ML13317A617
| ML13317A617 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 05/18/1981 |
| From: | Baskin K Southern California Edison Co |
| To: | Lainas G Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8105210310 | |
| Download: ML13317A617 (5) | |
Text
Southern California Edison Company P. 0. BOX 800 2244 WALNUT GROVE AVENUE ROSEMEAD, CALIFORNIA 91770 K. P. BASKIN MANAGER OF NUCLEAR ENGINEERING, MELEP(35 1
SAFETY, AND LICENSING
.1
, 9[3524 Director of Nuclear Reactor Regulation Attention:
Gus C. Lainas, Assistant Director MAY 20 1981 Safety Assessment caaGUsA7om Division of Licensing U. S. Nuclear Regulatory Commission Washington, D.C. 20555 Gentlemen:
Subject:
Docket No. 50-206 Environmental Qualification of Safety-Related Electrical Equipment San Onofre Nuclear Generating Station Unit 1 Your letter of February 17, 1981 transmitted the report entitled, "Partial Review, Equipment Evaluation Report by the Office of Nuclear Reactor Regulation," for San Onofre Unit 1. The report contains the preliminary results of the NRC staff's review of the environmental qualification of safety related electrical equipment at San Onofre Unit 1 and identifies potential equipment deficiencies in Appendices B and C of the report.
You requested that we review these potential equipment deficiencies and provide our overall finding regarding continued safe operation of Sari Onofre Unit 1.
The purpose of this letter is to advise you that we have completed our review of the aforementioned report. In all cases for the equipment identified in Appendices B and C, with the exception of FCV 11150, E and F, the information provided by our October 31, 1980 submittal is still applicable and provides sufficient basis for continued operation.
With respect to FCV 111SD, E and F, it was identified in our April 16, 1981 letter that the non-metallic parts of the valve's operators and solenoids and the valve's regulators would be replaced. The need to replace these items constituted a reportable occurrence and the letter served as written notification to the NRC regional office. Additional details regarding these valves was provided in the two week follow-up report provided to the regional office by letter dated May 4, 1981.
In addition to reviewing the potential equipment deficiencies in Appendices B and C of the aforementioned report, our review has considered information developed as part of the continuing review of this matter as reflected in our October 31, 1980 submittal for equipment not addressed in the aforementioned report. Based on this review, the information provided in our October 31, 1980 submittal for equipment not addressed in the aforementioned report, with the exception of terminal blocks and cable, is still applicable and provides sufficient basis for continued operation. As indicated in our April 16, 1981 letter, we proceeded with an evaluation of the safe shutdown b
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May 18, 1981 terminal blocks and cable which we indicated may be adversely affected by the post-accident environment. That evaluation is now complete and the results are provided in the paragraphs below.
Our evaluation of information obtained since our October 31, 1980 submittal, concerning the terminal blocks located inside containment, including the tests documented in the Sandia report, NUREG/CR-1682, Electrical Insulator in a Reactor Accident Environment, concludes that operability of the terminal blocks in the post-accident environment cannot be assured and that replacement of safety-related circuits on the terminal blocks with qualified in-line splices is warranted. Our April 28, 1981 letter provided written notification to the NRC regional office of this matter which constitutes a reportable occurrence and indicated replacement of the safety-related circuits on the terminal blocks would be done prior to the startup of the unit from the current outage.
Our evaluation of the cable (i.e., teflon, polyethylene and PVC type) which may be adversely affected by the post-accident environment included a review of the information obtained since the October 31, 1980 submittal, the determination of the components served by the cable, the safety function required to be performed, the potential failure mode of the cable and postulated accident scenarios. Based on the evaluation, the operability of the components served by the cable in the post-accident environment can be assured and no remedial actions or corrective measures are warranted. The details of our evaluation are provided in Enclosure 1 of this letter.
In addition to reviewing the potential equipment deficiencies in Appendices B and C of the aforementioned report, we have reviewed the NRC staff actions contained therein to assure the adequacy of our environmental qualification program. Our comments and clarifications concerning the actions are provided in Enclosure 2 of this letter. The organization of our comments and clarifications correspond to the sections of the aforementioned report.
We trust that the information provided above is responsive to your February 17, 1981 letter.
Subscribed on this day of
, 1981.
By K. P. Baskin Manager of Nuclear Engineering Safety, and Licensing Subscrib d and sworn to before me on this
/ VTA-day of 1980.
SDONA MARY WILCOMBO OFFICIAL SEAL of6o Th~NOTARY PtJBLIC.CALIMONIA f
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PRINCIPAL OFFICE IN Cut LOS ANGL COUNT'Y Notary Public in d for the County of
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Myommisskm Expes June18,1981 Los Angeles, Sta of California Enclosures
ENCLOSURE 1 EVALUATION OF TEFLON, POLYETHYLENE AND PVC CABLE INSIDE CONTAINMENT Teflon Cable This cable insulation has a radiation limit of 4 x 106 rads above which cracking of the insulation occurs. However, it maintains its dielectric properties up to 109 rads. In the event of a large break LOCA, the components supplied by this cable (i.e., three pressurizer pressure and three pressurizer level transmitters) will perform their safety function prior to being exposed to the harsh environment. In the event of a small break LOCA or secondary line break inside containment, the post-accident radiation environment is not expected to exceed 4 x 106 rads. Therefore, the components will be capable of performing their safety function during a small break LOCA and a secondary line break.
In addition, a sump level.alarm is supplied by the teflon cable. This alarm is utilized by the operator to initiate recirculation; however, the operator also utilizes three other instruments, two of which are located in non-harsh environments. Therefore, the operator can utilize other components to initiate recirculation with the loss of the sump level alarm.
Therefore, as discussed in our October 31, 1980 submittal, the components served by this cable will perform their safety function prior to experiencing a harsh environment or other components are available to perform a similar function in the event the components fail.
Accordingly, no remedial actions or corrective measures are warranted.
Polyethylene Cable Polyethylene cable varies by type and will melt in a range of temperature of 202oF to 2830F. The component supplied by this cable is the containment radiation monitor.
As discussed in our October 31, 1980 submittal, this component does not perform a safety function to shut down the reactor. In addition, portable detectors or an appropriate Environmental Radiation Monitoring System channel outside containment can be used to determine the radiation level inside containment. Accordingly, no remedial action or corrective measures are warranted.
PVC Cable Based on industry test results provided in IPCEA S-61-402, NEMA WC5, PVC cable with a continuous allowable temperature rating of 750C or greater will function in the post-accident environment. Based on available purchase order documentation and evaluation of typical company requirements for cable installation, it is concluded that the PVC cable at San Onofre Unit 1 has a continuous allowable temperature rating of 750C or greater.
However, based on test results provided in the Sandia report (SAND-79-092 CK),
the cable may experience unacceptable embrittlement when exposed to radiation in excess of 4 x 106 rads in combination with the post-accident tempera tures. In the event of a large break LOCA, components supplied by this
-2 cable are required to ensure adequate core cooling during long-term recirculation (i.e., two control valves in the Hot Leg Recirculation System).
However, an alternate hot leg recirculation path is available to perform a similar function in the event the components fail.
In the event of a small break LOCA or secondary line break inside containment, the post-accident radiation environment is not expected to exceed 4 x 106 rads. Therefore, the components supplied by the cable (i.e., two control valves in the Hot Leg Recirculation System, reactor coolant temperature detectors, three steam flow transmitters and their power supplies and Residual Heat Removal flow control valve positioners, temperature detector and flow transmitter) will perform their safety function in the post-accident environment.
As discussed in our October 31, 1980 submittal, the components served by this cable will perform their required safety function in the harsh environment or other components are available to perform a similar function in the event the components fail.
Accordingly, no remedial actions or corrective measures are warranted.
Notwithstanding the above information, we initiated efforts to replace the cable in parallel with proceeding with our evalution of the cable to minimize the impact on the current outage. At the time we completed our evaluation, new cable had been pulled for one reactor coolant tempeature detector on each reactor coolant system cold-leg. Since this cable has been pulled, we have decided to complete the final electrical connection of this cable prior to startup of the unit from the current outage.
ENCLOSURE 2 COMMENTS AND CLARIFICATIONS CONCERNING NRC STAFF ACTIONS CONTAINED IN EQUIPMENT EVALUATION REPORT PARTIAL REVIEW 3.2 Service Conditions The staff assumed and requires that the Licensee verify that the containment spray system is automatic and not subject to a disabling single component failure and therefore, satisfies DOR Guideline requirements of Section 4.2.1.
Our October 31, 1980 letter indicated that the containment spray system is automatic; however, no discussion was provided concerning the single failure criteria. The containment spray system was evaluated as part of the Single Failure Analysis of the ECCS which was provided to the NRC by letter dated December 21, 1976. The evaluation identified various administrative controls which have been implemented to ensure the containment spray system cannot be disabled by a single component failure. Therefore, the large break LOCA environment conditions have been used to envelope the main steam line break environmental conditions in accordance with the requirements of Section 4.2.1.
3.3 Temperature, Pressure, and Humidity Conditions Inside Containment Contrary to the NRC staff's recommendations, we have concluded that an increase in the minimum temperature profile to account for higher than average temperature in the upper regions of the containment that can exist due to stratification is not warranted. The basis for this conclusion is that the inherent conservatisms in our containment pressure and temperature analysis and the conservatisms included by the DOR Guideline requirements of Sections 4.0 and 6.0 assure adequate margins to the maximum containment conditions.
3.5. Submergence Contrary to the NRC staff's statements, we have identified equipment below the maximum submergence elevation of 3 feet 11 inches. The NRC staff is directed to our June 18 and October 31, 1980 letters for a listing of the equipment.
3.6 Chemical Spray The NRC's referenced value for our chemical concentrations for the containment spray system is not appropriate. Our October 31, 1980 letter indicated that the chemical concentrations are contained in Amendment No. 52 submitted in Docket No. 50-206 by letter dated December 3, 1975.
3.7 Aging As requested by the NRC staff, our June 18 and October 31, 1980 letters identify measures initiated to ensure that equipment is not susceptible to significant degradation.
3.8 Radiation (Inside and Outside Containment)
As requested by the NRC staff, we have justified the gamma and beta radiation service conditions inside containment used in our environmental qualification program. Table 3 of the enclosure to our October 31, 1980 letter indicates that the DOR Guideline requirements for gamma and beta radiation have been utilized.