ML20210P472
| ML20210P472 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 04/30/1986 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20210P467 | List: |
| References | |
| NUDOCS 8605130401 | |
| Download: ML20210P472 (5) | |
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UNITE 3 STATES n
't, NUCLEAR REGULATORY COMMISSION 3
i WASHINGTON, D. C. 20666
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT N0.123T0 FACILITY LICENSE NO. DPR-62 CAROLINA POWER & LIGHT C0f;PANY BRUNSWICK STEAM ELECTRIC PLANT, UNIT N0 2 DOCKET N0. 50-324
1.0 INTRODUCTION
By letter dated December 20, 1985 as supplemented March 28, 1986 (Reference 1, NLS-85-415 and NLS-86-097) the Carolina Power & Light Company (CP&L, the licensee) submitted proposed changes to the Technical Specifications appended to. Facility Operating License No. DPR-62 for the Brunswick Steam Electric Plant (BSEP), Unit No. 2.
CP&L, in the meeting with the staff on March 18, 1986, gave a technical presentation about the changes for Cycle-7. The technical information discussed in the March p
18, 1986 meeting was provided formally in the March 28, 1986 submittal.
The proposed amer.dment would change the Technical Specifications (TSs) to permit operation of Unit 2 for Cycle 7.
The changes incorporate revised minimum critical power ratio (MCPR) values and delete references to 8X8 fuel which is totally removed from the core. The Cycle-6 operating MCPR values are increased by " ADDERS" ranging from 0.04 to 0.07 AMCPR for Cycle-7.
The licensee has relied on the results presented in the approved GE topical report NEDE-24011. " General Electric Standard Application for Reactor Fuel", or GESTAR II (Ref. 3) for safety analyses of postulated transients and accidents, as well as for the core-related areas of fuel design, thermal-hydraulic design, nuclear design (including power distributions and reactivity analyses) and their safety analyses.
2.0 EVALUATION 2.1 Fuel System Design - Fresh Fuel Assemblies BP80RB299 Fresh fuel assemblies (BP8DRB299), which are prepressurized 8x8 retrofit barrier fuel assemblies with an average enrichment of 2.99 w/o in U-235, will be loaded for Cycle 7 operation. Since (1) the prepressurized 8x8 retrofit barrier fuel has been previously approved (Ref. 3), and (2) the average enrichment of the fresh fuel is less than that of the approved maximum enrichment stated in Reference 3, we conclude that the fuel assemblies are acceptable for Cycle-7 operation.
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2-2.2 Nuclear Design l
The nuclear design and analysis of the proposed reload has been performed by the methods described in Reference 3.
Reference 3 has been approved for use in the design and analysis of reloads in BWR reactors and its use i
is acceptable for this. reload. We have reviewed the results of the j
nuclear design analysis for Brunswick Unit 2 Cycle-7 and have determincd that since they are consistent with those for similar reloads and are l
done with acceptable methods, they are acceptable.
2.3 Thermal Hydraulic Design l
The objective of the review of the thermal-hydraulic design of the core l
for Cycle-7 operation is to confirm that the thermal-hydraulic design has been accomplished using acceptable methods, and to assure an acceptable margin of safety from conditions which could lead to fuel damage during normal operation and anticipated transients, and to assure that the core is not susceptible to thermal-hydraulic instability.
The review includes the following areas:
(1) operating. limit minimum f.
critical power ratio (MCPR) and the related changes to the Technical Specifications, and (2) thermal-hydraulic stability. Discussion of the review concerning the thermal-hydraulic design for Cycle 7 operation follows.
i 2.4 Operating Limit MCPR and the Related Technical Specification Changes j
Ihe licensee performed an evaluation to establish minimum critical power ratio (MCPR) operating limits for Cycle-7. The licensee reviewed the previous reload analyses for Units 1 and 2.
From that they established j
the limiting
- transients for ODYN options A and B and for different fuel exposure levels. The uncorrected aCPR value as calculated by~GETAB for the limiting transients for BP/P8X8N fuel type was used as the base value.
l To this base val.ur. " ADDERS" were imposed as follows:
(a) a 0.01 oCPR to l
account for the GETAB round-off process; (b) a 0.01 ACPR to account for l
mid-cycle exposure shape and scram reactivity differences between Cycles 6 and 7; (c) a 0.02 ACPR ADDER to provide assurance, without an adverse
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j impact on operations, that the proposed MCPR limits bound any reasonable variation in Cycle-7 designs and potential abnormal modes of operation and i
)
(d) 0.0 to 0.03 ACPR to account fnr different fuel' types. The " ADDERS"~
thereforevariedintotalfrom0.M4to0.07dependingsponthefueltype, i
i exposure level and the type of transient. ODYN correction factors were r
l then superimposed on the (GETAB uncorrected ACPR + ADDERS) values to l
determine the operating limit MCPR. The previous reload analyses indicate that the maximum observed cycle to cycle variation in MCPR operating j
limits for the limiting transients is only.0.02 ACPR. The proposed i
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" ADDERS" (0.04 to 0.07 ACPR) are therefore conservative and are acceptable.
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l We find that since approved methods (Ref. 3) were used and the results show an acceptable margin of safety from conditions which could lead to fuel damage during any anticipated operational transient, that the j
thermal-hydraulic design of the Cycle-7 core is acceptable. The corresponding Technical Specification (3/4.2.3) changes are also acceptable since they are consistent with the Cycle-7 safety analysis.
2.5 Thermal-Hydraulic Stability t
The results of thermal-hydraulic analyses show that the maximum core stability decay' ratio is 0.78 for Cycle-7. We find that (1) the 4
i calculated decay ratio for Cycle-7 is less than that for similar-reload cores and (2) the Technical Specifications prohibit normal operation in l
the natural circulation mode in which the core would be less stable. We therefore conclude that the thermal-hydraulic stability results are acceptable for Cycle-7 operation.
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2.6 Transient and Accident Analyses The Postulated Uncontrolled Rod Withdrawal Error, Fuel Misorientation Event and Rod Drop Accident have been analyzed for this cycle. The cycle
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specific Rod Drop Accident analysis was necessary because-certain l
parameters (accident reactivity shape function and scram shape function 1
in the cold startup mode) were not bounded by the generic analysis. The results of the cycle specific analysis meet our acceptance criterion (220 calories per gram peak enthalphy) for this event and are therefore acceptable.
On the basis that approved methods have been used to perform the analyses
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and to obtain input parameters for them and that the results of the i
accident analyses are acceptable for Cycle-7, we conclude that the i
analyses of the three cited events are acceptable. Core-wide transient j
j analyses are discussed in Sections above.
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2.7 Technical Specification Changes Various revisions to the Technical Specifications have been proposed.
j The results of our review are as follows:
Section 3/4.2.3 and Table 3.2.3.2-1 of the Technical Specifications have been revised to include the proposed operating limit MCPRs for Cycle-7 i
operation. We find that the proposed operating limit MCPRs have been established using approved methods to avoid violation of the safety limit MCPR during any anticipated operational transient. We conclude that the Technical Specification changes related to the operating limit MCPRs are acceptaole 4
based on the discussion in Section 2.4 of this SE.
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. A note is added to Technical Specification 3.3.6.2 to indicate that during current Cycle operation the E0C recirculation pump trip system will be inoperable. This is acceptable since no credit is taken for this trip in the plant safety analysis.
The other changes are editorial in nature.
3.0 EVALUATION
SUMMARY
i From the basis of our review which is described above, we conclude that the Brunswick-2 reactor may be operated for Cycle-7 with the new fuel without undue risk to the health and safety of the public. This conclusion is based on the fact that acceptable methods and procedures were used to perform the design and analysis of the cycle and that the Iechnical Specifications have been correctly based on the results of that analysis.
4.0 ENVIRONMENTAL CONSIDERATION
S The amendment changes a requirement with respect to installation or j
g use of a facility component located within the restricted area as F
defined in 10 CFR Part 20 and changes surveillance requirements. The 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 consnent on such finding.
Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
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.
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5.0 CONCLUSION
We have 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, and (2) such activities will be conducted in compliance with the Commission's regulations and the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
i Principal Contributor: George Thomas Dated: April 30, 1986 i
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5.0 References 1.
Letters from CP&L to NRC, Request for License Amendment Fuel Cycle No.7-Reload Licensing, December 20, 1985, March 28, 1986 (NLS-85-415, NLS-86-097).
2.
23A1765, Supplemental Reload Licensing Submittal for Brunswick Steam Electric Plant Unit 2, Reload 5, May 1984.
3.
NEDE-24011-P-A-7-US, General Electric Boiling Water Reactor Generic Reload Fuel Applications, August 1985.
4.
R. E. Engel (GE) letter to T. A. Ippolito (NRC) dated May 6,1981.
5.
R. E. Engel (GE) letter to T. A. Ippolito (NRC) dated May 28, 1981.
6.
L. S. Rubenstein (NRC) memorandum for T. M. Novak (NRC) on " Extension of General Electric Emergency Core Cooling System Performance Limits" dated June 25, 1981.
7.
P. W. Howe (CP&L) letter to D. B. Vassallo (NRC) dated June 7,1982.
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