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Brunswick Nuclear' Project i' ''

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[ Southport, NC 28461-0429>

December 27, 1989 ;

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-FILE: -B09-135100 -10CFR50.73SERIAll , BSEP/89-1119

.U.S. Nuclear--Regulatory Commission#: ATIN:- Document-Control Desk .

Washington DCD 20555 ;:

-BEUNSWICK STEAM ELECTRIC PLANT UNIT 2DOCKET NO. 50-325 a

LICENSE NG. DPR-71 -!'' LICENSEE EVENT REPORT 1-89-024' .c

^ Gentlemen:g

y :Ini accordance wit!h Title 10 to the Code of Federal Regulations, the enciased ,

: Licensee Event Report 1s' submitted. -This report fulfil.ls the requirement for|a written report within thirty (30), days of.a reportable ~ occurrence and is in i

;accordance with the format set.forth in NUREG-1022, September 1983. ?!

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Very truly yours, [t\

J. . Harness, General Manager f<

Brunswick Nuclear Project't

TMJ/jlh-

'h- Enclosure-

,cc: -Hr. S. D. Ebneter.,

L c Hr. E. G. TourignyBSEP NRC Resident Office

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Brunswick Steam Electric Plant Unit 1 0 |5 | 0 |0 | 0 D | 2 !5 1 |0F|0 15T"''''' Failure.to Test Seventeen Primary Containment 1 solation Valves Per Tech Spec 4.6.1.1.a ,

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L6C80s&Et CONTACT FOR TH78 ttF H3)j( NAMt TELEPHONE NUMeth

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- Amt A COD 4| Theresa M. Jones, Regulatory Compliance Specialist

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On November 27 1989, it was determined as reportable that s e nteen primary|'| containment isolation valves were not being tested in accordance with !

|; Technical Specification (T/S) Surveillance Requirement 4.6.1.1.a. Sixteen of :

L the seventeen valves were installed by Plant Modifications (PMs) and theseventeenth was removed as a temporary repair and then reinstalled. The root

i cause of the f ailure to identify' appropriate testing requirements for thevalves installed by PMs was determined to be a failure to identify applicabledesign criteria and regulatory commitments in the design basis documentation.The root cause for the failure to incorporate the seventeenth valve back intothe appropriate test has not yet been determined. An investigation is still

' underway into the root cause for the failure to develop appropriate designbasis documentation and failure to reincorporate the resinstalled valve backinto the test. A revision to incorporate these valves into the appropriate

'' test has been completed. A review of each primary containment penetration isunderway to ensure appropriate testing is being carried out on the remainderof the valves. A supplement to this report will be issued by June 30, 1990.This event had minimal safety significance.

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NihN PAPEmwo na v i ON 3 toOF MANAGEMENT AND SvDOff.WASHINOTON.DC 20603.

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Event-

On November 27, 1989, it was determined as reportable that seventeen primary 1containment isolation valves were not being tested in accordance with |Technical Specification (T/S) Surveillance Requirement 4.6.1.1.a. 1

Initial Conditions / Event Description

Unit 1 reactor was at 100*. power and Unit 2 was shut down in Refuel Outage No. 8.During e review of common'(i.e., for both Unit 1 and Unit 2) Periodic Test02.2.4a Primary Containment Integrity Verification-Containment External, '

Operations personnel determined that seventeen valves were not being verified '

by'the PT that appeared to meet the requirements for verification inaccordance with' Technical Specification Surveillance Requirement 4.6.1.1.a. A

! nelf-identified Significant Condition Adverse to Quality (SCATQ), 89-045, wasinitiated on October 12, 1989, in accordance with Plant Procedure (PLP) 04,

'Corrective Action' Program.

Event Investigation

'As a result of SCATQ 89-045, Nonconformance Report (NCR) S-89-102 was .initiated on October 25, 1989. The NCR is based on the fact that contrary tothe design criteria for code "D1" valves, listed in System Description (SD) 12,Primary Containment Isolation System (PCIS) (EIIS/JM), the subject valves (seeAttachment A) were not being checked per PT-02.2.4a. An investigationdetermined that sixteen of the seventeen valves had been installed via PlantModifications (PMs) (see Attachment A) without identification of the designcriteria or applicable T/S. The PMs were installed by the Nuclear EngineeringDepartment (NED) and the former Brunswick Engineering Support Unit (BESU).(BESU responsibilities have now been divided between NED and plant Outage .

Management and Modification (0&M) group.) In addition, the PM review byOperations and Technical Support Group personnel failed to identify thesurveillance requirements applicable to the involved valves.

The remaining valve, 2-E11-V127 (the body drain isolation valve to the ResidualHeat Removal B loop Suppression Pool Cooling Isolation valve), had been removed

- as a temporary repair in accordance Engineering Evaluation Request 86-0224 and'the requirement to test the valve had consequently been removed fromPT-02.2.4a. However, the valve was replaced by Direct Replacement 87-195during the following outage without revising PT-02.2.4a to reincoporate thenecessity to test the valve.'

The root cause of the failure to identify the appropriate surveillancerequirements and include the valves in PT-02.2.4a was determined to be the"

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failure of the PMs design basis documentation to identify the applicable

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design criteria and regulatory related commitments per Engineering Procedure I

(ENP) 03, Plant Modification Procedure. Contributing to the failure of j

Operations-and Technical Support personnel to identify the applicable'

surveillance requirements was a'1985 Technical Specification Interpretation ;

(TSI), 85-01, that defined PCIS valves as those listed in ENP-16, Procedure for j

iAdministrative Control of Inservice Inspection Activities, which lists valvesrequiring Local'' Leak Rate Testing (LLRT). . This TSI indicated that if a valve

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did not require LLRT it was not a primary containment concern.. It is feltthis contributed to the failure of personnel to identify the testing )requirements'while performing. associated procedure revirion requirement'

|& views. On November 6, 1987, the TSI was reissued to correctly refer toSu-12.

Root Cause.

The root cause of the failure of the PMs design basis documentation to |identif; the design criteria and applicable regulatory commitments has not yet

~been determined. An investigation is continuing and a supplement to this'report will be issued by June 30, 1990.

.

Corrective Actions ,

Revision 16,'of PT-02.2.4a, implemented on October 17, 1989, incorporated thesubject valves.

The Technical Support group is currently reviewing each primary containmentpenetration against SD-12, PT-02.2.4 and 02.2.4a, and the Final Safety

, Analysis Report (FSAR) to determine and initiate any additional revisions to'

these documents. This review is scheduled to be complete by June 1, 1990.

NED is currently investigating this event in accordance with its DesignDeficiency Program, NED procedure 3.18. This item is scheduled to be resolvedby February 1, 1990.

An investigation into the cause of the failure to reincorporate 2-E11-V127fr.to PT-2.2.4a is continuing. The results will be included in the supplement.

Event Assessment

This event had minimal safety significance. The containment isolation system'. is designed to prevent or limit the release of radioactive material that may

result from postulated accidents. During this event personnel would have beenable to identify a PCIS problem and take appropriate actions to mitigate arelease to the Reactor Building. Secondary containment is designed to prevent'*

a release to the environment. In addition, the Unit 1 and 2 CAC (EIIS/BB)

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valves, CAC-V164, V166, and V169 each have inboard redundant-isolation valveswhich are and were verified by.PT-02.2.4a. The reactor vessel head inner sealifes between the reactor and the Unit 1 or 2 B21-V83. If both the inner sealand the B21-V83 were to leak, the leakage would be identifiable by condensationand sound associated with steam escaping to the Reactor Building fifty-footatmosphere. The 2-E11-V127 is located on the outboard side of the 2-E11-F024Bdisc, The E11-F024B-acts as a redundant isolation and is tested in accordance

with LLRT procedures. In addition, the line associated with the E11-F024B-discharges below suppression pool water level which results in a water sealnormally being present;. The , remaining. valves, Unit 1 and 2 B21-V160, V161,V162i and V163, exist on fluid' filled lines supplying the N026A and B reactor1cvel transmitters. Leakage associated with these valves would be apparent by- |

1evel-indications problems and associated water leakage would have been .Iidentified by personnel in the Reactor. Building.

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ATTACilMENT An.

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' . 1-CAC-V164,-V166, V16980-133

80-134 2-CAC-V164, V166, V169

82-287C 1-B21-V83

~ 82-288C 2-B21-V83

86-007 1-B21-V160, V161, V162, V163-

86-008 2-B21-V160, V161, V162, V163i!- Direct Replacement

87-195 2-E11-V127 ,

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Valve No. -Valve Name

21(2)'-CAC-V164 CAD N Inj ction Line Vent Valve (EIIS/BB/VTV)2

l'(2)-CAC-V169 CAD N ;Injecti n Line Vent Valve (EIIS/BB/VTV) !2

l- 1(2)-CAC V166' Suppression Pool Purge Exhaust Line Vent Valven (EIIS/BB/VTV)' !

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1(2)-B21-V83. Test Valve for Excess Flow Check Valve (EFCV) B21-F008 -!

(EIIS/IJ/TV)L !

[ 1(2)-B21-V160 Test Valve .for EFCV B21-IV-2456 (EIIS/IG/TV) )

~ 1(2)-B21-V162 . Test Valve for EFCV B21-IV-2456 (EIIS/IG/TV).:

I|; .1(2)-B21-V161 Test Valve for EFCV B21-IV-2455 (EIIS/IG/TV)

1(2)-B21-V163 Test Valve for EFCV B21-IV-2455 (EIIS/IG/TV) '

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2-E11-V127 Valve E11-F024B Body Drain Isolation Valve (EIIS/B0/*/ISV)'

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