ML20246N032
| ML20246N032 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 07/11/1989 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20246N023 | List: |
| References | |
| GL-83-28, NUDOCS 8907190282 | |
| Download: ML20246N032 (9) | |
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ENCLOSURE SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION GENERIC LETTER 83-28, ITEM 1.2 TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR POWER PLANT, UNITS 1 AND 2 DOCKET N05.50-327 AND 50-328
1.0 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 was terminated manually by the reactor operator about 30 seconds after the initiation of the automatic trip signal. The failure of the circuit breaker was determined to be related to Ahe sticking of the undervoltage trip attachment. Prior to this accident, 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 startup. 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 (E00) directed the staff to investigate the generic implications of these occurrences at Unit 1 of the Salem huclear Power Plant. The results of the staff's inquiry into the generic implications of these events are reported in huREG-1000, " Generic Implications of the ATWS Events at the Salem Nuclear Pcwer Plant." As a result cf this investigation, the Commission (NRC) requested (by Generic Letter 83-28 dated July 8, 1983) all licensees of operating reactors, epplicants for an operating license, and holders of construction permits to respond to generic issues raised by the analyses of these two anticipated transients without scram (ATWS) events. This report is an evaluation of the responses submitted by Tennessee Valley Authority (TVA) for Item 1.2 of Generic Letter (GL) 83-28 for Sequoyah Nuclear ; Plant Units 1 and 2. The actual documents reviewed as part of this evaluation are listed in the reference section of this report. 2.0 DISCUSSION For Item 1.2 of GL 83-28, " Post-Trip Review, Data And Information Capability," licensees and applicants shall have or have planned a capability (1) to record, recall and display data and information to permit diagnosing the causes of unscheduled reactor shutdown, prior to restart, and (2) for ascertaining the proper functioning of safety-related equipment. 3907190282 890711 PDR P ADOCK 05000327 PDC
e g s . The intent of the post-trip review requirements in Item 1.2 is to ensure that the licensee has adequate data and information sources to understand the cause(s) and progression of a reactor trip. This understanding should go beyond a simple identification of the course of the event. It should include the capability to determine the root cause of the reactor trip and to determine whether safety limits have been exceeded and if so to what extent. Sutficient information about the reactor trip event should be available so that a decision on the deceptability of a reactor restart can be made. The equipment that provides the digital sequence of events (SOE) record and the analog time history records of an unscheduled shutdown should provide d reliable source of the necessary information to be used in the Each plant variable which is necessary to determine the cause(s) post-trip review. and progression of the event (s) following a plant trip should be monitored by at least one recorder [such as an SOE recorder or a plant process computer for digital parameters; and strip charts, a plant process computer or analog recorder for ' analog (time history) variables]. Each device used to record an analog or digital plant variable should be described in sufficient detail so that a determination can be made as to whether the following performance characteristics are met-
- 1. Each SOE 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 the plant final safety analysis report (FSAR)
Chapter 15 Accident Analyses. The recomanded guideline for the SOE time discrimination is approximately 100 msec. If current SOE recorders do not have this time discrimination capability the licensee or applicant should show that the current time discrimination capability is sufficient for an adequate reconstruction of the course of the reactor trip. As a minimum, this should include the ability to acequately reconstruct the accident scenarios presented in Chapter 15 of the plant FSAR.
- 2. 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 or applicant should be able to reconstruct the course of the accident sequences evaluated in the accident analysis of the plant FSAR (Chapter 15).
The recommended guideline for the sample interval is 10 seconds. If the time history equipment does not meet this guideline, the licensee or applicant should show that the current time history capability is sufficient to accurately reconstruct the accident sequences presented in Chapter 15 of the FSAR.
- 3. 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.
.__--________--___a
- 4. The information gathered by the SOE recorder and time history data .,
collectors should be stored in a manner that will allow for retrieval and analysis. The data may be retained in either hardcopy (computer printout, strip chart output, etc.) or in an accessible memory (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).
- 5. 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 sources used need not be safety related.
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 can be reconstructed. The parameters monitored should provide sufficient information to determine the root cause of the reactor trip, the progression of the reactor trip, and the response of the plant parameters and systems to the reactor trip. 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. Tfie parameters deemed necessary, as a minimum, to perform a post-trip review (one that would determine if the plant remained within its design envelope) are presented in the attached Table. If the applicant's or licensees' SOE recorders and time history recorders do not monitor all of the parameters suggested in the table, the applicant or licensee should show that the existing set of monitored parameters are sufficient to establish that the plant remained within the design envelope for the appropriate accident conditions; such as those analyzed in Chapter 15 of the FSAR. Information gathered during the post-trip review is required input for future post-trip reviews. 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 future unscheduled shutdowns. It is therefore necessary that information gathered during all post-trip reviews be maintained in an accessible manner for the life of the plant. 3.0 EVALUATION TVA responded to the requirements of Item 1.2 for Sequoyah with submittals dated November 7, 1983; and August 27 and September 7, 1984. Based on these submittals, the staff concluded that the post-trip review data and information capabilities are acceptable in the following areas:
- 1. The recorded data is output in a readable and meaningful format.
- 2. The SOE recorders meet the minimum performance characteristics.
l l l ) ___ - __ ____ ___ O
This is documented in the staff's letter dated August 15, 1985. The staff's evaluation is based on a Technical Evaluation Report (TER)-performed by the staff's contractor, Science Applications International Corporation. The staff has reviewed the TER (SAIC-85/1524-18) and agrees with the conclusions in the TER that Item 1.2 areas 1 and 2 listed above are acceptable. A copy of this TER is attached to this evaluation. The staff.also concluded that the
.information supplied indicates that the post-trip review data and information capabilities are not acceptable in the following areas:
- 1. Twelve of the parameters specified in the Table are not recorded for use in a post-trip review,
- 2. Time history recorders do not meet the minimum performance characteristics, and
- 3. The data retention procedures may not ensure that the information recorded for the post-trip review is maintained in an accessible manner for the' life of the plant.
The 12 parameters specified in the Table that are not recorded are the follow-ing: containment isoldtion, control rod position, containment radiation, containment sump injection flow and level, pump pressurizer
/ valve level,steam status, main primary system isolation flow, valve safety) position.
(MSIV l auxiliary feedwater system flow and pump / valve status AC and DC system status, diesel generator status and pressurizer power operated relief valve (PORV)-
- position. TVA cid not present sufficient information in its submittals to l justify not recording these parameters for use in a post-trip review.
In its submittals dated May 2, 1986 and December 14, 1987, TVA addressed the three issues listed above which the staff concluded were not acceptable for Sequoyah. 3.1 Parameters Recorded l l TVA stated that the Phase A and Phase B containment isolation signals, containment sump level, primary system ficw, safety injection flows (Train A and B), individual MSIV position (open/ closed) and auxiliary feedwater flows and valve status have been added to the Technical Support Center (TSC) computer for historical data storage. The containment radiation is monitored and the information is available on a continuous analog recorder. The pressurizer level exceeding the high level reactor trip setpoint is recorded on the SOE recorder and the pressurizer level is recorded on an analog level chart recorder. The status of the diesel generators is en orded using the electrical board Terminet 300 printer, which is located in the main control room. 1 l: R_________.__.___o
. 1 .
For control rod position, TVA explained that the reactor trip causal functions associated with the control rod malfunctions (i.e., overpower excursions, startup excursions, and rate trips) are monitored by the nuclear instrumentation ' I system and recorded by the SOE program on the plant process computer. Rod bottom lights are verified as an immediate operator action following a reactor trip to ensure that all control rods are fully inserted. TVA stated that because the trip functions associated with control rod malfunctions are monitored and recorded and the proper tripped condition is verified, recording the individual control rod positions as a function of time would not provide a significant benefit over the existing system. TVA stated that, during an anticipated trip without a scram event, the immediate action steps require the operator to physically open the reactor trip breakers in the auxiliary building. The SOE program records the time the trip breakers are opened, and the operator verifies the " rods at bottom" lights. TVA stated that, because of the high level of confidence that rods reach their "at-bottom" condition following the opening o' the reactor trip breakers, recording individual control rod positions as a function of time would not provide significant additional information for amlyzing the root cause of the trip or proper equipment operation. For the safety injection (SI) valve status data, TVA stated that this data is available on the TSC computer for the high head and low head SI flow paths. Valve status information for the intermediate head SI flow path is not provided. Valves within this flow path are not required to change positions upon an SI or reactor trip signal. The emergency procedures only require these valves to be operated 15 hours after switchover to the containment sump following a loss of coolant accident (LOCA). The addition of these valves to the historical data base would not aid in identifying improper valve operation following' a reactor trip, nor would it support on analysis for determining the root cause of a reactor trip. For the SI and auxiliary feedwater pump status (on/off), TVA stated that pump flow is indicative of pump status. Historical data exists within the TSC com-puter for pump flow; therefore, the acdition of pump status would not provide significant benefits for determining the rout cause of the reactor trip or for analyzing proper equipment operation. l For the AC/DC system status (bus voltages) TVA explained that the AC buses associated with safety equipment at SQN are the 6900-volt buses within the 6900-volt shutdown boards and the DC buses associated with safety equipment are the 125-volt vital power buses within SQN's 125-volt battery board. TVA stated that both sets of buses are munitored and are provided with " time tagged" alarms. Any abnormal condition (overvoltage, undervoltage, or failure) on these sets of buses is recorded on the electrical board ferminet 300 printer located in the main control room. The printer records the time of the event and identifies the bus that is experiencing the abnormal condition. TVA stated that once the voltage of the bus undergoing the abnormal condition is restored to normal status, the printer will again record the time and identify the bus that was restored. The addition of AC/DC voltages on a time
n history recorder would not provide significant benefits over the existing time tagged alarm printer system. The power supply for the logic system that inputs to the printer is the 48-volt de plant battery board. The control room printer receives power from the 120-volt ac preferred power system with battery backup power. Both power supplies provide noninterrupted power. For the PORV indication, TVA stated that the following five parameters, which are indicative of the PORV position, are stored on the TSC computer:
- 1. Acoustic Monitoring: single point digital,
- 2. Tailpipe Temperatures: Two analog signals--one for each line downstream of each PORV,
- 3. Pressurizer Relief Tank (PRT) Temperatures: One analog signal provides indication of PRT temperature,
- 4. PRT Level: One analog signal provices indication of PRT level, and
- 5. PRT Pressure: One analoi signal provides indication of PRT pressure.
TVA stated that each of these five parameters is an indirect function of PORV position, the addition of direct PORY position indication to the 50E recorder or a time history recorder would not provide significant benefits over the existing set of paramaters. The staff evaluated the information, in TVA's letters dated May 2, 1986 and December 14, 1987, on the parameters to be recorded to provide sufficient information about the reactor trip event so that a proper decision can be made on the acceptability of a reactor restart. TVA is recording all of the parameters listed in the Table either directly or indirectly. Where the parameter is measured indirectly, the indication is acceptable to the staff. Based on this, the staff concludes that Sequoyah has an acceptable list of recorder parameters to provide sufficient information about the reactor trip event. This addresses the staff's concerns about parameters being recorded in Item 1.2 of GL 83-28. 3.2 Alternative Recorders Performance Characteristics l The recording is on the SOE recorder, a time history recorder (the plant i process computer) or an acceptable alternative. The TSC computer has the j following sample interval and time history capability: (1) A 10-second sample interval extends 5 minutes pre-trip and 5 minutes post-trip, (2) a 1-minute sample interval extends 36 minutes post-trip, and (3) a 24-hour time history exists post-trip with a 5 minute sample interval. Based on these capabilities, the staff concludes that an additional 5 minutes of post-trip data with a 10-second sample interval would not significantly support the post-trip analysis for determining the cause of the trip. The containment radiation and pressurizer level is recorded on a continuous analog recorder. The Terminet 300 printer records the time (1) when the diesel. generator starts h - - - - - . -m---m___m.m. __m _ . _ _ _ _ _ . _ _m____.[)___
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-7 or stops, (2) when there is an abnormal event on the 6900-volt buses and 125-volt buses, identifying the bus and (3) when the bus was returned to normal status.
The staff concludes that the TSC computer; containment radiation and pres-surizer level analog recorder; and that the Terminet 300 printer meets the staff's concerns about equipment recording capability in Item 1.2 of GL 83-28 and are acceptable. 3.3 Time History Recorder The time history recorder, the piant process computer, had a sample interval of two or eight seconds, but did not have a storage capacity of five minutes for the pre-trip and ten minutes for the post-trip. The sample interval was provided in the TVA letter dated November 7, 1983. TVA stated that the plant computer storage capacity has been extended to five and 10 minutes, respectively. Therefore, the staff concludes that the time history recorder is acceptable and addresses the staff's concerns about equipment recording capability in Item 1.2 of GL 83-28. 3.4 Data Retention Capability TVA stated that the reactor trip report is made using an administrative instruction which ensures that all applicable recorder charts and computer printouts are included in the report. The report is stored for the life of the plant as a retrievable quality assurance record. The staff concludes that l this addresses the staff's concerns about data retentien in Item 1.2 of GL 83-28.
4.0 CONCLUSION
Based on the above, the staff concludes that TVA has acceptably addressed the staff's concerns in Item 1.2 of GL 83-28.
5.0 REFERENCES
- 1. Letter from D. G. Eisenhut, (NRC) to all Licensees of Operating Reactors, Applicants for Operating Licenses, and Holders of Construction Permits,
Subject:
Required Actior< 53ased on Generic Implications of Salem ATWS Events (Generic Letter 83 28), dated July 8, 1983. I
- 2. Letter from L. M. Mills (TVA) to E. Adensam, (NRC), dated November 7, 1983.
- 3. Letter from L. M. Mills (TVA) tc E. Adensam (NRC), dated August 27, 1983.
- 4. Letter from L. M. Mills (TVA) to H. R. Denton (NRC), dated September 17, 1984. ,
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* : 5. . Letter from R. Gridley. (TVA) 'to'B. Youngblood. (NRC), dated May 2,1986.
- 6. Letter from R. Gridley -(TVA) to NRC, dated December ~14,1987.-
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' Principal Contributor: J..Dononew.
Dated: July 11 1989 1 , :. \'
r, ? ;_ . g , .O .. }* .: TABLE PWR Parameter List
- SOE Time History Recorder Recorder Parameter / Signal <
X- Reactor Trip (1)X Safety Injection X- Containment Isolation (1)X Turbine Trip
.X Control Rod Position 1)X X Neutron Flux, Power X X Containment Pressure (2) Containment Radiation X Containment Sump Level 1)X X Primary System Pressure-1)X X Primary System Temperature X Pressurizer Level X- Reactor Coolant Pump Status X Primary System Flow Safety Injection System: Flow, Pump / Valve Status X MSIV Position X X Steam Generator Pressure 1)X X Steam Generator Level 1)X X Feedwater Flow I'X X Steam Flow (3) Auxiliary Feedwater System:
Flow. Pump / Valve Status
'X AC and DC System Status (Bus Voltage)
X Diesel Generator Status (Start /Stop,On/Off) X 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
- Sequoyah is a Westinghouse PWR
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Ef f',1 :.' . : REVIEW.0F. LICENSEE AND APPLICANT RESPONSES f/ : ' *.? . . T0 NRC GENERIC LETTER 83-28
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. Salem ATWS Events). Item 1.2
- V. ' c " POST-TRIP REVIEW: . DATA AND INFORMATION CAPABILITIES" FOR
'i.'y ". .SEQUOYAH NUCLEAR PLANT, UNITS 1 & 2 (50-327, 50-328)
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1This' report contain.:s the technical evaluation of the Sequoyah Nuclear
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i .F. . on Generic Implications of Salem ATWS Events), Item 1.2 " Post Trip Review: I
$.;-.. ' Data and Information Capabilities.".
fbER For the purposes of this evaluation, the review criteria, presented in Y j4 " part 2 of this report, were divide'd into five separate categories. These '
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- 1. The parameters' monitored by the sequence of events and the time y,k.i[ ....
1 . histery recorders, i[h'd 2. ' The pt: formance" characteristics of the sequence of events k[/ recorders,
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[;(+f - The long term data retention capability for post-trip review material,
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All available responses to Generic Letter 83-28 were evaluated. The l.'l . plant for which this report is applicable was found to have adequately 3 .'
. responded to, and met, categories 2 and 4
'c. The report describes the specific methods used to determine the cate-
;. gerization of the responses to Generic Letter 83-28. Since this evaluation
' report was intended to apply to more than one nuclear power plant specifics regarding how each plant met (or failed to meet) the review criteria are not I" presented. Instead, the evaluation presents a categorization of the i,.. resportses according to which categories of review criteria are satisfied and l
[li- which are not. The evaluations are based on specific criteria (Section 2) [,t derived from the requirements as stated in the generic letter. G.
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- 1. Background. . . . . . . ... . . . . . . . . . . . . . . . . .
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Review Criteria . . . . . . . . . . . . . . . . . . . . . . . 3
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- 3. . Evaluation. . . . . . . . . . . . . . . . . . . . . . . . . . 8
- 4. Conclusion. . . . . . . . ... . . . . . . . . . . . . . . . . 9
. S. R e fe re nc e s . . . . . . . . . . . . . . . . . . . . . . . . . . 10 A a g l
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- i SAIC has reviewed the utt11ty's response to Generic 1.etter 83 28, item {
j;se$$'.+.'j jj1 y,. 1.2 " Post Trip Review: Data and Information Capability." The response (see
,;;/yTp references) contained sufficient information to deters.Ine that the data and ;
information capabilities et these plants are acceptable in the following
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e The sequence of events recorder (s) performance charac.
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. e The output f.ormat of the recorded data.
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( However, the data and information capabilities, as described in the
' ,t submittal, either fail to meet the review criteria or provide insufficient
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? I information to allow determination of the adequacy of the data and information capabilities in the following areas.
e The parameters monitored by both the sequenr.e-of events and time history recorders, e Tre time history recorder (s) performance characterts-tics.
'f. , ,1 e The long term data retention, record keeping, capa-7,: ' '.! bility, t..a
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, On February 25,'.1984 both of the scram circuit breakers at Unit 1 of
.I'd h j 'the. . . Salem
, : . Nuclear
, 9 .-Power Plant. failed t'o open upon an automatic reactor trip
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signal from the reactor;.'p'retection system. . This incident occurred during j 3 {$ ,-, the piant startup and the reactor.,has tilpp'ed manually by the operator about I M' 30 seconds after the initiation of'the automatic trip signal. The failure i
'f'r.4...i hf. rl . of the circuit breakers 'has b'een determined to be related to the sticking of f40
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j^ i ~ was 1 ,,generated the based under voltage trip attachment.. Prior 1983; at Unit 1 of the Salein Nuclear Powe'r Plant an automatic trip signal on steam generator lolw low level during plant startup.
$, ,' ' In this case the r'eactor was tripped manually by the operator almost coinct-
[; dentally with the automatic trip. At that time, because the utility did not S,0 f have a requirement for the systematic evaluation of the reactor trip, no "Ch ,'
, investigation was performed.to determine whether the reactor was tripped g* Automatt': ally as expected or manually.. The utilities' written procedures '
p, [ requiret only that the cause of the trip be determined and identified the
'('lJ j responsiC a personnel that could authorize a restart if the cause of the '
% '; trip is known. Following the 'second trip which clearly indicated the 5 i problem with the trip breakers, the question was raised on whether the f circuit breakers had functioned properly during the earlier incident. The most useful source of information in this case, namely the sequence of
, events printout which would have indicated whether the reactor was tripped automatically or manually during the February 22 incident, was not retained
, af ter the incident. Thus, no judgment on the proper functioning of the trip system during the earlier incident could be made.
( Following these incidents; on February 28, 1983; the NRC Executive 7,l , Of rector for Operations (E00), directed the staff to investigate and report
.'I ' e - on the generic implications of these occurrences at Unit 1 of the Salem
/M Nuclear Power Plant. .The results of the staf f's inquiry into the generic y $,
implications of the Salem Unit incidents is reported in NUREG 1000.
- Generic I
g.,l Implications of ATWS Events at the Salem Nuclear Power Plant." Based on the f", d ' results of this study, a set of required actions were developed and included T.51 in'Generi'd 1.etter 83 28 which was issued on July 8,1983 and sent to all JJ licensees of operating reactors, applicants for operating license, and EJ 'N i- construction permit holders. The required actions in this generic letter l $"
.Jo; consist of four categories. These are (1) Post Trip Review. (2) Equipment l
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Classification and Vender Interface, (3) Post Maintenance Testing, and (4) Reactor Trip System Rentability Improvements. The first required action of the generic letter, Post Trip Review. Is the subject of this TER and consists of action item 1.1 " Program Description and Procedure" and action item 1.2 " Data and Information Capability." In the next section the review criteria used to assess the adequacy of the utilities' responses to the requirements of action item 1.2 will be discussed.
- 2. Review Criteria The intent of the Post Trip Review requirements of Generic Letter 83-28 is to ensure that the licensee has adequate procedures and data and information sources to understand the cause(s) and progression of a reactor trip. This understanding should go beyond a simple identification of the course of the event. It should include the capability to determine the root cause of the reactor trip and to determine whether safety limits have been exceeded and if so to what extent. Sufficient information about the reactor trip event should be available 50 that a decision on the acceptability of a reactor restart can be made.
The following are the review criteria developed for the requirements of Generic Letter 83 28, action item 1.2: The equipment that provides the digital sequence of events (SOE) record and the analog time history records of an unscheduled shutdown should pro-vide 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(s) and progression of the event (s) 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 plant process computer or analog recorder for analog (time history) variables). Each device used to record an analog or digital plant variable should be described in sufficient detail so that a determination can be made as to whether the following performance characteristics are met: 3
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C i e Each sequence-of-events recorder should be capable of detecting and recording the sequence of events with a suf ficient time discrimination capability to ensure that the time responses asso-ciated with each monitored safety-related system can be ascer-tained, and that a determination can be made as to whether the time' response is within acceptable limits based on FSAR Chapter 15 Accident Analyses. The recommended guideline for the SOE time discrimination is approximately 100 msec. If current SOE recorders do not have this time discrimination capability the licensee or applicant should show that the current time discrimi.
nation capability is suf ficient for an adequate reconstruction of the course of the reactor trip. As a minimum this should include the ability to adequately reconstruct the accident scenarios pre-sented in Chapter 15 of the plant FSAR. e Each analog time history data recorder should have a sample inter-val small enough so that the incident can be accurately ' reconstructed following a reactor trip. As a minimum, the licensee or applicant should be able to reconstruct the course of the accident sequences evaluated in the accident analysis of the plant FSAR (Chapter 15). The recommended guideline for the sample interval is 10 sec. If the time history equipment does not meet this guideline, the licensee or applicant should show that the current time history capability is sufficient to accurately recon-struct the accident sequences presented in Chapter 15 of the FSAR. e 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. o The information gathered by the sequence-of-events and time history data collectors should be stored in a manner that will allow for retrieval and analysis. The data may be retained in either hardcopy (computer printout, strip chart output, etc.) or in an accessible memory (magnetic disc or tape). This information should be presented in a readable and meaningful format, taking 4 L - -- --- -- - -
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into consideration good human factors practices (such as those outlined in NUREG-0700). e- All equipment used to record sequence of events and time history information sho'uld be powered from a reliable and non-interruptible power source. The power source used need not be safety related. , 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 can be reconstructed. The parameters monitored should provide sufficient information to determine the root cause of the reactor trip, the progression of the reactor trip, and the response of the plant parameters and systems to the reactor trip. 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 , tr o review. The parameters deemed necessary, as a minimum, to perform a post-trip review (one that would determine if the plant remained within its design er.velope) are presented on Tables 1.21 and 1.2-2. If the appli-cants' or lice'. tees' SOE recorders and time history recorders do not monitor all of the parameters suggested in these tables the applicant or licensee should show that the existing set of monitored parameters are sufficient to establish that the plant remained within the design envelope for the appro-priate accident conditions; such as those analyzed in Chapter 15 of the plant Safety Analysis Report. Information gathered during the post trip review is required input for future post trip reviews. Data from all unscheduled shutdowns provides a f valuable reference source for the determination cf the acceptability of the l plant vital parameter and equipment response to future unscheduled shut-downs. It is therefore necessary that information gathered during all post trip reviews be maintained in an accessible manner for the life of the plant. 5 i
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. Tabl e 1.2-1. PWR Parameter List 1
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% SOE Time History -
,. Recorder Recorder Parameter / Sional i x Reactor Trip (1) x Safety injection x containment Isolation )
(1) x Turbine Trip l x Control Rod Position-(1)x x Neutron Flux, Power x x Containment Pressure I. (2) Containment Radiation j x Containment Sump Level (1)x x Primary System Pressure (1) x x Priniary System Temperature (1) x , Pressurizer Level , (1)x Reactor Coolant Pump Status (1) x x Primary System Flow (3) Safety Inj.; Flow. Pump / Valve Status x MS!V Position x r Steam Generator Pressure (1) x x Steam Generator Level (1)x x Feedwater Flow (1) x x Steam Flow (3) Auxiliary Feedwater System; Flow. Pump /Value Status x AC and DC System Status (Bus Voltage) x Diesel Generator Status (Start /Stop, On/Off) x PORY 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 In SOE recorder (b) system flow recorded on a time history recorder, or (c) ' equipment status recorded on an SOE recorder.
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Table 1.2 2. BWR Parameter List 1
,; SOE Time History !
Recorder Recorder Parameter / Signal
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x Reactor Trip , x Safety Injection I x containment Isolation x Turbine Trip
, x control Rod Position l x (1) x Neutron Flux, Power x (1) Main Steam Radiation (2) Containment (DryWell) Radiation x (1) x Drywell Pressure (Containment Pressure)
(2) Suppreeston Pool Temperature
-x (1) x Primary System Pressure x (1) x Primary System Level x MS!Y Position .
x (1) Turbine Stop Valve / Control Valve Position x Turbine Bypass Valve Position x Feedwater Flow x Steam Flow (3) Recirculation; Flow. Pump Status x (1) Scram Discharge Level x (1) , Condenser Yacuum x AC and DC System Status (Bus Voltage) (3)(4) Safety Injection; Flow Pump / Valve Status x Diesel Generator Status (On/Off,
, Start /Stop)
(1): Trip parameters. (2): Parameter may be recorded 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)equipmentstatusrecordedonanSOErecorder. (4): Includes recording of parameters for all applicable systems from the following: HPCI, LPCI, LPCS, !C, RCIC. 7 O ut ., .. . ... .... . . '
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..j 3. Evaluation
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The parameters identified in part 1 of this report as a part of the
;I review criteria are those deemed necessary to perform an adequate post-trip
, review. The recording of these parameters on equipment that meets the guidelines of the review criteria will result in a source of information that can be used to determine the cause of the reactor trip and the plant response to the trip, includir.g the responses of important plant systems.
The parameters identified in this submittal as being recorded by the
'. sequence of events and ting history recorders do not correspond to the parameters specified in put 2 of this report. '
The review criteria require that the equipment being used to record the sequence of events and time history data required for a post-trip review meet certain performance characteristics. These characteristics are intended to ensure that, if the proper parameters are recorded, the record-ing equipment will provide an adequate source of information for an effec- ' tive post-trip review. The information provided in this submittal does not indicate that the time history equipment used would meet the intent of the performance criteria outlined in part 2 of this report. Information supplied in the submittal does indicate that the SOE equipment meets the performance criteria specified in part 2 of this report. The data and information recorded for use in the post-trip review should be output in a format that allows for ease of identification and use of the data to meet the review criterion that calls for information in a readable and meaningful format. The information contained in this submittal indicates that this criterion is met. The data and information used during a post-trip review should be retained as part of the plant files. This information could prove useful during future post-trip reviews. Therefore, one criterion is that infor-mation used during a post-trip review be maintained in an accessible manner for the life of the plant. The information contained within this submittal does not indicate that this criterion will be met. 8 1 _ _ _ _ _ - - - - _ _ - - ._-___------__a-.
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The information supplied in response to Generic Letter 83-28 indicates that the current post-trip review data and information capabilities are ] adequate in the following areas- 1 a
- 1. The recorded data is output in a readable and meaningful format.
- 2. The sequence of events recorders meet the minimum performance characteristics.
The information suppited in response to Generic Letter 83-28 does not indicate that the post-trip review data and information capabilities are adequate in the following areas:
- 1. Based upon the information contained in the submittal, all of the parameters specified in part 2 of this report that should be ,
recorded for use in a post-trip review are not recorded.
- 2. Time history recorders, as described in the submittal, do not meet the minimum performance chaNeteristics.
- 3. The data retention procedures, as described in the submittal, may not ensure that the information recorded for the post-trip review is maintained in an accessible manner for the life of the plant.
It is possible that the current data and information capabilities at this nuclear power plant are adequate to meet the intent of these review criteria, but were not completely described. Under these circumstances, the licensee should provide an updated, more complete, description to show in more detail the data and information capabilities at this nuclear power plant. If the information prov' 'ad accurately represents all current data and information capabilities, then the licensee should show that the data
, and information capabilities meet the intent of the criteria in part 2 of this report, or detail future modifications that would enable the licensee .
to meet the intent of the evaluation criteria. 9
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. ,E REFERENCES NRC Generic Letter 83 28. " Letter to all licensees of operating reactors ' applicants for operating license, and holders of construction j permits regarding Required Actions Based on Generic Implications of l Salem ATWS Events.' July 8, 1983.
NUREG-1000, ' Gen'eric Implications of ATWS Events at the Salem Nuclear _. Power Plant April 1983.
' Letter' from L.M. Mills. Tennessee Valley Authority, to E. Adensam. NRC, dated November 7,1983, Accession Number 8311150111 in response to Generic Letter 83 28 of July 8,1983, with attachment.
Letter from L.M. Mills. Tennessee Valley Authority, to E. Adensam NRC. I dated August 27, 1984 Accession Number 8409050421 providing response ) to the NRC SER transmitted to the Westinghouse Owners Group. l l Letter from L.M. Mills. Tennessee Valley Authority, to H.R. Denton, ! NRC, dated September 17, 1984 Accession Number 8409250381 f.Oviding supplemental response to Section 2.2.2. 1 9 4 e
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