ML20087B441
| ML20087B441 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 01/02/1992 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20087B439 | List: |
| References | |
| NUDOCS 9201130131 | |
| Download: ML20087B441 (6) | |
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asog Io, UN11 ED STATES
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NUCLEAR REGULATORY COMMISSION
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-a WASHINGTON, D. C. 20555
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT N05.132 AND 111 TO FAC'LITY OPERATING LICENSE N05. DER-70 AND DPR-75 PUBLIC SERVICE tLECTP.IC & GAS COMPANY PHILA9ELPHIA ELECTRIC COPPANY l
DELMARVA POWR H!D LIGHT 07 @ANY i
ATLANTIC CITY ELECTRIC COMPANY SALEP MUCLEAD. GENERATING STATION, UNIT NOS. 1 AND 2 DOCKET N05. 50-??? AND 60-311 INTFCDUCTION letter dated April 2, 1990, the Public Service Electric and Gas Company, padelphia Electric Company, Delmarva Power and 1ight Company and Atlantic Flectric Company (the licensees) submitted a request for changes to the im Nuclear Generating Station, Unit Nos. I and ?, Technical Specifications
).
The reonested changes would increase the allowable isolation times
)ciated *:;th the f eedwater control valves in TS Tabic 3.3-5 and 3.6-1.
The iges are proposed due to difficulty in meeting the current TS response time airements and to be consistent between Units 1 and 2 for functionally 1tical feedwater systems. Specifically, the licensees propose to increase respoese time in Table 3.3-5 from 7 seconds or less to 10 seconds or less
- all feedwater isolation functions except for steam flot in two steam lines a coincident with loop average temperature (Tayg) low-low.
Fce the above
-m flow in two stum lines, a response time of 15 seconds or less is iosed from the present 10.75 seconds or less because of Tavg total sensor time of 5 seconds.
In Table 3.6-1, the licensees propose to change the lwater control valve response time associated with the contsinn.ent isolation
- tion to 9 seconds or less from the current 5 seconds or less for Unit 1 and conds or less for Unit 2.
The proposed revised closure times acknowledge time requirements associated with the electronics and ensures that the neered Safety Features Actuation System (E0FAS) resprpse time is not 4eded.
EVALUATION hstrument Response T m Except for the new Resistance Temperature Datectors (RTDs) installed in the primary coolant system hot legs and r legs to determine Tavg, the instrument response times are the same e Ae assumed in the licensees'
' approved licensing analysis. The liccmsee state that the Tavg RTDs kkh6d? /$$
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. have a 5 second total sensor lag (response) time. This is consistent with the licensees' submittal to the NRC dated April, 1987 (Pef. 1) which provided supporting documentation for the Saleni Unit I and 2 RTD Bypass Manifold removal project. As part of its review, the staff found tie RTD response time to be acceptable (Ref. 2). Consequently, the staff accepts the licensees' use of 5 seconds for tF Tavg RTD response time.
The licensees assunte the electronics components have a response time of one second. This assumption is consistent with the value used in the licensees' approved licensing analyses.
The instrun:ent response time and electronics respense time portions of the licensees' request for TS revision are consistent with previously approved licensing analyses.
Consequently, the staff finds the instrumentation and control system; r.spects of the licensees' submittal to be acceptable.
B.
l.oss of Coolant Accident (LOCA) and Non-LOCA Analysis Westinghouse performed a safety analysis to determine if an increase in feedwater control valve closure time could be supported by the current licensing basis safety analysis. Westinghouse evaluated the effect of the increase in feedwater -/alve closure times for LOCA and non-LOCA analyses.
In addition, an analysis of the consequences of a compicte tailure of a feedwater control valve to close was also performed by Westinghouse.
(1) Increase in feedwater valve cicsure time.
During small and la.'ge break LOCAs, an extension in the time required to isolate feedwater would increase the decay heat removal capability slightly and result in a small benefit during these events. The failure of-a feedwater control valve to close results in the same small benefit and it is bounded by the sing 1s failure assumed in the Salem licensing basis LOCA analysis.
Past analyses performed for steamline break core protection purposes indicate that a small increase in core power (maximum of 1%) would result due to the increase in feedwater control valve closure time.
The departure from nucleate boiling ratio (DNBR) penalty associated with this slight core power ir. crease does not exceed the design limit value of DNBR. Thus, the consequences and conclusions of the existing Salem steamline break core protection analysispre still applicable.
(2) Failure of a feedwater control valve to :: lose.- (This is an additional evaluation performed by Westinghouse for this anendment.)
The design basis steamlire break core analysis currently assumes the limiting single failure of a safeguard train, which minimizes the boron injection tapability to terminate the event.
If the single j
failure was assumed to be the failure of a feedwater control-valve to close, a 30-second delay in feedwater isolation would be imposed because this is the closure time for the feedt:ater isolation valve which is in series with the ferdwater control valve. Continued feedwater addition at a rate of 125% of full feedwater flow for 30 seconds was evaluated. The results showed that the positive reactivity insertion resulting from the additionel cooldown prior to_feedwater isolation would be less than the negative reactivity from boron injection provided by a second safeguard train. Therefore, the single failure of the feedwater control valves to close would be less limiting than the failure of a safeguard train.
In summary, the conclusions of the current Salem licensing basis _ analysos for LOCA and non-LOCA events would be unchanged if the feedwater isolation ESFAS response time was increased as proposed. The single failure of a feedwater control valve to close is bounded by the single failure assumptions used for the S&iem licensing basis LOCA'and non-LOCA related analyses.
21 sed on the licensees' evaluation of LOCA and non-LOCA events for an increase in feedwater isolation control valve isolation response times, the staff concludes that the proposed TS changes are acceptable.
C.
Containment Integrity Analysis The licensees indicated that the-Salem design basis containment analyses-considered ths short and long-term mass and energy release for postulated LOCAs, containment response analyses following a LOCA or steamline break inside containment, and subecmpartment pressure transient analyses.
The licensees stated that increasing the feedwater control valve closure time would have no effect on the calculated results for short-term ness and energy-release and subcompartment pressure analyses because the-transient has a duration of 5 seconds or less. The long-term mass and energy release and containment pressure response fellowing a LOCA would improve with increased feedwater isolation closure times because of the reduction in steam generator secondary side temperature as-the mass increases and thus it will reduce secondary to primefy heat transfer occurring during a LOCA. -The staff agrees with the above discussion that the increased closure time will have no negative effect on the
~l short-term and long-term mass and= energy releases, shott-term subcompartment analysis and the containment pressure. response following a postulated LOCA.
The licensees indicated that the increase in valve closure time can affect the steamline break containment analysis slightly. The current Salem design basis containment analysis include multiple failure assumptions.
The existing most limiting containment pressure occurs for a 0.944 square
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feet split ru valve (MSIV) pture at 30% power, with tie failure of a main steam isolation-and a containment safeguard train, resulting in a peak pressure of 46.4 psig. The most limiting analyses were reevaluated with the feedwater closure time increased to 10 seconds with all single failures.
This resulted in a peak pressure of 46.53 psig. Therefore, the containment pressure will be maintained below the design pressure of 47 psig for all single failures analyzed.
The licensees also indicated that the existing most limiting containment temperature occurs for a 0.6 square feet double ended rupture initiated at 102% power, with failures of an MSIV, feedwater control valve, feedwater pump runout protection, and a containment safeguard train.- The associated peak temperature is 345.5 'F.
The most lim Ding analyses were reevaluated with feedwater control valve closure time increased to 10 seconds with all single failures. This resulted in a peak temperature of 338.3 'F.
Therefore, the containment temperature will be maintained below 340 'F for all single failures analyzed.
Based on the above discussion, the~ stiff agrees that the )roposed increase in feedwater control valve closure time does not affect tie containment integrity as the containment design pressure and temperature will not be exceeded.
3.0 STATE CONSgtTATION-1 in-accordance with the Comission's regulations,-the New Jersey State official was notified of the proposed issuance of the. amendments. The-State official.
had no comments.
4.0 ENVIRONMENTAL CONSIDEPATION The amendraents change-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 amendments; involve no-significant increase in the amounts, and no significant change i_n the types, of any effluents that may be released offsite, and that there-is no-significant increase in individual or cumulative occupatSnal 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 f!nding (56 FR 51930).: Accordingly, the amendments-l meet the eligibility cr.%ria for categorical exclusion set forth' in 10 CFR 51.22(c)(!).- 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 endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Connission'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:
M. Waterman, SICB R. Goel, SPLB K. Desai, SRXB Date: January 2,1992 4
9 6-RFFERENCES 1.
Licensing Report S-87-05, " Licensing Report for New Narrow Range Temp ;J: ire Measurement System (RTL Bypass Elimination, PSE&G, Salem, Unit; 1 end 2," April, 1987.
2.
Safety Evaluation that accompanied f.mendments 84 and 56 for Selem, Units 1 and 2, respectively, " Technical Specification Changes Due to RTD Bypass System Modifications" dated November 16, 1987.
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