ML20137Y601
| ML20137Y601 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 11/21/1985 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20137Y598 | List: |
| References | |
| NUDOCS 8512110084 | |
| Download: ML20137Y601 (6) | |
Text
{{#Wiki_filter:', j$'## %,h UNITED STATES 4 ",j f NUCLEAR REGULATORY COMMISSION y WASHINGTON, D. C. 20555 ,,,,s* t. SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT N0. 90 TO FACILITY OPERATING LICENSE NO. DPR-69 BALTIM0RE GAS AND ELECTRIC COMPANY CALVERT CLIFFS NUCLEAR POWER PLANT, UNIT NO. 2 DOCKET NO. 50-318
1.0 INTRODUCTION
By applications for license amendments dated August 29 and August 30, 1985, Baltimore Gas and Electric Company (BG8E) requested changes to the Technical Specifications (TS) for Calvert Cliffs Unit 2. The proposed changes to the TS reflect changes in analysis to accommodate Unit 2 Cycle 7 operation. The analyses also support proposed changes to the TS for steam line safety valve setpoints. 7 reload design is nearly identical to that of the Unit 1 The Unit 2 Cyc1 Cycle 8 design.9 Accordingly the latter cycle is taken as the reference cycle for the Unit 2 Cycle 7 reload. Since the two units are essentially the same this is appropriate and acceptable. The August 30, 1985 application, which contains the reload analysis, provides a description of the proposed reload design. The cycle length is increased over current cycles, which results in a more positive Beginning-of-Cycle moderator temperature coefficient and a more negative End-of-Cycle coefficient. A revision to the control rod group assignments is made to maintain the required shutdown margins. In the application dated August 29, 1985, BG&E has proposed that the range of allowable lift settings on the steam line safety valves be increased. Maximum proposed values in the TS would range from 995 to 1065 psig compared to a present range of 985 to 1035 psig. More flexibility would also be permitted for the range of setpoints for individual valves; These changes were previously approved for the reference cycle. r O P
2.0 EVALUATION Fuel Mechanical Design The mechanical design of the new fuel to be inserted for Cycl'e 7 is identical to that which was inserted for Cycle 6. All fuel to be employed for Cycle 7 has been examined to ascertain that adequate shoulder gap clearance exists. Analyses were performed with approved models and it was concluded that all shoulder gap and assembly length clearances are adequate for Cycle 7. A generic analysis of the phenomenon of clad collapse in modern CE fuels was submitted by the licensee as part of the reload submittal for the reference cycle. Reference 1 presents the results of the staff review of that analysis. It was concluded that a cycle specific annlysis of clad collapse was not required for Cycle 7 of Unit 2. We conclude that the analysis of the fuel mechanical design is acceptable. Nuclear Design The nuclear parameters of the Cycle 7 core have been determined with the same methods and techniques as those used for the reference cycle. Due to the similarity of the two cycles the values of the two sets of parameters are very nearly the same. In general the differences are well within the uncertainties assumed for the parameters. Control rod patterns are similar in the two reloads and analyses show that the shutdown margins are met in both Cycle 7 and the reference cycle with nearly the same margin in both cases. Power distributions have been e calculated for the End-of-Cycle 6 burnup which yields the largest Cycle 7 peaking factors. The expected power distributions are bounded by the values used in the safety analyses. The interpellet gap augmentation factors are being eliminated from the analysis as described in Reference 1. A fifteen percent negative bias is being applied to the Fuel Temperature (Doppler) coefficient to make it consistent with the power coefficient bias used in the ROCS /DIT code. This bias is applied conservatively to power increase transients. We find the nuclear design analysis to be acceptable. This conclusion is based on the fact that previously used and approved design methods are employed and the results are essentially the same as those for the reference cycle. Thermal Hydraulic Design The thermal hydraulic parameters of Cycle 7 are essentially the same as those of the reference cycle. Small differences in the total heat.. transfer area due to a lower number of shims in Cycle 7 are accounted for in the analyses. The safety limit DNBR value has been reduced from 1.23 to 1.21 for Cycle 7. This reduction was made possible by the approval (Reference 2) of CE's topical report on the CE-1 CHF correlation with non-uniform axial power distribution. The 1.21 value includes a rod bow penalty of 0.006 for burnups to 45 GWD/T and 5 percent uncertainty increase in the CHF correlation and in the thermal hydraulics code. I
3 We conclude that acceptable thermal hydraulic analyses have been performed for Cycle 7. Transient Analyses f-The Cycle 7 analysis contains a comparison of the core parameters input to safety analysis to those used for the reference cycle (Cycle 8 for Unit 1). These parameters are identical except that the safety limit DNBR ratio is 1.21 for Cycle 7 instead of 1.23 which was used in the reference cycle (see discussion above). In view of this fact and of the close similarity of the two units no reanalysis of the transient events was performed. For similar reasons the rod ejection event was not reanalyzed. ECCS analyses were perfonned to demonstrate compliance with the requirements of 10 CFR 50.46 with the Peak Linear Heat Generation Rate of 15.5 kW/ft. Staff approved codes were used for the analysis and both large break and small-break analyses were performed. These analyses confirmed that adequate margins to acceptance criteria exist even when account is taken of the recogtly discovered potential nonconservatism in the large break analysis We conclude that acceptable analyses of the transients and accidents have been performed. Technical Specifications TS associated with start-up testing and Cycle 7 operation were reviewed as described herein. E 1. TS 3/4.1.1.1-Shutdown Margin: The shutdown margin is being lowered from 4.3% a k/k to > 3.5% a k/k to accommodate the effects of extended burnup. The Boron Dilution, Excess Load and Steam Line Break events have been reanalyzed with the revised shutdown margin and show acceptable results as described in Reference 1. We conclude that this change is acceptable. 2. TS 3/4.1.1.4 Moderator Temperature Coefficient: The range of the allowed moderatog temperature coefficient (MTC) is -10~ggextendedfrom.5to7x10-bei ak/k'F and from -2.5 to -2.7 ak/k'F to accomodate the longer fuel cycle and extended burnup. Safety analyses have been changed to include the extended range with acceptable results and we conclude that the change is acceptable. O i
. 3. TS 4.2.1.4 and Figure 4.2-1: The flux peaking augmentation factors have been deleted afrom this TS and the axial fuel densification and thennal expansion factor have been reduced from 1.0 percent to 0.2 percent. The justification for these changes is discussed in Section 1 of Reference 1. We find these changes to be acceptable. The measurement-calculational uncertainty factor has been reduced from 7.0 to 6.2 percent. This is consistent with the latest approved evaluation of this quantity and is acceptable. 4. TS Figure 3.2-3b The radial peaking factor is being increased and made to be identical with the Unit 1 TS in accordance with the Unit 2, Cycle 7 setpoint analysis. 5. TS 3/4.7.1.1 and Table 4.7-1: A footnote has been added which will permit entry into Mode 3 for the purpose of detennining safety valve operability with a minimum of two operable safety valves per steam generator. Analyses have been performed to demonstrate that sufficient relief capacity exists in Mode 3 with only two operable main steam line safety valves per steam generator. We conclude that this change is acceptable. Table 4.7-1 has been revised to define the revised values of the allowable lift setpoints for the steam generator safety valves. The revised values are consistent with or conservative with respect to the values assumed in the safety analysis and are acceptable. A footnote has been added to TS Table 4.7-1 to allow flexibility among safety valve settings while preserving the overall relief capability. This change is consistent with the safety analyses and is acceptable. 6. Bases Changes to the TS Bases have been proposed to reflect changes in the TS, as described herein, and changes made to the analysis for Unit 2 cycle 7. These changes are acceptable.
3.0 ENVIRONMENTAL CONSIDERATION
This amendment involves a change in the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The staff has determined that the amendment involves no si'gnificant increase in the amounts, and no significant change in the types, of apy effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously published a proposco finding that the amendment
. involves no significant hazards consideration and there has been no public comment on such finding. Accordingly, the amendment meets the' eligibility criteria for categorical exclusion set forth in 10 CFR 651.2f(c)(9). Pursuant to 10 CFR 551.22(b), no environmental impact statemdht or environmental assessment need be prepared in connection with the issuance of the amendment.
4.0 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. Date: November 21, 1985 Principal Contributors: D. Jaffe W. Brooks O e 9 ~ n - --+ - -- - ~--.
References 1. Letter, D. Jaffe (NRC) to A. E. Lundvall (BG&E), Amendment 104 to Facility Operating License DPR-53, May 20,1985. 2. Letter, C. O. Thomas (NRC) to A. E. Scherer (CE), " Acceptance for Referencing of Topical Report CENPD-207", November 2, 1984. 3. Letter, J. A. Mihalcik (BG&E) to D. H. Jaffe (NRC), "Large Break LOCA ECCS Performance Evaluation", July 25, 1985. 1 k i l 4 -{ D e i i e i 7.,, y ..__,____.m_7 ,_... -..}}