ML17262A225
| ML17262A225 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 11/05/1990 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17262A224 | List: |
| References | |
| GL-83-28, NUDOCS 9011070356 | |
| Download: ML17262A225 (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 ROCHESTER GAS AND ELECTRIC CORPORATION R. E.
GINNA NUCLEAR POWER PLANT DOCKET NO. 50-244 1.
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 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 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 occur rences 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.
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.
90ii070356 90ii05 PDR ADOCK 05000244 PDC 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.
Ne 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 shut down 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, 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 event time discrimination is approximately 100 milliseconds.
If current sequence of event recorders do not have this time discrimination capability, the licensee should show that the current time discriminiation capability is sufficient for an adequate reconstruction of the course of the reactor tip 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 FSAR.
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.
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B.
C.
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 envelope are presented in Table 1.
They were selected on the basis of staff engineering judgement 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 these tables, it should be shown that the existing set of monitored parameters are sufficient to establish that the plant remained within the design envelope for the accident conditions analyzed in the plant FSAR.
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, 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 consider ation 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 resource 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.
EVALUATION AND DISCUSSION By letter dated November 4, 1983, Rochester Gas and Electric Corporation provided information regarding its post-trip review program data and information capabilities for the R. E.'inna Nuclear Power Plant.
We have evaluated the licensee's submittal against the review guidelines described in Section III.
Licensee deviations from the guidelines of Section III wer e reviewed with the licensee by telephone on June 19, 1989 A brief description of the licensee's response and the staff's evaluation of the responses against each of the review guidelines are pr ovided 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.
Subsequently, the licensee installed a
new plant process computer.
The new computer has a
one second sample interval for time history data and retains this detailed information covering the period from 10 minutes prior to the trip until ten minutes after the trip.
Further, this computer is powered from a one second sample interval for time history data and retains this detailed information covering the period from 10 minutes prior to the trip until ten minutes after the trip.
Further, this computer is powered, from a reliable and non-interruptible power source.
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
'identifed in Table 1.
The licensee does not record all of the parameters recommended in Section IIIB; however, alternate parameters may be used to implicitly determine the recommended parameter.
- Further, as noted below, the new plant computer records parameters over and above those included in the licensee's original response.
I Me find that the containment isolation signal is not recorded directly but is indirectly available by consulting containment pressure.
Control rod position is not recorded for all rods; however, the control rod bank position is now recorded by the plant computer as part of the analog data base.
Both containment sumps are fully instrumented and are now recorded on the plant computer.
MSIV position is now monitored on the plant
. computer.
Steam flow is,recorded as an analog signal on the plant computer, while the SOE records steam flow-related trip signals.
AC and DC bus voltages are not monitored directly; however, breaker status can be used as an indication that the AC electrical buses are energized.
DC buses have Normal/Off-Normal status indication.
The emergency diesel generators (EDGs) start on a safety injection signal which is recorded on the SOE.
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 of the EDG output breakers.
Although not required, recording of bus voltage and EDG output breaker status would be desirable.
In summary, most of the desirable plant parameters needed for post-trip review are now recorded by the licensee.
Alternative data 'sour ces 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 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. 'ased on our review, we find that the program for the retention of data conforms to the guide-lines of Section IIID.
V.
CONCLUSION Based on the foregoing discussion, the staff concludes that the licensee's post-trip review data andinformation capabilities for the R.
E. Ginna Nuclear Power Plant are acceptable for Item 1.2 of Generic Letter 83-28.
- However, recording of additional parameters as discussed in Section IVB, would serve to improve and expedite post-trip reviews.
Dated:
November 5, 1990 Principal Contributor:
W. H. Swenson
TABLE 1 PWR PARAMETER LIST SOE Recorder x
(1) x x
(1) x x
(1) x x
(2)
- 1) x 1I x (1) x 1
x (3)
Time History Recorder Parameter / Si nal Reactor Trip Saf ety Inject ion Containment Isolation Turbine 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 x
x
- 1) x 1
x 1
x (3)
HSIV Position Steam Generator Pressure Steam Generator Level Feedwater Flow Steam Flow Auxiliary Feedwater System; Flow, Pump/Valve Status AC and DC System Status (Bus Voltage)
Diesel Generator Status (Start/Stop, On/Off)
PORV Position I) 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:
~i NRC
& Local PDRs PDI-3 R/F S. Varga, w/o enc.
B. Boger, w/o enc.
N. Rushbrook A. Johnson R. Messman OGC (Ini'o. Only)
E. Jordan ACRS (10)
J. Johnson, RI M. Swenson, HS13E21 A. Chu E. Adensam, w/o enc.
R. Eckenrode, HS10D24