ML20134P962

Responds to RAI Re Removal of cycle-specific Parameter Limits from Tech Specs

ML20134P962
Person / Time
Site:	Byron, Braidwood
Issue date:	10/24/1996
From:	Stanley H COMMONWEALTH EDISON CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9612020005
Download: ML20134P962 (5)

Text

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Commonwealth Ediwn Company Braidwood Generating Station Route c), Ilox 81 Ilraceville, llA) 60%19 Tel HI5-15&2M01 October 24,1996 United States Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555-0001

Subject:

Additional Information Regarding the Removal of Cycle-Specific Parameter Limits from Technical Specifications Byron Nuclear Power Station Facility Operating Licenses NPF-37 and NPF-66 NRC Docket Nos. 50-454 and 50-455 Braidwood Nuclear Power Station, Facility Operating Licenses NPF-72 and NPF-77 NRC Docket Nos. 50-456 and 50-457 j

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References:

1.

M.T. Lesniak (Comed) letter to Document Control Desk (USNRC),

Expansion of Operating Limits Report, dated December 21,1995 2.

R. R. Assa (USNRC) letter to D. L. Farrar (Comed), Request for Additional Information on the Operating Limits Report, dated June 7, 19 %

In Reference 1, Comed provided a proposed amendment request to expand the current Byron /Braidwood Operating Limits Report (OLR) to include the limits suggested by Generic Letter (GL) g8-16, ' Removal of Cycle-Specific Parameter Limits from Technical Specifications. The cycle-specific parameters for Shutdown Rod Insertion Limit, Control Rod Insertion Limits, Axial Flux Difference Target Band, Heat Flux Hot Channel Factor, and Nuclear Enthalpy Rise Hot Channel Factor were proposed to be included in the OLR. Currently, the OLR contains a cycle-specific limit for the radial peaking factor, F, and the cycle-specific Moderator Temperature Coefficient.

y Reference 2 is a Request for Additional Information from NRC review of Reference 1. Comed is supplementing our initial submittal with the requested additional information. Specifically, the INSERT B page in Attachment B of Reference I requires further clarification of the methodologies presented in Technical Specification 6.9.1.9. Attachment A of this letter pro 5 ides a response to the questions from Reference 2.

Attachment B of this letter provides a revision to the previously submitted INSERT B of Reference 1. Comed requests that you replace Insert B of Reference I with Attachment B of this letter.

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9612O20005 961024 PDR ADOCK 05000454 P

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A l'nicom Company

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Document Control Desk Page 2 The proposed changes have been resiewed and approved by both on-site and off-site review in accordance with Comed procedures. Comed has reviewed this infonnation in accordance with 10 CFR 50.92 (c) and has determined that the original no significant hazards determination of Reference 1 remains valid.

Comed is notifying the State ofIllinois of this information by transmitting a copy of this letter and the associated attachments to the designated State Official.

To the best of my knowledge and belief, the statements contained in this document are true and i

correct In some respects, these statements are not based on my personal knowledge, but on information furnished by other Comed employees, contractor employees, and/or consultants. Such information has been resiewed in accordance with company practice, and I believe it to be reliable.

Please direct any questions to Marcia Lesniak, Nuclear Licensing Administrator, at (630)663-6484.

Sincerely,

/

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. Gene Stanley Site Vice-President Braidwood Generating Station HGSfrWS-pb/fb/m4.

Attachments cc:

A. B. Beach, Regional Administrator - Region III G. F. Dick, Byron Project Manager - NRR R. R. Assa, Braidwood Project Manager - NRR S. D. Burgess, Senior Resident Inspector - ByTon C. Phillips, Senior Resident inspector - Braidwood Oflice of Nuclear Facility Safety - IDNS Signed before nw on this M day AM7/thdd

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by NA O1Y llYM

!! MIt'!fAL SE4L I!

(( TINA M TAMAYO-SANTOLIN ff h NOI ARY PUBLIC ST ATE OF ILLINOIS >

J ATTACHMENT A i

NRC RAIltem 1 Reference to "the latest approved version" of the documents listed in Technical Specification (TS)

Section 6.9.1.9 is not legally acceptable The specific approved version and date of the methodology used to obtain the operating limits report (OLR) parameter must be listed. Please modify proposed TS Section 6.9.1.9, accordingly.

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Response

TS Section 6.9.1.9 has been modified to include the specific approved version and date of the

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methodologies used to obtain the operating limits report parameters. The revised TS Section 6.9.1.9 is included as Attachment B.

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NRC RAIItem 2 1

i What is the approved neutronics method for calculating the Moderator Temperature Coefficient? This should be referenced in 6.9.1.9.

Response

j the approved neutronics method for calculating the Moderator Temperature Coefficient is included in NFSR-0081, " Commonwealth Edison Company Topical Report on Benchmark of PWR Nuclear j

Design Methods Using the Phoenix-P and ANC Computer Codes," dated July 1990.

Moderator Temperature Coefficient has been added after " Methodology for Specification:" in reference 7 of Technical Specification Section 6.9.1.9. This change is included in Attachment B.

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NRC RAIltem 3 i

Reference 2 has a typo; it should state "... Constant Axial Offset Control " instead of".. Constant Control Offset Control."

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Response

The correction is included in Attachment B.

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ATTACHMENT B l

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l INSERTB i

1 i

OPERATING LIMITS REPORT i

j 6.9.1.9 Operating limits shall be established and documented in the OPERATING LIMITS i

REPORT (OLR) before each reload cycle or any remaining part of a reload cycle for the following:

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1. Moderator Temperature Coefficient for Specification 3.1.1.3,

2. Shutdown Bank Insertion Limit for Specification 3.1.3.5,

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3. Control Bank insertion Limit for Specification 3.1.3.6,

4. Axial Flux Difference Limits, Target Band for Specification 3.2.1, j

5. Heat Flux Hot Channel Factor and K(Z) for Specification 3.2.2, j

6. Nuclear Enthalpy Rise Hot Channel Factor, and Power Factor Multiplier for Specification j

3.2.3, and

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7. F, Radial Peaking Factor for Specification 4.2.2.2.

The analytical methods used to determine the operating limits shall be those previously reviewed and approved by the NRC in the following documents:

1. WCAP 9272-P-A, " Westinghouse Reload Safety Evaluations Methodology," dated July 1985 (Westinghouse Proprietary).

(Methodology for Specification: Shutdown Bank insertion Limit, Control Bank Insertion Limit, Axial Flux Difference, Heat Flux Hot Channel Factor, and Nuclear Enthalpy Rise Hot Channel Factor)

2. WCAP-8385, " Power Distribution Control and Load Following Procedures-Topical Report,"

dated September 1974 (Westinghouse Proprietary). (Methodology for Specification:

Axial Flux Difference, Constant Axial Offset Control)

3. WCAP 9220-P-A, " Westinghouse ECCS Evaluation Model-1981 Version," Revision 1, dated February 1982 (Westinghouse Proprietary). (Methodology for Specification: Heat Flux Hot Channel Factor) 4.' WCAP 9561-P-A, "BART A-1: A Computer Code for the Best Estimate Analysis of Reflood Transients," including Addendum 3,

"Special Report - Thimble Modeling in Westinghouse ECCS Evaluation Model," Revision 1, dated July 1986 (Westinghouse Proprietary). (Methodology for Specification: Heat Flux Hot Channel Factor)

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5. WCAP 10266-P-A, "The 1981 Version of the Westinghouse ECCS Evaluation Model i

Using the BASH Code," Revision 2, dated March 1987, includog Addendum 1 " Power Shape Sensitivity Studies," Revision 2-P-A, dated December 15,1987, and Addendum 2

" BASH Methodology improvements and Reliability Enhancements" Revision 2 dated May 1988 (Westinghouse Proprietary). -(Methodology for Specification: Heat Flux Hot Channel Factor)

6. NFSR-0016, " Commonwealth Edison Company Topical Report on Benchmark of PWR Nuclear Design Methods," dated July 1983. (Methodology for Specification: Shutdown Bank Insertion Umit, Control Bank Insertion Limit, Axial Flux Difference, Heat Flux Hot Channel Factor, and Nuclear Enthalpy Rise Hot Channel Factor)

7. NFSR-0081, " Commonwealth Ediwn Company Topical Report on Benchtnark of PWR Nuclear Design Methods Using the Phoenix-P and ANC Computer Codes," dated July 1990. (Methodology for Specification: Shutdown Bank Insertion Limit, Control Bank insertion Limit, Axial Flux Difference, Heat Flux Hot Channel Factor, Nuclear Enthalpy Rise Hot Channel Factor, and Moderator Temperature Coefficient)

8. WCAP 10079-P-A, "NOTRUMP, A Nodal Transient Small Break and General Network Code," dated August 1985 (Westinghouse Proprietary). (Methodology for Specification:

Heat Flux Hot Channel Factor and Nuclear Enthalpy Rise Hot Channel Factor)

9. WCAP 10054-P-A, " Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code," dated August 1985 (Westinghouse Proprietary). (Methodology for Specification: Axial Flux Difference, Heat Flux Hot Channel Factor, and Nuclear Enthalpy Rise Hot Channel Factor)

10. Comed letter from D. Saccomando to the Office of Nuclear Reactor Regulation dated December 21,1994, transmitting an attachment that documents applicable sections of WCAP-11992/11993 and Comed application of the UET methodology addressed in

" Additional Information Regarding Application for Amendment to Facility Operating i

Licenses-Reactivity Controls Systems".

The operating limits shall be determined so that all applicable limits (e.g. fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as SHUTDOWN MARGIN, and transient and accident analysis limits) of the safety analysis are met.

The OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident inspector.

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