ML14183A222
| ML14183A222 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 07/15/1992 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML14183A221 | List: |
| References | |
| NUDOCS 9207220190 | |
| Download: ML14183A222 (7) | |
Text
rkkREG&Z ull 0UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 141 TO FACILITY OPERATING LICENSE NO. DPR-23 CAROLINA POWER & LIGHT COMPANY H. B. ROBINSON STEAM ELECTRIC PLANT. UNIT NO. 2 DOCKET NO. 50-261
1.0 INTRODUCTION
BY letter dated January 7, 1991, (Ref. 1), as supplemented by letters dated April 16, 1992 (Ref. 2), and June 4, 1992, Carolina Power & Light Company (CP&L or-the licensee), submitted a request for changes to the H. B. Robinson Steam Electric Plant, Unit No. 2 (HBR2) Technical Specifications (TS).
The proposed changes would modify TS having cycle-specific parameter limits by replacing the values of those limits with a reference to a Core Operating Limits Report (COLR) containing the values of those limits.
The proposed changes also include the addition of the COLR to the Definitions section and to the reporting requirements of the Administrative Controls section of the TS. Guidance on the proposed changes was developed by NRC and provided to all power reactor licensees and applicants by Generic Letter (GL) 88-16, dated October 4, 1988 (Ref. 3).
The April 16, 1992, and June 4, 1992, letters provided clarifying information and updated TS pages reflecting pages changed by recent amendments and did not change the initial determination of significant hazards consideration as published in the Federal Register.
2.0 EVALUATION The proposed changes to the TS are in accordance with the guidance provided by GL 88-16 and are addressed below:
(1) The Definition section of the TS was modified to include a definition of the COLR that requires cycle/reload-specific parameter limits to be established on a unit-specific basis in accordance with NRC-approved methodologies that maintain the limits of the safety analysis.
The definition notes that plant operation within these limits is addressed by individual specifications.
(2) The following TS were revised to replace the values of cycle-specific parameter limits with a reference to the COLR that provides these limits.
(a) TS 3.1.3.1 and 3.1.3.3 The moderator temperature coefficient (MTC) limits for these TS are specified in the COLR.
920 720T90 920715 PDR' ADOCK 05000261 P
-2 (b) TS 3.10.1.2 The shutdown rod insertion limit for this specification.is specified in the COLR.
(c) TS 3.10.1.3 and 3.10.1.4 The control rod insertion limits for these specifications are specified in the COLR.
(e) TS 3.10.2.1, 3.10.2.2, 3.10.2.2.1, and 3.10.2.2.2 The heat flux hot channel factor (F.) limit at rated thermal power (F RTP), and the normalized F0 limit as a function of core height K(z) for these specifications are specified in the COLR.
(f) TS 3.10.2.1 The nuclear enthalpy rise hot channel factor (FA ) limit at rated thermal power (F R.) and the power factor multip ier (PFz ) for thi specification is specified in the COLR.
(g) IS 3.10.2.2, 3.10.2.2.1, 3.10.2.2.2, 3.10.2.7, 3.10.2.9,- and
- 2.
- 10. 2.
11 The axial flux difference limits, the target band, and the axial variation function corresponding to the target band V(z) for these IS are specified in the COLR.
The bases of affected IS have been modified by the licensee to include appropriate reference to the COLR. Based on our review, we conclude that the changes to these bases are acceptable.
(3) TS 6.9.3.3 is revised to include the COLR under the reporting requirements of the Administrative Control section of the nS.
This TS requires that the COLR be submitted, upon issuance, to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector. The report provides the values of cycle-specific parameter limits that are applicable for the current fuel cycle. Furthermore, this TS requires that the NRC-approved methodologies be used in establishing the values of these limits for the relevant IS and that the values be consistent with all applicable limits of the safety analysis.
The approved methodologies are the following:
I--30
-3 (a) XN-75-27(A), latest revision and supplements, "Exxon Nuclear Neutronics Design Methods for Pressurized Water Reactors," Exxon Nuclear Company, Richland, WA 99352.
(Methodology for TS 3.1.3.1 - Moderator Temperature Coefficient, 3.10.1.2 - Shutdown Bank Insertion Limits, 3.10.1.3 and 3.10.1.4 Control Bank Insertion Limits, 3.10.2.1, 3.10.2.2, 3.10.2.2.1, 3.10.2.2.2 - Heat Flux Hot Channel Factor, 3.10.2.1 - Nuclear Enthalpy Rise Hot Channel Factor, and 3.10.2.2, 3.10.2.2.1, 3.10.2.2.2, 3.10.2.7, 3.10.2.9, and 3.10.2.11 - Axial Flux Difference).
(b) XN-NF-84-73(P), latest revision and supplements, "Exxon Nuclear Methodology for Pressurized Water Reactors: Analysis of Chapter 15 Events," Exxon Nuclear Corporation, Richland, WA 99352 (Accepted by the NRC for HBR2 in the SE related to Amendment No. 87 to Facility License No. DPR-23, November 7, 1984).
(Methodology for TS 3.1.3.1 - Moderator Temperature Coefficient, 3.10.1.2 -
Shutdown Bank Insertion Limits, 3.10.1.3 and 3.10.1.4 Control Bank Insertion Limits, 3.10.2.1, 3.10.2.2, 3.10.2.2.1, 3.10.2.2.2 - Heat Flux Hot Channel Factor, 3.10.2.1 - Nuclear Enthalpy Rise Hot Channel Factor, and 3.10.2.2, 3.10.2.2.1, 3.10.2.2.2, 3.10.2.7, 3.10.2.9, and 3.10.2.11 - Axial Flux Difference).
(c) XN-NF-82-21(A), latest revision, "Application of Exxon Nuclear Company PWR Thermal Margin Methodology to Mixed Core Configurations," Exxon Nuclear Company, Richland, WA 99352.
(Methodology for TS 3.10.2.1 - Nuclear Enthalpy Rise Hot Channel Factor).
(d)
XN-NF-84-93(A), latest revision and supplements, "Steamline Break Methodology for PWR'S," Exxon Nuclear Corporation, Richland, WA 99352.
(Methodology for TS 3.1.3.1 - Moderator Temperature Coefficient, 3.10.1.2 - Shutdown Bank Insertion Limits, 3.10.1.3 and 3.10.1.4 Control Bank Insertion Limits, 3.10.2.1 - Nuclear Enthalpy Rise Hot Channel Factor).
(e) XN-75-32(A), Supplements 1, 2, 3, 4, "Computational Procedure for Evaluating Rod Bow," Exxon Nuclear Company, Richland, WA 99352.
(Methodology for Specifications 3.10.2.1, 3.10.2.2, 3.10.2.2.1, 3.10.2.2.2 -
Heat Flux Hot Channel Factor, 3.10.2.1 -
Nuclear Enthalpy Rise Hot.Channel Factor).
-4 (f) XN-NF-82-49(A), latest revision, "Exxon Nuclear Company Evaluation Model EXEM PWR Small Break Model," Exxon Nuclear Company, Richland, WA 99352.
(Methodology for TS 3.10.2.1, 3.10.2.2, 3.10.2.2.1, 3.10.2.2.2 Heat Flux Hot Channel Factor, 3.10.2.1 - Nuclear Enthalpy Rise Hot Channel Factor, and 3.10.2.2, 3.10.2.2.1, 3.10.2.2.2, 3.10.2.7, 3.10.2.9, and 3.10.2.11 - Axial Flux Difference).
(g) EXEM PWR Large Break LOCA Evaluation Model as accepted in Letter, D. M. Crutchfield (NRC) to G. N. Ward (ENC), "Safety Evaluation of Exxon Nuclear Company's Large Break ECCS Evaluation Model EXEM/PWR and Acceptance for Referencing of Related Licensing Topical Reports," July 8, 1986.
EXEM PWR LBLOCA Model includes the following references:
XN-NF-82-20(P), latest revision and supplements, "Exxon Nuclear Company Evaluation Model EXEM/PWR ECCS Model Updates," Exxon Nuclear Company, Richland, WA 99352.
XN-NF-82-07(A), latest revision, "Exxon Nuclear CompanylECCS Cladding Swelling and Rupture Model," Exxon Nuclear Company, Richland, WA 99352.
XN-NF-81-58(A), latest revision, "RODEXZ Fuel Rod Thermal-Mechanical Response Evaluation Model," Exxon Nuclear Company, Richland, WA 99352.
XN-NF-85-16(P), Volume 1 and supplements, Volume 2, latest revision and supplements, "PWR 17x17 Fuel Cooling Test Program," Exxon Nuclear Company, Richland, WA 99352.
XN-NF-85-105(P), and supplements, "Scaling of FCTF Based Reflood Heat Transfer Correlation for other Bundle Designs," Exxon Nuclear Company, Richland, WA 99352.
(Methodology for TS 3.10.2.1, 3.10.2.2, 3.10.2.2.1, 3.10.2.2.2 Heat Flux Hot Channel Factor, 3.10.2.1 - Nuclear Enthalpy Rise Hot Channel Factor, and 3.10.2.2, 3.10.2.2.1, 3.10.2.2.2, 3.10.2.7, 3.10.2.9, and 3.10.2.11 - Axial Flux Difference).
(h) XN-NF-78-44(A), latest revision, "Generic Control Rod Ejection Analysis," Exxon Nuclear Company, Richland, WA 99352.
(Methodology for TS 3.10.1.2 - Shutdown Bank Insertion Limits, 3.10.1.3 and 3.10.1.4 - Control Bank Insertion Limits, 3.10.2.1, 3.10.2.2, 3.10.2.2.1, 3.10.2.2.2 - Heat Flux Hot Channel Factor).
-5 (i) XN-NF-621(A), latest revision, "XNB Critical Heat Flux Correlation,"
Exxon Nuclear Company, Richland, WA 99352.
(Methodology for TS 3.10.2.1 - Nuclear Enthalpy Rise Hot Channel Factor).
(j)
ANF-1224(A), "Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel," Advanced Nuclear Fuels Corporation, Richland, WA 99352.
(Methodology for TS 3.10.2.1 - Nuclear Enthalpy Rise Hot Channel Factor).
(k) XN-NF-82-06(A), latest revisions and supplements, "Qualification of Exxon Nuclear Fuel for Extended Burn-up," Exxon Nuclear Corporation, Richland, WA 99352.
(Methodology for TS 3.10.2.1, 3.10.2.2, 3.10.2.2.1, 3.10.2.2.2 Heat Flux Hot Channel Factor, 3.10.2.1 - Nuclear Enthalpy Rise Hot Channel Factor).
(1) Meyer, P. E. and Kornfilt, J., "NOTRUMP, A Nodal Transient Small Break and General Network Code," WCAP-10080-A, August 1985.
(Methodology for TS 3.10.2.1, 3.10.2.2, 3.10.2.2.1, 3.10.2.2.2 Heat Flux Hot Channel Factor, 3.10.2.1 - Nuclear Enthalpy Rise Hot Channel Factor, and 3.10.2.2, 3.10.2.2.1, 3.10.2.2.2, 2.10.2.7, 3.10.2.9, and 3.10.2.11 - Axial Flux Difference).
(m) Lee, N., Tauche, W. D., Schwartz, W. R., "Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP code," WCAP-10081-A, August 1985.
(Methodology for TS 3.10.2.1, 3.10.2.2, 3.10.2.2.1, 3.10.2.2.2 Heat Flux Hot Channel Factor, 3.10.2.1 - Nuclear Enthalpy Rise Hot Channel Factor, and 3.10.2.2, 3.10.2.2.1, 3.10.2.2.2, 3.10.2.7, 3.10.2.9, and 3.10.2.11 - Axial Flux Difference).
(n) Bordelon, F. M., et al., "LOCA-IV Program:
Loss of Coolant Transient Analysis," WCAP-8301 (Proprietary), and WCAP-8305 (Nonproprietary), June 1974 (accepted by the NRC in the SE related to WCAP-8472-A, "The Westinghouse ECCS Evaluation Model:
Supplementary Information," April 1975).
(Methodology for TS 3.10.2.1, 3.10.2.2, 3.10.2.2.1, 3.10.2.2.2 Heat Flux Hot Channel Factor, 3.10.2.1 - Nuclear Enthalpy Rise Hot Channel Factor, and 3.10.2.2, 3.10.2.2.1, 3.10.2.2.2, 3.10.2.7, 3.10.2.9, and 3.10.2.11 - Axial Flux Difference).
-6 (o) "Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment No. 87 to Facility Operating License No.
DPR-23, Carolina Power & Light Company, H. B. Robinson Steam Electric Plant, Unit No. 2, Docket No. 50-261," November 7, 1984.
(Methodology for TS 3.1.3.1 - Moderator Temperature Coefficient, 3.10.1.2 - Shutdown Bank Insertion Limits, 3.10.1.3 and 3.10.1.4 Control Bank Insertion Limits, 3.10.2.1, 3.10.2.2, 3.10.2.2.1, 3.10.2.2.2 - Heat Flux Hot Channel Factor, 3.10.2.1 - Nuclear Enthalpy Rise Hot Channel Factor, and 3.10.2.2, 3.10.2.2.1, 3.10.2.2.2, 3.10.2.7, 3.10.2.9, and 3.10.2.11 - Axial Flux Difference).
(p)
ANF-88-054(P), "PDC-3: Advanced Nuclear Fuels Corporation Power Distribution Control for Pressurized Water Reactors and Application of PDC-3 to H. B. Robinson Unit No. 2," Advanced Nuclear Fuels Corporation, Richland, WA 99352, latest revisions and supplements.
(Accepted by the NRC for H. B. Robinson Steam Electric Plant, Unit No. 2, in the NRC SE related to Amendment No. 128 to Facility License No. DPR-23, Docket No. 50-261, August 22, 1990).
(Methodology for TS 3.10.2.2, 3.10.2.2.1, 3.10.2.2.2, 3.10.2.7, 3.10.2.9, and 3.10.2.11 - Axial Flux Difference).
Finally, the TS requires that all changes in cycle-specific parameter limits be documented in the COLR before each reload cycle or remaining part of a reload cycle and submitted upon issuance to NRC, prior to operation with the new parameter limits.
On the basis of the review, the NRC staff concludes that the licensee provided an acceptable response to the items in GL 88-16 on modifying cycle-specific parameter limits in TS. Because plant operation continues to be limited in accordance with the values of cycle-specific parameter limits that are established using NRC-approved methodologies, the NRC staff concludes that this change is administrative in nature and there is no impact on plant safety as a consequence. Accordingly, the NRC finds that the proposed changes are acceptable.
As part of the implementation of GL 88-16, the staff has also reviewed a sample COLR provided by the licensee and concludes that the format and content of the sample COLR are acceptable.
3.0
SUMMARY
We have reviewed the request by CP&L to revise the HBR2 TS by removing the specific values of some cycle-dependent parameters from the TS and placing the values in a COLR referenced by the TS. Based on this review, we conclude that these revisions are acceptable.
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4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the State of South Carolina official was notified of the proposed issuance of the amendment. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20.
The NRC staff has determined that the amendment involves no 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 significant increase in individual or cumulative occupational radiation exposure.
The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding (56 FR 6868). The amendment also changes recordkeeping, reporting, or administrative procedures or requirements.
Accordingly, the amendment meets the eligibility 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 environmental assessment need be prepared in connection with the issuance of the amendment.
6.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
7.0 REFERENCES
- 1.
Letter (NLS-90-248) from G. E. Vaughn (CP&L) to NRC, dated January 7,
.1991.
- 2.
Letter (NLS-92-101) from R. B. Starkey, Jr. (CP&L), dated April 16, 1992.
- 3.
Generic Letter 88-16, "Removal of Cycle-Specific Parameter Limits from Technical Specifications," dated October 4, 1988.
Principal Contributor: T. Huang Date:
July 17, 1992