the

response of the cabinet is essentially unaffected by the difference in

mounting. By letter dated May 17, 1984, the applicant reiterated his

position. Since no new information has been presented, the request for

an in situ test is still felt appropriate. Discussion under equipmentspecific Item 5, "Main Control Board," in Section 3.10.2 is pertinent for

the resolution of this issue. This issue is open.

(4) Many safety-related equipment items, such as the insulation of motors,

transformers, and other electric devices, are age sensitive with respect

to their seismic performance. To ensure that safety-related equipment is

seismic resistant throughout the plant life, a detailed program of surveillance and maintenance should be provided for staff review and approval.

Watts Bar SSER 3 3-5
The applicant's submittal of May 17, 1984, did not address the equipment

located in a mild environment. Therefore, this issue remains open.

(5) The applicant was requested to provide a statement describing how damping

values were used in qualifications for seismic input as obtained from TVA

floor response spectra. The applicant stated that Westinghouse uses a 5%

critical damping value in the development of test response spectra. The

TRS is then compared to an RRS of equal or less damping to show acceptability. This approach complies with the recommendations of Institute of

Electrical and Electronics Engineers (IEEE) Standard 344-1975 and is therefore satisfactory. This concern is closed.

(6) Because nozzle loads exerted on safety-related equipment were frequently

not mentioned in the qualification documentation, the applicant was

requested to show how nozzle loads were considered in qualification. The

applicant responded that the nozzle loads criteria imposed on equipment

suppliers ensure that the equipment can sustain specified loads. These

same criteria are imposed on the piping analysis in the form of load limits

at the equipment interface. Problems that arise in the imposition of these

criteria are resolved on a case-by-case basis. This approach ensures that

acceptable values for nozzle loads are not exceeded. This concern is

closed.

(7) The status of safety-related equipment that is not yet seismically qualified

should be periodically updated and fully qualified before fuel loading.

The applicant has confirmed his completion of seismic qualification of

safety-related equipment by letter dated May 17, 1984. This issue is

closed.

3.10.2 Specific Concerns

The status of the specific concerns is:

(1) Reactor Trip Switchgear

(a) The applicant was requested to demonstrate that the welded field

mounting is structurally as sound, as the bolted lab mounting. Resolution of this item relies on the response to Item 3 in Section 3.10.1

and will be evaluated under that generic issue.

(b) A box-shaped cable support beam that extends downward from the floor

above and contacts the cabinet near its top should be addressed in

the cabinet qualification. By letter dated February 9, 1984, the

applicant stated that adequate clearance between the beam and cabinet

has been provided. In addition, the staff requested that a walkdown

audit on a sampling basis be conducted before fuel loading to confirm

that field modifications of this type have been made. This walkdown

audit (including all corrective actions) should be completed before

fuel loading, and the results should be maintained by the applicant

in an auditable manner and be available for NRC review. The applicant

was requested to provide written confirmation that this has been performed before fuel loading. The applicant committed to provide the

results of the walkdown in the May 17, 1984, submittal. Written

Watts Bar SSER 3 3-6
confirmation of the completion of these actions was provided in the

applicant's May 25, 1984, submittal. This issue is closed.

(c) The RRS should be peak broadened to adequately qualify the equipment.

In his submittal, the'applicant showed that the TRS sufficiently envelopes an RRS that has been broadened 10%. This issue is closed.

(d) The applicant was requested to provide an explanation as to how

damping was used to qualify for operational basis earthquake (OBE)

and SSE seismic input. The applicant's response was adequately encompassed by his response to Item 5 in Section 3.10.1. This issue

is closed.

(2) Reactor Protection System Cabinet

(a) The applicant was asked to demonstrate that field mounting is as

adequate as lab mounting. Resolution relies on the response to

Item 3 in Section 3.10.1 and will be evaluated under that generic

issue.

(b) The applicant was requested to evaluate the effects of mounting the

cabinets close together. The applicant responded that a minimum of

1/2-in. clearance between cabinets is required. The cabinets have

been inspected and adjusted accordingly. By letter dated February 9,

1984, the applicant provided written confirmation that this work has

been completed. This issue is closed.

(c) The applicant was requested to evaluate the degree of amplification

that occurred in the cabinet response motion during tests to clearly justify single-frequency testing. The applicant replied by referring

to his response to Item 1 in Section 3.10.1. The applicant also

stated that the RRS, when peak broadened by 10%, is still enveloped by the TRS. Resolution of this item relies on a satisfactory response

to Item 1 in Section 3.10.1 and will be evaluated under that generic

issue.

(3) Charging/Safety Injection Pumps

(a) The applicant was asked to provide a detailed analysis of the suction

nozzle connection using an approach appropriate for the nozzle geometry. The applicant responded in the December 1, 1982, submittal

that the suction nozzle was analyzed according to the requirements of ASME Code, Section III, Appendix A2212. This approach does not,

however, sufficiently address effects from externally applied nozzle

loads or consider the nozzle geometry. The applicant was therefore

requested to further evaluate the nozzle. The June 10, 1983, submittal identified external loads that were applied to the nozzle, which

included seismic, deadweight, and operating loads. The submittal did

not, however, clarify the acceptability of using the Bijlaard method

to perform a stress analysis at the connection when the nozzle geometry does not conform to guidelines established for the use of this

technique. Thus, the applicant was requested to justify the use of

the Bijlaard technique for this case. In the May 17, 1984, submittal,

Watts Bar SSER 3 3-7
the applicant responded that the Bijlaard technique was not used.

This being the case, justification for the technique is not necessary.

Furthermore, the concern about the shortcomings of the reference to

ASME Code, Section III, Appendix A-2212, was alleviated by a description of the nozzle analysis performed that was included in the June 10,

1983, submittal. This issue is closed.

(b) The applicant was requested to provide a comparison of the nozzle

loads used in the pump analysis and those obtained from piping

analysis. The applicant's December 1, 1982, submittal stated that

piping analysis results showed that piping loads imposed on the

pumps exceeded the nozzle load limit criteria provided to the pumps'

vendor. An evaluation by the vendor of the revised nozzle loads

showed the pumps to be acceptable. Therefore, this issue is closed.

(4) Control Rod Drive Mechanism

(a) The qualification documentation did not identify the load combinations used. The applicant replied that the faulted condition used

included loss-of-coolant-accident (LOCA), safe shutdown earthquake

(SSE), and deadweight loads. The LOCA and SSE loads were combined

by the square-root-of-the-sum-of-the-squares method, and the resultant

was added absolutely to the deadweight loads. Because this is an

acceptable methodology for determining load confirmation, this issue

is closed.

(b) The applicant was requested to verify that Westinghouse response

spectra envelope the Watts Bar floor-level response spectra. The

applicant's December 1, 1982, submittal stated that an analysis of

the control rod drive mechanism (CRDM) using Watts Bar response

spectra was performed in the same manner as the generic analysis.

The resulting stresses in the CRDM from the plant-specific analysis

are enveloped by those from the generic analysis. Thus, the CRDM

qualifies according to the Watts Bar seismic criteria. Therefore,

this issue is closed.

(c) The applicant was requested to provide a comparison of natural frequencies identified during pluck tests on the CRDM with calculated

natural frequencies. The applicant's December 1, 1982, submittal

stated that pluck tests were performed on a Westinghouse Model 1-105

CRDM. The natural frequencies identified by tests were thus somewhat

higher than those calculated for the longer L-106A model. The

difference in frequencies was accounted for by amplifying the response

of the longest CRDM by the ratio between the peak acceleration value

and the acceleration value used in the original analysis. The staff

requested that the applicant supply further information on the qualification safety margin so that the significance of any error introduced

by this approach could be assessed. In the June 10, 1983, submittal,

the applicant explained that the responses at all frequencies in the

4.26- to 10.8-Hz range were amplified by the ratio between the peak

spectrum acceleration in this range and the spectrum acceleration at

4.26 Hz (the lowest calculated natural frequency). This was done to

account for uncertainties in the determination of natural frequencies

in this range.

Watts Bar SSER 3 3-8
Although this approach is likely to be conservative, the applicant

was requested to provide the safety margin for this piece of equipment. The May 17, 1984, submittal indicated the margin is 1.10. The staff

finds this acceptable and considers this issue closed.

(5) Main Control Board

(a) For the qualification performed the panel was assumed to be fixed at

its base. However, the panel is attached to the floor with spot welds along only the inside edge of an angle-shaped member at the

base of the panel. The applicant was requested to consider potential flexibility that could be introduced by freedom of the outside edge of this member. The December 1, 1982, submittal stated that the

critical natural frequency of the panel was calculated to change from 21.1 Hz to 19.7 Hz as a result of.this flexibility, which does

not affect the panel qualification. The applicant was requested to

furnish these calculations for review. In'the June 10, 1983, submittal, the applicant furnished these calculations for review. In

these calculations, the stiffness of an equivalent rotational spring

at the cabinet base was calculated under the assumption that the

angle-shaped member was fixed along the inside edge and guided along

an outside edge. Were this member more conservatively assumed to be

fixed on the inside edgeand free on the outside edge, the frequency

calculation would yield 14.7 Hz rather than 19.7 Hz. A difference

of this magnitude could result in a significant effect on cabinet

response. This possible nonconservatism in the applicant's calculations is, however, countered to some degree by his assumption that

the angle-shaped member receives no support from the floor. To clear

the concern about flexibility at the base of this panel, the applicant

was requested to justify the assumption that the outside edge of the

angle-shaped member is guided. A justification was provided in the

May 17, 1984, submittal. However, the question discussed in the submittal of which value is correct (TVA has stated the conservative calculation would yield 16.7 Hz rather than 14.7 Hz) is not particularly pertinent. If the applicant's value Of 16.7 Hz is used, the shift in

frequency from 21.1 Hz represents a decrease in frequency of 21%.

This represents the bounding decrease, with smaller decreases being

based on engineering judgment. If this decrease percentage is applied

to the minimum frequency for the full control board (14.3 Hz), a value

of 11.32 Hz is obtained. Further, if a broadened RRS is compared with the response spectra (RS) of the analysis time history, the

broadened RRS is not enveloped by the time history RS under 12 Hz.

Therefore, the qualification of the control board cannot be ensured

without the exercise of engineering judgment. This being the case,

an in situ test of the board is requested to gain a measure of confidence for the minimum horizontal natural frequency of the board.

Note that such a test could also close Generic Concern No. 3. Thi.s

issue remains open.

(b) The applicant was requested to provide justification for not peak broadening the RRS. The applicant's submittal maintained that peak

broadening of the RRS for this piece of equipment was unnecessary.

The natural frequencies of the equipment are sufficiently removed,

from response spectrum peaks so that equipment response is not

Watts Bar SSER 3 3-9
affected by the shape of the spectrum near the peaks. The staff

agrees with the applicant's justification and considers this issue

closed.

(6) Electrical Penetrations

The applicant was requested to supply a comparison of site-specific RRS

and the TRS that shows adequate envelopment of the RRS by the TRS. The

applicant responded that the equipment was qualified to the generic RRS.

The generic RRS envelopes the Watts Bar floor-level RRS. Thus, qualification to the generic criteria ensures qualification to site-specific

criteria. This issue is closed.

(7) 125-V dc Vital Batteries

(a) The applicant was requested to supply a description of surveillance

programs that will be instituted to maintain seismic capability of

the vitalbatteries throughout plant life. As part of the applicant's

December 1, 1982, submittal, an annual fi-eld inspection procedure

for maintaining the batteries was provided. This procedure details

a maintenance program that, if followed, should ensure continued

seismic capability of the batteries. Therefore, this issue is

closed.

(b) The applicant was requested to provide verification that batteries

will have spacers installed, as was done during qualification tests.

By letter dated June 19,1984, the applicant stated that spacers

will be installed by December 1, 1984. Spacers are in place in the

emergency diesel generator battery assembly. As with Item (1)(b) in

this section, this item will be closed on confirmation of the field

modification. Because this confirmation has not yet been submitted,

this issue is open.

(c) The applicant was requested to explain whether the vendor had been

given release from the requirement that the battery cells have

positive anchorage. The December 1, 1982, submittal stated that the

Watts Bar battery specification does not require holddown hardware,

for anchorage. Side-and end rails located above the center of gravity

of the batteries, as used at Watts Bar, have been demonstrated by

qualification tests to sufficiently restrain the batteries. Therefore,

this issue is closed.

(8) Diesel Generator Control and Protection Relay Panel

(a) The applicant was requested to provide a comparison of an RRS and

the TRS and justification for the adequacy of single-frequency input

motion. In the December 1, 1982, submittal, the applicant provided

TRS curves for the single-frequency tests that envelope the RRS with

a large margin at the natural frequencies of the equipment. These

test results along with results from tests conducted at RRS peak

frequencies indicate that the anticipated equipment response to SSE

conditions was well exceeded during tests, even if multimodal excitation effects are considered. For this specific item, this issue

is closed.

Watts Bar SSER 3 3-10
(b) The applicant was requested to address the consequence of observed

relay chatter. The applicant responded that the relay exhibiting

chatter is used only during the testing phases of the diesel generator system and will not prevent proper operation of the system

during a seismic event. Therefore, this issue is closed.

(c) The applicant was requested to demonstrate that all relays (including

differential relays) will be functional in both the open and closed

positions. As part of the December 1, 1982, submittal, the applicant

furnished a copy of the test procedure that was followed during seismic

testing of the relays. The procedure indicates that the lockout and

auxiliary relays were tested in both-the energized and deenergized

modes. The protective relays that were tested in only the deenergized

mode because they do not perform any safety-related function in the

energized mode were reviewed carefully to determine that no safety

concerns arose as a result of a change of state from energized to

deenergized condition while the diesel generator is providing emergency

power. By letter dated May 17, 1984, the applicant indicated that these

relays could prevent safe operation of the diesel generators only if

other, fully qualified relays tripped'improperly. This issue is closed.

(9) Metal Clad Switchgear

The applicant was requested to describe the mounting method used during

testing and to show that the field mounting is as adequate as the test

mounting. The December 1, 1982, submittal stated that the test specimen

was attached to the test table with 10-1/2-in. bolts and that the cabinet

was field mounted in the same manner: Thus, the testing configuration

was representative. For this specific piece of equipment, this issue is

closed.

(10) Main Steam Isolation Valve

(a) The applicant was requested to provide a comparison of nozzle loads

with allowable values. In the December 1, 1982, submittal, the applicant supplied allowable values for bending moment and axial loads at

the nozzle that well exceed the values obtained from the piping

analysis. The applicant was requested to furnish similar information

on torsion and shear loads. In the June 10, 1983, submittal, the

applicant supplied a value for the-torsional moment that was applied

during qualification tests on this valve. By letter dated May 17,

1984, the applicant quoted values for the torsional moment and shear

force obtained from the piping analysis to ensure that the shear

stresses were enveloped by the test. The staff has reviewed these

values, finds them acceptable, and considers this issue closed.

(b) The applicant was requested to supply the maximum acceleration levels

at the valve location as obtained from the piping analysis results.

The applicant's submittal stated that the seismic response acceleration calculated for the main steam isolation valves (MSIVs) has a

minimum resultant magnitude of 2.68g, which lies well within the test

acceleration levels. This concern is closed.

Watts Bar SSER 3 3-11
(c) The applicant was requested to describe tests performed on the

32-in. MSIV and explain how these ensure operability of the valve

under Watts Bar seismic conditions. In the December 1, 1982, submittal, the applicant indicated that the valve was subjected to

internal pressure, the nozzle loads described in Item (1O)(a) above,

and static seismic loads in each direction. The seismic loads were

applied at the operator center of gravity. The valve was previously

shown to behave rigidly, and the applied loads exceed loads incurred

in the Watts Bar installation. Provided that the torsion and shear

loads mentioned in Item (1O)(a) above are properly addressed, these

tests adequately demonstrate qualification of the valve. This issue

is closed.

(11) Essential Raw Cooling Water Pump

The applicant was requested to perform an evaluation of the effects of

impact loads resulting from nonlinear supports along the pipe column.

The applicant responded by performing a conservative simplified calculation that showed the "downhole" portion of the pump and its gapped supports

would impact the impeller bearings, but, the impact load was well within

allowable loading on the impeller bearings. Operability of the pump in a

seismic event is, therefore, not impaired by impact loads. Therefore, this

concern is'closed.

(12) Diesel Combustion Air Intake Filter

Because the qualification report for this item was not available during

the plant audit, the applicant was requested to submit the qualification

documents for review. The applicant's December 1, 1982, submittal contained qualification documents that indicate that the filter was qualified

by stress analysis for seismic and deadweight loads. Calculations showed

that the filter, together with six vertical stiffeners, responds rigidly

to seismic loads. Seismic loads derived from the Watts Bar response spectra were applied statically to the filter, and resulting stresses were

found to lie well within acceptable levels. The documentation verified

that this item was adequately qualified. The vertical stiffeners were,

however, considered to be 4 x 4 x 3/8 in. in size in one document and

2 x 3 x 3/8 in. in another. The applicant was requested to resolve this

inconsistency. In the June 10, 1983, submittal, the applicant indicated

that the angle stiffeners were verified by field inspection to be

2 x 3 x 3/8 in. in size. He also identified the material (American Society

for Testing and Materials (ASTM) A36 steel) from which they are constructed.

The staff finds this acceptable and considers this issue closed.

(13) Barksdale Pressure Switch/Square D Relay

(a) The applicant was requested to supply details of biaxial tests and

items involved therein. The December 1, 1982, submittal indicated

that biaxial sine beat tests at SSE input levels were performed on

the Barksdale switch at all resonance frequencies. After initial

biaxial tests, the specimen was rotated 90' horizontally and retested.

Watts Bar SSER 3 3-12
No biaxial tests, however, were performed on the square D relays. This was justified on the basis that electrical relays inherently reflect a single axis failure mode, since contact chatter is due to

excitation along the axis parallel to that of contact motion. In

addition, a wide safety margin in seismic input to the panel during the relay tests alleviates concern that directional coupling in

panel response could increase loading on the relays to unacceptable levels. Therefore, the concern on single and biaxial testing is

closed.

(b) The applicant was requested to supply a justification for single- frequency tests. The December 1, 1982, submittal presented rationale

for this type of testing. The square D relays and the Barksdale

pressure switch were qualified in separate test programs. The

relays were contained in a control panel that was subjected to

high-level sine beat tests over a range of frequencies and at equip- ment natural frequencies. Single-frequency tests were justified by noting that one frequency predominates in the panel RRS. Because

equipment natural frequencies are well removed from this critical

frequency, the potential for multimodal excitation is minimal. Additionally, the TRS accelerations far exceed RRS accelerations at equip- ment natural frequencies. This justification for single-frequency

tests on the relays is reasonable.

The pressure switch was tested in a similar fashion. Again, single- frequency tests were justified by noting that the RRS exhibits one

predominant frequency. In this case, however, at least two of the

equipment natural frequencies lie at or near the RRS peak. This

situation creates the potential for, multimodal excitation and the

possibility that the single-frequency TRS do not adequately envelope

a broadened RRS. The applicant was requested to further evaluate

this issue. The June 10, 1983, submittal did not supply any new

information relative to this concern about multimodal excitation for

the pressure switch. The May 17, 1984, submittal would be acceptable if the applicant had demonstrated, using the unbroadened spectra, that multimodal response is not possible. The fact that the response spectra had been broadened does not, by itself, ensure that the

unbroadened spectra will not generate multimodal response. Th.is

issue remains open.

3.10.3 Summary

On the basis of the Seismic Qualification Review Team site audit and the submittals from the applicant, the staff concludes that an appropriate seismic

and dynamic qualification program has been defined and substantially implemented, with the exception of the above open issues. The open issues for both generic and equipment-specific items must be resolved before fuel loading.

Completion of the applicant's seismic and dynamic qualification program for

safety-related equipment is required to satisfy applicable portions of GDC 2, 4, 14, and 30 of Appendix A to 10 CFR 50 and Paragraphs XI of Appendix B to 10 CFR 50 and VI(a)(1) and (2) of Appendix A to 10 CFR 100 as they relate to

qualification of equipment.

Watts Bar SSER 3 3-13

4 REACTOR

4.4 Thermal-Hydraulic Design

4.4.5 Loose Parts Monitoring System

In the SER, the staff stated that before an operating license is issued, the applicant should provide a commitment to supply a report on the loose parts monitoring system (LPMS) that contains:

(1) an evaluation of the LPMS for conformance to RG 1.133, Revision 1

(2) a description of the operator training program

(3) a description of the system hardware operation and implementation of the loose parts detection program after startup testing

In a letter dated February 25, 1982, the applicant provided a brief description of the operator training program and the LPMS hardware and an evaluation of conformance to RG 1.133. By letter dated November 10, 1982, the applicant

Amendment 

submitted additional information. The Watts Bar FSAR was also revised in 49 to provide a more detailed description of the LPMS. The appli- cant's evaluation has shown that the LPMS meets Positions 1 through 4 in Sec- tion C of RG 1. 1 3 3 regarding the LPMS characteristics, alert level establishment, data acquisition, and safety analysis report content. With regard to Position 5, the limiting conditions for operation and surveillance requirement with respect to the LPMS have been incorporated into Technical Specifications 3.3.3.11 and 4.3.3.11. Position 6, "Notification of a Loose Part," has not been addressed

mit 

by the applicant. However, 10 CFR 50.73 requires that the applicant shall sub- a licensee event report (LER) within 30 days after the discovery of "any event or condition that resulted in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded, or that resulted in the nuclear power plant being: (a) in an unanalyzed condition that significantly compromised plant safety and (b) in a condition that was outside

primary 

the design basis of the plant." Because the presence of loose parts in the' system belongs to this category, an LER would be required according to 10 CFR 50.73. Because 10 CFR 50.73 became effective January 1, 1984, and replaces all existing requirements for applicants to report reportable occur- rences as defined in individual plant Technical Specifications, Position 6 is covered by this regulation.

On the basis of its review of the applicant's submittals, the staff has found that the LPMS is acceptable. However, to ensure compliance with RG 1.133, the license will be conditioned to require the applicant to submit to the NRC within 90 days following completion of the startup test program the results of the preoperational test and the alert level setting of the LPMS.

Watts Bar SSER 3 4-1

6 ENGINEERED SAFETY FEATURES

6.2 Containment Systems

6.2.1 Containment Functional Design

6.2.1.1 Containment Structure

Protection Against Damage From External Pressure

The containment vessel is designed for an external pressure of 2.0 psig. Inadvertent operation of the spray system and/or air return fan system during normal

plant operation would cause a reduction in the containment pressure. The applicant has stated that such transients would not cause containment external pressure to exceed the design value of 2.0 psig. The staff requested the applicant

to furnish the analysis that supports this statement. A simplified analysis by

the staff showed that the containment external design pressure value would not

be exceeded by the transients mentioned above. As a result, the staff considered this matter to be a confirmatory item.

By letter dated June 4, 1982, the applicant provided the requested analysis,

with which the staff concurs. The analysis shows that inadvertent actuation

of one or both trains of the containment spray system during normal plant

operation would reduce containment pressure to -1.3 psig, which is within the

design capacity of -2.0 psig.

The applicant has further calculated that inadvertent operation of.one air return fan during normal plant operation results in a containment internal pressure of -2.0 psig after about 790 sec (13 min). Because fan operation is

alarmed in the control room, operators would be alerted to the need to take

corrective action.

The staff considers 13 min to be sufficient time for the plant operators to

determine that inadvertent fan actuation has occurred and to terminate fan

operation.

Inadvertent operation of both air return fans is not considered likely because

of the redundancy in the electrical power supplies. This could only occur as a

result of a spurious containment high-high pressure signal or phase B isolation

signal. Nevertheless, a 10-min delay in fan operation is designed into the

signal, after which it would take another 6.5 min for containment pressure to

drop to -2.0 psig. Again, operators would be alerted and would have sufficient

time to terminate the transient.

The staff, therefore, concludes that the Watts Bar'design provides adequate protection against damage from external pressure transients and considers Confirmatory Item (17) resolved.

Watts Bar SSER 3 6-1
6.2.4 Containment Isolation System

The SER stated that the Watts Bar plant conforms with all the provisions of

Item II.E.4.2, "Containment Isolation Dependability," of NUREG-0737, "Clarification of TMI Action Plan Requirements," except for Section (6) concerning

containment purging and venting during normal operation. It was further

stated that the applicant had not fully addressed the provisions of Branch

Technical Position (BTP) CSB 6-4, "Containment Purging During Normal Plant

Operations" (NUREG-0800). The applicant has since provided the necessary

information, addressing point-by-point the provisions of BTP CSB 6-4. The

applicant has shown conformance with the provisions of BTP CSB 6-4, contingent

on a favorable finding by the staff as to the operability of the containment

purge/vent isolation valves, that is, that they are qualified to close against

LOCA-induced pressure transients inside containment if they are open at the

onset of an accident. Therefore, subject to resolution of the contingency

stated above, the staff finds that the purge/vent, system design conforms with'

the provisions of Item II.E.4.2 of NUREG-0737 and BTP CSB 6-4 and, therefore,

considers Open Item (8) resolved. Operability of the containment purge/vent

isolation valves is currently under review and will be addressed in a future

supplement to the SER under License Condition (8).

Watts Bar SSER 3 6-2
7 INSTRUMENTATION AND CONTROLS

7.3 Engineered Safety Features Actuation System

7.3.5 IE Bulletin 80-06

Office of Inspection and Enforcement (IE) Bulletin 80-06 calls for the review

of all safety equipment to determine which, if any, safety functions might be

unavailable after reset and what changes could be implemented to correct any

problems. In the SER, the staff concluded that the applicant's proposed design

modifications are acceptable subject to review of the electrical schematics,

which were not available at that time. In .FSAR Amendment 48, the applicant

submitted the electrical schematics that implement the following changes:

(1) For feedwater isolation valves (FCV-3-33, FCV-3-47, FCV-3-87, and

FCV-3-100), feedwater check valve bypass valves (FCV-3-185, FCV-3-186,

FCV-3-187, and FCV-3-188), and upper tap main feedwater isolation valves

(FCV-3-236, FCV-3-239, FCV-3-242, and FCV-3-245), a new reset switch and

a relay have been added for each steam generator loop. When the engineered

safety feature (ESF) signal is reset, the individual valve will not change

state until both the loop and the ESF train reset switches have been reset.

(2) For steam generator blowdown isolation valves (FCV-43-54D, FCV-43-56D,

FCV-43-59D, FCV-43-63D, FCV-43-55, FCV-43-58, FCV-43-61, and FCV-43-64),

the ESF signal is sealed in by means of a valve-mounted limit switch.

The individual valve will not-change state until a hand switch in the

sample room is used to reopen the individual valve.

(3) For residual heat removal heat exchanger outlet flow control valves

(FCV-74-16 and FCV-74-28), the ESF signal is sealed in by the limit

switch. A new reset switch has been added *at the control room control

board. When the ESF signal is reset, the individual valve will not

change state until the individual reset switch has been reset.

The staff has reviewed the electrical schematics and finds the design modification acceptable.

The applicant has also identified all the safety-related equipment that does

not remain in its emergency mode after ESF reset. The applicant evaluated

this equipment and determined that it does not impact the safety of the plant

or the ability to achieve and maintain safe shutdown. The staff has reviewed

the applicant's response and finds that the applicant's justification is

acceptable. Therefore, the staff considers Confirmatory Item (24) resolved.

Watts Bar SSER 3 7-1

8 ELECTRICAL POWER SYSTEMS

8.2 Offsite Electric Power System

As noted in Section 8.2.2.2 of the SER, by letters dated October 9, 1981, and January 7, 1982, the applicant proposed a system design change that adds two

service transformers with related distribution system components to the offsite

power system. In Amendments 48 and 49 to the FSAR, the applicant presented additional description and analysis relating to the proposed new design for

the offsite power system.

The following items address concerns raised during the staff review and their

resolution.

(1) Section 8.2.2 of the FSAR states that, with degraded voltage conditions

on the 6.9-kV shutdown board (Class 1E bus), offsite power is automatically transferred from the normal to the first alternate to the second

alternate to the onsite diesel generator source. The same section of the

FSAR also states in contradiction that, on degraded voltage, offsite

power is only transferred from the normal to the diesel generator source.

This contradiction was discussed with the applicant during telephone conference calls. on October 27 and November 1, 1983. On the basis of these telephone conference calls, the staff concluded that the Watts Bar

design provides immediate automatic transfer to the preferred offsite circuits on degraded grid voltage. However, the Watts Bar design shown

in FSAR Figure 8.3-5a is not consistent with the applicant's verbal

description of the design provided during the above referenced telephone conference calls. This item will continue to be pursued with the applicant, and the results of the staff review will be reported in a future supplement.

(2) The routing of the two offsite circuits from the service transformers to

the Class 1E distribution system is through separate conduit and cable

trays of a single commonly supported raceway system. With a single commonly supported raceway system being used for both offsite circuits, the circuits are not located, in accordance with GDC 17, so as to minimize to

the extent practical the likelihood of their simultaneous failure under

operating and postulated accident conditions. The Watts Bar design, however, has other offsite circuits that are physically separated and that

can be made electrically independent and available to supply power on a

delayed basis. With the delayed availability of these other circuits, the staff concludes that this part of the Watts Bar design meets GDC 17

and is acceptable.

(3) Section 8.2.1.6 of the FSAR indicated only that control power associated

with the two offsite power circuits is provided by a 125-V dc supply for

the station service transformer and switchgears C and D. Information on

the physical and electrical separation of control power and independence between control power associated with onsite and offsite sources had not'

been described or analyzed in the FSAR in accordance with the requirements of GDC 17. By Amendment 49 to the FSAR, the applicant provided the required description and analysis, as discussed below:

Watts Bar SSER 3 8-1
The 125-V dc control power for the offsite power circuits is provided from

the onsite Class 1E power system. The control power cables are treated as

associated circuits and are routed in separate raceways. The staff finds

that this routing meets the independence requirements of GDC 17 for offsite

circuits and is acceptable.

In regard to supplying control power from the onsite Class 1E power system,

it was the staff's concern that design provisions were not provided, in

accordance with the requirements of GDC 17, to minimize the probability

of losing electrical power from the offsite system as a result of or concurrently with loss of the onsite power system supplies. By Amendment 49

to the FSAR, the applicant indicated that loss of control power will cause

failure of the automatic transfer system used to transfer offsite circuits

from the normal to the preferred offsite power sources and may necessitate

reestablishment of offsite circuits by manually closing breakers. Although

it may be preferable for control power to be from a supply independent of

the Class 1E system, the staff concludes that the Watts Bar design, which

requires loss of two independent Class 1E dc power supplies in order to

cause loss of both offsite circuits, minimizes the probability of losing

offsite power in accordance with the requirements of GDC 17 and is, therefore, acceptable.

8.3 Onsite Power Systems

8.3.2 Onsite DC System Compliance With GDC 17

8.3.2.2 DC System Monitoring and Annunciation

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

indications and alarms that are provided for the 125-V station battery system

be provided for the 125-V diesel generator battery system or, as a minimum,

the following abnormal conditions shall be alarmed in the control room:

(1) battery circuit open

(2) battery charger circuit open

(3) dc bus ground fault

(4) dc bus undervoltage

(5) dc overvoltage

(6) battery charger failure

(7) battery discharge

In addition, the staff required that the following parameters also be monitored:

(1) dc bus voltage

(2) battery circuit input current

(3) battery circuit output current

(4) battery charger output current

By letter dated January 15, 1984, the applicant indicated that the Watts Bar

design fully satisfies the above requirements. However, on the basis of

information presented in Amendment 48 to the FSAR, the staff finds that the

following items have not been included in the Watts Bar design.

(1) Battery circuit input current is not monitored.

(2) DC bus undervoltage is not alarmed in the control room.

Watts Bar SSER 3 8-2
Justification for not including these items will be pursued with the applicant,

and the results of the staff review will be reported in a future supplement.

8.3.3 Common Electrical Features and Requirements

8.3.3.1 Compliance With GDC 2 and 4

8.3.3.1.1 Submerged Electrical Equipment as a Result of a Loss-of-Coolant

Accident

In the SER, the staff indicated that the design for the automatic deenergization

of circuit loads as a result of a loss-of-coolant accident would be verified

as part of the staff's site visit/drawing review. This item will be discussed

with the applicant, and the results of the staff review will be reported in a

future supplement.

8.3.3.2 Compliance With GDC 5

8.3.3.2.2 Sharing of AC Distribution Systems and Standby Power Supplies

Between Units 1 and 2

In the SER, the staff indicated that sharing of onsite ac and dc systems had

not been adequately described or analyzed in Section 8.3 of the FSAR. By letter dated January 7, 1982, the applicant provided the subject description and

analyses. On the basis of this letter, the staff concluded in the SER that the

design meets the guidelines of RG 1.81 and was found acceptable pending revision

of the FSAR that reflects requirements of the shared safety systems. By Amendment 48 to the FSAR, the applicant partially documented the description and

analysis presented in the January 7, 1982, letter. By letter dated January 17,

1984, the applicant submitted preliminary copies of an amendment to the FSAR in

order to provide the remaining description and analysis.

The staff has determined that information presented in Amendment 48 and the

January 17, 1984, letter is consistent with information presented in the applicant's letter of January 7, 1982. This item is, therefore, acceptable pending

confirmation that the information in the January 17, 1984, letter is incorporated into the FSAR.

8.3.3.2.4 Possible Sharing of DC Control Power to AC Switchgear

In the SER, the staff required, as a condition in the license, that all possible interconnections between redundant divisions through normal and alternate

power sources to various loads be identified in the FSAR regardless of the

source of power and that these interconnections meet the following positions:

(1) The circuit breaker for the alternate feed shall be kept open and be

alarmed in the control room when closed.

(2) The manual transfer switch shall be alarmed in the control room when it is

in the alternate supply position.

By letter dated January 17, 1984, the applicant documented that al.l interconnections are identified in Tables 8.3-9 and 8.3-10 of the FSAR and meet the staff

Watts Bar SSER 3 8-3
position, except as noted and accepted in Section 8.3.3.2.4 of the SER. This

item is acceptable, and, therefore, License Condition (14) is no longer required.

8.3.3.3 Physical Independence (Compliance With GDC 17)

(2) Associated Circuits

The staff's evaluation of this item is included in Section 8.3.3.3(3) of this

report.

(3) Separation Criteria Between Class 1E and Non-Class 1E Circuits

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

devices (used to isolate non-Class 1E from Class 1E circuits) be of a high

quality commensurate with their importance to safety and be periodically

tested. By Amendment 48 to the FSAR, the applicant provided the results of a

reliability study to demonstrate that a single circuit breaker that is periodically tested has reliability equivalent to each of the following protective

device configurations:

(a) two series connected circuits breakers that are not tested

(b) a circuit breaker in series with a fuse that is not tested

(c) a single fuse that is not tested

On the basis of the results of this reliability study, the staff concludes

that each of the above configurations provides equivalent protection and is

acceptable. The applicant by letter dated January 17, 1984, has stated:

(a) The protective devices of the 160 associated circuits either have or are

being modified to have one of the above protective device configurations.

(b) The protective devices of non-Class 1E circuits that are routed closer to

Class 1E circuits than allowed by RG 1.75 have been identified and will

be periodically tested, except those that have one of the above protective

device configurations.

The staff concludes that this item is acceptable, and, therefore,.License

Conditions (15) and (16) are no longer required.

8.3.3.4 Compliance With NUREG-0737 Items

Emergency Power for Pressurizer Equipment (II.G.1)

To meet the guidelines of both TMI Item II.G.1 and Branch Technical Position

RSB 5-2, the staff in the SER required as a condition to the license that the

power supplies for the power-operated relief valves (PORVs) and block valves

be from the same power train but from different buses. By letters dated June 9,

1983, and January 17, 1984, the applicant stated that the Watts Bar design does

not comply with the above requirements and indicated his disagreement with the

requirement because of safety reasons. The basis presented in the applicant's

June 9, 1983, letter is inconsistent with the current design being implemented

at the Watts Bar facility. By letter dated October 24, 1984, the staff informed

TVA of the basis of its position. This item will be pursued with the applicant,

and the results of the staff review will be reported in a future supplement.

Watts Bar SSER 3 8-4
8.3.3.6 Compliance With GDC 50

In the SER, the staff required a reevaluation of the penetrations' capability

to withstand, without seal failure, the total range of available time-current

characteristics assuming a single failure of any overcurrent protective device.

By Amendment 48 to the FSAR, the applicant documented the results of the reevaluation. On the basis of these results, discussions with the applicant,

and information submitted by letter dated January 17, 1984, the staff concludes

that all circuits (Class 1E, non-Class 1E, normally energized, and normally

deenergized) that pass through containment electric penetrations have been

designed with the required capability, contain the required primary and backup

protective devices (except low-energy instrument circuits), meet the guidelines

of Position 1 of RG 1.63, and are, therefore, acceptable pending confirmation

that information in the January 17, 1984, letter is incorporated into the

FSAR.

Watts Bar SSER 3 8-5

9 AUXILIARY SYSTEMS

9.1 Fuel Storage Facility

9.1.4 Fuel-Handling System

In the SER, the staff stated that the applicant would submit the results of the

review against the guidelines of NUREG-0612, "Control of Heavy Load at Nuclear

Power Plants," by August 1982. The staff further stated that the applicant's

commitment to implement the interim actions of NUREG-0612 before issuance of

an operating license provides reasonable assurance of safe handling of heavy

loads until NUREG-0612 can be fully implemented. By letters dated February 6,

and March 20, 1984, the applicant submitted the required information regarding

Sections 5.1.1 and 5.3 (Phase I) and Sections 5.1.2 through 5.1.6 (Phase II) of

NUREG-0612. This information is currently under review and will be the subject

of a future safety evaluation. However, to ensure completion of Phase I in a

timely manner, the staff will require that the following license condition be

placed in the license:

The applicant will meet the guidelines of Sections 5.1.1 and 5.3 of

NUREG-0612 (Phase I) before the first refueling outage.

9.3 Process Auxiliaries

9.3.2 Process Sampling System

Postaccident Sampling Capability (II.B.3)

In the SER, the staff determined that the postaccident sampling system (PASS)

met 5 of the 11 criteria for Item II.B.3 in NUREG-0737. The following six

criteria were unresolved.

(1) Criterion (2) - Provide a procedure for relating radionuclide gaseous and

ionic species to estimated core damage.

(2) Criterion (5) - Verify that chloride analysis can be completed within 4 days

following an accident which requires postaccident sampling.

(3) Criterion (8) - Demonstrate the capability of analyzing the grab samples

and verify that equipment provided for backup sampling shall be capable

of providing at least one sample per day for 7 days following onset of the

accident and at least one sample per week until the accident condition no

longer exists.

(4) Criterion (9)(a) - Verify that the sensitivity of onsite liquid sample

analysis capability is such as to permit measurement of nuclide concentration of the range from approximately I pCi/g to 10 Ci/g.

and

Watts Bar SSER 3 9-1
Criterion (9)(b) - Verify that provisions are available to restrict background radiation levels such that the sample analyses will provide results

with a range of accuracy within a factor of 2.

(5) Criterion (10) - Describe the procedures for onsite radiological and

chemical analyses and provide the accuracy, range, and sensitivity of these

analyses in an accident chemistry and radiation environment (that is, the

presence of large amounts of fission products and a high radiation field

in the samples).

Provide information on testing frequency and type of testing to ensure

long-term operability of the postaccident sampling system and on operator

training requirements for postaccident sampl~ing.

(6) Criterion (11)(a) - Verify that the residues of sample collection will be

returned to containment or to a closed system.

and

Criterion (11)(b) - Verify that the ventilation exhaust from the sample

station will be filtered with charcoal adsorbers and high-efficiency

particulate air filters..

By letters dated October 29, 1981, September 20, 1983, December 19, 1983, and

July 13, 1984, the applicant provided additional information to address these

items, as discussed below.

Criterion (2)

NUREG-0737 requires that the applicant establish an onsite radiological and

chemical analysis capability to provide, within a 3-hour time frame, quantification of the following:

(1) certain radionuclides in the reactor coolant and containment atmosphere

that may be indicators of the degree of core damage (e.g., noble gases,

iodines and cesiums, and nonvolatile isotopes)

(2) hydrogen levels in the containment atmosphere

(3) dissolved gases (e.g., H2 ), chloride (time allotted for analysis subject

to discussion below), and boron concentration of liquids

or have inline monitoring capabilities to perform all or part of the above

analyses.

The applicant has the capability to obtain and analyze samples (within 3 hours

of the time a decision is made to sample) from the reactor coolant system, containment sump, and containment atmosphere under accident conditions, and to

obtain grab samples for offsite analyses. The applicant also has onsite radiological and chemical analysis capabilities to quantify hydrogen levels in containment atmosphere; radioactive isotopes (noble gases, iodine and cesium

isotopes, and nonvolatile isotopes); and dissolved gases, chloride, and boron

concentrations in liquid samples.

Watts Bar SSER 3 9-2
In a letter dated December 19, 1983, the applicant provided an interim procedure for estimating the degree of reactor core damage from measured and predicted postaccident radionuclide concentrations from failed fuels. The applicant indicated that, as a member of the Westinghouse Owners Group, the generic methodology for estimating core damage that is being prepared by the Westing- house Owners Group will be adopted for the Watts Bar facility.

The staff has determined that these provisions meet Criterion (2); therefore,

the procedure for estimating core damage is acceptable on an interim basis.

The license will require the applicant to provide a final plant-specific proce- dure to estimate the extent of core damage on the basis of radionuclide concentrations and taking into consideration other physical parameters such as

core temperature data, sample location, and containment radiation levels and

hydrogen concentrations before restart following the first refueling outage.

Criterion (5)

NUREG-0737 states that the applicant shall provide for a chloride analysis within 24 hours of the taking of a sample (1) if the plant's coolant water is

seawater or brackish water and (2) if there is only a single barrier between

primary containment systems and the cooling water. In all other cases, the

applicant shall provide that the analysis be completed within 4 days. The

chloride analysis does not have to be done on site.

The applicant has stated that the chloride analysis will be performed within

4 days after an accident that requires postaccident sampling. This analysis will be done in line, with undiluted samples, using an ion chromatograph. The

staff finds these provisions meet Criterion (5) and are, therefore, acceptable.

Criterion (8)

NUREG-0737 requires that the applicant shall provide backup sampling through

grab samples if in-line monitoring is used for any sampling and analytical capability as specified, and shall demonstratethe capability of analyzing the

samples. Established planning for analysis at offsite facilities is acceptable. Equipment provided for backup sampling shall be capable of providing at least

one sample per week until the accident condition no longer exists.

The applicant has stated that a diluted and undiluted reactor coolant grab sample and undiluted containment atmosphere grab sample will be obtained for

analyses of boron, dissolved hydrogen, pH, chloride, and radioisotopes in the

reactor coolant and hydrogen, oxygen, and radioisotopes in the containment

atmosphere. The staff finds that these provisions meet Criterion (8) and are,

therefore, acceptable.

Criterion (9)

NUREG-0737 requires that the applicant's radiological and chemical sample

analysis capability shall include provisions to:

(1) Identify and quantify the isotopes of the nuclide categories discussed

above to levels corresponding to the source term given in RGs 1.3 or 1.4

and 1.7. Where necessary and practicable, the ability to dilute samples

to provide capability for measurement and reduction of personnel exposure

Watts Bar SSER 3 9-3
should be provided. Sensitivity of onsite liquid' sample analysis capability

should be such as to permit measurement of nuclide concentration in the

range from approximately 1 pCi/g to 10 Ci/g.

(2) Restrict background levels of radiation in the radiological and chemical

analysis facility from sources so that the sample analysis will provide

results with an acceptably small error (approximately a factor of 2).

This can be accomplished through the use of sufficient shielding around

samples and outside sources, and by the use of a ventilation system design

that will control the presence of airborne radioactivity.

The applicant has stated that'the radionuclides in both the primary coolant and

the containment atmosphere will be identified and quantified. Provisions are

available for diluted reactor coolant samples to minimize personnel exposure.

The PASS can perform radioisotope analyses at the levels corresponding to the

source terms given in RGs 1.4, Revision 2, and 1.7. These analyses will be

accurate within a factor of 2. The staff finds that these provisions meet

Criterion (9) and are, therefore, acceptable.

Criterion (10)

NUREG-0737 requires that accuracy, range, and sensitivity shall be adequate to

provide pertinent data to the operator in order to describe the radiological

and chemical status of the reactor coolant systems.

The applicant has stated that the accuracy, range, and sensitivity of the Watts

Bar PASS instruments and analytical procedures are consistent with the recom--

mendations of RG 1.97, Revision 3, and the clarifications of NUREG-0737,

Item II.B.3. Therefore, they are adequate for describing the radiological and

chemical status of the reactor coolant. The analytical methods and instrumentation were selected for their ability to operate in the postaccident sampling

environment. The applicant proposes that equipment used in postaccident sampling and analyses be calibrated or tested annually and retraining of operators

for postaccident sampling be performed as needed. Every 6 months, one-half of

the chemistry technicians will both operate the PASS and actually take samples

of the fluids in pertinent systems., At the same time,,identical samples will

be taken in the hot sample room. This will verify that the PASS is functioning

properly. By using this timetable, the operator will be retrained on a yearly

basis, and the PASS will be tested every 6 months.

The staff has concluded that these provisions meet Criterion (10) and are,

therefore, acceptable.

Criterion (11)

In the design of. the postaccident sampling and analysis capability, NUREG-0737

states that consideration should be given to the following items:

(1) Provisions for purging sample lines, for reducing plateout in sample lines,

for minimizing the sample loss or distortion, for preventing blockage of

sampling lines by loose material in the reactor coolant system or containment, for appropriate disposal of the samples, and for flow restrictions to

Watts Bar SSER 3 9-4
limit reactor coolant loss from a rupture of the sample line. The postaccident reactor coolant and containment atmosphere samples should be representative of the reactor coolant in the core area and the containment atmosphere following a transient or accident. The sample lines should be as

short as possible to minimize the volume of fluid to be taken from containment. The residues of sample collection should be returned to containment

or to a closed system.,

(2) The ventilation exhaust from the sampling station should be filtered with

charcoal adsorbers and high-efficiency particulate air (HEPA) filters.

The-applicant has addressed provisions for purging to ensure samples are representative, size of sample line to limit reactor coolant loss from a rupture of

the sample line, and ventilation exhaust from the PASS filtered through charcoal

adsorbers and HEPA filters. Excess samples are flushed to a radwaste system.

Heat tracing of the containment atmosphere sample line is provided to aid in

obtaining representative samples. The staff has determined that these provisions meet Criterion (11) of Item II.B.3 of NUREG-0737, and are, therefore,

acceptable.

Conclusion

The staff has concluded that the postaccident sampling system now meets all of

the 11 criteria of Item II.B.3 of NUREG-0737, and, therefore, the postaccident

sampling system is acceptable. The proposed procedure for estimating the degree

of reactor core damage is acceptable on an interim basis. Before restart following the fi.rst refueling outage, the applicant will be required to provide a final

procedure for estimating the degree of core-damage.

9.5 Other Auxiliary Systems

9.5.7 Emergency Diesel Engine Lubricating Oil System

The SER stated:

The auxiliary keep-warm Iubrication system for the diesel engine is

composed of a continuously operating oil circulating ac motor-driven

pump which prelubricates the turbocharger bearings only and circulates

oil to the lube oil cooler for preheating. The other wearing parts of

the engine do not receive any lubrication until after the engine starts,

and the engine driven lube oil pumps reach full speed. This is not

acceptable....

In letters dated October 9, 1981, and-December 14, 1982, the applicant proposed

using the manufacturer's modification to alleviate the staff's concern regarding

dry diesel engine starting. The staff reviewed the manufacturer's modification,

EMD MI-9644, and determined that the modification would provide continuous

prelubrication to the lower portions of the engine, but not the upper portions,

that is, rocker arms, camshaft, and associated wearing parts, before an emergency diesel engine start. The staff informed the applicant that adequate

justification for the nonautomatic prelubrication of the upper rocker arm assembly wearing parts be provided or prelubrication be provided periodically.

Watts Bar SSER 3 9-5
By letter dated January 12, 1984, the applicant submitted the manufacturer's

justification for not prelubricating the upper rocker arm assembly of the

emergency diesel engines at the Watts Bar facility. The staff in its review

of this information found:

(1) The bearings at the camshaft and rocker arms are lightly loaded during

startup, are not subjected to marginal conditions that require prelubrication, and will not experience damage or distress even if they are not

lubricated during startup.

(2) The unfilled portion (volume) of the camshaft and upper rocker arm assembly

lube oil system, after implementation of EMD MI-9644, is small (1/2 to 1 gal

depending on engine size) compared with the total volume of the engine lube

system. Because of the small volume involved, this portion of the system is

rapidly filled during the first few revolutions of the engine by the enginedriven lube oil pump. The pump delivers 1/10 to 1/4 gal per engine revolution, depending on the engine size.

(3) Lube oil system operating pressure is rapidly established following startup

because most of the oil system is fully charged.

On the basis of this information, the staff finds the implementation of EMD

MI-9644 at the Watts Bar facility will alleviate the concern regarding.dry

engine starting and is an acceptable alternative to continuous or intermittent

prelubrication of the upper bearing surfaces on the diesel generators. The

modification will be installed before fuel loading.

On the basis of its review, the staff concludes that the emergency diesel

engine lubricating oil system meets the requirements of GDC 2, 4, 5, and 17;

meets the guidance of the cited regulatory guides and SRP Section 9.5.7; meets

the recommendations of NUREG/CR-0660 and industry codes and standards; can

perform its safety function; and is, therefore, acceptable.

Watts Bar SSER 3 9-6
13 CONDUCT OF OPERATIONS

13.5 Plant Procedures

13.5.3 NUREG-0737 Items

Report on Outages of Emergency Core Cooling System (II.K.3.17)

In the SER, the staff stated that the applicant has committed to develop and

implement a plan for collecting emergency core cooling system (ECCS) outage information. Since publication of the SER, the staff has received sufficient

information from operating facilities to perform its review of the TMI Action

Plan item. In a letter dated October 28, 1983, the applicant revised his

commitment to comply with this item. The staff concurs with the applicant's revised position that active participation in the nuclear power reliability data system and compliance with the requirements of 10 CFR 50.73 satisfy the

current requirements of Item II.K.3.17 of NUREG-0737. Therefore, the staff

will not impose a license condition on the Watts Bar facility in this matter

and considers License Condition (30) closed.

Watts Bar SSER 3 13-1

14 INITIAL TEST PROGRAM

In the SER, the staff identified open and confirmatory items resulting from its

review of FSAR Chapter 14 (as amended through Amendment 46). The applicant

has since provided additional information and FSAR amendments addressing the

staff's concerns. The following is a discussion of these items.

(1) Test Procedures

The staff reviews selected test procedures as part of the preoperational

phase inspection plan. The SER stated that the applicant plans to provide

the staff with copies of test procedures I month before testing. RG 1.68,

"Initial Test Programs for Water-Cooled Nuclear Power Plants," Appendix B,

states that preoperational test procedures should be available 60 days

before testing and startup test procedures should be available 60 days

before fuel loading. In a letter dated December 22, 1983, the applicant

indicated that he had discussed this issue with the NRC staff who had indicated that I month was sufficient. This issue is considered resolved.

(2) Conformance to RG 1.20, Revision 2; RG 1.52, Revision 2; and RG 1.79,

Revision 1

The SER stated that FSAR Table 14.2-3 does not reflect conformance of

preoperational tests with these regulatory guides. In FSAR Amendment 49

the table was appropriately revised. This item, therefore, is resolved.

(3) Preoperational Tests

The SER stated that the applicant's FSAR does not include preoperational

tests of the following systems:

(a) condensate system

(b) condenser circulating water system

(c) hotwell level control system

(d) condensate storage tank auxiliaries

(e) 48-V dc system

(f) failed fuel detection system

(g) chemical addition system for the secondary plant

(h) turbine gland sealing system and gland seal water system

(i) standby lighting system

Watts Bar SSER 3 14-1
(j) ventilation systems for the intake pumping station

(k) emergency core cooling (ECC) systems (leakage tests) and leak

detection and pumping systems provided to control leakage from ECC

systems

(1) turbine building area ventilation system

(m) raw cooling water system

In a letter dated April 21, 1982, the applicant indicated that the above systems

would be tested as part of the Non-Critical Systems (NCS) Test Program. NCS

tests have a lowerlevel of quality assurance (QA) overview than the preoperational tests covered by FSAR Chapter 14. NCS test procedures are not subject

to review by the Plant Operations Review Committee, and the NCS test program

is not subject to QA audit or review by the Nuclear Safety Review Board. This

graded approach to testing is consistent with GDC 1. The staff has reviewed

the NCS Test Program described in Procedure No. 1105.01 of the applicant's

Test Staff Program Manual and has concluded that NCS testing is acceptable

for Items (a), (b), (c), (g), (h), (1), and (m) above.: Test descriptions for

NCS are not provided in the FSAR; however, these test procedures will be subject

to review by the staff in accordance with the preoperational phase inspection

plan. The staff requires that the status of NCS tests be tracked and considered

in determining readiness for licensing in the same way as the preoperational

tests described in the FSAR.

The SER stated that the systems in Items (d), (e), (f), (i), (j), and (k) should

be preoperationally tested and that the staff would review test abstracts,'when

they are received, and report theresults of its review. In letters dated

April 20, 1983, March 28, 1983, and May 2, 1984, and in FSAR Amendments 48 and

49, the applicant provided additional information or test descriptions for these

items. A discussion of each test follows.

Item (d): It is the applicant's position that the condensate storage tank

auxiliaries should not have to be preoperationally tested because the condensate

storage tank is not an engineered safety feature (ESF) and is backed up by the -

essential raw cooling water system, which is an ESF and is preoperationally

tested. The staff accepts this position and considers testing as part of the

NCS Test Program adequate. Therefore, this item is resolved.

Item (e): The applicant stated that the 48-V dc system should not be preoperationally tested because it serves only as a backup power supply to a nonessential communications system. The 48-V dc system is a backup power supply to the

Southern Bell DIMENSION system. The DIMENSION system is a primary system for

offsite and interplant communications. However, it is not required for emergency shutdown, emergency response, or emergency notifications. Other communications systems that are tested will be available for emergency use. The staff

therefore concludes that the 48-V dc system need not be preoperationally tested

in accordance with FSAR Chapter 14 or as part of the NCS Test Program.

Items (f), (i), and.(j): These items are now described in the FSAR preoperational test abstracts. The staff has reviewed the abstracts and found them

acceptable.. These items are considered resolved.

Watts Bar SSER 3 14-2
Item (k): The applicant has provided a test description for the station drainage water system, but it was not clear that this preoperational test included

testing of the leakage detection systems of concern. However, a commitment was

made in FSAR Amendment 35, Chapter 6, page 6.3-23a, that the individual detectors

located in each emergency core cooling system (ECCS) pump compartment, in the

ECCS heat exchanger rooms, in the pipe gallery for each unit, and in the pipe

chase are to be preoperationally tested to verify initial operability. These

leakage detection systems monitor the residual heat removal (RHR) pump compartments as part of the ECCS. The RHR pumps also provide flow to the containment

spray system. The staff concludes that acceptable testing will be performed by

fulfillment of this commitment, and, therefore, Item (k) is resolved.

Although all of the staff concerns discussed in the SER have been resolved, new

concerns arose because of applicant changes to the previously reviewed test program (post-SER FSAR amendments). The natural circulation test, power coefficient

test, and control rod drop and plant trip tests had been deleted without justification. These concerns were addressed by the applicant in the letter dated

May 2, 1984, as discussed below:

Natural Circulation Test

The natural circulation test, which satisfies TMI Task Action Plan Item I.G.1,

had been deleted. By letter dated May 2, 1984, the applicant stated that this

test is to be reinstated. Therefore, this concern isresolved.

Power Coefficient Test

The applicant proposes performing the power coefficient-measurement only at full

power. The justifications offered for the modified power coefficient measurement are:

(1) The power coefficient is not measured directly.

(2) The measurement is time consuming compared with the value of the data.

(3) The measurement also was previously deleted in another plant (McGuire

Unit 2) because the measurement had been made in an identical sister plant.

Watts Bar Nuclear Plant Units 1 and 2 are nearly identical to the Sequoyah

Nuclear Plant Units 1 and 2 and the McGuire Operating Station Units 1 and 2.

The power coefficient or Doppler coefficient (the major component of the power

coefficient) was measured and compared with the design values in three of these

plants. The comparison between the measurements and design values was within

the acceptance criteria in all cases, demonstrating the ability to analytically

predict this design parameter. In addition, this modification to the power

coefficient measurement is consistent with the deletion of themeasurement in

the McGuire Unit 2 test program. Therefore, the staff concludes that the proposed reduced power coefficient measurements are acceptable.

Control Rod Drop and Plant Performance Test

The applicant requests deleting this test entirely from the test program on the

basis that the trip function and other test objectives are accomplished in the

Watts Bar SSER 3 14-3
Technical Specification surveillance testing and other startup tests. The original primary intent of this test was to demonstrate that control rods of low

reactivity worth and/or distant from external neutron detectors were capable of

activating the negative rate two-out-of-four power level detector trips. The

safety concern was that, at full power, if a control rod(s) was dropped while

in automatic control and a reactor scram did not occur, the automatic control

would try to reestablish the predrop power level and in the attempt would overshoot the steady-state power level. The dynamics of this power overshoot might

produce the conditions leading to violation of the departure from nucleate boiling (DNB) limit.

The nuclear steam supply system (NSSS) vendor (Westinghouse) has presented

analytical studies that define detection limits for control rod reactivity

worths and/or locations and have shown that' the DNB limit is not violated for

rods below the detection limit. In addition, the control rod drop tests performed at several similar plants have demonstrated that rods well below the

analytically defined detection limits are capable of scramming the plant.

The staff has reviewed the information presented by the applicant. On the basis

of the NSSS vendor's analyses, generic testing at similar plants, and the accomplishment of other test objectives elsewhere in the test program, the staff concludes that it is acceptable'to delete the; control rod drop and plant performance

test from the Watts Bar Nuclear Plant, Units 1 and 2, test program.

Sequence and Administrative Changes

In addition to the above items, the applicant proposed making the following

administrative and test sequence changes:

(1) Modify test sequence documents by removing Tests SU-3.8C, SU-4.1OA, and

SU-4.11, and make these tests individual test documents without changing

the objectives or method of performance.

(2) Incorporate Test SU-2.5, "Inverse Count Rate Monitoring for Core Loading,"

into Test SU-2.1, "Initial.Core Loading," and Test SU-3.1, "Inverse Count

Rate Monitoring for Approach to Critical," into Test SU-3.2, "Initial

Criticality for Ease of Test Performance."

(3) Incorporate Tests SU-3.4, SU-3.7, and SU-4.2 into plant technical instructions (TIs) that have been reviewed by the Plant Operations Review

Committee. Test SU-3.4, "Reactivity Computer," will be incorporated into

TI-7, "Reactivity Computer Checkout," and Tests SU-3.7, "Incore Moveable

Detectors," and SU-4.2, "Thermocouple Mapping," will be incorporated into

TI-41, "Incore Flux Mapping."

The staff finds the sequence and administrative changes acceptable since the

originally reviewed test objectives and methods are not changed.

Additional Concerns

The applicant has made other modifications to the Watts Bar initial test program

in additional amendments to the FSAR (through Amendment 53). By letter dated

December 19, 1984, the staff requested additional information regarding these

changes. These items will be pursued with the applicant, and the results of

the staff review will be reported in a future supplement.

Watts Bar SSER 3 14-4
Conclusions

The staff concludes that the Watts Bar initial test program described in the

application', with the exception of the above open items, will meet the acceptance criteria of SRP Section 14.2 (NUREG-0800) and that the successful completion of the program will demonstrate the functional adequacy of plant structures,

systems, and components. The results of the staff's review of the remaining

open items will be reported in a future supplement.

Watts Bar SSER 3 14-5

15 ACCIDENT ANALYSIS

15.3 Limiting Accidents

15.3.2 Steamline Break

In a letter dated May 2, 1984, the applicant proposed removal of the boron

concentration requirement of 20,000 ppm in the boron injection tank (BIT) and proposed changes to the Technical Specifications that would delete those

Technical Specifications that address the boron concentration and heat tracing for the BIT.

The BIT was incorporated in the plant design to mitigate the consequences of postulated steamline break (SLB) events. During these events, the high head safety injection pumps automatically align to discharge through the BIT, which

contains a highly concentrated boric acid solution (20,000 ppm). This solution is then flushed into the primary system to ensure adequate shutdown reactivity. The high head safety injection pumps take suction from the refueling storage tank, which contains borated water at a concentration of 2,000 ppm.

In the revised analyses to support the proposed design change, the BIT is assumed to contain a boron concentration of zero ppm. The applicant rean- alyzed the SLB accident using the same computer codes and assumptions as for the previous analysis. The results indicate that the reactor returns to power with a maximum heat'flux of approximately 18% of the design and a corresponding reactor coolant system pressure of approximately 800 psia. The minimum departure from nucleate boiling ratio remains above the 1.30 limit.

Although limited cladding perforation following an SLB event is permitted by the Standard Review Plan (NUREG-0800), the applicant has demonstrated that no cladding perforation is predicted to occur. On the basis of its review of the applicant's evaluation, the staff concludes that there is no significant change in the safety margin.

With regard to the proposed changes to the Technical Specifications and heat

tracing for the BIT, the applicant stated that the current requirement was due

to high boron concentration in the BIT and associated piping. Reduction of

the boron concentration requirement to zero ppm would eliminate all Technical

Specifications concerning BIT boron concentration, temperatures, and associated surveillance, including heat tracing, since heat tracing would only be required for boron concentrations above 4 weight percent corresponding to approximately

7,000 ppm.

On the basis of its review of the applicant's evaluation, the staff concludes

that there is no significant reduction in the safety margin. The supporting analysis also demonstrates compliance with SRP Section 15.1.5 (NUREG-0800). Therefore, the staff finds the applicant's proposal acceptable.

Watts Bar SSER 3 15-1
15.3.6 Anticipated Transients Without Scram

Generic Letter 83-28 was issued by the NRC on July 8, 1983, indicating actions

to be taken by applicants based on the generic implication of the Salem anticipated transients without scram (ATWS) events. Item 4.3 of the generic letter

requires that modifications be made to improve the reliability of the reactor trip system by implementation of an automatic actuation of the shunt attachment on the reactor trip breakers. By letter dated June 4, 1984, the applicant

provided responses to the plant-specific questions identified by the staff in

its SER on the generic Westinghouse design submitted by letter dated August 10,

1983. The staff has reviewed the applicant's proposed design for the automatic

actuation of reactor trip breaker shunt trip attachments and finds it acceptable.

Evaluation

The staff requested the applicant to submit information on the following required

plant-specific items on the basis of its review of the Westinghouse Owners Group

(WOG)-proposed generic design for this modification. Below is a discussion of

the applicant's responses.

(1) Provide the electrical schematic/elementary diagrams for the reactor trip

and bypass breakers showing the undervoltage and shunt coil actuation

circuits as well as the breaker control (e.g., closing) circuits, and circuits providing breaker status information/alarms to the control room.

The design of the electrical circuits for the shunt trip modification has been

reviewed and found consistent with the WOG generic proposed design, which was

previously reviewed and approved by the staff. However, the applicant's design

includes test jacks to facilitate response time testing during plant operation.

This addition to the WOG generic design consists of test jacks wired directly

to an auxiliary switch "a" contact and test jacks wired in series with 1-kilohm

(2.5-W) resistors across the undervoltage coil. Thus, test connections for an

undervoltage trip signal and breaker-tripped condition are available to perform

the response time test. The resistors in series with the test connections to

the undervoltage coil provide protection against potential accidental shorts or

grounds during response time testing to ensure that such events would not result

in an inadvertent breaker trip or overload on the protection system power source

for the undervoltage trip attachment. On the basis of its review of these plantspecific aspects of the applicant's design, the staff concludes that they do

not introduce a-safety-significant consideration, will facilitate on-line response time testing, and are, therefore, acceptable.

(2) Identify the power sources for the shunt trip coils. Verify that they are

Class 1E and that all components providing power to the shunt trip cOrcuitry are Class 1E and that any faults within non-Class 1E circuitry will

not degrade the shunt trip function. Describe the annunciation/indication

provided in the control room upon loss of power to the shunt trip circuits.

Also describe the overvoltage protection and/or alarms provided to prevent

or alert the operator(s) to an overvoltage condition that could affect

both the undervoltage (UV) coil and the parallel shunt trip actuation

relay.

Watts'Bar SSER 3 15-2
Redundant Class 1E power sources are used for the shunt trip actuation of the

reactor trip breakers and for the shunt trip of the bypass breakers. Class 1E

circuitry is separated from non-Class 1E circuitry in accordance with RG 1.75

and is, therefore, acceptable.

The breaker position status lights are used to supervise the availability of

power to the shunt trip circuits. The red light, which is connected in series

with the shunt coil and the "a" auxiliary contact, indicates that the breaker

is closed and that the power is available to the shunt trip device and, therefore, provides detectability of power failure to the shunt trip coil. Also,

normally open contacts of an auxiliary relay that is energized when the breaker

is closed provide breaker-status information to the plant computer. These contacts would change state if power for the shunt trip was lost. Main control

room annunciation is provided upon trip of any 125-V dc breaker in the battery

board. Also, undervoltage and overvoltage annunciation is provided for the

shunt trip circuit's 125-V dc power source. Annunciation is provided for the

loss of any one of the two 48-V dc power supplies in each train of the solidstate protection system (SSPS).

Normally, the shunt trip coils in the reactor trip breakers are in the deenergized condition When the trip breakers are closed. The red lamp current

(approximately 50 ma) flows through the trip coil to monitor the circuit continuity, which is not large enough to actuate the trip coil armature. Since

the current through the shunt trip-coils is interrupted when the breaker

trips, energization of the shunt trip coil is only momentary. The maximum

available voltage occurs during a battery equalizing charge at a maximum

voltage of 115% of the nominal voltage. Because of the short-duty cycle of

the shunt trip coil, it can operate at this overvoltage condition without

harmful effects.

The added shunt trip circuitry is powered from the reactor protection logic

power supply. The reactor protection logic power supply consists of two

redundant 48-V dc power supplies for each train of the SSPS. Components in

the added shunt trip circuitry have been selected on the basis of their ability

to perform their intended function up to 115% of nominal voltage. The overvoltage protection in each redundant reactor protection logic power supply is

set at 115% of nominal voltage.

On the basis of its review, the staff concludes that appropriate consideration

has been given to the aspects of the design described above, and the design

is, therefore, acceptable.

(3) Verify that the relays added for the automatic shunt trip function are

within the capacity of their associated power supplies and that the relay

contacts are adequately sized to accomplish the shunt trip function. If

the added relays are other than Potter & Brumfield MDR series relays

(P/N 2383A38 or P/N 955655) recommended by Westinghouse, provide a description of the relays and their design specifications.

The design at Watts Bar includes the Potter & Brumfield MDR series P/N 955655

relays as specified in the WOG generic design for the automatic shunt trip

function. The relay contacts are adequately sized to accomplish the shunt

trip function. The staff finds this aspect of the design acceptable.

Watts Bar SSER 3 15-3
(4) State whether the test procedure/sequence used to independently verify

operability of the undervoltage and shunt trip devices in response to an

automatic reactor trip signal is identical to the test procedure proposed

by the Westinghouse Owners Group (WOG). Identify any differences between

the WOG test procedure and the test procedure to be used and provide the

rationale/justification for these differences.

The applicant notes that the steps used to independently confirm the operability

of the undervoltage trip and shunt trip devices in response to an automatic

reactor trip signal are identical to the test procedure proposed by the WOG to

NRC by letter No. OG-101 dated June 14, 1983. This procedure will be incorporated into Surveillance Instruction (SI) 3.1.26 before Unit 1 fuel load. The

staff finds this acceptable.

(5) Verify that the circuitry used to implement the automatic shunt trip function is Class 1E (safety related), and that the procurement, installation,

operation, testing, and maintenance of this circuitry will be in accordance

with the quality assurance criteria set. forth in Appendix B to 10 CFR 50.

The applicant confirmed that the circuitry used to implement the automatic shunt

trip function is Class 1E (safety related) and the procurement, installation,

operation, testing, and maintenance of this circuitry will be in accordance with

the TVA nuclear power station quality assurance program and Watts Bar-specific

quality assurance procedures which satisfy Appendix B to 10 CFR 50. The staff

finds this acceptable.

(6) Verify that the shunt trip attachments and associated circuitry are/will

be seismically qualified (i.e., be demonstrated to be operable during and

after a seismic event) in accordance with the provisions of Regulatory

Guide 1.100, Revision 1, which endorses IEEE Standard 344, and that all

nonsafety-related circuitry/components in physical proximity to or associated with the automatic shunt trip function will not degrade this

function during or after a seismic event.

The applicant notes that the shunt trip attachments and associated circuitry

will be seismically qualified in accordance with IEEE Standard 344. The WOG

is working with Westinghouse to obtain seismic qualification of the shunt

trip attachments. Nonsafety-related circuits are isolated from safety-related

circuits by qualified isolators. The shunt trip circuitry is located within

seismically qualified Class 1E reactor trip switchgear. The staff finds this

acceptable pending verification of the seismic qualification of the shunt trip

attachments.

(7) Verify that the components used to accomplish the automatic shunt trip

function are designed for the environment where they are located.

The applicant has verified that the plant-specific environmental conditions

defined in Table 1 of the WOG generic design package envelope Watts Bar Units 1

and 2. The staff finds this acceptable.

(8) Describe the physical separation provided between the circuits used to

manually initiate the shunt trip attachments of the redundant reactor

trip breakers. If physical separation is not maintained between these

Watts Bar SSER 3 15-4
circuits, demonstrate that faults within these circuits cannot degrade

both redundant trains.

Physical separation between the circuits used to manually initiate the shunt

trip attachments of the redundant trip breakers is maintained by routing the

field cabling from the main control board and reactor protection logic to redundant train A and train B reactor trip switchgear as train A and train B

circuits. Coil-to-contact isolation is provided within reactor trip switchgear, and metal-braid-enclosed cabling is used for train A and train B wiring

where a 6-in. air gap is not maintained. Dual section manual reactor trip

switches with metal barriers are used between redundant train switch decks.

The staff finds this meets the requirements of RG 1.75 and is, therefore,

acceptable.

(9) Verify that the operability of the control room manual reactor trip switch

contacts and wiring will be adequately tested prior to startup after each

refueling outage. Verify that the test procedure used will not involve

installing jumpers, lifting leads, or pulling fuses and identify any deviations from the WOG procedure. Permanently installed test connections (i.e.,

to allow connection of a voltmeter) are acceptable.

The applicant notes that all control room manual reactor trip switch contacts

and wiring will be tested before startup after each refueling outage'by the

performance of SI-3.1.1. The test procedure used does not involve insta. 1 ing

jumpers, lifting leads, or pulling fuses. The UV coil voltage will be monitored by the voltmeter permanently installed on the SSPS test panel. The shunt

coil voltage will be monitored by temporarily connecting a voltmeter across the

combination of shunt trip coil and the series connected "a" auxiliary contact.

This procedure will be incorporated into SI-3.1.1 before Unit 1 fuel loading.

The staff finds this acceptable.

(10) Verify that each bypass breaker will be tested to demonstrate its operability prior to placing it into service for reactor trip breaker testing.

The applicant notes that before the reactor trip breaker test during power

operation is performed, the operability of the bypass breaker required for

testing will be verified by racking the bypass breaker into test position and

tripping the breaker open. This procedure will be incorporated into SI-3.1.26

before Unit 1 fuel loading. The staff finds this acceptable. However, the

staff will require that the bypass breaker undervoltage trip attachment be

demonstrated operable at a refueling outage frequency.

(11) Verify that test procedures used to determine reactor trip breaker operability will also demonstrate proper operation of the associated control

room indication/annunciation.

The applicant notes that the reactor trip breaker tests will verify the proper

operation of the main control room reactor trip breaker indication. The staff

finds this acceptable.

(12) Verify that the response time of the automatic shunt trip feature will be

tested periodically and shown to be less than or equal to that assumed in

the FSAR analyses or that specified in the Technical Specifications.

Watts Bar SSER 3 15-5
The applicant notes that the response time of the automatic shunt trip will be

tested and used in-demonstrating that the reactor trip system instrumentation

response times specified in the Technical Specifications are within their limits.

The staff finds this acceptable.

(13) Propose Technical Specification changes to require periodic testing of the

undervoltage and shunt trip functions and the manual reactor trip switch

contacts and wiring.

The applicant notes that Technical Specifications require a manual reactor trip

actuating device operational test be performed at least once every 18 months

and a reactor trip breaker trip actuating device operational test be performed

on each train at least every 62 days on a staggered test basis. The applicant

further notes that the Technical Specifications require that at least once every

18 months and following maintenance or adjustment of the reactor trip breakers,

the trip actuating device operational test shall include independent verification of the UV and shunt trips. The staff will review the plant Technical

Specifications to confirm that they include surveillance requirements consistent

with the applicant's commitments. This review will be completed before the

operating license for Unit 1 is issued.

Conclusion

On the basis of the review of the applicant's response to the plant-specific

questions identified in the staff's evaluation of the WOG generic design modifications, the staff finds that these modifications are acceptable.

However, the staff will require that the applicant submit confirmation that the

seismic qualification of the shunt trip attachment has been successfully completed

as noted in Item (6). Further, the staff requires that proposed Technical Specifications be submitted, which are responsive to the staff requirements noted in

Items (10) and (13), following implementation of this modification.

15.4 Radiological Consequences of Accidents

15.4.3 Steam Generator Tube Rupture

In the Watts Bar FSAR, the applicant made general, unverified assumptions concerning system performance following a complete severance of a single steam

generator tube. In addition, the FSAR assumed that the break flow was terminated within 30 min of the event by operator actions to equalize the primary

and secondary pressures. In the SER, the staff addressed the accident, including the sequence of events and the radiological consequences, land found them

acceptable. However., the actual steam generator tube rupture (SGTR) event that

occurred at Ginna indicated that more than 30 min could be required for pressure

equalization, implying that the Watts Bar analysis was nonconservative with

respect to assumed operator actions.

As a result, by letter dated May 3, 1984, the staff requested additional information, including an evaluation of operator action times, as to whether liquid

can enter the steamlines and the effects on the integrity of steam piping and

supports. The staff also requested that a reactor systems analysis be performed

for natural circulation cooldown with an SGTR, including.the effect of the worst

Watts Bar SSER 3 15-6
single failure of a system that is either required or expected to operate during

the event.

By letter dated June 11, 1984, TVA stated that it has joined the Westinghouse

Owners Group, which is investigating the SGTR issue on a generic basis. The

owners group plans to issue a report late in 1984. The applicant has committed

to implement all corrective actions recommended by the owners group as approved

by the NRC staff before startup following the first refueling outage for Unit 1.

As to whether there is adequate assurance that the Watts Bar plant can operate

safely for one cycle of operation until the SGTR issue is satisfactorily resolved, the staff notes the following:

(1) All components necessary for mitigation of the design-basis SGTR are safety

related.

(2) The plant emergency procedures for an SGTR have been reviewed and approved

by the NRC staff.

(3) There is a low probability of a design-basis SGTR in the first cycle of

operation.

Therefore, the staff concludes that there is sufficient assurance that the

Watts Bar plant can operate safely for one cycle until this issue is resolved.

The staff will condition the license to require satisfactory resolution of this

issue before the startup following the first refueling outage of Unit 1.

Watts Bar SSER 3 15-7

December 22, 1980

February 18, 1981

January 27, 1982

February 25, 1982

August 12, 1982

June 3, 1983

June 10, 1983

June 29, 1983

August 10, 1983

APPENDIX A

CHRONOLOGY OF RADIOLOGICAL REVIEW OF

WATTS BAR NUCLEAR PLANT, UNITS 1 AND 2,

OPERATING LICENSE REVIEW

Letter .to applicant regarding control of heavy loads

(NUREG-0612).

.Letter to applicant concerning post-TMI requirements for

the emergency operations facility (Generic Letter 81-10).

Letter from applicant transmitting Revision 0 to Radiological Emergency Plan.

Letter from applicant regarding loose parts monitoring

program.

Letter from applicant regarding NUREG-0737, Items II.B.1,

II.B.2, II.F.1, and II.F.2.

Letter from applicant concerning responses to Seismic

Qualification Review Team (SQRT) reviewer concerns.

Letter from applicant concerning responses to SQRT

reviewer concerns.

Letter from applicant regarding reactor vessel 1 ,el

indication system installation and operation.

Letter to Westinghouse transmitting SER on generic

Westinghouse design of the shunt attachment on reactor

trip breakers.

Letter from applicant providing additional information

concerning NUREG-0737, Item II.F.2.

Letter from applicant concerning automatic trip of

reactor coolant pumps.

Letter to applicant concerning NRC use of the terms,

"important to safety" and "safety related" (Generic

Letter 84-01).

Letter from applicant forwarding safety parameter

display system safety analysis.

Letter to applicant concerning notice of meeting

regarding facility staffing (Generic Letter 84-02).

A-1

January

January

January

January

January

4,

5,

5,

5,

6,

1984

1984

1984

1984

1984

Watts Bar SSER 3
January

January

January

January

January

January

January

January

January

January

January

January.

January

6,

9,

9,

10,

11,

12,

12,

12,

12,

12,

12,

13,

13,

1984

1984

1984

1984

1984

1984

1984

1984

1984

1984

1984

1984

1984

Letter from applicant forwarding Amendment 49 to FSAR.

Letter from applicant concerning compliance with General

Design Criterion (GDC) 51.

Letter from applicant forwarding drawings showing vital

area designations.

Letter from applicant updating status report for items

listed in Appendix D of SER.

Letter from applicant concerning equipment qualification

program.

Letter to applicant concerning facility staffing survey.

Letter to applicant concerning diesel generator auxiliary

systems design deficiencies.

Letter from applicant concerning diesel engines.

Letter from applicant concerning milestone schedules for

the prompt notification system.

Letter from applicant concerning the fire protection

program.

Letter from applicant concerning the issue of liquefac- tion potential.

Letter to applicant concerning Physical Security Plan -

designation of vital equipment.

Letter to applicant concerning availability of NUREG-0933,

"A Prioritization of Generic Safety Issues" (Generic Letter 84-03).

Letter from applicant concerning NRC Power Systems Branch status list of open and confirmatory items and

license conditions.

Letter from applicant concerning equipment qualification

program.

Letter from applicant concerning supervision of fire

alarm and detection circuits.

Letter from applicant informing that fuel loading is now

anticipated to be June 1984 for Unit 1 and December 1985

for Unit 2.

Letter from applicant concerning installation of highrange noble gas monitors on the auxiliary building vent.

January 17, 1984

January

January

January

17,

19,

23,

1984

1984

1984

January 24, 1984

Watts Bar SSER 3 A-2
January

January

January

January

January

January

26, 1984

30, 1984

30,

30,

30,

31,

1984

1984

1984

1984

February 1, 1984

February

February

February

February

February

February

February

February

February

February

3, 1984

6, 1984

6, 1984

7, 1984

7, 1984

9, 1984

9, 1984

10, 1984

14-16, 1984

15, 1984

Letter from applicant forwarding Amendment 49 to FSAR.

Letter from applicant concerning Underwriters Laboratory

test report for the 1-hour-fire-rated barriers being used to separate redundant safe shutdown circuits.

Letter from applicant responding to NRC concerns regarding the Physical Security Plan.

Letter to applicant issuing Supplement 2 to Safety Evaluation Report.

Letter from applicant concerning compliance with Regulatory Guide 1.97.

Letter from applicant concerning changes to Technical

Specification surveillance requirements for the diesel

fuel impurity level tests.

Letter to applicant concerning safety evaluation of

Westinghouse topical reports dealing with elimination

of postulated pipe breaks in PWR primary main loops

(Generic Letter 84-04).

Letter to applicant concerning submittal of offsite dose

calculations manual.

Letter to applicant concerning Physical Security Plan - designation of vital equipment.

Letter from applicant concerning control of heavy loads

at nuclear power plants, Generic Letter 81-07.

Letter from applicant concerning limiting tank concentration value.

Letter from applicant submitting "TVA Watts Bar Nuclear.

Plant Auxiliary Feedwater System Independent Review."

Letter from applicant concerning facility staffing survey.

Letter from applicant concerning updated status on fire

protection and miscellaneous commitments.

Meeting with applicant to discuss welding codes used.

(Summary issued March 1, 1984)

Meeting with applicant at site to audit equipment

qualification program. (Summary issued March 14, 1984)

Letter to applicant concerning review status of the

application.

Watts Bar SSER 3 A-3
February 15, 1984

February

February

February

February

March 6,

March 6,.

17,

21,

27,

28,

1984

1984

1984

1984

1984

1984

March

March

March

March

March

March

March

March

March

March

March

7, 1984

8, 1984

9, 1984

13, 1984

13, 1984

14, 1984

14, 1984

14, 1984

15-16, 1984

20, 1984

21, 1984

Letter from applicant providing additional information

concerning the Radiological Emergency Plan emergency

action levels.

Letter to applicant concerning compliance with

Item II.K.3.17 of NUREG-0737.

Letter from applicant concerning status of outstanding

and confirmatory items.

Letter from applicant concerning calculation of vertical

temperature gradients.

Letter from applicant forwarding Revision 4 to Physical

Security Plan.

Letter to applicant concerning diesel generator auxiliary

systems design deficiencies.

Letter to applicant requesting additional information

concerning use of containment high-range monitor readings

as an indicator of core damage.

Letter from applicant concerning various NRC positions

described in the SER.

Letter to applicant requesting additional information

concerning use of containment high-range monitor readings.

Letter to applicant requesting additional information

regarding the initial test program.

Letter to applicant concerning changes to the FSAR.

Letter to applicant concerning deletion of home telephone

numbers; unlisted utility numbers, etc. from emergency

plans.

Letter to applicant concerning equipment qualification

audit.

Letter from applicant concerning 175-ton polar crane.

Letter to applicant concerning compliance with GDC 51.

Meeting with applicant to conduct the confirmatory site

visit.

Letter from applicant concerning control of heavy loads.

Letter from applicant concerning qualification of the

diesel generator auxiliary system piping.

Watts Bar SSER 3 A-4
March

March

March

March

March

March

21,

26,

27,

28,

28,

29,

1984

1984

1984

1984

1984

1984

March 29, 1984

Letter from applicant forwarding revision to the Radiological Emergency Plan.

Letter to applicant issuing order extending construction

completion dates for Unit 1 to January 1, 1985, and

Unit 2 to July 1, 1986.

Letter from applicant concerning proposed modifications

to the draft Technical Specifications.

Letter from applicant providing additional information

relating to the installation and preoperational testing

schedules for the reactor vessel level instrumentation

system.

Letter from applicant concerning modifications to the

radiological environmental monitoring program.

Meeting with applicant to discuss proposed changes to

TVA's Nuclear Safety Review Board. (Summary issued

April 11, 1984)

Letter from applicant concerning TVA's evaluation of the

Black and Veatch independent review of the Watts Bar

plant.

Letter from applicant concerning radiographic inspection

of the wedge located in each of the main feedwater isolation valves.

Letter to applicant forwarding change to NUREG-1021,

"Operator Licensing Examiner Standard" (Generic Letter

84-05).

Letter to applicant concerning interim procedures for

NRC management of plant-specific backfitting (Generic

Letter 84-08).

Letter from applicant concerning remedial action to

resolve the issue of potential liquefaction.

Letter from applicant concerning environmental qualification of the safety-related mechanical equipment.

Letter from applicant responding to NUREG-0737,

Item II.K.3.10.

Meeting with applicant to discuss the Black and Veatch

independent design review. (Summary issued May 22, 1984)

Letter from applicant forwarding Revision 6 to Physical

Security Plan.

March 30, 1984

April 2, 1984

April 4, 1984

April

April

April

April

April

6, 1984

6, 1984

6, 1984

9, 1984

10, 1984

Watts Bar SSER 3 A-5
April 13, 1984

April 13, 1984

April 13, 1984

April 18, 1984

April 19, 1984

April 23, 1984

April 24, 1984

April 25, 1984

April 25, 1984

April 26, 1984

April 30, 1984

May 1, 1984

May 2, 1984

May 2, 1984

May 3, 1984

Letter to applicant concerning buckling criteria for

Class 2 and 3 supports.

Letter from applicant forwarding revision of Technical

Instruction to Preservice Inspection Program.

Letter from applicant forwarding revision to Implementing

Procedures Document.

Letter from applicant concerning the prompt notification

system.

Letter from applicant responding to Office of Inspection

and Enforcement (IE) Bulletin 80-06 regarding.controls

for feedwater isolation and containment purge isolation

valves.

Letter to applicant requesting additional information

regarding deliberate ignition hydrogen control.

Letter from applicant concerning the offsite power automatic transfer scheme, undervoltage alarms, and potential submergence of electrical equipment.

Letter to applicant concerning supplement regarding the

seismic and dynamic qualification of safety-related

electrical and mechanical equipment.

Letter to applicant concerning use of ASME Code Case

N-32-4 for hydrostatic testing of-Section III, Class 3,

embedded piping.

Letter to applicant concerning administration of operating tests before initial criticality (10 CFR 55.25)

(Generic Letter 84-10).

Letter to applicant concerning compliance with 10 CFR 61

and implementation of the Radiological Effluent Technical Specifications (RETS) and attendant process control program (PCP) (Generic Letter 84-12).

Letter from applicant forwarding Amendment 50 to FSAR.

Letter from applicant concerning deletion of startup

tests from FSAR.

Letter from applicant concerning Westinghouse study,

"Report for the BIT Concentration Reduction/BIT Elimination Study for Watts Bar Units 1 and 2."

Letter to applicant concerning Technical Specification for

snubbers (Generic Letter 84-13).

Watts Bar SSER 3 A-6
May 3, 1984

May 3, 1984

May 4, 1984

May 7, 1984

May 8, 1984

May 8, 1984

May 8, 1984

May 8, 1984

May 9, 1984

May 9, 1984

May 10, 1984

May 10, 1984

May 14, 1984

May 14, 1984

May 15, 1984

May 1.6, 1984

May 16, 1984

Letter to applicant requesting additional information

concerning steam generator tube rupture.

Letter to applicant concerning compliance with GDC 51.'"

Letter from applicant forwarding Revision 2 to Radiological

Emergency Plan IPD and revision to Radiological Emergency

Plan.

Letter from applicant concerning status of various commitments concerning control room modifications.

Letter from applicant concerning proposed changes to draft

Technical Specifications.

Letter from applicant concerning current status list for

outstanding and confirmatory issues.

Letter from applicant providing updated status on certain items related to fire protection and updated information on miscellaneous commitments.

Letter from applicant concerning requests for exemptions

from requirements of NUREG-0737, Item II.F.1, and requirement to install the high-range noble gas monitors on the

auxiliary building vent.

Letter to applicant requesting additional information

regarding the financial qualifications review.

Letter from applicant forwarding Amendment 51 to FSAR.

Letter from applicant advising that fuel load date for

Unit 1 is now July 1984 and for Unit 2 is now June 1986.

Meeting with applicant at site to discuss resolution of

open items remaining in the Physical Security Plan.

Letter from applicant concerning buckling criteria for

Class 2 and 3 supports.

Letter from .applicant transmitting proposed modifications

to draft Technical Specifications related to testing of

circuit breakers.

Letter to applicant providing comments on the proposed

offsite dose calculation manual.

Letter from applicant concerning containment design in

comparison with current buckling studies.

Letter to applicant requesting additional information

regarding the Physical Security Plan.

Watts Bar SSER 3 A-7
May 17, 1984

May 25, 1984

May 25, 1984

May 29, 1984

May 30, 1984

May 30, 1984

June 4, 1984

June 4, 1984

Letter from applicant concerning seismic qualification of

equipment.

Letter from applicant concerning hydrogen control for

degraded core accidents.

Letter from applicant concerning seismic qualification

of equipment.

Letter to applicant requesting additional information

regarding the financial qualifications review.

Letter to applicant concerning review of utility onshift

operating experience.

Letter from applicant requesting license contain exemptions for performing natural circulation tests.

Letter from applicant concerning NUREG-0737, Item II.B.2,

"Design Review of Plant Shielding and Environmental

Qualification of Equipment for Spaces/Systems Which

May Be Used in Postaccident Operations."

Letter from applicant concerning Generic Letter 82-28,

"Required Actions Based on Generic Implications of Salem

ATWS Events."

Letter to applicant concerning TVA's reorganization.

Letter to applicant concerning vital area definition

reports.

Letter from applicant concerning overpressure protection.

Letter from applicant concerning feedwater isolation

valve wedges.

Letter from applicant concerning the analysis and mitigation of steam generator tube rupture events.

Letter from applicant providing Revision 7 to Physical

Security Plan.

Letter from applicant concerning audit of electrical

equipment qualification program.

Letter from applicant concerning deletion of the rod

drop and plant trip test.

Meeting with applicant to discuss TVA's generic control

room design review program.

Letter from applicant concerning cement mortar lining

in the essential raw cooling water piping.

June

June

June

June

5,

6,

7,

7,

1984

1984

1984

1984

June 11, 1984

June 11, 1984

June 13, 1984

June 13, 1984

June 14, 1984

June 15, 1984

Watts Bar SSER 3 A-8
June 18,

June 19,

1984

1984

June 19, 1984

June 25, 1984

June 25, 1984

June 26,

June 26,

1984

1984

June 27, 1984

July 3, 1984

July 6, 1984

July 6, 1984

July 10, 1984

July 11, 1984

July 13, 1984

July 19, 1984

July 20, 1984

July 24, 1984

Letter from applicant forwarding Amendment 52 to FSAR.

Letter from applicant providing comments/proposed modifications to'Unit 1 Technical Specifications.

Letter from applicant concerning seismic qualification

of equipment.

Letter from applicant responding to Generic Letter 83-10c

concerning automatic trip of reactor coolant pumps.

Letter from applicant concerning NRC review of confirmatory items, license conditions, and miscellaneous items.

Letter from applicant forwarding Amendment 53 to FSAR.

Letter from applicant concerning status of control room

modifications required for fuel loading.

Letter to applicant concerning adequacy of onshift operating experience for near-term operating license applicants (Generic Letter 84-16).

Letter to applicant concerning annual meeting to discuss recent developments regarding operator training,

qualifications, and examinations (Generic Letter 84-17).

Letter from applicant concerning experience levels of

operating crew personnel.

Letter from applicant concerning financial qualifications.

Letter from applicant concerning ASME Code, Section XI,

Preservice Inspection Program Technical Instruction TI-50A.

Letter from applicant providing Revision 8 to Physical

Security Plan.

Letter from applicant concerning postaccident sampling

system.

Letter from applicant concerning relationship between the

containment high-range radiation monitor reading and the

radioactivity uniformly dispersed in the containment

atmosphere.

Letter from applicant concerning NUREG-0737, Item II.D.1.1,

"Integrity of Systems Outside Containment."

Letter from applicant concerning audit of electrical

equipment qualification files.

Watts Bar SSER 3 A-9
July 24, 1984

July 27, 1984

July 27, 1984

August 6, 1984

August 7-8, 1984

December 19, 1984

Letter from applicant specifying new fuel load date for

Unit 1 as October 1984 and for Unit 2 as August 1986.

Letter from applicant concerning various comments/

proposed modifications to Unit 1 draft Technical

Specifications.

Letter from applicant concerning NUREG-0737, Item I.C.7, "NSSS Vendor Review of Procedures."

Letter to applicant concerning availability of Supplement

to NUREG-0933, "A Prioritization of Generic Safety Issues"

(Generic Letter 84-19).

Meeting with applicant to discuss the Technical Specifi- cations for Unit 1.

Letter to applicant requesting additional information

concerning the initial test program.

Watts Bar SSER 3 A-IO0
APPENDIX B

BIBLIOGRAPHY

Letter, Apr. 24, 1981, from L. Mills (TVA) to J. P. O'Reilly (NRC), Subject:

Improper Class of ERCW Piping and Components.

--- , June 8,1981, from L. Mills (TVA) to J. P. O'Reilly (NRC), Subject:

Improper Class of ERCW Piping and Components.

--- , July 14, 1981, from L. Mills (TVA) to J. P. O'Reilly (NRC),. Subject:

Improper Class of ERCW-Piping and Components.

--- , Sept. 2, 1981, from L. Mills (TVA) to J. P. O'Reilly (NRC), Subject:

Improper Class of ERCW Piping and Components.

--- , Dec. 9, 1981, from L. Mills (TVA) to J. P. O'Reilly (NRC), Subject:

Improper Class of ERCW Piping and Components.

--- , Feb. 17, 1982, from L. Mills (TVA) to J. P. O'Reilly (NRC), Subject:

Improper Class of ERCW Piping and Components.

Apr. 1, 1982, from D. Hoffman (PWR Owners Group) to H. R. Denton (NRC).

--- , Aug. 11, 1982, from L. Mills (TVA) to J. P. O'Reilly (NRC), Subject:

Improper Class of ERCW Piping and Components.

--- , Oct. 19, 1982, from L. Mills (TVA) to J. P. O'Reilly (NRC), Subject:

Improper Class of ERCW Piping and Components.

--- , Apr. 27, 1983, from L. Mills (TVA) to J. P. O'Reilly (NRC), Subject:

Improper Class of ERCW Piping and Components.

--- , June 14, 1983, from J. Sheppard (Westinghouse Owners Group) to

D. G. Eisenhut (NRC), Subject: Generic Design Package for Incorporation of

Automatic Shunt Trip Feature Into Westinghouse Reactor Protection System.

--- , Sept 28, 1983, from L. Mills (TVA) to J. P. O'Reilly (NRC), Subject:

Improper Class of ERCW Piping and Components.

Lysmer, J., T. Udaka, H. B. Seed, and L. Hwang, "Computer Program for Complex

Response Analysis of Soil-Structure Systems," Apr. 1974.

Lysmer, J. T. Udaka, C. F. Tsai, and H. B. Seed, "Computer Program for Approximate 3-D Analysis of Soil-Structure Interaction Problems," Nov. 1975.

Memorandum, Aug. 26, 1976, from J. P. Knight (NRC) to R. C. DeYoung, Subject:

Seismic Demonstration Program for Electrical Equipment.

Watts Bar SSER 3 B-1
Seed, H. B., and I. M. Idriss, "Simplified Procedure for Evaluating Soil Lique- faction Potential," ASCE, Journal of the Soil Mechanics and Foundations Division,

Sept. 1971.

Tennessee Valley Authority, Topical Report TVA-TR 75-1, Rev. 5, "TVA Quality Assurance Plan," July 11, 1984.

--- , "Final Safety Analysis Report for Watts Bar Nuclear Plant, Units I 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" (includes General Design Criteria), U.S. Government Printing Office,

Washington, D.C.-

U.S. Nuclear Regulatory Commission, NUREG-0588, "Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment," Nov. 1979.

--- , NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants, Resolution

of Generic Technical Activity A-36," July 1980.

--- , NUREG-0654/FEMA-REP-1, "Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power

Plants," Rev. 1, Nov. 1980.

--- , NUREG-0737, "Clarification of TMI Action Plan Requirements," Nov. 1980.

--- , NUREG-0800 (formerly NUREG-75/087), "Standard Review Plan for Review of

Safety Analysis Reports for Nuclear Power Plants - LWR Edition," July 1981

(includes Branch Technical Positions).

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

Bar Nuclear Plant, Units 1 and 2," June 1982; Supplement No. 1, Sept. 1982; Supplement No. 2, Jan. 1984.

--- , NUREG-0966, "Safety Evaluation Report Related to the D2/D3 Steam Generator

Design Modification," Mar. 1983.

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

--- , NUREG/CR-2165, "An Investigation of Buckling of Steel Cylinders With Circular Cutouts Reinforced in Accordance With ASME Rules, "Los Alamos Scientific

Laboratory, July 1981.

U.S. Nuclear Regulatory Commission, Office of Inspection and Enforcement, IE Bulletin 79-02, "Pipe Support Base Plate Designs Using Concrete Expansion

Anchor Bolts,". Mar. 8, 1979.

--- , IE Bulletin 80-06, "Engineered Safety Features (ESF) Reset Controls,"

Mar. 12, 1980.

Westinghouse Topical Report WCAP-7769, "Overpressure Protection for Westing- house Pressurized Reactors," Rev. 1, Oct. 8, 1971.

Watts Bar SSER 3 B-2
--- , WCAP-8624, "General Method of Developing Multifrequency Biaxial Test

Inputs for Bistables," Sept. 1975.

--- , wcAP-8673, "Multifrequency and Direction Seismic Testing of Relays,"

Dec. 1975.

WCAP-8694, "Seismic Qualification of the Rotary Relay for Use in the

Solid-State Protection System," Jan. 1976.

Watts Bar SSER 3 B-3

APPENDIX C

NUCLEAR REGULATORY COMMISSION

UNRESOLVED SAFETY ISSUES

This appendix provides an update to the NRC staff's evaluation of one unresolved safety issue (USI) that is applicable to the Watts Bar facility.

A-46 Seismic Qualification of Equipment in Operating Plants

The scope of Task A-46 is limited to dealing with the seismic qualification of

equipment in currently operating plants. The staff's evaluation of the Watts

Bar seismic qualification of equipment is discussed in Section 3.10 of this

report. The evaluation will not be handled under USI A-46 because it is being handled on a case basis; therefore, the open item identified under this USI

should be deleted.

Watts Bar SSER 3 C-1

APPENDIX E

PRINCIPAL CONTRIBUTORS

NRC Personnel Branch

G.

D.

F.

M.

R.

D.

P.

L.

Y.

W.

S.

R. S.

J.

H.

W.

C.

J. 3.

G.

Bagchi

Becker

.Cherny

Duncan

Giardina

Gupta

Hearn

Hel ler

Hsii

Jensen

B. Kim

Kirkwood

Kirslis

Knox

Li

Long

Nichols

Pulsipher

Rajan

Staley

Seismic Qualification

Procedures and Systems

Mechanical Engineering

Licensing

Power Systems

Geotechnical Engineering

Auxiliary Systems

Geotechnical Engineering

Core Performance

Reactor Systems

Structural Engineering

Mechanical Engineering

Chemical Engineering'

Power Systems

Instrumentat ion and Control

Procedures and Systems

Effluent Treatment

Containment Systems

Mechanical Engineering

Environmental and Hydraulic

Engineering

Chemical Engineering

Review Area

Seismic Qualification

Geotechnical Engineering

-J. Wing

Contractor

EG&G Idaho, Inc.

U.S. Army Corps of

Engineers, Tulsa

District

Watts Bar SSER 3 E-1

/

NRC 1(777)U.S. FORM'335 NUCLEAR 1. REPORT NUMB,, (Assignedby DOC) REGULATORY COMMISSION UE

BIBIOGRAPHIC DATA SHEET SURem-084 -- ~ em tNo3 4. TITLE AND SUBTITLE (Add V me No., if appropriate) 2. (Leave bla

Safety Evaluation Rep t related to the operation of /

Watts Bar Nuclear Plan Units I and .2 3. RECIP NT'S ACCESSION NO.

7.DnuE REPORT COMPLETED

NTH YEAR

9. PERFORMING ORGANIZATION NAME AN MAILING ADDRESS (Include Zip Code)

Division of Licensing

Office of Nuclear Reactor Regu tion

U.S. Nuclear Regulatory Commiss n

Washington, D. C. 20555

12. SPONSORING ORGANIZATION NAME AND MAILI% ADDRESS (Include Zip Code) I

Same as 9 above /

13 TYPE OF REPORT P 4P3OD C

Safety Evaluation Report, Supplement No. 3 /

iDATE REPORT ISSUED

? MONTH IYEAR

January 1985

6. (Leave blank)

8. (Leave blank)

10. PROJECT/TASK/WORK UNIT NO.

11. CONTRACT NO.

OVERED (inclusive dares)

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 Evaluationilejirt, NUREG-0847 (June 1982), Sunplement

No. 1 (September 1982), and Supplement No. 2 ,(Jan~r, 1984) issued by the Office'of Nuclear

Reactor Regulation of the U.S. Nuclear Regu tory Cmmission with respect to the application filed by the Tennessee Valley Authorit ., as app"•icant and owner, for licenses to

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

facility is located in Rhea County, Tenne pee, near th` Watts Bar Dam on the Tennessee

River. This supplement provides recent i.formation reg ding resolution of some of the

open and confirmatory items and license tpnditions ident ied in the Safety Evaluation

Report.

17 E WRSAN OUMN NAYI 17a. DESCRIPTORS

17ti. IDENTIFIERS/OPEN-ENDED TERMS

18. AVAILABILITY STATEMENT . 9.ECURITY CLASS (Thareport) 21. NO. OF PAGES

Unlimited 2 22 PRICE

nC a ssl le S

NFIC FORM 335 17.77)