ML20154K911

Insp Repts 50-259/88-16,50-260/88-16 & 50-296/88-16 on 880501-0611.Violations Noted.Major Areas Inspected: Operational Safety,Maint Observation,Surveillance Testing Observation,Restart Test Program & Q-list Concerns

ML20154K911
Person / Time
Site:	Browns Ferry
Issue date:	09/09/1988
From:	Little W, Paulk G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20154K885	List: ML20154K880 ML20154K889 ML20154K911
References
50-259-88-16, 50-260-88-16, 50-296-88-16, NUDOCS 8809260043
Download: ML20154K911 (20)

See also: IR 05000259/1988016

Text

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y*p* no ,k UNITED STATES

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NUCLEAR REQULATORY COMMISSION

REGION 11

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\-e,,,e *[ 101 MARieTTA ST, N.W.

ATLANTA. GeOAOlA 3e323

Report Nos.: 50-259/SS-16, 50-260/53-16, and 50-296/55-16

I Licensee: Tennessee Valley Authority

6N 38A Lookout Place

1101 Market Street

Chattanooga, TN 37402-2801

i Docket Nos.: 50-259, 50-260, and 50-296

License Nos.: OPR-33. OPR-52, and OPR-6S

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Facility Name: Browns Ferry Nuclear Plant

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Inspection at Browns Ferry Site near Decatur Alabama

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Inspection Conducted: May1-Junj ll, 1988

Inspecto / N 90 0

G. L. Paulk, 61er Resident Inspector Ditt (1gned

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Accompanied by: C. R. Brooks, Resident Inspector

i E. F. Christnot, Resident Inspector

W. C. Bearcen, Resident Inspector

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7 .F

J o h n so n,,

roject Engineer

. Approved by: _ //

W. S. Little /Section Chief,

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at Signed

j Inspecticn Programs,

TVA Projects Division

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SUvyARY

l Scope: This routine inspection was in the areas of operational safety,

i rnaintenance cbservation, surveillance testing observation, restart

test program, Q-List concerns, fuel reconstitution, seismic analysis

of the Standby Gas Treatment Building, employee concerns program,

Unit 2 drywell fire followup report, and licensee action on previous

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inspection findings,

l Results: Inspector Followup Item (259,260,296/SS-16-01): Control of systems

wnile testing per the Restart Test Program (RTP) is in progress.

l (Restart Item)

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Unresobed Item (260/SS-16-02): Quality requirements for components

l not on the Q-List. (Restart Item)

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Violation (259,260.296/SS-16-03): Six examples of failure to comply

with procedures. (Restart It'm)

Inspector Fo11cwup Iten (260/SS-16-04): Verify process to review

ECP investigation reports for reportability. (Restart Item)

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REPORT DETAILS

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1. Licensee Employees Contacted

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, G. Walker, Plant Manager

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P. J. Spiedel, Project Engineer

j "J. D. Martin, Assistant to the Plant Manager i

] *R. M. McKeon, Operations Superintendent i

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f. Ziegler, Superint9ndent - Maintenance i

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D. C. Mims. Manager - Technical Services Supervisor

• J, G. Turner, Manager - Site Quality Assurance

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l M. J. May, Manager - Site Licensing l

"J. A. Savage, Compliance Supervisor i

A. W. Sorrell, Site Radiological Control Superintendent  !

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R. M. Tuttle, Site Security Manager  !

L. E. Retzer, Fire Protection Supervisor  !

) *H. J. Kuhnert. Office of Nuclear Power, Site Representative l

! *T. C. Valetzano, Director - Restart Operations Center  ;

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,0ther licensee employees contacted included licensea reactor operators.

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aaxiliary operators, craftseer, technicians, public safety officers, l

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quality assurance, and design and engineering personnel.  !

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i "Attenced exit interview.  !'

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{ 2. Operational Safety (71707, 71710) .

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The inspectors were kept informed of the overall plant status and any

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{ ,. significant safety matters reitted to plant operations. Daily discussions {

were

staff.held with plant manage. ment and various members of the plant operating i,

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The inspectors Nade routine visits to the control rooms when an inspector

was on site.

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. Observations included instrument readings, setpoints  ;

i and recordings; status of operating systems; status and alignments of

5 emergency standby systems; onsite and of f site emergency power sources

4 .kvailable for automatic operation; purpose of temporary tags on equipment  !

1 controls and switches; annunciator alarm status; adherence to procedures; }

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adhtrence to l irciting conditions for operations; nuclear instruments  :

i operable; temporary alterations in effect; daily tournals and logs; stack i

i monitor recorder traces; and control room manning. This inspection

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activity also included numerous inforcal discussions with operators and

their supervisors.

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General plant tours were conducted on at least a weekly basis. Portions

of the turbine building, each reactor building and outside areas were

visited. Observations included valve positions and system alignment;

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snubber and hanger conditions; contain: rent isolation alignments; instru-

ment readings; housekeeping; proper power supply and breaker, alignments;

radiation area controls, tag contrcls on equipment; work activities in

progress; and radiation protection controls. Informal discussions were  %

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held

tours.

with selec ed plant personnel in their functional areas during these 4j

Within this area no violations or deviations were found.

3.

Maintenance Observation (62703)

Plant maintenance ac+ivities of selected safety-related systems and

components

re observed /reviened to ascertain that they were conducted in

accc

this

h raquirements. The following items were considered during

were a

the limiting conditions for operations were met; activities

.., f i shed usi ng approved procedures; functional testing and/or

calibrations were performed prior to returning components or system to

service; quality control records were maintained; activities were

accomplisheJ by qua'ified personnel; parts and materials used were

properly certified; proper tagout clearance procedures were adhered to;

Technical Specification adherence; and radiological controls were

implemented as required.

Maintenance requests were reviewed to determine the status of outstanding

jobs and to assure that priority was assigned to safety-related equipment

maintenance which might affect plant safety.

Within this area no violations or deviations were found.

4 Restut Test Program (1TP)

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The inspector attended RTP status meetings, reviewed RTP test procedures,

observed RTP Tests and associated tests perfcrmances, reviewed RTP Test

results and attended selected Restart Operations Center (War Room) and

Joint Test Group teetings. The follow.ng are the RTP activities and

associated

reporting period:

activities tronitored and status of testing during this

Observations were made of the LOP /LOCA tests for Unit 2 startup.

Details of the observations will be osered in the next resident

report.

Within this area no violations or deviations were fourd.

5. Cuality "arveillance Report Reviews

The inspector reviewed the follow ag Quality Surveillance Monitoring

Reports:

a. Report C W-S-SS-0'36, dated May 12, 19S8. The Quality Monitoring

Inspector documented the fact that dJring Special Test 88-17,

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Diesel Generator "B" was started with the cylinder vent valves being

left open.

into the test.TheOperations

test was stopped approximately 45 to 50 seconds

generated critique report n aber 88-025 to

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acdress the vent valve problem.

b. Report QBF-S-88-0455, Dated May 8-17, 1938. The Quality Monitoring

Inspector documented the fact that during Special Test 88-17, Diesel

Generator "B"

maximum. was started the load limiter set at zero instead of

These two items are being

(259,260,296/38 *6-01) pending t. racked as an Inspector Followup Item (IFI)

reports.

review of the licensee's response to the

6.

Fuel Reconstitution (60710)

Fuel Reconstitution activities continued throughout the month. The

inspector made weekly visits to the refuel floor to observe the recon-

stitution activities and cor. duct discussions witn the inspection

personnel.

There have been some personnel errors during the initial

inspection and reconstitution activities. These were detected either by

the contractor sunervision or licensee supervision during their reviews of

the paperwork following reconstitution of fuel bundles. The problems can

be categurized :n three arean

a.

Typographical or transcription errors. Rod transfers within the same

bundle and from donor bundles have been erroneously documented on the

fuel buridle Matrix Sheets and rod movement sneets. Changes have been

nade to the sequence of rod movement documentations in order to

prevent recurring deficiencies of this type.

b.

Use of donne fuel rods which haven't been inspected. When the second

donor bt.ndle was selected, rods were used in the reconstitution

process which had not been inspected. A reconstituted bundle was

actually finished and replaced back into the storage ra:ks with rods

that wei'e of unknown quality. This error was attributed to a lack of

f amiliarity of the process by a contractor QC inspector. Retraining

of the inspectors was conducted to prevent recurrence.

c.

Some rods which were determined to be Visual Standard 5 (VS-5) and

unacceptable durir.g the bundle inspection were not removed from the

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seconstituted bundle. This was attributed to an cversight by the

contractor QC inspector who directed rod swaps. The three qt.estion-

able bundles were reinspected and the QC inspectors reinstructed.

The inspectors discussed all of these errors with the licenset's

supervisor of Test Directors and tne contractor managar in charge of

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the reconstitution proc.ess.

CAQR 880377 was written to document the

problems and the corrective action taken. The problems are considered to

be licensee identified and therefore no violations will be issued.

7. General Electric Contractcr Recommendations

(0 pen) Unresolved Item (259,260,296/85-39-04) Licensee Resolution of GE

Report Safety Related Items.

The licensee had contracted with several

outside consultants to perform various evaluations as part of the

Regulatory Derformance Improvement Program (RPIP). The RPIP was imposed

by Confirmatory Order (EA 84-54) on July 13, 1984. In July 1985, the

resident inspector followed up on the General Electric (GE) NSSS recom-

mendations and docueented the results in Inspection Report 259, 260,

296/85-33. Basically, the inspection found that the licensee had not

developed a coordinated program for resolution of numerous deficiencies

ani recommendations identified t,y GE. Subsequent to this, in a NRC

Request for Information pursuant to 10 CFR 50.54(f), dated September 17,

1985, the NRC asked for an evaluation and proposed disposition of

contractor recommendations. TVA responded to this request in the Browns

Ferry Nuclear Performance Plan (Volume 3) Aopendix B, Evaluation of

Contractnr Recommendations.

A followup inspection of the implementation of the above commitments was

conducttti by the resident inspectors and documented in Inspection Report

259,260,296/87-20. During that reperting period the inspectors identified

larious problems with the licensee's program for resolution of the

contrac tor recommendations. These problems included failure to classify

items as restart, failure to include all contractor findings on computer

tracking lists, failure of the Plant Operating Review Committee (PORC) to

review recommendations, and the lack of timely resolution on itams that

had been tracked for extended time periods.

During this reporting period the inspector reviewed the status of the

resolution of GE contractor recommendations. This progra,a included items

associated with 22 NSSS and safety related systems for Unit 2. The items

were reviewed for applicability to Browns Ferry, desirability, and if a

determination of requirenient for restart was performed. As of May 30,

1988, there were a total of 676 iten.s in the tracking program which were

divided into 5 separate categories as shown below:

Category / tion

Total Itemt Comp'eted

A Rewired for restart 49 22

B Applic able/ Desirable 459

'; 260

Applicable /Not Desirable 42 ---

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U Not Applicable 23 ---

Other 103 75

Totals 673 3D

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The licensee stated that the computer tracking list and assignment of each

category haa been reviewed and approved by PORC. Additionally a system

documentation file is prepared for the closecut of each item and will  !

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receive management review o cer the resolution of each recommendation.

This management review effort is approximately 25% complete.

Additionally, the licensee has committed in the Nucle /.r Performance Plan '

to perform a Quality Assurance Surveillance on the system review plan.

The program is currently scheduled for completion by August 1, 1988.

The 1.icensee stated that the QA surveillance would be complete shortly

thereafter.

The inspector reviewed licensee memos dated August 10, 1987 (R40 870810

976) and November 12, 1987 (R40 87110 997) which piovided the criteria and

additional guidelines for determining Category A (restart) items. The

inspector feels that the guidance contained in the above two memos

generally contains adevjate cetail to support the proper classif' cation of

each item. However, the following concerns Lxist:

GE rer,ommendations concerning GE design specs are automatically

classified as Category C (not desirable). The inspector questions

the adequacy of this assumption without evaluating each item on a

case by case basis.

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Any item being worked / tracked / completed by another TVA program such

as a drawing discrepancy or ECN is classified as Category E. The

inspector questf or.s the adequacu of this assumption especially when ,

alternate tracking items such as ECNs can be cancelled. ,

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The inspectors will look at these specific concerns cnd continue to follow

the progress of resolution of contractor recommendatiers in the next

reporting period. There still remains a considerable amount of effort to

resolve the recommendations. This item will remain open pending further ,

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review by the inspectors.

Within this area no violations or deviations were found.

3. Q-List Program Implementation

The inspect-* reviewed the implementation of the Unit 2 Q-List program as

identified a I.E. Inspection Reports 88-05 and 88-10. Follow-up manage-

ment meetings with site program management indicated numerous program and

procedural changes have been implemented to correct the inspector

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identified deficiencies.

To fulfill the commitments in the NPP Volume III, the BFN Unit 2 Phase I

Q-List was implemented on February 26, 1988, listing nuclear sa fe ty-

related components, systems, and structures. The Q-list was imi.lemented ,

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by issuing design drawing #47A302-1, Unit 2 phase I Q-List, and SDSP 3.10,

Use of the Q-List, and deleting Unit-2 components from BF 1.11, Critical

Structures, Systems, and Components (CSSC) List. Some system components

are only required fo the mitigation of abnormal operating transients and

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special events and are not included in the Q-List because of the present

Q-List definition. Components that were on the CSSC list that should have

been included on the Q-li st are s dby licuid control system pumps,

valves, tanks and controls; the vacuv. breaking system; shutdown cooling

mode components of the RHRS; and the fuel pool cooling system. The

following steps were being taken to alleviate concerns regarding those

components:

(1) A review of the Q-List Design Review File, the BFN Safe Shutdown

Analysis (SSA), and the associated System Requirements

Calculations shall be performed to outermine the operating modes

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(and components) not included in ,e Q-List because they were

re quired to funccion in the miti ition of abnormal operating

transients and special events.

(2) For those systems which have operating modes (and components)

for the mitigation of abnormal operating transients and special

events that are determined not to be included on the Q-List

because they are not s a fe ty-re l a t ed , the system designations

i shall be compared to the BFN CSSC to determine that all systems

originally specified on the CSSC are considered in this

evaluation.

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(3) A comparative review and evaluation of components within the

operating modes of steps 1 and 2 will be performed to reduce the

total

set due to any components that appear common to safety-

related operating moces.

(4) The set of components developed through step 3 will be added co

the Q-List on a systematic revision basis with definition of

limited QA program requirements.

(5) A review of the general boundaries of the CSSC and the included

operating modes of the SSA shall be performed to de termine

whether the Q-List for each system is enveloped by the CSSC. If

not, CAQRs will be generated as appropriate.

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(6) Once all systems have been considered, as indicated in steps 1

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through 5 above, 0-List procedures will be revised to indicate

the Unit 2 Q-List will stand alone independent of the Unit I and

3 CSSC list.

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TVA intends to do the following to resolve existing weaknesses in the

Q-List and make the Q-List usable:

(1) A training program is in progress and upper level management

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emphasis has been provided to the appropriate organizations.

The interfacing requirements of SDSP-3.10 and BF 1.11 will be

included in this training.

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(2) SDSP-3.10 was revised to provide direction to users when system

components cannot be located on the Q-List.

(3) SDSP-3.10 and PI 87.52, Development and Control of BFNP Unit 2

Phase I Q-List, were revised to clarify the language regarding

adheience to 10 CFR 50, Appendix 8.

(4) In addition, an evaluation program of system operating modes

(and equipment) for systems of special significance and limited

QA requirement: will be completed and components added to the

Q-List as appropriate. This will eventually ieplace any depen-

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dence on BF 1.11 (and the CSSC listing).

On May 6, 1988, temporary change No. 10 was issued to BF 1.11 to reinstate

the CSSC list for Unit 2.

that were on the CSSC list is an unresolved item.The failure of the Q-list to incl

(260/88-16-02) The

licensee will be asked to respond to this item describing how they ensured

thatCSSC

the the c.uality

list butcontrol activities

were ,;ot on thefor the components that had been on

Q-list were properly specified and

implemented from February 26, 1988 until May 6, 1988.

The inspector learned that the Q-List was prone to misuse and that train-

ing was required for Q-List end-users to alleviate the potential problems

with the Q-List not being a stand alone document. The inspector attended

one of the training classes conducted on May 19, 1988. The pitfall of

"default classification" was stressed many times during the session. A

default classification would be to assume that a component was not safety-

related or not under the program of limited QA controls if it could not be

found

it on the Q-List. Because the Q-List is not a stand alone document and

is being

The only thing issued

thatin acan

phased approach, this assumption cannot be made.

be interpre,ed from the Q-List is that if a

component is on the list, it is safety-related. If a component is not on

the list, a request must be made to DNE to perform a component-specific

classification.

Part of the reason for this is that the Q-List was

developed using DNE "as-designed" drawings. This was prior to the Design

Baseline Verification Program (DBVP) which was to reconcile the deviations

between the as-constructed and as-designeJ drawings. Thus there were

known problems with the Q-List inputs and therefore a lack of conf fderce

exists in the Q-List itself. The inspector found that the training could

have been enhanced by use of several example cases where a comronent would

be selected and the Q-List consulted for a safety classification. This

feedback was provided to the instructor.

During a review of the Q-List Equipment Data Packages (QEOP), the inspac-

tor identified a noncompliance with the lict.asee's procedure. Section 5.7

of BFEP PI 87-52, Development and Control of the Browns Ferry Unit 2

Phase ! Q-List, identifies the info mation required to be placed in the

QEDP.

This includes drawings, commitment /requiremcnt data sheets classi-

fication derivations, correspondence, design baseline program resu,lts, and

other miscellaneous data. Step 5.7.3 of PI 87-52 requires that QEOPs

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shall be controlled as QA Records upon completion. The inspector's review

of the QEDP for System 001, Main Steam, detected a violation of the QA

Records requirements in the Q-List Data Entry, Update and Input Sheets

contained in Pickoff.

Component Tab B1 Analysis Component Pickoff and the Tab B1/82 Analyses

The following examples expressly prohibited by

Section 6.1, QA Records Administration, of the NQAM Part III, Secti0n 4.1,

Quality Assurance Records, were detected:

a.

A majority of the QA record was not in black ink (some entries were

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in light blue and red ink).

b.

A majority of the corrections were not made by marking a single line

through the item to be changed, marking the new entry,

the dated initials of the person making the correction. and entering

c.

There was no name or date included in the reviewer block on a

majority of the (N/A).

not applicable input sheets, nor were these blocks marked as being

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tion asproblems

a violationwere identified to the licensee's compliance organiza-

of the NOAM during the first week of May 1988.

These

B, are considered

Criterion to be examples of a violation of 10CFR 50, Appendix

V, Instructions, Procedures ard Drawings. (259,260,296/

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88-16-03).

9. Employee Concern Program (ECP) ,

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The inspector reviewed the employee concern program at Browns Ferry to

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determine program adequacy and procedural controls. Site Director

Standard 15.1, Employee Concern Program, and Site Director Standard

Practice 15.5, Employee Concerns Handling Procedure, were reviewed for i

adequacy and adherence to regulatory requirements.

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A generic deficiency was identified by the inspector that should be

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addressed

address thebyreview

a program procedure change. SDSP 15.5 does not specifically

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process required of ECP completed investigations to

evaluate the findings for reportability requirements, This concern was

thoroughly discussed with licensee program management. The licensee

J committed to review all issued (26) ECP investigation reports for reporta- *

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bility and correct SDSP 15.5 to address this concern. This item will be

j listed as an inspector followup item to verify completion of this task

(260/88-16-04). This item was identified during the review of ECP Inves-

tigation Report 87-BF-897-Pl.

Witnin this area no violations or deviations were identified.

10. Unit 2 Orywell Fire Followup Report

The

of November NRC conducted

2, 1987. a special in3pection of the fire in the Unit 2 drywell

Details of the inspection are delineated in Inspec-

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tion Report 87-43.

Specific concerns identified in the inspection report

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were responded to by the licersee. This followup report identifies

violations identified by the NRC during the inspection, but held in

abeyance until the final licensee Serious Accident Analysis Report of the

fire was issued. The inspectors have reviewed the final issued Serious

Accident Investigation Team Reports of December 7, 1987, and April 13,

1988. Management concerns identified during the inspection will be

reviewed during management reviews required for Unit 2 restart.

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The following violations were identif;ed during the November 1987 inspec- l

tion:

a. A temporary alteration control form (TACF) was not used to authorize

temporary connections through penetration EE for recirculation system

valve controls and drywell bicwer controls performed under Mainte-

nance Requests (MR) A793993 and A775468. Plant Managers Instruction

(PMI) 8.1, Temporatry Alterations, recuires that long term altera-

tions shall be controlled using a TACF in lieu of other mechanisms

(such as an MR) which are approved for only short term alterations.

These MRs were performed in May and October 1987, and should have

been considered long term alterations. (See Report 87-43, Details,

Section 8.a. page 12)

b. Nuclear Quality Assurance Manual, (NQAM) Part III, Section 4.1,

requires that QA Records shall have all blanks filled in or marked

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N/A. Many MRs were found witt. signatures and data missing. Examples

of these were MR No. A775468 which was missing signature's for

"Raychem acceptable" on 6 pages and signatures for "QC Verification

of Standard Test 1. "on 5 pages; and MR No. AS22017 which was missing

an entry on blocks 26 through 28 wnich should have documented work

performed and cause of failure,

c. PMI 6.2, Conduct of Maintenance, Section 4.4.13, requires that

post-maintenance testing be performed on all plant process equipment

following all corrective maintenance, and some preventive maintenance

and troubleshooting activities that might have impaired proper

functioning of the component.

No electrical checks of any nature were performed as post-maintenance

testing following completiu of the temporary electrical splices

installed under MRs 793993 and 775468. This was attributed to

inadequate controls in Modification / Addition Instruction MAI-45,

Cable Terminating and Splicing fo- Insulated Cables up to 15,000

volts. Also, Electrical Maintenance Instruction (EMI) 7.2, Test

Procedure for Initial Installation and Troubleshooting of Molded Case

Circuit Breakers, was found to be deficient in that it failed to test

the motcr starter portion of the breakers. The starters contain the

thermal overload elements which perform a necessary function for some

modes of end-device failures.

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d.

Three of the six fire brigade members who e..tered the drywell for

fire fighting operations were not eligibse for fire brigade duty due

to failure to comply with the training and qualification requirements

of FPP-1, Fire Protection Program Plan. Additionally, 67 of 126 fire

brigade members assigned to five operating crews were ineligible for

fire brigade duty for the same reasons. (See Report 87-43, Details,

Section 9.a, page 16).

These concerns are considered further examples of the violation of 10 CFR

50, Appendix

260, 296/88-16-03). B, Criterion V, Instructions, Pr;cedures, ano Drawings (259,

11.

Seismic Analysis of the Standby Gas Treatment Building

During a review by the resident inspector of the closure of Unresolved

Item 259,260,296/87-27-03, it was noted that CAQR 87-180 had been issued

in October 1987, rAotive to the concern that the Standby Gas Treatment

Building seismic rasponse spectra in the original plant design basis was

underpredicted.

The inspector noted during the review process that this significan'

concern

licensee.had not been reported to the NRC or adequately evaluated by the

was received. inspector

The reviewed CAQR 87-180 to determine why no rep 0"t

The following deficiencies were noted that violated the

requirements of Site Director Standard Practice 3.7, Corrective Action.

a.

The management reviewer is required to identify if operability at the

nuclear unit could be potentially affected using criteria in Attach-

ment b of the SOSP 3.7. One of the criteria states that any CAQR

citing a plant /FSAR discrepancy shall be reviewed for operability. A

plant /FSA1 discrepancy is defined as a discrepancy between the

as-built facility and the applicable description in the FSAR. A

second criterta in Attacrment 5 states that the technical specifica-

tion required

technical equipment must meet the operability definition in the

specifications.

The reviewer decided this item did not

affect unit operability on the CAQR appropriate section. This was an

incorrect evaluation.

b.

The Plant Operations Review Staff did not complete its evaluation of

the CAQR for ef fect on operability. P0RS had reviewed SCR 86-29

which originally identified this deficiency. P0RS requested followup

information from design in March 1987, in accordance with SOSP 15.2.

No information has been fo, thcomi'ng to present. No effective

followup by P0RS, required procedurally each quarter, was apparent.

c.

The responsible organization is required to evaluate the CAQR for

significance in accordance with criteris in Section 4.12 of 50SP 3.7.

The

and notdesign

reportable. organization determined that the CAQR was not significant

The PORS group did not evaidate the CAQR. The

specified criteria in Section 4.12 states that any event that could

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11

prevent the fulfillment of the safety function of a system needed to

control the release of radioactive material should be noted as

significant.

In fact, the basis of this CAQR was the transfer of

this known deficiency from a previous licensee significant condition

report (SCR 86-29) of February 1937.

a

~ ilure to conduct an adequate CAQR review and evaluation is another

example of the violation against 10 CFR 50, Appendix B, Criterion V and

Site Director Standard Practice 3.7 (260/83-16-03). In response to this

violation TVA is requested to address thcir confidence that CAQR operabil-

ity and reportability determinations have been properly made in the past.

12.

Licensee Action on Previous Enforcement Matters (92802)

(CLO5ED) Violation 259,260,296/85-45-02,

tion activity without the update of drawings. Improper closecut of modifica-

determine the reason for the .iolation; however, The licensee was unable to

the drawings that were

in error have been updated to reflect the as-built configuration. The

valves, and components for drawinginspector toured the area and p

problems.

This item is closed. verification and found no further

1

13.

Followup of Open Inspection Items (92701)

(CLOSED)

ciated with Unresolved Item (259,260,296/86-28-04) Labeling Problems asso-

shutdown board normal -

emergency control nower selector

switch.

The inspectors had identified a concern with adequate circuit

, breaker

board identification labeling in both the 3A and 3B 480 volt .iC shutdown

rooms.

i

The inspector reviewed maintenance request 859228 which

orrected

cion report.the specific discrepancies as described in the original inspec-

Additionally, the inspector toured the 4KV and 480 VAC

shutdown

discrepancies. board rooms in Units 2 and 3 and noted no apparent labeling

adhesive to prevent new labels were noted which appeared to have adequate

Many

loose. recurrence of the problem of many labels becoming

The inspector considers that the licensees ef forts with componeat

identification walkdowns as part of the Design Baseline Verification

Program (08VP)

labeling in this area.has resulted in an improvement of the overall condition of

This item is closed.

(CLOSED) Unresolved Item (259,260,296/87-27-02), Transportation, control

of cor.tamination, and inadequate radiological surveys associated with a

Carbon-14 Tracer used at the Browns Ferry Biothernal Research Facility.

,

A followup inspection :as conducted by NRC Region II on August 4, 1987,

i

to evaluate this item. The details of the inspection are delineated

in Inspection Report 01-16821-02/87-01 issued January 6, 1988, which

satts'actorily closed this issue.

(OPEN)

Unresolved Item (259,260,296/86-28-02), Discrepant scram valve

opening times. In July 1986,

mance of Special Test 86-10, thatthe licensee discovered during the perfor-

several scran inlet and outlet valves

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delayed opening for up to 20 seconds. The licensee researched GE Service

Information Letters 441 and 373 and NRC Information Notice number 86-78 in

an attempt to resolve the anomaly. Another special test (86-26) was

performed on October 22, 1986, to determine the effect of rebuilding the

scram pilot valves on scram valve opening times. The test determined that

opening times improved by up to 15 secor.ds; however, the licensee's

analysis of the data (RIMS R40 870407 930) which was presented to the

Plant Operations Review Committee (PORC) documented that even after

solenoid pilot valve rebuilding, the times remain up to "4 seconds greater

than expected." The licensee considers the anomaly resolved; however,

since General Electric representatives have indicated that four other

plants have experienced similar delays, the anomaly was not specific to i

BFNP. The inspector's review of the data and other reference material l

1. icated that a potential problem exists with the spring tension adjust- I

ment of the scram inlet and outlet valves and a possible excessive

pressure drop across the scram pilot valves. The following issues should

be addressed in closing this item

a,

Acceptance criteria for scram pilot valve timing upon scram air

header blowdown should be addressed. The data already accumulated

supports compliance with this time or perform followup tests to

demonstrate compliance should be considered,

b. Perform e.ither single rod scram testing prior to plant startup or

scram valve time tests prior to plant startup for each scram solenoid

pilot valve that has been refurbished in accordance with the GE

recommendations in SIL No. 441. This is to ensure HCU operability

and to detect further anomalies.

c. The licensee should check the adjustment of all scram valve opening

air pressures which have indicateo a potential for noncompliance with

the recommended spring tension settings in GE SIL No. 373.

(CLOSED) Inspector Followup Item (259,260,296/86-25-02), Control room

emergency ventilation walkdown deficiencies. This IFI was opened to track

numerous deficiencies discoverr.d during a walkdown of the CREV system.

All of the items except two were corrected and dispositioned in Inspection

Report 259, 260, 296/87-46. The remaining concerns were; 1) operator

,

knowledge of damper locations which were required to be checked shut upon

a control room isolation and 2) accessibility of dampers for manual  ;

operation. The licensee revised 01-31 and A01-31, Control Bay Emergency i

Pressurization Operating Instructions and Abnormal Operating Instructions,

t provide damper locations to the operator. The licensee also evaluated

the necessity for remote reach-rod linka es to the dampers but concluded

that since the actuators are not expected to f ail by remote operation, no

need existed for manually operated reach rods. This IFI is closed.

.

(CLOSED) Unresolved Item (259,260,296/85-28-09), Secondary containment  ;

blowout panel deficiencies. The licensee discoverad that an unauthorized I

modification had been made to some secondary containment blowout panels

that would have preventad them from relieving at 26 pounds per !quare foot

!

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13

differential pressure. Thus, secondary containment and various other

safety systems could have been compromised in the event of a tornado

depressurization or a steam break outside prima ry containment. The

lice see repaired the blowout panels and inspected all other blowout

panels for similar problems. An engineering evaluation was performed in

order to evaluate the as-found deficient condition. The evaluation

concluded that failure of certain block walls could have occurred during a

tornado depressurization resulting in

Equipment Cooling Water (EECW) headers.f This ailuretogether

of one of with

the Emergency

a

failure on the other EECW header could have resulted in a total single

EECW. loss of

Since the problem was licensee identified and corrective action has

been completed

this iter- with periodic inspections to prevent re,:urrence included,

is closed.

(OPEN) Inspector Followup Item (259,260,296/86-32-03) Reactor protection

system (RPS) calibration frequency. This item concerned a discrepancy

identified by the inspector between the safety analysis which supported

Technical Specification changes for the new RPS Analog Transmitter and

Trip Units (ATTU) and actual plant practice. The licensee assembled a

documentation package on this concern which would resolve it to their

satisfaction. The inspector noted; however, that QIR EEd BFN 88070, which

was contained in the package, still reported that "present Technical i

Specifications have an 18 month calibration cycle. This is not support-

able for the TOBAR transmitters." The package also noted that calcula-

tions for the calibration frequency of PT-68-95 and PT-68-96 are not yet

completed but will be issued prior

to restart of Unit 2. A licer.se

representative was in formed that this

resolution of these outstanding discrepancies.

IFI will remain open pending

(OPEN)

of drywell Inspector Followup

electrical Item (260/87-33-0G) Post-modification testing

penetrations.

require an inspection of the electrical This concern related to a failure to

containment penetration welds during the

restart. integrated leak rate tests (CILRT) to be performed prior to

A licensee representative stated that it was planned to perform

a soap-bubble leak inspection of the new welds while the containment was

pressurized for the CILRT. This item remains open pending completion of

the planned inspection.

(CLOSED)

296/87-02-02),Unresolved Item Gear

Limitorque (260/87-02-02),

Ratios. The (OPEN) Unresobed Item (259,

inspectors documented a

concern that the Unit 2 High Pressure Coolant Injection (HPCI) system

steam isolation valve, 2-FCV-73-2, may not have been able to close against .

design

ratio.

differential pressure due to improper Limitorque Operator gear

The operator had a 33:1 ratio rather than the reoutred 60:1 ratio.

The 1.'censee had idectified the incorrect gear ratio during review of

valve requirements for IE Bulletin 85-03. As part of the corrective

actions for this item, the licensee has performed an evaluation to deter-

mine any additional valves with unexplained timing differences which could

be due to other errors in gear ratios. In addition to 2-FCV-73-2, valves ,

,

_ _

_ _ _ _ .

_ _ _ -_ _ _ _ _

.. .

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14

3-FCV-69-2 and 3-FCV-69-12 were found to have different timing and gear

ratios.

ECN E-2-P7054 has been worked to change the gear ratio on the valve

2-FCV-73-2 to provide a 60:1 ratio. Since no other Unit 2 valves were

identified which had errors in operator gear ratio the inspector concluded

that the licensee has taken adequate corrective actions to address the

original concerns associated with Unit 2. However, this item associated

Mith Units 1 and 3 will remain open pending review of corrective actions

to correct any other valves with identified gear ratio errors. This item

is closed for Unit 2 only.

(OPEN) Unresolved Item (259,260,296/87-27-03) Standby Gas Treatment

System (SGTS) Blower - Train C Seismic Qualification. The SGTS Blower -

Train C is moanted on a steel frame and held by six vibration isolator

spring mounts. The resident inspector identified that the tait had no

lateral support. TVA evaluated the mounting configuration of the Blower-

Train C and discovered that the Blower was not adequately supported to i

prevent damage during a seismic event. Design change notice B00033C was

issued to correct this deficiency. The inspector reviewed the design

change notice, the associated Work Plan 3303/88, the associated USQD, and

the Bechtel calculation package J.N. 19106. No deficiencies were noted.

The field installation was observed by the inspector. One of the neoprene

pads added during the seismic modification was nottd to be defective

during the field walkdowns. The glue used to attach the neoprene pads had

lost its adherence on one of the six s u pp o .'t s . A CAQR was initiated to

address this deficiency (CAQR 88-385) . This unresolved item will be

evaluated for possible enforcement action

(OPEN) Unresolved Item (260/87-46-04), Update on Adequacy of Heat Tracing

For the Residual Heat Removal Service Water And (RHRSW) Emergency Equip-

l ment Cooling Water (EECW) Systems. A meeting was held between the

l resident inspectors and the BFN Plant Manager on December 7, 1987, at BFN

i

to identify NRC concerns dealing with the heat tracing system / components

j

for the RHRSW/EECW systems located in the intake pumping station. The

inspectors feel that the heat tracing in question thould be considered

important to safety-related due to TVA's commitment to I.E. Bulletin

79-24.

There are several instruments located in the pumping station that are

required for accident mitigation and whose functions would be impaired if

t the small lines providing their inputs were to freeze. Pump discharge

pressure switches PS-67-001, 005, 008, and 011 actuate at a system pres-

sure of 20 psig upon pump actuation to energize the strainers (0-STN-67-A,

B, C, and D) and open the corresponding strainer backwash valves

(0-FCV-67-001, 005, 008, and 011). Pressure differential switches

PDS-67-001, 005, 008, and 011 do nut perform a safety function, but their

failure could prevent the pressure switches from operating. This informa-

tion was derived from three main sources: The Master Component Electrical

List (MCEL), Baseline's System Requirements Calculaticn, and the Q-List.

. _ _ _ _ _ _ _ _ _ . _ _ _ _

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15

All these documents include the above mentioned items as "safety-related".

The MCEL and the Q-List classify the heat tracing temperature switches

TS-23-70, 71, 72, an) 73 as safety-related, but are being revised under

CAQR

from the 870018

MCEL for to remove

these these switches. The Q-List receives its informa

switches.

Upon a visual

station, it was observed that the piping and instrument lines (as w

the strainers themselves) were protected with heat tape and controlled by

thermostats. All lines appeared to be insulated. The possibility of

these lines freezing is the subject of MEB Calculation BWR-M2-751-1. It

has been determined by this calculation that the possibility of the large

diameter (greater than 14 iaches) pipes freezing when the pumps are

running does not exist, but the smallest instrument lines could conceiv-

ably freeze if subjected to extreme cold. This is a credible event since

the pumps, piping, equipment, and specialties are essentially located in

an outdoor environment. This calculation is being revised under the

corrective action for CAQR BFPS70018. The revision was prompted by

concerns that the parameters used in the calculation were overly conserva-

tive or improper (incorrect temperature gradients, no credit taken for

insulation, etc.). Therefore, it is questionable whether the small

instrument lines would be subject to freezing, aspecially with one or nore

of the pumps in each compartment running--contrary to the conclusions of

the existing MEB calculation. The results of the revised calculation will

reveal whether or not freezing of any instrument lines will occur, and the

significance of this freezing (i.e. , loss of EECW flow, room flooding,

etc.).

Special precautions have been taken by TVA to annunciate the condition of

the heat tracing in the evntrol room for proper operator actions to occur. ,

The heat trace automatically initiates at a temperature of 39 F. At a

temperature of 35"F, annunciators TA-23-70, 71, 72, and 73 for pumps A, B,

C, and D respectively inform the operator that the heat tracing equipment

in the pumping station has failed to prevent the piping from reachi..g

potential freezing conditions.

The operator then takes action according

to the Browns Ferry Alarm Response Procedures (BFARP) for panel 9-20. The

actions that are prescribed include a visual inspection of the equipment.

This w>uld allow maintenance to iden t i fy any freezing / rupture problems '

that may exist or have the potential to develop. If the cause of the

failure cannot be determined, the field personnel are instructeo to refer r

'

to GOI-200-1 (Browns Ferry General Operating Instructions), and to issue a

Maintenance Request (MR) on any affected equipment or instruments that are

found to be frozen due to inoperative heat tracing. GOI 200-1 in turn

guides

Corrective the actions

operator toto the Electrical

prevent / correctMaintenance Instructions (EMI-46).

station could then be initiated. freezing of piping in the pumping

No specific actions are described by the BFARP for panel 9-20 other than *

writing an MR against af fected equipment. It is left to the discretion

of field personnel / shift supervisor to determine what corrective actions

are appropriate.

In this respect, the existing response procedures are

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16

deficient and should be revised to provide more explicit corrective

actions when freezing situations are encountered.

The RHRSW/EECW heat tracing system is basically corarised of thermostats

set at predetermined setpoints that regulate the protected piping. In

order to ensure reliable service from this rather simple configuration,

ecrtain features

maintenance planhave been incorporated inte the system and a preventative

devised. The system receives Class 1E power from the

l 480V Diesel Auxiliary Boards. Failure of the heat tracing is annunciated

in the control room, as previously mentioned, when the thermostats fail to

actuate

summarily at oispatched.

the desired temperature and an Assistant Unit Operator (AV0) is

The operator can then monitar the heat tracing

system on a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> basis if necessary. The hr.at tracing is inspected

annually by implementation of the EMI-46 Freeze Protection Program as part

of the Preventative Maintenance schedule performed by Electrical Main-

tenance.

TVA feels their program commitment to NRC Bulletin 79-24,

regarding the heat tracing for the RHRSW/EECW systems, is currently being

met with the existing freeze protection programs in existence.

The heat tracing comprises an important part of the freeze prevention

program for the RHRSW/EECW systems. However, TVA feels credit can be

taken for operator actions during an emergency since the time frame

,

l

involved regarding freezing and possible rupture or RHRSW/EECW related

l'.nes is of an extended nature. As mentioned earlier, no specific actions

are delineated by procedures once a potential freezing situation arises.

TVA considers no portions of the heat tracing system to be safety-related.

However, the inspector regards the system as important-to-safety and

operational procedures are required to more specifically instruct the

operator on the necessary corrective actions that would need to be taken

to preclude freezing of RHRSW/EECW lines.

The folloiving need to be addressed prior to closing this item.-

a.

Procedures should be upgraded to reference operator actions required

if a freezing event occurs,

b.

The system must be evaluated in light of the Q-List program for

applicability to IMPORTANT-TO-SAFETY and/or LIMITED QA components.

c.

Calculation

reviewed forMEB-BWR-M2-751-1

credibility. and associated results should be

(CLOSED)

'/icinity ofUnresolved the intakeItem (259/260/296/86-25-03), Heavy Loads in the

structure. The licensee, in response to the

delineated concern, responded b;-

a.

Evaluating placement of a fully loaded crane over the point of

minimal earth cover and the resulting surface loading;

b.

Calculation

ture; of the transmission of that load to the conduit struc-

_______ __ _ _ _ _ _ _ _ _ _ _

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c.

Summation of the crane's load along with the other loads including

dead load, hydraulic gradient, eartn pressure, etcetera;

d. Demonstration that the load imposed on the structure is less than

the allowable stresses;

e. Calculation of the maximum live load that may be placed on the

surface and which the conduit structures will support along with the

margin of safety associated with that load;

f. Delineation of the area boundaries appearing on drawing 37N200

betweer, the pumping station and the oil storage tanks to which that

maximum live load apples;

g.

Identification of any other sub grade structures, conduits or piping

which are more limiting within that area.

Design charge 3473 was conducted to address these concerns. The inspector

reviewed the calculation package and action items completed by the

licensee.

Calculations revealed that load limits could be increased at

the intake from 35 ton to 150 ton. Also, the recommended loading distance

from to the intake well was increased from 3 feet to 10 feet. A FSAR

change was submitted to update the FSAR to the correct amounts. This item

is closed.

(CLOSED) Unresolved Item (259,260,296/81-37-03), Containment atmosphere

dilution

pancies.

system valve identification and Technical Specification discre-

It was identified by the licensee that valves FSC-84-8A, 8B, SC,

SD are containment isolation valves; however, they are not PCIS group 6

isolation valves. They receive no logic input for automatic closure as a

result of a PCIS group sigr.il . The use of these valves is administra-

tively controlled.

They are normally closed valves. The only time these

valves are open is for the injection of nitrogen into primary containment

for combustible gas control af ter a LOCA has occurred. Therefore, there

is no reason to perform a closure time test on these valves.

FCV-84-19 does not close on a group 6 containment isolation signal. This

outboard valve is normally closed, and its use is administratively

,

controllea by the shif t supervisor with a keylock switch. Venting con-

!

tainment for pressure control during normal operation is accompli.1ed via

FCV-84-20, per 01-64

I When operated with HS-64-35 as specified in 01-64,

FCV-84-20 will close upon receipt of a PCIS group 6 containment isolation

{ signal. Furt he rmo re , to vent through FCV-84-19 would require either

FCV-64-29 or FCV-64-32 to be open; however, these valves are interlocked

closed with the mode switch in the run position.

FCV-84 23, -20, 8A, -88, -8C, -SD are cycled monthly in SI 4.7.G.1.A, CAD

sy stem operability test.

4.7.0. FCV-84-20 is tested for closure time in SI

l

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TVA has submitted to Site Licensing a proposed change to the Technical

Specification tables for containment isolation valves in response to

inspector followup item 86-40-07. The proposed changed would delete

Tables 3.7.8 - 3.7.H, and revise table 3.7. A to include all containment

isolation valves including those specified above. Table 3.7.A was

originally intended to identify only those valves which were part of a

specific PCIS group logic. TVA has committed te revise and update Tech-

nical Specification 3.7.A prior to startup of Ucit 2, PORC approval is

scheduled for June 1988, with NSRB approval by July 14, 1988. Therefore,

this item will be closed and tracked under IFI 86-40-07.

(CLOSED) Unresolved Item (259,260,296/87-26-04), Special Nuclear Material

shipment deficiencies noted. On June 18, 1987, the licensee reported an

irregularity involving a shipment of Special Nuclear Material (SNM) to

another licensed facility. The licensee shipped what was thought to be

five intermediate range monitors (IRM), each containing 1 milligram of

Uranium-235 to Peach Bottom Nuclear Station on June 16, 1987. Peach

Bottom personnel informed Browns Ferry via telephone on June 18, 1987,

that a sixth IRM was received in the shipment. A followup inspection by

Regional inspectors was conducted on July 27-29, 1987, as detailed in I.E.

Report 87-29. Violations of regulatory requirements ware identified.

Therefore, this item will be closed out and tracked under the open items

listed in I.E. Report 87-29,

14. Exit Interview

The inspection scope and findings were summarized on June 10, 1988, with

those persons indicated in paragraph 1 above. The inspectors described

the areas inspected and discussed in detail the inspection findings listed

below. The licensee did not identify as proprietary any of the material

provided to or reviewed by the inspectors during this inspection. Dis-

senting comments were not received from the liccnsee.

Item Number Description and Reference

259,260,296/88-16-01

IFI - Control of Systems while testing per

,

the Restart Test Program (RTP) is in pro-

gress, paragraph 5.

1

260/88-16-02 Unresolved item - Quali+.y requirements for

l

components not on Q-list, paragraph 8.

259,260,296/88-16-03 Violation - Six examples of failure to

comply with procedures, paragraph s 8, 10

and 11.

260/88-16-04

IFI - Verify the process for reviewing ECP

investigations reports for reportabil:ty,

paragraph 9.