View source for BECO-90-136, Application for Amend to License DPR-35,revising Tech Specs to Upgrade Safety Limit Min Critical Power Ratio & Revise Operating Limit Min Critical Power Ratio.Revs Will Maximize Fuel Utilization During Fuel Cycle

{{Adams
| number = ML20062E353
| issue date = 11/08/1990
| title = Application for Amend to License DPR-35,revising Tech Specs to Upgrade Safety Limit Min Critical Power Ratio & Revise Operating Limit Min Critical Power Ratio.Revs Will Maximize Fuel Utilization During Fuel Cycle
| author name = Bird R
| author affiliation = BOSTON EDISON CO.
| addressee name = 
| addressee affiliation = NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
| docket = 05000293
| license number = 
| contact person = 
| document report number = BECO-90-136, NUDOCS 9011200130
| package number = ML20062E357
| document type = OPERATING LICENSES-APPLIATION TO AMEND-RENEW EXISTING, TEXT-LICENSE APPLICATIONS &amp; PERMITS
| page count = 10
| project = 
| stage = Request
}}

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                                    ,                                g,                                                    10CFR50.90
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00$1%EDi$ON Pdghm Nuclear FOwer stationi                                                                           =j Roch Hill Rc.ad       .
Plymouth, Massachusetts'02360                                                                                '
                                                                                                                                                                   .J L                Ralph G. Bird                                                                                                                  '
                                                                                                                                                                   '3 senior Vice Prestaent - Nudear                                                                           .BECo 90-i36'                                l November 8. 1990 U.S. Nuclear Regulatory-Commission ~                                                                                                                  a Document Control Desk'                                                                                                                                y Hashington, DC 20555                                                                                                                                 j
                                                                                                                         ' License'DPRl 35' Docket 50-293-PROPOSED TECHNICALLSPECIFICATION-                                                          .
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CHANGE MINIMUM CRITICAL r3HER RAJJ.Q.                                                                                !
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            . Boston- Edison -Company proposes the attached revisions ;to. Appendix A of.                         -
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   ,          Operating License DPR-35 for the' Pilgrim Nuclear' Power Station in-accordance.                                                                       H with-10CFR50.90. The' pro' posed revision :to :the1 Technical' Specifications                                                                         ~
upgrades the Safety Limit Minimum Critical Power Ratio and revises ,thet                                                                                J Operating Limit Minimum Critical! Power A. ''                                                                                                          ;
These. proposed revisionsnwill extend the ,e of: spectral! shift to maximize fuel                                                                      !
utilization during the present fuel cycle. Thisuchange tis <not _ nee'ded-if our
_ proposed change of August .21, a1990-is- approvedifirst F '                                                     .
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l Then personally appeared before me, George H.; Davis,-who'being duly. sworn,-did-state that he =is Vice President -_ Nuclear' Admi.nistration of Boston Edison Company and that hefis. duly authorized'to executeland file the; submittal contained hereiniin the.nameland on sbehalf'ofiBoston EdisonLCompany and that                                                                          '
the' statements in said-submitta1Jare truelto the best ofchis-:                                            knowledge and-                            i belief.                                                                                   "
My commission en               s:JO              M /995~
                                                          .DATE '                   ;7'
                                                                                                                  )#             ub NOTARYP0pIC n
Attachments: . A.< Description'.of' Propose'[Chinges 1
d                                                                                          :
                                   'B. Replacement Technical l Specification Pages e"
C. Marked-up Technica1LSpecif,1 cation;Pages D. Supplemental Reload Licensing Submitted for Pilgrim Nuclear'                             '
Power ' Stationereload l 7o Cycl e. 8,123A4800 - Rev .- l.-                       -
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l'sigscd original and 37Lcopies '               L h.
cc:    See.nex.t page                                        ,
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                ^w         9011200130 901108                     &,                              is    't r     >
PDR          ADOCK 05000293            %,                                                                                          /g, p                           PDG       4;3lp'    's'                                                                         +:        .j
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3                        T   5 BOSTONLEDISON COMPANY'                 ,
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                                 - U.S. Nuclear Regulatory Commission                                                                                                                                                ]
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                                                                                                                                                                                                                  ,          i 4             s cc:      Mr. R. Eaton, Project Manager                                  .
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Divis_ ion of.' Reactor _ Projects'- I/II Office of Nuclear Reactor Regulationi                                                                                                                                             l Mail Stop: .14D1                                   1
                                                                                                                                                                                                                     ,l U. S. Nuclear Regulatory Commission                                                              ,
1 Hhite Flint North:                                                                                                                                                             $
11555 Rockville' Pike-Rockville, MD 20852                                                                                                                                                             1 i
U.:S. Nuclear Regulatory Commission':                                                                     4 i
Region I      _
475 Allendale Road ,              .
King of-Prussia,'PA':19406;
                                           . Senior NRC Resident Inspector-
                                          . Pilgrim Nuclear Power. Station:
Mr.-RobertM.:Hallisey,;Directorj Radiation Control Program                 . .               -
                                                                                                                                                                                                 ,                         ~
Massachusetts Department of Public Health'.
150 Tremont Street, 2nd' Floor _..                                                                                                       ,
3 Boston, MA' 02111-
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                                                                                                        ^       d Atta'chment A'.-to BECo. 90-136                                         j Qginintion' of Proposed Chance                                                  1 1
Pronosed Chad 221-                                                                                             .
Boston-Edison Company proposes to upgrade' the Safety Limit Minimum Critical Power Ratio (MCPR)-and revise the'0perating Limit Minimum Critical Power <
Ratio. The revision;to the' Safety Limit Minimum Critical Power Ratio reflects                                '
improved fur,) de:1gns in the core _ and the General Electric Boiling. Hater:                          '
Reactor Ger.eric- Reload Fuel, Application, NEDE-24011-P.                                                     ,
The revision to the Operating Lim',OMinimum'CriticalL Power Ratio is being                           .
proposed to be consistent with the upgraded Safety Limit'MCPR and to' reflect spectral' shift operation. Spcctral. shift. operation resultsfin' top peaked l flux' distriLution patterns <toward the end of.the fuel' cycle.- A. top. peaked flux:
distribution pattern increases the change in the Critical Power. Ratio. (ACPR)'
associated with'the analyzed operational trans.ients. The proposed-revision l
maintains conservative operating limits.
The upgraded Safety Limit -MCPR and. thetrevised Operating' Limit MCPR_ ensureLthe.
plant will be operated safely and willinot ' pose an,unoue: risk to the health and-                              I safety of the public. These revisions willzallow the plant to be. operated more;
                                                                                                  ~                     ~
efficiently.
Basis for Changg-The NRC approved the upgraded Safety Liinit MCPR in its' safety. evaluation for-                                 1 Amendment 14 to NEDE-240ll-P-A, " General Electric ' Standard Application for                                 J Reactor Fuel'.', dated December:27,1987; Incorporation of the upgraded. Safety                                  i Limit-MCPR is appropriate -for BWR's with.0-lattice- fuel assemblies:provided:7,_                               l
: 1) the fuel has a beginning of life'R-factor greater 1than4or equal:to1 1.04'and.                                l H                         consists of! fuel -types P8x8R BP8x8R,3 GE8x8E or GE8x8EB;- 2) the fuel is at L                         least 2.80 weight percent,U-235 bund _le average enrichment; and 3)1 the lower enrichment bundles in the core hav'efoperated for_-at:least 2 cycles'..
                        'The fuel assemblies presently. in:the core of the Pilgrim Nuclear: Power: Station'                                l satisfy the conditions; evaluated in NEDE-24011-P-A. _These conditions;wil_1lalso                                I be satisfied for future reloads',
The Supplemental Reload Licensing Submittal' for Pilgrim' Nuclear. Power Station,                                 l Reload 7 Cycle 8. submitted by.BECo letter 87.081: dated:May 22~ 1987-has been.     ,
revised to reflect operati_on:Using?a spectral shift fuel managementLstrategy.
a                      The' revise.d analysis also, reflects an' assumption that the Turbine Bypass Valves;                        a
:do not open in the analysisLof the Feedwater Controller Failureiwith maximum,                                    !
                        ' demand. This-assumption was made becausetprecise. measurements of the opening time are not available.         The' assumption _ is. conservative and"it results in a                        3 more severe transient analysis.and:aimore conservative . Operating; Limit MCPR.                                 ,
This limit'is incorporated in the proposed Technical Specification. The-                                       +
proposed Operating Limit MCPR willoensure the MCPR does not' decrease-below the~
proposed Safety Limit MCPR at anyttime during any abnormal operating transient.
!-                        as defined in the FSAR.
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                                                                                                                  ,                 l Determination of No Sionificant Hazards Considerations The code of Federal Regulations (10CFR50.91). requires-licensees requesting an                                              I amendment-to provide an analysis, using the standards!in 10CFR50.92,' that                              -
J determines whether a significant hazard consideration exists. The following                                              :l analysis is provided in-accordance with-10CFR50.91 and-10CFR50.92'for the' proposed amendment.to Pilgrim's Minimum Critical: Power Ratio.
: 1.         Upgraded Safety Limit MCPR.                                                                                      i A.          The proposed change does not involve 6!significant increase'in the.                             .
                                                                                                                                    )
probability or consequences of an accident lpreviously evaluated because the.                              =l proposed change to the Safety Limit MCPR does not change ~ plant equipment,                                  H operating procedures, or governing design. criteria _used-to protect >the; plant against the initiati_on of any analyzed accident or used to mitigate-the consequences of any' analyzed' accident.;
B.          The proposed change does.not create the. possibility of a new or!different                                      l kind of accident from any previously analyzed'because.the proposed change                                       I does-_not change plant et,'uipment,' operating procedures .or governing design criteria-and the change toithe Safety Limit MCPR:provides the same level of:
protection as the existing' Safety Limit MCPR.againstLfuel cladding-failureL during an abnormal operational transient. The proposed Safety. Limit' MCPR            .
                 . therefore provides equal: assurance against a. release'of radioactive l
material in excess of.10 CFR 20 limits. during abnormal operational  '
                                                                                                                                  +
transients and a new event. sequence leading to an accident is not2 created.
q C.,         The following design requirement ensures an' adequate safety margin is                                     i  '
maintained:
Abnormal operational transients caused by aLsingle operator error or                         w equipment malfunction shall be limited such that,-:considering                                   :
uncertainties in manufacturing and moni_toring,the core operating                             i state, more than 99.97.Lof the: fuel l rods:would be expectedLto avoid.                       a boiling transition.                                             ''
Theprposedchangedoesnotfinvolve.asignificantreduction;inthemargin-1 of safety because this design' requirement which governs fuel cladding .                                      ,
integrityandmaintainsthedefense-in-depthphilosophy,hasinotchanged.
: 2.          Revised Operating' Limited MCPRL                                                                          'l A. - The proposed change does noCinvolve.a:significant increase in thel                                                 j probability;or consequencesLof.an accident previously, evaluated because the
                  - proposed change'to the;0peratingl Limit MCPR doe's.not change plant equipment, operating procedures,-or; governing. design'criteriaiused to-protect the plant against the'initiationfof any analyzed l accident orcused'                                  a
                   - to mitigaterthe conseque_nces-_of any analyzed accident.
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B. The proposed change does not create.the possibilityTof(a-new or;differentL                                                                                                                          -             '
kind.of accident from any'accider.tfpreviously evaluated'because the ;
e                                      -proposed change:does:not change plant equipment,-operating procedures',4or?
governing design criteria and'the: changes to Operating Limit MCPRs;providei                                                                                                                                      j the same-level of' assurance thatsthe- Safety Limit MCPR will' not:be' exceeded during.an' abnormal operational transient thereby assuring a release of; '
;'                                           radioactive material = in' excess ofL10 CFR 20 limits ~will not: occur'during;                                                                                                                 ,
abnormal operational transients and a' new event sequence / leading-to>an:
accident has not.i.een created                                                                   ,                       .m.
                                                                                                                                                                         ..                                                                                            i
                                                                                                                                                            . .                          . . i..                          ..                          * .'''9 C. 'The' proposed change doesinot involve;aLsignificantJreduction:in the margin l=
                                           'of3 safety because the conservative Operating Limit MCPR:ensuresLthe most:                                                                                                                                        -,
limiting transient will not violate; the SafetyfLimit MCPR..,                                                                          .+
B1 guested Schedule-                                                        ,
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The requested change /is not needed                   ! untildhe,present fuel (burnup;reac'hes 7200;
                                    -HH9/T, about January 31,.1991'. , Additionally,s approval .off this'. t equest willo noti
                                                ' ed if our proposed change;of August 21.-L1990 isiapproved first. >
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                                                                                        . Attachment;B'to BECo 90-136'-
List of Effective Pages-                                                                       '
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                                            '                                                                                                                                                                                                                                                                                     l Revised Pages                                                                                                                                                                                  ,                  j
                                                                                                                        <     6,                                                                                                                                                                                           -1 12-                                                                                                                         4
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p LI ' SAFETY LIMIT -                                                               2.1             LIMITING SAFETY SYSTEM SETTING _ .                                                                                       ;
l 1'1        FUEL CLADDING INTEGRITY'                                            '2.1 ~ FUEL CLADDING INTEGRITY 5 ADDIicabilitV:                                                              '
ADDIiCabiIitV                                                                                                          j L                                                        .
                                                                                                                                                                        ..                                        . -. .                            j L
Applies to the' interrelated                                                          Applies to tripJsettings of the-                                                                                       j variables associated with fuel            '
instruments and devices which?are-                                                                                   n L                                    thermal behavior.                                                                     provided to prevent the reactort                                                                                         l K                                                                                                                          system: safety limits,from being.                                                                     '                  1 L                                    Obiective:                                                                            exceeded.                                  ,
                                                                                                                                                                                                                                                 ~l L                                   To establish limits below which                                                        Obiective:s L                                    the integrity of-the fuel clade                                                                   .E.             .
                                                                                                                                                                                                # .                                               R
!                                    ding'is preserved.-                                                             lTo define the level of the process                                                                                             I L,
variables at.which automatic; ,                                                                                       1 Soeci fi ca ti.QD:.              .
protectite_-action is initiated toi 7                   prevent the fuel claddingt                                                                                            1 Reador Pressure >800 osia'and                                                    ' integrity.saf ety limits from tieing?
                        .A-Core Flow >10% of Rated.                                                              exceeded.-                                                                                                                l The existence:of a minimum                                                            Speci f s ca' tion:-
critical . pcwer ratio (MCPR) less '                                      .       .
l
(         than 1.04'sh611l constitute '.                                       A.-              Neutt'on' Flux' Scram;                                               '
                                  ',integrity violationsafet'y   of the'fuelicladding limit., A HCPR'                                      -
                                                                                                                                                                                                                    . 1.
l        of 1.04 is hereinafter referredL                                                       The limiting safety; system trip?                                                                                      ]
                                                                                                                  ,4
: l.                                   to as the Safety: Li.mit MCPR.;                                                   ~ settings shall be'as specified                           .
                                                                                                                                                                                                                                                =}
: below: -                                                                                                                ';
B.         Core Thermal' Power Limit 1(Reactor                                                                                                    .              .                 .
4 Pressure 1800 psia.and/or; Core.                                                  . l.-Neutron Flux' Trio-Settinas                                                                                            r Flow 110%).                                                                                 .                                       .
j
                                                                                                                                .a.,=APRM Flux Scram Trioc e' .
When the reactor pressure                                                                                 Settina1Run Mode)-                                                                                 J
                                  -1 800 psia or core-flow is,is-                  less                                                                  ,                   .. ,                 .           .          . . .
than-or equal to?l0% of rated,                                                                           When the ModetSwitch is in-                                                                          i the' steady state core' thermal                                                                          the:RUN position ktheLAPRM power shall not exceed 25% of                         '
ifluxi scram: trip. setting                                                                         ci design 1 thermal power.                                                                                 =shall be:                                                                                           1 Power Transient' b                        C.                                                                                                                          S f .58H~+ 62% 2 1000' a
The safety l limit shall be assumed                                                                    'Hhere:                                                                                              i u                                 sto be exceeded when scram is                         .
                                                                                                                                             ..                     .          .                                                                  J j'                                  known-to have been accomplished                                                                         cS = Setting in percent of                                                                             I f                                    by a means other;than the                                                                                                   rated thermalzpower expected scram signal?unless.-                                                                                                                                                                             N b'                                                                                                                                                            L(1998 MHt)-
L                                    analyses--demonstrate-that the                                                                                                                   .                 .,                                         i K                                    fuel cladding:. integrity l safety:                                                                      H - Percent-of drive flow:                                                                         ;[
e                                    limits defined. in' Specifications -
Jto produce'a rated; core-                                           '
1
                                  'l.1A and l'.lB sare not': exceeded                                      ,                                                    flow of-69 H lb/hr.   '                                                            !
b                                   during the actual transient.                                                                                                                              ' '
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                                                                                                                                                                                                            ~6                                    i j             ,
Amendment No. 721                                                                                                                                                                                  '
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i                                         i BASES:
                                                                                                                                                                              '                       1
: j.                                         The statistical: analysis used to-determine th'e MCPR safety: limit is-                                                                        - '
                                                                                                                                                                                                       )
based on a model of the BHR corr which simulates theLprocess computer                                                                                   3 l:                                          function. The reactor. core selected for these analyses,was a large 76C                                                                              1 assembly, 251 inch reload core; Results from:the large reload-core-                                                                                      ;
L                                           analysis apply for all operating. reactors:for all? reload cycles,                 c                                                ,
;                                           including equilibrium cycles. ' Random Monte Carlo selections-of all-                                                                                      l operating' parameters; based. on= the ' uncertainty. rangesiof manufacturing =
tolerances, uncertainties in-measurement of. core operating parameters,                                                                               J L                                           calculational uncertainties, and. statistical uncertainty' associated with?                                                                              ,
the critical power correlations are imposed'upon the' analytical                                                                                         '
                                         . representation'of the coreJand-the resulting-bundle critical power ratios. Details.of thisfstatistical. analysis are presented lin Reference 2.-                                                                                                                                                    ;
B. Core Thermal Power Limit-(Reactor: Pressure <-800 osia or Core Flow,                                                                                     :7
                                             < 10% of Rated)                                                                                                        ,                                ,
1 The use of the GEXL1 correlation is not valid'.for the critical power 1 ..                                                                           "
O calculations'at pressures-below:800;psig.or cor.e flows less1 than1 10% ofc rated. Therefore, the' fuel claddinglintegrity safety' limit is;                                                                                    _;
established by c,ther means.:DThis is done by; establishing a. limiting                                                              <
condition'ofcore; thermal-poweroperation.withnthe-followingbasis..                                             .                                     3 Sitce the pressure. drop inithe bypass regionsis essentially all elev6 tion head whichiis 4;56.psiitheicore pressure: drop at lowLpower/and-                                                                               '
all-flows wil.1: iways be greater than 4.56 psin Analyses show,that wi.th                                                              '
a flow of 28x10 lu./hrl bundletflow, -bundle pressure drop-is nearly-                                                                                 4 inden ndent-of bundle power:and hasca)valueiof:3.5 psi. Thus the                                                                                          ]
bundle flow with a 4.56; psi' driving -head willibe
* greater than '28x103 :
c lbs/hr irrespective:of- totalfc6te:flowland independent of bundle power                                                                                    ;
for:the range of, bundle-powersjofLeoncern. Full: scale ATLAS test. data y                                            taken at. pressures 1from 14.7 psia tto 800 psia 1 indicate that the fuel:
assembly critical powerJat this flow ris ?approximately 3.'35:MWt'. ; Hith'-                                                                           s the design peaking; factors the.3.35lMHt; bundle power corresponds to'a:                                                                          ';
core thermallpower :of more than150%L Therefore: a core t'iermal: power -
limit of 25% for reactor: pressures ;below 800 psia, or_ core flow less,                                                                             ,
than 10% is conservativ'e; j
                                                                                                                                                    ,                                             1 1
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d BASES:                                                   <
C. Power Transient,                                                          .
Plantsafetyanalyseshaveshown[thatthescrams.causedbyexceeding.
any safety setting will assure that:the Safety < Limit lof Specification                                                _l 1.lA or 1.1B will.not be. exceeded.:' Scram. times are: checked
                                    -periodically to assure tht insertion times are adequate . The thermal                   .
power transient resultingLwhen'a scram is accomplished other;than;by                                                         ;
Jthe expected scram signal (e.g., scram from. neutron flux followingL                                                          -
closures of the~ main ' turbine stop valves) does notJnecessarily cause '                                               1' fuel damage. However,; for: this specification la Safety Limit violation:
will be assumed when a seramLisionly; accomplished by means:of a,backupL                                        '
feature of the' plant design.- The concept of,not: approaching'aiSafety                                                     ;
umit provided scram signals are;operableLis1 supported by.the. extensive" plant safety analysis.
The computer 'provided with Pilgrimidnit l' has isequ(nce annunciationi
                                     . program which will indicate lthe sequence ~in.which events-such'.as: scram,.                                             .i
: APRM trip initiationh pressurefscramiinitiation, etc.,' occur. ;This -
program also, indicates when the' scram setpok nis cleered. .This-will!
provide-information.on.howLlong;a scram condition' exists,and thust                                                         .
providesome(measureof'the'energyadded-during(a' transient.                                                                  !
                               -D. Reactor Water Level (Shutdown Condition)
During periods' when :the' reactortis' shutdown',jconsideraGon must alsoEbe                                               1 given to water level requirements due.Lto the:effect.of decay heat.4If reactor water level .shouldsdrop below ,the top of the : active fuel during                                               i this time,' the abilityito coo 1Lthe' core is reducedt TThis1 reduction in                                                >
core cooling capabilityfcould:leaditotelevatedicladding temperaturesi                                                       ;
and clad perforation.! ..The core can1be cooled'sufficiently;should the,                                                   j water level be1 reduced toEtwo-thirdsithe coretheight.3 3Establishmenti of.                                                i the safety.limitiatal2'inchesiabove the; top ofmthel fuel provides;
                                                                                          ~
adequate margin. This level will"be{ continuously monitored.
References                                                       ,                                            ,
y
      .                                                                .r                                                                                        1 1.GeneralElectric:ThermalAnalysisiasis((GETAB):
Data,uCorrelation and                        '
l-                                   . De si gn Appli cati on',1 General El ectri c Co ,q BHR ' Systems l Department ,
L                                      November 1973 (NEDO-10958).                                                                                               (
: 2. General Electric' Boiling Hater Reactor Generic Reload / Fuel? Applicatio'n,                                                "
                                     'HEDE-24011-A.
                                                                                               ,                         a:.
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TABLE 3.11-1
                                                                                                         -' OPERATING LIMIT MCPR VALUES-                                                                                                   ,;
A.           MCPRL0perating Limit from Beginning of Cycle (BOCF to BOC E7,953 MHD/ST..
P8x8R/BP8x8R For all values of t                                                                                                  - 1.48 -
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B.- MCPR Operating Limit from BOC-+ 7,953 MWD /ST.tW End;of' Cycle.'.                                                                                                                                           ?
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0.0 < t 1 0.1-                                                                               ,               -1?42-                                                                       ;
0.1.< t 1 0.2'                                                                                                 1.43-                                                              q 0.2.<~t 1:0.3;                                                                                                 1.44'.                                                       -
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0.4.<.t 1 0.5                                                                                                  1.46 0.5'<.t 1:0.6                                                                                                  1.47 0.6 < t 1 0.7.-                                                                                                 1.48' n
0.7.< t 1 0'8                 .
                                                                                                                                                                           -1.49c   -
l 1;50                                                            j 0.8 < t 1 0.9 0.9 < 2 1 1.0' .
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