ML17329A121
| ML17329A121 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 08/13/1991 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17329A120 | List: |
| References | |
| GL-83-28, NUDOCS 9108150251 | |
| Download: ML17329A121 (8) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO GENERIC LETTER 83-28 ITEM 1.2 - POST-TRIP REVIEW DATA AND INFORMATION CAPABILITY INDIANA MICHIGAN POWER COMPANY DONALD C.
COOK NUCLEAR PLANT UNIT NOS.
1 AND 2 I.
INTRODUCTION On February 25, 1983, both of the scram circuit breakers at Unit 1 of the Salem Nuclear Power Plant failed 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 fai lure 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 generated 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) directed the staff to investigate and report on the generic 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 Event at the Salem Nuclear Power Plant."
As a result of this investigation, the Commission (NRC) requested (by Generic Letter 83-28 dated July 8, 1983) 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 Review; (2)
Equipment Classification and Vendor Interface; (3) Post-Maintenance Testing; and (4) Reactor Trip System Reliability Improvements.
The licensee submitted a response to Generic Letter 83-28 on November 4, 1983, with additional information provided by letters dated March 30, 1984 and December 17, 1986.
This safety evaluation (SE) addresses only the licensee's response to Action Item 1.2, Post-Trip Review - Data and Information Capability.
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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.
III. REVIEW CRITERIA The following review guidelines were developed after initial evaluation of the various utility responses to Item 1.2 of Generic Letter 83-28 and incorporate the best features of these submittals.
As such, these review guidelines in effect represent a "good practices" approach to post-trip review.
We have reviewed the licensee's response to Item 1.2 against these guidelines:
A.
The equipment that provides the digital sequence-of-events (SOE) records and the analog time history records of an unscheduled shutdown should provide a reliable source of the necessary information to be used in the post-trip review.
Each plant
- 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, and 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 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-events time discrimination is approximately 100 milliseconds.
If current sequence-of-events 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 adequately reconstruct the transient and accident scenarios presented 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 the accident analysis of the plant FSAR.
The recommended guideline for the sample interval is 10 seconds.
If the time history 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 USAR.
To support the post-trip analysis of 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 five minutes prior to the trip until at least ten minutes after the trip.
All equipment used to record sequence-of-events and time history information should be powered from a reliable and non-interruptible power source.
The power source used need not be safety related.
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 other 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 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 envelop are presented in Table 1.
They were selected on the basis of staff engineering judgment following a complete evaluation of utility submittals.
If 'the licensee's sequence-of-event and time history recorders do not monitor all of the parameters suggested in this table, it should be shown that the existing set of monitored parameters are sufficient to establish that the plant remained within the design envelop for the accident conditions analyzed in the plant FSAR.
C.
The information gathered by the sequence-of-events and time history recorders should be stored in a manner that will allow for data retrieval and analysis.
The data may be retained in either hardcopy (e.g.,
computer printout or strip chart record) or in an accessible memory (e.g.,
magnetic disc or tape).
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.
D.
Retention of data from all unscheduled shutdowns provides a valuable 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.
IV.
EYALUATION AND DISCUSSION By letters dated November 4, 1983, March 30, 1984, and December 17, 1986, the Indiana Michigan Power Company provided information regarding its post-trip review program data and information capabilities for the Donald C.
Cook Nuclear Plant, Unit Nos. I and 2.
We have evaluated the licensee's submittal against the review guidelines described in Section III.
Licensee deviations from the guidelines of Section III were reviewed with the licensee by telephone on October 24, 1990.
A brief description of the licensee's responses and the staff's evaluation of the responses against each of the review guidelines are provided below:
A.
The licensee has described the performance characteristics of the equipment used to record 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.
Information supplied in the licensee's original submittal of November 4, 1983, indicated that the SOE recorder met the guidelines noted above but that the analog time history data. recorder did not.
Although not taken credit for in their post-trip review process, the licensee also has Technical Support Center (TSC) computer facilities which enhance their post-trip review capabilities.
The TSC computer records time history data at one minute intervals for the period from 30 minutes prior to the trip until 30 minutes after the trip.
In addition, for the first 5 minutes following a reactor trip, the recording interval is 10 seconds.
The TSC computer is powered from a reliable power source.
Although less stringent than the guidelines, the licensee meets the intent of the criteria.
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 parameters selected by the licensee include most of those identified in Table 1.
The licensee does not record on the SOE or time history recorder all of the parameters recommended in Section III.B. however, alternate parameters may be used to implicitly determine the recommended parameters.
- Further, as noted below, the TSC computer records parameters over and above those included in the licensee's formal post-trip review.
C.
D.
We find that the containment isolation signal is not recorded directly but is indirectly available by consulting containment pressure.
Furthermore, containment isolation information is available from the TSC computer.
Control rod position is not recorded for all rods; however, the control rod bank position is recorded by the plant computer as part of the analog data base.
Containment sump level is recorded on the TSC computer.
The licensee has flow indicators on all safety injection pumps and limit switch indicators on key emergency core cooling system (ECCS) valves.
These indicators are monitored by the TSC computer.
Hain steam isolation valves (HSIV) position is not recorded directly but can be obtained indirectly from the reactor trip safety injection steamline pressure/flow steamline isolation signal.
Auxiliary feedwater flow is available from all auxiliary feedwater pumps and is monitored on the TSC computer.
Finally, power-operated relief valve (PORV) position is monitored on the TSC computer.
In summary, most of the desirable plant parameters needed for post-trip review are recorded 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 III.B and is acceptable.
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 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 guidelines of Section III.C.
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 guidelines of Section III.D.
V.
CONCLSION Based on the foregoing discussion, the staff concludes that the licensee's post-trip review data and information capabilities for the Donald C.
Cook 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.
Principal Dated:
Contributor:
W.
H. Swenson
TABLE 1 PWR PARAMETER LIST SOE Recorder Time History Recorder Parameter / Si nal x
(1) x x
(1) x x
(1) x x
(2)
(1) x (1) x (1) x (1) x (1) x (3)
(1) x (1) x (1) x (3)
Reactor Trip Safety Injection Containment Isolation'urbine Trip Control Rod Position Neutron Flux, Power Containment Pressure Containment Radiation
. Containment Sump Level Primary, System Pressure Primary System Temperature Pressurizer Level Reactor Coolant Pump Status Primary System Flow Safety Inj. Flow, Pump/Valve Status MSIV Position Steam. Generator Pressure Steam Generator Level Feedwater Flow Steam Flow Auxiliary Feedwater System;
Diesel Generator Status (Start/Stop, On/Off)
PORV Position (1)
Trip parameters (2)
Parameter may be monitored by either an SOE or time history recorder.
(3)
Acceptable recorder options are:
(a) system flow recorded on an SOE recorder, (b) system flow recorded on a time history recorder, or (c) equipment status recorded on an SOE recorder.
DISTRIBUTION NRC PDR Local PDR B. Boger J. Zwolinski E.
Greenman T. Harsh W. Long T. Colburn W. Swenson i3/p/4 OGC (For inform. Only) 15-B-18 E. Jordan HNBB-3701 ACRS (10)
P-315 PD31 R/F
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