ML20215K009
| ML20215K009 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 05/06/1987 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20215J998 | List: |
| References | |
| NUDOCS 8705110030 | |
| Download: ML20215K009 (4) | |
Text
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[k/AQ ONITED STATES
?o NUCLEAR REGULATORY COMMISSION
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.E WASHINGTON, D. C. 20555
. SAFETY EVALUATI0ll BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 50 TO FACILITY OPERATING LICENSE NO. NPF-11 AND AMENDMENT NO.33 TO FACILITY OPERATING LICEllSE NO. NPF-18 COMMONWEALTH EDISON COMPANY LA SALLE COUNTY STATION, UNITS 1 AND 2 DOCKET N05. 50-373 AND 50-374
1.0 INTRODUCTION
By letter dated October 23, 1986, (Ref. 1), as supplemented November 5, 1986 (Ref. 2) and March 6, 1987 (Ref. 3), Commonwealth Edisco Company (licensee) proposed a change to the La Salle. County Station, Units 1 & 2 Technical Specifications (TS). The main purpose of this change is to re-duce challenges to the safety relief valves (SRVs). This would be ac-complished by changing the water level setpoint for closure of the rhain steam isolation valves (MSIVs) and main steam line drain valves (MSLDVs) from level 2 to the lower Level I.
The probability of closing.an MSIV due to variation of water level following a scram would thus be reduced.
With the MSIVs open and the main condenser available, prime mover steam fer the reactor feed pump turbines will also remain available to allow continued operation of the condensate and nain feedwater system. This will aid in water level recovery before the low level isolation setpoint is reoched.
Should Level I be reached, MSIV closure will cause reactor pressure to rise causing the SRVs to open and discharge to the suppression pool.
The new heat load for the suppression pool will be reduced by the amount of heat picked up by the main condenser.
The MSIVs fail closed upon loss of pneumatic (nitrogen) supply. The nitrogen supply is presently isolated from the MSIVs at Level 2; and, therefore, its setpoint will also have to be lowered to isolate at Level 1.
The Novenber 5, 1986 and the March 6, 1987 submittals provided analyses to support the proposed changes. Thus the action noticed in the Federal Register on November 19, 1986 and the staff's prcposed no significant hazards consideration determination were not affected by these submittals.
2.0 EVALUATION The licensee request to lower the water level trip setpoint for MSIV closure is consistent with staff solutions for NUREG-0737, item II.K.3.16,
" Reduction of Challenges and Failures of Relief Valves".
In addition, the General Electric Infornation Letter, SIL No. 367, also recorrends the lower trip setpoint. The staff evaluation of the change in setpoint is discussed below.
8705110030 870506 PDR ADOCK 05000373 P
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.> ce y, 2.1 LOSS OF COOLANT ACCIDENT (LOCA) Af;D MAIN STEAf! LINE BREAK (MSLB) 2.1.1 LOCA To justify the setroint change, LOCA analyses for the Design Basis Accident (DBA), small break, and feedwater line break were performed. At the request of the staff, the licensee submitted a report (1-83-008) which included the LOCA analyses (Ref. 2). Our review of the report revealed that the emergency core cooling system (ECCS) evaluation used the RETRAN-02, M002 computer code. This code is not staff approved as an Appendix K-ECCS evaluation trodel.
In addition, the analysis was performed by Quadrex; it also is nut an approved user of the code. The staff stated that the portion of the Quadrax report relating to the above breaks would be ac-cepted, if, on a LOCA comparison basis, the peak cladding temperature (PCT) with the lower MSIV level trip point did not increase. The remaining alter-native was to use a staff approved ECCS evaluation model.
The licensee chose the latter alternative.
Ey letter dated March 3,1987 (Ref. 3), the licensee submitted a new LOCA safety evaluation to justify changing the MSIV water level isolation set point.
Presently, the most limiting LOCA, the one that results in the highest peak cladding temperature and determines the maximum average planar linear heat generation rate (ItAPLHGR) limit, is the recirculation suction line break DBA. ECCS calculations were performed using the staff approved codes of SAFE, REFLOOD and CHASTE. The effects of the proposed lower setpoint for large, interrrediate and small break LOCAs were considered.
The licensee stated that large and intermediate LOCA events would not be dffected by the setpoint change.
For these events, there would be a rapid depressurization and inventory loss within the reactor vessel resulting in a fast actuation of the MSIVs.
It was concluded that the lower MSIV set-point would not significantly increase the reactor core inventory loss, the total core uncovery time or subsequent fuel heatup, or the radiation release to the environrcent. Thus, the setpoint change would not affect the conse-quences of design basis accidents. The staff accepts these findings.
For a small break LOCA there is a potential of initittion of MSIV closure at the propused lower level setpoint which results ir r61 sing the peak cladding temperature (PCT). This event was analyzed. The results show that increase in PCT is less than 30 F.
The highest small break LOCA PCT would be substantielly less than the 2200 F limit.
The results of the LOCA analyses show that the DBA remains unchanged. Therefore, the MAPLHGR will not be changed. We find this acceptable.
2.1.2 MSLB MSLB was not analyzed.
The Quadrex report stated that the liSIV closure is initiated by high steam flow rather than by water level. We find this acceptable.
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- 2.2 ABf10RMAL OPERATI0f1AL TRANSIENTS Abnormal transient events were not evaluated by the litersee. Based on previous reviews with a lowered MSIV setpoint level, the staff does not consider these events to have any adverse effect with regard to the re-actor safety performance.
2.3 NITR0 GEN SUPPLY To allow the MSIVs to isolate at Level 1, the nitrogen supply to t'5IVs must also actuate Level 1.
This actuation is also changed from Level 2 to Level 1.
We find this to be acceptable.
2.4 TECHNICAL SPECIFICATION (TS) CHANGES We have reviewed the following TS changes:
TABLE 3.3.2-1 for isolation actuation instrumentation TABLE 3.3.2-2 for isolation actuation instrumentation set-points TABLE 3.3.2-3 for isoletion system instrumentation response time TABLE 4.3.2.1-1 for isolation actuation instrunentation sur-veillance requirements TABLE 3.6.3-1 for primary containment isolaticn valves Bases for figure B 3/4 3-1 for reactor vessel water levels We conclude that the proposed changes are acceptable.
3.0 ENVIRONMENTAL CONSIDERATI0fi These amendments invche changes in the installation and use of a f acility corponent located within the restricted area as defined in 10 CFR Part 20 and changes in surveillance requirements.
The staff has determined that these erendments involve no significant increase in the amounts, end no significant change in the types, of any effluents that may be released off-site, and that there is no significant increase in individual or cumulative occupational radiation exposure.
The Commission has previously issued a proposed finding that these amendments involve no significant hazards con-sideration and there has been no public comment on such finding. Accord-ingly, these emendments 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 prepareo in connection with the issuance of these anendn;ents.
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4.0 CONCLUSION
The Commission made a proposed determination that the emendments involve no significant hazards consideration which was published in the FEDERAL REGISTER (51 FR 41847) on November 19, 1986, and censulted with the statt of Illinois.
No public comments were received, and the state of Illinois did not have any comments.
The staff has concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the nealth and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Comaission's regulations and the issuance of these amendments will not be inimical to the common defense end security or to the health and safety of the public.
Prircipal Contributor:
D. Katze, RSB Dated: May 6,1987 REFERENCES 1.
Letter from C. Allen Commonwealth Edison to H. R. Denton, USNRC, dated October 23. 1986.
2.
Letter from C. Allen, Commonwealth Edison to H. R. Denton, USNRC, dated November 5,1986.
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3.
Letter fron C. Allen, Commonwealth Edison to H. R. Denton, USNRC, dated March 6,1987, 1
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