ML20056B671

Forwards Negative Declaration for Amend to License DPR-22

ML20056B671
Person / Time
Site:	Monticello
Issue date:	10/21/1975
From:	Regan W Office of Nuclear Reactor Regulation
To:	Ziemann D Office of Nuclear Reactor Regulation
References
NUDOCS 9102070696
Download: ML20056B671 (3)

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j j NUCLE AR REGULATORY COMMISSION w ASHIN GTO N, D. C. 20555 CCI : I UI.

Docket No. 50-263 Dennis Ziemann, Chief, Operating Reactors Branch 2, DRL NEGATIVE DECLARATION FOR TECHNICAL SPECIFICATION CHANGES TO MONTICELLO UNIT 1 Attschcd is the Environmental Impact Appraisal and Negative Declaration associated with proposed changes in Technical Specifications appropriate to implementation of the ECCS Acceptance Criteria.

1 33\.'lW-f r W:n. H. P,egan, Jr. , Chief Environmantal Projects Branch 4 Division of Reactor Licensing

Enclosures:

1. Environmental Impact I '

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Appraisal f j Njf E. Negative Declaration ' /

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V The bypass flow has bcen considered in the determination of the MCPR limit. Finger sprines have been attached to the lower end fittings of the reload fuel to taintain the core bypass flow within the range of the bounding analysis. In the bounding analysis, 12% bypass flow is assumed.

'!he uncertainty of this bypass flow is factored in the total core flow uncertainty that is used in the GETAL analysis.

The optrating linit 1:CPR is based on the rost limiting transient , a t ur N r, t rip ti:!c bypasr frc:: 90'. power and 100' f! cv.. conditions. The calculat ed decrease in MCPR during this transient is 0.27 for 7 x 7 fuel and 0.35 for b x E fuel . The resulting operating limit MCPR is 1.33 for "x 7 fuel :od 1.41 for S y F fuel.

The required operating limit MCPR is a function of the magnitude and location of the axial and rod-to-rod power penLing. In determining l

• e rn, ti red MCPR, axial and Ic cal pealing representative of beginning-of-cycit were assured. That is, R-factors of 1.075 for 7 x 7 fuel and 1.10.' for b x S fuel and an axial peaking factor of 1.57 at a point 1/4 of the hcated length below the top of the fuel were assumed. This is the most adverse set of local and axial peahing factors. During the cycle the luct.1 pealing, and therefore the R-fact or, is reduced while the peak ib :xial r t: ;,e :1r e : touard tF bot t cc cf the core. Althouph the epc: a t .i n li; n 'TL would be increased by npproxirviely li by the rtdecc. <

J-nf-cg l e P- fact or, thir is effrei by t Le reduction in MCPR ro ult i: f;.

thc : elocation of the axini peak t o below the r.idp]: m

2. Conclusi ens i:ecarding Accept abilit y of GETAB-Lased Technical 9,c c a 15 c a t i c:

lhe APiN scra: .nd : oJ block setting changes suggest ed in Mr. Mayer's July 10, 1975 let ter to D. L. Ziemann are not part of the GETAB-GEXL changes. A definitive stability analysis has not been presented for the APfJ' serem anj rod block setting changes so these changes cannot be accepted at this tiuc. Houcver, t he GETAD-GEXL changes are well de vn.:nt ed and are h:rhly desirable in view of the cuch improved dat:

base f or the GEXL over that for the previously used Hench-Levy ?XHF correlation. The proposed technical specification changes for incor-parating the GETAI-GEXL analysis are acceptable.

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h Lave conclu..., but d on the considerations discussed above, that 1 (1) there is reasonable assurance that the health and safety of the public j will not be endr.ncered by operation in the proposed manner, and (2) such  ;

activit ic s ti)) 1; conducted in compliance with the Coa. mission's j regulations and 1 2 issuance of this amendment will not be inimical to the l con. :en def e ns e nr. security or to the health and safety of the public.

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L. Ge al Lin ric Ther .a1 .nalysis 1 asis (Gl u M

1. Evaluation of GETAb-Based Technical Specifications ,

I The GE generic S x 8 fuel reload topical (5) describes the thermal- ~

hydraulic methods used to establish the thermal margins. However, based on our review of this topical we have found the GETAB application description to be incomplete. Therefore,wehagevaluatedtheMonticello thc rr.a1 cargins 1 ased on the NEDO-10958 report which the staff has previously found to be acceptabic end plant specific input information providt d by t he licensee in it s application dated March 12, 1975, as supplerented by NSP letters dated July 10, 1975 and July 24, 1975 The fuel cladding int egrity safety limit MCPR for both the 6 x 6 and

< x / : n 1 i s 1. 0 f . It is based on the GETAB statistical analysis wh4ch assures that 99.91 of the fucI rods in the core are expected to avoid boiling transition. The uncertainties in the core and system rating parameters and the GEXL correlation (Table 4-1 of NEDO-20694) . combined with the relative bundle power distribution in the core form the basis for the GhTAh stat ist ical det ermination of t he r afety linit MCPR. The bases for th(se uncertainties are reported in NEDO-20340 b) and are acceptable.

Tl e bv;.dle powe: di s tri but io: used in the GETAB analysis conserv:.tively assu::e> .aore hir,h power bundles than would be expected during operation of the reacter.

In conparing the tabulated lists of uncertainties for Monticello uith those in NEh0-1095E we have found only one difference. The Monticello r,ndah. deviation for the TIP readings uncertainty is 5.7'. where:.s the Cl:Tal: ' EDD-10955 report shows 6.3%. The increase in uncertainty for Monticello is a consequence of the increase in uncertainty in the measure--

ment of power in a reload core. A TIP reading uncertainty of 6.3's would bc applicabic if this were the initial core. In both cases the TIP reading uncertrintics are based on a symmetrical planar power distribution and are acceptabic.

5). "G t n c- ' lc cirics' in Generic Relo id Application for S x 8 Fuel,"

NED3-' Eevi sion 1, Novenber, 1974.

6). "Generc4 cetric BWR Thermal Basis (CETAB): Data, Correlation and Design Application," NEDD-10P58, 73NED9, Class I, November ,1973.

7). " General Electric EC Reload No. 3 Licensing Submittal for Dresden Unit 3," NFDD-20094, Decenber, 197.'.

8). " Process Computer Performance Evaluation Accuracy," and Amendment 1, NEDD-20340 and NEDD-20340 1,- dated June, 1974 and December, 1974.

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4-The st eaniline beca1. accident j (by rcference to Quad Cities 1 analysiy)s t 2 presented by the is acceptabic licensee based on car I generic review of NEDO-20360. l

2. Technical Specification Chances to Innlement Conformance to

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The proposed Limiting Conditions of Operation present two limitatiens en pc cr dist ribution relat ed to the LOCA analysis. These are the liniting  :

assctbly MiPIGIIR and MCPR. The MCPR value used in the LOCA analysis was l 1.16 and this value is less than the value dctermined from the transient analysis which has been incorporated in the proposed lechnical Specificatiorn. [

The bases for establishing the limiting value of MAPLhGK are 2ndicated  !

al ove in Section 2.0. A.I . - i t

1he licensee did not include the coualiter line area in the LOCA analysis, l therefore, the Technical Specifications wi11 ' require that the equalizer l line valves remain closed at all tines during reactor operation. The ,

LOCA analysis did not ad 1ress one loop operation, therefore the Technical i Specifications will not allow continuous operation with one. loop otn of service.

The LOCA analysi s assumed a)) Aut onatic Depressuri:ation Syst er (ADS) valves operM t e for sna11 line breal:s t-it h HTC) failure. Therefo;;, the i Technical Speci fications vill not perrit continuous operation with an,.  !

AliS valve out 01 service except as with other LCCS equipment one valve '

tay be out of service for 7 days.

3. Conclusi cns Re""rdine Conformance to a)) Recuirements of j Appendix E t o 10 Ci d 50 l I

On the basis c: our review of the information prov2ded by the 12censee t

{ for Montice]Io, we conclude that the sniety analyses are acceptable with l l respect t o confornance with all requirements of paragraph 50.46 of ,

i 10 CFl: Part 50 after the referenced MAPLilGR and MCPR technical

specification changes are incorporated. 1

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' 4). Status Report on the Licensing lepical Report " General Electric  !

Boiling h'ater Reactor Generic Reload Application for S x 8 Fuel," '

NED3-20360, Revision 1 and Supplement 1 by Divisien of Technical  !

i Review, Office of Nuclear Reactor Regulation, U. S. Nuclear llegulat ory Cont.issien, April,1975. .

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Thead{ionalanalyses (performed on the lead plant , Quad Cities Unit No. 2 and incorporated by reference) supported the earlier submittal which concluded that the worst break was complete severence of the recirculation line. These additional calculations provided further details with regard to the limiting location and size of break as well as the worst single failure for the Monticello design. The limiting break continues to be the complete severence of the recirculation line assuming a failure of the LPC) injection valve.

We have reviewed the e aluation of ECCS performance submitted by Northern States Power Company for Mont cello and conclude that the evaluation has i

been performed wholly in conformance with requirements of 10 CFR 50.46 (a).

Therefore, operation of the reactor would meet the requirements of 10 CFR 50.46 provided that operation is limited to the maximum average planar linear heat generation rates (tMPLHGR) of figures 3.11.1-A, 3.11.1-B, 3.11.1-C, 3.11.1-D and 3.11.g~ F of the Northern States Power Company 1 citer dated August 4, 1975 3, and to a minimum critical power ratio (MCPR) greater than 1.1S.

Ilowever, certain changes must he made to the proposed Technical Specificat ions t o conform with the evaluation of ECCS performance. 1he j larp ~ t recircu: .iira I-cal crca assumed in the evaluat ion uns T. r eon e f 'et . 1his break size is based cn operation with a closed valve in the i equ li: t: line betre:n t he two rceirculat ion loops. Therefore, thc

'iechnicc) Spe c i fi e r.t i e ns have been :..odified to limit reactor operation for a peri.od not t o exceed 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> unless the valve in the equalizer l line is closed. This change was discussed with and found acceptable by l the licensec.

The ICCS perfornance analysis assumed that reactor operation will be limitcJ to a MCPR of 1.18. liowever, a more restrictive technicc1 specification limits operation of the reactor to a MCPR of 1.33 for 7 x 7 fuel and 1.41 for S x 8 fuel based on consideration of a turbine trip transient with failure of bypass valves.

The Technical Specifications have been modified to require the licensce to report as a reportr51e cccurrence, operation in excess of the limiting MAPLH3R values even if corrective action was taken upon discovery. The change was discussed with and found acceptable by the licensee.

An evaluat ion was not provided for ECCS performance during reactor eperatica : it h one recirculatien locp cut of service. Therefore, continuous operation in excess of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> under such condition will not be permitted until the necessary analyses have been performed, evaluated and determined acceptable.

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I Cn July 9, 1975 the licensee gibritted an evaluation of the LCCS perforrance for Panticc110. An amendncnt requesting changes to the I lechnical Specification: forMonticellotoinp3pgynt The the results of l the evaluat ion wn sP it t ed on Au;,ust 4, 1975. licensec  ;

incorporated further infermation relating to the details of the I{fS on )

evaluation by re6 renec to the Quad Cities Unit No. 2 submittal ECCS evaluation as an appropriat e lead plant analysi s to shoe: conpliance to t he 3 0 CFR 50.46 crit eria and Appendix K t o 10 CFR Part 50. The Ordt' 20: lbdific at jun of hicense issued Dece:..ber 27, 1974, stated that evn]u:. tion of I CLS cooling perforrance may be. based on the vendor's evalu"t ion r.odel as modified in accordance with the changes described in the staff Saf et y i valuation Report of the Monticello Nuclear Generating Plant dated Decenber 27, 1974.

The br ci. ground of t he st aff revier of t he General Electric (CE) ECCS codelf . . I their application's to Ibnticello is described in the Staff Safct y Fra]unt j en Report (SER) for that facilit y dated December 27, 1974 issued in connection with the Order. The bases for acceptance of the principal portions of the evaluation model are set forth in the staff's Sirius Pep-t of October 1974 and t he Supplement to the Status Report of

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• iF d > ch are w fe?cnced an th: Decer,her 27, 1974 SER. The It o . 1: A c ri M , IL v. . m c;.anges ru g: ire? in t h<

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a i Topether the Decembe 27, 1974 SER nnd the 0 :t m Etpo.: <

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ra r t able ECCS eva]uat ion n del ,!th F: sas for 1 st aff's accept ance of the model . 1he

!% i c < ' l a t 1.]u :t ion whi t i, is covered by this S1R properly conforme to the accepted radel.

1 iiit h re> p ct ic refle.C :n] refill co: pat : t ions , the Monticello r.nalysis 1 cs based on the r olified version of the SAi:E computer code, with l explicit consideration of t he staff recommended limitations, as described in the Decenber 27, 1974 SFL The Monticello evaluation did not attempt ,

t o include any furt her creJit for other potential changes which the  !

Dectnher 27, 1971 SLR indicat ed were under consideration by GE at that t im e.

During the course of our review, we concluded that additional individual l breat sires should be analy:cd to substantiate the breal spectrum curves submit t ed in connection with the evaluation provided in August 1974. We als requested ' hat other breat locations be studied to sub,+antiate that the lir.itin.c break location was the recirculation line. '

l 1). Monticello Nue) ear Power Station IOCA Analyses Conformance with 10 CFR 50 Appendix K (Jet Fe:tp Pl an t ) , J u1 y~ , 1975.

2J. License Amend::.ent Request Dated August 4, 1975, Monticello Kuclear Generating Plant.

3). Qaad Cities Unit 2, Specini Report No.15, Supplement C, April 8,1975, Aprj] 21, 1975 (proprietary), and July 21,1975 (non-proprietary version of April 21, 1975 filing).

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UNIT E D STATES flUC LE /.R R E G U L AT O R Y - CC h".11SSION W ALHIN G T O N, D. C. 205LL l

SAFETY ]TALUATION EY THE OFFICE OF NUCLEAR REACTOR REGULiTIO."

SUPPORTING A'1END'IENT NO. 14 TO FACILITY OFERATING LICENSE NO. DPR-22 (CIDNGE ';D. 22 TO THE TECHNICAL SPECIFICATIONS)

NORTHER'; STATES P01GR C051PANY MUNTICEL LO NUCLE /d GENERATING PIXT DOCEET NO. 50-263 1.0 I NTR0!'UCTI ON Northern St at es l'ouer Cc pny (NSP) has proposed to operate the Monticello Nac1 c ::r Cc erating P1nni teith additional 8 x S fuel assc=blies as requested in its application dated August 4, 1975, using:

(1 ) Modific d operat ing limits based on an acceptable crerrency core cooling syst e:: (ECCS) evaluation model that conforms vith Section

50. 4 6 of 10 (: ' int '" as request ed in NSP'.. aj plicat ion dat ed

/suru + :, 19 m, .4 1 supportive filings dated August 20, 1974, July 9, 1975 .:d Eept < o 16, 1975.

(2) Ope; a ang iinit r basc d on the General 1.lectric Thernal Analysis Ensis (GEI Al') a s requested in NFP's application dated Narch 12, 1975 and rupf o  :.t dat c.. J al) 10, 19 ;.

Since propased changes No. 3 and 4 as described in the March 12, 1975 application are not directly related to GETAB or ECCS analysis, they will be considered at a later dat e.

2.O EVAL 14T10N A. Emergency Core Coolinn Syst ces

1. Conformance t o all Recuircr'ent s of Appendix E t o 10 CFR 50 On Ivc e. ' c: 27, 1971, the istwric Encrgy Coc:nission issued an Order for MnJificat den of License inplementing the requirements of 10 CFR 50.46,

".,cc ept an.ce Crit eria and Emergency Core Cooling Syst er.s for Eight Kat er

,acicar Power Reactors." One of the requirements of the Order was that

". . .the lic cusee shall subnit a reevaluation of ECCS cooling performance calculat ed in accordance with an acceptable evaluation model which conforms wit h the provisions of 10 CFR Part 50, E50.46." The Order also required that the evaluation shall be accompanied by such proposed changes in

'lechnical Spe c i fi c a t i ons as may be necessary to impicment the evaluation  ;

results. ,

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