ML20244D387

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Review of Licensee & Applicant Responses to NRC Generic Ltr 83-28 (Required Actions Based on Generic Implications of Salem ATWS Events),Item 1.2, `Post-Trip Review:Data & Info Capabilities' for Bellefonte Nuclear Plant.
ML20244D387
Person / Time
Site: Beaver Valley, Millstone, Dresden, Davis Besse, Nine Mile Point, Palo Verde, Kewaunee, Wolf Creek, Hope Creek, Sequoyah, Pilgrim, Prairie Island, Turkey Point, River Bend, Vermont Yankee, Ginna, Waterford, Cook, Maine Yankee, Rancho Seco, Zion, Bellefonte, FitzPatrick, 05000000, Trojan, 05000246
Issue date: 07/10/1985
From:
SCIENCE APPLICATIONS INTERNATIONAL CORP. (FORMERLY
To:
NRC
Shared Package
ML20235V135 List:
References
CON-NRC-03-82-096, CON-NRC-3-82-96, FOIA-87-644 GL-83-28, SAIC-85-1524, NUDOCS 8507160408
Download: ML20244D387 (29)
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REVIEW OF LICENSEE AND APPLICANT RESPONSES'. , ,

.TO NRC. GENERIC: LETTER 83-28 . .

~vs (Required Actions Based on' Generic Implications of-Salem ATWS Events), Item:1.2: e

" POST-TRIP REVIEW: ' DATA AND INFORMATION CAPABILITIES *(FOR . ., T '

BEAVER VALLEY POWER STATION,: UNIT;1c(50-334)L. q BELLEFONTE NUCLEAR PLANT, UNITS.1 AND 2 (50-438,c50-439); t 1 DAVIS-BESSE NUCLEAR POWER. STATION, UNIT: l' (50-'346)1. L f

DONALD C.! COOK NUCLEAR POWER PLANT,: UNITS l' AND 2-(50-315, 50-316):

DRESDEN NUCLEARLPOWER STATION,LUNITS 24 AND 3. (50-237. 501249) ,

, a JAMES A. FITZPATRICK- NUCLEAR: PLANT:(50-333) ' i >

o1 ROBERT EMMET GINNA' NUCLEAR PLANTc UNIT l'(50-244)', ." /i n j , ,

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HOPE CREEK NUCLEAR STATION' , UNIT I'(50-354): ,

s KEWAUNEE. NUCLEAR POWER PLANT (50-305)c  :;. , H MAINE' YANKEE ATOMIC POWER PLANT..(50-309)'

MILLSTONE NUCLEAR POWER STATION,' UNITS .1,: 2 ; AND '3' (50-245, 50-335,! 50-423); ,

NINE MILE POINT NUCLEAR STATION. UNIT 11(504220))

1 PALO VERDE NUCLEAR STATION, UNITS 1,' 2, AND 3 (50-528; 50-529,~.50-530)) ,l PILGRIM STATION'(50-293) '

PRAIRIE ISLAND NUCLEAR GENERATING PLANT,; UNITS 1- AND 2-(50-282 . 50-306) ,

RANCHO SECO NUCLEAR GENERATING STATION (50-312)'

RIVER BEND STATION, UNIT l' (50-458) L ..

SEQUOYAH NUCLEAR PLANT, UNITS 1 AND 2:(50-327, 50-328).

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TROJAN NUCLEAR PLANT (50-344) .

TURKEY POINT PLANT, UNITS 3. AND 4' (50-250, 50-251) ,

VERMONTYANKEENUCLEARPOWERSTATION(50-271).

WATERFORD GENERATING STATION, UNIT, 3:(50-382) s WOLF CREEK GENERATING STATION:(50-482);

ZION NUCLEAR POWER STATION, UNITS 1 AND 2 (50-295,' 50-304) 1 l July 10, 1985 1

Technical Evaluation, Report Prepared by Science Applications International Corporation 4 1710 Goodridge Drive McLean, Virginia 22102 1 i K 4 Prepared for- ,

L U.S. Nuclear Regulatory Commission Washington, D.C. 20555"  :

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This TER, includes SAIC's review of the utility's- submittals in response to Generic Letter 83-28, Item -1.2. " Post-Trip Review: . Data and . Information Capability" for. the plants listed in Table 1.

Thest submittals contained- sufficient information to determine that the.

data and'information capabilities at these. plants are acceptableLin.the' -

following areas.. '

e The sequence-of-events' recorder (s) perfomance characteristics. .

0l m,. 1 The . output format;of the recorded data. ' \

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However, the data and information capabilities, as described in the '!q submittal, either fail to meet .the review criteria or provide insufficient?

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infoqsation to allow determination of the adequacy of---the data and -

infMmation 4 capabilities in- the' following areas..  !

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sThe parameters z recorded by both the'seque'nce-of-events;and time - M

/ history recorders. . >

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,I e The ,long-term data retention, record keeping, capability.  !!

s For your convenience., the results.of our review of these submittals'.are' j1 reproduced for bahn f ndiiidual plasti with' a . plant-specific cover sheet so j t

L' that these evaluations v 1.

c4f1 be transmitted m to each utility individually. The. d SAIC report numpic 3 associated with each individual plant levaluationLis 0 provided in Tablejl. Nditionally, the ' supporting information' showing"the' L

detailed review' df each's@mittal-is provideh in a separat'e TER, SAIC-'- 1 M'

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Table 1 Nuclear Power Plant SAIC heport' Number Beaver Valley Power Station,-Unit 1 <

SAIC-85/1524 J1 Bellefonte Nuclear Plant , Units 1 and 2 LSAIC-85/1524-2'

- Davis-Besse Nuclear' Power Station, Unit 1- SAIC-85/1524-3 Donald C. Cook Nuclear Power Plant Units.1 and 2 : SAIC-85/1524-4 Dresden Nuclear Power Station, Units 2 and 3 SAIC-85/1524-5 James A. FitzPatrick Nuclear Plant - ' SAIC-85/1524-6 Robert Emmet Ginna Nuclear. Plant, Unit 1 - SAIC-85/1524-7 .

Hope Creek Nuclear Station Unit 1 SAIC-85/1524-8 Kewaunee Nuclear Power Plant SAIC-85/1524-9 .

Maine Yankee Atomic Power Plant S AIC-85/1524-10 Millstone Nuclear Power Station, Units 1, 2 and 3 SAIC-85/1524-11 Nine Mile Point Nuclear Station, Unit 1 ' S AIC-85/1524-12; Palo Verde Nuclear Station, Units 1,'2 and 3- SAIC-85/1524-13 i Pilgrim Station S AIC-85/1524-14 '

Prairie Island Nuclear Generating Plant, Units 1 and 2' SAIC-85/1524-15:

Rancho Seco Nuclear Generating Station S AIC-85/1524-16 River Bend Station, Unit 1 SAIC-85/1524 Sequoyah Nuclear Plant, Units 1 and 2 .SAIC-85/1524-18 Trojan Nuclear Plant SAIC-85/1524-19 Turkey Point Plant Units 3 and 4 S AIC-85 /1524-20.

Vermont Yankee Nuclear Power Station . SAIC-85/1524 Waterford Generating Station, Unit 3 SAIC-85/1524-22 Wolf Creek Generating Station SAIC-85/1524-23 Zion Nuclear Power Station, Units 1 and 2 SAIC-85/1524-24 i

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REVIEW OF LICENSEE ANDLAPPLICANT RESPONSES.

.T0 NRC GENERIC LETTER 83-28> . , ,: -

. (Required Actions -Based.on: Generic Implications 'of- 1 y

' Salem ATWS Events) Item;1.2' . . .

.c POST-TRIP REVIEW: : DATA AND INFORMATION - CAPABILITIES"' FOR J -

BEAVER VALLEY ; POWER . STATION,' UNIT ~ 1 )(50-334) , ' ;n l i

l-l Technical Evaluation R& port Prepared by '

j Science Applications International Corporation  :

-l 1710 Goodridge Drivel /1 McLean, Virginia 22102,

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Prepared.for, 3. . ;,  ;

o U.S. Nuclear Regulatory Commission; Washington,l0.C. 20555 > <

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Contract No. NRC-03-82-096 ,

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FOREWORD This report contains the technical evaluation of the Beav' r Valley.

Power Station, Unit I response to Generic Letter 83-25 (Required Actions.

Based on Generic Implications of Salem ATWS Events). Item 1.2:" Post Trip Review: Data and Information Capabilities."'

For the purposes of' this eva'iuation, the review -criteria, presented 'in part 2 of this report, were divided into five separate categories. These are:

1. The parameters monitored by the sequence of' events.and the time =

history recorders,

2. The performance characteristics of the sequence of events recorders.

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3. The performance characteristics of the time history recorders,
4. The data output format, and
5. The long-term data retention capability fra post-trip review l material.

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All available responses to Generic Letter 83-28 were evaluated. 'The i

plant for which this report is applicable was found to have adequately responded to, and met, categories 2 and 4. 'l The report describes the specific methods used to determine the cate-gorization of the responses to Generic Letter- 83-28, 5'nce this evaluation -

report was intended to apply to more than one nuclear. p:wer plant specifics regarding how each plant met (or failed to meet) the redew criteria are not presented. Instead, the evaluation presents a . categorization of the responses according to which categories of review criteria are satisfied and ,

l which are not. The evaluations are based on specific :riteria (Section 2)

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derived from the requirements as stated in the generic letter.

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g 4 TABLE OF CONTENTS Section ,

. Pa ge .

Introduction.-... . . . . . . . . . . . . . .:.. .=. . .,. . . . '1'

1. Background. . . . . . . . . . . . . ... . . . . .:. . . . ... 2 2.- Review Criteria . . . . ... . . . . . . . . . ... ... . . .. . 3-
3. Evaluation. . . . . . . . . . . . . . . .

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4. Co n cl u s i o n . . . . . . . . . , . . . . . . . . . . . . . . . . . . . . 9
5. Re fe re n c e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10' l

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't INTRODUCTION l

SAIC has reviewed the utility's response to Generic Letter 83-28, item - ,

1.2 " Post-Trip Review: Data and Information Capability.* ' The' response. (see- l references) contained sufficient information to determine that the data and .

information capabilities at these plants are' acceptable in the following_

areas.

e The sequence-of-events recorder (s) performance charac-teristics. ,

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e The output format of the recorded data. .

However, the data and information capabilities, as described-1'n.the .q submittal, either fail to meet the review criteria or. provide insufficient j information to allow determination of the adequacy of. the' data and.

information capabilities in the following areas. ,

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l e The parameters monitored by both the sequence-of-events -

and time history recorders.

I e The time history recorder (s) performance characteris- 'q tics. i i

e The long-term data ' retention, record k'eeping,- capa- .

bility.

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1. Background On February 25, 1984, 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 startup 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 Febraary 22, 1983; at Unit 1 of the Salem Nuclear Power Plant an automatic'tri;. signal was generated based on steam generator. low-low level daring plant startup.
n this case the reactor was tripped manually by the operator almost coinci-dentally with the automatic trip. At that time, because the utility did not have a requirement for the systematic evaluation of the reactor trip, no investigation was performed to determine whether the reactor was tripped automatically as expected or manually. The utilities' written precedures required only that the cause of the trip be determined and identified the responsible personnel that could authorize a restart-if the cause of_ the trip is known. Following the second trip which clearly indicated the problem with the trip breakers, the question was raised on:whether the i circuit breakers had functioned properly during the earlier incide5t. The most useful source of information in this case, namely the sequence of events printout which would have indicated whether the reactor was tripped I automatically or manually during the February 22 incident, was not retained after 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 Director for Operations (EDO), directed the staff to investigate and report on the generic implications of these occurrences at Unit 1 of tre Salem i Nuclear Power Plant. The results of the staf f's inquiry into the generic  ;

implications of the Salem Unit incidents is reported in N'JREG-1000, " Generic-Implications of ATWS Events at the Salem Nuclear Power Plant." Based or the results of this study, a set of required actions were developed and included in Generic Letter 83-28 which was issued on July 8,1983 and sent to all licensees of operating reactors, applicants for operating licerse, and construction permit holders. The required actions in tnis generic letter-consist of four categories. These are: (1) Post-Trip Review, (2) E:;uipment 2

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Classification and Vender -Interface, (3) Post' Maintenance -Testing, and (4).-

' Reactor Trip System Reliability Improvements.,

l The' first . required' action o'f the generic letter, Post-Trip Review.. is the subject of this TER and consists of action item 1.1 " Program Description and Procedure"innd action item 1.2 " Data and Information Capability." 1.i the next section the review criteria used to assess.the adequacy of. the' utilities' responses .to the requirements ~ of action item; 1.2 will-'bei

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discussed.

2. Review Criteria The intent of the Post Trip Review requirements of Generic Letter 83-28 is to ensure that. the. licensee nasl adequate procedures .and; data '.and-information sources to understand the cause(s) andL 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 roote

- i cause of the reactor trip and to determine whether safety limits t.have: been:

exceeded and if so to what extent. Sufficient ~information~ about the ' reactor trip event should be available'so that.a decision on the acceptability of a reactor restart can be made, a The fol' lowing are the review criteria devel'oped .for the requirements 'of Generic Letter 83-28, action item 1.2:

The equipment that provides the digital sequence cf 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 bel described in sufficient detail so that a determination can be' made as to-whether the following performance characteristics are met:

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, ,1 e Each sequence-of-events recorder should be capable' of detecting ]

and recording the sequence' of events with a ' sufficient time- j discrimination capability to ensure .that~ the time . responses asso-:

ciated with eaci) monitored safety-related system'can be ascer ' .

tained, and that a determination can be made as to whether the I time response is within acceptable'11mits based on FSAR Chapter 15 '!

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Accident Analyses. The recommended guideline fo'r,.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- .f nation capability is sufficient for.an adequate reconstruction of the course of the reactor trip. As a minimum this, should include j the ability to adequately reconstruct the accident scenarios pre- 1 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 j licensee or applicant should be able to' reconstruct the course of - .

the accident sequences evaluated in the accident analysis of .the l i plant FSAR (Chapter 15). The recommended guideline-fcr the sample interval is 10 sec. If the time history. equipment . 60es .not ' meet-this guideline, the licensee or applicant should show' that the current time history capability is' sufficient to accurately recon- ;l struct the accident sequences presented in Chapter 15 of the FSAR.

o e To support the post trip analysis of the caus'e of'the trip and the -l proper functioning of involved safety related ' equipment, each- '

l analog tim history data recorder should be capable of updating- -J and retaining information' from approximately five:minates prior to' I the trip until at least ten minutes after the trip. i L e The information gathered by the sequence-of-events'. and time j history data collectors should be stored in a manner that will i allow for retrieval and analysis. The data may be retained in-either hardcopy (computer printout, strip chart out;ut, etc.) or i in an accessible memory (magnetic disc or. tape). This information should be presented in a readable and meaningful format, taking 4

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into consideration good human factors practices.(such as those -

outlined in NUREG-0700). -

o All equipment used to record sequence of events and time history information should' be powered. from 'a rel.iable and non-interruptible power source. The power source used need not~ be safety related.

The sequence of events and time history recording equipment shoul'd monitor sufficient digital and analog parameters, respectively to assure-that the course of the reactor trip can be reconstructed._ The parameters zunitored should provide sufficient'information to deternine 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. The para' meters deemed necessary, as a minimum sto. perform a '

post-trip review (one that would determine. if the plant remained within its design envelope) are presented on Tables 1.2-1 and 1.2-2. If the appli-cants' or licensees' SOE recorders and time history recor:!ers 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 .statdowns provides a- l valuable reference source for the determination of the a:ceptability 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 j'

plant.

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Table 1.2-1.. 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 (1) x Pressurizer Level (1) x Reactor Coolant Pump Status (1) x x Primary System Flow (3) Safety Inj.; Flow, Pur.;/ Valve Status x MSly Position x x 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 PORV- Position

.i (1): Trip parameters (2): Parameter may be monitored by either an SOE or time history recorder. j (3): Acceptable recorder options are: (a)' system flow r'ecorded on an SOE recorder (b) system flow recorded on a time history recorder, or (c) l equipment status recorded on an SOE recorder.

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e Table 1.2-2. BWR Parameter List 1 SOE Time History j Recorder Recorder Parameter / Signal 4 x Reactor Trip-x Safety Injection x Containment Isolation-x Turbine Trip. 4

, x control Rod Position j '

x (1) x Neutron' Flux ,- Power l

x (1) Main Steam Radiation (2) Containment (Dry Well) Radiation

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x (1) x DrywellPressure(ContainmentPressure)  !!

(2) Suppression Pool' Temperature  !

x (1) x Primary System Pressure x (1)- x Primary' System level- , j x MSIV Position '

-l 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 Vacuum -

x AC and DC System Status (Bus Voltage)

(3)(4) Safety' Injection; Flow. Pump / Valve Status x Diesel Generator Status:(On/0ff, 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) equipment status recorded on.an SOE recorder.

(4): Includes recording of. parameters for all applicable systems from .the.o following: 'HPCI,'LPCI, LPCS, IC, RCIC.

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3. Evaluation The parameters identified in part 2 of this report as'a part of the ]

review criteria are those deemed necessary to perform an adequate post-trip 1 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, including the responses of important plant systems. q The parameters identified in this submittal as being recorded by the sequence of events and time history recorders-do not correspond to the j parameters specified in part 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 seet 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 ,1 indicate that the time history equipment used would meet the intent of the i performance criteria outlined in part 2 of this report. Information supplied in the submittal does indicate that the'50E equipment meets the performance criteria specified in part 2 of this report. )'

i 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 'l readable and meaningful format. The information contained in this submittal j indicates that this criterion is met.  ;

1 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-nation 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.

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4. Conclusion The information sup; ad 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. The recorded data is output' in a readable _and meaningful format.-
2. The sequence of events recorders meet the . minimum performance

, characteristics.- t j

1 The information supplied 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 cf 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 subuittal, do not meet -

J l the minimum performance characteristics.

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3. The data retention procedures, as described in the submittal, may' not ensure that the information recorded for the post-trip review l is maintained in an accessible manner for the. life of the plant. 1 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 sh:iw in -

sore detail the data and information capabilities at this nuclear power plant. If the information provided accurately represents all current data 1 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 q to meet the intent of the evaluation' criteria.

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i REFERENCES NRC Generic Letter 83-28. ." Letter to all. licensees 'of-: operating reactors, applicants for operating license, and holders of construction 2

. permits regarding Required Actions Based on. Generic Implications of {

Sales ATWS Events." July 8, 1983. y NUREG-1000, Generic: Implications of ATWS Events at the Salem Nuclear

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Power Plant, April 1983. o Letter from J.J. Carey, Duquesne Light,'to D. G. Eisenhut! NRC,~ dated November 4,1983 in response to Generic Letter. 83-28. of July 8, .1983,-

with attachment. )

. Required Actions Based on Generic Implications of Salem ATWS! Events for-Beaver Valley Power Station, Unit 1. j Letter from J.J. Carey, Duquesne Light, to S.A. Varga, NRC,1 dated 'l October 30, 1984, Accession Number 8411060397 providing additional'  !

information to Generic Letter 83-28, with attachment. -)

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, - ; , . S A10-85 /1524 ,

N REVIEW OF LICENSEE AND APPLICANT; RESPONSES 'i TO NRC. GENERIC LETTER 83-28 (Required Actions Based on Generic ~ Implications of '

Sales ATWS Events),~ Item 1.2

" POST-TRIP REVIEW: DATA AND INFORMATION CAPABILITIES" FOR BELLEFONTE NUCLEAR PLANT ' . UNITS'1 &' 2 (50-438, 50-439) :t l

l Technical Evaluation Report Prepared by.

Science Applications International Corporation 1710 Goodridge Drive McLean, Virginia 22102~

Prepared ' for -

U.S. Nuclear Regulatory Commission Washington, D.C. 20555.

Contract No. NRC-03-82-096

L FOREWORD This report contains the technical evaluation of the Bellefonte' Nuclear.-

Plant, Units 1 & 2 response to_ Generic Letter 83-28 (Required Actions Based on Generic Implications of Salem ATWS Events), Item 1.2 " Post Trip Review:

Data and Information Capabilities."

For the purposes of this ~ evaluation, the review-criteria, presented in part 2 of this report, were divided into five separate categories. .These are:

1. The parameters monitored by the sequence of events'and the time history recorders,
2. The performance characteristics of the sequence of events '

recorders,

3. The performance characteristics-of the time history recorders,
4. The data. output format, and
5. The long-term data retention capability for post-trip review material.

All available responses to Generic Letter 83-28 were evaluated. The plant for which this report is applicable was found to have adequately responded to, and met, categories 2 and 4.

The report describes the specific methods used to determine the cate-gorization 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 presented. Instead, the evaluation presents a categorization 'of the -

responses according to which categories of review criteria are satisfied and which are not. The evaluations are based on specific criteria (Section 2) derived from the requirements as stated-in the generic letter.

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- Section Introduction. . . . . . . . . . ..i. . ...'. . . . . . ' '

.e. . E.c 1 1 Background. . . . . . . . . . . . ... .c. . . . . . ' .. : . . . . - . .: .

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2. Review Criteria ... .--. .:. . . . . . . . . . . .:. . . . . . 3:

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3. Evaluation.' . . . . ... . . . . . . . .'. . ...'. . .;. . . .-

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4. Conclusion.' . . . . . . . . . . . . . . . ... . . . .'. . ... -91 < "
5. R e fe re nc e s . . . . . . . . . . . . . . . - . . . . . . . . . . . . . .

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INTRODUCTION' SAIC has reviewed the utility's response to ' Generic 't.etter 83-28, ites 1.2 " Post-Trip Review: Data and Information Capability." The response-(see-references) contained sufficient information to. determine that the data' and information capabilities at these plants are acceptableLin the following.

areas.

I e . The sequence-of-events recorder (s) performance. charac-teristics..

e The output format of the recorded data.-

However, the data and information capabilities, as described in the submittal, either fail to meet the review criteria or provide . insufficient.  !

information to allow determination of the adequacy of: the' dataLand information capabilities in the following areas.

1 e The parameters monitored.by both. the ' sequence-of-events and time history recorders.

e The time history recorder (s) performance characteris-tics.

e The long-term data retention, record keeping, capa-bility.

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1. Background

.i On February 25, 1984, both of the scram circuit breakers at Unit 1 of I 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 startup 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 incider.t; on February 22, i 1983; at Unit 1 of the Salem Nuclear Power Plant an automatic trip signal f was generated based on steam generator low-low level during plant startup. f In this case the reactor was tripped manually by the operator almost coinci- 'l dentally with the automatic trip. At that time, because the utility did not have a requirement for the systematic evaluation of the reactor' trip, no I investigation was performed to determine whether the reactor was tripped automatically as expected or manually. The utilities' written procedures f

required only that the cause of the trio be determined and identified the i responsible personnel that could authorize a restart if the cause of the l trip is known. Following the second trip which clearly indicated the l problem with the trip breakers, the question was raised on whether the 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 after the incident. Thus, no judgment on the proper fur.ctioning of the trip system during the earlier incident could be made. '

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 'Jr.it 1 of the Salem '

Muclear Power Plant. The results of the staff's inquiry into the generic implications of the Salem Unit incidents is reported in NJREG-1000, " Generic Implications of ATWS Events at the Salem Nuclear Power Plant." Based on the results of this study, a set of required actions were developed and included in Generic Letter 83-28 which was issued on July 8,1933 and sent to all-licensees of operating reactors, applicants for operating license, and  ;

construction permit holders. The required actions in this generic letter consist of four categories. These are: (1) Post-Trip Review, (2) Equipment I

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Classification and Vender Interface, (3) Post Maintenance Testing. an::l(4) .

Reactor Trip System Reliability Improvements.

The first required action of the generic letter, Post-Trip Revien, is the subject of this TER and consists of action itam 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:

.l discussed.

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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 "I 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 rea: tor trip event should be available so that a decision on the acceptability of a reactor restart can be made.

l The following are the review criteria developed'for the requirements of Generic 1.etter 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- i vide a reliable source of the necessary information to be used in the post i trip review. Each plant variable which is necessary to determine the I l cause(s) and progression of the event (s) following a plant trip should be l

l sronitored by at least one recorder [such as a sequence-of-events recorder or i l

a plant process computer for digital parameters; and strip charts, a plant I

rocess computer or analog recorder for analog (time history)' variables].

Each device used to record an analog or digital plant variable should be cescribed in sufficient detail so that a determination can be made as to ,

whether the following performance characteristics are met:

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4 y 4 o 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 asso-ciated with each monitored safety-related. system can be ascer :

ta'ined, 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 50E' time -

discrimination is appr'oximately 100_ m sec. If current -SOEL recorders do not have this time discrimination capability the'-

licensee or applicant should show that the current time discrimi-nation capability is sufficient for an adequate reconstruction ~ of the course of the reactor trip. As a minimus this should include the ability to adequately reconstruct the accident scenarios ' pre-sented in Chapter 15 of the plant FSAR.

l e Each analog time history data recorder should have a sample inter-val small enough so that the incident can be- accurately-l reconstructed following- a reactor trip. As a minimum, the licensee or applicant should be able to reconstruct the course' of i the accident sequences evaluated in the accident analysis of the l

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.

o To support the post trip analysis 'of the cause o'f the trip. and the l 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..

e The information gathered by the sequence-of-events Land 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

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should be presented in a readable ~and meaningful format, taking l

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2 into consideration good; human factors 'practicesL(such'asithose ; '

outlined L in .NUREG-0700)i [;g 1

e . All equipment- used tot record Lsequencelof eventsiandLtimef history  ?' ,

information'should be Jpowered L from raireliable;'andi no'n-a interruptible power, source. The, power. source used;needLnot be; safety related.. ,

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w The' sequence ofievents;and. time history recording equi'pmentishould .

sonitor sufficient digitale andianalog .parameteps, .'r' respectively,L toiassureP that the course of the reactor trip can be' reconstructed.:.:The parameters? #

aonitored should provide sufficient:information to' detertiine. the? root cause '

of the ' reactor trip, the progression of the Lreactor. trip, and the response 1 -

of the plant para' meters and < systems: to :the reactorf trip.; j Specifica11y,Ua11l "

input parameters associated with reactor trips, : safety injections 7and 'other:

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safety-related systemstas well..as , outputL parameters sufficient .to re$ord the( y proper functioning of these systems should, be' recorded.forjuse in thef post' trip review.. The. parameters deemed?necessary,a 'a's :aimir.imum. to; perform .at .

post-trip review (one that would determine (if.the' plant remained!within its' , ,

design envelope) are. presented on TablesL I.2-1 and 1.2-2. ' !fLthetapplik i ..

l cants' or licensees' SOE recorders and time l history recorders doinot sonitorf '

all of the parameters suggested in.these. tables the ap;11 cant or licensee ? .

should show that' the existing set:of monitored l parameters arelsufficientIto j establish that the plant remained within the design .envelopeJfor;the appro .

L priate accident conditions; such as.those. analyzed ir.lCh' apter 15 of/the plant Safety Analysis Report.

Information gathered during the post trip. review:is required input for -

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future post trip reviews. Data from 'all unscheduled ' shutdowns. pr;ovidesl a, valuable reference source for the.' determination of the -acceptability of;the:

l riant vital parameter and equipment response to future unscheduled shute  ;

downs. It is therefore necessary tha' t information gathered.during aDl post 1 trip reviews be maintained in an accessible manner fo'rithe:11fe!cff the-l.

plant.

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5 Table 1.2-l. ' 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 t x Containment Sump Level' (1) x x Primary System Pressure-(1) x x Primary System Temperature (1) x' ,

Pressurizer Level (1) x Reactor Coolant Pump Status ,

l (1) x x Primary . System Flow.

(3) Safety Inj. t . Flow, Pump / Valve Status' ]

x MSIV Position 'j l x x . Steam Generator. Pressure-  !

(1) x x Steam Generator Level. ]

(1).x x Feedwater Flow:

1 (1) x x Steam Flow (3) Auxiliary Feedwater; System; Flow, l Pump / Val ue . Sta tus' x AC and DC System Status (Bus Voltage) x Diesel Generator Status (Start /Stop, l

On/Off) x PORV Position (1): Trip parameters (2): Parameter may be om' nitored 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.

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3 Table 1.2-2. BWR Parameter List SOE Time History Recorder gcorder Parameter / Signal' x Reactor Trip:

x Safety Injection x Containment Isolation x Turbine Trip-x Control. Rod Position x (1) x Neutron Flux.- Power x (1) Main Steam Radiation (2) Containment' (Dry Well) Radiation x (1) x Drywell Pressure (Containment Pressure).

(2) Suppression: Pool Temperature '

l x (1) x Primary System Pressure x (1) x Primary System level x MSIV Position x (1) Turbine Stop Valve / Control _ Yalve Position l x Turbine Bypass Valve Position: i x Feedwater Flow.

x Steam Flow (3) Recirculation; Flow Pump Status i x (1) Scram Discharge Leve1~

i x(1) ,

Condenser Vacuum l x AC and 'DC System Status. (Bus Voltage)

(3)(4) Safety Injection;- Flow. Pump / Valve Status x Diesel Generator Status (On/Off,-

Start /Stop) l (1):Tripparameters.

(2): Parameter may be recorded _by either an SOE or time history recorder.

(3): Acceptable recorder options are: (a) system flow recorded on an 50E-recorder, .(b) system flow recorded on a- time history recorder, or (c) equipment status recorded on an SOE recorder.

(4): Includes recording of parameters for all applicable systess' from, the .

following: HPCI. LPCI LPCS, IC, RCIC.

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3. Evaluation The parameters identified in part 2 of this report as a part of the review criteria are those deemed necessary to perform an adequate post-trip l review. The recording of these parameters on equipment that meets the q guidelines of the review criteria will result in a source of information i that can be used to determine the cause of the reactor trip and the plant response to the trip, including the responses of important plant systems.

The parameters identified in this submittal as being recorded by the sequence of events and time history recorders do not correspond to the i

parameters specified in part 2 of this report, j l

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 l 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 i supplied in the submittal does indicate that the 50E equipment meets the performance criteria specified in part 2 of this report. .

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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 j indicates that this criterion is met. l 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-sation used during a post-trip review be maintained in an accessible manner l

1 for the life of the plant. The information contained within this submittal does not indicate that this criterion will be met.

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4. Conclusion 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. The recorded data is output'in a readable and meaningful format'. ,
2. The sequence of events recorders meet the minimum performance. -'

characteristics. '

a l The information supplied in response to Generic Le:ter 83-28 does not l 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 l parameters specified in part 2 of this repert' that should be recorded for use in a post-trip review are not recorded.

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2. Time history recorders, as described in the subsittal, do not meeth the minimum performance characteristics. ,

J The data retention procedures, as described in the submittal, may:

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not ensure that the information recorded for the post-trip review; is maintained in an accessible manner for the life of'the plant.. '

-h It is possible that the current data and information capabilities at this: ,

nuclear power plant are adequate to meet the intent' of these- review . Y criteria, but were not completely described. Under these circumstances, the' licensee should provide an updated, more complete, description to'show in" ba 1

more detail the data and information capabilities'at this nuclear power L5) plant. If the information provided accurately represents- all current data;

( and information capabilities, then the. licensee. should show that the data- .j 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. :I

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' REFERENCES ' 1, .

NRC Generic Letter 83-28. ' Letter to all licens'ees of operating reactors, applicants for operating license. and holders of construction ,

permits regarding Required Actions Based on Generic Implication's of L)

Salem ATWS Events." July 8, 1983._ 4 q

NUREG-1000, Generic Implications 'of ATWS Events at the Salem Nuclear. I Power Plant,. April 1983. 1 Letter from D.S. Kammer, Tennessee Valley Aut'hority, to E. Adensas. l l NRC,.datM June 26,, 1984, Accession. Number 8406290135, in response to .

/ Generic utter 83-28 of July 8,1983, with: attachment. ', l i

L

.,6tqe,ric, Lett.e.' 83-28 Required Actions J.gfed on Generic Information of '

1 S.aled fr, JS tfents, for Bellefonte- Nuclear.flent.

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