ML20086C866
| ML20086C866 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 11/14/1991 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20086C863 | List: |
| References | |
| NUDOCS 9111250128 | |
| Download: ML20086C866 (5) | |
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NUCLE AR REGULATORY CC. MISSION 5
/. t W ASHINGToN, D. C. 20$$6 et SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.154 TO FACILITY OFERATING LICENSE NO. DPR-51 ENTERGY OPERATIONS. INC.
$ KANSAS NUCLEAR ONE. UNIT NO.1 DOCKET NO. 50-313
1.0 INTRODUCTION
Ey letter dated September 20, 1990, as supplemented by it.tters dated February 28, and August 14, 1991 Entergy Operations, Inc. (the licensee) submitted a request for changes to the Arkansas Nuclear One, Unit No. 1 (Ah0-1) Technical Specification (TS). The requested changes woulo revise the reactor coolent system TS pressure / temperature (P/T) operating limits for the first 15 ef f ective full power yeurs (EFPYs), using the methodology of F.egulatcry Guide 1.99, Revision 2.
The proposed amenduent would also revise the low-tet>perature overpressure protection (LTOP) enable temperature.
The February 28, and August 14, 1991, letters proviced clarifyir.t informtion that aid not change the ir.itial proposeo no significant hazards consideration ceterminution.
2.0 EVALUATION 2.1 Frerosed Fast Neutron Fluence The determination of the reactor coolant pressure boundan raaterial strength is required to comply with the provisions of Appendix G to 10 CFR Part 50.
Analyses for the P/T limits for 15 EFPYs for ANO-1 are cescribed in CAW-2106 by B&W Nuclear Service Corapany. The nuaximm inside pressure vessel exposure for 15 EFPYs is 0.488 E19 n/cm' and was estitaated in BAW-2075, Revision 1.
The nethodology ir EAW-2075 is based on BAW-1485, which is under NRC review.
However, the NFst staff has separately reviewed BAW-2075, Revision 1, for the specific AN1-C capsule analysis and results.
The transport calculations were
. carried out with the DOT-4.3 ctaputer code using an S angular quadrature and aP scattering ap?roxination. The calculation is ba$ed on the CASK cross seclion set with w 9tch DOT 4.3 has been benchmarked; thus, the calculation is acceptable. The dosimeters used ENDF/B-V based cross sections. A set of reasos.6ble uncertainies has been used for the estimetion of the fluence.
Therefore, the staff finds the information in BAW-2075, Revision 1, adequate to accept the fluence estimate for 15 EFPYs.
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2 2.2 Proposed PT__ Limits The proposed P/T limits are valid for 15 EFPYs. The proposed P/T limits were developed using Regulatory Guide (RG) 1.99, Rev. 2.
Generic Letter 88-11 reconrends that RG 1.99, Rev. 2, be used in calculating P/T limits, unless the use of different methods can be justified.
The P/T limits provide for the cperation of the reactor coolant system during heatup, cooldown, criticality, and hydrctest.
To evaluate the P/1 liinits, the staff uses the following NRC regalations and guidance: Appendices G and H of 10 CFR Part 50; the AS1H Standards and the ASME Code, which are referenced in Appendices G and H; 10 CFR $0.36(c)(2);
RG 1.99, Rtv. 2; Standard Review Plan (SRP) Section 5.3.2; and Generic Letter E8-11.
Each lices.ste authorized to operate a nuclear pcwer reactor is required by 10 CFR 50.36 to provide TSs for the operation of the plant.
In particular, 10 CFR EC.?6(c)(2) requires that limiting conditions of operation be includtd in the TSs.
The P/1 limits are anong the limiting conditions of operation in the TSs for all comerciel nuclear plants in the U.S.
Appendices G and H of 10 CFR Part 50 describe specific requireme nts for f racture toughness and reactor vt.ssel material surveillance that must be considereo in setting P/T limits.
An accepteble method for constructing the P/T limits is described in SRP Section 5.3.2.
Appendix G of 10 CFR Part 50 specifies fracture toughness and testinc requirerents for reactor vessel materials in accuraance with the ASHL Code and, in particular, specifits that the beltline materials in the surveillance capsules be tested in accorcance with Appendix H of 10 CFR Part 50. Appendix H, in turn, refers to ASTM Standards. These tests define the extent of vessel embrittlement at the tire of capsule withdravtal in tenns of the increase in reference terperature. Appendix G also requires the licensee tu predict the ef fects of neutron irradiation on vessel embrittlement by calculating the adjusted referente temperature (ART) and Charpy upper shelf energy (USE).
Generic Letter EC-11 requested that licensees and permittees use the methods in materials.ev. 2. to predict the effect of neutron irradittion on reactor vessel RG 1.99 R Th.s guide defines the ART as the sum of unirradiated reference-temperature, the increase in reference tergerature resulting from neutron irradiation, and a margin to account for uncertainties in the prediction rethod.
Appendix H of 10 CFR Part 50 requires the licensee to establish a surveillance program to periodically withdraw surveillance capsules frou the reactor vessel.
Appendix H refers to the ASTM Stanoards, which, in turn, require that the capsules be installed in the vessel before startup and that they contain test specimens made from plate, weld, and heat-affected-rone (HAZ) materials of the reactor beltline.
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3 The staff evaluated the ef fect of neutron irradiatier, enbrittlecent on each beltlit.e n6ateriL1 in the ANO-1 reactor vessel.
The an. cont of irradi6 tion enbrittlenefit was calculatec in accordance with RG 1.99, Rev. 2.
The staff has deterr; ired that the rieterial with the highest ART at 15 FFPYs was weld WF-18 with 0.29', copper (Cu), 0.55f nickel (Ni), ano ar initial reference tenperature (RTnot) nf -0 F.
The licensee has removed four surveillance capsules from ANO-1. The results fron, capsules E, B, A, and C were publishea in Babcock and Wilcox Roports LAK-1440, BAW-1698, BAK-1636, and BAW-2075, respectively. Surve111hr.ce capsules E, A, ono C cor.teined Charpy in.pect speciniens and tensile specimens rade frota be se n.ctal, weld n.etel, and HAZ metal. Surveil 16nce capsule B contained crily base metal anc HAZ metal sarrples. These surveillance data have bees, usso wherever uppliceble in calculatir,9 ART for beltlire traterials.
For the liruiting beltline t..aterial, welo WF-18, the staff calculated the AF.T to be 175.2* F at 1/dT (T = reactor vessel belt 11r4 thickness) and 133.9'F for 3/4T 2
at 15 EFpYs. The sttff used a neutrcr. fluence of 4.88E1E n/cm, which reducec to 2.94E18 n/tn/ at 1/4T and 1.06E18 n/cnf at 3/41. The ART was determined by using Section 1 ci RG 1.99, Rev. 2, because weld WF-16 was not in the survtillance capsules.
The 11censee used the method it, h61.99, Rev. 2, to calculate an ART of 183'T dt 15 EFFis at 1/4T for the liriitirig beltlir e welo. This value reduced to 173'F wher tre extra margin c1410'F ( 5%) was reroved.
(This extro conservatism wts iner.tichea in the 11cer.see's response to the sthff S request for additional information.) The staff )ucps that a difference ci 2.2'F between the licensee's ART of IW F (183*F-10*F) arid the staff's ART of 175.2'F is acceptable. Substituting the ART cf 175.2'F it.to equations in SRP 5.3.2, the stLf f verified ttti the proposeo F/T 11raits for heatop, couldown, ono hydrotest tret the beit11r4 r..aterial requirenents in Appenoix C cf 10 CFR part 50.
In acc..iun to beltlir.e t.iaterials, Arper.cix G of 10 CFR Far t 50 in.puses p/1 lin,1t., basec ct, the reference ten.perature 1or the reactor vessel closure flunse r.. ate ri e. l s.Section IV.2 of Apper dix ; states thut aben the pressure exceeds 2Cf of the preservice system hycrostatic test pressure, the temperature of the closure flange regions highiy stressed by the bolt preloao ruust exceed the reference teraperatuit cf the material in those regions by at least 120'F for ricrc.61 operation and by 90* F for todroststic pressure tests ar.o leak tests.
Besed on the tier.re reference teriperature of 10*F and the preservice systera hycrostutic test pressure of 317E psi (1.25x2500 psi), the staff has determined that the correspot.cir.g ten.perature of 220*F f roci the proposed F/1 limits I
satisfies Secticr. IV.2 of Appendix G.
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. The staf f fir.d5 that the propused P/T limits for the reactor coolant system for heatua, coolcown,1(at test, and criticality are valid ttrough 15 EFPYs because t1e lir.its conforra to the requirements of Appendices G ano H of 10 CFR Part 50. 1he licensee's submittal also setisfies Generic Letter 06-11 because the licensee used the n.ethod in RG 1.99, Rev. 2, to calculate the ART.
Hence, the proposed P/T li Lits may be incorporated into the AfiO-1 TSs.
Although the subject was net discussea by the lieensee in the sutr,ittal, the staf f notes that Section IV. A of Appendix C requires that the precicteo Charpy USE et ena-ct-lif e (EOL) be above 50 f t-lb, unless it is demonstrated in a nanner approveo by the director, Office of Nuc1 car Reactor Regulation, that lower values of Lil will provide n.argins (>f safety egtirist f racture equivalent to those requirtc by Appendix G of the A5P.E Cose. 1he material with the lucest Unirrachted USE and the highest Cu content is weld WF-18 with 0.29% Cu.
Since sut veillence data is not av6ilable f or this ty re of weld, the staff used figure 2 in RG 1.99 Rev. i directly and celtulated tie EOL USE at 1/4T (fluenct of 4
4.EPE18 n/crf) to be 44.9 f t-lb, which is below the reqLirec 50 f t-lb.
It utisfy the rt.quirements of Section V.C of Appendi> G, the licensee has tc propose a progt urs at le6st 3 years prior to the dote when the predicted frecture tought.ess 1 mis will no longer utisfy the requirer,ents of Secticr.
Y.B of this Appendix. Since f urther action is required, the staf f recuests ttet, within 60 hys of receipt oi th13 safety evaluation, the licensee respond r egarding its plcr.! to address this 1ssue.
2.2 Frcposed LTOP Limits Standard review Fian (SRP) 5.2.2 recemmends that LTOF P/T limits intet the frecture techcnics ciiteria in Appendix G to the ASt'E Code Section III.
However, the proposed Li0F lin.its were calculateo using the non-Appendix G eriteria that producec liinits less restrictive than the. limits celculated by the Apper.cix C ctittria.
Generic Letter LS-11, *NRC f csition on Radietion En.brittit;rner.t cf Reactor Vessel Pcterials ar.o its it.pect on Plant Operations," states that If changes con be it.plen,ented to show that the f requency of an LTCF event that would exceea Appenoix 611ruits is expected to be niuch less than one per reector lif etirae, then the staff would consider alternatius to Appendix G LTOP setpoints." The alternatives allow licensees to establish LT0F P/T linits that are hister (less restrictive) than the Appendix G F/T limits.
In the piopesed LTOP li.cits the licensee ccosidered that 1) an L10P event has not occurred in over 100 years of the L&W nuclear plants operating experience and is therefore not an anticipated operatioral occurence; ano
- 2) that the reettor coolant system in Ah0-1, a LD: reactor, has the benefit of the nitrogen /steem bubble in the pressurizer.
Such provision reduces the occurence of an LTOP by enabling the optrator to reduce the reactor coolont syster" pressure within an allowable time.
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' Considerir.g the operating experience at ANO-1 ar.d other B&W plants, the staff judges that the Ah0-1 LTOP lignits inay be calculated by using non-Apt er. dix G criteria. The L10P lin.its at Rarcho Seco were calculateo by using nor.-Appendix G criteria, and the staff appruved those lirrits on a plant-specific basis. The staff has evaluated the proposed AN0-1 L10P limits using the sana: criterie used in evaluating the Rancio Seco LTOP limits. The staff cor.cludes that the proposed ANO-1 LTOP limits will arovide adequate protection fo'r the reactor vessel up to the requested EFPYs.
lence, the proposed LTOP limits rnay be incorpurated in the ANO-1 TSs.
2.0 STATE C0hSULTAT10h In accordance with the Commissior.'s regulations, the Arkansas State of ficial was notifiec of the proposed issuance of the anendr.:er:t. The State official had r.e conrnents.
4.0 Et31koht<EhTAL C0hSIDERAT10h The anenoment changes 6 rcqcirement with respect to iristallation or use of a fccility comper.u.t locateo within the restricted area as defined in 10 Cfr.
Part :0 and changes ir surveillance requirements. The hPC staff has deterniined that the an.e:Aent involves no significant increase in the arcunts, and no significert change in the types, of any effluents that may be releaseo cfisite and that there is no signifieent increase in it.dividual or cunclative occupational redi6tior exposure. The Ccmission has previously issued a prcrosec finding that the amendment it. solves no significant hazards consideratic.n, and there hos bar. re public content on such finding (56 FR 8 00,'. Accordirigly, the sendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no envircraental impact statement or environc. ental ast.essn.ent need be prepored in ccnnection with the issuance of the amendrent.
5.0 C03_CLUSION 0
lhe Ccerission had conclucec, based on the considerations oiscussed above, that:
(1) there is reasonable assurance that the health and saf ety of the public will r.ot be endangerec Ly creration in the proposed tr.anner (2) such activities will be ccocucted in compliance with the Cons.ission's regulations, and (3) the issuar.ce of the amendrent will not be iniraical to the coiron defense and i
security or to the health and safety of the public.
Principal Ccntributors:
S. Sheng, Materials anc Chemical Engir.eering Branch J. Tsao, Materials and Chemical Engineering Branch L. Lois, Reactor Systen.s Branch Date: November 14, 1991