ML20096G232
| ML20096G232 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 01/18/1996 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20096G223 | List: |
| References | |
| NUDOCS 9601250052 | |
| Download: ML20096G232 (4) | |
Text
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'4 UNITED STATES
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NUCLEAR REGULATORY COMMISSION j
WASHINGTON. D.C. 3088H001 o
t, SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT N0.195 TO FACILITY OPERATING LICENSE NO. DPR-65 NORTHEAST NUCLEAR ENERGY COMPANY THE CONNECTICUT LIGHT AND POWER COMPANY THE WESTERN MASSACHUSETTS FLECTRIC COMPANY NILLSTONE NUCLEAR POWER STATION. UNIT NO. 2 00CKET NO. 50-336
1.0 INTRODUCTION
29, 1995, as supplemented November 9,1995, the By letter dated September Northeast Nuclear Energy Company (NNECO/the licensee) submitted a request for changes to the Millstone Nuclear Power Station, Unit No. 2 Technical Specifications (TS). The requested changes would provide three changes to the TS relating to the pressurizer safety valves (PSV) and the main steam safety valves (MSSV).
L The first change is to TS 3.4.2.1 and 3.4.2.2 and involves relaxing the as-found setpoint tolerance for the PSVs and the MSSVs from the current value of i
11% to 13%. Table 4.7-1 is also modified to correct the as-found tolerance for the MSSV from fl% to 13%. Notes are added to TS 3.4.2.2 and Table 4.7-1 which specify that the lift setting should be detemined at nominal operating conditions and should be set at *1% of the lift setting.
For the second change, Surveillance Requirement 4.7.1.1 and Table 4.7-1 are modified to eliminate the need to verify the orifice size of each MSSV.
The third change modifies the statement for TS 3.7.1.1 so that if a MSSV is inoperable and compensating action cannot be taken, the plant must be brought to hot shutdown (Mode 4) within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> instead of cold shutdown (Mode 5) in 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
The November 9,1995, letter provided clarifying information that did not change the initial proposed no significant hazards consideration determination.
9601250052 96o11s PDR ADOCK 05000336 P
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2.0 BACKGROUND
j Overpressure protection for the reactor coolant system (RCS) and the main steam system is provided, in part, by the PSVs and the MSSVs located on the pressurizer and on the four main steam lines, respectively.
The PSVs are credited for mitigation of RCS overpressurization events. The limiting RCS overpressurization event is the loss of electrical load.-
TS 3.7.1.1 requires all MSSVs to be operable in Modes 1, 2, and 3.
If the MSSVs are not operable, the action statement specifies that operation in Modes 1, 2, and 3 may continue provided that either the inoperable valves are restored to operable status or the high power level trip setpoint be reduced per Table 3.7.1 (which allows up to three of the eight valves on any one steam generator to be inoperable). Otherwise, the plant is required to be in Mode 3 within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in Mode 5, cold shutdown, within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. The proposed modification is to require that the plant be in hot shutdown (Mode 4) within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> instead of cold shutdown (Mode
- 5) in 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
3.0 EVALUATION In the telecon of November 1,1995, the staff had informed NNEC0 that in similar proposals by licensees with plants designed by Combustion Engineering, i
the Feedwater Line Break Event was frequently determined to be the limiting pressurization event for the RCS. NNECO's letter of November 9,1995 stated that the limiting RCS pressurization event for Millstone Unit 2 is the Loss of Load Event, as described in Section 14.2.1 of the FSAR. As stated in Section 14.2.8 of the FSAR, the Feedwater System Pipe Break Event is "not in the current licensing basis for Millstone Unit 2 and, therefore, is not analyzed."
The loss of electrical load was reanalyzed by NNEC0 with a +3 percent tolerance from the nominal setpoints for the pressurizer safety valves and the main steam safety valves. The analysis showed that RCS pressure remains below the ASME allowable of 110 percent of design. NNECO also stated that the RCS 1
pressure used for accidents where minimum departure from nucleate boiling ratio (DNBR) is a concern bounds the -3 percent lower bound on the PSV nominal setpoint. Thus, the allowance of a 13% tolerance has no impact on the pressure limit or minimum DNBR for the limiting DNBR transients. The staff, therefore, finds the 135 tolerance limit to be acceptable for the PSVs.
NNECO reanalyzed the steam generator tube rupture (SGTR) event to take into account the i3% as-found tolerance and to extend the margin for operator action to I hour. The results indicated that the combined effect of extended releases and a 13% tolerance on the MSSV setpoint has only a small effect on the calculated offsite doses which were found to be a small fraction of the 10 CFR Part 100 acceptance criteria.. The staff, therefore, finds the results for the reanalysis of the SGTR event to be acceptable.
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- l' l l By the letter of November 9, 1995, NNECO informed the staff that, with the l
exception of the SGTR event, the events analyzed in the Safety Analyses i
section (Chapter 14) of the Millstone Unit 2 FSAR already take into account a i
l 13 percent tolerance for the PSVs and MSSVs. This is verified in Table j
14.0.9-1 of the FSAR. As a result, the analysis provided with the proposed revision to the TS was limited to the change to the analysis of the SGTR l
}
event. Upon approval of the proposed TS change, NNEC0 stated that the new results of the SGTR event will be incorporated into the FSAR. The staff, l-therefore, finds that NNEC0 has provided acceptable reanalyses for the 1
j proposed changes and, therfore, the proposed changes are acceptable.
NNECO proposed to eliminate the orifice diameter of 4.515 square inches listed f
for the steam line safety valves in Table 4.7-1.
NNEC0 stated that the MSSV orifice size represents the smallest inside diameter of the safety valve i
nozzle, an internal part of the valve. The orifice diameter is not adjustable and can only be changed by replacement of the nozzle. Replacement of the i
nozzle requires removal and disassembly of the safety valve.
Further, only one size nozzle is available for these safety valves. Since there is no adjustment possible to the orifice size, and changes to the orifice requires a modification of the valve that would be covered under the design change The process, the specification of the orifice size in the TS is unnecessary.
staff agrees with this since removing the specification will have no impact on the plant configuration or operation and the safety analysis is unaffected by the change.
The proposed modification to require that the plant be in hot shutdown (Mode
- 4) within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> instead of cold shutdown (Mode 5) in 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> is on the following basis. The limiting condition for operation (LCO) does not require the MSSVs to be operable in Mode 4 (RCS average temperature less than 300 'F but greater than 200 'F). Thereferc, the action statement is being changed to be consistent with the LCO. This change.is also consistent with NUREG-1432, " Standard Technical Specifications - Combustion Engineering Pl ants." The staff, therefore, finds this modification to be acceptable.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Connecticut State official was notified of the proposed issuance of the amendment. The State official had no comments.
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 and changes surveillance requirements. 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 previously issued a proposed finding that the amendment involves no significant hazards
consideration, and there has been no public comment on such finding (60 FR 54723). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR SI.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 amendment.
6.0 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 amendment will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributor:
H. Balukjian Date: January 18, 1996 I