ML19210E568
| ML19210E568 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 10/26/1979 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19210E567 | List: |
| References | |
| NUDOCS 7912050300 | |
| Download: ML19210E568 (7) | |
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UNITED STATES
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 55 TO FACILITY OPERATING LICENSE NO. DPR-28 VERMONT YANKEE ~ NUCLEAR POWER CORPORATION VERMONT YANKEE NUCLEAR POWER STATION DOCKET NO. 50-271 D
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1.0 Introduction R
h\\jij gL Vermont Yankee Nuclear Fower Corporation (VYNPC or licensee) nas proposed changes to the Technical Specifications of the Vermont Yankee Nuclear Power Station (YY) in Reference 1 anc as supplementea oy Reference 2.
The proposec cnanges relate to tne replacement of fuel assemolies consti-tuting refueling of the core for Cycle 7 operation at power levels up to lo65 Nwt (luus power).
In supper; of the reloac application, tne licensee nas enclosec proposea Tecnnical Specification cnanges in Reference 1 anc the GE BWR supplemental licensing suomittal (Refer-ence 3).
This reload involves loacing of prepressurizec GE t:xe retrofit (PexeR) fuel.
The cescription of the nuclear anc mechanical cesigns cf exc retrofit 1s containec in References 4 anc 5.
Reference 4 also contains a complete set of references to topical reports wnicn cescrioe GE's analytical metnods for nuclear, nermal-nycraulic, transient anc acci-cent calculations, anc information regarcing the applicacility of tnese me:nocs to cores containing a mixture of bxb anc exd fuel. Tne use ano safety imolications of prepressurizec fuel are presentec in Appencix 0 to Reference 3 anc have oeen founc acceptable per Reference 6.
Tne conclusions of Reference 6 founc that tne metnoas of Reference 4 were generally applicaole to prepressurizec fuel. Therefore, unless otner-wise specifiec, Reference 4, as supportec by Reference o, is acequate justification for the current application of prepressurizec fuel.
Values for plan -specific cata such as steady s m Operating pressure, core flow, safety and safety / relief valve 1e: points, ratec tnermai O cw e r, ratec steam flow, and c ner cesign parameters are orovicea in
.se erence. nceiticnal plant anc cycle cepencent inf ormation 1s provicec in :ne reloac application (Reference 3) wnicn closely follows ne outline of Appencix A of Reference 4.
A::encix C o# Reference 4 incluces a cescription of tne staff's review, a:p rov al, anc concitions of approval for :ne plant-specific cata ac-
- essec la ;sfere ce. The accve-mentionec pl an -s:eci#ic ca:a r.a'.e oeen usec in :ne transien anc accicent analysis provicec with ne reload a;:011 cation.
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J[ Cur safety evaluation (Reference 4) cf the GE generic reloaa licensing tcpical report has also concludec that the nuclear ano mechanical cesign of the 8xdR fuel, anc GE's analytical metnocs for nuclear ano thermal-hydraulic calculations as appliec to mixec cores containing exo ano exer fuel, are acceptacle. Approval of tne application of tne analytical metnocs cia not incluce plants incorporating a pror.ipt recirculation pump trip (RPT) or Thermal Power honitor (TPM).
Because of our review of a large number of ganeric consicerations relateo. to use of bx6R fuel in mixec loacings, anc on the bas 1s of the evaluations wnicn have oeen presentec in Reference 4, only a limitea number of accitional areas of review have oeen inclucec in
- nis safety evaluation report. For evaluations of areas not specifi-cally accressec in tnis safety evaluation report, tne reacer is re-ferred to Reference 4.
This report also addresses proposed Technical Specification changes submitted by VYNPC in reference 7.
These changes, which concern surveillance of control rod hydraulic return line isolation valves, are discussed in Section 2.6 of this report.
2.0 Evaluation 2.1 Nuclear Characteristics For Cycle 7 operation of Vermont Yankee, 96 fresn PexbR fuel ouncles of type PbDPS2e9 will be loaoed into tne core (Ref. 3).
Tne remaincer of tne 3od fuel cuncles in tne core will oe oc eDEi74L ouncles,1Z4 bDB27*n Dundles,120 8DB219L buncles, and 60 8JPB2c9 ouncles. Inese are all previcusly irraciatec Dundles.
Based on tne data provioed in Reference 3, cotn :ne control roa system and the standoy liquid control system will nave acceptable snutdown capability during Cycle 7.
2.2 inermal Hycraulics 2.2.1 Fuel.Cladcina Intecrity Safety Limit MCPR As stated in Reference a, for %R cores whicn reloac with GE's retrofit ex:R fuel, tne safety limit minims. critical pc.,er ratic (SLMCPR) re-sulting f rca eitner core-wice or locali:ec abnor.ial cperaticnal tran-sients is ecual to 1.v7.
anen meeting tnis 5 :':PR c. ring a ransient,
- a. least 99.r. of tne ft.21 rocs in tne core are expectec to avoic aciling trar.siticq.
Tne 1.u7 S&R to ce usec for Cycle 7 is urcnarsec fr:;n Ine SLMCFR o-eviously a;oroveu for ycle 6.
Tne : asis fcr tnis safety limit is ac:resse: ', Reference 4,. nile ;ur generic a:;revai of tr.e limit is 91 <en in tne staff evaluation included in Reference 4.
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2.2.c C:eratina Limit MCPR Various transient events can recuce tne MCPR from its normal opera.ing level. To assure that the fuel claccing integrity SLMCPR will not be violatec curing any aonormal operational transient, tne most limiting transients nave oeen reanalyzec for tnis reloac cy tne licensee, in or:er to cetemine wnich event results in the largest recuction in the
.inimum critical p0wer ratio. These events nave seen analyzec for tne exposec oxo fuel anc the exposed anc fresn ex R fuel. Acaiticn or tne largest recuctions in critical power ratio to the.5LF.CPR estaclishes tne operating limits for eacn fuel type.
2.2.2.1 Transient Analysis Methocs Tne generi
.stnccs usec f:r.Pese cal:ulati:ns, in:1.:i n; cycle-incepencent initial concitions anc transient input parameters, are cescrioec in Reference 4.
Tne staff evaluation, inclucea as Appencix C of Reference 4, contains our acceptance of tne cycle-incecencent values.
Accitionally, Appencix C contains our evaluation of ne transient anal-ysis metnocs, togetner witn a cescription ano sunnary of the outstanding issues associated witn :nese methocs. Supplementary c, cle-incepencent initial concitions anc transient input parameters usec in tne transient analyses appear in the tales in Sections 6 anc 7 of Reference 3.
Our evaluation 'of the methocs used to cevelop these supplementary input values is also inclucea in Appendix C of Reference 4.
2.2.2.2 Transient Analysis Results The transients evaluatec were the limiting pressure anc power increase transient (turoine trip without bypass in tnis case), the limiting coolant temperature cecrease tran'sient (loss of a feecwater neater),
the feecwater controller f ailure transient, anc the c:ntrol roc with-crawal error transient.
Initial concitions anc transient input para-meters as specifiec in Sections 6 anc 7 of Reference 3 were assumec.
The results of these analyses are outlined in Reference 3 sections 9 anc 10. On tnis topic, Reference 6 founc it acce: acie if fue,1 specific coerating limits are estaclishec for prepressuri:ec ruel as nas Deen
- ne f:r VY.
On this casis, tne transien. analjsis rasel s are a::e;t-acie for use in tne evaluation of tne operating limit MCP.4.
Easec on-
- nis,.ne propcsec Tecnnical Specification macificaticns to operating ii.-it MCrR are acceptacle.
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Acci ce r.: Analyses 2.3.1 ECCS Accencix K Analysis For the previcus cycle, the licensee re-evaluatec :ne acequacy of VY's ECCS performance in connection with toe retr: fit t:xe r-lcac fuel cesign. Tne metnocs usea in :nis analysis.ere previously approvec Dy tne staff. For that cycle, we reviewea :ne ECCS analysis results suomittec by the licensee anc concluceo tnat VY woulc ce in conformance with all the requirements or lu CFR su.*o anc Aopenaix K to lu CFR bu wnen operated in accorcance witn tne MAPLnGR excR versus Average Planar Exposure values wnicn appearea in the proposec plant Tecnnical Specifi-cations.
In Reference 6, we have conclucec Inat !eLnuR limits for prepressurizec fuel is conservatively counc Oy T.ne values tcr ne non-prepressurizea fuels. VY nas conservatively usec tne non-prepressurizec values. Inerefore, casea on our conclusions of Reference 6, the pro-posec MAPLHGR limits are acceptaole.
2.3.2 Control Roo Droo Acciaent For VY Cycle 7, the accicent reactivity snape function (colo) coes not satisfy the requirements for ne councing analyses cescrioec in Refer-ence 4 Tnerefore, e was necessary for tne licensee to perform a plant anc cycle specific analysis for the control roc crop accicent.
The results of tnis analysis incicateo that tne peak fuel entnalpy for nis event woula be at mo t 130 calories per gram.W) Since this is well below the criterion of 2eo calories per gram, we finc tne results of tnis analysis to be acc eptaole.
2.3.3 Fuel Loacine Error Potential fuel loacing errcrs involving misorientec ouncles anc ouncles loacec into incorrect positions have Deen analyzec. Tnis bE me nod for analysis of miserientec and misloaced ouncles has seen reviewec ano approvc1 by the staff and is part of the Reference 4 metnocology.
In orcer to accress cur concerns on the fuel icacing errors for tne pre-vicus cycle, :ne licensee anc we agreec to an F.A ac;ustment on radio-logical incica:1cns of potential fuei loacin; errors. Tne iicensee nas proposed similar requirements for tnis cycle or.icn.e finc acceptacle.
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s 2.3.4 Lveroressure Analysis Tne overpressure analysis for the MSIV cicsure with hign flax scram,
- nich is tne limiting overpressure event, has oeen performec in accorcance witn tne requirements of Reference 4.
As specifiec in the staff evaluation incluceo in Reference 4, tne sensitivity of peak vessel pressure to failure of one safety valve has also teen evaluatec.
We agree that tnere i-sufficient margin oetween tne peak calculatec vessel pressure an 2 Ine cesign limit pressure to allow for tne failure of at least one valve. Therefore, the limi.ing over-pressure event as analyzed by tne licensee is consioereo acceptaole.
2.4 Thermal Hycraulic Stability Tne results of the thermal hycraulic stacility analysis (3) show that tne channel hycrodynamic and reactor core cecay ratios at the natural circulation - 1052 roc line intersection (wnicn is tne least stable physically attainacle point of operation) are Delow the sta-oility limit.
Because operation in the natural circulation moce at greater tnan 1*. rated tnermal power will De prohibited by Technical Specifica-tions, there will De acced margin to tne staDility limit and this is acceptaole.
2.5 Startuo Test Program The licensee has not changed his startup test program from that approved for the previous cycle. This program therefore remains acceptable.
2.6 Tecnnical Specifications The only change to the VY Technical Specifications involving core refueling that has not yet been discussed is the elimination of operating limit MCPR for 7x7 fuel and a statement that 7x7 fuel MCPR limits have not been established and that future use of 7x7 fuel would require further evaluation. L and the licensee have agreed to such a specification.
In reference 7 the VYNPC requested that Table 4.7.2.6 of the Technical Specifications be changed to delete valves V 3-110 and V 3-113 and add valve V 3-181 to the listing of valves subject to Type C leakage tes ts. This enange corrects Table 4.7.2.6 to delete two valves which no longer exist ir. the control rod hydraulic return line and adds an additicnal salve to the table. This change is acministrative in nature anc is accer. cable.
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. 3.0 Environmental Considerations We have determined that the amendment does not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact. Having made this detennination, we have further concluded that the amendment involves an action which is insignificant from the standpoint o.: environmental impact, and pursuant to 10 CFR Section 51.5(d)(4) that an environmental impact statement, or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance of the amendment.
4.0 Conclusion we have concluded, based on tne considerations discussed above, that:
(1, because the amendment does not involve a significant increase in the probability or consequences of accidents previously considered and does not involve a signi.~icant decrease in a safety margin, the amendment does not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be er, tangered 'by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.
Dated:
October 26, 1979 1500 177
s Referances 1.
Letter, D. E. Vancenourgh (VYNPC) to Office of Nuclear Reactor Regulation (USNRC), cated August 21, 1979.
2.
Letter, D. E. Vancenourgn (VYNPC) to Office of Nuclear Reactor Regulation (USNRC), cated Octocer 5,1979.
3.
" Supplemental Reload Licensing Suomittal for Vermont Yankee Nuclear Power Station Reloaa 6,"
NEu0-2420e, August 1979.
4
" General Electric Boiling Water Reactor Generic Reloaa Application,"
NEDE-24u11-P-A, May 1977.
5.
Letter, R. E. Engel (GE) to U. S. Nuclear Regulatory Ccamission, cated January 30, 1979.
6.
Letter, T. A. Ippolito (US'hRC) to R. Gricley (GE), April 10, 1979 ano encloseo sts.
7.
Letter, D. E. Vandenburgh (VYNPC) to Office of Nuclear Reactor Regulation dated October 5,1979, WVY 79-113.
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