ML15113A072
| ML15113A072 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 11/23/1983 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML15113A071 | List: |
| References | |
| NUDOCS 8312070350 | |
| Download: ML15113A072 (9) | |
Text
R EG UNITED STATES NU EAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 124 TO. FACILITY OPERATING LICENSE NO.
DPR-38 AMENDMENT NO. 124 TO FACILITY OPERATING LICENSE NO.
DPR-47 AMENDMENT NO. 121 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 8312070350 831123 PDR ADOCK 05000269 P
1.0 INTRODUCTION
By letters dated September 1, 1983 (Ref. 1) and September 14, 1983 (Ref. 2), Duke Power Company made application to modify the Oconee Nuclear Station Technical Specifications in support of Cycle 7 operation of Unit 2. The analysis performed and the resulting modifications to the Station's common Technical Specifications are described in the Unit 2 Cycle 7 reload report (Ref. 3).
The safety analysis for the previous sixth cycle of operation at Oconee Unit 2 is being used by the licensee as a reference for the proposed seventh cycle of operation. Where conditions are identified as limiting in the sixth cycle analysis, our previous evaluation (Ref. 4) of that cycle continues to apply. Our evaluation of the most recent reload submittal from the Oconee Station (Oconee Unit 1, Cycle 8 - Ref. 5) is also used as a basis for the findings described in the following sections.
1.1 Description of the Cycle 7 Core The Oconee Unit 2 Cycle 7 core will consist of 177 fuel assemblies, each of which is a 15X15,irray containing 208 fuel rods, 16 control rod guide tubes, and one incore instrument guide tube. Cycle 7 will operate in bleed-and-feed mode with core reactivity control supplied mainly by soluble boron in the reactor coolant and supplemented by 61 full length control rod assemblies (CRAs) and 64 burnable poison rod assemblies (BPRAs). In addition, 8 axial power shaping rods (APSRs) are provided for additional control of the axial power distribution. Seventy-two fresh assemblies having an initial enrichment of 3.24 weight percent U 235 will be loaded. The length of Cycle 7 is expected to be 421 effective full power days (EFPD) compared with 400 EFPD accumulated during Cycle 6.
The licensed core full power level remains at 2568 MWt.
2.0 EVALUATION OF THE FUEL SYSTEM DESIGN 2.1 Fuel Assembly Mechanical Design The 72 Babcock & Wilcox (B&W) Mark-B4 fuel assemblies loaded as Batch 9 at end of Cycle 6 (EOC 6) are mechanically interchangeable with the Batches SC, 7C and 8 fuel assemblies loaded previously at Oconee Unit 2.
The -Mark-B4 fuel assembly has been previously approved (Ref. 4) by the staff and utilized in other B&W nuclear steam supply systems. Batch 9 will contain one Advanced Cladding Pathfinder (ACP) assembly which, unlike the standard Mark-B4 design, is reconstitutable. This assembly.
contains 12 rods with special cladding, 6 z-ircomrium lined and 6 beta quenched, which are expected to provide improved resistance to water side corrosion and pellet-cladding interaction (PCI).
In other respects,
-2 the ACP assembly is nearly identical to the standard Mark-84 design and is not limiting for Cycle 7 operation. Because the ACP assembly d'oes not, in itself, result in any Technical Specification changes for the Cycle 7 core, and because the licensee has determined that its inclusion does not result.in any unreviewed safety question, this assembly may be incorporated into the Cycle 7 reload without NRC approval.
We are aware of a number of other recent changes to the B&W 15X15 fuel assembly design (e.g., a larger fuel assembly holddown spring, fuel pellets manufactured by an-alternate supplier). These changes have been approved for use in other operating B&W 177-fuel-assembly plants on a limited basis and may be incorporated into future cycles of operation at Oconee Unit 2. However, for the current cycle of operation, the licensee has identified no other changes in the fuel assembly mechanical design.
We find this acceptable.
In the course of our review, we have noted that a small number of holddown spring failures are continuing to occur at the Oconee station.
These springs are contained in the upper end fitting of the Mark-84 fuel assembly and are used to accommodate length changes due to thermal expansion and irradiation growth while providing a positive holddown force for the assembly. On May 14, 1980, a failed holddown spring was discovered by remote video inspection it Davis-Besse Unit 1. Further examination ultimately identified a total of 19 failed springs at that plant. Subsequent examination of spent fuel assemblies at other B&W reactors, including the Oconee station, revealed a small number of similar failures.
This issue was previously considered in our safety evaluation of the Oconee Unit 3 Cycle 6 reload (Ref. 6).
In that evaluation, we concluded that the holddown spring issue had been correctly analyzed and did not result in a safety concern for Unit 3 Cycle 6 operation. An inspection (Ref. 7) of all Unit 3 Cycle 6 assemblies revealed broken holddown springs in two assemblies due to be discharged. Another inspection (Ref. 8) revealed one broken holddown spring in Unit 1 Batch 4 fuel and three broken holddown springs in Oconee Unit 2 Batch 7 fuel.
More recently, 4 additional broken holddown springs were found:in Unit 1 (Ref.
9).
In all cases, the fuel was due to be discharged or the holddown springs were replaced prior to insertion.. We now conclude that a continuing program of detection and discharge/replacement of failed holddown springs is no longer adequate for the Mark-B4 fuel design.
Because our concern involves the use of the Mark-84 design in operating reactors in addition to the Oconee Station, we have reinitiated dis cussions of this problem with the fuel vendor, B&W.
2.2 Fuel Rod Desian The cladding stress, strain and collapse analyses for the fuel in the Cycle 7 core are generally bounded by conditions previously analyzed for Oconee Unit 2 or were analyzed specifically for Cycle 7 using methods and limits developed and used by the fuel vendor (B&W) and reviewed and
I approved by the NRC. In the case of Cycle 7, however, these same analyses were performed by the licensee rather than by the fuel vendor using a reload methodology (Ref. 10), which has also been approved (Ref. 11) by the NRC.
Duke Power Company previously applied this reload methodology, for the first time, in the Oconee Unit 3 Cycle 7 analysis (Ref. 12).
An exception to the general methodology has been identified by the licensee in the cladding stress analysis. The fuel rod total stress is not permitted to exceed the unirradiated yield strength (the previous limit) of the cladding. Two times the minimum unirradiated yield strength is now used as a criterion for the total stress calculation, as suggested by the ASME Boiler and Pressure Vessel Code (Ref. 13).
Primary membrane plus primary bending stresses continue to be limited to the unirradiated yield strength, and primary membrane stress continues to be limited to two-thirds of this value.
We have previously noted (Ref. 12) that the ASME Boiler and Pressure Vessel Code does not apply to fuel rod cladding, specifically Zircaloy cladding. The application of this code to the mechanical analysis of fuel rods is suggested by the NRC Standard Review Plan (Ref. 14), but only as general guidance. Stress limits similar to those proposed by the licensee have been accepted elsewhere, but these are usually based on both yield and ultimate tensile strength limits. Because the cladding stress analysis is not, and has not been, limiting for operation at Oconee, we accept the analysis as submitted.
The fuel thermal and material design analyses, including fuel rod internal pressure limits, continue to be analyzed with previously approved methods. The licensee has continued to rely upon fuel thermal performance from several B&W codes:
TAFY-3 (Ref. 15), TACO-1 (Ref. 16),
and TACO-2 (Ref. 17). A combination of TAFY-3 and TACO-2 analyses were used to generate the LOCA limits as described in Tables 7-2 and 7-3 of Reference 3. Three sets of bounding values for allowable LOCA peak linear heat rates are given as a function of core height.
These limits apply during the periods 0-25 EFPD, 25-65 EFPD and 65 EFPD to end-of cycle. We have determined (see Section 3.0) that these limits have been satisfactorily incorporated in to the Technical Specifications for Cycle 7 through the operating limits on rod index and axial power imbalance.
3.0 EVALUATION OF NUCLEAR DESIGN The nuclear characteristics of the core have been computed by methods previously approved for the Oconee Nuclear Station (Ref. 10).
Compari.
sons are made between the physics parameters for Cycles 6 and 7. The differences that exist between the nuclear characteristics are due to the increased cycle length, i.e., an estimated 421 EFPD for Cycle 7 vs
-4 an actual 400 EFPD for Cycle 6 (or an estimated 390 EFPD for Cycle 6).
In addition the peripheral assembly locations are occupied by once burned assemblies in order to minimize neutron leakage and increase cycle length. The critical boron concentration for beginning of Cycle 7 is higher at hot zero power, no xenon conditions, than at Cycle 6.
Changes in the radial flux and burnup distribution (peaked in the inner fuel assemblies and suppressed in the peripheral assemblies) accounts for the differences in control rod worths.
For example, Group 7 control rod worth at the beginning of cycle (hot full power) is 1.51 percent Ak/k vs 1.46 percent Ak/k for Cycle 6 and at the end of Cycle 7 is estimated as 1.64 percent Ak/k vs 1.53 percent Ak/k at the end of Cycle 6.
Safety criteria are met for stuck and ejected rod worths. Shutdown margin values are 3.88 percent Ak/k at beginning of cycle and 2.68 percent Ak/k at end of Cycle 7 compared to the required 1.0 percent Ak/k. The effective delayed neutron fraction both at beginning and end of Cycle 7 is practically unchanged-compared to the corresponding values of Cycle 6.
The values in.Cycle 6 and those in Cycle 7 are bounded by those calcu lated in the FSAR. Based on this 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.
3.1 Evaluation of Accident and Transient Analysis The key kinetics parameters for Cycle 7 have been compared to the values used in the FSAR and those calculated in Cycle 6. It is shown that in all cases Cycle 7 values are bounded by those used previously. The effects of fuel densification on the FSAR accidint results have been evaluated and are reported in Reference 19. However, the fuel as semblies of this reload contain fuel rods with a theoretical density higher than those considered in Reference 19, hence the conclusions are still valid. Considering the previously reviewed and approved design used in the FSAR and the results of the densification report (Ref. 18),
we conclude that the transients in Cycle 7 are bounded by the FSAR and the densification report and, hence, are acceptable.
4.0 THERMAL-HYDRAULIC Duke Power Company performed the thermal-hydraulic design analysis supporting Cycle 7 operation using the Oconee Reload Design Methodology (Ref. 10) that was previously reviewed and approved by the staff and employing the B&W approved methodology (Refs. 19, 20 and 21).
Cycles 6 and 7 are hydraulically and geometrically similar as shown in Table 1. The maximum core bypass flow for Cycle 6 was 7.6 percent of the total system flow versus 7.8 percent for Cycle 7. This value is less than the bypass flow value (8.2 percent) that is assumed in the generic thermal-hydraulic design analysis. Therefore, we find core bypass flow of 7.8 percent of the total system flow to be conservative for Cycle 7 operation.
-6 6.0 SUM1MARY le conclude that the Oconee 2 Cycle 7 iVill not adversel.! affect the capability to operate the plant safely. We also conclude that the proposed changes to the Technical Specifications discussed above for Oconee 2 Cycle are acceptable.
7.0 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 environ mental impact appraisal need not be prepared in connection with the issuance of these amendments.
8.0 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 C2) 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 23, 1983 The following NRC staff personnel have contributed to this Safety Evaluation:
L. Lois, A. Gill, J. Voglewede, J. Suermann.
- 7 TABLE 1 THERMAL HYDRAULIC DESIGN CONDITIONS OCONEE UNIT 2, CYCLE 7 Cycle 6 Cycle 7 Design power level, MWt 2568 2568 System pressure, psia 2200 2200 Reactor coolant flow, % design flow 106.5 106.5 Core bypass flow, % total flow 7.6 7.8 Vessel inlet/outlet coolant temp at 555.6/602.4 555.6/602.4 100% power, *F Ref design radial-local power 1.71 1.71 peaking factor Ref design axial flux shape 1.5 cosine 1.5 cosine Hot channel factors:
Enthalpy rise 1.011 1.011 Heat flux 1.014 1.014 Flow area 0.98 0,.98 Active fuel length, in.
(a)
(a)
Avg heat a1 x at 100% power, 103 176(b) 17 6(b)
Btu/h-ft2 a, CHF correlation BAW-2 BAW-2 Min DNBR with densification penalty 2.05
>2.05 (a)142.2 inches for Batches.6C and 7C, 141.8 for Batches 8 and 9.
(b)Heat flux based on a conservative minimum densified length of 140.3 in.
'8 REFERENCES
- 1. H. B. Tucker (Duke) letter to H. R. Denton (NRC) on "Oconee Nuclear Station Unit 2" dated September 1, 1983.
- 2. H. B. Tucker (Duke) letter to H. R. Denton (NRC) on "Oconee Nuclear Station Unit 2" dated September 14, 1983.
- 3. "Oconee Unit 2, Cycle 7 Reload Report," Duke Power Company Report DPC-RD-2002, September 1983. Attachment 3 to References 1 and (revised) 2 above.
- 4. L. S. Rubenstein (NRC) memorandum for T. Novak (NRC) on "SER Oconee Unit 2 Reload for Cycle 6" dated February 19, 1982.
- 5. J. F. Stolz (NRC) letter to H. B. Tucker (Duke) on Oconee Unit 1 Cycle 8 reload dated August 3, 1983 and transmitting Amendments No. 122, 122 and 119 to Facility Operating Licenses No. DPR-38, DPR-47 and DPR-55, respectively.
- 6. L. S. Rubenstein (NRC) memorandum for T. M. Novak (NRC) on "Oconee Unit 3 Cycle 6 Reload" dated February 2, 1981.
- 7. W. 0. Parker (Duke) letter to J. P. O'Reilly (NRC) dated July 23, 1982.
- 8. W. 0. Parker (Duke) letter to J. P. O'Reilly (NRC) dated February 16, 1982.
- 9. H. B. Tucker (Duke) letter to J. P. O'Reilly (NRC) dated July 21, 1983.
- 10.
"Oconee Nuclear Station Reload Design Methodology" Technical Report, Duke Power Company Report NFS-1001, Revision 4, April 1979.
- 11.
L. S. Rubenstein (NRC) memorandum for T. M. Novak (NRC) on "Duke Reload Design Methodology Technical Report Evaluation" dated May 26, 1981.
- 12. L. S. Rubenstein (NRC) memorandum for G. Lainas (NRC) on "SER
- Oconee Unit 3 Reload for Cycle 7" dated September 17, 1982.
- 13.
ASME Boiler and Pressure Vessel Code,Section III, "Rules for Construction of Nuclear Power Plant Components," American National Standard ANSI/ASME BPV-III (1980 Edition), July 1, 1980.
- 14.
U.S. Nuclear Regulatory Commission Standard Review Plan Section 4.2 (Revision 2), "Fuel System Design," U.S. Nuclear Regulatory Commission Report NUREG-0800 (formerly NUREG-75/087), July 1981.
- 15.
C. D. Morgan and H. S. Kao, "TAFY - Fuel Pin Temperature and Gas Pressure Analysis," Babcock and Wilcox Company Report BAW-10044, May 1972.
- 16.
R. H. Stoudt, et al., "TACO" Fuel Pin Performance Analysis," Babcock and Wilcox Company Report BAW-10087P-A, Rev. 2, August 1977.
- 17. Y. H. Hsii, et al., "TACO2: Fuel Pin Performance Analysis," Babcock and Wilcox Company Report BAW-10141P, January 1979.
- 18.
"Oconee Unit 2 Fuel Densification Report".BAW-1395, Babcock and Wilcox dated June 1973.
- 19.
"Normal Operating Controls" BAW-10122, Babcock and Wilcox, August 1978.
- 20. "Oconee Nuclear Station, Units 1, 2 and 3, Final Safety Analysis Report," Docket Nos. 50-269, 50-270 and 50-287, Duke Power Company, Charlotte, North Carolina.
- 21.
"Oconee Unit 2, Cycle 6, Reload Report" BAW-1691, Revision 1, Babcock and Wilcox, April 1982.
- 22. L. S. Rubenstein (NRC) letter to
. H. Taylor (B&W), "Evaluation of Interim Procedure for Calculating DNBR Reductions Due to Rod Bow,"
October 18, 1979.
- 23. W. 0. Parker, Jr. (Duke), letter to H. R. Denton (NRC), October 16, 1981.