ML20011E864
| ML20011E864 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 02/06/1990 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20011E861 | List: |
| References | |
| NUDOCS 9002220623 | |
| Download: ML20011E864 (4) | |
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't UNITED STATES f
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c; NUCLEAR REGULATORY COMMISSION i
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- WASHINGTON, D. C. 20606
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT N0. 48 TO FACILITY OPERATING LICENSE NO. DPR-80 i
AND AMENDMENT N0.40 TO FACILITY OPERATING LICENSE N0. DPR-82 PACIFIC GAS AND ELECTRIC COMPANY DIABLO CANYON NUCLEAR POWER PLANT. UNIT NOS. 1 AND 2 DOCKET N0. 50-275 AND 50-323
1.0 INTRODUCTION
By letters dated March 22, May 15, and December 26,1989(Reference LAR 89-03), Pacific Gas and Electric Com>any (PG8E or the licensee) requested amendments to the combined Tec1nical Specifications (TS) appended to Facility Operating License Nos. DPR-80 and DPR-82 for the Diablo Canyon Power Plant (DCPP), Unit Nos.1 and 2, respectively. The amendmentschangetheTSto(1)changeTS4.3.1.1. Table 4.3-1 Item 23, "SeismicTrip,"toincrearethesurveillancetestinterval(ST)(forthe seismic trip system actuating device operational test from "SA" at least once per 184 days or approximately 6 months) to "R" (at least onceper18monthstoeliminatetheneedtoperformseismictripsystem surveillance testing at power, and (2) Change TS 3.3.1, Table 3.3-1, Item 23, " Seismic Trip," to allow any one 01 the three seismic trip system channels to be bypassed for up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for surveillance testing or maintenance while operating at power.
The staff evaluation of these changes is given below and is based on the licensee's letters of March 22, May 15, and December 26, 1989. The information contained in the licensee's letter of December 26 1989 did l
not change the action noticed in, or alter the staff's propose,d l
determination of no significant hazards consideration published in the Federal Register on May 31, 1989 at 54 FR 23318.
2.0 EVALUATION The NRC staff has evaluated the proposed changes and finds them I
acceptable, based on the analyses and evaluations given by the licensee.
A discussion of each of the specific technical specification changes made by these amendments, is presented below.
(1)
Item 23 " Seismic Trip," of Table 4.3-1," Reactor Trip System Instrumentation Surveillance Requirements", of TS 3/4.3.1, " Reactor Trip System Instrumentation", is modified to increase the surveillancetestinterval(STI)fortheseismictripsystem actuatingdeviceoperationaltestfrom"SA"(atleastonceper184 days) to "R" (at least once per 18 months) to eliminate the need to perform seismic trip system surveillance testing at power.
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The licensee notes that operation of the seismic trip system is not required or assumed to mitipate the consequences of any accident i
analyzed in Chapter 15 of t se FSAR. The plant was Itcensed on the basis of its capability to safely shut down following the "Hosgri" and the " Double-Design Earthquake' without relying on the seismic l
trip system.
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The licensee has had an unavailability evaluation perfomed by.
Westinghouse to determine the effect on core damage frequency of changing the STI from 6 to 18 months. The evaluation indicated v
that the core damage frequency would be increased by a negligible amount-(approximately 0.012 percent).
The licensee also states that the operating history of the seismic trip system demonstrates that component failures would not have prevented a reactor trip had a seismic event of the prescribed l
magnitude occurred. Because the system design does not pemit reliable testing at power, two inadvertent reactor trips have occurred during testing. Such challenges to the reactor protection system cause an increase in core damage frequency.
Increasing the STI to allow testing to be performed during shutdown periods will 1
eliminate the risk of inadvertent reactor trips and establishing an out of service time will allow for maintenance or component replacement at power, Therefore, the licensee concludes that increasing the STI of the trip actuating device operational test from 6 to 18 months will result in a net increase in plant safety, because of the reduction in the number of unnecessary plant trips is more significant than the longer time between system surveillances.
Each' seismic trip sensor is provided with internal batteries to l
assure operation in the event the normal power source is lost.
The i
E normal power source is vital instrument power installed in accordance with Class 1E requirements with each of the three t
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sensors powered from a separate protection set.
Thus, the battery backup power supply for a sensor would be relied upon only if the protection set power supply for that sensor were to become E
inoperable. The battery vendor recomends that the battery be l
checked semiannually. To ensure that the seismic monitor battery check is performed in accordance with the manufacturer's recomendations, the licensee has committed to create a new i
surveillance test procedure which will require that the the L
batteries be checked semiannually. The licensee has also comitted L
to track battery surveillance using PHE's computerized recurring task schedular system to further assure that it is performed semiannually, as planned. The NRC staff has reviewed the licensee's comitment to semiannually check the seismic monitor batteries and finds it acceptable.
The NRC staff has reviewed the licensee's evaluations and analyses described above and agrees that the proposed increase in the STI will not adversely affect the safety of the plant.
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i (2) Item 23, " Seismic Trip " of Table 3.3-1,* Reactor Trip System Instrumentation", of TS 3/4.3,1, " Reactor Trip system Instrumentation", is modified to allow any one of the three seittic 4
trip system channels to be bypassed for up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for surveillance testing or maintenance while operating at power.
When one channel is bypassed, a two of two coincidence in one of the three axes would generate a reactor trip if the trip setpoint i
of 0.35g is exceeded. When no channels are bypassed, two of three coincidence is required to generate a trip.
The licensee states that this~ change does not result in a I
significant reduction in plant safety, based on an evaluation of the effect on core damage frequency of taking all three channels out of service for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> while operating at power. This study showed that the core damage frequency was increased by a negligible amount (0.002 3ercent). Since placing one channel in by> ass for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> will inve a much smaller effect than removing t1ree i
channels from service for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, the licensee concludes that no adverse safety consequences result.
The NRC staff has reviewed the licensee's evaluation and agrees that it will not have an adverse affect on safety, in summary the NRC staff has reviewed the request by the Pacific Gas and Electric Company to modify the combined Technical Specifications for Diablo Canyon Units 1 and 2 to change the STI for the seismic trip system to 18 months, and to allow one channel of the system to be placed in bypass for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> during power operation, and finds it acceptable, as discussed above.
3.0 ENVIRONMENTAL CONSIDERATION
These amendments involve changes to a requirement with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20 and a change in surveillance requirements. At Diablo Canyon, the restricted area coincides with the site boundary. We have 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 o fsite, and that there is d
no significant increase in individual or cumulative occupational 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 finding. Accordingly
. these amendments meet the elig)ibility criteria for categorical exclusion setforthin10CFR51.22(c)(9. Pursuantto10CFR51.22(b),no environmental impact statement or environmental assessment need be prepared in connection with the issuance of these amendments.
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4.0 CONCLUSION
j We have concluded, based on the considerations discussed above that:
(1) there is reasonable assurance that-the health and safety of the i
p(2) such activities will be conducted in compliance with theublic will not :'
Commission's regulations and (3) the issuance of these amendments will not be inimical to the common defense and security or the health and l
safety of the public.
i Principal Contributors: Harry Rood Robert Rothman i
Dated: February 6, 1990
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