ML15112A928
| ML15112A928 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 06/12/1980 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML15112A927 | List: |
| References | |
| NUDOCS 8007160445 | |
| Download: ML15112A928 (10) | |
Text
o UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 0 6 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 83 TO FACILITY OPERATING LICENSE NO. DPR-38 AMENDMENT NO. As TO FACILITY OPERATING LICENSE NO. DPR-47 AMENDMENT NO. 80 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 letters dated November 16, 1979, and March 12, 1980 (1, 2), Duke Power Company (DPC or the licensee) requested amendment to Appendix A of License No. DPR-47 for the Oconee Nuclear Station, Unit No. 2. The DPC submittal of March 12, 1980, was provided to substitute for the November 16, 1979 application in its entirety due to the extended Cycle 4 operation (350 Effective Full Power Days (EFPDs) vs. 297 EFPDs) provided by our License Amendment of January 4, 1980. This increased cycle length necessitated a complete core redesign. Therefore, the proposed amendment to the Technical Specifications (TSs) allowing full power operation of Unit 2 based on an extended Cycle 4, and the Babcock and Wilcox (B&W) Topical Report BAW-1565, Rev. 1 (2), were presented in support of Cycle 5 opera tions. The topical report describes the fuel system design, accident analyses, and the startup test program. The design length of the pro posed Cycle 5 operation is 360 EFPDs. At the end of the current cycle, Cycle 4, a total of 68 burned fuel assemblies will be discharged and 68 fresh fuel assemblies will be loaded in the core in a checkerboard pattern.
The Cycle 5 operational mode will be changed from rodded to feed-and bleed. This change is not regarded as a major change in operating mode since Oconee 2 was operated essentially in a rods-out configuration during the latter part of the previous cycle. A similar change fr9m rodded to feed-and-bleed mode was approved for Oconee Unit 3, Cycle 5(3).
Reactivity is controlled by 61 full-length Ag-In-Cd control rods, soluble boron shim, and 56 burnable poison rod assemblies (BPRAs). The latter is required to offset the increased excess reactivity built into the longer cycle length. In addition to the full-length control rods,' 'ight axial power shaping rods are provided for axial power distribution control.
The following sections present the evaluations of any changes to the fuel system design, accident and transient analysis, startup physics testing, nuclear and thermal-hydraulic design, and the proposed TS changes required for Cycle 5 operation.
By letter dated September 11, 1979(20) the licensee also requested an amendment to all three Oconee Units to incorporate a secondary water chemistry monitoring program in the body of the license.
6 0071O4
-2 2.0 Evaluation of Core Design Modifications 2.1 Fuel System Design 2.1.1 General To achieve the longer cycle (Cycle 5) length and to utilize an in-out-in fuel management scheme, the operational mode is being changed from rodded to feed-and-bleed and an average core fuel enrichment increase is applied.
To ensure that achieved core power distributions conform with values assumed in the safety and setpoint analyses, monthly incore power maps are to be compared with predicted distributions and deviations are to be reported to the NRC. In addition to the longer cycle length and fuel management changes, two fuel assemblies in Cycle 5 are demonstration 17 X 17 Mark-CR assemblies. The Mark-CR demonstration assemblies of batch 5 are mechani cally identical in function to the Mark-C assemblies of batch 4 described in Reference 4; these assemblies have been twice burned in previous cycles.
One Mark-BZ demonstration fuel assembly is included in batch 7. The Mark-BZ is a 15 X 15 fuel assembly similar to the Mark-B assembly described in Reference 5 except that six intermediate spacer grids are of Zircaloy material, and an Inconel 718 spring replaced the Inconel X750 holddown spring. The Mark-BZ assembly is described in Reference 6, which also states that reactor safety and performance are not adversely affected by the presence of the one Mark-BZ demonstration assembly.
2.1.2 Rod Design The fuel pellet end configuration has changed from a spherical dish for batches 1 through 6 to a truncated cone dish for batch 7. This minor design change facilitates manufacturing while maintaining the same end void volume. We conclude that fuel performance will not be adversely affected by this change and it is thus acceptable.
2.1.3 Cladding Creep Collapse Due to its longer accumulated incore exposure, the fuel of batch 5 is more limiting than the fuel in other batches. The batch 5 assembly power his tories were analyzed and the most limiting Mark-B and Mark-C assemblies were used to perform the creep collapse analysis using the CROV computer code and procedures described in Reference 7. The collapse time for the most limiting assemblies were both conservatively determined to be more than 30,000 effective full power hours (EFPHs), which is greater than the maximum projected residence time for Cycle 5.
operation. We conclude that cladding collapse has been adequately considered.
-3 2.1.4 Cladding Stress and Strain For design evaluation., the primary stress is less than two-thirds of the minimum specified unirradiated yield strength, and all stresses (primary and secondary) are less than the minimum specified unirradiated yield strength. The licensee states that the stress analysis has a margin in excess of 30%.
The fuel design criteria specify a limit of 1.0% on cladding circumferential plastic strain. The pellet design is established for plastic cladding strain of less than 1.0% at maximum design local pellet burnup (55,000 MWD/MTU) and heat generation rate (20.15 KW/ft).
Oconee 2 fuel will not operate up to these maximum allowable values. We conclude that the cladding stresses and strains to be experienced.by the Cycle 5 fuel are acceptable.
2.1.5 Thermal Design All fuel assemblies in this core are thermally similar. The fresh batch 7 fuel inserted for Cycle 5 operation introduces no significant differences in fuel thermal performance relative to the other fuel remaining in the core. Linear Heat Rate (LHR) capabilities are based on centerline fuel melt and were established using TAFY 3 code (8) with consideration for fuel densification. We conclude that the fuel thermal design is acceptable.
2.2 Nuclear Design The design core physics parameters for Cycle 5 are generated using the B&W version of PDQ07 (9, 10, 11) and are compared to the Cycle 4 parameters (2,
Table 5-1). The boron concentrations for Cycle 5 are higher due to the additional reactivity'necessary for the longer cycle which is not com pletely offset by the BPRAs. The control rod worth differs between cycles due to changes in burnup and radial flux distributions. Cycle 5 shutdown margin is calculated to be 3.27% AK/K and 2.46% AK/K for beginning of cycle (BOC) and end of cycle (EOC) conditions, respectively, with the maximum worth rod stuck out of the core. The required shutdown margin is 1.0% AK/K; thus, the BOC and EOC shutdown margins are acceptable.
2.3 Thermal-Hydraulic Design The incoming batch 7 fuel is hydraulically and geometrically similar to the fuel remaining from previous cycles. The thermal-hydraulic models and methodo logies used to support Cycle 5 operation are described in References 5, 12, and 13. The main differences between Cycle 5 and the reference Cycle 4 are discussed below.
2.3.1 Core Bypass Flow The maximum core bypass flow due to the removal of all orifice rod assemblies (ORAs) in Cycle 4 was 10.4%. For Cycle 5 operation, 56 BPRAs will be inserted, leaving 50 vacant assemblies, resulting in a decrease in calculated maximum core bypass flow to 8.1% (i.e., net increase in core flow). A flux/flow trip setpoint of 1.08 was established to compensate for the increase in core flow.
4 2.3.2 BPRA Retainers The retainers added to provide positive holddown of BPRAs introduce a small Departure from Nucleate Boiling Ratio (DNBR) penalty as discussed in Reference
- 14. However, the increase in core flow due to the BPRA insertion (Section 2.3.1 above) more than compensates for the decrease in DNBR due to the BPRA retainers.
2.3.3 Mark-CR and Mark-BZ Demonstration Assemblies The two Mark-CR demonstration assemblies will be limited to a design peak of 1.50, and the Mark-BZ low-absorption demonstration assembly will be limited to a 1.40 design peak. This will assure both peaking and DNBR margin for Mark-CR and Mark-BZ assemblies and certify that they are not limiting for reactor protection. The 1.71 design radial-local peak remains valid for all other assemblies.
2.3.4 Rod Bow DNBR Penalty The rod bow penalty applicable to Cycle 5, according to the licensee, was calculated using the interim rod bow penalty evaluation procedure approved in Reference 15. The limiting (maximum radial x local peak) fuel assembly for Cycle 5 is a batch 7 assembly at a projected burnup of 15,219 MWD/MTU.
The calculated rod bow penalty using this procedure is 0.5%. Utilizing the 1% Departure from Nucleate Boiling (DNB) credit for the flow area reduction factor, the actual penalty is zero; therefore, no penalty is applied to the DNB calculations and they are thus acceptable.
3.0 Evaluation of Accidents and Transients The licensee has examined each Final Safety Analysis Report (FSAR) accident analysis with respect to changes in Cycle 5 parameters to determine their effect on the plant thermal performance during the analyzed accidents and transients. The key parameters having the greatest effect on the outcome of a transient or accident are the core thermal parameters, thermal-hydraulic parameters, and physics and kinetics parameters. Fuel thermal analysis values are listed in Table 4-2 of Reference 2 for all fuel batches in Cycle 5. Table 6-1 of the same reference compares the thermal-hydraulic parameters for Cycles 4 and 5. These parameters are the same for both cycles with the exception of the higher value of design Maximum Departure from Nucleate Boiling Ratio (MDNBR) for Cycle 5 (2.05 as compared to 1.98 for Cycle 4). A comparison of the key kinetic parameters from the FSAR and Cycle 5 its provided in Table 7-1 of Reference 2. These comparisons indicate no significant changes (Table 4-1 of Ref.2 compared to Table 4-1 of Reference 12) or changes in the conservative direction (Tables 6-1, 7-1 of Reference 2).
The effects of fuel densification on the FSAR accident analyses have been evaluated in Reference 13.
A generic Loss of Coolant Accident (LOCA) analysis for the B&W 177-fuel assembly, lowered loop Nuclear Steam Supply System (NSSS) has been performed using the final acceptance criteria Emergency Core Cooling System (ECCS) evaluation model (Reference 17).
That analysis used the limiting values of key parameters for all plants in the 177-FA lowered loop category, and therefore is bounding for the Oconee 2 Cycle 5 operation.
-5 We conclude from the examination of Cycle 5 core thermal and kinetic prop erties, with respect to acceptable previous cycle values and with respect to the FSAR values, that this core reload will hot adversely affect the Oconee 2 plant's ability to operate safely during Cycle 5.
4.0 Emergency Core Cooling System An Exemption was granted on December 18, 1978, to 10 CFR 50.46(a), "Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors."
The Exemption provided for its own termination upon completion of the modifications required by the Exemption. By letter dated December 13, 1978 (18), we found the design of the modifications to be acceptable. DPC has installed the modifications at Oconee 2 (19) and prepared acceptable operating procedures; thus, we conclude that the as-modified ECCS required by the Exemption of December 18,:1978, is acceptable.
5.0 Startup Test Program Startup tests have been proposed by DPC to provide assurance that Oconee 2 has been loaded as intended. This test program is similar to that used at Oconee Nuclear Station and other B&W reactors. We have reviewed the test program and find it acceptable.
6.0 Technical Specification Changes Proposed modifications to"the Oconee 2 TSs needed to support Cycle 5 operation are described below (2):
(1) The effect of transient xenon on power peaking is conservatively accounted for by the xenon penalty factor-of 5%.
(2) Primarily due to the decrease in core bypass flow, the power-to-flow ratio has been increased to 1.08. Reactor trip setpoints of power based on flow have been correspondingly changed.
(3) Oconee 2 will be changed from a rodded to a feed-and-bleed mode of operation for Cycle 5. A similar change was approved for Oconee Unit 3 for Cycle 5 (3).
(4) The following limits have been changed:
- a. Axial power and reactor power imbalance safety limits and trip setpoints for 2, 3, and 4 reactor coolant pump operation.
- b. Rod position limits for 2, 3, and 4 reactor coolant pump operation for less than 150 EFPDs and for more than 150 EFPDs during the cycle life.
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- d. Operational power imbalance limits for 0 to 360 EFPDs.
We have evaluated the reload report for the Oconee 2 Cycle 5 operation and the proposed TS changes that reflect the changed parameters for the new cycle and find the revised TSs acceptable.
7.0 Boron Dilution Accidents Theseoaccidents were not addressed in the DPC March 12, 1980(2) license amend ment submittal as the licensee previously submitted his analyses of these accidents in the FSAR. The Oconee FSAR did not specifically include the analysis for moderator dilution after the reactor vessel was drained down to the bottom of the lowest nozzle to enable maintenance work, such as steam generator repairs or reactor coolant pump repairs, and subsequent refilling of the primary system.
The system during such work Is in the cold shutdown mode and the accident of concern is refilling the system with unborated water which would result in a return to criticality, possibly, with all control rod assemblies fully inserted.
While such an accident would not result in fuel failures nor release any radia tion outside the primary system and surely none to the environment it is still an undesirable event. Our review of the licensee's Technical Specifications 3.5.1 and 3.5.2 indicate that the source range nuclear instrumentation channel is maintained in an operable condition during the cold shutdown mode and that a shutdown margin of 2.5% Ak/k is maintained in this mode. Our position for plants currently receiving an operating license is for a shutdown margin of 1% Ak/k.
Based on the above, we conclude that the Oconee Station is operated to mitigate the effects of a boron dilution accident. DPC is engaged in a program to further diminish the probability of a boron dilution accident by strengthening the Technical Specifications and Station Operating Procedures. This program will be implemented prior to any maintenance work requiring drainage of the primary system to the nozzles for any of the three Oconee reactors.
8.0 Control Rod Guide Tube Wear By letter dated November 23, 1979 we requested DPC to inspect control rod guide tubes for wear. B&W performed the inspections for DPC on spent fuel assemblies in the Oconee spent fuel pools. The results of these tests, performed by eddy current techniques, indicated negligible wear. Similar inspections were per formed on spent fuel at the Rancho Seco plant with results that confirmed the Oconee results. We conclude that operation of Oconee 2 in Cycle 5 will not result in guide tube wear beyond design limits and is thus acceptable. Continued testing, particularly for fuel in the final cycle-of extended cycles may still be needed before we can complete our evaluation of this problem.
-7 9.0 Secondary Water Chemistry Monitoring By letter dated September 11, 1979 DPC requested amendment to the Facility Operating Licenses to incorporate the monitoring program for secondary water chemistry in the body of the license.
In 1976, we sent letters to the licensees who operate Pressurized Water Reactors (PWRs) regarding the control of secondaryiater themistry to inhibit corrosion of steam generator tubes. The letters requested the licensees to propose Technical Specification changes to incorporate limiting conditions for operation and surveillance requirements for secondary water chemistry parameters. This request was sent to DPC by letter dated August 18, 1976.
Many licensees objected to the Model Technical Specifications principally on the basis that they could unnecessarily restrict plant operation, The majority of these licensees submitted alternative approaches that were directed more toward monitoring and record keeping rather than specific limits on chemistry parameters. At the time of our request, we recognized that a major disadvantage of the Technical Specifications was a.potential decrease in operational flexi bility, but our request was motivated by an overriding concern for steam gener ator tube integrity. Our objective was to provide added assurance that licensees would properly monitor and control Secondary Water themistry to limit corrosion of steam generator tubes.
However, based on the experience.and knowledge gained since 1976, we concluded in mid-1979 that Technical Specification limits would not be the most effective way of accomplishing this objective, Due to the complexity of the corrosion phenomena involved, and the state-of-the-art as it exists today, we believe that a more effective approach would be to institute a license condition that requires the implementation of a Secondary Water chemistry monitoring and control program containing appropriate procedures and administrative controls.
The required program and procedures would be developed by the licensees, with any needed input from their reactor vendors or other consultants, and thus could more readily account for site and plant specific factors that affect chemistry conditions in the steam generators, In our view, such a license condition would provide assurance that licensees would devote proper attention to controlling secondary water chemistry while also providing the needed flexibility to allow them to more effectively deal with any off-normal conditions that might arise.
Consequently, by letter dated July 23, 1979, we requested the licensee to propose such a license condition for the Oconee Station. The licensee responded on September 11, 1979 to our request and agreed to implement the program within 60 days from the issuing date of the proposed amendment. The proposed amendment complies with the guidance we provided to the licensee in our July 23, 1979 request. The NRC staff has made minor changes to the wording of the proposed license condition for the purpose of clarification. These changes were discussed with and concurred in by the-licensee.
Based on our review, we have concluded that the addition of this license condition:
in conjunction with existing Technical Specifications on steam generator tube leakage and inservice inspection, will provide the most practical and comprehen sive means of assuring that steam generator tube integrity is maintained; and thus,.the proposed amendment is acceptable.
-8 10.0 Environmental Consideratioh We have determihed 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 concluided that the amendments involve an action which is insignificant from,the standpoint of environmental impact and, pursiunt to 10 CFR §51.5(d)(4), that an environmental impact statement, or negative declaration and environmental impact appraisal need not b4 prepared in connection with the issuance of these amendments.
11.0 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, hot 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: Junie 12, 1980
REFERENCES 1.,
Letter, W. 0. Parker.(Duke'Power Company) to H. R. Denton (NRC),
dated November 16,,1979.
- 2. Letter, W. 0. Parker (Duke Power Company) to H~. R. Denton (NRC),
dated March 12, 1980.
- 3. Letter, R. W. Reid, (NRC) to W. 0. Parker (Duke Power Company),
dated June 23, 1979, transmitting License Amendment for Oconee Unit 3, Cycle 5 reload.
- 4. Irradiation of Two 17 X 17 Demonstration Assemblies in Oconee 2, Cycle 2-Reload Report, BAW-1424, Babcock & Wilcox, Lynchburg,.
Virginia, January 1976.
- 5. Oconee Nuclear Station, Units 1, 2, and 3 -
Final Safety Analysis Reports, Dockets Nos. 50-269, 50-270, and 50-287, Duke Power Company.
- 6. Mark-BZ Demonstration Assembly - Licensing Report,' BAW-1533, Rev. 1, Babcock & Wilcox, Lynchburg, Virginia,..March 1980.
- 7. Program to Determine In-Reactor Performance of B&W Fuels - Cladding Creep Collapse, BAW-10084, Rev. 2, Babcock & Wilcox, Lynchburg, Virginia, October 1978.
- 8. C. D. Morgan and H. S. Kao, TAFY-Fuel Pin Temperature and Gas Pressure Analysis, BAW-10084, Rev. 2, Babcock & Wilcox, Lynchburg, Virginia, October 1978.
- 10.
Core Calculational Techniques and Procedures, BAW-10118, Babcock &
Wilcox, Lynchburg, Virginia, October 1977.
- 11.
Assembly Calculations and Fitted Nuclear Data, BAW-10116A, Babcock
& Wilcox, Lyntchburg, -Virginia, May 1977.
- 12.
Oconee Unit 2, Cycle 4 Reload Report, BAW-1491, Babcock & Wilcox, Lynchburg, Virginia, August 1978.
- 13.
Oconee 2, Fuel Densification Report, BAW-1395, Babcock & Wilcox, Lynchburg, Virginia, June 1973.
- 14.
BPRA Retainer Design Report, BAW-1496, Babcock & Wilcox, Lynchburg, Virginia, May 1978..
- 15.
L. S. Rubenstein (NRC) to J. H. Taylor (B&W), Letter, "Evaluation of Interim Procedure for Calculating DNBR Reductions Due to Rod Bow,"
October 18, 1979.
-2
- 16.
Letter, W. 0. Parker (Duke Power Company) to H. R. Denton (NRC),
Rpvised Pages to BAW-1565, Rev. 1, "Oconee Unit 2, Cycle 5 Reload Repprt, " dated April 30, 1980.
- 17.
ECCS Analysis of B&W's 177-FA Lowered-Loop NSS, BAW-10103A, Rev. 2, Babcock & W1lcox, Lynchburg, Virginia, Spetember 1977.
- 18.
Lettqr, R. W. Peid (NRC) to W. 0. Parker (DPC) dated December 13, 1978.
- 19.
Letter, W. 0. Parker (DPC) to H. R. Denton (NRC) dated May 29, 1980.
- 20.
Letter, W. 0. Parker (DPC) to H. R. Denton (NRC) dated September 11, 1979.