ML20199L179
| ML20199L179 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 11/19/1997 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20199L176 | List: |
| References | |
| NUDOCS 9712010334 | |
| Download: ML20199L179 (3) | |
Text
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j NUCLEAR REGULATORY COMMISSION WASHINGTON. D.C. 20666 0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR '1EGULATION RELATED TO AMENDMENT NO. 211 TO FACILITY OPERATING LICENSE NO. OPR-p1 NORTHEAST NUCLEAR ENERGY COMPANY THE CONNECTICUT LIGHT AND POWER COMPANY THE WESTERN MASSACHUSETTS ELECTRIC COMPANY MILLSTONE NUCLEAR POWER STATION. UNil.NO. 2 DOCKET NO. 50-336
1.0 INTRODUCTION
By letter dated September 16, 1997, the Northeast Nuclear Energy Company, et al. (NNEC0/ licensee) submitted a request for changes to the Millstone Nuclear Power Station, Unit No. 2, Technical Specifications (TSs). The proposed changes would modify the T3s for the main steam line American Soc'ety of Mechanical Engineers Code (Code) safety valves, hereinafter referred to as Code safety valves.
Specifically, NNECO proposes to: (1) delete TS Table 3.7.1, " Maximum Allowable Power Level-High Trip Setpoint with Inoperable Steam Line Safety Valves During s
Operation with Both Steam Generators," by not allowing operation in Mode 1 or 2 with inoperable Code safety valves while allowing operation in Mode 3 with up to three Code safety valves inoperable per steam generator, (2) modify the associated action statement in TS 3.7.1.1 to reflect tne operational thanges, and (3) update the TS Bases to reflect the proposed changes and include the correct amendment histary numbers to reflect previously approved c
amendments.
2.0 BACKGROUND
Overpressure protection for the steam generators (shell side) and the main steam line 91 ping (up to the turbine stop valves) is provided by 16 spring loaded Code safety valves.
Each of the two steam generators has eight Code safety valves that are designed to limit the pressure to 110 percent of the design pressure. These Code safety valves also provide reactor core heat removal and design-basis accident mitigation.
During its effort to verify the current design and licensing bases for Millstone, Unit 2, NNEC0 has determined that the current maximum allowable power level high trip setpoints with inoperable Code safety valves specified
'n Table 3.7-1 of TS 3.7.1.1 are incorrect. The trip setpoints were not changed to be consistent with a previously approved reduction in the maximum 9712010334 971119 PDR ADOCK 05000336 P
. power level high trip setpoint.
In addition, NNECO is also in the process of reanalyzing the inadvertent closure of the main steam isolation valve (MSIV) and the loss of electrical load (LOEL) events. The results of the reanalysis indicate that the MSIV event results in the highest peak pressure in the secondary system and that the formula currently contaned in the TS Bases for TS 3.7.1.1 may not result in the corract trip setpoints.
3.0 IVALUATION The proposed deletion of TS Table 3.7-1 and changes to TS 3.7.1.1 will remove the ability to continue to operate in Mode 1 or 2 with inoperable Code safety valves while allowing operation in Mode 3 with up to three Code safety valves inoperable per steam generator.
Four hours is still allowed to restore the inoperable Code safety valve (s) to operable status before power reduction to Hot Standby (Mode 3) is required while in Modes 1 and 2.
Continuous operation
.in Mode 3 is allowed provided no more than three Code safety valves per steam generator are inoperable.
If more than three Code safety valves on a single steam generator are inoperable the reactor must be in Hot Standby within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and Hot Shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
The operability of the Code safety valves ensures that the secondary system pressure will be limited to within 110 percent (1100 psig) of the design pressure of 1000 psig during the most severe anticipated operational transierit.
By letter dated October 6, 1980, the NRC issued Amendment No. 61 to Facility Operating License No. OPR-65 for the Millstone Nuclear Power Station, Unit No. 2.
The Amendment, among other things, modified the power level 5igh trip setpoint from 107 percent te 106.6 percent. The TS Bases reflected the change; however, the setpoint values in TS Table 3.7-1 were not recalculated to reflect the new maximum power level. Thus, the current values in the table are incort$ct for Modes 1 and 2 operation.
In addition, the reanalysis of the MSIV and L0EL events indicates that the formula currently contained in TS Bases 3/4.7.1.1 may not be conservative for establishing the reduced power level high trip setpoints.
NNECO has verified that oPration in Mode 3 with up to three inoperable Code safety valves per steam generator is acceptable. The reactor is at least 1 percent suberitical when operating in Mode 3.
During operation in Mode 3, the power level high trip setpoint is an automatically variable setnint and is at approxiuately 15 percent. The remaining five Code safety vah es per steam generator are chpable of removing the maximum possible decay heat load and maintaining the secondary system within 110 percent of the system design pressure for the most severe anticipated operational transient. NNEC0 also notes that the ability to operate in Mode 3 with inoperable Code safety valves provides f1mibility for maintenance or repairs on the valves.
NNECO noted that previous Amendment Nos. 52, 61, and 63 had resulted in changes to the TS Bases, pages B 3/4 7-1 and B 3/4 7-2, but had not been added to the pages as amendment numbers or retained as history numbers. This inadvertent error will be corrected by including the appropriate amendment history numbers.
B
. Therefore, on the basis of the previous discussion, the NRC staff has determined that the deletion of TS Table 3.7-1 and the proposed changes to TS 3.7.1.1 are acceptable. The staff has also determined that the changes to TS Bases pages B 3/4 7-1 and B 3/4 7-2 adequately reflect the changes and previous amendment history numbers.
4.0 STATE CONSULTATION
in accordance with the Cominion's regulations, the Connecticut State official was notified of the pNposed issuance of the amendment. The State r.,fficial had no coments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement 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 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 occupational radiation exposure.
The Commission has previoualy issued a proposed finding that the amendment involves no significant hazards consideration, and there n.u been no public commant on such finding (62 FR 52582 dated October 8, 1997).
Accordingly, the amend.nent meets the eligibility criteria for categorical exclusion set forth in 10 CFR 5;.22(c)(9).
Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared ;a connection with the issuance of the amendment.
6.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that:
(1) there is reason ele assurance th t the health and safety of the a
public will not be endangered by operation in the proposed manner, (2) such activities will be ccaducted in compliance with the Commission's regulations, and (3) the issuance of the amendment wili not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributor:
D. Mcdonald Date: November 19, 1997
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