ML16138A772
| ML16138A772 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 01/25/1994 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML16138A773 | List: |
| References | |
| NUDOCS 9402020366 | |
| Download: ML16138A772 (5) | |
Text
- Jk~ REG&
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.204 TO FACILITY OPERATING LICENSE DPR-38 AMENDMENT NO. 204TO FACILITY OPERATING LICENSE DPR-47 AND AMENDMENT NO. 201TO FACILITY OPERATING LICENSE DPR-55 DUKE POWER COMPANY OCONEE NUCLEAR STATION, UNITS 1. 2. AND 3 DOCKET NOS. 50-269, 50-270, AND 50-287
1.0 INTRODUCTION
By letter dated July 14, 1993, as supplemented August 24 and September 22, 1993, Duke Power Company, et al. (the licensee), submitted a request for changes to the Oconee Nuclear Station, Units 1, 2, and 3, Technical Specifications (TS).
The requested changes would revise TS 3.1.2.9 to clarify the role of High Pressure Injection and Core Flood Tank deactivation in maintaining pilot operated relief valve operability for low temperature overpressure protection (LTOP), add restrictions regarding applicability of controls which assure 10 minutes are available for operator action to mitigate an LTOP event, revise the pressure-temperature limits and associated LTOP setpoints, and make associated administrative changes. Also, the Bases would be revised to be consistent with the above changes. The August 24 and September 22, 1993, letters provided clarifying information that did not change the scope of the July 14, 1993, application and the initial proposed no significant hazards consideration determination.
1.1 Pressure-Temperature (P-T) Limits The existing P-T limit curves are revised because their applicability period, 15 effective full power years (EFPY), will be reached in February 1994, August 1994, and December 1994 for Units 1, 2, and 3, respectively. The proposed P-T limits are applicable for 21, 19, and 21 EFPY for Units 1, 2, and 3, respectively.
To evaluate the P-T limits, the staff used the following NRC regulations and guidance: Appendices G and H of.10 CFR Part 50; the ASTM Standards and the ASME Code, which are referenced in Appendices G and H; 10 CFR 50.36(c)(2);
Regulatory Guide (RG) 1.99, Rev. 2; Standard Review Plan (SRP) Section 5.3.2; Generic Letter 88-11; and Generic Letter 92-01.
Each Licensee authorized to operate a nuclear power reactor is required by 10 CFR 50.36 to provide TS for the operation of the plant. In particular, 10 CFR 50.36(c)(2) requires that limiting conditions of operation be included in the TS. The P-T limits are among the limiting conditions of operation in 9402020366 940125 PDR ADOCK 05000269 I
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-2 the TS for all commercial nuclear plants in the U.S. Appendices G and H of 10 CFR Part 50 describe specific requirements for fracture toughness and reactor vessel material surveillance that must be considered in setting P-T limits. An acceptable method for constructing the P-T limits is described in SRP Section 5.3.2.
Appendix G of 10 CFR Part 50 specifies fracture toughness and testing requirements for reactor vessel materials in accordance with the ASME Code and, in particular, that the beltline materials in the surveillance capsules be tested in accordance with Appendix H of 10 CFR Part 50. Appendix H, in turn, refers to ASTM Standards. These tests define the extent of vessel embrittlement at the time of capsule withdrawal in terms of the increase in reference temperature. Appendix G also requires the Licensee to predict the effects of neutron irradiation on vessel embrittlement by calculating the adjusted reference temperature (ART) and Charpy upper shelf energy (USE).
Generic Letter 88-11 requested that Licensees use the methods in RG 1.99, Rev. 2, to predict the effect of neutron irradiation on reactor vessel materials. This guide defines the ART as the sum of unirradiated reference temperature, the increase in reference temperature resulting from neutron irradiation, and a margin to account for uncertainties in the prediction method.
Appendix H of 10 CFR Part 50 requires the Licensee to establish a surveillance program to monitor embrittlement of reactor vessel materials. Appendix H refers to the ASTM Standards which, in turn, require that the capsules be installed in the vessel before startup and be removed from the reactor vessel periodically for testing. The capsules should contain test specimens made from plate, weld, and heat-affected-zone (HAZ) materials of the reactor beltline.
Generic Letter 92-01, "Reactor Vessel Structural Integrity," requires, from the Licensee, reactor materials information needed to assess compliance with the requirements in Appendices G and H to 10 CFR 50.
1.2 Low Temperature Overpressure Protection This change is to clarify the role of High Pressure Injection (HPI) and Core Flood Tank (CFT) deactivation in maintaining PORV operability for Low Temperature Overpressure Protection (LTOP). The licensee is requesting this change in accordance with the commitment for corrective actions in LER 270/92-03 and in response to Violation 92-08-01. In addition, the licensee is proposing to impose an additional TS requirement to control the high pressure nitrogen system to help ensure 10 minutes are available for operator action to mitigate an LTOP event.
2.0 EVALUATION 2.1 Pressure/Temperature (P-T) Limits The licensee calculated the ART for each beltline material in the Oconee Units 1, 2, and 3 reactor vessels in accordance with RG 1.99, Rev. 2. The limiting
-3 materials and associated calculation results for the proposed P-T limits are shown in Table 1. The staff confirmed that the initial RT and chemistry values in Table 1 are the same as the licensee's response o Generic Letter 92-01.
The staff performed an independent calculation and verified that the licensee's ART values are acceptable in accordance with RG 1.99.
Substituting the ARTs in Table 1 into equations in SRP 5.3.2, the staff verified that the proposed P-T limits for Units 1, 2, and 3 for heatup, cooldown, and leak test meet the requirements in Paragraphs IV.A.2 and IV.A.3 of Appendix G of 10 CFR Part 50.
In addition to beltline materials, Appendix G of 10 CFR Part 50 also imposes P-T limits based on the reference temperature for the reactor vessel closure flange materials.Section IV.A.2 of Appendix G states that when the pressure exceeds' 20% of the preservice system hydrostatic test pressure, the temperature of the closure flange regions highly stressed by the bolt preload must exceed the reference temperature of the material in those regions by at least 120OF for normal operation and by 90*F for hydrostatic pressure tests and leak tests. The staff has determined that the proposed P-T limits satisfy the closure flange requirement in Paragraph IV.A.2 of Appendix G.
Paragraph IV.A.1 of Appendix G requires that reactor vessel beltline materials maintain a Charpy upper shelf energy (USE) throughout the life of the vessel of no less than 50 ft-lbs unless it can be demonstrated that lower values of USE will provide margins of safety against fracture equivalent to those required by Appendix G of the ASME Code. The conformance of upper shelf energy to paragraph IV.A.1 will be determined pending the staff evaluation of licensee's response to Generic Letter 92-01.
The conformance of the surveillance program to Appendix H to 10 CFR 50 also will be determined under the staff evaluation of licensee's response to Generic Letter 92-01.
The NRC staff has performed an independent analysis to verify the licensee's proposed P-T limits. The staff concludes that the proposed P-T limits for Oconee Units 1, 2, and 3 are valid through 21, 19, 21 EFPY, respectively, because they conform to the requirements of Appendix G of 10 CFR Part 50.
Hence, the proposed P-T limits may be incorporated in the Oconee Units 1, 2, and 3 Technical Specifications.
The conformance of the upper shelf energy and reactor vessel material surveillance program to Appendices G and H will be determined pending the staff resolution of Generic Letter 92-01.
2.2 Low Temperature Overpressure Protection The licensee has submitted a proposal as a corrective action for Violation 92-08-01. The violation entailed deviation from LTOP procedures in that HPI pumps were energized, instead of deactivated as required by the Technical Specifications. Currently, if the second train of LTOP becomes inoperable (i.e., pressurizer is water solid) the TS requires that compensatory measures shall be provided within four hours to monitor for initiation of an LTOP
4-4
-4 event; however, this action is not sufficient. Additional action to deactivate the HPI system and CFTs is also necessary. If the HPI system is not deactivated or the CFTs are not deactivated, the PORV (i.e., the first train of LTOP) may not have sufficient relief capacity to mitigate postulated LTOP events.
To ensure that the HPI and CFTs are deactivated, the licensee proposed two changes to the TS. The first proposed change is a revision to TS 3.1.2.9 requiring that the HPI trains and the CFTs be deactivated for the two trains of LTOP to be operable. The second proposed change is an addition, specified as TS 3.1.2.9.5.a, to require the immediate deactivation of the HPI trains and CFTs if the HPI and CFTs have not already been deactivated when the LTOP trains are required. These TS changes are valid when the following conditions exist: (1) the temperature of one or more of the RCS cold legs is 5 325 "F, and (2) an RCS vent path capable of mitigating the most limiting LTOP event is not open.
The second train of LTOP consists of controls to ensure that 10 minutes are available for operator action to mitigate an LTOP event. The licensee also proposed a new line item, i.e., TS 3.1.2.9.4.e, to control the high pressure nitrogen system to ensure 10 minutes are available when there is no other vent path open that is capable of preventing or mitigating the most limiting LTOP.
Finally, the licensee proposed the addition of TS 3.1.2.9.5.c which specifies the required actions when the second LTOP train is inoperable and compensatory measures have not been provided within the required 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
All of the changes discussed above constitute additional restrictions not presently included in the TS. As stated by the licensee, these changes do not affect the design basis accident analyses since they are not contributors to the accident analyses addressed in the Oconee Final Safety Analysis Report.
2.3 LTOP Setpoints The licensee evaluated the current TS LTOP limits against the revised P-T limits and the new P-T limits were found to be equally or less restrictive than the present limits at almost all RCS temperatures below 325 "F. The evaluation also demonstrated that the Unit 3 PORV setpoint, 480 psig, and RCS pressure limit for temperatures < 220 OF are conservative and applicable to Units 1 and 2. The setpoints were lowered on Units 1 and 2 based on the Unit 3 analysis, which is applicable to Units 1 and 2 and therefore acceptable. Finally, the evaluation demonstrated that the 325 OF enable temperature remains conservative relative to the RTNDT + 90 OF enable temperature criterion.
The licensee has proposed changes to the TS and associated Bases for the LTOP system. Those TS that were modified include TS 3.1.2.9, 3.1.2.9.2 (3.1.2.9.4 in the new TS), 3.1.2.9.5.a (new), and 3.1.2.9.3 (3.1.2.9.5 in the new TS) and the associated Bases. These changes are in response to LER 270/92-03 and violation 92-08-01 and impose additional restriction to the current TS. These changes modify the current TS in such a way as to more clearly define the role of High Pressure Injection and Core Flood Tank deactivation in maintaining
-3 PORV operability for LTOP, and therefore address the problem that lead to violation 92-08-01.
For the above reasons the staff finds the changes acceptable.
3.0 STATE CONSULTATION
In accordance with the Commission's regulations, the South Carolina State official was notified of the proposed issuance of the amendments. The State official had no comments.
4.0 ENVIRONMENTAL CONSIDERATION
The amendments change requirements with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendments involve 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 occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding (58 FR 46228 dated September 1, 1993).
Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.
5.0 CONCLUSION
The Commission 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, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributors: S. Brewer J. Tsao Date:
January 25, 1994