View source for 05000272/LER-1983-023, Supplemental LER 83-023/01X-1:on 830602,notification Received from Westinghouse of Potential Unreviewed Safety Question Re Operation at Full Power Following Extended Reduced Power Operation.Procedures Implemented

{{Adams
| number = ML20083F259
| issue date = 12/13/1983
| title = Supplemental LER 83-023/01X-1:on 830602,notification Received from Westinghouse of Potential Unreviewed Safety Question Re Operation at Full Power Following Extended Reduced Power Operation.Procedures Implemented
| author name = Rupp J, Zupko J
| author affiliation = PUBLIC SERVICE ELECTRIC &amp; GAS CO. OF NEW JERSEY
| addressee name = Murley T
| addressee affiliation = NRC OFFICE OF INSPECTION &amp; ENFORCEMENT (IE REGION I)
| docket = 05000272
| license number = 
| contact person = 
| document report number = LER-83-023-01X, LER-83-23-1X, NUDOCS 8312300222
| document type = LICENSEE EVENT REPORT (SEE ALSO AO RO), TEXT-SAFETY REPORT
| page count = 4
}}
{{LER
| Title = Supplemental LER 83-023/01X-1:on 830602,notification Received from Westinghouse of Potential Unreviewed Safety Question Re Operation at Full Power Following Extended Reduced Power Operation.Procedures Implemented
| Plant = 
| Reporting criterion = 
| Power level = 
| Mode = 
| Docket = 05000272
| LER year = 2083
| LER number = 23
| LER revision = 
| Event date = 
| Report date = 
| ENS = 
| abstract = 
}}

=text=
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LICENSEE EVENT REPORT
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DATE
                ' EVENT DESCRIPTION AND PROBABLE CONSEQUENCES h
[TTT1 lIn June, notification was received from Westinghouse Research Division of a potential l ITTTl i USQ. According to the notice, operation at full power following extended reduced power l                                                                                        l 10 l 4 l leperation could result in increased power peaking in the bottom of the core and possibl{f y o l 3 l l exceeding' limits for the Heat Flux Hot Channel Factor. Neither unit had operated in                                                                                l 0 s        lthe manner identified; and steps were immediately taken to preclude such operation.                                                                             l
  -l 0 l 7 l l Prompt notification was made to the USNRC in accordance with 6.9.1.8h.                                                                                              g l
10181 l                                                                                                                                                                        80 7        8 9 SYSTEM              CAUSE           CAUSE                                                   COMP.           VALVE CODE .              CODE        SU8 CODE                 COMPONENT CODE                  SUSCODE          SUSCODE O 9 8
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SEQUENTIAL                       OCCURRENCE          REPORT                       REVtSION EVENT YE AR                                REPORT NO.                           CODE               TYPE                          NO.
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[g 48 lYl 42                43 g lWl1l2l0 44               47 CAUqE DESCRIPTION AND CORRECTIVE ACTIONS h The measured
[TTT1 1 Procedures were implemented that augment the Technical Specifications.                                                                                                  l
        , ,        l target' value is administratively controlled to ensure the plant is within the RSE                                                                              ,
        , 7        l design tolerance under all operating conditions. We no longer have a safety question..g g
g l Westinghouse concurred with the action.
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i a 8  9 1010101@l12     11 90 LCSS OF OR DAMAGE TO FACILITY TYPE       DESCRIPTION                                                       8312300222 831213 N/A                   PDR ADOCK 05000272                                                                  l 7
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J. L. Rupp NAME OF PREPARER _                                                                                PHON E:(609) 339-4309                                 ",

  ?O PSERG
       ..Public Service Electric and Gas Company. P.O. Box E Hancocks Bridge, New Jersey 08033 Salem Generating Station December 13, 1983 Dr.:. Thomas E. .Murley, Regional. Administrator 2USNRC.
   ; Region 1 -
1631 Park Avenue
   . King of Prussia,fPennsylvania                                           '19406
   ' Dear Dr.= MurleyL
> i LICENSE :NO. DPR-70 DOCKET NO. 50-272 iREPORTABLE OCCURRENCE 83-023/0lX     SUPPLEMENTAL REPORT
   ~ Pursuant to the requirements of Salem Generating Station              !
  -Unit.No. 1 Technical Specifications, Section 6.9.1.8h, Lwe.are submitting supplemental Licensee Event Report for Reportable Occurrence 83-023/01X-1.
Sincerely yours, J. M. .Zupko,    r.
General Manager -
Salem Operations iJR:kilf8/
CC:      Distribution T
tll T_he Energy Peop!e                                                                                   pp
,                                                                                                                       % 2 iB9 (20M) 1181

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                                     ~                                                                                                                                                           .
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             'Raporf Number:'                                83-023/01x-1 s             Raport 1 Date:                               12-13-83                                                                                                                               ;
~
: Occurrence:Date:- ~06-02-83 Fccili ty': -                               Salem Generating Station Unit 1
:Public Service' Electric & Gas Company' l         -
:Hancock's Bridge,'New Jersey 08038 IDENTIFICATION OF OCCURRENCE:
,               Power < Distribution Limits - Heat Flux Hot Channel Factor--
             .Potentially:Out-of-Specification.
            -CONDITIONS ~ PRIOR'TO' OCCURRENCE:
                             ~
Unit 1 - Mode 1                           : Rx Power 100 % - Unit Load 1125 MWe.
             ' Unit.2 - Mode 5 - Rx. Power 0% - Unit Load 05.
I#              DESCRIPTION OF OCCURRENCE:
            'On:J'ne,2,                  1983, during normal operation, notification was received-
                                           ~
.                              u from Westinghouse Research Division of a potential unreviewed safety Equestion involving = operation'at full power following extended reduced
: powerE operation.-- According to-the information received, continuous plant operation atiless than 85% power for. periods greater than 500
,          . MWD /MTU,ffollowed by a return-toffull power in the'same. cycle, could result in increased power peaking at the-bottom of the core.                                                                                                        This
            ' increase could potentially result in exceeding the Technical' Specification. limit ~for the Heat Flux Hot' Channel Factor (Fo (Z)).
Both Salem-Generating' Station units were potentially affected, c1though at the time of_ notification neither unit had.been operated in L               a pattern'of concern, and Unit 2 was shutdown for refueling.
            ' Prompt notification of the Resident NRC' Inspector was performed on June 2, with written confirmation ~ transmitted later that day. The J0perations Department wasfinformedtof the potential problem and work inmediatelyEcommenced on development of procedural changes to prohibit                                                                                                         :   -i
    ,           operation of the plant in a: fashion leading to excessive values of Ff(Z). JA followup! report was submitted on June - 15, 1983.-
APPARENT'CAUSE OF' OCCURRENCE:
Initial = investigation of the problem by the PSE&G Nuclear Fuels Group
            ' revealed-that the' reload safety evaluation (RSE) was based primarily
           -on?unrodded, full power-depletions.                                                                               In the'RSE perturbation analysis                                       i
            'offbounding operating conditions, peaking factors are compared to limits'forfan envelope around' predicted axial flux difference (AFD)
            . values. Subsequent plant operation'is bounded by the RSE~as long as the measured target flux condition is in good' agreement with the predicted. values.
             'Unrodded, : low power operation, however, causes the axial flux to be
            -significantly more positive than at full. power.                                                                                                          Continued operation ot low ~ power.followed by a return.to full power operation can l
i

CER'83-023/01X-1                                   PPAR,ENT CAUSE OF OCCURRENCE:'(cont'd) therefore result in a shift of the target flux position relative to that assumed in the RSE. If the shift is significant, the allowed plant operating AFD conditions may not be bounded by the reload
                    ~
analysis.                                                                                              l l
ANALYSIS OF OCCURRENCE:                                                                                I LThe limits on heat flux-and nuclear.enthalpy hot channel factors
  .snsure that the design limits on peak local power density and minimum                                  )
DNBR are not exceeded and in the event of a LOCA the peak fuel clad tcmperature will not exceed the 2200 F ECCS acceptance criteria limit.                                                                                                 )
  .B: sed on the information that was 'eceived, existing Technical Specification requirements ware insuffic'ent to prevent exceeding the limit for F o(Z). As noted, neither Salem core was operated in a Ernner outlined as being of concern, and mear.ures were immediately
          ~
t ken .to preclude such operation. Due to the presence of an oversight
  -in the- methods used in -the transient or accident analyses which could
  .possibly have resulted in operation in a manner less conservative than essumed'in the analyses, the event was reportable in accordance with
  -Technical Specification 6.9.1.8h.
CORRECTIVE ACTION:
As noted, operation of Salem Unit 1 was continued, and immediate nsasures were taken to insure that the Technical Specification bases were met. Procedures were drafted and incorporated in the Reactor Engineering Manual which now augment the Technical Specifications; the measured target'value is administratively controlled to insure the plant is within the RSE design tolerance under all operating conditions. We therefore no longer have a safety question.
  ' Subsequently, on August 12, 1983, we received a letter from the Wostinghouse Nuclear Fuels Division recommending the same procedural changes that we had already incorporated. These measures, combined with existing Technical Specification requirements will insure safe operation of both units. Based on the information we have at the present time,'a License Change Request is not necessary.
FAILURE DATA:
  -Not Applicable Prepared By     J. Ruon
[' #enerai Manager -
Salem Operations
  .SORC Meeting No. 89-149
                                . . _ _ _ _ _ _ .     . __ .__     _ _ _ _ _ _ _ . - _ _ . . _ _ _ _ .
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