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#REDIRECT [[05000278/LER-1990-004-03, :on 900430,Tech Spec Violation Occurred When MSIV Closure Timing Testing Not Performed in Required Surveillance Interval.Caused by Ambiguous Test Procedure. Surveillance Test Revised]]
| number = ML20043C578
| issue date = 05/30/1990
| title = LER 90-004-00:on 900430,Tech Spec Violation Occurred When MSIV Closure Timing Testing Not Performed in Required Surveillance Interval.Caused by Ambiguous Test Procedure. Surveillance Test revised.W/900530 Ltr
| author name = Fray J, Fulvio A
| author affiliation = PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
| addressee name =
| addressee affiliation = NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
| docket = 05000278
| license number =
| contact person =
| document report number = LER-90-004-03, LER-90-4-3, NUDOCS 9006050348
| document type = LICENSEE EVENT REPORT (SEE ALSO AO,RO), TEXT-SAFETY REPORT
| page count = 4
}}
 
=Text=
{{#Wiki_filter:.
CCN-90-14108-
                                            . PHILADELPHIA ELECTRIC COMPANY PEACll BOTIDM ATOMIC POWER STATION R. D 1, Box 208
,                                                        Delta, Itnnsyh ania !?314 esua mornw-nas rowsm or xcrit Nc                        (717) 456-7014 l
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May 30, 1990 Docket No. 50-278 Document Control Desk U. S._ Nuclear Regulatory Commission Washington, DC 20555
 
==SUBJECT:==
Licensee Event Report Peach Bottom Atomic Power Station - Unit 3                )
This LER concerns the untimely performance of a-Technical Specification    f Surveillance.                                                                  j
                                                                                                  \
 
==Reference:==
Docket No. 50-278 Report Number:            3-90-004                                          j Revision Number:          00                                                l Event Date:-              04/30/90                                        1 Report Date:              05/30/90 Facility:                Peach Bottom Atomic Power Station RD 1, Box 208, Delta, PA 17314 l
This LER is being submitted pursuant to the requirements of 10 CFR        ]
50.73(a)(2)(1)(B).                                                                !
i Sincerely,          l t
lant Manager      ,
cc:    J. J. Lyash, USNRC Senior Re-ident Inspector T. T. Martin, USNRC, Region I                                              1 l
i 9006050349 900530                                                      l PDR    ADOCK 05000278                                            \ \
S                      PDC                                          \  -
                                                                                                  ]
 
i NRC Form 386                                                                                                                              U $. NUCLE A715.E1ULATORY COMMIS$10N
,                                                                                                                                                    APMOVED OMO EO 315&O104 LICENSEE EVENT REPORT (LF.R)                                              ' " "a ' 8 ' ''
F ACILITY NAME til                                                                                                              DOCKE T NUMBER (2)                          P AGE i3' Peach Bottom Atomic Power Station - Unit 3                                                                              015 l 0 [0 l 0121718 1 lOFl o 13
    "'''''' Technical Specification Violation Due To Late Surveillance Caused By Test Omission                                                                                                I Because Of A Deficient Proceduro EVENT DATE (51                          LE R NUMBER 16)                    REPORY DATE (7)                        OYMER F ACILITIES INVOLVEG 181 MONT H      DAY      YEAR    YEAR            "      '
                                                            ,                  MONTH        DAY  YEAR              r ac4ut v hAMEs                DOCKEI NUMBEH!Si 015101010 1 I l
                                              ~                  ~
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OPE RATING
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J 20.406(aH1 Het                        60 73(a H2Hm!                      50.73isH2Hal LICENSEE CONT ACT FOR TH11 LER titi NAMt                                                                                                                                              TELEPHONE NUMBER A. A. Fulvio, Regulatory Engineer 7111 7 4 [5 l6 l -l 710 l 1l 4 COMPLETE ONE LINE FOR E ACH COMPONENT F AILURE DESCRISED IN THl$ REPQRT (131 REPORTA LE                                                                      C-                  LE CAUSE    SYSTEM      COMPONENT              MANyt                                            CAUSE SY ST E M  COMPONENT                          R E,POR gpq f A I        I I f                I I I                                                      I        I I I            I I I I        l l I                l l l                                                      l        l l l            1 l l
                                                      $UPPLEMENT AL REPORT E XPECTED (14l                                                                        MONTH    OAV      YEAR SUOM s 5SiON Y t$ tif y.g comer t. EXPECTEO sugwsSION DATfl                                      NO l      l        l
:.eST R ACT a,- a ro ,a n.c          . . .oaov e rn.n n. w.c. am            m.a..v osi                                                                                                ,
On 4/30/90, it was discovered that Technical Specifications (Tech Specs) were violated when Main Steam Isolation Valve closure timing testing was not performed in                                                                                                  ,
its surveillance interval which expired 4/29/90. Tech Specs require this testing to be performed every quarter during power operations. . Testing-is required to be-performed at a reduced reactor power of less then 75%. The cause of the event is an ambiguous. test procedure in that the test frequency requirement was stated as every 3 months when power is less than 75%. The test was incorrectly omitted on the basis that plant power was above 75%.                                        The surveillance test was performed satisfactorily on 5/1/90. There were no safety consequences as a result of this event.                                                                            The ambiguous surveillance test has been revised to clarify the test frequency requ rement. Other surveillances will be reviewed for similar ambiguities.
i There were 2 previous similar events identified.
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NRC Form astA                                    .
U.S. NUCLEAll KEiULAfoRY COMMISSION      /a
                ' } ,:
        .                                  .UCENSEE EVENT REPORT (LER) TEXT CONTINUATION                            senoveo oue no neo-oio4
      ,    ,.,-                                                                                                      EXPIT.ES: 8/31/*;3 ''
FACst:Tv Nanet up .                                              DOCKET NUMBER (23              LGR NUMetR (4)                        PAOt (3)
P:ach Bottom Atomic. Power Station                                                  "'"      TN              UsN
                                                                                                      ~
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  , itXT II rnere spece e reWrect, use ed*oone/ NAC Form J864W H7) .
Requirement: for the Report
        <This report is required per 10 CFR 50.73(a)(2)(1)(B) because.of a condition
: prohibited by Technical ~ Specifications (Tech Specs).
                                    ~
y          Unit Status at Time of the Event (4/29/90)
Unit 3 was-in the RUN-Mode at 99.8% rated power.
There~were no structures, components or systems that were inoperable at the time of
                                                                                  ~
the' event that contributed to the event.
Description of the Event On:4/30/90,-Operations Management personnel discovered that Tech Spec 4.7,0.1.b(2)
        - was violated when Main Steam Isolation Valve (EIIS:V) closure timing testing was not                                                          i performed within its surveillance interval. Surveillance Test (ST) 6.4, " Main-Steam Isolation Valve Closure Timing" is required to be performed every quarter during l power operation conditions. The 25% late grace period allowed by Tech Specs expired on 4/29/90 for this test. Tech Spec requires the performance of this surveillance to be done at a reduced Reactor (EIIS:RPV) power less than 75% of rated. A Reactor
        . power reduction was started 5/1/90, 0008 hours, and the surveillance was completed satisfactorily by 5/1/90, 0333 hours.
Cause of the Event The root cause of this event is an ambiguous surveillance test procedure that resulted in an error in omitting ST 6.4 from the test schedule. ST 6.4 is ambiguous                                                          3 Lin that the test frequency indicates that the test is required to be performed every three months when reactor power is less than 75%. Tech Spec 4.7,0.1.b(2) requires the test to be performed during power operations evfry quarter at a power-level less than 75%. The-test had been incorrectly omitted on 3/6/90 by.the Operations cognizant engineer who coordinates surveillance performance (Non-utility, Non-licensed). . The test was scheduled to be performed during the week starting 3/12/90.
The test was omitted on the basis that it was required to be performed when operating
                  ~
at a Reactor power of less-than 75% and although the-plant was shutdown on 3/6/90, it.
was expected.to be at full power before the scheduled week of performance.                                                                    .!
1A-contributing cause to this event is ambiguity within the Tech Spec surveillance requirement 4.7,0.1.b(2). This Tech Spec states "with reactor power less.than 75%
          ' trip main steam isolation valves individually and verify closure time". On 3/10/90,
        -the Operations Unit Coordinator (Utility, Licensed) reviewed the omission of this
        -test.          He consulted Tech Specs but believed that this Tech Spec required the testing only if the reactor is being operated c.t a power of less than 75%. Since it was expected that Reactor power would be greater than 75% by the scheduled performance, the Operations Unit Coordinator concurred with the omission.
row .                                                                                                          .v . . oc : i.e m o m 9 aooro
 
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senc Poem asea -                                                                                    - U.S. NUCLEam ativt TORY COMMises0N
                  ".                        LICENSEE EVENT REPORT ILER) TEXT CONTINUATION                          uenovEo ous No mo-oio4 y,      .    .%'                                                                                          (XPIRES: $/$1C FActLITY NAME (1)                                            DOCKET NUMeER (2)            LER NUMe(R le)                    PAGE(3)            i 1Paach Bottom Atomic Power Station                                                "'"    "UN#'          :
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ol0        ol3    oF ol 3 TEXT W mese spece 4 tr7pired, ano sapeanst 44C Form 354 W (1h
            , Analysis of the Event L)
There were no actual safety consequences that occurred as a result of this event.
              . Surveillance testing performed on 5/1/90 proved that the Main Steam Isolation Valves                                                      i (MSIVs) were operable. NRC' Generic Letter 37-09 states that it is overly conservative to assume that systems or components are inoperable when a surveillance                                                      ,
has not oeen performed. The purpose of ST 6.4 is to verify that the MSIVs are capable of closing in 3 to 5 seconds to limit the release of radioactive materials and' reactor coolant in design basis events.
Corrective Actions-ST 6.4 has bsen revised to clarify the test frequency requirements. A review of other surveillance tests will be performed to detect ambiguity in test. frequency requirements. Surveillance tests will be revised as-required,                                                                              j It is expected that the test frequency ambiguity that exists in Tech Specs will be corrected.in the Technical Specification improvement Program that is being pursued by                                                    .
PECo in accordance with the Boiling Water Reactor (BWR) Owner's Group's improved BWR'
: Tech Specs effort.
Previous-Similar Events There were 2 previous similar events identified in which surveillance tests were incorrectly placed on the omitted test report resulting in Tech Spec violations. LER
            '2-90-005 concerned omitting a Radiation Monitor functional Test Surveillance due to
            -personnel error involving a misinterpretation of Tech Specs. LER 2-89-032 concerned omitting a Control Rod Exercise Surveillance due to personnel error in failing to follcw procedures. Corrective actions for these LERs involved personnel counseling and enhanced omitted test report review. These corrective actions did not prevent this event because the ccuse-was an ambiguous test procedure.
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