ML20150E065

Forwards Slide Presentation Used During 880608 Enforcement Conference Held to Discuss Details of Insp Rept 50-285/88-15.Responses to Question Raised During Conference Also Encl

ML20150E065
Person / Time
Site:	Fort Calhoun
Issue date:	07/05/1988
From:	Morris K OMAHA PUBLIC POWER DISTRICT
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
LIC-88-537, NUDOCS 8807140304
Download: ML20150E065 (40)

Text

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Omaha Public Power District 1623 Harney Omaha. Nebraska 68102 2247 402/536 4000 July 5, 1988 LIC-88-537 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Mail Station Pl-137 Washington, DC 20555

References:

1.

Docket No. 50-285 2.

Letter from the NRC (L. J. Callan) to OPPD (R. L. Andrews) dated June 1, 1988 3.

Enforcement Conference - June 8, 1988 Gentlemen:

SUBJECT:

Inspection Report 88 June 8, 1988 Enforcement Conference On June 8, 1988, OPPD met with members of the NRC to discuss the detalis of an NRC inspection as documented in report 50-285/88-15.

In Attachment I are copies of the overheads used during the June 8, 1988 presentations.

In i are responses to questions that were raised during the conference.

OPPD appreciates the time given to discuss our position and actions relative to Inspection 88-15 and believes this meeting was very beneficial in communicating this information.

If you have any additional comments or questions concerning this matter, please do not hesitate to contact us.

Sincerely, Sbo J. Morris

$8 Division Manager

• O Nuclear Operations od gg KJM/me ec p

cc: LeBoeuf, Lamb, Leiby & MacRae om 1333 New Hampshire Ave., N.W.

$$o Washington, DC 20036 R. D. Martin, NRC Regional Administrator P. D. Milano, NRC Project Manager

/p' P. H. Harrell, NRC Senior Resident Inspector

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ATTACHMENT 1 SLIDES FROM JUNE 8, 1988 ENFORCEMENT CONFERENCE

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4 NRC MEETING June 8, 1988 Inspection Report 88-15 I.

OPENING REMARKS / INTRODUCTION (5 minutes)

K.'J. Morris II.

OVERVIEW 0F MEETING (5 minutes)

J. J. Fisicaro j

III.

POTENTIAL VIOLATION - FAILURE TO INSTALL A S. K. Gambhir COMPONENT THAT COM? LIES WITH ESTABLISHED DESIGN CRITERIA (285/8815-06) (25 Minutes)

Background, (System Configuration, Design Basis and Statement of Problem)

Statement of Problem Corrective Actions Taken/Results Achieved Future Planned Actions REVIEW 0F ACCUMULATOR TESTING Status of Inspection Report Items 1-10 IV.

POTENTIAL VIOLATION - CONTAINMENT INTEGRITY (285/8815-14) (25 Minutes)

T. L. Patterson Statement of Problem Corrective Action Taken/Results Achieved Future Corrective Action Planned Calculation of Radiation Levels (285/8815-13)

V.

MANAGEMENT INVESTIGATIVE SAFETY TEAM (10 minutes)

J. K. Gasper VI.

SAFETY SIGNIFICANCE (25 minutes)

T. L. Patterson Safety Analysis for Operability (SAO)

SA0 for Instrument Air System LOCA concurrent with CCW Line Rupture in containment (285/8815-07)

Hot-leg injection (285/8815-11)

A/HCV-742 (285/8815-15)

VII.

INSTRUMENT AIR STATUS (15 minutes)

R. L. Jaworski Update on Issue i

VIII. CLOSING COPMENTS (5 minutes)

K. J. Morris 4

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SIRWT LEVEL INSTRUMENTATION SCHEMATIC

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I CORRECI'lVE ACTIONS TAKBI AND RESULTS AQIIEVED

- REPLACDIEtTP OF GANE FIG #27 QiECK VALVES WUDI NEW IEAK TIGff (IUPRO B-8C-1) GiECK VALVES.

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- NEN VALVES INCIUDED IN THE ISI PROGRAM i

- INSTALIATION IS PROEOSED 'IO BE IURhfER MODIFIED 'IO FACIIITATE FUNCTIONAL TESTING (!R-FC-88-39)

- REVIEW OF OIHER ACCUtRA'IOR INSTALIATIONS I

4 (14 SIMIINI INSTALIATIONS IDBinntU - SEE TABIE - A)

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O TIBLE A FUNCTION OF CQE & NON-02E VALVES WITH AOCUNUIATORS WITH NON IEAK TIGIT QIECK VALVES VALVE VALVE FUNCTION FUNCTION OF ACGMHATOR CORRECTIVE ACTION IfV-383 1&2 SI WI HDR ISOL SAFEIY REIATED TH4IORARY IDDIFICATION IIAS BEEN Ct4PIETED.

illE VALVES API CDN-SIDERED OPERATIONAL.

HCV-2850 WP ISO VLVS IICV-2851 HCV-2852 SEE OSAR 87-10, P. 15 and 16 HCV-2874A W IIIR ISO VLVS ONLY REQUIRED IF IM LDIE EREAIG -

HCV-2875A CUISIDE F.C. DESIQI BASIS HCV-2876A HAND WFTTA ARE PROVIDED AND THE ALTERNATIVE HEAIG ARE HCV-2874B VALVES ARE ACCESSIBIE FOR IWUAL AVAIIABIE TO PEREDRM HCV-2875B OPERIiTION.

DESIGT BASIS IUTC-HCV-2876B TIONS. A IODIFICA-MS-291 AIR OPERATED SEE OSAR 87-10, P. 5 -

TION HAS uCEN MSS VALVES ACCUM. IS NOT REQUIRED FOR PERFORM-SGIEDULED.

MS-292

_ING ANY DESIQi BASIS SAFETi FUNCTION.

HCV-484 CCW EXIT VLV SDEyX SEE OSAR 87-10, P. 11 -

HCV-485

_ARE ACCESSIBLE FOR IRIUAL OPERATION 1

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PARTIAL SUIEARY OF OPHI/UIRESOLVED ITBd3 FICI DISPECTIOIT REIORT #88-15 SUr@ WRY

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OPDT ITH4 285/8815-01: SUH11T SCIEDUIE FOR FDIAL TESTDIG OF AIR AOCGULATORS.

OPEN ITB4 285/8815-02: SU&iIT NB1 REVISIQi OF IST PROGRAM 'IO DICUJDE ACCRUIA'IOR OIECK VALVES IDHTTIFIED DI OSAR 87-10.

OPDT ITHf 285/8815-03: 'IHIS ITB4 DEAIS WI'IH ' HIE SEISIIC QUALIFICATIOT 1

ISSUE REIATING 'IO AOCHUIAIOR/'IUBDIG A!!D CDNIROL VALVES.

OPHI ITH1 285/8515-04: SIIMT LEVEL 00tTIPOLIERS ISSUE FOH4AL CAILULATIGI TO VERIFY ACCacIATOPS ARE SIZED CORRECILY.

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SUFNARY OPSI ITai 285/8515-05: COMPIETE 12 HOUR PUNCTIONAL TESTDIG OF ACQDUIA'IORS ASSOCIATED WI'IH EURRTERS FOR SIIUr IEVEL INDICATIOli.

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OPDT ITDI 285/8815-06: USE OF NON IEAK TIQff GIECK VALVE - IUTE!EIAL VIOIATION FAIIURE 'IO MEET PIRTP DESIGI CRITERIA - SIIMT IIVEL 00t7fROLIERS.

OPEN ITEM 285/8815-07: HCV-438 B&D - EVAIDATE AND DEIERMDE IF OPPD MJSI' PERPORM AN ANAIESIS ASSUMING IOCA 00NQJRREt7f WI'IH CCW IJNE RUPIURE IN C0tTTAINiENT.

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SUIEARY OPDT ITB4 285/8815-08: HCV-238 & 239 - SDICE ACCUMUI1 DOR QECK VALVES ARE DISIDE BIO-SHTvin, IT IS lior APPARBTP HOW OPPD EtrBIDS 'IO TEST 'IIE QiECK VALVES QUARTERLY.

i OPEN ITEM 285/8815-09: HCV-238 & 239 - PREPARE SEISMIC CM.CUIATIOIIS.

OPEN ITEM 285/8815-10: HCV-240 - PREPARE SEIS4IC CAIGIATIONS.

OPEN ITEM 285/8815-06: IG-383-1 & 2 - ONLY ONE ACCOMUIATOR IS PROVIDED FOR TWO VALVES. 'IHIS DOES NOT MEET SHIGIE FAIIURE CRITERIA.

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OUTLINE OF CONTAINMENT INTEGRITY EVENT PRESENTATION e STATEMENT OF PROBLEM e

SUMMARY

OF TECIINICAL/ LICENSING COMMITMENTS NOT MET e STATEMENT OF OPERABILITY e SAFETY IMPLICATIONS e RESULTS OF HUMAN PERFORMANCE EVALUATION SYSTEM REVIEW e CORRECTIVE ACTIONS e CONCLUSIONS

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o SEOUENCE OF EVENTS l

3/27/87 - PERFORMED ST-CONT-3 TEST 1 3/28/87 - PERFORMED ST-CONT-3 TEST 2 5/4/87 - NOZZLE DAMS REMOVED, PARTS STORED IN RM 59 5/27/87 - PERFORMED PENETRATION VERIFICATION USING M.O.

1. 872525

~6/15/87 - NOZZLE DAMS REMOVED FROM RM 59 4/18/88 - DISCOVERED MISSING CAP PC-743 (~1515) 4/19/88 - I&C FOREMAN CONFIRMS CTMT LEAKAGE (~0815) 4/19/88 - CAP REPLACED AND LEAK CHECKED (~0945) 4/19/88 - MAINTENANCE SUPERVISOR CONVENES PRC (~1030) 4/19/88 - PRC DECLARES EVENT REPORTABLE, ASSIGNS FOLLOWUP ACTIONS TO INVESTIGATE OTIIER POTENTIAL PROBLEMS 4/20/88 - COMPLETE WALKDOWN OF ALL NON-WETTED PENETRATIONS 4/28/88 - MIST ACTIVATED 4/29/88 - SP-CONT-3 ISSUED 5/2/88 - SP-CONT-3 COMPLETED 5/12/88 - MIST REPORT ISSUED 5/26/88 - DRAFT HPES REPORT ISSUED 5/27/88 - CONSEQUENCES ANAINSIS COMPLETED BY CE

s CFC IRC CAPS REMOVED A/HCV-742 9

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FIGURE 2 j

TECHNICAL / LICENSE ISSUES POTENTIALLY VIOLATED e

ST-CONT-3 IS PERFORMED TO SATISFY TECHNICAL SPECIFICATION 3.5, PARAGRAPH (4), "CONTAINMENT ISOLATION VALVES LEAK RATE TESTS (TYPE C TESTS)".

e TECH SPEC 2.6, PARAGRAPH (1)a. STATES "CONTAINMENT INTEGRITY SHALL NOT BE VIOLATED UNLESS THE REACTOR IS IN THE COLD SHUTDON:I CONDITION".

e TECH SPEC 2.8, PARAGRAPH (1) REQUIRES "ALL AUTOMATIC ISOLATION VALVES SHALL BE OPERABLE OR AT LEAST ONE VALVE IN EACH LINE SHALL BE CLOSED" DURING ANY REFUELING OPERATIONS.

TECHNICAL SPECIFICATION 2.15, TABLES 2-2 AND 2-3, IDENTIFIES e

THE NUMBER OF M7.NIMUM OPERABLE CHANN %S FOR THE CONTAINMENT HIGH PRESSURE (CPHS) AS 2.

GDC-19 AND SRP 6.4 CRITERIA ESTABLISHES 5 REM WBE AND 30 REM e

THYROID / BETA SKIN DOSE LIMITS FOR CONTROL ROOM.

CONCLUSION 1

1 THE FORT CALHOUN STATION WAS IN VIOLATION OF SPECIFICATIONS 2.6, 2.8 AND 2.15.

9

SAFETY ANALYSIS FOR M-38 PENETRATION OPENING PERFORMED A CONSERVATIVE CONSEQUENCES ANALYSIS TO DEFINE e

EXPECTED IMPACT ON SAFETY AS A RESULT OF ADDED LEAKAGE FROM CONTAINMENT.

e PERFORMED INDEPENDENTLY VERIFIED CALCULATION OF EXPECTED LEAKAGE FROM TIIE OPEN CAP.

LEAKAGE WAS DETERMINED USING THE CONTAINMENT PRESSURE e

PROFILE USED FOR EEQ.

LEAKAGE FROM M-38 WAS ADDED TO ALLOWED TECHNICAL e

SPECIFICATION LEAKAGE FROM ALL OTIIER SOURCES TO DETERMINE TOTAL EXPECTED LEAKAGE FOR DESIGN BASIS LOCA.

USING TIIE CALCULATED TOTAL LEAKAGE, RADIOLOGICAL e

CONSEQUENCES WERE DETERMINED USING DESIGN BASIS SOURCE TERMS CONSISTENT WITH REG GUIDE 1.4 CRITERIA.

TOTAL DOSES USING TIIE ABOVE ASSUMPTIONS INDICATE TIIAT THE e

0-2 HOUR VALUE IS 6.4 REM WHOLE BODY AND 258.6 REM THYROID AT THE SITE BOUNDARY.

THESE VALUES ARE WITHIN THE 10 CFR 100 LIMITS OF 25 REM AND 300 REM, RESPECTIVELY.

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ROOM LAYOUT FIGURE 3 O

PC-743 k

AREA 3 one ac 1 OPEN CAP FOR TESTING 6 in OFF THE FLOOR BELOW INSTRUMENT 4-INSTRUMENTATION SUPPORT RACK O

A/PC-765 AREA 2 AREA 1

__O A/PC-744-1 A/ P C -744,-2 A/HCV-74 13 CAPSN V A-219X VA-228 It 3 OPEN CAPS VA-275

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HPES FINDINGS (1)

EVALUATED THREE SEPARATE INAPPROPRIATE ACTIONS:

TWO PAST EVENTS /ONE RECENT (2)

PRIMARY CAUSAL FACTOR FOR TWO ACTIONS WAS INADEQUATE WRITTEN INSTRUCTIONS.

PRIMARY CAUSAL FACTOR FOR THIRD CONCERN WAS COMMUNICATIONS DEFICIENCY (3)

SECONDARY CAUSAL FACTORS INCLUDED (TWO CONCERNS).

e INADEQUATE PRE-JOB BRIEFING l

e UNFAMILIARITY / LACK OF TRAINING ON SPECIFIC EVOLUTIONS IN PROGRESS e

INADEQUATE TRAINING ON WORK PRACTICES (SELF-CHECKING PROCESS) i ENVIRONMENTAL CONDITIONS (SLOPPY WORK AREA, WINTER o

STORM, AND STORAGE OF MATERIALS IN AREA) e INADEQUATE TRAINING ON SYSTEMS / SAFETY FUNCTIONS

CORRECTIVE ACTIONS ALREADY IN PROGRESS AT TIME OF INCIDENT e

ST-CONT-3 WAS UNDERGOING MAJOR UPGRADE TO ENSURE ALL CAPS /

COMPONENTS PROPERLY ADDRESSED e

INITIATED PROCEDURES UPGRADE PROGRAM FOR ALL OPERATING MANUAL PROCEDURES e

ADDED REQUIREMENT FOR PRE-JOB BRIEFING FOR ALL SURVEILLANCE TESTS TIIAT ARE PERFORMED ON AN INFREQUENT BASIS HELD SERIES OF MEETINGS WITII PLANT PERSONNEL TO STRESS e

IMPORTANCE OF VERBATIM COMPLIANCE AND ATTENTION TO DETAIL IMPLEMENTED ACCREDITED PERFORMANCE-BASED TRAINING FOR e

CRAFTSMEN AND OTHER TECHNICAL STAFF MEMBERS RESTRUCTURED MAINTENANCE DEPARTMENT TO ADD FIELD SUPERVISORS e

AND PLANNERS FOR EACII CRAFT INSTITUTED PRE-REVIEW REQUIREMENT ON INFREQUENT CALIBRATION e

PROCEDUR,ES DEVELOPING STRUCTURED TRAINING FOR PRC MEMBERS e

I

l ADDITIONAL MEASURES PLANNED AS A RESULT OF CONTAINMENT INTEGRITY e

A FORMAL GUIDANCE DOCUMENT FOR PERFORKING PRE-JOB BRIEFS IS TO BE DEVELOPED.

THE NEED TO BROADEN REQUIREMENTS FOR PRE-JOB BRIEFINGS BEYOND CURRENT REQUIREMENTS WILL DE EVALUATED l

WILL REVIEW THE PROCESS FOR CERTIFYING THE QUALIFICATION OF e

PERSONNEL PERFORMING SAFETY-SIGNIFICANT MAINTENANCE e

A FORMAL PROCEDURE FOR PRE-STARTUP VERIFICATION OF ALL CONTAINMENT PENETRATIONS WILL BE DEVELOPED EEAR ISSUED TO INSTALL CAGE IN ROOM 59 TO SEPARATE OUTAGE e

STAGING AREA FROM SAFETY-RELATED COMPONENTS IMPROVED PROCESS FOR POST SURVEILLANCE TEST REVIEW TO e

ENSURE FIRST LINE SUPERVISOR REVIEW AND VERIFICATION OCCURS WITHIN 24 HOURS OF COMPLETION FORMAL TRAINING FOR IMPROVED SAFETY AWARENESS WILL BE e

DEVELOPED FOR ALL FIRST LINE SUPERVISORS

t SAFETY ANALYSES FOR OPERATION AS APPLIED AT THE FORT CALHOUN STATION A SAFETY ANALYSIS FOR OPERATION IS REOUIRED e

WHEN AN EXISTING EVENT OR CONDITION IIAS BEEN IDENTIFIED THAT HAS PLACED TIIE PLANT OUTSIDE ITS DESIGN BASIS, i

THE EVENT OR CONDITION CANNOT BE IMMEDIATELY CORRECTED, AND e

THE EVENT MUST BE REPORTED PURSUANT TO TIIE REQUIREMENTS OF e

THE CODE OF FEDERAL REGULATIONS, CIIAPTER 10, PARTS 50.72 OR j

50.73.

BASIS i

THE SAFETY ANALYSIS FOR OPERATION IS REQUIRED BY 10 CFR 50.73(B), CONTENTS.

SPECIFICALLY, PARAGRAPH (3) REQUIRES "AN ASSESSMENT OF THE SAFETY CONSEQUENCES AND IMPLICATIONS OF THE EVENT.

THIS ASSESSMENT MUST INCLUDE THE AVAILABILITY OF OTHER SYSTEMS OR COMPONENTS THAT COULD HAVE PERFORMED TIIE SAME FUNCTION AS THE COMPONENTS AND SYSTEMS THAT FAILED DURING THE EVENT."

SAO PROCESS F L O L4 CHART 1 C 5

ACCEPT?

HO i!El3 8 223 9025 CONSEbUENCES.

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e SAO WAS FIRST INITIATED IN APRIL 1987.

CONSERVATIVE IN NATURE IN THAT AT TIIAT TIME, NO SPECIFIC DEFICIENCIES OTl!ER TIIAN IICV-385/386 IIAD BEEN IDENTIFIED.

e SAO ASSUMED THAT ALL SAFETY RELATED ACCUMULATORS AND THEIR ASSOCIATED CHECK VALVES FAILED.

USING TIIIS ASSUMPTION, OTIIER PLANT FEATURES WERE IDENTIFIED TO ASSESS SAFETY SIGNIFICANCE.

e APRIL 6, 1988 TIIE OSAR FOR ACCUMULATORS WAS COMPLETED.

BASED ON TIIE FINDINGS FROM TIIIS EVALUATION A ONE IIOUR REPORT WAS MADE TO TIIE NRC ON LCV-383-1/2, IICV-238,239 AND i

240, AND HCV-438B/D.

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MID-APRIL 1988, DECISION WAS MADE TO COMPLETE TESTING OF THE ACCUMULATOR FOR THE CCW VALVES TO CONTAINMENT COOLERS AND TO TEST THE SIRWT BUBBLER CHECK VALVES.

TESTING IS STILL REQUIRED FOtt IICV-385/386, SIRWT BUBBLERS, e

AND liCV-2987.

MODIFICATIONS OR OTHER CORRECTIVE MEASURES ARE NEEDED FOR REPORTED DEFICIENCIES.

FOR HCV 438B/D AND IICV-400 SERIES, EITl!ER MODIFICATION e

REQUIRED OR DESIGN BASIS MUST BE REVISED AND SUPPORTED.

APPEARS THAT MR IS MOST LIKELY FIX FOR 438 B/D.

e HCV-304, 305, 306 AND 307 NEED A CONFIRMING EVALUATION TO VERIFY THE ADEQUACY OF HOT LEG INJECTION FLOW WITH VALVES IN CURRENT CONFIGURATION.

t EVALUATION OF LCV-383-1/2 SINGLE FAILURE OUALIFICATION e

THE FORT CALIIOUN STATION UPDATED SAFETY ANALYSIS REPORT (USAR) SECTION 6.2.5, DESIGN EVALUATION, STATES IN PARAGRAPH h, "THE SAFETY INJECTION SYSTEM HAS BEEN DESIGNED TO MEET TIIE 1

SINGLE FAILURE CRITERION."

I THE USAR, APPENDIX G, CRITERION 21, SINGLE FAILURE e

DEPINITION, STATES "THE DESIGN OF THE FORT CALHOUN STATION IS BASED ON THE CONCEPT THAT NO SINGLE FAILURE OF ACTIVE COMPONENTS WILL INHIBIT NECESSARY SAFETY ACTION WHEN REQUIRED."

THE CODE OF FEDERAL REGULATIONS, CHAPTER 10, PART 50, e

APPENDIX A, CRITERION 21, PROTECTION SYSTEM RELIABILITY AND TESTABILITY, REQUIRES THAT "REDUNDANCY AND INDEPENDENCE DESIGNED INTO THE PROTECTION SYSTEM SHALL BE SUFFICIENT TO ENSURE THAT (1) NO SINGLE FAILURE RESULTS IN THE LOSS OF PROTECTION FUNCTION".

LCV-383-1 AND LCV-383-2 ARE NORMALLY OPEN, FAIL OPEN AIR e

OPERATED VALVES THAT MUST BE OPEN TO PROVIDE A BORATED WATER SOURCE FOR THE SAFETY INJECTION AND CONTAINMENT SPRAY PUMPS EARLY IN A DESIGN BASIS LOCA.

ONCE LEVEL IN THE SAFETY INJECTION REFUELING WATER TANK (SIRWT) REACHES 16" LEVEL, TH2 VALVES RECEIVE A RECIRCULATION ACTUATION SYSTEM (RAS) SIGNAL TO CLOSE.

THE VALVES ARE CLOSED TO PREVENT BACKFLOW OF ACTIVATED COOLANT FROM THE CONTAINMENT FLOOR FROM ENTERING THE SIRWT.

LCV-383-1 AND LCV-383-2 RELY ON ONE AIR ACCUMULATOR TO CLOSE e

WITH LOSS OF INSTRUMENT AIR TO PERFORM THE RAS FUNCTION.

THE TWO VALVES HAVE THEIR INDEPENDENT SOLENOID VALVES DOWNSTREAM OF THE ACCUMULATOR AND ITS ASSOCIATED CHECK VALVE.

THERE IS NO SINGLE FAILURE OF THE ACCUMULATOR OR CHECK VALVE, ACTIVE i

OR PASSIVE, THAT WOULD PREVENT THE VALVES FROM GOING TO THEIR FAILED POSITION.

PASSIVE FAILURE OF THE ACCUMULATOR AND ASSOCIATED TUBING OR AN ACTIVE FAILURE OF THE CHECK VALVE COULD PREVENT BOTH VALVES FROM BEING CLOSED.

AN ADDITIONAL SINGLE FAILURE OF ONE OF THE CHECK VALVES, e

SI-140 OR SI-139, WOULD HAVE TO OCCUR BEFORE THE SAFETY FUNCTION (I.E., PREVENTING BACKFLOW) IS INHIBITED.

LCV-383-1 AND LCV-383-2 ARE IN PARALLEL.

THIS DESIGN o

ARRANGEMENT ENSURES THAT WITH ANY SINGLE FAILURE, AT LEAST ONE PATH OF FLOW FROM THE SIRWT TO THE SI/CS PUMPS WOULD BE AVAILABLE.

ALSO, SINCE THE VALVES ARE IN PARALLEL, ONE VALVE FAILING TO CLOSE OR BOTH VALVES FAILING "'O CLOSE ON RAS, WOULD HAVE THE SAME EFFECT ON THPTR ABILITY TO 'ERFORM THF.IR SECONDARY SAFETY FUNCTION, THUS, CHECK VALVES SI-139 AND SI-140 ARE REQUIRED TO MEET SINGLE FAILUh* CD1TER.L\\ FOR PREVENTING BACKFLOW.

CONCLUSION TEiE FACT THAT LCV-383-1.AND LCV-383-2 SHARE AN ACCUMULATOR DOES NOT CONSTITUTE A VIOLATION OF FORT CALHOUN STATION'S OR CURRENT NRC SINGLE FAILURE CRITERIA.

Safety injection and Refueling Water Tank Air Supply i

LCV-383-2M FOM FO L C V-3 8 3-1 9

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SPRAY PUMPS 4

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MANAGEMENT INVESTIGATIVE SAFETY TEAM (MIST)'

PURPOSE Collect and preserve event information Establish event documentation files Analyze event significance Identify root cause(s)

Recommend and review corrective action PERMANENT MEMBERSHIP J. K. Gasper (Designated Team Leader)

R. L. Jaworski S. K. Gambhir METHODOLOGY Identify additional team members Review available documents and data Interview personnel involved Document reviews and interviews ACTIVATED April 28, 1988 by R. L. Andrews SCOPE SIRWT Bubbler Check Valve Failures Missing Cap On Containment Pressure Sensing Line REPORT TO DIVISION MANAGER, MANAGER - FORT CALHOUN, AND SARC MEMBERS REQUESTED BY MAY 12, 1988 REPORTS ISSUED MAY 12, 1988 1

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es SIRWT BUBBLER CHECK VALVE FAILURE ADDITIONAL MEMBERS

-R. C. Kellogg W. O. Weber COLLECT AND PRESERVE INFORMATION "As found" condition of check valves destroyed after testing due.to way M.0. w., written EVENT DOCUMENTATION INCLUDED WITH REPORT EVENT SIGNIFICANCE During a LOCA with loss of offsite power, instrument air would be lost and failure of the check valves could cause the bubblers to fail causing an inadvertent RAS.

This "false" RAS could cause SI pumps failure and consequential core damage.

ROOT CAUSES Inadequate scope of ST and IST programs

• Inadequate CQE list

• No systematic assessment to assure completeness Misapplication of Crane, Model No. 27, horizontal lift check valve CORRECTIVE ACTION RECOMMENDATIONS Limit "testing" procedures to testing and document as found condition Replace similar Crane check valves in other air accumulator applications.

Perform fault tree analysis on Safety Injection (SI) System to determine adequacy of CQE list, and ST and IST programs for the SI System.

Expand to other systems if discrepanc cs identified.

i Place air accumulator check valves in IST program 2

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• e MISSING CAP ON CONTAINMENT PRESSURE SAFETY LINE ADDITIONAL MEMBERS M. W. Butt (HPES Coordinator)

J. D. Kecy INFORMATION AND DOCUMENTATION INCLUDED IN REPORT EVENT SIGNIFICANCE Loss of containment integrity Greater than Tech Spec leak rate

- Post accident doses less than 10 CFR 100 limits

• Control Room habitability limits probably exceeded *

• Design Basis calculations using R.G. 1.4 assumptions ROOT CAUSE(S)

Inadequate procedure and work package (primary)

Lack of system familiarity (secondary)

IMMEDIATE CORRECTIVE ACT10N April 29, 1988 letter Develop and execute a procedure to ac omplish a complete and documented dual verification of all non-wetted penetration "line ups".

Completed May 2, 1988 MAJOR CORRECTIVE ACTION RECOMMENDATIONS (Includes HPES Evaluatior Recommendations)

Development and implemt.t procedure (s) to double verify ail penetration "line ups" Ensure STs and cps return instruments to proper configuration Evaluate acceptability of crimped and soldered line seals 3

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MISSINGCAPONCONTAINMENTPRESSURESAFETYLINE(continued)

Evaluate method tc reduce human error potential of these and similar penetrations Qualify ^ method of installing Swagelock caps to serve as 60 psi boundary Eliminate test tee caps as single boundary by installing an isolation valve as a second isolation boundary Training on safety significance of events and decisions Clearly define person serving as Acting Manager - Fort Calhoun Station Separate temporary equipment from permanent equipment in Room 59 (HPES)

Training on planning and organizing maintenance (HPES)

Centralize scheduling and implement ILS O

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MIST STATUS t

ADDENDUM TO BE ISSUED THIS WEEK IN RESPONSE TO QUESTIONS LETTER REQUESTING COMMENTS AND QUESTIONS BY JULY 1, 1988 TO BE SENT T0' ALL RECIPIENTS THIS WEEK ASSIGNMENT OF RECOMMENDATIONS TO BE MADE AND TRACKED REPORT TO BE PROVIDED TO NEXT SARC MEETING ON JULY 13, 1988 D

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M SLIDE 1 INSTRUMENT AIR SYSTEM STATUS REPORT O

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ITEMS COMPLETE SLIDE 2 A.

INSTRUMENT AIR SYSTEM SPECIFIC 1.

Operational a.

Assignment of System Engineer 2.

Procedural a.

Procedure Change to MP-FP-7 to Ensure Check Valve Operability b.

Review of IAS Operating Procedures c.

Upgrade Incident Reporting Process d.

Development of Operating Instructions for High Dew Points e.

Instrument Air System Line-up/ Tag Out 3.

Maintenance a.

Preventive Maintenance Program Upgrade b.

Air Dryer Desiccant Replacement c.

Revise Air Dryer Preventive Maintenance Schedule 4.

Surveillance a.

Preventive Maintenance / Surveillance Testing of Air Operated Dampers b.

Particulate Monitoring Program c.

Dew Point Sampling Program 5.

Evaluations a.

System Functional Inspection b.

Evaluation of Interface Valves and Bubblers 4

c.

Consideration of Check Valve Failures in the Safety Analysis fcr Operability 6.

Engineering a.

Common Mode Failure Consideration B.

MANAGERIAL IMPROVEMENTS 1.

Evaluations a.

SARC Evaluation of the Water Intrusion Event b.

Event Investigative Team 2.

Reporting a.

Emergency Plan Implementing Procedure (EPIP) Upgrade b.

Safety Analysis for Operability (SAO) 3.

Personnel a.

Employee / Management Conferences 4.

Procedures a.

Policy Statement Concerning Equipment Operation by Non Operators

II.

ITEMS IN PROGRESS SLIDE 3 A.

INSTRUMENT AIR SYSTEM SPECIFIC

1. Operational a.

Additional Walkdowns ei Instrument Air System 2.

Procedures a.

Abnormal Operating Procedure (A0P-17)

Upgrade in Accordance with Writers Guide 3.

Maintenance a.

Clean / Flush the Instrument Air System b.

Valve Teardown Program 4

5urveillance a.

Testing of CQE Non-ISI Valves b.

Monthly Valve Cycling for Inservice Test Program c.

Cycling of Inservice Inspection (ISI) Valves d.

Inservice Inspection Program 5.

Engineering a.

Diesel Generator Air Damper Replacement b.

Assessment of Instrument Air System Design to Current Industry Standards c.

System Functional Inspection - Follow-up d.

Development of Instrument Air System Design Basis Documentation i

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Identification and Installation of Air Filter Additions for Valve Operators and Air Dryer f.

Removal of Plant Air System and Fire Protection System Crosstie B.

MANAGERIAL IMPROVEMENTS 1.

Evaluations a.

Nuclear Operations Plan Development and Appraisal 1

b.

Instrument Air Steering Committee 2.

Training a.

General Employee Training (GET) Upgrade for Safety Awareness b.

Additional Field Observation Training c.

Upgrade of Lesson Plans d.

Pre-Job Briefings for Surveillance Tests

SLIDE 4 INSTRUMENT AIR SYSTEM CORRECTIVE ACTION ITEMS 50

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I44 TOTAL COMMITTED g

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ITEMS COMPLETED e

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10 NUMBER OF CORRECTIVE ACTION ITEMS TOTAL / COMPLETED 7

ys COMPLETION TIME JUNE 1,

1988 0

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DEC J AN FEB MAR APR MAY JUN JUL AUG SEP OCT NOV DEC

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1987 1988 COMPLETION TIME

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SLIDE 5 REVIEW 0F INSTRUMENT AIR PROJECT ACTIVITIES RELATED TO SIRWT BUBBLER CHECK VALVE AND CONTAINMENT INTEGRITY EVENTS o

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,. a SLIDE'6 CONCLUSIONS OF REVIEW:

1.

The two events were not the result of ineffective lor incomplete Instrument Air actions 2.

Remaining actions improve reliability and operability of the Instrument Air System 3.-

Completed Instrument Air System action have improved OPPO responsiveness to events o

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SLIDE 7 4.

Questions arise on broad issues relating the two recent events to the instrument air event:

a.

Is the scope of the Surveillance Test Program and CQE list i

properly defined? [0 pen - Action in Progress) b.

Are personnel who are performing Surveillance Tests properly-trained and briefed? [ Closed - Action Initiated and Complete]

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ATTACHMENT 2 This attachment provides detailed information regarding several questions raised during the enforcement conference.

1.

During the discussion, the topic of a Component Cooling Water line rupture inside containment concurrent with a Loss of Coolant Accident was raised.

The question raised was:

If such an event were to occur, would the amount of componert cooling water dumped in containment be sufficient to dilute critical?

As discussed at the June 8,1988 conference, this would require two concurrent accidents, which is outside the bounds of what is considered in an accident analysis.

This event was discussed in general terms.

As noted at the conference, sufficient time would exist for operator action to address the problem before any criticality would occur.

It was also noted that the operator would detect a rise in containment level with the Recirculation Actuation Signal activated.

Boric acid injection could still be accomplished via the charging pumps. A formal calculation to document that a return to critical would not occur is currently underway.

This calculation will be completed by July 31, 1988.

2.

As a result of the enforcement conference, the questic. was raised as to how Fort Calhoun Station met General Design Criterion 19 during the time the cap was off.

OPPD held follow-up conversations with the Senior Resident Inspector on June 10 related to the topic of doses in the Control Room Post-LOCA with the instrument cap missing. These conversations were held to ensure there was an understanding of where Fort Calhoun Station stood relative to General Design Criterion 19 during the time that the cap on the 3/8" con-tainment penetration line was not installed.

The purpose of our discussion relative to the General Design Criteria was to assess the "safety signifi-cance" of the condition if a design basis accident would have occurred during this time frame. As noted, there was never an intent to provide results by calculation which demonstrated that we could license the facility in the as-found condition.

However, to assess the plant's interim position from a licensing perspective, the General Design Criteria provide a basis from which to perform that assessment.

Subsequently, to clarify the results of comparing the radiological consequences between postulated and realistic conditions, OPPD met with the Senior Resident Inspector to discuss the conservatisms in the analysis which was presented.

The following information was discussed with the Senior Resident Inspector.

The NRC-required conservatisms in the analysis are compared to OPPD's best estitnate of a "real life" situation.

a.

The source term used in the analysis was developed based upon the assumptions presented in Revision 2 of Regulatory Guide 1.4, dated June 1974, which has been recognized as unrealistically conservative in the years since the Three Mile Island Unit 2 accident in March 1979.

In reality, the source term research conducted since the TMI 2 accident has shown that much smaller fractions of the core fission product inventory are actually available for release to the environ-ment (Reference NUREG-0956, dated July 1986).

In addition, OPPD has

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, s' continued estimated a maximum source term of 5% core inventory Xenon isotopes, 15% core inventory Krypton isotopes, and 1.25% core inventory Iodine isotopes would be available for release to the environment.

This is based upon cladding failure only, with no fuel celting, pursuant to Fort Calhoun Station USAR Section 14.15.

Resultant doses from a more realistic source term would be less by a factor of 10 for whole body and more for thyroid dose, b.

No mixing was assumed in the auxiliary building atmosphere, whereas mixing would take place.

In addition, operator action mandated by Fort Calhoun Station Abnormal Operating Procedures (A0P-9) would isolate the auxiliary building ventilation system and trap released fission products inside the building.

c.

A direct release to the environment from the point of containment closest to the control room ventilation intake was assumed, whereas the actual release would be to the auxiliary building which would be isolated as noted in b. above.

d.

Atmospheric mixing due to only the building wake (turbulence) was allowed, whereas additional mixing would occur due to the ventilation exhaust stack height of 51 ft. above the control room ventilation intake. Approximately 84% of any environmental release would be from the ventilation exhaust stack.

e.

A 1 meter per second wind speed was utilized to ct.lculate the dispersion factor, X/Q, whereas annual average wind speeds at the Fort Calhoun Station are on the order of 1.8 to 3.6 meters per second.

Higher wind speeds result in better dispersion of released gases.

Resultant dose reductions would be on the order of 44%-72% due to better dispersion of gases after release from the auxiliary building.

f.

Stay time of operators in the control room for the design basis calculation were assumed as follows:

100% for 0-24 hours, 60% for 1-4 days, and 40% for 4-30 days, whereas actaal stay times would be a maximum of 33% over the 30 day period if an operator worked every day for 30 days.

g.

10 CFR 20 allows up to 3 Rem / quarter, 12 Rem / year, if not greater than the PAD, whereas OPPD administratively limits worker doses to 4.5 rem per year.

In addition, OPPD's Radiological Emergency Response Plan places a Health Physics technician in the control room to monitor dose rates in the event of an accident which would ensure the OPPD administrative limit is not exceeded.

h.

Fission product leakage from containment was assumed constant at 0.05%

of the containment free volume (1985 ILRT results) for the period 0-24 hours and 0.025% for the remaining 29 days in addition to the conser-vatively calculated pressure driven release through the uncapped in-strument line. The pressure driven release through the uncapped line was based upon the containment pressure transient developed in 1984 for EEQ.

In reality, actual release volumes would be less due to a lower, shorter duration pressure transient inside containment

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,e continued with containment pressure suppression systems operating as designed (i.e., two spray headers and four air coolers operating within 20-30 seconds.)

Based on this discussion, a better understanding of how Fort Calhoun assessed the implicationr. of the event was reached.

3.

The use and application of the Safety Assessment for Operability document was discussed.

OPPD will arrange a meeting with Region IV to explain the process and provide the NRC with OPPD's perspective on when and how we determine inoperability. We will contact you to arrange for this meeting.

4.

OPPD is evaluating whether the SIRWT Bubbler functional testing can be performed before the 1988 refueling outage.

If operationally feasible, the testing of the valves will be performed prior to the outage.

The bubblers are considered vperable based upon the testing performed on the check valves in May and verification that the accumulators are ad,equately sized.

5.

A review to determine locations of check valves installed in applications similar to that of the SIRWT bubbler check valves was performed.

Thirteen (13) such valves were identified, and are installed in non-safety-related applications. As noted at the conference, these valves will be replaced.

The schedule for replacement will be determined by July 31, 1988.

6.

LCV-383-1 and LCV-383-2 were included in the original ISI program. This testing did not include the instrument air accumulators and associated check valves.

Testing of LCV-383-1 and LCV-383-2 IA check valves is included in the proposed revisions to the ISI program.

This testing and other updates to the ISI program, including changes to address other concerns from Inspectior. Report 50-285/88-15 will be submitted prior to the end of July 1988.

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