ML20248D404
| ML20248D404 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 05/28/1998 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20248D402 | List: |
| References | |
| NUDOCS 9806020424 | |
| Download: ML20248D404 (4) | |
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SAFETY EVALUATION BY THE OFFICE OF NUCI FAR REACTOR REGULATION RELATED TO AMENDMENT NO. 125 TO FACILITY OPERATING LICENSE NO. DPR-80 AND AMENDMENT NO. 123 TO FACILITY OPERATING LICENSE NO. DPR-82 PACIFIC GAS AND ELECTRIC COMPANY DIABLO CANYON NUCI FAR POWER PLANT. UNITS 1 AND 2 DOCKET NOS. 50-275 AND 50-323
1.0 INTRODUCTION
By application dated December 23,1997, Pacific Gas and Electric Company'(or the licensee) requested changes to the Technical Specifications (Appendix A to Facility Operating License Nos. DPR-80 and DPR-82) for the Diablo Canyon Nuclear Power Plant, Units 1 and 2. The proposed changes change the combined Technical Specifications (TS) for the Diablo Canyon Power Plant (DCPP) Units 1 and 2 to revise TS 3/4.7.11, Table 3.7-1," Maximum Allowable Power Range Neutron Flux High Setpoint With inoperable Steam Line Safety Valves." The power range (PR) neutron flux high setpoints would be revised based on revised calculational methodologies for 1,2, or 3 inoperable main steam safety valves (MSSVs) per steam generator (SG). The proposed TS change would lower the PR neutron flux high setpoints when 2 or 3 MSSVs are inoperable per loop such that the maximum power level allowed would be within the heat removing capability of the remaining operable MSSVs. Although the method for calculating the maximum power level allowed when one MSSV per loop is inoperable has been revised, the results have not and the limit remains the same. The associated Bases are also revised.
By letter dated January 20,1994, Westinghouse Electric Corporation (Westinghouse) issued Nuclear Safety Advisory Letter (NSAL)94-001, which notified the licensee of a deficiency in the basis of the TS 3/4.7.1.1, which allows the plant to operate at reduced power levels with a specified number of MSSVs inoperable. Westinghouse determined that the maximum allowable power range neutron flux high setpoints given in TS Table 3.7-1 may not be low enough to prevent a secondary side overpressurization during a loss of load / turbine trip. In NSAL 94-001, Westinghouse reported their determination that the maximum allowable initial power level is not a linear function of available MSSV relief capacity, it was further determined that the current TS provisions for reduced reactor power levels with inoperable MSSVs may not preclude the secondary side pressure from exceeding 110 percent of its design value during a loss of main feedwater transient, particularly at lower power levels. The licensee proposed to amend this section of the TS as recommended by Westinghouse to specify the correct maximum allowable reactor thermal power operation with inoperable MSSVs.
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2.0 DESCRIPTION
AND EVALUATION There are 20 MSSVs on each unit or 5 for each SG. The MSSVs provide emergency pressure relief for the SGs following an imbalance between the reactor power and turbine load. The MSSVs are designcd to prevont steam pressure from exceeding 1210 psia, or 110 percent of the 1085 psig SG design pressure. The MSSVs have a rated capacity of 105 percent of the design steam flow at an accumulation pressure net exceeding 1210 psia.
With a reduced number of operable MSSVs, TS 3.7.1.1, Action a., allows the plant to operate at a reduced power level as determined by TS Table 3.7-1. This reduction in the allowed power level is dependent upon the number of inoperable MSSVs, and is based on the reduced heat removal capacity.
The events that challenge the relieving capacity of the MSSVs are those characterized as decreased heat removal events, of which loss-of-load / turbine trip (LOUTT) is the limiting anticipated operational occurrence and may result in overpressurization of the main steam system when operating in accordance with TS 3.7.1.1. This event is analyzed in Chapter 15.2.7 of the Final Safety Analysis Report (FSAR).
Westinghouse states in NSAL 94-001, that at lower initial power levels a reactor trip may not be' actuated early enough in the transient. An overtemperature AT trip is not generated since the core thermal margins are increased at lower power levels. A high pressurizer pressure trip is not generated if the primary pressure control system functions normally. The reactor eventually trips on low steam generator water level, but this may not occur before steam pressure exceeds 110 percent of the design value if one or more MSSVs are inoperable in accordance with the current TS. This may occur due to the longer time in which primary heat is transferred to the secondary side before the trip.
Westinghouse recommended reducing the maximum power level (heat input from the primary system) allowed for operation with inoperable MSSVs below the heat removing capability of the operable MSSVs. This will prevent exceeding 110 percent of the design pressure of the secondary system.
Using RETRAN-02 and the Westinghouse algorithm provided in NSAL 94-001, the licensee calculated reduced power levels for TS Table 3.7-1. The aew PR neutron flux high setpoints of 53 percent and 35 percent rated thermal power (RTP) (without instrument and channel uncertainties), corresponding to 2 and 3 MSSVs unavailable per SG,'respectively.
The MSSVs are assumed to have two active and one passive failure modes. The active failures are spurious openbg and failure to close once opened. The passive failure mode is failure to open on demand. At Diablo Canyon, the passive failure is not postulated to occur for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following an accident. Therefore, when developing the algorithm and determining the required reduction in power level, Westinghouse does not assume that an MSSV falls to open on demand after the PR neutron flux high trip setpoint is reduced.
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The original algorithm was based on a linear function which did not bound the actual power versus steam flow curve at all points. The revised algorithm bounds the curve at all points and results in conservative PR neutron flux high trip setpoints. The revised Westinghouse algorithm calculated a maximum allowable PR neutron flux setpoint of 89 percent RTP (without instrument and channel uncertainties) when all MSSVs are available. To resolve this issue, PG&E performed an overpressure protection analysis using RETRAN-02.
The limiting FSAR Condition ll accident for overpressure concems is a LO11TT. This event was reanalyzed using RETRAN-02 for 1 MSSV inoperable per SG. The analysis indicated that the remaining 4 operable MSSVs per SG are sufficient to protect the SG and main steam system from overpressurization. Therefore, there is no need to reduce the maximum allowable power for one inoperable MSSV per SG. However, the licensee proposed to maintain the current maximum ellowable PR neutron flux high idp setpoint of 87 percent RTP for one inoperable MSSV per SG. The 13 percent RTP conservatism envelopes instrument and channel uncertainties, which are estimated to be 6 percent in Westinghouse WCAP -11082, Revision 5,
" Westinghouse Setpoint Methodology for Protection Systems, Diablo Canyon Units 1 and 2,24 Month Fuel Cycle Evaluation."
l The licensee recalculated the maximum PR neutron flux high trip setpoints for 2 and 3 inoperable MSSVs based on the revised algorithm. Incorporating the instrument and channel inaccuracy of 6 percent RTP, the maximum allowable PR neutron flux high trip setpoints would be reduced to 47 percent RTP and 29 percent RTP for 2 and 3 inoperable MSSVs per SG, respectively.
The staff has reviewed the TS changes and finds that the new setpoints for the neutron flux high reactor trip setpoints are lower, more conservative, and provide a greater margin of safety than the current setpoints and are therefore acceptable.
3.0 STATE CONSULTATION
in accordance with the Commission's regulations, the Califomia State official was notified of the proposed issuance of the amendments. The State official had no comments.
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4.0 ENVIRONMENTAL CONSIDERATION
l These amendments change a requirement with respect to the installation or use of a facility j
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 ir dividual or cumulative occupational radiation exposure. The Cornmission has previously issued a proposed finding that the amendments involve no I
significar,t hazards consideration, and there has been no public comment on such finding l
(63 FR 19975). 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.
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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 endangerod 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 Contributor: S. Bloom Date:
May 28, 1998 1
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