ML20155D031

Reload Info & SAR for Oyster Creek Cycle 12 Reload

ML20155D031
Person / Time
Site:	Oyster Creek
Issue date:	08/30/1988
From:	Alammar M, Dougher J, Walsh P GENERAL PUBLIC UTILITIES CORP.
To:
Shared Package
ML20155D014	List: ML20155D010 ML20155D015 ML20155D019 ML20155D031
References
TR-049, TR-49, NUDOCS 8810110036
Download: ML20155D031 (6)

Text

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TR 049 i Rev. I  !

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RELOAD INFORMATION AND ,

SAFETY ANALYSIS REPORT ,

FOR OYSiER CREEK CYCLE 12 RELOAD TR-049 Rev. 1 i

BA No. 335400 H. A. ALAMMAR J. O. DOUGHER AUGUST, 1988 l

APPROVALS:

I v & b AS t'!/9 28 Manager, Oyster Crefk Fuel Projects '

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Oh Nuclear Analysis t, fuels Director 9l/lb

' ~01 GPU NUCLEAR l Upper Pond Road l Parsippany, New Jersey 07054 '

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. . 1 TR No. 049 i Rev. I i Page 6 of 47 l 1

1.0 INTRODUCTION

AND

SUMMARY

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This report justifies the operation of the Oyster Creek Nuclear J Generating Station through the upcoming fuel reload Cycle 12. It is planned to operate the Oyster Creek reactor *. Cycle 12 beginning in l December 1988 with a partial core loading of fresh P8X8R and GE8X8EB fuel I bundles supplied by General Electric Company (GE). l The Cycle 12 reference loading pattern is designed to ensure compliance with Technical Specification limits and safety analysis criteria. The j reload dependent analyses are performed with methodology developed by GPUN. These methods were previously submitted to the N'<C for approval (References 1 through 7). The loss-of-coolant accident, rod drop ace.ident and stability analyses were performed by the fuel vendor using previously approved methods.

The GE8X8E8 fuel design vill be introduced for the first time into the 1

1 Oyster Creek core for Cycle 12. This fuel has been designed to accomrnodate higher enrichments, longer fuel cycles and will reduce or 1

eliminato PCI related fuel failures. The Cycle 12 fuel design has an l average enrichment of 3.21% and is described ir Appendix B of Reference 9.

l The transient and accident analyses presented in this report dencistrate I

i that based on the Turbine Trip Without Bypass transient, the Technical l l

Specification CPR operating limit will have to be raised free 1.45 to ,

1 1.51 for Cycle 12. The safety limit HCPR is determined using the General l l

1 l Electric Ccepany Thermal Analysis Basis, GETAB' with the GE critical {

! quality (X) boiling length (L) GExt, correlation.

l 3100C l l l l

TR No. 049 Rev. 1 Page 31 of 47 The second analysts uses a transient rod location that will  ;

result in a poor response of the APRM system and also has a j high rod worth. The most limiting results for Cycle 12 were ,

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obtained in the highest worth control rod analysts. Figure 5.8 provides the control rod pattern used in this analysis, j i

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The APRM response, and hance the rod block effectiveness, f versus transient rod position will vary based upon the number f of available LPRMs feeding the Al'RM. Three APRM status i conditions have been defined and the APRM response to control I

rod withdrawal is displayed in Figure 5.9. The results of l the most limiting APRM response (Status 1) are shown in Table 1

5.3. Since the RWE is not the 11alting CPR transient for Cycle 12 all three APRM status conditions will have the same I I

operating CPR limit (1.51). The peak MLLHGR remains well within limits.

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TR No. 049 <

Rev. 1 Page 44 of 47 6.0 OPERATING LIMIT MCPR t The operating limit MCPR for each transient is presented in Table 6.1 and 5 that the limiting event is the Turbine Trip Without Bypass. The required ;

HCPR operating limit for Cycle 12 is 1.51 based on a safety limit of 1.07, a ACPR of 0.37, and a statistical multiplier of 1.049. {

l 7.0 STABILITY ANALYSIS According to Reference 17, Oyster Creek (as a low power density BWR/2) is {

l exemptfromthecurrentrequirementtosubmitacyjlespecificstability  ;

analysis for its reload fuel. Ample stability margins to the 1.0 decay ratto criteria, as shown in the stability analysis for Cycle 10',

are typical for tne Oyster Creek Plant.

l 8.0 TECHNICAL SPECIFICATIONS f Based on the Cycle 12 reference core design and safety analysis provided  ;

in this report, the following sections in the Technical Specifications l will require modification.  !

Section 3.10.A (Average Planar LHGR): Add new limits for GE8X8EB fuel and revite limits for P8X8R fuel designs. Four and five loop operation .

l will use same MAPLHGP figures, f

Section 3.10.B (Local LHGR): Add reference for new fuel design (GE8X8EB) tc include LHGR limit of 1 13.4 KW/ft.

Section 3.10.C (Minimum Critical Power Ratio): Change MCPR LiNit from j 1.45 to 1,51 for each of the three APRM status levels.

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! The appropriate bases sections will also require modification.

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r-TR No. 049 Rev. 1 Page 45 of 47 TABLE 6.1

- Cycle Operating Limit MCPRs Transient MCPR fuel Loading Error (Hislocated) 1.28 fuel Loading Error (Misorientated) 1.32 Loss of Feedwater Heating 1.20 Rod Withdrawal Error 1.45 FH Controller Failure 1.40 Turbine Trip w/o Bypass 1.51 f

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. . . . 1 TR No. 049 Rev. 1 Page 46 of 47

9.0 REFERENCES

l 1.0 H. Fu, R. V. Furia, "Hethods for the Analysis of Bolling Water Ractors Lattice Physics," TR 020-A, Rev. O, January 1988. j 2.0 Letter from J. N. Donohew, Jr. (NRC) to P. B. Fiedler (GPUN) dated l November 14, 1986, "Reload Topical Report TR 020 (TAC 60339)."

l 3.0 R. V. Furia "Hethods for the Analysts of Boiling Water Reactors Steady State Physics," TR 021-A, Rev. O, January 1988.  ;

4.0 Lstter from A. W. Dromerick (NRC) to P. B. Fledler (GPUN) dated [

4 September 27, 1987, GPU Nuclear Corporation (GPUN) Topical Report l 3 TR 021, Revision 0, "Hethods for the Analysis of Bolling Water (

j Reactors Steady State Physics." '

l 5.0 0. E. Cabrilla, et al., "Hethods for the Generation of Core f

] Kinetics Data for RETRAN-02," TR 033A, Rev. O, May 1988.  ;

6.0 E. R. Bujtas, et al., "Steady-State and Quast-Steady-State Methods i Used in the Analysis of Accidents and Transtents," TR 040A, Rev. O, l

May 1983.

l 7.0 H. A. Alammar, et al., "BWR-2 Transient Analysis Model Using the  !

RETRAN Code," TR 045, Rev. O. September 3, 1987. ,

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8.0 Letter from H. Berkow (NRC) to J. S. Charnley (GE) dated .

December 3, 1985 "Acceptance for Approval of Fuel Designs i 1 Described in Licensing Topical Report NEDE-24011-P-A-6, Amendment l 10 for Extended Burnup Operation."

9.0 "0yster Creek Nuclear Generating Station SAFER /CORECOOL/GES1R-LOCA f Loss-of-Coolant Accident Analysis," NEDC-31462P, August 1987.  !

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10.0 "General Electric Reload Fuel Applica..on for Oyster Creek,"  !

l NE00-24195, (As Amended).  !

l i 11.0 "General Electric Pandard Application for Reactor Fuel," l j NEDE-24011-P-A-8, May 1986, j 12.0 "General Electric BWR Thermal Analysis Basis (GETAB): Data, f j Correlation and Design Application," NE00-10958-P-A, January 1977. j 13.0 "Banked Position Withdrawal Sequence," NE00-21231, January 1977.

14.0 "Guidelines for Generating OPL-3 Inputs," NEDE-22061, Feb. 1982.

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15.0 Letter from H. A. Paulson (NRC) to P. B. Fiedler (GPUN), dated  !

August 27, 1984, "Core 10 Refueling." l b

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