05000271/LER-1993-010

LER 93-010-00:on 930803,fire Protection Valve V76-420 Has Not Been Cycled Through Complete Cycle of Full Travel,Per TS 4.13.B.d.Caused by Human Error.Applicable Surveillance Procedure Revised to Include V76-420.W/930831 Ltr

ML20056H057
Person / Time
Site:	Vermont Yankee
Issue date:	08/31/1993
From:	Wanczyk R VERMONT YANKEE NUCLEAR POWER CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-93-010, LER-93-10, NUDOCS 9309080149
Download: ML20056H057 (4)

LER-2093-010,

Event date:
Report date:
2712093010R00 - NRC Website

text

{{#Wiki_filter:i VEllMONT YANKEE NUCLEAR POWER CORPORA 1 ION f . (~p*"<]\~~3} P O. Box 157. Governor Hunt Road

   !]f ' '            Ver non, Vermont 05354-0157 (602) 257-7711 A ~\-

N' .%c: ^- ! . )\ t/ August 31, 1993 U.S. Nuclear Regulatory Commission Document Control Desk Washington, D.C. 20555 ,

REFERENCES:

Operating License DPR-28  ! Docket No. 50-271 , Reportable Occurence No. LER 93-10  ; I Dear Sirst j As defined by 10 CFR 50.73, we are reporting the attached Reportable  ; Occurrence as LER 93-10.  ; i Very truly yours, j VERMONT YANKEE NUCLEAR POWER CORPORATION Robert J. Wanczyk  ; Plant Manager j t cct Regional Administrator USNRC Region I , 475 Allendale Road King of Prussia, PA 19406 ,. l

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             -                                                                                                                                                                                                l hRC form 366 U.S. CUCLEAR REGULATORY COMMISSION                                                               APPRotfED OMS CO. 3150-0104 (6-87)                                                                                                                 EXPIRES 4/30/92 ESTICATED BURDEN PER RESPONSE TO COMPLY WITH TGIS INFORMATION COLLECTION EEOUEST: 50.0 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS AND REPORTS MANAGEMENT $ RANCH (P-350), U.S. NUCLEAR REGULATORY COEMIS$10N, WASHINGTON DC 20555, AND 10 THE PAPERWORK REDUCTION PROJECT (3160-0104), OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20603.

FACILITY KAME (1) DOCKET Wo. (2) PAGE (3) TITLE (4) VERMONT YANKEE NUCLEAR POWER STATION Ol5l0l0l0l2l7l1 0 1 OF 0 3 FAILURE TO PERFORM ANNUAL VALVE CYCLING AS REQUIRED BY PLANT TECHNICAL SPECIFICATION 4.13.B.1.d FOR A VALVE IN THE VITAL FIRE SUPPRESSION WATER SYSTEM FLOW PATH DUE TO PROCEDURAL OMISSION. EVENT DATE (5) LER NUMBER (6) REPORT DATE (7) OTHER F ACILli!ES INVOLVED (8) MONTH DAY YEAR YEAR SEQ # REV # MONTH DAY YEAR FACILITY NAMES DOCKET NO. (S) N/A 0 5 0 0 0 0 8 0 3 9 3 9 3 - 0 1 0 - 0 0 0 8 3 9 3 1 N/A S 0 0 0 {U OFERATING TH15 REPORT IS SUEMITTED PURSUANT TO RE0'MTS OF 10 CFR i: CHECK ONE OR MORE (11) MODE (9) N 20.402(b) 20.405(c) 50.73(a)(2)(iv) 73.71(b) POWER 20.405(e)(1)(1) 50.36(c)(1) 50.73(a)(2)(v) 73.71(c) LEVEL (10) 0l9l5 20.405(a)(1)(ii) 50.36(c)(2) 50.73(a)(2)(vil) OTHER:

                    . . .... .........                           20.405(a)(1)(iii)                 x 50.73(a)(2)(i)s                    50.73(a)(2)(vi ll )( A)
                    ....... ..........                           20.405(a)(1)(iv)                    50.73(a)(2)(ii)                    50.73(a)(2)(viii)(B)
                    .. ......... ....                            20.405(a)(1)(v)                     50.73(a)(2)(lit)                   50.73(a)(2)(x)

LICENSEE CONTACT FOR THIS LER (12) NAME TELEPHONE NO. AREA CODE ROBERT J. WANCZYK, PL ANT MANAcER 8l0l2 2l5l7l-{7l7l1l1 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THis REPORT (13) CAUSE SYST COMPONENT NFR REPORTABLE CAUSE SYST COMPOWENT MFR REPORTABLE TO NPRDS .... TO NPRDS

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l I II i i i l i I I Il -- SUPPLEMENTAL REPORT EXPECTED (14) EXPECTED M0 DAY YR SUBMIS$10N DATE (15) YES (If yes, conplete EXPECTED SUBNIS$10N DATE) X N0 N/A f l lll ABSTRACT On 08-03 93, with the plant operating at 95% power, during a review of fire protection system valves (Ells Identifier = KP) that are cycled annually to meet the requirements of Technical Specificaton 4.13.B.1.d. it was determined that valve V76 420 had not been cycled annually since its installation in 1981. The valve is normally scaled open and provides manual isolation for three Turbine Building hose stations. The root cause was determined to be human error in the f ailure to include the valve in the applicable surveillance procedure. Immediate corrective actions included: revising the applicable surveillance procedure to include V76-420: successfully cyc,ing the valve ton 08-06-93); and, initiating a review of the vital fire suppression water system and surveillance procedures to determine whether there were any other valves not being cycled as required by plant Technical Specifications. This review has not identified any additional valves that require cycling. One simitt nt was reported to the Commission within the last five years (LER #90-02). NBC Form 366 (6-89) (. . . . . . . .

NRC Foro 366A U.S. NUCLEAR REGULATC27 COMMIS$1CJ APPROVED OMS No. 3150-0104 (6-69) EXPIRES 4/30/02 ESTIMATED BURDEN PER RESPONSE TO CoMPLV WITH THIS INFoRMATION COLLECTION REcuEST: 50.0 HRS. FORWARD LICENSEE EVENT REPORT (LER) COMMENTS REGARDING BURDEN EST1 HATE TO THE RECORDS AND TEXT CONTINUATION REPORTS MAkAGEMENT BRANCH (P-350), U.S. NUCLEAR REGULATORY COMMISSloN, WASHINGTON DC 20555, AND TO THE PAPERWORK REDUCTION PROJECT (3160-0104), OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20603. i FACILITY NAME (1) DOCKET WO (2) LER WUMsER (6) PAGE (3) YEAR SEO # REV # VERMONT YANKEE NUCLEAR POWER CORPORATION 9 3 - 0 0 - 0 0 0 0 0l5l0l0l0l2l7l1 1 l2 OF l3 TEXT (if more space is required, use additionai NRC Form 366A) (17) . DESCRIPTION OF EVENT On 08-03-93, with t..e plant operating at 95% power, an evaluation of the valves that need to be cycled per Technical Sper.ifi-cation 4.13.B.1.d was in progress, in response to a ques-ion raised by American Nuclear insurers (ANI). Tech Spec 4.13.B.1.b requires that "The Vital Fire Suppression Water System shall be demonstrated operable at least once per twelve months by cycling each testable valve in the flor path through at least one complete cycle of full travel". The evaluation verified that fire protection valve V76-420 (Ells identi' er KP) is in the flow path from the vital fire suppression water loop to hose stations governed by plant Technical Specific 6tions, and that the valve is not included in the procedure controlling vital fire suppression water system valve cycling, OP 41f 5, " Fire Protection Systems Surveillance". The plant Fire Protection Coordinator was informed of these findings, who determined that the failure to cycle this valve is not consistent with Technical Specification 4.13.B.1.d. Investigation determined that the valve is normally sealed in an open position which is verified monthly as required by Technical Specification 4.13.B.1.b. Thus, an operable flow path to the hose stations was confirmed and no immediate compensatory actions were required. Procedure OP 4105 was revised to include V76-420 and the valve was successfully cycled on 08-06-93. An evaluation was initiated to determine whether other valves were not being cycled as required by Tech Specs. It was determined that this was an isolated incident. CAUSE OF EVENT { V76-420 was installed under Plant Design Change Request (PDCR) 8102 when a portion of the fire water system outer loop was relocated to accomodate a building addition. Previously, the three Turbine Bldg. hose stations were supplied by a single line from a header connected directly to the outer fire water loop with no provision for header isolation. The PDCR appropriately concluded that the affected fire water supply components provide protection for safety-related equipment. The appropriate documents requiring revision due to the PDCR were identified, including the procedures for annual valve cycling (OP 4020) and monthly position verification (OP 2186). The check-off sheet which verified document update was properly signed. The required changes for V76-420 were effected in OP 2186 for valve position check but were not incorporated in OP 4020 for cycling. Interviews with available individuals have not resulted in a determination of root cause. i in summary, the design modification and review process was adequate at the time of the PDCR to identify applicable fire protection requirements. The design package was reviewed by appropriate personnel and the applicable surveillance procedures were identified for revision. PDCR documentation verified that required procedural changes were initiated. However, the valve

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was not added to the procedure as required. j The apparent cause for this incident is, therefore, determined to be human error in the f ailure to include the valve in the appropriate surveillance procedure to ensure compliance with the Technical Specification cycling requirement when the valve was added to the system in 1981. Additionally, a system review for similar discrepancies was performed in 1990 as corrective action for the missed monthly valve position check and annual valve cycling for valve FP-6, reported under LER #90-02 No other discrepancies were identified during that review. The root cause for the f ailure to identify valve V76-420 at that time could not be determined. However, a potential contributing cause is a drawing error in which a boundary identification symbol differentiating vital fire system components from non-vital system components is misplaced, erroneously indicating that the three Turbine Building hose stations are all non-vital components. Thus, V76-420 may have been viewed as a boundary valve not in the operable flow path of the vital fire suppression water system. NRC Form 366 i (6 89) '

       'NEC Eorm 366A U.S. NUCLEAR REOJLAVC2Y COMMIS$10N                                  APPROVED OMS CO. 3150-0104 (6-09)                                                                                 EXPIRES 4/30/92 EST! MATED BURDE0 PER RESPOWsE 70 COMPL7 UITH VQIS INFORMATION COLLECTION REQUEST: 50.0 HRS. FORWARD LICEWSEE EVENT REPORT (LER)                            COMMENTS REGARDING BURDEN ESilMATE TO THE RECORDS AND TEXT CONTINUATION                                   REPORTS MANAGEMENT BRANCH (P-350), U.S. NUCLEAR REGULATORY COMMIS$10N, WASHlWGTON DC 20555, AND TO THE PAPERWORK REDUCTION PROJECT (3160-0104), OFFICE OF MAhAGEMENT AND BUDGET, WASHINGTON, DC 20603.

FACILITY KAME (1) DOCKET NO (2) LER Ni.MBER (6) PAGE (3) TEAR SEQ # REV # 9 or VERMONT YANKEE NUCLEAR POWER CORPORATION 3 0 0 0 0 0 l3 0 l3 1 0lSl0l0l0l2l7l1 TEXT (if more space is required, use additional NRC Form 366A) (17) ANALYSIS OF EVENT Technical Specification 3.13.B.1.c requires "an operable flow path capable of taking suction from the Connecticut River and transferring the water through the distribution piping with operable sectionalizing control or isolation valves to the yard hydrant curb valves and the hose station isolation valves" Technical Specification 4.13.B.1.d requires that "the Vital Fire Suppression Water System shall be demonstrated operable at least once per twelve months by cycling each testable valve in the flow path through at least one complete cycle of full travel.' Contrary to the requirement of Technical Specification 4.13.B.1.d, fire protection valve V76-420, which is in the flow path to three Turbine Building hose stations, has not been cycled through at least one complete cycle of full travel at least once per twelve months since it was installed. V76-420 has been inspected monthly to verify its sealed-open position since it was installed. Consequently, an operable flow path to the concerned hose stations has existed, and these stations have been inspected and tested in accordance with Technical Specification 4.13.C.1.a-e. Therefore, the fire system design function for the affected areas has been operable and the intended protection provided. Accordingly, based upon the established operable flow path to the hose stations, operable hose stations, and the additional plant fire protection features, this incident did not degrade plant vital fire system operability or pose a significant threat to the health and safety of the public. CORRECTIVE ACTIONS

1. The surveillance procedure for valve cycling (OP 4105, which supercedes OP 4020) was revised and valve V76-420 was successfully cycled.

2. A review of plant drawings will be performed to verify the Vital Fire Suppression Water System boundaries. This will be i accomplished by October,1993. Corrective updates to the drawings will then be initiated as necessary based upon the  :

results of the review.  ;

3. A review of all fire protection valves subject to the requirements of Technical Specification 4.13.B.1.d will be completed by October,1993.

ADDITIONAL INFORM ATION A similar event occuring within the last five years was reported to the Commission as LER #90-02. N3C Form 366 (6-89) 1 1 }}