ML20211J595
| ML20211J595 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 10/27/1986 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20211J577 | List: |
| References | |
| NUDOCS 8611110184 | |
| Download: ML20211J595 (7) | |
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UNITED STATES
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8' NUCLEAR REGULATORY COMMISSION o
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i SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 111 TO PROVISIONAL OPERATING LICENSE NO. DPR-16 GPU NUCLEAR CORPORATION AND JERSEY CENTRAL POWER & LIGHT COMPANY OYSTER CREEK NUCLEAR GENERATING STATION DOCKET NO. 50-219
1.0 INTRODUCTION
By letters dated June 17, September 17, and October 13, 1986, GPU Nuclear (the licensee) requested an amendment to Provisional Operatin9 License No.
DPR-16 for the Oyster Creek Nuclear Generating Station (Oyster Creek).
This amendment would authorize changes to Section 2.3, limiting Safety System Settings, and to Section 3.10, Core limits, of the Appendix A-Technical Specifications (TS) to account for the Operating Cycid 11 reload.
The changes to Section 2.3 would increase (1) the neutron flux scram i'
setting for the average power range monitors (APPM) and (2) the neutron flux control rod block setting. The changes to Secticn 3.10 would increase the minimum critical power ratio (MCPR) limits and revise the maximum allowable average planar linear heat generation rate (MAPLHGR) for i
five loop and four loop operation in Figures 3.10-4 and 5, respectively.
i The changes to the figures would replace the MAPLHGR for the existing fuel l
type P80RB256L by that for the new fuel type P8DRB299. The MAPLHGR for the existing fuel types P80RB239 and P8DRB265H in Figures 3.10-4 and 3.10-5 are not.being changed by this amendment.
Included with these changes are changes to the Bases for TS Sections 2.3 and 3.10.
2.0 DISCUSSION By letter dated June 17, 1986 (Referencs 1), the licensee proposed to change the TS in the areas of the APRM Scram and Rod Block Lines and the MCPR and MAPLHGR limits for Oyster Creek to accomodate the Operating Cycle 11 reload. The submittal references the staff-reviewed and approved NED0-24195, " General Electric Reload Fuel Application for Oyster Creek" which provides the bases for the TS changes necessary for the Cycle 11 operation and the attached Appendix D (to NE00-24195) which is the summary of the results of the Cycle 11 reload core design and safety.
analysis.
i The Cycle 11 core will retain 372 irradiated fuel assemblies of Exxon Type VB, and GE types P80RB239 and P8DRB265H from the Cycle 10 and will add 188 fresh fuel assemblies (about 34% of the fuel) of GE Types P8DRB265H, P7DRB299tA and P8DRB299H (References 1, 2, 3). The reload is based on a Cycle 10 exposure of 15.769 GWD/t and the loading will be a conventional scatter pattern with low reactivity fuel on the periphery.
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1 The safety evaluation for the Cycle-11 reload included staff comparison of NEDO-24195 " General Electric Reload Fuel Application for Oyster Creek" with NEDE-24011, " General Electric Boiling Water Reactor Generic Reload Fuel Application" which was previously reviewed and approved by the staff for reference in the safety analysis of the GE Boiling Water Reactors. The-staff concluded that the methodology and procedures employed in the ' reload design and analysis are essentially the same as those described in the previously approved NED0-24011 and are acceptable. The procedures used to establish operating limits are similar to those previously approved and are acceptable. The safety analyses performed in-support of the Cycle 11 core design use the methods described in the NED0-24195. MAPLHGR values provided for the new reload fuel assemblies of P8DRB265H and P8DRB299 are provided based on LOCA analyses using approved methodologies and parameters. (Reference 4)
The projected End of Cycle 11 Maximum Batch Average Exposure was 27600 MWD /MTU.
This GE fuel design has been previously approved and operated well beyond this burnup range. A sumary of the results of the Cycle 11 reload core design and safety analyses are given in Appendix 0 of the Cycle 11 Reload Submittal.
In the core-related areas of fuel design, thermal-hydraulic design, nuclear design and safety analyses of the postulated accidents and _ transients, the
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licensee has relied on the results presented in the approved GE topical report NEDE-24011 " General Electric Standard Application for. Reactor Fuel (GESTAR II)," (Refere'nce 4).
In addition, the licensee submitted a supplemental reload
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licensing document (Reference 3) which provides the results of other analyses -
necessary to justify Cycle 11 operation but which are not included in GESTAR ~
l II.
3.0 EVALUATION 3.1 Fuel Design
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A total of 188 fresh GE type fuel assemblies of P80RB65H, P80RB299LA and i
l P80RB299H (References 1, 2, 3), which are pressurized 8x8 retrofit barrier
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fuel assemblies will be loaded for the Cycle 11 meration. Since the new pressurized 8x8 retrofit barrier fuel has been previously staff-reviewed and approved (Reference 5), we conclude that the fuel assemblies are
- acceptable for the Cycle 11 operation. The new fuel assemblies will~
reside with 372 irradiated 8x8 fuel assemblies of prior Exxon and GE designs presently in the core. The fuel designs for the irradiated assemblies of Exxon Type VB, and GE Types P80RB239 and P80RB265H have been l
previously approved and operated.
(References 4, 5) 3.2 Nuclear Design The nuclear design ~and analysis of the proposed reload has been performed by the methods described in Reference 5.
Reference 5 has been approved i
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3-for use in the design and analysis of reloads in BWR reactors and its use is acceptable for this reload. We have reviewed the results of the nuclear _ design analysis for the Oyster Creek Cycle 11 and have determined
-that-they are acceptable since the nuclear parameters are within the range of those normally obtained for similar cores and were obtained with acceptable methods.
3.3 Thermal-Hydraulic Design-The objective of the review of the thermal-hydraulic design of the Core Cycle 11 operation is:to confirm that the thermal-hydraulic design has been accomplished using acceptable methods, and to assure an acceptable margin of safety from conditions which could lead to fuel damage during normal operation and anticipated transients and to assure that the care is not susceptible-to thermal-hydraulic instibility.
The review included the following areas:
1 (1) Minimum Critical Power Ratio (MCPR) and-the related changes to the TS.
(2) Maximum Average Planar Linear Heat Generating Rate (MAPLHGR) and the related changes to the TS.
A safety limit value of MCPR is imposed to assure that 99.9 percent of the fuel rods in the core will not experience boiling transition during normal operation and anticipated operational transients.' As stated in Reference 2, the approved safety limit MCPR for the Oyster Creek P8x8R reload core is 1.07. The safety limit of 1.07 was used for the Cycle 11 analysis.
The licensee has proposed that two MAPLHGR curves for the fresh fuel bundles of P8DRB299 be added to the Oyster' Creek Technical Specifications to replace a reference to the same curves in a proprietary General Electric Topical Report (Reference 4). This is an administrative change which we find to be acceptable and appropriate.
4 3.4 Transient and Accident Analyses The corewide transient and accident analyses for Turbine Trip without i
Bypass, Loss of 100 F Feedwater Heating, Feedwater Controller Failure, MSIV Closure and Rod Withdrawal Errors (RWE) were performed using' approved methods described in Reference 4 and the results of the accident analyses are acceptable for Cycle 11.
(Appendix D to NED0-24195) i
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_ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ - - _ _ _ _ - The Turbine Trip Without Pypass was the most limiting transient for Cycle 11 with a maximum MCPR of 1.41 as calculated by ODYN option A and 1.36 by ODYN Option B.
This compares to 1.40 for Cycle 10 where the most limiting transient was the RWE (Rod Withdrawal Error) case. The licensee l
has conservatively selected an operating limit MCPR value of 1.45. This will not create any operating difficulties since Cycle 11 is expected to operate with an MCPR margin of 20% or greater.
3.5 Thermel-Hydraulic Stability The assurance that the reactor is stable and has adequate stability design margin is demonstrated analytically when a core stability decay ratio less than 1.0 is calculated using approved methods. For Oyster Creek Cycle 10, a limiting stability decay ratio of 0.67 for 0.87 including approved uncertainty values for the calculation method) was calculated (Reference 9). This was not reanalyzed for the Cycle 11 Core. However, operating conditions fcr Cycle 11 are essentially the same as for Cycle 10. Changes in stability margin due to the difference in core characterisitics are small. The licensee has concluded that there is sufficient margin to assure thermal-hydraulic stability for Cycle 11 (References 7, 8, 9). We Letter 86-02 (Reference 8)yster Creek (a BWR 2) in accordance with Generic find this e.cceptable for O 3.6 Technical Specifications Change The proposed revision of the APRM Scram and Rod Block lines to provide greater flexibility during startup and power escalation tc rated conditions was reviewed. The staff determined that the revision is acceptable since the methodology and procedures used are staff-reviewed and approved in~ References 3 and 4 The MCPR Limits are revised from 1.40 to 1.45 and maximum allowable average planar LHGR curves for 5 and 4 loop operations as described in Appendix D of. Reference 1 are added. The staff has reviewed the proposed TS changes for Cycle 11 and concludes that they are acceptable.
3.7 Extended Burnup Evaluation In response to a staff request, the licensee provided information that the Projected End of Cycle 11 Maximum Assembly Exposure is 29300 MWD /MTU
('leference 9). This GE fuel design has been previously approved and
operated well beyond this burnup range (Reference 5). Thus, extended burnup is not a factor in operation of the Cycle 11 core.
3.8 Fuel Performance In' Licensee Event Report (LER) No.86-016, dated July 30, 1986, the licensee reported fuel' clad failures associated with 47 fuel bundles
- during Cycle 10 operation.,(The offgas radiation level continually) and increased during the cycle from 50,100 mci /sec to 224,000 mci /sec the I-131/I-133 fission product ratio also increased (from.069 to.144).
All leaking fuel bundles and all other' fuel bundles in the same control cell were removed from the Cycle 11 reload. Nearly all of the failures (45 of 47) occurred in the same EXXON fuel batch. Although the licensee's investigation into the cause of the fuel. cladding failures is incomplete, it is believed that the failure mechanism involved defective cladding or pellet / clad interaction, possibly aggravated by failure to follow the fuel preconditioning recomendations of the fuel-supplier.
Concerns regarding the fuel failures and the progress of the licensee's investigation were reviewed by the NRC regional office as part of a safety inspection (Reference 10). The licensee identified several actions-intended to preclude the repetition of excessive fuel failures during Cycle 11, including:
(1) the acquisition of a new load line limit computer analysis which will permit maneuvering at high power by_ adjusting recirculation pump speed to minimize the contribution of rod position changes to pellet / clad interaction; (2) revision of the licensee's core monitoring computer program to eliminate errors which had made the program ineffective for conformance to fuel preconditioning operating recommendations; (3) the addition of a new computer program to track and trend the weekly offgas reactivity levels and the reactor coolant fission product ratios to aid in early identification of fuel failures during Cycle 11 operation.
In addition, the licensee has committed to keep the NRC informed of the
. fuel failure investigation results.
The staff concludes that the licensee is acting prudently to reduce the probability of fuel failures and to minimize the activity release in the offgas during Cycle 11 operation, and we find this acceptable. We will continue to follow the investigation of the Cycle 10 fuel failures and will pursue any additional actions that may be indicated with the licensee.
. L 3.9. Conclusion The staff concludes, as discussed above, that acceptable methods and procedures were used to perform the design and analysis of the Oyster Creek reactor reload for Cycle 11 operation and that the the licensee's proposed amendment is correctly based on the results of that design and analysis. Therefore, the staff concludes that this amendment is acceptable.
4.0 ENVIRONMENTAL CONSIDERATION
This' amendment involves a change-to a requirement with respect to 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 significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite and that there is no significant increase in individual or cumulative occupaticnal radiation exposure. The Commission has previously issued a proposed finding that this amendment involves no significant hazards consideration and there has been no public comment on such finding.
Accordingly, this amendment meets the eligibility criteria fcr categorical exclusion set forth in 10 CFR 51.22(c)(91. Pursuant to 10 CFR 51.22(b) no environmental impact statement nor environmental assessment need be ~
prepared in connection with the issuance of this amendment.
5.0 CONCLUSION
The staff has 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 this amendment will not be inimical to the common defense and security nor to the health and safety of the public.
6.0 REFERENCES
1.
Letter, P. B. Fiedler (GPUN) to J. A. 7wolinski (NRC), Oyster Creek Nuclear Generating Station Docket No. 50-219 Technical Specification Change Request (TSCR) No. 149, dated June 17, 1986.
2.
W. H. Hetzel (Oyster Creek) to R. B. Lee (GPUN), "0yster Creek Bundle Name Changes," dated February 28, 1986.
3.
Letter, Peter B. Fiedler (Oyster Creek) to John A. Zwolinski, Oyster Creek Nuclear Generating Station Docket No. 50-219, Technical Specifi-cation Change Request (TSCR) No.149. Revision 1 of Appendix D to NEDO-24195, dated September 17, 1986.
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" General Electric Reload Fuel Application for Oyster Creek," NEDO-24195 79NED288, Class I,-August, 1979.
5.
GESTAR II, General ' Electric Standard Application for Reactor Fuel, NEDE-24011-P-A-6-US Class III, April,1983, and its Proposed Amendment 13 (to Revision 6 to NEDE-24011-P-A) submitted September 24, 1985 and approved March 26, 1986.
6.
Telecopy from J. Lachermayer (GPUN) to J..Donohew (NRC), " Pro.iected End of Cycle 11 Fuel Exposures," dated ' July 29, 1986 7.
General Electric Service Information letter No. 380, Revision 1, February 10, 1984.
8.
Generic letter No. 86-02, " Technical Resolution of Generic-Issue B-19-Thermal Hydraulic Stability," January 23, 1986.
9.
. Letter, R' F. Wilson (GPUN)~to John A. Zwolinski, Oyster Creek Nuclear Generating Station Docket No. 50-219 Technical Specification Change Request (TSCR) No. 149, dated October 13, 1986,
- 10. NRC Region I Inspection Report Number 50-219/86-25 (on 4.0 Fuel Clad Failures) dated September.18, 1986.
Principal Contributor:
U. Cheh Dated:
October 27', 1986 p
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