ML15113A051
| ML15113A051 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 11/30/1981 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML15113A050 | List: |
| References | |
| NUDOCS 8112220275 | |
| Download: ML15113A051 (5) | |
Text
0 UNITED STATES 0
NUCLEAR REGULATORY COMMISSION 00 WASHINGTON, D. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 105 TO FACILITY OPERATING LICENSE NO. DPR-38 AMENDMENT NO. 105 TO FACILITY OPERATING LICENSE NO. DPR-47 AMENDMENT NO. 102 TO 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 29, 1981, Duke Power Company (Duke or licensee) submitted an application to change the common Oconee Nuclear Station (ONS) Technical Specifications (TSs) to support the full power operation of Unit 1 during Cycle 7 operation. Included in this application was a change to the high pressure injection system TSs to reflect the present design of the ONS systems. Another application was submitted by Duke on August 25, 1981, requesting that the minimum temperature of the Borated Water Storage Tank be increased from 40 to 500F to lessen the potential for thermal shock of the reactor vessel during high pressure injection system operation. Additional changes to the TSs were determined to be needed as a result of reevaluation of the Reactor Protective System accuracies. These changes were submitted by a sup plemental application dated October 16, 1981.
2.0 Background
The ONS Unit 1 core contains 177 fuel assemblies. The fuel assemblies for Cycle 7 operation include 68 new assemblies designated as Batch 9 and previously loaded assemblies designated as Batch 7B, 8A, 8B and 4E.
The Batch 9, 7B, 8A and 8B assemblies are mechanically interchangeable, Babcock and Wilcox (B&W) Mark B4 designs, while the single Batch 4E assembly is an older Mark B3 design. Reactivity control will be accomplished through the use of 69 full length Ag-In-Cd control rods, 60 burnable poison rod assemblies (BPRAs) and soluble boron shim. The BPRAs have the redesigned holddown latching mechanism which was previously approved.by the NRC staff.
3.0 Evaluation 3.1 Fuel Assembly Design Although all fuel assemblies for Cycle 7 operation are Mark B design, the Batch 9 and some Batch 8 assemblies have a slightly higher initial fuel density (94 to 95 percent theoretical density) as a consequence of a modified fuel fabrication process. In addition, four Batch 9 assemblies 2220275 811130 PDR ADOCK 05000269 DR
-2 have Zircaloy rather than Inconel intermediate spacer grids and are designated Hark BZ design, Batch 9B.
The Mark BZ design demonstration fuel assemblies were described in B&W's Report, BAW-1661P, submitted by letter dated April 10, 1981.
We have reviewed these changes in the fuel assembly design and find them to be relatively minor and not limiting for Cycle 7 operation.
The fuel assemblies were analyzed by the licensee for cladding collapse, stress and strain using methods and limits previously reviewed and approved by the NRC and were found to be bounded by either previously analyzed, or specifically analyzed for Cycle 7 conditions.
Fuel rod internal pressure was evaluated in accordance with approved methods and found to remain below normal system pressure for all assemblies except the Batch 4E assembly. We find this acceptable, however, because (1) the consequences of underestimating fuel rod internal pressure would be limited to the single assembly, and (2) the relative power density of this assembly (due to the higher burnup) is significantly lower than the average for the core and is not limiting for postulated transients and accidents. The differences in the computer codes used to calculate internal pressures may result in values that are too low at the beginning of core life and since these values are used to determine swelling and rupture behavior during a Loss of Coolant Accident (LOCA), reduced KW/ft limits at low core elevations during the first 50 effective full power days have been included in the TSs.
We have reviewed the factors related to fuel assembly design and find that they have been acceptably considered for Cycle 7 operation.
3.2 Core Physics The licensee described the core loading to be used in Cycle 7. Sixty-eight fresh assemblies having an initial enrichment of 3.28 weight percent U-235 will be loaded. A single-high burnup assembly will be located at the core center for its fifth cycle. Cycle 7 is to have an extonded length of 427 effective full power days.
For this reason burnable poison assemblies are used to limit the required beginning of cycle soluble boron concentration.
The nuclear characteristics of the core have been computed by methods previously used and approved for B&W reactors. 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 concentrations, stuck and ejected rod worths and moderator, coefficients. All safety criteria are still met. Shutdown margin values at beginning and end of cycle are 3.89 and 2.40 percent Ak/k, respectively, compared to the required 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.
-3 The key kinetics parameters for Cycle 7 have been compared to the values used in the Final Safety Analysis Report (FSAR) and densification report.
It is shown that in all cases Cycle 7 values are bounded by those previously used. We conclude that the FSAR transient and accident analyses are valid.
We have reviewed the proposed TSs for Cycle 7. The limiting safety systems settings and the limiting conditions for operation have been established by previously used and approved methods. The rod withdrawal limits for the various pump combinations and times in life are presented.
On the basis that previously approved methods were used to obtain the limits, we find them acceptable.
The effects of the recently discovered under-estimate of the errors in certain modules of the reactor protection system have been included.
By letter of September 10, 1981, the nuclear overpower trip setpoint was reduced from 105.5 to 104.9 percent full power. This is the same change that has been made on other B&W plants and is acceptable. Like wise the high reactor coolant temperature trip has been reduced from 619 to 618 degrees Fahrenheit. By letter dated October 16, 1981, the flux-flow-imbalance safety system setpoints were revised to restrict operation to narrower imbalance limits. On the basis that these.
setpoints were established by previously accepted methods, we conclude that the revised limits are acceptable.
3.3 Core Thermal-Hydraulics The thermal-hydraulics design conditions for Cycle 7 operation were compared to the Cycle 6 values in Duke's May 29, 1981, application (Table 6.1) and were shown to be identical.
The major differences of thermal-hydraulic concern between Cycle 6 and Cycle 7 are related to the Mark BZ demonstration assemblies which are discussed above and the rod bow Departure from Nucleate Boiling Ratio (DNBR) compensation.
The rod bow DNBR compensation for Cycle 7 operation was calculated using approved interim evaluation procedures which demonsttated that the Batch 9 fuel assemblies are the most limiting. The burnup used to calculate the rod bow penalty was the highest assembly burnup in Batch 9, (17,669 MWd/MTu) which contains the limiting (maximum radial peaking factor) fuel assembly. The resultant net rod bow penalty, after inclusion of the 1% flow area reduction factor credit, is a 0.2% reduction in DNBR.
We have reviewed the details of these calculations provided by letter dated October 16, 1981, and have concluded that the thermal-hydraulics design for Cycle 7 includes a margin of greater than 0.2% above the minimum acceptable DNBR and is therefore acceptable.
-4 3.4 Startup Physics Testing Included in the May 29, 1981,application, was a revision to the "Oconee Nuclear Station Startup Physics Test Program" which had been approved by the NRC letter dated March 23, 1981.
The proposed test would use the incore detector outputs to calculate the quadrant power tilt at 17% full power. Data are presented which show that the quadrant power tilts are consistent with those obtained from the previously used test.
The proposed test is a better measure of core symmetry and it provides information to help determine and correct the cause of a possible core asymmetry. The only disadvantage is that the proposed test cannot be performed until core power is at about 17% of full power. Because the proposed test would provide more usable data and because no core limits would be approached at 17% full power even for a large core asymmetry, we find the proposed test an acceptable substitute for the zero power rod swap asymmetry test for this and subsequent startup testing on all Oconee Units.
3.5 Technical Specifications Included in the May 29, 1981 application, were proposed revisions to the high pressure injection (HPI) cross connect system, the required volume of the boric acid storage tank (BAST) and the concentration of the boric acid solution in the borated water storage tank (BWST). The changes to the HPI sys'tem specification reflect completion of modifications which were approved by NRC letter dated December 13, 1978. The proposed specification had been approved, by Amendments 81, 81.and 78 issued on February 22, 1980, for Unit 3 and is therefore acceptable for the other Units. The changes to BAST volume (from 995 to 1020 cubic feet of 8700 ppm boron solution) and BWST concentration (from 1800 to 1835 ppm boron solution at a minimum volume of 46 ft.) were required to increase the minimum available boric acid solution to provide assurance that the reactor can be borated to an adequate subcritical margin during Cycle 7.
Since these changes maintain previously approved design bases, we find them to be acceptable.
By letter dated August 25, 1981, Duke proposed increasing the minimum temperature of the BWST from 40 to 50'F. This change would ensure that a higher temperature fluid would be injected; should the HPI system be operated. An increased injection temperature would lessen the thermal shock to the reactor pressure vessel.
We have reviewed this change and find it acceptable since it will provide added protection against thermal shock.
The TS changes related to full power operation of Unit 1 for Cycle 7 were reviewed as discussed in Section 3 of this evaluation and were found acceptable.
-5 One additional TS page was revised to remove a source of confusion. The last sentence of Ndte 5 of Table 4.4-1, on page 4.4-12, is confusing since it could be interpreted to conflIct with the requirements contained in Table 4.4-1.
To remove this confusion, this sentence was removed. Since no technical content of the requirements was changed by this action, we consider this to b n administrative action and find it to be acceptable.
4.0 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 551.5(d)(4),
that an environmental impact statement, or negative declaration and environ mental impact appraisal need not be prepared in connection with the'issuance of these amendments.
5.0 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 accidents previously considered and do not involve a signi ficant decrease in a safety margin, 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 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:
November 30, 1981