ML20237G554
| ML20237G554 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 08/27/1987 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20237G534 | List: |
| References | |
| TASK-2.D.1, TASK-TM TAC-51590, TAC-64905, TAC-64906, NUDOCS 8709020303 | |
| Download: ML20237G554 (3) | |
Text
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NUCLEAR REGULATORY COMMISSION o
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.E WAEHWGTON, D. C. 20555
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 17 TO FACILITY OPERATING LICENSE NO. DPR-80 AND AMENDMENT N0.16 TO FACILITY OPERATING LICENSE NO. DPR-82 PACIFIC GAS AND ELECTRIC COMPANY-DIABLO CANYON NUCLEAR POWER PLANT, UNIT NOS. 1 AND 2 DOCKET NOS. 50-275 AND 50-323
1.0 INTRODUCTION
l By letter dated March 13, 1987, as supplemented July 9, 1987, Pacific Gas and Electric Company (PG&E or the licensee) requested amendments to the Technical Specifications appended to Facility Operating License Nos. DPR-80 and DPR-82 for the Diablo Canyon Nuclear Power Plant, Unit Nos. I and 2.
The proposed amendments would change the reactor trip setting for steam generator low water level from 25% to 15% of the narrow instrument span.
2.0 DISCUSSION AND EVALUATION The Diablo Canyon Nuclear Power Plant has two reactor trips derived from steam generator water level:
low-low level, set at 15% of narrow i
instrument span; and low level set at 25% of span in coincidence with steam flow / feed flow mismatch (feed flow less than steam flow by a set amount).
Diablo Canyon has reported several reactor trips which occurred during j
startup when control of steam generator level is performed manually.
l The trips were caused by the 25% level setting in coincidence with false misn'atch signals due to the sensitivity of the instruments for detecting steam and feedflow. Since this circuit was causing spurious reactor i
trips, the licensee has evaluated the safety implications of changing the level setting to 15% and thus change this reactor trip to occur at the same level setting as the low-low level trip. This would provide the operators controlling level during startup a greater allowable range l
before the trip setting and should improve operations, leading to less frequent challenges to safety systems and greater reliability.
The licensee reports that, as shown in WCAP-10948*, low steam generator level in coincidence with low feedwater flow (steam /feedwater flow mismatch) trips were the third most prevalent cause of automatic reactor trips in U.S.
Westinghouse-designed plants in the 1980-1985 time frame, accounting for 154 out of 1279 total trips that were analyzed, or about 12 percent of the automatic trips in the 38 Westinghouse plants surveyed.
8709020303 870827 PDR ADOCK 05000275 P
PDR-
- WCAP-10948, "U.S. Westinghouse Inadvertent Plant Trip Experience: A Historical Review of Information from January 1980 through September 1985. March 1986.
(Pro 3rietary)
From the safety standpoint, the steam /feedwater flow mismatch signal in coincidence with a steam generator low level reactor trip set at its j
current setting of 25% is not used or relied upon in the transient and accident analysis for Diablo Canyon, but the trip is included in
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technical specification Table 2.2-1 to enhance the reliability of the i
reactor trip system (discussed further below). The trip provides a measure of diversity and redundancy to the low-low level trip. The licensee does not propose to delete it. The low-low level trip actuates i
the auxiliary feedwater pumps. The steam /feedwater flow mismatch / low level signal only trips the reactor, and has no other control function.
j Since Diablo Canyon has only three level channels for each steam j
generator level, the low level in coincidence with low feedwater flow reactor trip is needed in order to satisfy IEEE-279, " Criteria for Protection Systems for Nuclear Power Generating Stations." Paragraph 4.7.3 of this standard discusses specic1 requirements for nostulated failures where, as here, control and protection functions are combined. Since one protection channel per steam generator provides a-control input to j
the steam generator level control system as well as providing protection, paragraph 4.7.3 requires that two random failures be postulated. Since the low-low reactor trip utilizes 2-out-of-three logic, 2 postulated failures (such as failing high) could prevent protective action.
Additional protection is required to cause reactor trip which is provided
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by the low level circuit. Since the low level circuit uses one-out-of-two logic, and neither channel is used for control functions, this channel would trip on low steam generator level (in coincidence with low
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feedwater flow). Thus, the requirements of IEEE-279 are satisfied. The licensee proposes to change the setpoint to 15% of the narrow instrument range which is the same value as the low-low steam generator level reactor trip. Therefore, IEEE-279 would continue to be satisfied under the 4
proposed change.
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1 Newer Westinghouse-designed plants being licensed today do not have this low level / mismatch reactor trip. Rather, the requirements of IEEE-279 are satisfied instead by having four channels of protection for each steam generator and a 2-out-of-four trip logic.
In this way, a reactor trip would still occur even if two channels should fail.
The transients or accidents in the FU.R that are potentially affected by this setpoint change are:
- loss of feedwater
- loss of off-site power
- rupture of a main feedwater pipe
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This safety analyses contained in the Diablo Canyon FSAR Update do not low-low level trip (15%) p (25% setting), but instead rely on the rely on the low level tri to provide a reactor trip. Therefore, the safety analyses for Diablo Canyon would be unaffected by the proposed change.
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1 i
. 1 Historically, the original safety analysis for the main feedwater pipe rupture used the low level trip as part of a set of initial conditions in assumed to be dry (empty) generator (affected by the pipe rupture) was the analysis. One steam at the same time the other three steam generators were assumed to generate a reactor trip at the low level (25%)
setting. These assumptions were made in order to minimize steam generator fluid inventory and thereby maximize the resultant heatup of the reactor coolant system.
This analysis method for the rupture of a main feedwater aipe has been superseded by the analysis contained in the FSAR Update w1ich uses different assumptions, initial conditions, and methods developed in response to NUREG-0737 Item II.D.1, " Safety and Relief Valve Testing."
The revised analysis of this accident was submitted by PG&E on November 25, 1985. The analysis methods were accepted by the NRC staff as part of the resolution of Item II.D.1 for Diablo Canyon on January 27, 1986.
The specific results contained in licensee's November 25, 1985 have been l
reviewed and found acceptable, both with respect to this amendment request and NUREG-0737 Item II.D.1.
In summary, we conclude that the licensee's request to revise the low level trip setpoint to 15% of narrow instrument span is acceptable on i
the basis that the setpoint of 15% is the value relied upon in the safety analyses. Retaining the low level / low flow mismatch trip at this revised setting will continue the Diablo Canyon design conformance with j
l
3.0 ENVIRONMENTAL CONSIDERATION
These amendments involve changes in the installation or use of a facility l
component located within the restricted area as defined in 10 CFR Part 20.
l We have determined that the amendments involve no significant increase in l
the amounts, and no significant 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 Comission has previously issued a proposed finding that the amendments involve no significant hazards consideration and there has been no public coment on such finding. Accordingly, these amendments meets the eligibility l
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 environ-mental assessment need be prepared in connection with the issuarice of these amendments.
4.0 CONCLUSION
l We have concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the healtn 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 Comission's regulations and (3) the issuance of these amendments will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributors:
R. C. Jones, J. Mauck and C. Tramell Dated: August 27, 1987 j