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
| ML20062E353 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 11/08/1990 |
| From: | Bird R BOSTON EDISON CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20062E357 | List: |
| References | |
| BECO-90-136, NUDOCS 9011200130 | |
| Download: ML20062E353 (10) | |
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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- .
( CHANGE MINIMUM CRITICAL r3HER RAJJ.Q. ! l
. Boston- Edison -Company proposes the attached revisions ;to. Appendix A of. -
i
, 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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( G.{Bnm. rd ; RAH /njm/4820 V w.9' ; Li Commonwealth of Massachusetts) ..r ! g County of Plymouth- .) L ,
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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.- - 1 t l'sigscd original and 37Lcopies ' L h. cc: See.nex.t page , l 6 ft # g[ f
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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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l ( 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 . . -
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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;
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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. 1 ( ,1-5
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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.
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. . . . 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- , m m , 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- ' f 1 E
' l Revised Pages , j
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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 .
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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. .
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! 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' -
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l of 1.04 is hereinafter referredL The limiting safety; system trip? ]
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- l. to as the Safety: Li.mit MCPR.; ~ settings shall be'as specified .
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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. ' ' a n, i p ' r- Revision: .. '
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Amendment No. 721 ' j c g- > a [j q-. > u in ' ' l w < u' p n x . __ 4 . , (I Y i ? , QE - .QM';.)%;g %pp , Vi i'M
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- j. The statistical: analysis used to-determine th'e MCPR safety: limit is- - '
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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
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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;
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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.
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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.'. ?
,1 P8x8R/BP8x8R' t m
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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'. - l t 0.3 < t ILO.4' 1.'45 ? ,
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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 .
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