ML20138L685

Forwards Science Applications Intl Corp Draft Technical Evaluation Rept Re Util Response to Generic Ltr 83-28,Item 1.2 on Salem ATWS Events.Response Incomplete in Five Areas. Addl Info Requested to Complete Review

ML20138L685
Person / Time
Site:	Braidwood
Issue date:	10/24/1985
From:	Youngblood B Office of Nuclear Reactor Regulation
To:	Farrar D COMMONWEALTH EDISON CO.
References
GL-83-28, NUDOCS 8510310381
Download: ML20138L685 (1)

Text

-

s Docket Nos.: STN 50-456 OCT 2 41985 and STN 50-457 Mr. Dennis L. Farrar Director of Nuclear Licensing

Commonwealth Edison Cocpany Post Office Box 767 Chicago, Illinois 60690

SUBJECT:

DRAFT TECHNICAL EVALUATION REPORT (TER) FOR SALEM ATWS ITEM 1.2 (GENERIC LETTER 83-28)

Re: Braidwood Station, Units 1 and 2 The staff has completed a preliminary review to assess the completeness and adequacy of licensee responses to Generic Letter 83-28, Item 1.2. For Braid-

, wood Station, your response was found to be incomplete in all five areas

! evaluated. The enclosed TER provides a technical evaluation representing the staff's initial judgment of the areas evaluated.

In order to preserve our present review schedule, we would appreciate your cooperation in obtaining additional information that will permit us to complete our review. It would appear that the needed information on your facility could be obtained by telephone conference within one week of your receipt of the DRAFT TER. Your project manager will be working with you to arrange an acceptable time to conduct the necessary conference call.

Sincerely, B. J. Youngblood, Chief Licensing Branch No. 1 Division of Licensing

Enclosure:

DRAFT TEP,on Salem ATWS Item 1.2 cc w/ enclosure:

See next page

. DISTRIBUTION: '

d Docket File: MRushbrook EJordan NRC PDR " G JStevens TBournia Local PDR OELD PRC System ACRS (16)

NSIC JPartlow LB#1 R/F BG es LB#1/DL LB#f/

JSteven ac BJYd g)1ood j0/f73/ / B5 8510310381 851024 PDR ADOCK 05000456 A PDR

ENCLOSURE SAIC-85/1521-1 REVIEW OF LICENSEE AND APPLICANT RESPONSES TO NRC 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 BRAIDWOOD STATION, UNITS 1 AND 2 (50-456, 50-457)

Technical Evaluation Report l 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 f r- I ); i j f_ m i t ~ ^

o y - ~r s (qt (f i N l

f

- - = - -- -p g-, w -o-- -

w ---y ,v-- -,m-w - - ,- ,

FOREWORD This report contains the technical evaluation of the Braidwood Station, Units 1 and 2 response to Generic Letter 83-28 (Required Actions Based on Generic Imp!ications 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.

For this plant no information was provided in response to item 1.2 of Generic Letter 83-28.

I 1

3 . .

TABLE OF CONTENTS Section Page Introduction. . . . . . . . . . . . . . . . . . . . . . . . . I

1. Background. . . . . . . . . . . . . . . . . . . . . . . . . . 2

2. Review Criteria . . . . . . . . . . . . . . . . . . . . . . . 3

3. Evaluation. . . . . . . . . . . . . . . . . . . . . . . . . . 8

4. References. . . . . . . . . . . . . . . . . . . . . . . . . . 9 l

l

INTRODUCTION SAIC has reviewed the material prepared in response to Generic Letter 83-28. The response (see references) failed to provide any information regarding the post trip review data and information capabilities at this plant.

t

' e i

1

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. T11s 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 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 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 procedures 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 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 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 (E00), directed the staff to investigate and report on the generic implications of these occurrences at Unit 1 of the Salem Nuclear Power Plant. The results of the staf f's inquiry into the generic implications of the Salem Unit incidents is reported in NUREG-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 8,1983 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 2

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

Reactor Trip System Reliability 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) crd progression of a reactor trip. This understanding should go beyond a simple identification of the course of the event. I,t 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 so 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 I

9--

e 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-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 sufficient 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 rea c t'o r 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.

e 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 r tape). This information should be presented in a readable and iaeaningful format, taking 4

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 should 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 ard other safety-related systems as well as output parameters sufficient to ren:~i 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 (onetthat would determine if the plant remained within its v' design envelope) are presented on Tables 1.2-1 and 1.2-2. If the appli-cants' or licensees' 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 valuable reference source for the determination of the acceptability of the 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.

l 5

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 i (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, Pump / Valve Status x MSIV 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) l x Diesel Generator Status (Start /Stop, i 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.

6 l

W Table 1.2-2. BWR Parameter List SOE Time History Recorder Recorder 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 x (1) x Primary System Pressure x (1) x Primary System level ,

x MSIV 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 Vacuum 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 l (c) equipment status recorded on an SOE recorder.

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

7

_= -

o

3. _ Evaluation i Additional information is needed before a post-trip performed. review data and information capabilities efor thn adequate evalu plant can be to action item 1.2 of Generic Letter 83-28To date, ponse little or n Any information provided by the licensee should address th criteria set forth in part 2 of this report. e evaluation how the data and information capabilities at this nuclThe information shou ear power plant ful-

fill the intent of the evaluttion criteria.

meet the intent of the evaluation criteria, the licensee shouldIf curr that the data and information capabilities are sufficient either t show o meet the intent i,

of the evaluation criteria in part 2 of this report or detail f t

! u ure modifi-cations that will enable the licensee to meet these criteria .

e a

ud i

O w

l P 44 4-* 4 l

8 z -- -  : -

i

. 'f

REFERENCES NRC Generic Letter 83-28. " Letter to all licensees of operating reactors, applicants for operating license, and holders of construction permits regarding Required Actions Based on Generic Implications of Salem ATWS Events." July 8, 1983.

NUREG-1000, Generic Implications of ATWS Events at the Salem Nuclear Power Plant, April 1983.

Letter from P.L. Barnes, Commonwealth Edison, to H.R. Denton, NRC, of November 5, 1983, Accession Number 8311090213 in response to Generic Letter 83-28 of July 8,1983, with attachment.

Part 6. Braidwood Station Response to Generic Letter 83-28.

Letter from P.L. Barnes, Commonwealth Edison, to H.R. Denton, NRC, dated June 1, 1984, Accession Number 8406050418 transmitting supple-mental response to Generic Letter No. 83-28 of July 8,1983, with attachment.

Letter from G.L. Alexander, Commonwealth Edison, to H.R. Denton, NRC, dated June 30, 1984, Accession Number 8407060235 in response to Generic Letter 83-28 of July 8,1983. Attachment not received.

l 9

. n Mr. Dennis L. Farrar Braidwood Station Commonwealth Edison Company Units 1 and 2 cc:

Mr. William Kortier Ms. Lorraine Creek Atomic Power Distribution Route 1, Box 182 Westinghouse Electric Corporation Manteno, Illinois 60950 Post Office Box 355 Pittsburgh, Pennsylvania 15230 Jane M. Whicher, Esq.

109 N. Dearborn Street Joseph Gallo, Esq. Chicago, Illinois 60602 Isham, Lincoln & Beale 1120 Connecticut Ave., N. W. Rebecca J. Lauer, Esq.

Suite 840 Isham, Lincoln & Beale Washington, D. C. 20036 Three First National Plaza Suite 5200 C. Allen Bock, Esq. Chicago, Illinois 60602 Post Offices Box 342 Urbana, Illinois 61801 Erie Jones, Director Illinois Emergency Services Thomas J. Gordon, Esq. and Disaster Agency Waaler, Evans & Gordon 110 East Adams 2503 S. Neil Springfield, Illinois 62705 )

Champaign, Illinois 61820 Ms. Bridget Little Roram Appleseed Coordinator 117 North Linden Street Essex, Illinois 60935 Mr. Edward R. Crass Nuclear Safeguards and Licensing Division Sargent & Lundy Engineers 55 East Monroe Street Chicago, Illinois 60603 U. S. Nuclear Regulatory Commission Resident Inspectors Office RR#1, Box 79 Braceville, Illinois 60407 Regional Administrator, Region III U. S. Nuclear Regulatory Commission 799 Roosevelt Road Glen Ellyn, Illinois 60137