ML15112B065
| ML15112B065 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 09/29/1982 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML15112B064 | List: |
| References | |
| NUDOCS 8210190161 | |
| Download: ML15112B065 (6) | |
Text
o0 UNITED STATES 0
NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 113TO FACILITY OPERATING LICENSE NO.
DPR-38 AMENDIENT NO. 113TO FACILITY OPERATING LICENSE NO.
DPR-47 AM1ENDMENJT NO. 110TO FACILITY OPERATING LICENSE NO.
DPR-55 DUKE POWER COMPANY OCONEE NUCLEAR STATION, UNITS NOS. 1, 2 AND 3 DOCKETS NOS. 50-269, 50-270 AND 50-287 1.0 Introduction By letter dated May 3, 1982, Duke Power Company (Duke or the licensee) submitted a license amendment application for the Oconee Nuclear Station CONS) common Technical Specifications (TSs) to support full power operation of Oconee Unit 3 during fuel Cycle 7.
Since changes to the common TSs are involved, a license amendment for all three units is necessary.
The applica tion had been compiled on an assumed fuel Cycle 6 length of 365 Effective Full Power Days (EFPD), but Unit 3 was shutdown 17 EFPD early due to un related, steam generator problems. This early shutdown required a number of reanalyses to account for the reduced fuel usage. Therefore, by letter dated August 11, 1982, Duke submitted a revised report which replaced the earlier submittal in its entirety.
Additional information, requested in a telephone conference held on August 12, 1982, was provided by letter dated August 16, 1982.
The report attached to the August 11, 1982 submi'ttal (Reload Report) was compiled using the "Oconee Nuclear Station Reload Design Methodology", Duke Technical Report NFS-1001, which was approved by NRC letter dated July 29, 1981.
Additionally, the startup testing will be performed in accordance with the "Oconee Nuclear Station Generic Startup Physics TESt Program" which we approved by letters dated March 23 and May 29, 1981.
8210190161 820929 PDR ADOCK 05000269 P
( )-2 The core consists of 177 fuel assemblies, each of which is a 15 by 15 array containing 208 fuel rods, 16 control rod guide tubes, and one incore instrunent guide tube.
Cycle 7 is to have an extended length of approximately 440 EFPD.
For this reason, burnable poison assemblies are used to limit the required beginning-of-cycle (BOC) soluble boron con centration. Cycle 7 will operate in a rods-out, feed-and-bleed mode as did Cycle 6.
2.0 Evaluation 2.1 Fuel Assembly Design Although all batches in the Oconee 3 Cycle 7 core will utilize the same Babcock and Wilcox 15x15 fuel design, the Batch 9 assemblies will be of the Mark B5, as opposed to the previously-loaded Mark 84, fuel design. The Mark B5 fuel assembly is identical to the Mark 84 except its upper end fitting has been redesigned to provide a positive holddown of fixed control components such as burnable poison rod assemblies (BPRAs),
neutron source rod assemblies, and orifice rod assemblies. Oconee 3 Cycle 7 is the first application of the new design. We have determined that no special treatment of the Mark B5 fuel assembly is necessary because the thermal-hydraulic and fuel rod mechanical analyses are unaffected.
Although the Oconee 3 Cycle 7 core will contain both Mark 84 and Mark 85 fuel assemblies, the fuel rods used in both assemblies are virtually identical.
The results of the linear-heat-rate-to-melt analysis show slightly different densification characteristics for the new, Batch 9 fuel as opposed to previous batches. However, the resulting linear heat rate (LHR) values are the same for all batches in the Cycle 7 core. We regard such design changes as within the range of expected fuel rod design variationand, therefore, find them acceptable.
Fuel rod cladding collapse, stress and strain, fuel rod internal pressure and fuel rod bowing were all acceptably analyzed.
0 0 2.2-Nuclear Design Comparisons were made between the physics parameters for Cycles 6 and 7.
The differences that exist between the parameters are due to the increased cycle length, which tends to increase values of critical boron concen trations.
Changes in the radial flux and burnup distributions between cycles also account for the differences in control rod worths, including ejected and stuck rod worths.
All safety criteria are still met.
Shut down margin values at beginning and end of cycle are 3.67 and 2.26 percent Ak/k, respectively, compared to the minimum required value of 1.0 percent.
Beginning of cycle radial power distributions Show acceptable margins to limits.
Based on our review, we conclude that approved methods have been used, that the nuclear design parameters meet applicable criteria and that the nuclear design of Cycle 7 is acceptable.
2.3 Thermal-Hydraulic Design In order to confirm that the thermal-hydraulic design of the reload core has been accomplished using acceptable methods and provide acceptable margin of safety from conditions which could lead to fuel damage during normal and anticipated operational transients, comparisons of the differences between Cycle 7 and Cycle 6 were performed.
The main differences are decreased core bypass flow and fuel rod bow compensation and have been shown to be of little consequence for Cycle 7.
We, therefore, conclude that the available margin for Cycle 7 has been demonstrated and that the thermal hydraulic design is acceptable.
2.4 Accident Analyses The important kinetics parameters for Cycle 7 have been compared to the values used in the Final Safety Analysis Report (FSAR).
The initial condi-
-4 tions of the transients in Cycle 7 are bounded by those assumed in the FSAR, and the safety analyses of Cycle 7 are, therefore, bounded by pre viously accepted analyses.
Two sets of bounding values for allowable Loss of Coolant Accident (LOCA) peak LHRs are given as a fun:tion of core height. The first set, which covers the first 50 EFPD, includes reduced LOCA kW/ft limits at low core elevations and are based on the interim LOCA LHR limits. The second set, which covers the balance of the cycle, are the Finil Acceptance Criteria LOCA LHR limits.
Those limits are identical to those approved for the previous cycle and are satisfactorily incorporated into the TSs for Cycle 7 through the operating limits on control rod position and axial power imbalance.
2.5 Technical Specification Modifications Oconee Unit 3, Cycle 7 TSs have been modified to account for minor changes in power peaking and control rod worths due to the transition to an 18-month, lumped burnable poison cycle.
We have reviewed the proposed TS revisions for Cycle 7. These changes con cern the (1) Core Protection Safety Limits of Specification 2.1; (2) Protec tive System Maximum Allowable Setpoints of Specification2.3; and (3) Rod Position Limits of Specification 2.5.2.
The limiting safety system settings and the limiting condition for operation have been established by approved methods.
Changes which reflect the core thermal-hydraulic response still maintain the safety limit Departure from Nucleate Boiling Ratio (DNBR) criterion of 1.30.
The control rod withdrawal limits for the various pump combinations and times in core life are presented as well as part length axial power shaping rod position limits.
On the basis that previously
e 9
-5 approved methods were used to obtain the limits, we find these TS modifications acceptable.
Editorial changes were also made for the bases for Units 1 and 2 (pages 2.1-2 and 2.1-3b) to correct referenced Figures.
Since these are editorial only, we find them acceptable.
2.6 Summary We have reviewed the fuels, physics, thermal-hydraulic and accident analyses information presented in the Oconee 3 Cycle 7 reload report.
We find the proposed reload and the associated modified TSs acceptable.
3.0 Environmental Consideration We have determined that the amendments do 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 amendments involve an action which is insignificant from the standpoint of environmental impact and, pursuant to 10 CFR 951.5(d)(4),
that an environmental impact statement, or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance of these amendments.
4.0 Conclusion We have concluded, based on the considerations discussed above, that: (1) because the amendments do not involve a significant increase in the probability or consequences of an accident previously evaluated, do not create the pos sibility of an accident of a type different from any evaluated previously and do not involve a significant reduction in a margin of safety, the amendments do not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by
( )-6 operation in the proposed manner, and (3) such activities will be conducted in compliance with the Commission's regulations and the issuance of these amendments will not be inimical to the common defense and security or to the health and safety of the public.
Dated:
September 29, 1982 The following NRC staff personnel have contributed to this Safety Evaluation:
P. C. Wagner, L. Kopp, A. Gill, J. Voglewede and S. Sun.