ML20137X116
| ML20137X116 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 04/16/1997 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20137X084 | List: |
| References | |
| NUDOCS 9704210163 | |
| Download: ML20137X116 (4) | |
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UNITED STATES E
NUCLEAR REGULATORY COMMISSION If WASHINGTON, D.C. 2066H001
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 88 TO FACILITY OPERATING LICENSE NO. NPF-37, l
AMENDMENT NO. 88 TO FACILITY OPERATING LICENSE NO. NPF-66, AMENDMENT N0. 80 TO FACILITY OPERATING LICENSE NO. NPF-72, AND AMENDMENT NO. 80 TO FACILITY OPERATING LICENSE NO. NPF-77 COMONWEALTH EDISON COMPANY BYRON STATION. UNIT NOS. 1 AND 2 BRAIDWOOD STATION. UNIT NOS. 1 AND 2 DOCKET NOS. STN 50-454. STN 50-455. STN 50-456 AND STN 50-457 i
1.0 INTRODUCTION
By letter dated December 21, 1995, as supplemented October 24, 1996, and i
March 24, 1997, Commonwealth Edison Company (Comed, the licensee) requested changes to the Byron and Braidwood Technical Specifications (TS) to expand the current Operating Limits Report (0LR). Specifically, the cycle-specific parameters for Shutdown Rod Insertion Limit, Control Rod Insertion Limits, i
Axial Flux Difference Target Band, Heat Flux Hot Channel Factor [F,(z)], and Nuclear Enthalpy Rise Hot Channel Factor (FL) ing information that did not would be included in the OLR.
The March 24, 1997, submittal provided clarify change the initial proposed no significant hazards consideration i
determination.
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l The staff's evaluation of the proposed changes follows.
2.0 EVALUATION a
l The' licensee requested TS changes in accordance with 10 CFR 50.90. The specific changes are as follows:
1)
Specification 3.1.3.5 The LCO for the Shutdown Rod Insertion Limit will be revised to refer to the insertion limit specified in the OLR. The Action statement and Surveillance Requirement (SR) 4.1.3.5 will be revised accordingly.
2)
Specification 3.1.3.6 The LCO for the Control Rod Insertion Limits will be revised to refer to the insertion limits specified in the OLR. Figure 3.1-1, " Rod Bank Insertion 9704210163 970416 PDR ADOCK 05000454 P
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. Limits versus Thermal Power - Four Loop Operation" will be deleted from the TS i
and placed in the OLR. The Action statement will be revised accordingly.
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Specification 3.2.1 The Axial Flux Difference (AFD) target bands specified in the LC0 will be relocated to the OLR.
Figure 3.2-1, " Axial Flux Difference Limits as a Function of Rated Thermal Power" will be deleted from the TS and placed in the OLR.
4)
Specification 3.2.2 The equations for the Heat Flux Hot Channel Factor Fe(z) specified in the LC0 will be relocated to the OLR. The specific numerical limits will be replaced with F Figure 3.2-2, "K(Z) - Normalized F z) as a Function of Core will be removed from the TS and placed,(in the OLR.
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Height 5)
Specification 6.9.1.7 The NRC staff noticed that the title for TS 6.9.1.7 of the Braidwood TS was incorrectly written as " SEMIANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT" instead of " ANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT" as written in the Byron TS. The staff contacted Comed on April 10, 1997, and confirmed that Braidwood's TS 6.9.1.7 should read " ANNUAL RADI0 ACTIVE EFFLUENT RELEASE 1
REPORT" as requested by Comed in their June 13, 1994, application for j
amendments and erroneously issued by the staff in Amendment Nos. 59 dated february 2, 1995. The staff corrected this error.
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Specification 6.9.1.9 TS 6.9.1.9 will be revised to provide a listing of the individual TS that address cycle-specific operating limits relocated to the OLR. TS 6.9.1.9 will also be revised to list the specific NRC-approved analytical methods used to determine the OLR parameters.
These methods are documented in the following:
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).
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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 4
1986 (Westinghouse Proprietary).
(Methodology for Specification: Heat' i
Flux Hot Channel Factor).
5.-
WCAP-10266-P-A, "The 1981 Version of the Westinghouse ECCS Evaluation Model Using the BASH Code," Revision 2, dated March 1987, including j
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 i
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Proprietary).
(Methodology for Specification: Heat Flux Hot Channel i
Factor).
6.
NFSR-0016, " Commonwealth-Edison Company Topical Report on Benchmark of i
PWR Nuclear Design Methods," dated July 1983.
(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).
7.
NFSR-0081, " Commonwealth Edison Company Topical Report on Benchmark 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).
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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 Licenses-Reactivity Controls Systems."
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SUMMARY
The staff had reviewed the above items and concludes that the licensee provided an acceptable response to those items addressed in the NRC guidance in Generic Letter 88-16 on modifying-cycle-specific parameter limits in TS.
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. Plant operation'of the Byron and Braidwood Stations continues to be limited in
.accordance with the values of cycle-specific parameter limits that are established using NRC-approved methodologies. This will ensure 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.
i Therefore, the NRC staff concludes that' these changes to the TS are acceptable.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Illinois State official was notified of the proposed issuance of the amendments. The State official had no comments.
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5.0 ENVIRONMENTAL CONSIDERATION
The ambndments change 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 and change surveillance requirements. The NRC staff has determined i
that the amendments involve no significant increase in the amounts, and no
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significant change in the types, of any effluents that may be released i-offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a i'
proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding (62 FR i
7804). The amendment also involves a change to record keeping, reporting or administrative requirements. Accordingly, the amendments meet the eligibility l
criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9) and (c)(10). Pursuant to 10 CFR 51.22(b), no environmental. impact statement or l
environmental assessment need be prepared in connection with the issuance of the amendments.
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6.0 CONCLUSION
i The Commission has concluded, based on the considerations discussed above, I
that:
(l) there is reasonable assurance that the health and safety of the
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public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, i
and (3) the issuance of the amendments will not be inimical to the common i
defense and security or to the health and safety of the public.
Principal Contributor:
L. Kopp Date: April 16, 1997 i
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