ML20204G963
| ML20204G963 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 03/20/1987 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20204G873 | List: |
| References | |
| NUDOCS 8703260511 | |
| Download: ML20204G963 (3) | |
Text
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8 UNITED STATES 8'-
N NUCLEAR REGULATORY COMMISSION y
-E WASHINGTON, D. C. 20555 k..v /
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION un-37 RELATED TO AMENDMENT NO. 6 TO FACILITY OPEPATING LICENSE FO.-DPR AND AMENDMENT NO. 6 TO FACILITY OPERATING LICENSE NO.
-4 C0FMONWEALTH EDIS0N COMPANY BYRON STATION, UNITS 1 AND P DOCKET N05. STN 50-454 AND STN-50-455 INTRODUCTION In a "05000454/LER-1985-061, :on 850624,reactor Trip Occurred Due to Steam Generator 1A LO-2 Level.Caused by Attendant Inadvertent Activation of Local Overspeed Trip Bar for [[system" contains a listed "[" character as part of the property label and has therefore been classified as invalid. Pump 1C.Trip Lever Handles Removed|letter dated November 26, 1985]], the licensee requested a change to Technical Specification (TS) 3/4 5.2 to allow certain valves in the safety injection (SI)/ residual heat removal (RHR) systems to be temporarily closed during check valve leakage tests required in T.S. 4.4.6.2.2 Specifically T.S.
4.4.6.2.2 requires check valves SI 8818 A, B, C & D, located in the cold leg in.iection lines, to be checked for back leakage into the RHR system during Mode 3 at certain intervals, or after a valve malfunction or other unusual occurrence. The requested change would allow closure of either RHR injection valve SI 8809 A or B in order to perform leakage tests for the above check valves, since performance of these tests with the corresponding RHR injection valve in the open position could result in false readings, could significantly increase the time required to perform this surveillance, and could make it more difficult to determine which of several check valves is leaking.
Closure of one RHR in,iection valve (SI 8809 A or B) isolates injection flow from the RHR pumps into two RCS cold legs. The FSAR accident analysis assumes injection flow into all 4 cold legs during a large break LOCA. The licensee submitted additional analyses which would indicate that sufficient flow from the accumulators, charging, SI & RHR pumps would be available with the proposed system line-up to maintain the downcomer level completely full for the large break LOCA, and therefore little or no penalty in calculated peak clad temperature (PCT) would occur.
In a conference call of January 7,1986, the staff requested that the licensee '
provide assurance that the accumulators would be available during performance of these tests, since the accumulators would normally be isolated in the pressure range in which the tests would be performed (800 to 1000 psi).
In a conference call of March 11, 1986, the licensee agreed to this but further informed the staff that the comon SI pump discharge line isolation valve 8703260511 870320 PDR ADOCK 05000454 P
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SI-8835 also has to be closed during performance of these tests. Thus, the SI pumps would not be available in the event of a LOCA during check valve leak test performance and the large break LOCA analysis in the "05000454/LER-1985-061, :on 850624,reactor Trip Occurred Due to Steam Generator 1A LO-2 Level.Caused by Attendant Inadvertent Activation of Local Overspeed Trip Bar for [[system" contains a listed "[" character as part of the property label and has therefore been classified as invalid. Pump 1C.Trip Lever Handles Removed|November 26, 1985 letter]] was therefore invalid.
In a letter dated June 16, 1986, the licensee submitted a modified amendment, including a new LOCA analysis which reflected the unavailability of the SI pumps due te valve SI-8835 closure during the leak tests.
It was assumed that the accumulators would be available with their isolation valves either open or closed with the valve motors energized.
In this condition, the accumulators would be automatically aligned to the RCS in the event of an SI signal. The licensee further indicated that the 3.4 psig containment high-1 SI actuation setpoint would be exceeded during a large break LOCA in mode 3, thus automatically actuating SI, " based on detailed Westinghouse calculations on a similar plant", and that the PCT would be bounded by the FSAR analysis of LOCA at full power.
In subsequent conference calls, the staff requested the folloving additional information from the licensee: identification and a comparison of Byron with the plant referenced in the Westinghouse calculations; what indications and alarms would alert the operator to manually initiate SI in the event of a small break LOCA for which the containment pressure would not rise high enough to automatically actuate SI; and whether the core would stay covered in the event of a small break LOCA since the RCS pressure would not necessarily decrease sufficently for accumulator and RHR in,iection and the SI pumps wculd be isolated. The licensee identified the referenced plant as Millstone 3, which has a slightly smaller containment than Byron.
In a letter dated November 24, 1986 the licensee indicated that the probability of a large break LOCA is lower at shutdown than at operating conditions because of lower temperatures and pressures. The licensee provided the results of estimates for breaks less than G inches during mode 3 two heurs after shutdown, For breaks up to 3 inches at least 20 minutes would be available to initiate flow from one of the charging pumps, which is expected to limit PCT to less than the design case, although some core uncovery might i
result. For breaks between 3 8 6 inches operator action to start one charging pump would be required in about 10 minutes. Additional operator action may be required, depending on break size, within one hour after LOCA initiation to start an additional charging or SI pump, or depressurize the RCS using the steam generators and to start an RHR pump.
The licensee provided a list of alcrms and indications that would alert the operator regarding occurrence of a small break LOCA. The licensee also indicated that the check valve surveillance test would probably be performed during startup after a refueling outage or after the plant has been in Mode 5 for longer than 3 days. Thus, the decay heat would be lower than assured in the above estimates. The licensee also informally proposed the following footnote to the accumulator TS:
"When either SI 8809 A or SI 8809 R is closed and pressurizer pressure is be, low 1000 psig, the accumulators shall be OPERABLE with their isolation valves either closed but enercized or open."
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- In the "Significant Hazards Consideration" the licensee indicated that implementation of the proposed TS changes would result in an overall increase in the margin of safety since it would provide increased assurance of the integrity of the RHR injection check valves, thus reducing the probability of an intersystem LOCA. The licensee has also provided reasonable assurance that a LOCA occurring during these check valve leakage tests can be mitigated. The staff concurs with the licensee's conclusions and finds the proposed T.S.
changes, namely closure of SI 8835 and either SI 8809 A or B during leak checks of SI 8818 A, B, C & D, acceptable providing the above footnote to the accumulator T.S. is also added and procedures are available to mitigate the LOCA which may occur while the valves are closed.
The licensee's submittal of November 24, 1986 was made as a result of NRC staff reouest to clarify the original submittal dated November 26, 1985, supplemented June 16, 1986.
ENVIRONMENTAL CONSIDERATION These amendments involve a change in the installation or use of the facilities components located within the restricted areas as defined in 10 CFR 20.
The staff has determined that these amendments involve no significant increase in the amounts, and no sionificant 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 these amendnents involve no significant hazards consideration and there has been no public comment on such finding. Accordingly, these amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR Sec 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 these amendments.
CONCLUSION We have 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, and (2) such activities will be conducted in compliance with the Commission's regulations and the issuance of these amendments will not be inimical to the common defense and security or to the health and safety of the public.
Dated:
March 20, 1987 PRINCIPAL CONTRIBUTORS:
B. Mann L. Olshan I
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