ML20127N317
| ML20127N317 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 01/21/1993 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20127N314 | List: |
| References | |
| GL-88-16, NUDOCS 9301290169 | |
| Download: ML20127N317 (4) | |
Text
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SAFETY EVALVATION BY THE OfflCE OF NUCLEAR REACTOR REGULATION R[ LATED 10 AMENDMEN.T NO. Ins TO FACILITY OPERATING LICENSE NO. NPF-29 ENTERGY OPERATIONS. INC.. ET AL.
GRAND GULF NUCLEAR STA110N. UNIT 1 DOCKET NO. 50-416
1.0 INTRODUCTION
By letter dated October 29, 1992, the licensee (Entergy Operations, Inc.),
submitted a request for changes to the Grand Gulf Nuclear Station, Unit 1 (GGNS) Technical Specifications (TS).
The proposed changes would modify specifications having cycle-specific aarameter limits by replacing the values of those limits with a reference to 11e Core Operating Limits Report (COLR).
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 TS.
Guidance on the proposed changes was developed by NRC on the basis of the review of a lead-plant proposal submitted on t1e Oconee plant docket that was endorsed by the Babcock and Wilcox Owners Group.
This guidance was provided to all power reactor licensees and applicants by Generic Letter 88-16, dated October 4, 1988.
2.0 EVALVATION The licensee's proposed changes to the TS are in accordance with the guidance provided by Generic Letter 88-16 and are addressed below.
(A) The Definitions 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 an NRC-approved methodology that maintains the limits of the safety analysis.
The definition notes that plant operation within these limits is addressed by l
individual specifications, i
l (B) The following specifications were revised'to replace the values of cycle-specific parameter limits with a reference to the COLR that provides these limits.
(1)
Specification 3.2.1 and Bases 3/4.2.1 The Average Planar Linear Heat Generation Rate (APLHGR) limits for this specification and for these Bases are specified in the COLR.
I 9301290169 930121 PDR-ALOCK 05000416 P
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(2) Specification 3/4 1.3 The Minimum Critical Power Ratio (MCPR) limits for this specification are specified in the COLR.
(3) Specification 3/4.2.4 and Bases 3/4.2.4 P
The Linear Heat Generation nde (LHGR)he COLR. limits for this specification and for these Bases are specified in t (C)
Specification 6.9.1.11, COLR, was added to the resorting requirements of the Administrative Controls Section of the TS. 111s specification requires that the COLR be submitted, upon issuance, to the NRC Docunent Control Desk with copics to the Regional Administrator and Resident inspector.
The report arovides the values of cycle-specific parameter limits that are applicaale for the current fuel cycle.
Furthermore, this specification requires that the values of these limits be established using NRC-approved methodologies and be consistent with all applicable limits of the safety analysis. The approved methodologies are the following:
(1) XN-NF-79-71(P), {_xxon Nuclear Plant Transient Methodoloav for Boilina Water Reactors, Exxon Nuclear Company, Inc., Richland, WA.
Approved by NRC letter dated October 24, 1986.
(2)
XN-NF-80-19(P)(A), Volume 1, Exxon Nuclear Methodoloav for Boilina Water Reactors - Neutronic Methods for Desion and Analysis, Exxon Nuclear Company, Inc., Richland, WA.
(3)
XN-NF-80-19(P)(A), Volume 1, MyJnced Nuclear Fuels Methodoloav for Boilina Water Reactors 1._ Benchmark Results for the CASMO-10/MICR0 BURN-B Calculation Methodoloav, Advanced Nuclear Fuels Corporation, Richland, WA.
(4) XN-NF-80-19(P)(A), alume 3, Exxon Nuclear Methodology for Boilina Water Reactors THERMEX: Thermal Limits Methodoloav Summary Description, Exxon Nc: lear Company, Inc., Richland, WA.
(5) ANF-913(P)(A) Volume 1, COTRANSA2: A_ Computer Proaram for Boilina Water Reactor Transient Analysis, Advanced Nuclear _ Fuels-Corporation, Richland, WA.
(6) ANF-1125(P)(A), ANFB Critical P_ower Correlation, Advanced Nuclear Fuels Corporation, Richland, WA.
(7)
XN-NF-84-105(P)(A), Volume 1, XCOBRA-T: A C_pmouter Code for BWR Transient Thermal' Hydraulic Core Analysis, Exxon Nuclear Company, Inc., Richicnd, WA.
(8) XN-Nf-573(P), B/diEEX Pellet Clad _intenclionlvalutilon Code for Egwrr_R3ni, Exxon Nuclear Company, Inc., Richland, WA.
Approved by NRC letter dated August 28, 1990.
(9)
XN-NF-81-58(P)(A), RQDEX2: Fuel Rod Thermaj-Me danical R upan n
[yalvation Model, Exxon Nuclear Company, Inc., Richland. WA.
(10) XH-NF-85-74(P)(A), 80R[1?L (fWR): fy.q)__Epd Thermal-Mechan _Lul Etiponse Evaluation Model, Exxon Nuclear Company, Inc., Richland, WA.
(11) XN-CC-33(P)(A), BOXY: A Etnenlind Multirod Heatuo Code with
~
10Lf_RiO_Aprendix K Heatup Option, Exxon Nuclear Company, Inc.,
Richland, WA.
(12) XN-NF-825(P)(A), BWR/6 Generic Rod Withduwal Error Analysis. MCPRo f or PlanL_Qpfation Within the Extended Opint_ing_Qomain, Exxon Nuclear Company, Inc., Richland, WA.
(13) XN-NF-81-51(P)(A) LQCA-Seismic Struct_ uni Respuse of an Exxon Nucltar Company BWR Jet Pump fuel Assembly, Exxon Nuclear Company, Inc., Richland, WA.
(14) XN-NF-84-97(P)(A), LQCA-Seismic Struttural Response of an ENC 9x9 BWR Jet Pumn fuel Assembiv, Advanced Nuclear fuels Corporation, Richland, WA.
(15) XN-NF-86-37(P), EU)eric.1QCA Break Spectrum Aqtlysis for BWR/6 Plants, Exxon Nuclear Company, Inc., Richland, WA.
Approved by NRC letter dated October 24, 1986.
~
(16) XN-NF-82-07(P)(A), Exxon Nuclear Comp 3ny ECCS Claddina Swellina and Ruplure Model, Exxon Nuclear Company, Inc., Richland, WA.
(17) XN-NF-80-19(A), Volumes 2, 2A, 2B, & 2C, Exxon Nucitar Method _oloav for Boilina Wtter Reaclors EXEM BWR ECCS Evaluation Model, Exxon Nuclear Company, Inc., Richland, WA.
(18) XN-NF-79-59(P)(A), tif_thodoloov for Calculation of Pressure Drop in DWR fuel Assemblies, Exxon Nuclear Company, Inc., Richland, WA.
References (1), (8), and (15) were not published by the vendor as approved topical reports; therefore, the references have been clarified by adding the NRC approval dates.
ff ally, the specification 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.
' s In addition, the licansee has revised the description of fuel assembly design features in TS Section 5.3, Reactor Core.
The reference to the initial core design is deleted and that core designs be developed and analyzed using NRC-approved codes and methods. Also, the design description requires fuel assembly types to be identified in the COLR.
On the basis of the review of the above items, the NRC staff concludes that the licensee provided an acceptable response to those items as addressed in the NRC guidance in Generic letter 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 an NRC-approved methodology, the NRC staff concludes that this change is administrative in nature and there is no impact on plant safety as a consequence.
Accordingly, the staff finds the proposed changes acceptable.
3.0 11 ATE CONSULTATION in accordance with the Commission's regulations, the Mississippi State official was notified of the proposed issuance of the amendment.
The State official had no comments.
4.0 ENVIRONMENTAL CONSIDERATION
This amendment involves a change to a requirement with respect to the instal-lation 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 Commissiot, has previously issued a proposed finding that this amendment involves no significant hazards consideration and there has been no public comment on such 11nding (57 FR 55578). This amendment also changes recordkeeping, reportins or administrative procedures or requirements.
Accordingly, with respect to these items, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 551.22(c)(9) and (10),
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.
5.0 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 vill not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributors:
T. Huang, SRXB/ DEST T. Dunning, OTSB/00EA H. Rathbun, PDIV-1/DRPW Date: January 21, 1993 t___m