ML20062F136

SE Accepting Util Responses to Generic Ltr 83-28,Item 1.2, Post-Trip Review - Data & Info Capability

ML20062F136
Person / Time
Site:	Point Beach
Issue date:	09/25/1990
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML20062F134	List: ML20062F130 ML20062F136
References
GL-83-28, NUDOCS 9011270108
Download: ML20062F136 (5)

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SAFETY EVALVATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO GENERIC LETTER 83-28. ITEM 1.2 - POST-TRIP REVIEW DATA AND INFORMATION CAPABILITY WISCONSIN _ ELECTRIC POWER COMPANY POINT BEACH NUCLEAR PLANT. UNIT NOS.1 AND 2 DOCKETNOS.50-266AND_50-33

!. INTRODUCTION On February 25, 1983 both of the scram circuit breakers at Unit 1 of the Salem Nuclear Power Plant f ailed to open upon an automatic reactor trip signal from the reactor protection system. This incident occurred during the plant start up and the reactor was tripped manually by the operator about 30 seconds after the initiation of the automatic trip signal. The failure of the circuit breakers has been determined to be related to the sticking of the under voltage trip attachment. Prior to this incident, on February 22, 1983 at Unit 1 of the Salem Nuclear Power Plant, an automatic trip signal was ge,nerated based on steam generator low-low level during plant start up. In this case, the reactor was tripped manually by the operator almost coincidentally with the automatic trip.

Following these incidents, on February 28, 1983, the NRC Executive Director for Operations-(EDO)directedthestafftoinvestigateandreportonthegeneric implications of these occurrences at Unit 1 of the Salem Nuclear Power Plant.

! The results of the staff's inquiry into the generic implications of the Salem unit incidents are reported in NUREG-1000 " Generic Implications of the ATWS EventattheSalemNuclearPowerPlant."dsaresultofthisinvestigation the Counission (NRC) requested (by Generic Letter 83 28 dated JuTy 8,19e3) all ,

licensees of operating reactors, applicants for an operating license, and holders of construction permits to respond to certain generic concerns. These concerns are categorized into four areas: (1) Post-Trip Reviews (2) Equipment Classification and Vendor Interface, (3) Post-Maintenance Testings and (4)

Reactor Trip System Reliability Improvements. The licensee submitted a response to Generic Letter 83-28 on November 7,1983, with additional information provided on June 1, 1984, December 28, 1984, and February 28, 1985.

This safety evaluation (SE) addresses only the licensee's response to Action Item 1.2. Post-Trip Review Data and Information Capability.

II. PROPOSED CHANGES The licensee's response to Generic Letter 83-28 was reviewed to ensure that the licensee has the capability to record, recall and display data and information which will permit diagnosing of the causes of unscheduled reactor shutdowns and for ascertaining the proper functioning of safety-related equipment.

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III. REVIEW CRITERIA The following rev ew guidelines were developed after initial evaluation of the i

l various utility responses to item 1.2 of Generic Letter 83-28 and incorporate

! the best featurt.s of these submittals. As such, these review guidelines in l effect represer.t a " good practices" approach to post-trip review. We have i

reviewed the 'icensee's response to item 1.2 against these guidelines:

A. The equip.. W inat provi e s the digital sequence of events ($0E) records and the analog time history records of an unscheduled shutdown should provide a reliable source t.f the necessary information to be used in the post-trip review. Eacie 01Jnt variable, which is necessary to determine the cause and progression of the events following a plant trip, should be monitored by at least one recorder (such as a sequence-of-events recorder or a plant process computer) for digital parameters, and strip charts, a process computer or analog recorder for analog (time history) variables.

Performance characteristics guidelines for sequence of events and time history recorders are as follows:

Each sequence of events recorder should be capable of detecting and recording the sequence of events with a sufficient time discrimination 4 capability to ensure that the time responses associated with each monitored safety-related system can be ascertained, and that a determination can be made as to whether the time response is within acceptable limits based on FSAR Accident Analyses. The recommended guidelines for the sequence of event time discrimination is approxi-mately 100 milliseconds. If current sequence of event recorders do not have this time discrimination capability, the licensee should show that the current time discrimination capability is sufficient for an adequate reconstruction of the course of the reactor trip and '

post-trip events. As a minimum, this should include the ability to l adequately reconstruct the transient and accident scenarios presented l in the plant FSAR.

Each analog time history data recorder should have a sample interval small enough so that the incident can be accurately reconstructed following a reactor trip. As a minimum, the licensee should be able to reconstruct the course of the transient and accident sequences  ;

evaluated in t!.e accident analysis of the plant FSAR. The recommended guideline for the sample interval is 10 seconds. If the time history I

equipment does not meet this guideline, the licensee should show that the time history capability is sufficient to accurately reconstruct the transient and the accident sequences presented in the FSAR. To support the post-trip analysis o' the cause of the trip and the proper functioning of involved safety-related equipment, each analog time history data recorder should be capable of updating and retaining information from approximately 5 minutes prior to the trip until at least 10 minutes after the trip.

All equipment used to record sequence of events and time history informa-

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tion should be powered from a reliable and non-interruptible power source.

The power source used need not be safety related.

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l 3 B. The sequence of events and time history recording equipment should monitor sufficient digital and analog parameters, respectively, to assure that the course of the reactor trip and post-trip events can be reconstructed. The parameters monitored should provide sufficient information to determine the root cause of the unscheduled shutdown, the progression of the reactor trip, and the response of the plant parameters and protection and safety systems to the unscheduled shutdowns. Specifically, all input parameters associated with reactor trips, safety injections and othcr safety-related ,

systems as well as output parameters sufficient to record the proper '

functioning of these systems should be recorded for use in the post-trip l review. The parameters deemed necessary, as a minimum, to perform a post-trip review that would determine if the plant remained within its '

safety limit design envelope are presented in Table 1. They were selected on the basis of staff engineering judgement following a complete evaluation of utility submittals. 11 the licensee's sequence of event and time I history recorders do not monitor all of the parameters suggested in these i tables, it should be shown that the existing set of monitored parameters I

is sufficient to establish that the plant remained within the design l envelope for the accident conditions analyzed in the plant FSAR, C. The information gathered by the sequence of events and time history i recorders should be stored in a manner that will allow for data retrieval and analysis. The data may be retained in eitherhardcopy,(e.g.,

computer printout, strip chart record), or in an accessible memory (e.g., i magneticdiscortape). This information should be presented in a readable and meaningful format, taking into consideration good human factors practices such as those outlined in NUREG-0700, i D. Retention of data from all unscheduled shutdowns provides a valuable I reference source for the determination of the acceptability of the plant vital parameter and equipment response to subsequent unscheduled shutdowns.

Information gathered during the post-trip review is to be retained for the life of the plant for post-trip review comparison of subsequent events.

I IV. EVALUATION AND DISCUSSION By letters dated November 7,1983, June 1,1984, December 28, 1984, and February 28, 1985, the Wisconsin Electric Power Company (WEPCO) provided information regarding its post-trip review program data and information capabilities for the Point Beach Nuclear Plant. We have evaluated the licensee's submittals against the review guidelines described in Section III. l Licensee deviations from the guidelines of Section III were reviewed with the l licensee by the project manager during a site visit on August 28, 1989. A  !

brief description of the licensee's response and the staff's evaluation of the 1 responses against each of the review guidelines are provided below:

A. The licenser; has described the performance characteristics of the equipment used to rerord the sequence of events and time history data needed for '

post-trip review. Based on our review, we find that the sequence of events and time history recorder characteristics conform to the guidelines described above and are acceptable.

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g I Information supplied in the licensee's original submittals indicated that '

neither the analog time history recorders nor the SOE recorders met the c guidelines noted above. Subsequently, the licensee installed new plant l process computers. The new computers have a 5-second sample interval L for time history data and retain this detailed information well beyond the L guideline covering the period from 5 minutes prior to the trip until 10 minutes after the trip. The new plant process computers have e time discrimination capability on the order of I-2 milliseconds for SOE data.

Further, these computers are powered from vital buses with fast transfer between power supplies.

B. The licensee has established and identified the parameters to be monitored and recorded for post-trip review. Based on our review, we find that the sarameters selected by the licensee include most of those identified in Table 1. The licensre does not record all of the parameters reconnendad in Section IIIB; however, alternate parameters may be used to implicitly determine the recommended parameter. Further, as noted below, t1e new plant computers record parameters over and above those included in the licensee's original responses.

! ' Safety injection signals are recorded on the plant computers. We find that the containment isolation signal is not recorded directly but is indirectly available by consulting the safety injection signal. Control rod position is not recorded for all .ods; however, the control rod bank position is recorded by the plant computers as part of the analog data base. Containment radiation is rec,rded as part of the analog data L 9 e.

The containment sump is fully instrucented and recorded by the plant computers, pressurizer level is monitored as a time history variable and the pressurizer high level tri Reactor coolant pumpstati.s(changeofstate)pisrecordedontheSOE.

is monitored on the SOE recorders. Primary system flow is measured for each loop. Safety injection flow is recorded E , as-a time history variable. Power operated relief valve (PORV) position (open/ closed) now is recorded or, the SOE recorders. Main steamline isolation valve (MSIV) pos;' ion is not re 3rded but is indirectly available since signals generating MS!v closure are recorded. Both wide and narrow range steam generator levels are recorded as part of the analog time history data base with the low-low steam generator level reactor trip signal recorded on the SOE. Auxiliary feedwater flow is recorded by the plant computer as a time history variable.

AC and DC bus voltages are not monitored directly; however, the SOE recorders monitor the 4.16 kV bus status, providing indication that the AC electrical buses are energized. The emergency diesel generators (EDGs) start on a safety injection signal which is recorded by the plant computers.

However, the licensee cannot verify that the EDGs have loaded since there is no indication of undervoltage on the off-site power grid or the position l of the EDG output breakers. Although not required, recording of AC/DC bus voltage and EDG output breaker status would be desirable.

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In summary, most of the desirable plant parameters needed for post-trip review are now rea rded by the licensee. Alternative data sources for those parameters not recorded are available for the post-trip review.

Consequently, we find that the licensee's selection of parameters meets the intent of the guidelines described in Section IIIB and is, therefore, acceptable.

C. The licensee has described the means for storage and retrieval of the information gathered by the sequence of events, time history and analog data base recorders, and for the presentation of this information for L post-trip review and analysis. Based on our review, we find that this information is being presented in a readable and meaningful format, and that storage, retrieval and presentation conform to the guideline of Section IIIC.

D. The licensee has described the retention capability of the data gathered by the plant computer and the time history records. Based on our review, we find that the program for the retention of data conforms to the guide-lines of Section 1110.

I V. CONCLUSION 1

l- Based on the foregoing discussion, the staff concludes that the licensee's

! post-trip review data and information capabilities for the Point Beach Nuclear

!- Plant, Unit Nos. I and 2, are acceptable for Item 1.2 of Generic Letter 83-28.

However, recording of additional parameters, as discussed in Section IV B, would serve to improve and expedite post-trip reviews.

Dated: September 25, 1990 Principal Contributor: W. H. Swenson 1

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