ML20205T082

Safety Evaluation Supporting Amend 129 to License DPR-16

ML20205T082
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Site:	Oyster Creek
Issue date:	10/31/1988
From:	Office of Nuclear Reactor Regulation
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NUDOCS 8811140107
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UNITED STATES NUCLEAR REGULATORY COMMISSION n

WASHIN'1 TON. D. C. 20665 4 *...*

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 129 TO PROVISIONAL OPERATING LICENSE NO. OPR-16 GPU NUCLEAR CORPORATION AND JErSEFtlTf)tWPDWtTTitift@PANY OYSTER CREEK NUCLEAR GENERATING STATION DOCKET NO. 50,,2,19 INTRODUCTION By letters dated Varch 30, 1988. April 12, 1988 and September 22, 19P,P. (Ref. 1),

GPU Nuclear Corporation (GPUN) made application to amend to Technical Specifica-tions of the Provisional' Operat.ing License DPR-16 for the Oyster Creek Plant in order to operate for Cycle 12.

In support of this application, the licensee also provided a reload analysis submittal (Ref. 2) and supporting analyaes (Ref. 10).

The reload application involves 3 fuel-design related issues:

(1) the replacerrent of 172 spent fuel asserblies with 20 General Elettric P8XSR and 152 GE8xfEB (extended burnup) fuel assemblies, (2) the analys.'s of safety considerations involved in the determination of Cycle 12 operating limits, and (3) the incorporation of new and extended maximum average pianar linear heat generation rate (MAPLHR) limits. The NRC staff has reviewed these subnittals as follows.

2.0 FUEL DESIGN The Oyster Creek Cycle 12 core (see Table 1 for the fuel inventory) will retain 28 Ixson Type YB assemblies and 360 General Electric (GE) P8xC.,

r assemblies from the previous cycle and add 20 unirradiated GEP8x8R fuels and 1

152 new GE8xBE8 fuels, which are 3.21 percent average U235 enriched fuel i

assemblies. The GE8x8EB fuel type was ap (Ref.13) for Amendment 10 to GESTAR II (proved in the Safety Evaluattan Ref.14) and has been used in l

many existing SE plants. The specific descriptions of this fuel were also included in Amensnent 18 to GESTAR !! (Ref.16) which was previously approved by NRC (Ref. 15). LOCA analyses have been dond for the retained and reload GE fuel using the SAFER /GESTAR-LOCA methods approved by the NRC (see Section 6.0).

Since the MAPLHGR values for the fuel assemblies have been calculated with approved methodology (GESTAR II, Reference 14. Section 2 of Volume 1) they are acceptable.

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The proposed average linear heat generation rate (LHGR) limit for the GE8x8EB fuel is 13.4 Kw/ft, which is less than the limit of 14.4 Kw/ft approved by the NRC for the Gi extended burnup fuel (Ref.13). The LFGP is, therefore, acceptable for the GE8x8EB fuel in the Oyster Creek Cycle 12 core.

3.0

,N,U, CLEAR DESIGN The nuclear design for Oystec Creek Cycle 12 has been perfonned by GPUN with the approved methodologies, which include fuel lattice methods (Refs 3 and 4) and 3-dimensional core steady state methods (Refs. 5 and 6).

The fuel lattice athods are used to calculate fuel bundle nuclear parameters such as reactivities, relative rod powers and 2 or 4 group cross sections. The 3-dimensional reactor code calculates power and exposure distributfor.s. core thennal-hydraulic cl.v-acteristics, and cold shutdown margin.

The results of the reload analyses are given in Reference 2.

The results are within the range of those usually encountered for BWR reloads.

In particular, the shutdown mar exposure of minimum shutdown margin thus, fully ngin is.0161 delta K at the eeting the required 0.01 delta K.

The Standby Liquid Control System also meets shutdown margin requirements with a shutdown margin of 0.034 delta K.

Since the Oyster Creek Cycle 12 nuclear desigr.

parareters have been obtained with previously approved methods and fall within expected ranges, the nuclear design is acceptable, 4.0 THEPW L AND,,H,Y0,RAU,L,1,C,,0E,51,G,N The objective of the review is to confirm that the thertal-hydraulic design of the core has been accorplished using acceptable methods, and that it prnvides an acceptable e,argin of safety fru conditions which could lead to fuel damage during transient conditions.

The review includes two areas:

safety limit n.inimum critical gwer (MCPR) and operating MCPR limits.

The licensee has submitted the analysis report for Cycle 12 operation at rated l

flew conditions (Ref. 2). Discussion of the review concerning the therr.al-hydraulic design for the Cycle 12 operation follows:

4.)

S a fe ty L i,mi,t, f,C,F,R, A safety limit MCPR has been imposed to assure that 99.9 percent of the fuel rods in the core are rot expected to experience boiling transition during operational transients. As stated in Reference 20, a safety limit MCPR of 1.04 was approved to be applied to the second successive reload core of P8x8R, BP8x8R, GE8x8A or GE8x8EB fuel designs with an initial bundle R factor greater than or eque.1 to 1.04.

To provide more safety margin, the safety limit of 1.07 is used by GPUN for Oyster Creek Cycle 12 reload analysis.

4.2

,0pera,tjng1.imitMCPR r

The most limiting events have been analyzed by the licensee to detennine which i

event could potentially induce the largest eduction in the initial critical power ratio (RCPR). The RCPR values given in Table 5.1 of Reference 2 are plant specific values calculated by the methods including NODE-B (Ref. 8) and RETRAN (Ref 9) methods, which were previously approved (Refs. 17 through 19) by NRC for the Oyster Creek reload applications. The turbine trip without bypass event was identified as the worst case with the largest RCPR of 0.37.

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3-The p.'eposed operating limit MCPR of 1.51, a more restrictive operating limit as compared to 1.45 for Cycle 11, was determined by adding the safety limit MCPR of 1.07 to a maximum RCPR of 0.37 with inclusion of statistical uncertainty factcr of 1.049. Since the approved methods were used to determine the operating limit MCPR to avoid violation of the safety limit MCPR in the event of any anticipated transients, we conclude that the proposed operating limit MCPR of 1.51 is acceptable for incorporation into the Technical Specifications.

5.0 TRANSIENT ANALYSES The transient analysis methodology used for the Oyster Creek Cycle 12 core described in References 8 and 9 were previously approved (Refs. 17, 18 and 19) by NRC. The CPM and NODE-B codes and methods de:cribed in Reference 8 were used to analyze the non-pressurization events including the loss of feedwater event. The RETRAN code and methods described in Re1erence 9 were used to analyze the. pressurization transients including the turbine trip without bypass, loss of feedwater heater, feedwater controller failure and main steam isolation valve closure without scram. The limiting MCPR event for the Oyster Creek Cycle 12 core is the turbine trip without bypass and is discussed in Section 4.0.

Compliance with overpressurization criteria was demonstrited by analysis of main steam isolation valve closure without credit for the first safety grade scram signals.

Maximum vessel pressure was 1305 psia, which is well under 110 percent (1390 psia) of design pressure.

Since bank position withdrawal sequence and rod pattern are used for Oyster Creek, a cycle specific control rod drop accident analysis is not required. The basis for this posit wn and NRC approval is presented in Amendment 9 to Reference 14 We find that the approved nethodologies and anolyti:al results for both the pressurization and non-pressurization events were used to show that the analytical results fall within the safety limit to achieve the fuel and pressure boundary integrity during transients, therefor *, we conclude the trtnsient analyses are acceptable.

6.0 LOCA ANALYSES ihe LOCA enalyses for the Oyster Creek Cycle 12 core were performed using the SAFER /COREC00L/GESTR LOCA methodology (Ref.11), which has been approved by the staff (Ref.12) and used and approved in several jet purp BWR reload applications E

(e.g., Duane Arnold Cycle 9. Quad Cities 1 Cycle 10).

In Reference 12, the staff has specified conditions for demonstrating applicability of the SAFER /COREC00L/

GESTR LOCA methodology. These conditions are:

1.

Calculation of a sufficient number of plant specific Peak Clad Temperature (PCT) points based on both nominal input values and Appendix K values to verify the shape of the PCT curves versus break size.

2.

Confirmation that plant specific operating parameters have been bounded by the models and input used in generic calculations.

The licensee has provided the resdits of thost analyses (Refs. I and 10) which are required to meet these conditions.

Specifically, the analyses include break sizes from 0.05 ft' to the design basis accident (DBA) red rculation

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discharge line break (4.66 ft ).

Six different break sizes were analyzed in 4

confunctionwithECCSfailurecombinations. A total of 15 cases were evaluated to establish the trend of PCT curves (for both the nominal and l

Appendix K conditions) versus break size.

The input parameters for be'.h the nominal and Appendix K cases are within those used in the approved generic analyses. The results show that the DBA recircula-tion line disciarge break with automatic depressurization system (ADS) valve failure is the limiting case. Tre calculated PCTs are 1831*F and 1714'F for low and high exposures, respectively, when nominal input values are used. The correspording PCTs for this break size with Appendix K input values were calcu-lated to be 2196'F and 2027'F for the low and high exposures respectively.

Because the approved methods were used in the analysis, the Input parameters and the cases analyzed to establish the trend of PCT versus break sizc meet the staff requirements given above, and the Appendix K SAFER /COREC00L results (which are.

i 2196*F and 2027'F) bound the generic upper limit PCTs (which are 2088'F and 1999'F.

and approved by NRC in Reference 11), we conclude that these analyses are acceptable.

l 7.0 TECHNICAL SPECIFICATION CHANGES Various changes to Technical Specification (TS) 3.10 have been proposed in order for GPUN to operate the Oyster Creek Plant, Cycle 12 core. These changes, l

indicated in the pro and 3.10-11 (Ref.1) posed 15 f rom pages 3.10-1 through 3.10.6, pages 3.10-10

, range from miscellaneous changes (i.e., adding new references, etc.) to new and extended MAPLHGR limits, and operating limit MCPR

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for Cycle 12 fuels.

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The Technical Specifications changes are:

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Section 3.10.A: Add new limits for GE8x8EB fuel and revise

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limits for P8x8R fuel design. The new and revised MAPLHGR will be applied to both four loop (with the inactive loop suction and discharged valves unisolated) and five loop operation, j

2.

Section 3.10.B: Add reference for new fuel design (GE8x8EB to include LHGR limit of equal to or less than 13.4 kw/ft.

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Section 3.10.C: Change operating limits MCPR from 1.45 to I

j 1.51 fo? each of the three APRM status levels.

Il We find that all of these TS changes reflect the characteristics of fuel, in Cycle 12 and are supported by the analytical results that demonstrate no violation to the fuel integrity acceptance criteria and the fuel performance acceptance criteria of 10 CFR 50.46, and therefore, we conclude the TS changes l

acceptable.

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8.0 CONCLUSION

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i We have reviewed the reports submitted for the Cycle 12 reload of Oyster Creek j

with GE fuel, and the GE methodology and analysis for LOCAs, and the GPUN methodology and analysis for transients.

Based on this review, we concluded i

s that appropriate material was submitted and that the fuel desig, nuclear design, thermal hydraulic design and transient and accident ana yses are acceptable. The TS changes submitted for this reload suitably reflect the use of acceptable methodologies. The operating limits associated with those changes and reload parameters are acceptable.

9.0 ENVIRONMENTAL CONSIDERATION

1 This amendment changes a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The staff has determined that the amendment involves nc significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no signif-icant increase in individual or cumulative occupational radiation exposure.

The Comission has previously issued a proposed finding that this amendment involves no significant hazards consideration and there has been ne public coment on such t in' ding.

Accordingly, the amendment meets the eli criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9)gibility Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of this amendment.

10.0 CONCLUSION

r The staff has concluded, based on the considerations discussed above, that (1) there is reasonab h assurance that the health and safety of the will not be endangered by operation in the proposed manner, and (2) public such i

activities will be conducted in compliance with the Cossnission's regulations, and the issuance of the amendment will not be inimical to the i

comon defense and security nor to the health and safety of the public.

11.0 REFERENCES

1.

Letters from R. F. Wilson (GPU) to Document Control Center of NRC, Oystor CreekNuclearGeneratingStation(DocketNo.50-219)-TechnicalSpecifica-tion change Request (TSCR) No. 166, dated March 30, 1988, April 12, 1988 and September 22, 1988.

2.

GPU TR-049 (Rev. 0), Reload Information and Safety Analysis Report for Dyster Creek Cycle 12 Reload, Narch 1988.

3.

GPU TR-020-A (Rev. 0) Methods for the Analysis of Bo!11ng Water Reactors Lattice Physics, January 1988.

4 LetterfromJ.Donohew(NRC) top.Fiedler(GPU),ResponseTopicalReport TR 020 (TAC 60339), Novesber 14, 1986.

5.

GPU-TR 021-A (Rev. 0), Methods for the Analysis of Boiling Water Reactors l

Steady State Physics, January 1988.

6.

Letter from A. Dromerick (NRC) to P. Fiedler (GPU), dated September 27, 1987.

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.GPU-TR 033 (Rev. 0), Methods for Generation of Core Kinetics Data for RETRAN-02, February 1987.

8.

GPU-TR 040 (Rev. 0), Steady-State and Quasi-Steady State Methods Used in the Analysis of Accidents and Transients, February 199.

9.

GPU-TR 045 (Rev. 0), BWR-2 Transient Analysis Model Using the RETRAN Code Septesber 1987.

10. NEDC-31462P and NEDC-31462, Oyster Creek Nuclear Generating Station SAFER /CORECOOL/GESTR-LOCA Loss-of-Coolant Accident Analysis, August 1987.

11.

MEDE-30996-P, Volume II, "SAFER Model for Evaluation of Loss-of-Coolant Accidents for Jet and Non-Jet Power Plant."

12.

LetterfromA.Thadani(NRC)toH.Pfefferlen(GE),datedMay 11, 1987, 13.

Letter from C. Thomas (NRC) to J. Charnley (GE), dated May 28, 1985.

14. GESTAR 11 NEDE-24011, Revision 8. "Genu al Electric Standard Application for Reactor Fuel."

15. Letter from A. Tt.adani (NRC) to J. Charnley (GE), dated May 12, 1988.

16. Letter from J. Charnley (NRC) to G. Leinas (NRC), "Proposed Amendment 18 to GE Licensing Topical Report NEDO-24011-P-A, dated October 31, 1986.

17.

LetterfromA.Dromerick(NRC) top.Siedler(GPU),datedMarch 21, 1988 (TACs65138and65139).

18. LetterfromA.Thadani(NRC) tor.Furia(GPU),datedOctober 15, 1988.

19. Letter from W. Hodges (NRC) J. Stolz ( W C), dated September 23, 1988,

20. LetterfromA.Thadani(NRC)toJ. Charnley (GE),datedDecember 27, 1987.

Dated:

October 31, 1988 Principal Contributor:

S. Sun i

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e TABLE 1 Oyster Creek Cycle 12 Fuel Bundles Fuel Type Average Exposure Number (GWD/MT)

Irridiated Exxon VC 17.96 28 P80R8239 18.33 112 P80RB265 16.30 64 i

P80RB299-7GZ2 10.45 136 P80RB299-7GZ1 8.76 48 New P8DRB-321(EB) 0.0 152 P80Rb299-7GZ2 0.0 20 Total "TEU

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