ML20151A610

Safety Evaluation Supporting Amend 26 to License DPR-3

ML20151A610
Person / Time
Site:	Yankee Rowe
Issue date:	06/02/1976
From:	Office of Nuclear Reactor Regulation
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ML20151A604	List: ML20151A602 ML20151A606 ML20151A610 ML20151A613
References
NUDOCS 8011030807
Download: ML20151A610 (5)

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UNITED STATES' g k NUCt. EAR REGULATORY COMMISSION

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SAFETY EW.LUATIO:I BY THE OFFICE OF NUCLEAR REACTOR PICULATIO::

SUPPORTING AMENDMENT NO. 26 TO FACILITY OPERATING LICENSE NO.' DPR-3 YANKEE ATOMIC ELECTRIC COMPANY YANKEE NUCLEAR POWER STATION- (YA2EE-ROWE)

DOCKET No. 50-29 Introduction By a lication dated February 20, 1976,(1)and supplement dated April 29, 1976 ) Yankee Atemic Electric Company (the licensee) requested an atendment to License No. DPR-3 for the Yankcc-Rowe reactor. The acendeent uould change the Core XII allowab3c peak linear heat generation rate in Figure 8-3 in the Technical Specifications.

- Discussion .

Presently, the licensee is authorized to operate Yankec-Rouc in the 4-loop mode within the limits for the Linear Keat Generation Rate (LHGR) in Figure 8-1 in the Technical Specifications. We have previously cencluded in our Safety Evaluation dated Dece=ber 4, 1975(3) that operation of Yankee-Rowe with Core XII within these limits assures compliance with 10 CFR Part 50,5 50.46. Our conclusion was based on the results of the licensee's Core XII ECCS perf ormance. analysis donc by the Exxon Nucicar Cocpany Inc.

(ENC) using the evaluation codel which we had previously approved for the H. B. Robinson f acility. (4) Specifically, with respect to the liniting break spectrum analyzed for Core XII (the largest doubic-ended cold leg 2uillotine (DECLG) break and the equivalent double-ended cold leg split (DECLS) break), we found that the licensee had provided acceptable bounding calculations, had identified the most limiting break and that the Technical Specification limits for the LHGR would acceptably restrict operation of Yankee-Rowe within the bounds of the analysis results.

However, we required the licensee to submit additional analysis to confirm that the trends predicted for the H. B. Robinson facility are appropriate far reference by Yankee-Rowc and that the DECLS break is the most limiting break size for Yankee-Rowe. The licensee's application for licence amendment (1)(2) provided the additional analysis in support of a proposed change of'the limits-for the LHGR in Figure 8-1 in the Technical Specifications. This safety evaluation su:carizes the results of our review of the licensee's proposal.

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Evaluation i.

1. ECCS Evaluation llodel

'The ECCS evaluation model for Yankee-Rowe described in the licensee's Februcry 20, 1976 submittal is identical to that which we have previously revleucd nnd approved (3). With respect to the nothod used for ccicuintin3 in'.c t rubcooling, uc hcve further invcctigated and detailsuhichhavebeenconfirmedbythelicensce:(grifiedthefolic.:ing

a. In calculating heat transfer to the ECCS water in the 3cwar plenum of the reactor vesael during the refill period, the condut:ivity for stain 3ena steel was used for the vcosel material. Sinty thic hent trnnaf;r is c6nduc.Livlty liuited nnd beccuse the thcr- .1 conductivity of nt tinicsc steel is significantly lo.rer th a tut of car'eca st ce) in the reactor ressel, uc find t he heat trans.N r calculations to be acceptably conserrcti ve.

b. In the cedsiylvi.ty studie i condu: ted en the subcoolin.:. efftces, the f2 ceding rate and tha carryove.r rate fraction were alle ad co vary as the inlet cubcool:'ng vas changed. Uc agrce with chis apprcach and find that ENC has properly identified and conservatively calculated the effects of inlet subcoolinc, .

In cur previous Safety Evaluation (3) we concluded that the E::C evaluaticn model u.ed for the Core ::II ECCS core cooling performance analy.is m ets the ree;utrcreats of 10 CFlf Part 50,G 50.46. Our conclusion rc mins unchanged with respect to the upc 'of the same model L r the add cie::a*

Core XII cnaly ds provided by the licensec with the February 20, 1975 appliention for license anendaicut.

2. Dreah Scectrun Prior to our approval of operation of Yankce-Rouc with Core XII, the licensee subnitted LOC.\ analyces censisting of two(b eaks for (1.0 x DECLG and 1.0 x DECLS)(b) . haa]onous calculationsdonc AJ the H. B. Robinson facility had identified these breaks as the most li uitin r..

We found this approach to be accep, tabic, but recuired that more bren:s be analyzed for the Y:nkee-Rcec reactor to confirm that the vorst breck had been identified. The licensec's additional analyses (l)have instead identified the 0.6 x DECLC as tha limiting break, but with tha peak clad ter..perature only 180F higher. In addition to the limiting 0.6 x DECLC break, the licensee hcs analyzed the 0.6 x DECLS brack, the 0.4 x DECL guillotine and split breaks, and the 1.0 x DECL r,uillot i n e and nolit breaks. Because th: lin ting break is net th laracst, is virnin the range of Mocuy u"Itipliers of 0.6 to 1.C, and is bruided ca hoth sides by less sever < brc kc, we have cencluda.:

that the break sg.c:trur.: rahmitted by the 11ctn:ce is acceptable.

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3. Containment ' Backpressure l i

A constant value of 11.8 psig has been used for containment backp'ressure.

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i in the core heatup calculations for Core XI!(5). This was an acceptable

-value based on calculations that we had accepted for the ECCS analysis of Core XI and a comparison of the mass and energy release to the i containment for the Core XI analysis and the Core XII analysis (3) . We l had compared this value with our independent calculation of the con-tainment pressure as a function of time based on Exxon's calculated mass and cnergy relcase data and the containment input data for the Yankee-E,we plant. This analysis demonstrated the value of 11.8 psig to be conservative to a time of 115 seconds after the accident.

The containment pressure analysis was conservative for the two breaks submitted previously(3}because the peak cladding temperature was _

calculated to occur before 115 seconds. However, the 0.6 x DECLS and l 0.4 x DECLS breaks yield peak cladding temperatures which occur significantly later than 115 seconds, and pressure has by this time  ;

dropped to 11.5 psig. -To account for this, the licensec has  !

analyzed the additional breaks (l) assuming a constant lower value of I 11.5 psig(2), thus retaining tha conservative margin' in the calculations.

We find that the licensec has acceptably accounted for the change of-containment bachpressure in the analysis of the additional breaks.

4. Technien1 Specifications and Sufficiency of Incore Monitoring -

Capability , .

The licensec has proposed (I)two curves of limiting linear heat generatica rate (LHGR) versus exposure to be, incorporated into the Technical Specifications. One of these curves is for the onec-burned Gulf fuel assemblics and the other is for the newer Exxon fuel assemblics. The curve for the Gulf , fuel is essentially unchanged from that presently included in the Technical Spe'cifications. HovcVer, the Exxon fuel, excep't for a brief period at the very beginning of the cycle while the pelict-cladding gap is decreasing, has a lower fission gas inventory, a higher gap conductance, lower stored energy, and therefore a higher permissibic LHCR than the Gulf fuel. From our review of the licensee's February 20, 1976, submittal, we find that the proposed changes implement the additional cooling performance evaluation results and provide acceptable limits for the safe operation of Yankee-Rowe with Core XII.

In order to take credit for the higher limiting LHGR for the Exxon fuel, the method used for monitoring maximum LHGR must be able to differentiate between the two types of fuel. Yankee-Rowe utilizes movable incore detectors t.o map the core power distribution at least monthly. Appropriate penaltics are included in the Technice'. Specifications to account for ,

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Xenon redistribution, control rod insertion, and other effects which perturb the power distribution between maps. The incore system is

, based on 90 decree rotational (not reflective) syenetry effectively providing inctrumentation in approximately half the essemblies. The instrumentation tube layout is such that all peripheral uninstrumented assemblics are adjacent to at least one instrumsn+.cd assembly (or symmetric equipment and all interior uninstrunented assembifes are adjacent to st least two opposed instrumented casomblics (or symmetric equivalents). Because of the unusually complete incore system of Yankec-Rowe ve find that this existing instrumentation is adequate to determine the m:::inum linear heat generation rate separately ior the Gulf and Exxon fuel.

5. Other Accidents and Transients The eininun D: 3 ratios predicted for the various abnormal operating transients all exceed 2.0 for decign hot channel conditions of 12.9 kw/ft and Pt n of 1.81(5). The prescut LOCA analyses (l)(3)show that r.n c limiting het channel conditiens will be lower than the design cond; tic.ns used for the thermal-hydraulic design. Therefore, the less-of-ecolant accident remains limiting.

We have detere.intd that the anendnent does not authorizn a change in ef fluent types or total caounta nor an inerccsc in poper level nr.d vill not result in any significant environmental impact. Having acdc this deternination, we have further cencluded that the amendment involves an action which is insignificant frc tha standpoint of environizanthi impact and percu,nt to 10 CFR S 51. 5(,d)(4) that en environmental stcte- .

ment, negative declaration, or environments.1 impact appraiscl need net be prepared in connection with the iscu~ ace of this cuende. ant.

Conclucien

• We have concluded, based on the considerations discussed above, that: i (1) bccause the change does not involve a significant increase in the l probability or censeqeences of accidents previously censidered and dcas l not involve a significant decrease in a safety margin, the change dess not i involve a significant hasards cor.siderations, (2) there is reasonabla l assurance that the health and safety of the public vill not be endangered I by operation in the proposed menner, and (3) such activities will

• e conducted in cecpliance uith the Commission's regulations and the issusnee of this amendment will not be inimical to the common defense and security or to the health and safety of the public.

Date: June 2,1976 l

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- - 5-Referen~ces '

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- Letter from W. R. Johnson (YAEC) to USNRC dated ' February 20, 1976, enclosing Appendix B, Additional.LOCA Analyses 2

Letter from L. !!. lleider (YAEC) to USNRC dated April 29, 1976 3.

Safety Evaluation Amendment No. by the Office Nucicar Reactor Regulation Supporting 21 to Facility Operating License No. DPR-3 (Change No.

126 to the Technical Specifications). Yankee Atomic Electric Company-Yankee Rowe CoreNuclear XII Reload)Power' Station (Yankee-Rowe)

- dated December 4, 1975. Docket No. 50-4.

Safety Evaluation Report Regarding Review of 'the' Exxon Nuclear Company PWR ECCS Codes and the 11. B. Robinson Reactor. ECCS Evaluation Model for Conformance to All Requirements of Appendix K to 10 CTR 50-by the Office of Nuclear Reactor Regulation, USNRC, September 11, 1975 5.

Letter from D. E. Vandenburgh (YAEC) to USNRC dated Nover.ber 26, 1975 enclosingtittachegn,t, sonolement No. 5 of Propened Chnnec i:o. 125,.

Proposed Inter _in Techni_eni Specifications, and Anpendi:* A. Loss of Coolant Accident Station (October 10, Anafenis 1975) for Core XII of the Yankee Nucicar Pcwer w

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