ML19209C823
| ML19209C823 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 09/21/1979 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19209C820 | List: |
| References | |
| NUDOCS 7910180347 | |
| Download: ML19209C823 (6) | |
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UNITED STATES y#c'gg NUCLEAR REGULATORY COMMISSION 5 ?\\ C WASHINGTON, D. C. 20555 \\f.M.... / SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT N0. 25 TO FACILITY OPERATING LICENSE NO. DPR-71 CAROLINA POWER AND LIGHT COMPANY BRUNSWICK STEAM ELECTRIC PLANT, UNIT NO. 1 DOCKET N0. 50-325 1.0 Introduction and Chronoloay In a letter dated September 11, 1979, Carolina Power and Light Company requested a Technical Specification change for MCPR limits at the Brunswick Nuclear Power Plant Unit No.1. The reason for the change in MCPR limits was because a fuel bundle loading error had been discovered in the cycle 2 core while reviewing the core loading tapes. The chronology of the actions taken following this event are as follows: On August 28, 1979, Carolina Power and Light Company (CP&L) identified the presence of a misoriented fuel bundle in the core for cycle 2 of the Brunswick Steam Electric Plant, Unit No.1 (BSEP-1). At that time the plant was operating at 90.5% power (Reference 1). LF%L immediately imposed an administrative 9.7 kW/ft Linear Heat Generatio,. Rate Limit (LHGR) and increased the operating limit MCPR to 1.45 on the misoriented bundle. CP&L contacted General Electric (GE) on the same day. GE recommended observing an LHGR limit to 9.7 KW/ft, and Minimum Critical Power Ratio (MCPR) of 1.38 pending a more detailed evaluation of the problem. CP&L imposed the GE recommended opera-tional limits, which resulted in 88% power, on August 28, 1979. On August 29, 1979, an NRC I&E inspector contacted NRCs Project Manager of BSEP-1 to verify NRC acceptance of the proposed GE limits (Reference 2). A staff review of the reload analysis for BSEP-1 cycle 2 operation (Reference 3) showed that the proposed GE limits included adequate conservatism to account for a " worst case" misloading. The NRC Project Manager informed the NRC I&E inspector on August 29, 1979 that the proposed GE limits were acceptable and that these limits should be maintained until a more detailed analysis and review was compl eted. On August 30, 1979 the NRC staff discussed the BSEP-1 misloading event with the CP&L staff (Reference 4). This discussion revealed that fuel bundle LJ0197 located at core position 29-10 was rotated 180 from its correct orientation. The misoriented fuel assembly was not a fresh fuel bundle and therefore was bounded by the " worst case" misloading analyzed in the reload analysis. CP&L stated that they expected the results of j]/Q 900 GEs reanalysis in early September. 7910180 [
. On September 12, 1979. CP&L telecopied Reference 5 to the NRC staff. Reference 5 contained the results of GEs reanalysis of the CP&L mis-loading event, and the proposed Technical Specification change to remedy the effects of the misoriented fuel bundle. 2.0 Discussion The Fuel Loading Error (FLE) discussed herein addresses only the mis-oriented bundle observed at BSEP-1. The FLE analyzed in the reload submittal is considered a low probability accident condition. However, in this case a FLE actually occurred. Thus, we have reassesied the effeTtih-of this particualr FLE on the steady state operations, the most severe operational transients, and other hypothetical accidents. The following sections describe our evaluation of the reanalysis sub-mitted by CP&L. In addition to discussing the reanalysis submitted by CP&L, this report references other specific information obtained via telephone conversations with the CP&L staff. 3.0 Evaluation 3.1 Nuclear Characteristics 3.1.1 Shutdown Margin CP&L stated that control rod (30-11) which is adjacent to the mis-oriented fuel assembly w(as not exercised in the startup program tests of the shutdown margin Reference 1). The closest rod exercised during the shutdown margin was two diagonal fuel bundles removed from the misoriented fuel bundle. The loose neutronic coupling of the cor? will minimize the effects of this misoriented fuel bundle on the shutdown margin test results. The staff agreed with this detemination by GE and CP&L in support of continued operation at the reduced limits. The results of GEs reanalysis confirmed our earlier determination (Reference 6, 9). 3.2 Thermal Hydraulics 3.2.1 Operating Limit MCPR Although a FLei is an accident condition, it is sufficient for safety to treat it against transient criteria. Thus, it is acceptable that the Minimum Critical Power Ratio (MCPR) resulting from a fuel loading error be no less than the safety limit MCPR, e.g.,1.07. The safety limit MCPR assures that during transients 99.9% of the fuel rods in the core will avoid transition boiling, and that transition boiling will not occur during steady state operations. 1178 201 GE has determined that the fuel loading error at BSEP-1 involving a 180 rotation of fuel bundle LJ0197, located at core position 29-10, does not significantly affect the core-wide transients. However GE has determined that local ACPR adjustments to the calculated CPR of the misoriented fuel bundle will maintain this fuel bundle within acceptable safety limits. GE calculated exposure-dependent ACPRs which include the 0.02 ACPR allowance required by NRC to allow for the axially varying water gap. The operating MCPR for fuel bundle LJ0197 will be adjusted by sub-tracting the following aCPRs from the calculated MCPRs for that fuel bundle. Exposure Adjustment ACPR (B0C-2) to 0.11 (EOC-2) - 2000 mwd /T (E0C-2) - (2000 mwd /T) to 0.12 (E0'-2) - (1000 mwd /T) (E0C-2) - (1000 mwd /T) to 0.13 (E0C-2) The resulting MCPR (after adjustments) for fuel bundle LJ019 shall be equal to, or greater than, the exposure dependent MCPRs shown in section 3.2.3 of the Technical Specifications. The core-wide MCPR limits, exclusive of fuel bundle LJ0197, and exclusive of the above adjustments, will be equal to, or greater than, the exposure dependent MCPRs shown in section 3.2.3 of the Technical Specification (no change). The staff finds the proposed ACPR adjustment to fuel bundle LJ0197 acceptabl e. 3.2.3 Operating Limit LHGR The LHGR of fuel bundle LJ0197 should be modified to reflect the increase in local peaking due to the rotation of the fuel assembly. CP&L pro-poses to increase the calculated LHGR for fuel bundle LJ0197 by 20% until the modified peaking factors have been determined. The adjusted LHGR for fuel bundle LJ0197 will remain limited by the current Tech-nical Specification LHGR limits of 13.4 KW/ft. Since the 13.4 KW/ft Technical Specifkation limit on LHGR is below the safety limit LHGR of 17.5 KW/ft (Reference 7), we find the proposed adjustment to the calculated LHGR for fuel bundle LJ0197 acceptable. 1178 202
. 3.3 Rod Withdrawal Error Control rod 30-11, which is adjacent to the misoriented fuel bundle, has a rod worth of less than 60% of the worth of the limiting control rod analyzed in the cycle 2 RWE analysis. Therefore a RWE involving control rod 30-11 would be less severe than the limiting condition approved in Reference 8. We find this acceptable. 3.4 Accident Analysis ~ 3.4.1 ECCS Appendix K Analysis In Reference 8, it was concluded that the proposed operating limit MCPRs are more limiting than the 1.2 MCPR assumed, and found accept-able, in the Loss-of-Coolant Accident (LOCA) analysis. Likewise the 13.4 KW/ft LHGR is more limiting that the MAPLHGR verses Average Planar Exposure calculated in the cycle 2 LOCA analysis. For the misoriented fuel bundle at position 29-10, GE has proposed that BSEP-1 increase the locally calculated LHGR by 20% and decrease the locally calculated MCPR by the exposure dependent ACPRs shown in Section 3.2. These locally peaked values would still be maintained within the more restrictive limitr currently required for reasons not connected with the LOCA analysis. With the above adjustments to the Limiting Conditions of Operations (LCOs), we have concluded that the plant will be in conformance witn all requirements of 10 CFR 50.46 and Appendix K to 10 CFR 50.46. 3.4.2 Control Rod Drop Accident 'he cold reactivity worth of control rod (30-11) was determined to be 0.95% AK compared with 1.21% AK for the maximum worth control rod. Even considering local peaking at the misoriented fuel bundle, the peak enthalpy resulting from a CRDA of control rod (30-ll)would be less than 208 cal /gm. Since this would be less severe than the CRDA analyzed and approved in the cycle 2 reload analysis, we find this condition acceptable. 4.0 Technical Specification Modifications The Licensee has proposed adjustments to the calculated MCPR for the misoriented fuel bundle LJ0197 located at core position 29-10. The adjustment involves subtracting the exposure dependent ACPRs shown in Section 3.2 from the calculated MCPRs for bundle LJ0197. The lower adjusted MCPRs would then be required to be greater than, or equal to, the exposure dependent MCPRs shown in Section 3.2.3 of the Technical Specifications. 1178 203
., The 13.4 KW/ft LHGR limits which are currently specified in the Tech-nical Specifications remain unchanged. This is acceptable since the 20% adjustment factor is to be applied to the calculated Local LHGR. Therefore the locally peaked LHGR will be maintained within the current 13.4 KW/ft LHGR limit. We find the above Technical Specification changes acceptable along with the procedures for accommodating the misoriented fuel bundle. Based on the abcve considerations, we conclude that the proposed change to Specification 3.2.3 is acceptable as written. 5.0 Environmental Consideration We have determined that the amendment does not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact. Having made this determination, we have further concluded that the amendment involves an action which is insignificant from the standpoint of environmental impact and pursuant to 10 CFR Section 51.5(d)(4) that an environmental impact statement or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance of the amendment. 6.0 Conclusion We have concluded, based on the considerations discussed above, that: (1) because the amendment does not involve a significant increase in the probability or consequences of accidents previously considered and does not involve a significant decrease in a safety margin, the change does not involve a significant hazards consideration, (2) there is reasonable assuranc.e that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such act.vities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public. Dated: September 21, 1979 1178 204
References 1. Telephone conversation, J. McQueen (CP&L) to R. Riggs (NRC), dated September 14, 1979. 2. Telephone conversation, K. Julian (I&E/NRC) to J. Hannon (NRC), dated August 29, 1979. 3. " Supplemental Reload Licensing Submittal" for Brunswick Unit 1, Reload 1 NED0-24166, December 1978. 4. Telephone conversation, Reactor Safety Branch (NRC) to H. Bowles (CP&L), dated August 30, 1979. 5. Telecopy, Serial No. GD-79-2289, transmitting CP&L request for Amend-ment, General Electric's analysis results, and proposed Technical Specification Changes, dated September 12, 1979. 6. Telephone conversation, H. Bowles (CP&L) to R. Riggs (NRC), dated September 18, 1979. 7 Safety Evaluation of the General Electric Generic Reload Fuel Application (NEDE-240ll-P), April 1978. 8. Reactor Safety Branch SER supporting BSEP-1 Cycle 2 Operation, March 28,1979. 9. CP&L letter dated September 18, 1979, Rotated Fuel Bundle: Technical Documentation for MCPR Limit Revision. 1178 205}}