NUREG-0847

Supplement No. 1

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

September 1982
NOTICE

Availability of Reference Materials Cited in NRC Publications

Most documents cited in NRC publications will be available from one of the following sources:

1. The NRC Public Document Room, 1717 H Street, N.W.

Washington, DC 20555

2. The NRC/GPO Sales Program, U.S. Nuclear Regulatory Commission,

Washington, DC 20555

3. The National Technical Information Service, Springfield, VA 22161

Although the listing that follows represents the majority of documents cited in NRC publications,

it is not intended to be exhaustive.

Referenced documents available for inspection and copying for a fee from the NRC Public Document Room include NRC correspondence and internal NRC memoranda; NRC Office of Inspection

and Enforcement bulletins, circulars, information notices, inspection and investigation notices;

Licensee Event Reports; vendor reports and correspondence; Commission papers; and applicant and

licensee documents and correspondence.

The following documents in the NUREG series are available for purchase from the NRC/GPO Sales

Program: formal NRC staff and contractor reports, NRC-sponsored conference proceedings, and

N RC booklets and brochures. Also available are Regulatory Guides, NRC regulations in the Code of

Federal Regulations, and Nuclear Regulatory Commission Issuances.

Documents available from the National Technical Information Service include NUREG series

reports and technical reports prepared by other federal agencies and reports prepared by the Atomic

Energy Commission, forerunner agency to the Nuclear Regulatory Commission.

Documents available from public and special technical libraries include all open literature items,

such as books, journal and periodical articles, and transactions. Federal Register notices, federal and

state legislation, and congressional reports can usually be obtained from these libraries.

Documents such as theses, dissertations, foreign reports and translations, and non-NRC conference

proceedings are available for purchase from the organization sponsoring the publication cited.

Single copies of NRC draft reports are available free upon written request to the Division of Technical Information and Document Control, U.S. Nuclear Regulatory Commission, Washington, DC

20555.

Copies of industry codes and standards used in a substantive manner in the NRC regulatory process

are maintained at the NRC Library, 7920 Norfolk Avenue, Bethesda, Maryland, and are available

there for reference use by the public. Codes and standards are usually copyrighted and may be

purchased from the originating organization or, if they are American National Standards, from.the

American National Standards Institute, 1430 Broadway, New York, NY 10018.

GPO Printed copy price: _$4.50
NUREG-0847

Supplement No. 1

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

September 1982

ABSTRACT

This report supplements the Safety Evaluation Report, NUREG-0847, issued June

1982 by the Office of Nuclear Reactor Regulation of the U.S. Nuclear Regulatory

Commission with respect to the application filed by-the Tennessee Valley Authority, as applicant and owner, for licenses to operate the Watts Bar Nuclear Plant,

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

County, Tennessee, near the Watts Bar Dam on the Tennessee River. This supplement provides recent information regarding resolution of some of the open items

identified in the Safety Evaluation Report and discusses recommendations of the

Advisory Committee on Reactor Safeguards in its report dated August 16, 1982.

Watts Bar SSER 1 i ii

TABLE OF CONTENTS

Page

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

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

1.1 Introduction ........................................... I-I

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

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

3 DESIGN CRITERIA - STRUCTURE, COMPONENTS, EQUIPMENT AND

SYSTEMS ..... ............................................... 3-1

3.10 Seismic and Dynamic Qualification of Seismic

Category I Mechanical and Electrical Equipment ......... 3-1

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

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

5.4.2 Steam Generators ................................ 5-1

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

13.6 Physical Security Plan ................................ 13-1

13.6.1 General ...................................... 13-1

13.6.2 Physical Security Organization ................ 13-1

13.6.3 Physical Barriers ............................. 13-1

13.6.4 Access Requirements ........................... 13-2

13.6.5 Detection Aides ............................... 13-3

13.6.6 Communications ................................. 13-3

13.6.7 Test and Maintenance Requirements ............. 13-3

13.6.8 Response Requirements ......................... 13-4

13.6.9 Conclusions.................................... 13-4

19 REPORT OF THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS....... 19-1

APPENDICES

A CHRONOLOGY OF RADIOLOGICAL REVIEW OF WATTS BAR NUCLEAR PLANT,

UNITS 1 AND 2, OPERATING LICENSE REVIEW

B BIBLIOGRAPHY

E PRINCIPAL CONTRIBUTORS

F ACRS REPORT

G ERRATA TO WATTS BAR SAFETY EVALUATION REPORT

Watts Bar SSER 1 V

1 INTRODUCTION AND DISCUSSION

1.1 Introduction

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

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

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

the Watts Bar Nuclear Plant, Units I and 2. This report is the first supplement to that Safety Evaluation Report (SER).

This supplement provides more recent information regarding resolution of some

of the open items identified in the SER. This supplement also provides and discusses the recommendations of the Advisory Committee on Reactor Safeguards (ACRS)

in its report on the Watts Bar Nuclear Plant, dated August 16, 1982. 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 identified in the SER.

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

same as the section or appendix of the SER that is being updated, and the discussions are supplementary to and not in lieu of the discussion in the SER unless

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

of the safety review. Appendix B is an updated bibliography. Appendix E is a

list of principal contributors to this supplement. Appendix F is a copy of the

ACRS report. Appendix G is a list of errata for the SER. No changes in SER

Appendices C and D have been made by this supplement.

The Project Manager is Thomas J. Kenyon; he may be reached on (301) 492-7266.

1.7 Summary of Outstanding Issues

In Section 1.7 of the SER the staff noted that certain items remained outstand- ing at the ti-me the report was issued. This supplement updates some of those

items for Which the staff has completed part or all of its review. These items,

and the sections of this supplement discussing the staff's review conclusions,

are

(4) Seismic qualification of equipment (3.10)

/(7) Model 0-3 steam generator preheater tube degradation (5.4.2.2)

(15) Physical Security Plan (13.6)

1.9 License Conditions

In Section 1.9 of the SER the staff identified 37 license conditions. There

is an additional issue for which a license condition may be desirable to ensure

that staff requirements are met during plant operation. This item, with the

appropriate reference. to its section in this supplement, is

(38) Physical Security Plan (13.6.4)

Watts Bar SSER I 1-1

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

3.10 Seismic and Dynamic Qualification of Seismic Category I Mechanical and

Electrical Equipment

The following evaluation applies to Unit 1 of the Watts Bar facility only. The

Unit 2 review and evaluation will be performed at a later date, and the results

will be reported in a future supplement to the SER.

The staff's evaluation of the adequacy of the applicant's program for qualification of safety-related electrical and mechanical equipment for seismic and

dynamic loads consists of (1) a determination of the acceptability of the procedures used, standards followed, and the completeness of the program in general,

and (2) an onsite audit of selected equipment items to develop the basis for the

staff judgment on the completeness and adequacy of the implementation of the

entire seismic and dynamic qualification program.

The Seismic Qualification Review Team (SQRT) has reviewed the equipment dynamic

qualification information contained in the pertinent Final Safety Evaluation

Report (FSAR) Sections 3.9.2 and 3.10. The SQRT made a site visit on April 26

through April 30, 1982 to determine the extent to which the qualification of

equipment (as installed in Watts Bar Unit 1) meets the current licensing criteria as described in Regulatory Guides 1.92 and 1.100 and the Standard Review

Plan (SRP) Section 3.10 (NUREG-0800). A representative sample of seismic Category I mechanical and electrical equipment, as well as instrumentation, included

in both nuclear steam supply system (NSSS) and balance-of-plant (BOP) scopes,

were selected for the plant site review. The review consisted of field observations of the actual equipment configuration and its installation, followed by

the review of the corresponding test and/or analysis documents.

In instances where components have been qualified by test or analysis to other

than current licensing criteria such as Regulatory Guides 1.92 and 1.100 and

SRP Section 3.10, the applicant is currently re-evaluating the acceptability of

the qualification program in the light of current criteria.

In the SQRT onsite audit exit meeting and the trip report of the SQRT site

visit, the staff concluded that in order to complete the review, it would need

additional information and clarification of the details of the qualification

for some pieces of equipment. The trip report includes both generic and

equipment-specific concerns. The applicant will be required to address and

resolve the specific concerns before initial fuel load.

The generic concerns are significant, in that they apply to all safety-related

equipment and can potentially affect a large number of equipment items. Therefore, an acceptable approach and a plan to implement the resolution of generic

issues will also be required before fuel load. A list of the generic concerns

is summarized below:

Watts Bar SSER 1 3-1
Generic Open Items

(1) Single-axis and single-frequency tests were performed to qualify equipment. For equipment in the flexible range (below 33 Hz) these tests may

not challenge the multi-axis and multi-frequency response of the equipment. The extent to which this issue can affect the qualification status

of equipment at the plant is the thrust of this concern.

(2) In numerous cases the required response spectra (RRS) were not broadened

at the peaks to account for the uncertainty in the prediction of natural

frequencies of the supporting structures. Also, sufficient margins must

be included in the test response spectra (TRS) to account for the uncertainty in manufacturing process and the test apparatus.

(3) In numerous cases the field mounting of the equipment is by welding of

various lengths, whereas the mounting for the qualification testing is by

bolting. Changes in the field mounting could alter the dynamic characterristics of the system invalidating the results of the original qualification testing. Devices within electrical cabinets could be more susceptible

to this type of problem.

(4) Many safety-related equipment are age sensitive with respect to their

seismic performance; for example, the insulation of mot.ors,.transformers

and other electric devices, and diaphragms for valves may limit the qualified life.. To ensure that seismic resistance of safety-related equipment

is available throughout the plant life, a detailed program of surveillance

and maintenance must be developed.

Resolution of the specific and generic items as they progress will be reported

in a future supplement to the SER.

As discussed in Section 3.10 of the SER, the seismic hazard for the Watts Bar

site was redefined by the staff during the operating license review. The seismic hazard is now specified in terms of a site-specific spectrum, which corresponds to the 84th percentile spectrum shape derived from a number of recorded

time histories of ground motion. Whether or not the seismic input used for

qualification of safety-related equipment is conservatively bounded by the input

actually used for qualification was a concern. By submittals dated December 9,

1981, the applicant provided a comparison between the Watts Bar spectrum against

the 84th percentile spectrum. These comparisons were made for higher values of

damping representative of the structural damping and they established the adequacy of the Watts Bar spectrum. Because equipment damping values are somewhat

lower than the structural damping values, the applicant, in a submittal dated

April 21, 1982, provided additional comparisons between the 84th percentile spectrum and the Watts Bar spectrum for lower damping values representative of equipmenit damping values. During the site audit, the staff examined the selected

equipment input spectra incorporated in the purchase specification against the

84th percentile spectrum and confirmed that the Watts Bar spectrum was properly

used and that it adequately covers the 84th percentile spectrum. Thus there is

reasonable assurance that the seismic input used for the qualification for

safety-related equipment is at least as severe as can be expected from the

redefined seismic hazard for the Watts Bar site.

Watts Bar SSER 1 3-2
5 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS

5.4 Component and Subsystem Design

5.4.2 Steam Generators

5.4.2.2 Westinghouse Model D Steam Generator Tube Degradation Potential

As discussed in the SER, a generic problem concerning the potential for tube

degradation caused by flow-induced vibration in the preheater section of Westinghouse Model D steam generators has been identified. Because the Watts Bar

facility uses the Model D-3 steam generator, the applicant is working with Westinghouse, Duke Power Company, and South Carolina Electric and Gas Company (the

latter two being owners of facilities with Model D-2/D-3 steam generators) in

resolving this concern.

During a June 30 - July 1, 1982 meeting, Westinghouse made presentations regarding their proposed modifications to the Model D-2/D-3 steam generators to the

Design Review Panel (DRP) with the NRC staff in attendance. The 19 member DRP

was established by TVA, Duke Power Company, and South Carolina Electric and Gas

Company to (1) evaluate the final proposed modifications to the preheater to

ensure that applicable regulatory criteria are met, (2) verify the functional

adequacy of the proposed design, and (3) ensure the safety acceptability of the

design. The DRP is comprised of representatives of the above utilities (or

their consultants) with expertise in such disciplines as materials engineering,

structural mechanics, radiation exposure,, quality assurance, cooling and insulation engineering, and thermal-hydraulics.

On August 25, 1982, the staff again met with representatives of the DRP and

Westinghouse to discuss the scope and content of the DRP's safety evaluation

of the proposed Westinghouse modifications to the model D-2/D-3 steam generators. The NRC staff made recommendations to the DRP to assist them with their

review of these modifications. During these and other meetings with the DRP

and Westinghouse, the staff has monitored the analysis and development work by

Westinghouse to resolve this matter. The vendor has developed a manifold to be

installed in the feedwater inlet nozzle of the steam generator, based on the

model testing and analytical programs. The manifold will disperse the-fluid

flow so that impingement of the water on. the rows of tubes in the preheater

region will be reduced and thereby eliminate the vibration responsible for the

accelerated wear of the steam generator tubes. The final design review of'this

manifold and related .modifications is anticipated to be completed by the DRP in

mid-September and will be presented to the staff at that time.

The procurement of materials, the tooling, and the procedures for the proposed

modifications already have been completed by Westinghouse. Training is expected

to be complete by mid-September 1982.

During the August 13, 1982 ACRS full committee meeting, TVA stated that the

utility has determined that it would be to their advantage to install the manifold and complete any necessary modifications before hot functional testing at

Watts Bar SSER 1 5-1
the Watts Bar facility. However, it has not been decided which facility (Watts

Bar, Summer, or McGuire) will be the first domestic facility to receive the

modifications.

The staff will continue to monitor developments of the Westinghouse program and

will evaluate the proposed modifications and the DRP report upon final submittal.

The staff will address this matter in a future supplement to the SER.

Watts Bar SSER 1 5-2
13 CONDUCT OF OPERATIONS

13.6 Physical Security Plan

In the SER the staff stated that appropriate revisions to the Watts Bar Physical

Security Plan had not been filed. By letter dated July 29, 1982, TVA submitted

their formal revisions to the plan. The following is the staff's evaluation of

TVA's security program plans and a summary of how the applicant has provided

for meeting the requirements of 10 CFR 73.

13.6.1 General

TVA filed the following security program plans with the NRC for the Watts Bar

Nuclear Plant Units 1 and 2.

(1) "Watts Bar Nuclear Plant Physical Security Plan," dated June 3, 1982 as

revised by letter of July 29, 1982

(2) "Watts Bar Nuclear Plant Contingency Plan," dated December 21, 1980 with

revision dated March 30, 1981

(3) "Watts Bar Nuclear Plant Guard Training and Qualification Plan," dated

August 17, 1979 with revision dated January 24, 1980, May 21, 1980,

October 1, 1980 and March 9, 1981

13.6.2 Physical Security Organization

To satisfy the requirements of 10 CFR 73.55(b), TVA has provided a physical

security organization that includes a Public Safety Service Shift Supervisor

who is onsite at all times with the authority to direct the physical protection

activities. To implement the commitments made in the physical security plan,

guard training and qualification plan, and the safeguards contingency plan,

written security procedures specifying the duties of the security organization

members have been written and are available for inspection.

The training program and critical security tasks and duties for the security

organization personnel are defined in the "Watts Bar Nuclear Plant Guard Training and Qualification Plan," which meets the requirements of 10 CFR 73, Appendix B, for the training, equipping, and requalification of the security organization members. The physical security plan and the training program provide

commitments that preclude the assignment of any individual to a security-related

duty or task before the individual is properly trained, equipped, and qualified

to perform the assigned duty in accordance with the approved guard training and

qualification plan.

13.6.3 Physical Barriers

In meeting the requirements of 10 CFR 73.55(c), TVA has provided a protection

area barrier that meets the requirements defined in 10 CFR 73.2(f)(1). To permit observation of activities along the barrier, an isolation zone of at least

Watts Bar SSER 1 13-1
20 ft is provided on both sides of the barrier with certain specified exceptions. The staff has reviewed these exceptions and determined that the security measures in place are satisfactory and continue to meet the requirements

of 10 CFR 73.55(c).

Illumination of 0.2 ft-candles is maintained for the isolation zones, protected

area barrier, and external portions of the protected area. In areas where illumination of 0.2 ft-candles-cannot be maintained, special procedures are applied.

Vital equipment is located within vital areas that are located within the protected area and requires passage through at least two barriers, as defined in

10 CFR 73.2(f)(1) and (2), to gain access to the vital equipment. Vital area

barriers are separated from the protected area barrier.

Patrols of the protected area are performed at random intervals to detect the

presence of unauthorized persons, vehicles, and materials.

The control room and central alarm station are provided with bullet-resistant

walls, doors, ceilings, floors, and windows.

13.6.4 Access Requirements

In accordance with 10 CFR 73.55(d), all points of personnel and vehicle access

to the protected area are controlled. The individual responsible for controlling the final point of access into the protected area is located in a bulletresistant structure. As part of the access control program vehicles (except

under emergency conditions), personnel, packages, and materials entering the

protected area are searched for explosives, firearms, and incendiary devices by

electronic search equipment and/or physical search.

Vehicles admitted to the protected area, except applicant designated vehicles,

are controlled by escorts. Applicant designated vehicles are limited to onsite

station functions and remain in the protected area except for operational, maintenance, repair, security, andemergency purposes. Positive control over these

vehicles is maintained by personnel authorized to use the vehicles or by the

escort personnel.

A picture badge/key card system, using encoded information, identifies individuals that'are authorized unescorted access to protected and'vital areas; it is

used to control access to these areas. Individuals not authorized unescorted

access are issued non-photo badges that indicate an escort is required. Access

Authorizations are limited to those individuals who have a need for access to

perform their duties.

Unoccupied vital areas are locked and alarmed. Access to the reactor containment(s) is'positively controlled to ensure that only authorized individuals are

permitted to enter. In addition'all doors and personnel/equipment hatches into

the reactor containment(s) are locked and alarmed. Keys, locks, combinations,

and related equipment are changed on an annual basis. In addition, when an

individual's access authorization has been terminated as a result of the lack

of reliability or trustworthiness, or for poor work performance, the keys,

locks, combinations, and related equipment to which that person has access are

changed.

Watts Bar SSER 1 13-2
Section 9.1 of TVA's physical security plan allows designation of the containment as a nonvital area when the fuel is out of the core during major refueling

and major maintenance. The staff finds this is in violation with the regulations and is, therefore, unacceptable. The license will be conditioned to

ensure that the containment shall not be designated as nonvital under any

circumstances.

13.6.5 Detection Aides

In satisfying the requirements of 10 CFR 73.55(e) the licensee has installed

intrusion detection systems at the protected area barrier, at entrances to

vital areas, and at all emergency exits. The licensee has exceeded the regulation by providing two perimeter-intrusion detection systems at the protected

area barrier. Alarms from the intrusion detection systems annnunciate within

the continuously manned central alarm station and a secondary alarm station

located within the protected area. The central alarm station is located so

that the interior of the station is not visible from outside the perimeter of

the protected area. In addition, the central alarm station is constructed so

that the walls, floors, ceilings, doors, and windows are bullet-resistant.

The alarm stations are located and designed so that a single act cannot interdict the capability of calling for assistance or responding to alarms. The

central alarm station contains no other function or duty that would interfere

with its alarm response function.

The transmission lines and associated alarm annunciation hardware for the intrusion detection systems are self-checking and tamper-indicating. Alarm annunciators indicate the type of al.arm and its location when activated. An automatic

indication of when the alarm system is on standby power is provided in the

central alarm station.

13.6.6 Communications

As required in 10 CFR 73.55(f) the licensee has provided for the capability of

continuous communications between the central and secondary alarm station operators, guards, watchmen, and armed response personnel through the use of a conventional telephone system and a security radio system. In addition, direct

communication with the local law enforcement authorities is maintained through

the use of a conventional telephone system and two-way FM radio link. All nonportable communiction links, except the conventional telephone system, are provided with an uninterruptable emergency power source backed up by diesel

generators.

13.6.7 Test and Maintenance Requirements

In meeting the requirements of 10 CFR 73.55(g) the applicant has established a

program for the testing and maintenance of all intrusion alarms, emergency

alarms, communication equipment, physical barriers, and other security-related

devices or equipment. Equipment or devices that do not meet the design performance criteria or have failed to otherwise operate will be compensated for by

appropriate compensatory measures as defined in the "Watts Bar Nuclear Plant

Physical Security Plan" and onsite procedures. The compensatory measures

defined in these plans will ensure that the effectiveness of the security system is not reduced by failures or other contingencies affecting the operation

of the security-related equipment or structures.

Watts Bar SSER I 13-3
The intrusion detection systems are tested for proper performance at the beginning and end of any period that they are used for security. Such testing will

be conducted at least once every 7 days.

Communication systems for onsite communications are tested at the beginning of

each security shift. Offsite communications are tested at least once each day.

Audits of the security program are conducted once every 12 months by the Quality

Assurance and Audit Staff, which is independent of site security management and

supervision. The audits, focusing on the effectiveness of the physical protection provided by the onsite security organization in implementing the approved

security program plans, include, but are not limited to (1) a review of the

security procedures and practices, (2) system testing and maintenance programs,

or (3) local law enforcement assistance agreements. The Quality Assurance and

Audit Staff prepares a report documenting their findings and recommendtions and

submits it to the Director of Nuclear Power and the plant superintendent.

13.6.8 Response Requirements

In meeting the requirements of 10 CFR 73.55(h) TVA has provided for armed

responders immediately available for response duties on all shifts consistent

with the requirements of the regulations. In addition, to provide additional

response support in the event of security events, liaison with local law

enforcement authorities has been established and documented.

The applicants' safeguards contingency plan for dealing with thefts, threats,

and radiological sabotage events satisfies the requirements of 10 CFR 73, Appendix C. The plan identifies appropriate security events that could initiate a

radiological sabotage event and it identifies TVA's preplanning, response resources, safeguards contingency participants, and coordination activities for

each identi'fied event. Through this plan, response activities using the available resources would be initiated upon:the detection of abnormal presence'or

activities within the protected or vital areas. The response activities and

objectives include the neutralization of the existing threat by (1) a requirement that the response force members interpose themselves between the adversary

and their objective, (2) instructions to use force commensurate with that used

by the adversary, and (3) authority to request sufficient assistance from the

local law enforcement authorities to maintain control over the situation.

To assist in the assessment/response activities a closed circuit television

system, providing the capability to observe the entire protected area perimeter,

isolation zones, and a majority of the protected area, is provided to the security/ organization.

13.6.9 Conclusions

Based on a review of the applicants' security program plans listed in Section 13.6.1 of this report and on visits to the site, the staff has concluded

that the protection provided by TVA against radiological sabotage at the Watts

Bar Nuclear Plant meets the requirements of 10 CFR 73, provided the license condition discussed above is imposed. Accordingly, the protection provided will

ensure that the health and safety of the public will not be endangered.

Watts Bar SSER 1 13-4
19 REPORT OF THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS

During its 268th meeting on August 13, 1982, the Advisory Committee on Reactor

Safeguards (ACRS) reviewed the application of the Tennessee Valley Authority

for licenses to operate the Watts Bar Nuclear Plant, Units 1 and 2. This application also was considered at Subcommittee meetings held on April 30, 1982 in

Knoxville, Tennessee, and on August 10, 1982 in Washington, D.C. Members of

the Subcommittee toured the facility before the April 30, 1982 meeting. Transcripts of each of these meetings are available from Alderson Reporting, 400

Virginia Avenue, S.W., Washington, D.C. 20024. Copies of transcripts also are.

available for review at the NRC Public Document Room at 1717 H Street, N.W.,

Washington, D.C. 20555, and at the Chattanooga-Hamilton County Bicentennial

Library, 1001 Broad Street, Chattanooga, Tennessee.

A copy of the ACRS Report on the Watts Bar Plant, Units 1 and 2, dated August 16,

1982, is included as Appendix F of this supplement. The report indicates the

ACRS' belief that, subject to the satisfactory completion of construction, staffing, and preoperational testing and if due regard is given to the items in the

letter, there is reasonable assurance that the Watts Bar Nuclear Plant, Units 1

and 2, can be operated at core power levels up to 3411 MWt without undue risk

to the health and safety of the public. A discussion of the items mentioned in

the August 16, 1982 letter follows below. Another supplement to the Watts Bar

SER will be issued before fuel loading of Unit 1 to provide additional information on those items, -as appropriate.

(1) Design and Construction Quality Assurance

The Committee noted that the effects of the breakdown of TVA's quality assurance

program in the design and construction areas still persist. Consequently, the

Committee has requested to be kept informed of the results of TVA's major quality assurance programmatic changes; including the applicants' plans to have an

independent contractor review the design and construction of a typical "vertical

section" of the plant to confirm the adequacy and safety of the as-completed

plant.

Region II has just completed an inspection effort (see Inspection Report 50-390/

82-05) devoted to determining how responsive the staff of the TVA Office of

Engineering Design and Construction (OEDC) have been in resolving TVA and NRC

audit findings. The inspection effort entailed review of the TVA Nuclear. Safety

Review Staff (NSRS), OEDC, QA, and NRC inspection report findings related to

OEDC activities. In addition, followup inspections were performed to determine'

the extent of problem identification, documentation, investigation, evaluation,

corrective action, and implementation of the corrective action.

The TVA audit group findings indicated that several management control programs

were not sufficiently adequate to ensure that requirements or commitments of the

QA program would be met. The majority of the problems were judged by TVA to be

programmatic. The most serious of the deficiencies were found in the areas of

QA program requirements, QA program applicability, engineering procedures for

the control of the design process and changes to the design, interface control,

Watts Bar SSER 1 19-1
and construction activity planning. TVA findings indicate that implementation

deficiencies also were identified; however, these specifically identified problems were not considered by the audit group to be as serious as those involving

the program. The OEDC has developed, during 1981, an extensive 1982 action plan

for quality improvement as a direct result of the problems identified by the TVA

audit groups and the NRC.

TVA's 1982 Action Plan is a program to perform an aggressive, indepth investigation and evaluation of the identified problem areas. The plan is designed to

focus corrective action on the root causes of the problems that have adversely

affected the quality assurance program. The TVA Action Plan consists of separate tasks for the OEDC QA staff, Division of Engineering and Design (ENDES),

and Division of Construction (CONST). Certain root causes of inadequacies in

the problem areas have been identified in the following categories: (1) positive attitude and approach, (2) authority and responsibility, (3) timeliness

and responsiveness, (4) procedures, (5) commitment control and requirements

definition, and (6) retaining experienced personnel. Each group having an

assigned task in the Action Plan (OEDC QA, ENDES, CONST) has developed a master

schedule to provide traceability relating to each root cause, action element,

and time schedule. Much of the 1982 Action Plan activities remain in the investigative and evaluation stages. All indications are that the investigation,

evaluation, and determination of corrective action activities will be completed

by late fall of 1982. Implementation of most corrective action elements should

be started by the first of 1983.

The staff has concluded that the past 12-month effort by OEDC to investigate and

evaluate the problem areas, identify root causes, develop the TVA Action Plan,

and determine the course for corrective action has been a very aggressive and

monumental task. Further review of these areas will continue during future NRC

inspections.

Based on its evaluation of the TVA audit findings and the extensive ongoing

effort associated with the TVA Action Plan, the staff has concluded that the

plan is comprehensive and commendable and, when completed and implemented,

should result in a well defined interdivisional program that will provide reasonable assurance for both the adequacy of design and construction of future TVA

facilities. However, with respect to Watts Bar, the staff's review of. both the

internal audit findings and TVAAction Plan Task Group findings, to date, as

well as previously identified NRC concerns, has reinforced the conclusion that

programmatic deficiencies did and possibly still do exist to varying degrees in

the following areas: (1) design control, (2) transmission of design information

to construction, (3) the construction process, and (4) the ability of QA to

verify the adequacy of the as-built product. The staff does not believe the

corrective actions associated withithe TVA Action Plan alone, when implemented,

will provide the necessary assurance that the as-built product at Watts Bar is

actually in accordance with the specified design requirements. Therefore, the

staff has concluded that further design verification of the as-built conditions

at the Watts Bar facility is necessary.

In Section 1.1 of the SER, the staff stated that TVA proposed to have the Institute of Nuclear Power Operations (INPO) perform the audit, but the staff reserved

judgment as to the acceptability of this proposal. During the August 13, 1982

ACRS Committee meeting, TVA stated they no longer believed that the INPO review

program was appropriate for use on the Watts Bar facility.

Watts Bar SSER 1 19-2
By letter dated September 9, 1982 and during a September 17, 1982 meeting with

the staff, TVA committed to having an independent organization (Black and Veatch)

perform an audit on the safety-related auxiliary feedwater system of the Watts

Bar facility to ensure that the system has been designed and constructed in

accordance with the license application and license commitments. The review is

intended to provide a comprehensive assessment of TVA's design and construction

activities at the Watts Bar facility and to provide confirmation of the adequacy

of the system. This independent design verification will be compiled with

broader, more comprehensive programmatic reviews (e.g., NSRS Review of Watts

Bar, United Engineers Design Verfication Program Review, and Theodore Barry and

Associates Review of OEDC) to verify that the Watts Bar facility is designed

and constructed adequately even though deficiencies in the QA program have been

identified.

The auxiliary feedwater system was chosen by TVA to be a representative system

of the facility because

(1) it was essential to plant nuclear safety

(2) it was primarily of TVA's design

(3) there was a clearly defined design basis

(4) it was representative of safety-related equipment

(5) the design and construction of the system required interfacing with

several disciplines

(6) essential areas have been completed at the plant site

(7) it is moderately complex both physically and functionally

The staff considers these criteria to be reasonable and the selection of the

auxiliary feedwater system to be appropriate for this aspect of the design

review. The independent design verification is expected to be completed by

Black and Veatch by the end of December 1982. Upon receipt and review of this

information, the staff will address this matter in a future supplement to the

SER.

(2) Model D-3 Steam Generator Preheater Tube Degradation

The Committee noted that the Watts Bar units have Westinghouse Model D-3 steam

generators. As discussed in Section 5.4.2.2 of the SER, steam generators of

this design have experienced tube degradation in the preheater region. The

Committee indicated its desire to be kept informed of the steps being taken to

resolve this problem.

The present status of this open item is discussed in Section 5.4.2.2 of this

supplement. New developments will be addressed in a future supplement to the

SER.

Watts Bar SSER 1 19-3
(3) Surveillance Requirements for ERCW Cement Mortar Lining

The Committee recommended that periodic inspections and tests of the cement

mortar lining in the essential raw cooling water (ERCW) pipeline be carried

out to ensure the ERCW system will not be subject to sudden entrainment of

debris resulting from the deterioration of the lining.

Discussion of the seismic qualification of the lining can be found in Section

9.2.1 of the SER.

The staff currently is assessing the need for TVA to develop these surveillance

requirements. The staff will address this matter in a future supplement to the

SER.

(4) Hydrogen Detection and Mitigation System

The Committee noted that they expect to review the Sequoyah and Watts Bar

hydrogen mitigation systems in the near future. By letters datedApril 23,

1982 and June 14, 1982, TVA presented additional information regarding their

research program on hydrogen control. TVA has selected a 120-V ac thermal

ignitor to be installed in place of the 12-V ignitor system currently installed

in the Sequoyah facility. Staff review of the information received is expected

to be completed by November 1982, at which time a supplement to the Sequoyah

SER (NUREG-O011) will be issued. TVA intends to use the same mitigation system

in both the Sequoyah and Watts Bar facilities. As stated in Appendix C of the

SER, the staff has requested additional justification for TVA's contention that

the work performed on this matter for Sequoyah is directly applicable to Watts

Bar. The staff will consider the additional justification in its continuing

review of this issue for the Sequoyah and Watts Bar facilities.

The Committee recommended that specific attention be given to ensuring the reliability of the hydrogen monitors used in conjunction with this system. The regulatory requirement for providing hydrogen monitors is prescribed in 10 CFR 50.44.

The staff's recommended performance features for the hydrogen monitors are contained in Regulatory Guide 1.97. Additional guidance is provided in Item II.F.1

of NUREG-0737, "Clarification of TMI Action Plan Requirements."

Staff reviews for compliance with Item II.F.1 are completed in conjunction with

the operating license (OL) reviews for all new plants. These same reviews are

scheduled for completion by December 1982 for all the operating plants. The

staff's evaluation of this item for Watts Bar Units 1 and 2 appears in Section

6.2.6 of the SER.

The ice condenser containments for the Watts Bar facility will be equippedwith

a distributed ignition system to deal with the hydrogen releases associated with

postulated degraded core accidents. The distributed ignition system will be

manually actuated by the operator upon receipt of a safety injection signal.

The hydrogen monitors will provide data for use in management of theplant following the onset of an accident and for use in postaccident analysis efforts.

However, these data are not essential for the operator's decision to actuate the

distributed ignition system. nor are they essential for the successful performance

of the system.

Watts Bar SSER 1 19-4
The staff has had a program to evaluate the performance of hydrogen monitors

for some time. The staff believes that there is no technological problem that

might prevent compliance with the requirements. Recent licensee event reports

(LERs) indicate that some licensees are experiencing difficulties with their

hydrogen monitoring system. The staff will review these reports to determine

whether any change to NRC licensing requirements is warranted. If such a

change is needed, the staff will report on this matter in a supplement to the

SER.

Watts Bar SSER 1 19-5

June 15, 1982

June 17, 1982

June 21, 1982

June 21, 1982

June 23, 1982

June 23, 1982

June 24, 1982

June 30-

July 1, 1982

July 9, 1982

July 9, 1982

July 12-15, 1982

July 19, 1982

July 19, 1982

July 20, 1982

APPENDIX A

CHRONOLOGY OF RADIOLOGICAL REVIEW

OF WATTS BAR NUCLEAR PLANT, UNITS 1 and 2,

OPERATING LICENSE REVIEW

Generic Letter 82-12-- Nuclear Power Plant Staff

Working Hours.

Generic Letter 82-13 -- Reactor Operator and Senior

Reactor Operator Examinations.

Letter from applicant concerning submergence of

electrical equipment.

Letter from applicant concerning options for

compensation of the steam generator reference leg.

Letter from applicant concerning buried seismic

Category I piping systems and tunnels.

Letter from applicant concerning askarel-insulated

transformers.

Letter to applicant concerning revised SER regarding

the boron dilution event concern.

Meeting with applicant to review geotechnical

engineering information.

SER issued.

Letter from applicant concerning compliance with

Appendix R.

Meeting with applicant to confirm the findings of the

Power Systems Branch in SER.

Letter from applicant concerning technical instructions

related to the leakage reduction program.

Letter from applicant forwarding generic security

training and qualification plan.

Letter from applicant concerning implementation of the

requirements of 10 CFR 73.21.

WATTS BAR SSER APP A A-1
July

July

21,

26,

July 29,

July 30,

July 30,

August 9,

August 23

August 25

August 31

September

1982

1982

1982

1982

1982

1982

1982

1982

1982

9, 1982

Letter to applicant forwarding flow diagrams for the

pump and valve inservice testing program.

Letter from applicant concerning proposed modifications

to the draft radiological effluent technical

specifications.

Letter from applicant concerning the Physical Security

Plan.

Letter from applicant concerning the preservice

inspection program.

Letter to applicant concerning loose parts monitoring

program.

Generic Letter 82-14 -- Submittal of Documents to the

Nuclear Regulatory Commission.

Letter to applicant transmitting ACRS Letter of

August 16, 1982.

Letter from applicant concerning TVA schedule for

completion of the evaluation for control of heavy

loads.

Letter from applicant concerning schedule for process

control program for radioactive waste solidification.

Letter from applicant concerning TVA's proposal to per- form an independent design review for the Watts Bar

facility.

Meeting with applicant to discuss TVA proposal for

the QA independent design verification.

September 17, 1982

WATTS BAR SSER APP A A-2
APPENDIX B

BIBLIOGRAPHY

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

Plant, Units 1 and 2, October 4, 1976.

U.S. Nuclear Regulatory Commission, NUREG-O011, "Safety Evaluation Related to

Operation of Sequoyah Nuclear Plants, Units 1 and 2," March 1979.

--- , NUREG-0737, "Clarification of TMI Action Plan Requirements," November

1980.

--- , NUREG-0800, "Standard Review Plan for Review of Safety Analysis Reports for

Nuclear Power Plants---LWR Edition" (includes Branch Technical Positions),

July 1981.

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

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

Region II Inspection Report 50-390/82-05.

, Regulatory Guide 1.92, "Combining Modal Responses and Spatial Components

in Seismic Response Analysis," Rev. 1, February 1976.

Regulatory Guide'l.100, "Seismic Qualification of Electric Equipment for

Nuclear Power Plants," Rev. 1, August 1977.

Watts Bar SSER 1 B-1

APPENDIX E

PRINCIPAL CONTRIBUTORS

NRC Personnel

M. Duncan

C. Gaskin

J. Pulsipher

D. Quick

G. Bagchi

Branch

Licensing Branch No. 4

Physical Security Licensing Branch

Containment Systems Branch

Region II

Equipment Qualification Branch

Watts Bar SSER I E-1

Appendix F

ACRS REPORT

Watts Bar SSER1

"V,, UNITED STATES

~ NUCLEAR REGULATORY COMMISSION ~ .1 -'--ADVISORY COMMITTEE ON REACTOR SAFEGUARDS

WASHINGTON, 0. C. 20555

110

August 16, 1982

Honorable Nunzio J. Palladino

Chai rman

U.S. Nuclear Regulatory Commission

Washington, D.C. 20555

Dear Dr. Palladino:

SUBJECT: ACRS REPORT ON WATTS BAR NUCLEAR PLANT, UNITS 1 AND 2

D uring its 268th meeting, August 12-14, 1982, the Advisory Committee on Reactor Safeguards reviewed the application of the Tennessee Valley Authority

(TVA) for authorization to operate the Watts Bar Nuclear Plant, Units 1 and

2. The project was considered at ACRS Subcommittee meetings in Knoxville,

Tennessee on April 30, 1982, and in Washington, D.C. on August 10, 1982.

Members of the Subcommittee toured the facility on April 30, 1982. In its

review, the Committee had the benefit of discussions with representatives of

TVA, Westinghouse Electric Corporation, and the NRC Staff. The Committee

also had the benefit of the documents listed. The Committee commented on

the construction permit application for the Watts Bar Nuclear Plant in a

report dated September 21, 1972.

The Watts Bar Nuclear Plant is located in Rhea County in southeastern

Tennessee, about 45 miles north-northeast of Chattanooga, Tennessee. Each

of the two identical units uses a Westinghouse nuclear steam supply system

with a rated core power of 3411 MWt and has an ice-condenser containment

wi th a design pressure of 15 psig. TVA estimates that Watts Bar Nuclear

Plant, Units 1 and 2 will *be ready for fuel loading by August 1983 and

August 1984., respectively.

A number of items have been identified by the NRC Staff as Outstanding

Issues, Confirmatory Issues, and License Conditions. These matters should

be resolved in a manner satisfactory to the NRC Staff.

Late in the construction program a serious quality assurance breakdown was

identified - principally in the construction area, but also in the design

area. 'the effects of the breakdown persist, and corrective work -on the

plant will continue at least throughout 1982. TVA invoked major quality

,assurance programmatic changes, including plans to have an independent

contractor review the design and construction of a typical "vertical section"

of the plant, to confirm the adequacy and safety of the as-completed plant.

This issue should be resolved in a manner satisfactory to the NRC Staff.

We wish 'to be kept informed.

82-07 180072

Watts Bar SSER 1F- F-1
Honorable N. J. Palladino -2 - August 16, 1982

Both Watts Bar Nuclear Plant units have Westinghouse Model D-3 steam aenerators. Steam generators of this design have experienced tube failures,

apparently related to flow-induced vibrations in the preheater region. TVA

has stated that this problem is being worked on by Westinghouse and that a

resolution involving internal modifications is expected before the projected

fuel load date for Unit 1. We wish to be kept informed.

TVA is using a cement mortar lining in the essential raw cooling water

system piping to reduce the pressure drop from corrosion-induced roughness.

We believe that periodic inspections and tests of this lined piping should

be carried out so that, if the .bonding or quality of the coating should

unduly deteriorate, the system will not be subject to sudden entrainment of

debris.

TVA is developing a hydrogen ignition system using controlled distributed

ignition sources. The system to be used at the Watts Bar Plant will be of

the same design as the permanent system to be installed .at the Sequoyah

Nuclear Plant. We expect to review that system in the near future.. We

recommend that specific attention be given by the NRC Staff to assuring the

reliability of the hydrogen monitors used in conjunction with this system.

Acceptability of this system has been designated as a License Condition by

the NRC Staff.

The ACRS believes that, if due regard is given to the items mentioned above,

and subject to satisfactory completion of construction, staffing, and

preoperational testing, there is reasonable assurance that the Watts Bar

Nuclear Plant, Units 1 and 2 can be operated at core power levels up to 3411

MWt without undue risk to the health and safety of the public.

Additional comments by ACRS member D. Okrent are presented below.

Sincerely,

P. Shewmon

C-hairman

Additional Comments by ACRS Member D. Okrent

With regard to the seismic design, I recommend that TVA and the NRC Staff

conduct studies to-evaluate the margins available to accomplish safe shutdown, including long-term heat removal, following an earthquake of somewhat

greater severity and lower likelihood than the safe shutdown earthquake.

I believe it is important that there be considerable assurance that the

Watts Bar SSER 1 F-2
Honorable N. J. Palladino - 3 - August 16, 1982

combination of seismic design basis and margins in the seismic design is

such that this accident source represents an acceptably low contribution to

the overall risk from this plant.

References:

1. lennessee Valley Authority, "Watts Bar Nuclear Plant Final Safety

Analysis Report," with Amendments 1-46.

2. U. S. Nuclear Regulatory Commission, "Safety Evaluation Report Related to the Operation of Watts Bar Nuclear Plant, Units 1 and 2,"

NUREG-0847, dated June 1982.

Watts Bar SSER 1 'F-3

APPENDIX G

ERRATA TO WATTS BAR SAFETY EVALUATION REPORT

Section Page Change

1.3 1-8 Change quantity listed for "containment volume,

ft 3 "' for Watts Bar to 1,191,500

1.3 1-8 Change quantities listed for "Total reactor

coolant flow-rate, lb/hr" to 144,800,000 for

Watts Bar, 138,100,000 for Sequoyah, and

144,800,000 for McGuire

1.3 1-8 Change quantity listed for "Secondary steam

flow-rate, lb/hr" for Sequoyah to 15,140,000

1.7 1-12 In Item (10), change "(6.3)" to "(5.2.2)"

1.8 1-13 Add "(40) Submergence of electrical equipment

as result of LOCA (8.3.3.1.1)"

1.9 1-14 In Item (26), change "(13.1.2)" to "(13.1.3)"

4.4.6 4-21 Change quantity listed for "Coolant flow-total

flow rate (106 lb/hr)" for Sequoyah to 144.8

6.2 6-4 In third paragraph, change "1,191,000 ft 3"' to

"1,191,500 ft 3

" and "110,400 ft 3 '' to

"110,500 ft 3

"

6.2 6-5 Change quantities listed for "Ice Condenser"

Volume to 110,500 and "Total Containment Volume

(ft 3 )" to 1,191,500

Watts Bar SSER 1 G-1

NRC FORM 335 1. REPORT NUMBER (Assigned by DDC)

(7-77) U.S. NUCLEAR REGULATORY COMMISSION NUREG-0847

BIBLIOGRAPHIC DATA SHEET Supplement No. I

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

Safety Evaluation Report related to operation of the

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

7. AUTHOR(S) 5. DATE REPORT COMPLETED

MONTH -7YEAR

September 1982

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

U. S. Nuclear Regulatory Commission MONTH YEAR

Office of Nuclear Reactor Regulation September 1982

Washington, D. C. 20555 6. (Leave blank)

8. (Leave blank)

12. SPONSORING ORGANIZATION NAME AND MAILING ADDRESS (Include Zip Code) 10. PROJECT/TASK/WORK UNIT NO.

Same as 9 above

11. CONTRACT NO.

13. TYPE OF REPORT PERIOD COVERED (Inclusive dates)

15. SUPPLEMENTARY NOTES 14. (Leave blank)

Docket Nos. 50-390 and 50-391

16. ABSTRACT (200 words or less)

This report supplements the Safety Evaluation Report, NUREG-0847, issued June 1982 by the

Office of Nuclear Reactor Regulation of the U.S. Nuclear Regulatory Commission with

respect to the application filed by the Tennessee Valley Authority, as applicant and

owner, for licenses to operate the Watts Bar Nuclear Plant, Units 1 and 2 (Docket Nos.

50-390 and 50-391). The facility is located in Rhea County, Tennessee, near the Watts

Bar Dam on the Tennessee River. This supplement provides recent information regarding

resolution of some of the open items identified in the Safety Evaluation Report and discusses recommendations of the Advisory Committee on Reactor Safeguards in its report

dated August 16, 1982.

17. KEY WORDS AND DOCUMENT ANALYSIS 17a. DESCRIPTORS

17b. IDENTIFIERS/OPEN-ENDED TERMS

18. AVAILABILITY STATEMENT 19, SECURITY CLASS (This report) 21. NO. OF PAGES

Unlimited Unclassified 20. JECUJRITY CLASS (Thispage) 22. PRICE

Inc assified s

NRC FORM 335 (7-77)