ML20083B549

Monthly Rept for June 1990 for FIN B-5699, ISLOCA Research Program

ML20083B549
Person / Time
Site:	Davis Besse
Issue date:	07/13/1990
From:	Bickel J EG&G IDAHO, INC., IDAHO NATIONAL ENGINEERING & ENVIRONMENTAL LABORATORY
To:	Burdick G NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES)
Shared Package
ML20081D760	List: ML20083B473 ML20083B549 ML20083B658 ML20083B696 ML20083B767
References
CON-FIN-B-5699, FOIA-90-512 JHB-40-90, NUDOCS 9109250120
Download: ML20083B549 (37)

Text

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INEL Idaho her O*sel i ng Neermy inborof or e July 13, 1990 Dr. Gary R. Burdick Special Assistant Division of Safety Issue Resolution US Nuclear Regulatory Commission 56ED Nicholson Lane Rc ckville, MD 20852 JUNE MONTHLY REPORT TO NRC/RES/DSR FOR FIN B5699, ISLOCA RESEARCH PP.0 GRAM - JHB-40-90 9

Dear Dr. Burdick:

The monthly report describing work.under FIN B5699, being performed by EGLG Idaho, Inc. is attach'ed. Technical and fiscal status are included, and a status summary is provided in this letter. This report covers the accounting period beginning on May 28, 1990, and ending on June 24, 1990.

Technical Status Sumary A meeting was held with the NRC Project Manager to finalize the workscope, budget, and schedule for additional work to be included in the ISLOCA program. The following were discussed and agreed upon: additional tasks to finalize the results for the Davis-Besse plant, the workscope that relates to completion of the evaluations for Catawba, and all workscope for evaluation of an ISLOCA at the Waterford plant. The NRC Form 173 was -

received for the finalized workscope. Additional analyses needed to finalize the Davis-Besse ISLOCA results were initiated and analysis of the data obtained from the plant visit to Catawba continued. Preparations were made for the initial information gathering visit to the Waterford plant.

Fiscal Status Summary The fiscal status is summarized in the attached table.

Very truly yours, he Dr. John H. Bickel, Manager NRC Risk Analysis Unit 9109270120 91n '

' $o PDR FOIA WILLIu1?O-512 Pop - I EGcG w ~ P.O. Box 162S Idaho falls. ID 8341S

4 Dr. Gary R. Burdick July 13, 1990 JHB-40-90 Page 2 cc: DOE-ID N. S. Burrell R. S. Bostian R. M. Stallman S. L. Zeigler LGL R. J. Barrett, NRC/NRR/ DRAB F. D. Coffman, NRC/P.ES/HFB M. S. Farr, NRC/RES/DSR J. A. Murphy, NRC/RES/DSR J. A. O'Brien, NRC/RES/SSEB T. G. Ryan, NRC/RES/HFB M. W. Weston, NRC/0C/BUDB

Rf K*i DATE : 06/2C/90 l

l ' IANCE AND FROGRESS ST ATUS SUWARY fOR MONTH Of JUNE f ('G f> R A N C et ; (i ")

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Y1D t St-T N T AGNSI T3TAL P. OrO',f D $ Y -1% * * '90 f Y-19% CURR ANTICIPATED FY.1990 TY-1999 JAC  %  %

OBGL 09GL ADDil!ONAL MGL CACPYOvfR Y . f .0. O/S T.I.N. F.I.N. STAT Y . T .D .

AufM 00ttAn$ AuYM fv-1990 COSTS SPNT (N VARIANCE Cre41 T AUTH CODE BtOGE T COSTS . . ......... . ... ....

Ntr8tfR T1ftE . .. .. ...... ......... .

17,C/A.L 50.0 11,066.6 43% 43%

S48.5 51,060.6 /.4% E6.0 0 1484.1 1462.2 121.9 IIV 99 I St oC A S FOP Vaei ANCES DCIEDING 125K.

• ALI 189'S Us iM tf SS THAN 1500K ##JST HAVE EXPT AN AY10NS FOR VAal AwCis EnCEIDf NG 110K AND 189'S FOR %)pt TMAM V, doe; pa5T NAvi i W V AR! Ahtf ST Af f MENIS ARF INCLUDED IN INE TECHNICAL REPORY.

A . Uht E SS INDICAff D OTPERu1SE Att Pf RCENT AGES IN THis Cott>N ARE THE RAY!O of CottFN (2) TO Cottsw (10).

J ',d ANCI AL lbf 0G" AllCA W11HIN TMi$ REPORY MAY VARY SLIGHTLY DUE TO ROUNDING.

l B5099

!' B5699: ISLOCA Research Proaram

< EG&G Program Manager: D. J. Hanson EG&G Principal Investigator: W. J. Galyean DOE Technical Monitor: S. L. Zeigler NRC Project Manager: G. R. Burdick The objective of this project is to provide the NRC with qualitative and quantitative information on the hardware, human factors, and accident consequence issues that dominate nuclear power plant risks for Interf acing System Loss Of Coolant Accidents (ISLOCA). This information is to be used in:

Developing a PRA framework for evaluating the ISLOCA and identifying insights with respect to the rist contribution fron both hardware and human error issues along with recommendations for risk reduction.

H;ghlighting the effects of specific types of human errors and their root causes on ISLOCA risk along with rccommendations for risk reduction.

Evaluating the fragility of low pressure systems when exposed to a high pressure, high temperature reactor coolant system. This evaluation will include identification of likely failure locations and their probabilities of failure.

Identifying and describing potential ISLOCA segaences with respect to sequence timing, possible accident management strategies and effects of ISLOCAs on other equipment and systems.

istimating the consequences associated with postulated ISLOCA eve. 's, ircluding estimates of source terms and offsite consequences. Again, important issues will be identified and recommendations will be made on possible consequence reOuction actions.

Real and potential ISLOCA problems considered in this program are limited to the containment bypass variety,

1. Summary of Work Performed Durina June 1990:

A meeting was held with the NRC Project Manager to finalize the workscope, budget, and schedule for additional work to be included in the ISLOCA program. The following topics were discussed and agreed upon: (1) additional tasks to finalize the results for the Davis-Besse plant, (2) the workscope that relates to completion of the evaluations for Catawba, and (3) all workscope necessary for an evaluation of the Waterford plant. The NRC Form 173 was received for the additional workscope.

The additional analyses needed to finalize the Davis-Besse ISLOCA results were initiated. Comparisons with existing large and small break calculations to establish the timing for the ISLOCA events at i

B5699 I i Davis-Besse were begun. Information was obtained on the external event analysis for Draft NUREG-1150 to ensure that a consistent approach is used. Commitments were obtained for the assignment of manpower to begin most tasks in early July.

Analysis of the data obtained from the plant visit to Catawba continued. The preliminary event trees and fault trees developed prior to the visit were refined ant additional infor.ation was incorporated.

Preparations were made to initiate the detailed human reliability analysis.

Preparations were made for the initial visit to the Waterford plant to obtain preliminary information.

2. Summary of Work to be Perforrea Durina July 1990:

The initial estimates for timing of the ISLOCA events for Davis-Besse will be completed. Evaluation of the capability to close the valves at the high system pressures and flow rates involved in an ISLOCA at Davis-Besse will begin. An abbreviated plant visit will be arranged to collect information on external events, mitigative actions for an ISLOCA, and release pathways from the auxiliary building. Evaluation of the results obtained from this plant visit will begin.

Evaluation of the Catawba results will continue. RELAP5 models of the piping systems will be developed and calculations initiated to provide estimates of the pressure distribution in the 1 ., pressure piping systems during an ISLACA. Plant information will by transmittet to Impell so that the component and piping failure analysis can begin.

Detailed analysis of the human reliability events on the event trees will continue.

An initial visit to Waterford will be completed. Preliminary event trees will be developed to help prioritize the gathering of information during the expended plant visit. The extended plant informatier, gathering visit will begin.

3. Ergblems and Potential Problems:

None

4. Cost Breakdown:

Total cost is shown on the finance and Progress Status Summary.

5. Variance Explanation The under expenditure shown for June does not reflect the new work scope defined for the three plants. A new work breakdown structure and associated budget and schedule are being developed to incorporate the new work scope. This will be developed in July but will not be included in the computerized tracking system until mid August. As a result, the budgetary portions of the monthly reports for June and July will not provide an accurate comparison of the spending rates and the budget.

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NRC COST CATEGORIES FOR

. ; JUNE 189 NO. B5699

---

( $0.0 K )-------

CURRENT COST CATEGORIES MONTH YEAR-TO-DATE DIRECT SALARIES $ 6.0 5 116.0 MATERIALS, SERVICES AND OTHER COSTS 0.1 5.1 ADP SUPPORT 0.0 1.6

$UBCONTRACTS 0.0 70.0 TRAVEL 0.1- 26.0 INDIRECT LABOR COSTS 9.9 189.5 GENERAL AND ADMINISTRATIVE 2.5 54.0 CAPITAL EQUIPMENT 0.0 0.0 T0TAL5 5 18.4 5 462.2

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Mj PROJECT AND BUDGET PROPOSAL FOR NRC WORK W Clev,sson No 2 P'oiect Ittle

!" B5699 15LOCA RESEARCH PROGRAM '"C """"* "

Nnc otoce Off'ct Division of of Nuclear Systems Regulatory r.esearchRasp)rch (0$K (RES), 9 60 19 30 040 -

conica: tor Accour i oOr tor contrt:daho, Inc. Numeer 1 1056 EG5G I """ '

IDAHO N ATION AL ENGINEERING LABORATORY (INEL) 0 10 01 C60 ORG ANtZ ATION FTS PHONE NUMBER PERFORM ANCE COGNIZANT PERSONNEL j statnng Date N AC Proiect Manager NRC/RES 492-3509 I 08 01-89 G. R. Burdick Cococ' 04*

, cce, N Ac Tec ,n.cai sian spC/RES I 492-3965 492 3520 09-30-91 M. A. Cunningham / F. D. Coffman, Jr. NRC/RES I oCE Protect Manager S. L. Zeigler DOE-ID 583-1546 Cen*f actor Pro,ect Manager EG5G Idaho 583 9751 D. J. Hanson ctDai invest:qatortSt EG5G Idaho 583-0627 W. J. Galyean ST AFF YE ARS OF EFFORT tRou n c 10 ned*est teatn et a ve sq FY .1989 ' CY .1990 ! FY .}gg} l FY .1992 l FY .}gg3 o,reci s:.e-onc iecnn>cai r 1.0 5.4 l 2.9 I  !

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279 9Ui ONC" N 0 d '0" D* c e"'C J aN 4"Y 18 ' 0' u a ~v l Ma't" py 9 ' 90 MONTHLY FOREC AST 44 59 59 104 1 51 I 77 i

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EXPENSE August se;1emte' Af ht May June { My 64 ]} g} q __g4 7;c ic'a 5 cre:ast E voease 1 333$_ __ - oa'.e i APPROV AL AUT@RITY-SIGN ATURE . ,

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,.. . o B5699 PROJECT AND BUDGET PROPOSAL FOR NRC WORK o te August 1990 pro:ect 1,1 e ISLOCA RESEARCH PROGRAM ocE PfcpCs'*Q C gandat<0" IDAHO OPERATIONS OFFICE (ID)

FOREC AST M:LESTONE CHART: Scheduled to Start 4M,, Completed t$hown in Overter Y eari PROVioE ESTIM ATED 00LL AR COST FOR EACH T ASK FOR E ACH FISCAL YE AR FY IO#O 1" ISOO i' *1qql !h ,jgqp h 19g3 TASK :ne d ei41r 1st ir ci rci atn

$st ': ci rci a:* *sti;-:.y:iata $ siiim y:i 4'a t t s-PWR Portion of S: eoa Program Cest l -

Task 1: Config-uration Review 5:reev e ! M  ! I C:s l 53K I I i

Task 2: Operational Data Assessment s *e:- -l -

M l i C:$t 37K l l I i

4 Task 3: Engineering s:*ees e l <- -

Analysis ---

C at l l l 110K SSK Task 4: Human s:*ee e l c ,= <

i iaktors/HumanRe 1 ability Analysis c ;3, I 228K 54K  !

, Task 5: Develop s:'*: ' ti N l jIrproued Meihocology  ;

j C:s' l 66K 1

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! Task 6: Trial

Application s - e

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= l i ces: l l 261K l 72K i  !

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'.5LOCAEva$velop uation j S *eca l  ! bW k Procedures 68K l {

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i i i i Task 8: Program 5:*e: ' ! l- . = I I Management ,

st  ! 74K  ! 30K I 41K j , i i l

Task 9: Davis Besse is:"*: 'I l- =  ! '

Analysis Completion ,

c:s: ! j 150K l 10K I I i

Task 10- 15LOCA s:: '  ! "f- -

External Events AnaIysis j  ::3: i 50K SOK

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. PROJECT AND BUDGET PROPOSAL FOR NRC WORK August 1990 Pre;ect Titie ISLOCA RESEARCH PROGRAM DOE Proposing Organization IDAHO OPERATIONS OFFICE (10)

FOREC AST MILESTONE CHART. Scheduled to Start . M . Completed (Shown in Owarter Year)

PROVIDE ESTIMATED DOLLAR COST FOR E ACH T ASK FOR EACH FISC AL YE AR ry-1993 Fv-1989 -

sv 1990 fy-1991 jFv 1992 .

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Task 11: Inside senesse ZL W, (gainmentISLOCA n ysis c:st 35K Task 12: C BenefitAnayst-ysis seme as e l [M ces' 120K l l

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TOTAL EstiM ATED PROJECT Cost 915K 375K OK CK 279K

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,. ,a. e B5699 o,te PROJECT AND BUDGET PROPOSAL FOR NRC WORK August 1990 Pract Tit e ISLOCA RESEARCH PROGRAM oOE P'eces.ng Organuanon ID AHO OPERATIONS OFFICE (10)

PROJECT DESCRIPTION: IPrence earratae concri;Uons et too reawree tec cs in woocai orcer if an nem is eset accocame, bst titi6 ano so state 1 (1) OBJECTIVE OF PROPOSED WORK

a. Eackunnd Interf acing Systems Loss-Of-Coolant Accidents (15LOCAs) have been identi-fied in some Probabilistic Risk Assessments (PRAs) as major contributors to risk at Nuclear Power Plants (NPPs). They have the potential to lead to core melt and in some instances bypass of the containment which leads to early release of large quantities of fission products. Recent events at several operating plants have been identified as precursors to an ISLOCA which have raised concerns ever the frequency of occurrence, poten-tial initiators, and means of idena fying and mitigating this potential accident. This program was initiated in response to the June 7, 1989 memorandum from Thomas E. Murley to Eric S. Beckjord, " Request for Office of Nuclear Regulatory Research (RES) Support for resolution of the ISLOCA issue",

b. Qb_iective The objective of this program is to provide the Nuclear Regulatory Commis-sion 'NRC) with qualitative and cuantitative information on the hardware, l

human factors, and accident consequence issues that are significant with

{

respect to risk of an ISLOCA. This information is to be used in:

I . Developing a PRA framework for evaluating the ISLOCA and identifying i

l' insights with respect to the risk contribution from both hardware and human error issues along with recommendations for risk reduction.

. Highlighting the effects of specific types of human errors and their root causes on 15LOCA risk, along with recommendations for risk reduction.

. Evaluating the fragility of low pressure systems when exposed to high pressure, high temperature Reactor Coolant System (RCS). This evalu-ation will include identification of likely failure locations and their nrobabilities of failure.

! . Identify and describe potential ISLOCA sequences with respect to se-quence timing, possible accident management strategies and effects of ISLOCAs on other equipment and systems.

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,aa we B5699 oai, PROJECT AND BUDGET PROPOSAt. FOR NRC WORK August 1990 Pr: ect Tit e ISLOCA RESEARCH PROGRAM _

occ ore:es:n; Organaat cn ID AHO OPERATIONS OFFICE (10)

U an nem is rot a:pocat.e.

PROJECT DESCRIPTION:iprence narrat.ve cescriptions of tre ee:;stec teoics in neencat creer Itst ittle and so sta'.t )

(1) OBJECTIVE OF PROPOSED WORK (Continued)

b. Obiettive (Continued) _

• Estimating the consequences associated with postulated ISLOCA events, including estimates of source terms and off-site consecuences. Again, j

important issues will be identified and recommendations will be made on possible consequence reduction actions.

Real and potential ISLOCA problems considered in detail in this program are limited'to those that could lead to core damage and could bypass the containment. However, the work associated with this program will be per-formed with a view to possible future expansions to cover ISLOCA problems inside the containment. Preliminary information will be generated in the near term to support an Office of Nuclear Peactor Regulation (NRR) ISLOCA pilot audit program and later, more completc information will support an NRR evaluation of industry safety with respect to ISLOCA.

(2) SUMY*RY OF PRIOR EFFORTS FY-19c9:

! The following Tasks were performed during Fiscal Year (FY)-19c9. _

Task 1: Conficuration Peview Existing Light Water Reactor (LWR) PRAs have focused on the residual heat removal systems and the check valves which separate these low pressure systems

from the reactor primary coolant systems. However, the number and type of I reactor operating events (problems) which have cccurred in recent years sug- l gest not only that the check valve reliability may be less than usually as-suced (because of both mechanical and human problems), but that systems other than the residual heat removal system may be involved in an ISLOCA. It was, therefore, necessary to identify the plant specific characteristics of the systems and equipment which can be involved in an ISLOCA.

Tne purpose of is task was to identify and characterize (group) systems and ecuipe.ent that may be involved in an ISLOCA. The information available from the NRC sponenred inservice Testing (IST) program reviews of most of the i

United States (U.S.) NPPs was used. Much of the needed component information l was cbtained from the Piping and Instrumentation Diagrams (P&lDs) obtained during this program. Additional information was obtained from the plant final i

j Safety Analysis Reports (FSARs).

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=a:PopW 4e U s. NUCLE AR REGULATORY COMuisslON ** B5699

, f., e. o PROJECT AND BUDGET PROPOSAL FOR NRC WORK oais August 1990 Prosect Tit's ISLOCA RESEARCH PROGRAM oCE Proposing Organ 4ano" IDAHO OPER ATIONS OFFICE (10) w PROJECT DESCRIPTIOk Pi rovice riarrat.ve coscr$tions of tre recwoo tec cs in n mencat orcer. if an item is not acontac<e.

iai titie ano so state >

(2)

SUMMARY

OF PRIOR EFFORTS (Continued)

FY-1989: (Continued)

Task 1: Confiouration Review (Continued)

The described information was used to develop the following:

1. All plants were grouped according to Vendor and Architect Engineer (AE).

These groupings were further divided into specific plant configurations.

Fo- example: two loop Westinghouse, three loop Westinghouse, etc.

NUREG-1350 Vol. I was utilized to identify these groupings.

2. Six representative plants were selected from the vendor /AE groups and compared to other plants / vendors to verify that their interfacing systems were representative of as many plants as possible.

All Pressure Isolation Valves ((P!Vs), i.e., those valves that are nor-I 3.

mally closed and separate the reactor coolant system from low pressure and/or low temperature piping) were identified for the six plants. The list of PlVs were then broken into lists of ISLOCA valves (those valves whose failure would result in a loss of reactor coolant outside containment) and Non-lSLOCA valves (those valves whose failure would re-i sult in a loss of reactor coolant inside containment). Those valves in the Non-ISLOCA group were not considered further in these studies.

I

! 4 Based on the plant P&lDs, one line drawings of the affected systems in the representative plants were produced showing interfacing system configurations. These drawings focused only on tne relevant components, showing line sizes, class changes, relief valve location, flanges, and expansion joints. Valve sizes, types, operators, and other unique fea-tures were also included. The identified systems from the six plants, l

i four systems per plant, were then utilized for detailed review. As much as possible, the plants were selected to correspond with the plants se-lected for the NRR pilot inspection program.

5. Additional technical information (i.e. r.ot typically found on the plant P&lDs) concerning the representative systems was collected wnere sources could be found. This information included piping size, schedule, temperature, and pressure rating. For a limited number of systems the additional information included: pump ratings, heat exchanger ratings, fallef valve inlet size, and setpoints. This information was added to the simplified drawings.

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=ac scaw its U.S. NUCLE,AR REGULATORY COMMISSION F'N:

ieun 85699 PROJECT AND BUDGET PROPOSAL FOR NRC WORK cais August 1990 pre;ect Tit <e ISLOCA RESEARCH PROGRAM __

doe Preces:n; Organnation 10AHO OPERATIONS OFFICE (10)

PROJECT DESCRIPTION:(provice naframe coscriptions of tee roovired tcpics in nwme4 cal ofcer if an item is not a phCacie, hst letle anc so state )

(2)

SUMMARY

OF PRIOR EFFORTS (Continued)

FY-1989: (Continued)

Task 2: Operational Data Assessment Potential precursors to an ISLOCA have occurred and these will need to be adequately represented in the PRA analysis. Events that must be modeled in-clude both mechanical f ailures of equipment and human errors in addition to the hypothetical valve failure events currently considered in the ISLOCA analysis. To aid in ensuring that the PRA analysis is as complete as possible, a review of all the relevant historical experience, specifically Licensee Event Reports (LERs), was performed to identify the specific modes and mechanisms that have contributed to the occurrence of an ISLOCA. Both primary component (equipment) f ailures and human influenced failures were considered. The primary equipment failure modes considered included: back-leakage through a check valve as a function of time, and a check valve sticking open given there had been flow through it. The review of potential human error events included: a) errors of commission, b) maintenance errors (e.g., valves assembled incorrectly, interlocks miswired, and c) errors duri:g testing (e.g., the recent Pilgrim event). The completed data set established from these reviews was evaluated to determine the frequency of events with failures in the identified systems.

Task 8

Procram Manace-ent

' Overall management of this Form 189 was provided, including the necessary I administrative and management functions to coordinate the activities related to the ISLOCA Research Program. These included, but were not limited to, the following: (1) tracking of program costs, preparation of monthly reports and other documents that were required by the NRC Project Manager (PM) to imple-ment the program; (2) technical management of the activities of all Idaho

! National Engineering Laboratory (INEL) program participants; (3) preparation

! of a Statement of Work (50W) and initiation of a Form 189; and (4) travel to support program needs.

(3) W0cv TO BE PERFORMED AND EYPECTED RESULTS

a. Wnrk Recuired i FY-1990:

The work required for FY-1990 comprises two major efforts. First, the INEL personnel will obtain human factors and hardware configuration infor-mation from the pilot audits being performed by NRR.

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PROJECT DESCRIPTION;iprovice na' rat *ve oescriptions of tre reasites socies en n ame<>cai orcer if an item is not accocae e list title 4"O so state )

(3) WORK TO BE PERFORMED AND EXPECTED RESULTS (Continued)

a. Work Recuird (Continued)

FY-1990: (Continued)

Second, the INEL will perform an ISLOCA risk evalbation that will identify the significant contributors, in terms of hardware design and human fac-tors issues, to the risk posed by the possibility of an ISLOCA. Prelimi-nary information will be shared with the NRR pilot audit teams by review and comment on the audit guidance packages, delivered through the NRC PM.

These potentially significant contributors will be developed through the analysis of six specific NPPs and subsequently utilized to compile a set of potential inspection guidelines that can be used to evaluate the risk for additional plants. An additional product will result from the plant specific evaluations in the form of a library of interfacing system models, which could be used to quantify the ISLOCA core melt frequency for the remaining plants.

Although the individual tasks are planned to be performed for a total of six plants, the first plant evaluated will be used as a test case. The experience gained during the review of this plant will be used to deter-mine the extent of data collection required during plant visits for the remaining plants. In addition, upon completion of the analysis of the third plant, the results will be reviewed and a decision will be made as to whether analysis of the remainder of the plants will be performed. The information or costs, schedule, and deliverables are applicable to the analysis necessary for these three plants which include Davis Besse, Catawba, and Waterford. For these .hree plants the following tasks will be performed unless otherwise noted.

Task 3: Enaineerina Analysis i This task will identify the most likely failure locations, median failure pressures (and associated untertainties), and leak areas based on best-estimate realistic analysis. This information will be directly uti-lized in the PRA quantifications of pipe and component failure probabil-ities as a function of pressure. From Task I the component types (pipes, flanges, valves, pumps, heat exchangers, etc.) that could potentially f ail will be selected for the representative systems of plants analyzed. Best es*.imate failure and leak models will be developed for each component type using a consistent methodology. A consistent methodology should be used to assure the calculated relative pressure retaining capacities of the com-ponents reflect their actual relative strengths. If different methods are page 8 gc(eg +

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PROJECT DESCRIPTION 'P'ooce na"atue oescriptions of tme rewec icocs in n urneticat croer it an item is not accocatre list lit:e and so state )

(3) woDK TO BE PERFORMED AND EXPECTE U ESULTS (Continued)

a. Work Reouired (Continued)

FY-1990: (Continued)

Task 3: Enaineerina Analysis (Continued) used for the various components, the calculated relative pressure retaining capacities may reflect differences in the analytical techniques rather than the actual relative strengths.

The method expected to be used is limit analysis; however, previous relative analyses by Brookhaven National Laboratory (BNL), Industry Degraded Core Rulemaking Program (IDCORP) and others will be considered in selecting the method and the results from these analyses will be utilized where appropriate.

The limit analysis method assumes the material is elastic until the yield l

stress is reached and then becomes perfectly plastic so that the struc-I tural elements deform elastically until the limit stress is reached then

- deform without increased load carrying capacity so that additional load is

! supported by the meatsrs not yet at the yield stress. When the load is j sufficient so that all members in the failure path are at the yield stress j l the structure is considered to have failed. The yield stress at temper- l .

i ature used in the analysis will be adjusted to account for strain hard- i ening where appropriate, sucn as for stainless steel components. The

, technique also provides a method of selecting the probable failure path and is therefore useful in developing the expected failure loads for con-tinuum structu us such as pipes, flange rings, etc. l l

Analysis of failures of structures with large elastic follow-up, such as bending failures of short pipe sectiorg connected to long unrestrained pipe l-sections, will include stresses from thermal expansion caused by the hot RCS coolant entering the cold low pressure system. In some cases the f i relative dimensions of the parts will influence the development of the I models. This information is expected to be available from Task 1 and detailed component drawings. In developing the models potential unde-tected flaws such as cracks that have not propagated through the wall will be considered and incorporated where appropriate. The actual method of estimating flaw size will be developed as part of this work but is ex.

pected to be based on the minimum flaw size detectable by inservice inspec-tions (ISI).

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a. Work Reauired (Continued)

FY 1990: (Continued)

Task 3: Enoineerina Analysis (Continued)

Some of the models will not be in the form of classic limit load equations even though developed from limit load principles. For example, the model for a bolted flange is envisioned as a model with the bolts elongating while below their limit stress reducing the support for the gasket until the combination of reduced support and increased press.re causes the gasket to reach its limit stress and extrude opening a leak path. The limit load would be calculated as the pressure at which the gasket ex-trudes. The leak area would be estimated based on the resultant flow path through the space where the gasket was. The model would include the method for calculating the increased leak area as pressure increases beyond that which would initiate a leak. The models for all the com-

! ponents would be similar in that they would identify the failure pressure and would provide the method for estimating the resultant leak area.

The component information of dimension, wall thickness, bolt size, etc.,

i from Task 1 and detailed component drawings along with material property I

information will be used to develop numerical values for the terms of the ,

j models. The effects of aging such as reduced wall thickness from corrosion, increased crack size from fatigue, changes in material proper-ties from thermal embrittlement, etc., will be considered in developing i the numerical values. The actual methods for estimating these effects ,

i will be developed as part of this task with input from the NRC aging research.

i Best estimates of the pressures that would initiate the leak and the leak sizes at the maximum expected pressure for the system would be calculated for each component. The calculated results would be compared to known behavior where available from test results or from actual plant experi-ences and the models adjusted if necessary to assure realistic results.

Most pressurizations of the interfacit.g systems are expected to be rela-tively slow. They are exoected to resbit from leaking valves or misposi-tioning of the valves. Leaks, of course, result in low flow rates which may, in most cases, be accommodated by the thermal relief valves. Hispost-tioning (opening) of valves will take several seconds and therefore will also result in a relatively slow pressure rise except perhaps for a par-tially filled system.

Sudden failure of a valve,.although unlikely, could j result in a rapid pressure rise. Also for a system with an initial steam

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(3) WORK TO BE PERFORMED AND EXPECTED RESULTS (Continued) a.

Work Recuired (Continued)

FY-1990: (Continued)

Task 3: Enaineerino Analysis (Continued) pocket which would collapse when pressurized may result in a dynamic loading.

the rate ofAny gas products pressure rise. from release of dissolved gas would mitigate Appropriate dynamic load f actors will be devel-oped appropriate. based on previous analysis of similar systems and will be used where A family of pressure dependent fragility curves representing the statis-tical distributions of the terms in the failure and leak models (part dimensions, material properties, etc.) will be determined from consider-ation of the manufacturing processes and material property test (these distributors reports). are expected to be available from standard tests and Estimates of the distribution of the effects of the aging mecha-nism (wall thinning from corrosion, crack growth from fatigue, change of materid properties from thermal embrittlement, etc.) will be included as model uncertainties.

The statistical distributions of the terms will be combined by the relationships defined by the model to provide a statis-tical of pressure. distribution of the failure pressures and leak areas as a function These distributions will serve as input to the thermal hy- ,

draulic analysis and will be combined with the flow through the thermal -l relief valves to provide the leak rates for the plant analysis. The com-bined of Taskresult 6. will provide the information necessary for the PRA analysis Task 4:

Human Factors / Human Feliability Analm jl The primary objective of Task 4 is to support development of evaluation guidelines, and identification of recommendations for remedial actions (e.g., training, procedures, management oversight).

Intermediate objectives of Task 4 are: (1) to develop a taxonomy of human errors (omission, commission) and their related causal (contributing) factors involved in ISLOCA; and (2) to support development and testing of a methodology for fully integrating human, hardware, engineering, and design ISLOCA.

considerations in the reliability and risk assessment process for

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August 1990 Pro,ect Titie ISLOCA RESEARCH PROGRAM DOE Procesing O';amzabon PROJECT DESCRIPTION: torevice marrauve cesen bons of tae reosec topics im Nmencat orce H aa ne s not a:poca e e., noe ano so staie i (3) WORK TO BE PERFORMED AND ErPECTED RESULTS (Continued)

a. Work Reouired (Continued)

FY-1990: (Continued)

Task 4: Human Factors / Human Reliability Analysis (Continued)

To accomplish primary and interw diate objective: of this task, three interrelated subtasks will be curied out. These are:

1. A retrospective analysis of existing information about ISLOCA events in the area of human behavior, specifically to determine factors which can cause and or contribute to the ISLOCA event.

2. Field data collection to su.1 port all other Task 4 activities, and other tasks of this form 189,

3. Human Reliability Analysis (HRA), as required within the context of the ISLOCA quantification methodology (Task 6).

l Note: Portions of this task involve a novel application of HRA to provide input to the PRA portion of this program. Due to it's unique nature. the  ;

HRA approach describec herein will be evaluated upon completion of the j Davis-Besse analysis (Davis-Besse being the first plant to be analyzed  ;

under Task 6). The evaluation criteria could include, but not be limited j to: repeatability (objectivity), measurability, applicability, j defensibility, utility, cost, and acceptability. Following evaluation, a  !

decision will be made by RES program and line management on how to proceed with human error quantification on the remaining plant (s).

Subtask 4,1: Retroscettive Analysis l

The purpose of this subtask is to conduct a review of existing information and data concerning human involvement in ISLOCA, and develop a taxonomy of human errors of omission and commission and their measurable causal (i.e.

contributing) factors for input to the methodology development (Task 6) i and quantification (Task 7).

1 l

Primary information for conducting this subtask will be obtained:

1 (1) from a search of the Sequence Coding and Search System. (2) from an initial review of ISLOCA involved LERs by Office for Analysis and Evalu-ation of Operational Data ( AE00) reported in the memorandum, " Requested List of LERs Possibly Relevant to Interf acing Systems ISLOCA Entry," from Thomasson (AE00) to Diab (NRR), dated May 12, 1989, (3) from published l PRAs, and (4) from Generic issue (GI) 105 documentation prepared by BNL.

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(3) WORK TO BE PERFORMED AND EYPECTED RESULTS (Continued)

a. Work Recuirqd (Continued)

FY-1990: (Continued)

Task 4: Human Factors / Human Reliability Analysis (Continued)

Suotask 4.1: Getrosrettive Analysis (Continued)

These sources will be supplemented by plant-specific information and data collected under Subtask 4.2, and eventually by quantitative and quali-tative results obtained from Subtask 4.3 below.

The taxonomy will be presented as a matrix of human actions and causal (contributing) factors which attempts to account for their interrelation-ships, This subtask will provide input for the analysis of the individual plants.

! This subtask will be performed coincident with Tasks 1 through 7, and should involve excnanges of information with Subtasks 4.2 and 4.3 of Task 4.

The product of this subtask will be input to the methodology development l l

i and cuantification (Tasks 5 and 6) and the taxonomy of human actions and i i causal f actors and their interrelatecness. This will include a written explanation of the taxonomy, Subtask 4.2: Field Data Collection f The purpose of this subtask is to gather plant-specific human factors l

information and data to support other subtasks under Task 4, and thereby, provide insights regarding human factors issues for developing ISLOCA evaluation guidelines (Task 7).

This subtask will involve visits to participating plants selected as part of Task 1. It is assumed that three plants will be visited in conjunction '

i with the NRR audit program and three plants will be visited as part of the l research related tasks.

It is assumed that data for the total program will be collected in two 1

pnases. The first phase will be a short plant visit where the list of j potential human failures will be expanded using questionnaires and

walk-through/ talk-throughs (where possible) to identify other mechanisms of failure. The second phase will be a detailed data collection effort using the following process

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(3) WORK TO BE PERFORMED AND EXPECTED RESULTS (Continued)

a. Work Reauired (Continued)

FY 1990: (Continued)

Task 4: Human Factors /Hu*an Reliabilitv Analysis (Continued)

Subtask 4.2: Field Data Collection (Continued)

1. Obtain a list of potential human failures from Task 6. This list will represent the integration of all the prior tasks including the prior short plant visit.

2. Produce a task analysis form incorporating the list of required infor-mation for input to the following HRA Models: THERP, HCR, NUCLARR, MAPPS, or SLIM MAUD. A determination will be made based upon the failure being investigated as to which information categories will be collected at the plant visits.

l

3. A list of additional queries will be developed to tap other mechanisms of failure and other nontraditional performanct shaping factors (i.e.

safety culture). These queries will be made of the personnel involved in the ISLOCA tasks.

[ 4. Collect and review relevant plant cocumentation (procedures, PRA, incidents reports etc.).

5. Conduct plant personnel interviews and walk-throughs to support task analysis.

6. Perform observations at the plant.

For the purposes of developing a concrete cost and schedule estimate for the di a collection effort it is assumed that this analysis will encompass 20 huma; failures per plant (five f ailures for each of four systems). An example of a human failure is "f ailure to perform logic test properly such that the pressure boundary is violated". It is assumed that each failure i will be considered in terms of one context. This effort includes one week

! of preparation and a maximum of two weeks data gathering at the plant for each of two analysts.

The products of this subtask will include:

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Work Reevired (Continued)

FY-1990: (Continued) i Task 4: Human Factors / Human Reliability Analysis (Continued) l Subtask 4.2: Field Data Collection (Continued) 1.

Lists of potential human failure mechanisms, completed task analysis forms, and results of interviews with operations personnel for input to the quantification task (Task 4.3).

g 2. Series of insights to guide revision of the emerging NRR evaluation guidelines and written input to the NRR evaluation reports.

This subtask should be performed coincident with and be guided by inputs from Task 1 (Configuration Review), Task 2 (Operational Data Assessment),

Task 3 (Engineering Analysis), and Task 5 (Improved ISLOCA Quantification Methodology). Subtask products should feed into other subtasks under Task 4, Task 5, and Task 6.

Subtask 4.3: Human Reliability Analysis The purpose of this subtask i'; to conduct HRA of pertinent human task actions initiating, otherwise contributing to, or imoortant to recovery of l the error. That is, this subtask will consider human task actions which are a part of the task sequences as identified in the Task 6, Trial Application. These results will be used, when incorporated into the risk assessment, to help improve guidelines for evaluating licensee policies, practices and procedures relevant to ISLOCAs.

Primary sources of information for completing this subtask will come from Subtask 4.1 and 4.2 of Task 4, and from Task 1 (Configuration Review),

Task 2 (Operational Data Assessment), Task 3 (Engineering Analysis),

Task 5 (Improved ISLOCA Methodology), and Task 6 (Trial Application of Methodology).

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(3) WORK TO BE PERFORMED AND EXPECTED RESULTS (Continued)

a. Work Recuired (Continued)

FY-1990: (Continued)

Task 4: Human Factors / Human Reliability Analysis (Continued)

Subtask 4.3: Human Reliability Analysis (Continued)

It is anticipated that this subtask should involve human error quantifi-cation tools (such as the SLIM-MAUD, HCR, THERP, and MAPPS computer codes), and generic data sources [such as the Nuclear Computerized Library for Assessing Reactor Risk (NUCLARR) data management system], for the analyses. It will also involve the application of current documentation on the TALENT HRA/PRA integrating concept being prepared by Lawrence Liver-more National Laboratory (LLNL), where appropriate, to insure the proper level of HRA and PRA integration.

There should be three products emerging from this subtask:

1. Quantitative estimates and qualitative insights estimating the contri-bution of human task actions to ISLOCA-centered reliability and risk j (support Task 6).

2. A set of recommendations for evaluation procedures and remedial actions based on HRA quantitative and qualitative results (support Task 7).

3. Procedures for improving the integration of HRA into PRA for ISLOCA (support Tasks 5, 6 and 7).

Task 5: Develoo Irrreved Methodology The objective of this task is to develop a state of-the-art PRA analysis method for quantifying the frequency of an ISLOCA and for performing sensi-tivity calculations aimed at identifying effective ISLOCA prevention, recovery, and mitigation actions.

Subtask 5.1: Develoo Format for Analyzina Risk Task 5 will begin with a detailed literature search (a preliminary search was performed during the preparation of this Form-189). Using input from the information obtained from the literature search, an initial format for pa;e 16

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Work Recuir?d (Continued)

FY-1990: (Continued)

Task 5: DeveloD froroved Methodolocy (Continued)

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Suttask 5.1: DeveloD For at for Analyzina Risk (Continued) analyzing ISLOCA risk will be developed. Subsequently, as information frem the other tasks of this effort Decome available, it will be incorpo-rated into the ISLOCA Analysis Method format. (Note: in order to be as responsive as possible to the timing requirements of NRR's inspections, Tasks 1, 2, 3, and 4 will progress in parallel. The final form of the methodology and inspection guidelines, which will incorporate all of the resu'its and insights of the other tasks, will be available upon the comple-tion of Task 6).

Subtask 5.2: Event Tree Develor ent in this initial effort, reDresentative high/ low pressure interfaces will be modeled using a component-level, multi-branched event tree to identify those combinations of human errors and equipment faults that result in a possible accident sequence. The event trees will be expanded to reflect j the effects of mitigating system operation and to incorporate important  ;

time-dependent events, dependent events (i.e., common cause and vital area '

ever.ts), and mutually exclusive events (such as: preexisting conditions, fast / slow equipment failure mechanisms, test and maintenance errors, recov-ery actions, etc.) in the model. Each interface would require an event tree for each reactor operating mode that will be included in the project scope.

Subtask 5.3: Fault Tree Develoc-ent in this task, the individual event tree headings will be expanded to ac-count for different fault mechanisms. These models will most likely be in the form of simple fault trees.

Therefore, Task 5 will function as a central collection point for the information generated in all other tasks. The important plant attributes, cethod format and insights list will be updated during the performance of Task 6 with a final update in Task 1. The product of this task will re-flect the integration of information generated in Tasks 1 through 'i and xm n e.. u v p,;e 17

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(3) WORK TO BE PERFORMED AND EXPECTED RESULTS (Continued)

a. Work Recuired (Continued)

FY-1990: (Continued)

Task 5: Devaloo Iroroved Methodology (Continued) i Subtask 5.3: Fault Tree Develcoment (Continued) will include a number of generic ISLOCA models .'nd a list of plant spe-cific informational requirements (both hardware and human factors related) for evaluating 15LOCA risk.

Task 6: Trial Aeolication The objective of this task is to apply the methodology developed in the i previous task in full scale applications on the plants chosen in Task 1.

An evaluation of the HRA methodology will be conducted following comple-tion of the first plant and a method selected for the remaining plants.

i Following completion of the evaluation of the third plant, a review will

! be performed and a decision will be made as to whether analysis of addi-tional plants will be performed. Execution of this task will require

detailed HRA using plant specific procedures, plant specific thermal hy-

draulic analysis, and plant specific component capacity analysis developed in the previous tasks.

Subtask 6.1: Thermal-Hydraulic Evaluation 1

! It is assumed that Thermal / Hydraulic (T/H) calculations are available or l' can be extrapolated from existing calculations and that only simple or minimal additional calculations will be required in this project.

, Specifically, given the distribution of possiDie break sizes produced in i Task 3, simple T/H calculations will be performed to estimate the leak rates. These leak rates will then be compared to calculations performed.

for past analyses to estimate the time to core uncovering, core damage and time available for recovery actions.

Subtask 6.2: Consecuence Analysis As a part of this task, the plant specific consequences portion of the j risk calculations, will be performed using either MACCS or CRAC2

' (preferably MACCS) and a plant specific input deck. The basic approach will be to develop a single site information deck, estimate plant specific l

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a. Work Recuired (Continued)

FY-1990: (Continued)

Task 6: Trial Aeolication (Continued)

Subtask 6.2: Consecuence Analysis (Continued) source term from existing literature, and run the MACCS Code. The re-quired source term inputs for the pertinent ISLOCA sequences will be esti-mated based on existing literature. This is a reasonable approximation since by definition, these releases bypass containment and therefore are unaffected by variations in ccatainment system designs.

Subtask 6.3: ISLOCA Ouantification As the models are bG ng developed for each plant, particular attention will be given to the identification of opportunities for errors of commission. Preliminary model development and identification of such opportunities will be done prior to plant visits to help focus activities i under Subtask 4.2, Field Data Collection _ ~

This task will calculate ISLOCA sequence frequencies for the represen-tative sample of plants. Also, as part of this task, the plant specific l

! analyses will be extended in an evaluation of various mitigative strategies. This will be done through the mechanism of sensitivity calcu-lations on the plant specific models. Any promising mitigative strategies will be incorporated into the final procedures document.

Task 8: Procram Manacement Overall management of this Form 189 will be provided, including the neces-sary administrative and management functions to coordinate the activities related to the ISLOCA Research Program. These include, but are not lim-ited to, the following: (1) tracking of program costs, preparation of i

monthly reports and other documents that are required by the NRC PM to

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implement the program; (2) technical management of the activities of all INEL program participants; (3) review of all materials prepared for the

- program; (4) travel to support program needs; and (5) participation in and

! organization of technical seminars and services as requested by the NRC PM.

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! Subtask 9.4: Sensitivity Evaluation

! For comparison purposes, determine Davis-Besse ISLOCA sequence sensitivity to typical PRA values.

I Subtask 9.5: Evaluation of Decontamination f actors Determine decontamination factors for the Davis-Besse ISLOCA sequences.

! Evaluate the effects of such factors on risk estimates.

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a. Work Reouired (Continued)

Fi-1990: (Continued)

Task 9: Davis-Besse Analysis Comoletion (Continued)

Subtask 9.6: HRA Analysis Review Engage a human reliability expert as a consultant on the HRA approach on Davis-Besse. The consultant should critique the approach with respect to: repeatability (objectivity), measurability, traceability, applica-bility (e.g., to errors of commission), defensibility, utility, cost (compared to alternative equally accurate approaches), and acceptability.

Subtask 9,7: Documentation Revision incorporate the results of Subtasks 9.1 through 9.5 into the Davis-Besse report. Prepare a presentation on how the report differs from the orig-inal draf t. The presentation is to include a discussion of reviewers' comments to date and how each comment has affected the revised report.

( Group comments by similarity to the extent possible. Anticipate an Advi- -

sory Committee on Reactor Safeguards (ACRS) review of the revised report prior to finalization.

i Subtask 9.8: IMPELL NURES Provide to the NRC PM a camera-ready copy of the IMPELL Davis-Besse engi-neering analyses for publication as a NUREG/CR report.

Task 10: ISLOC A External Events Analysis Perform a simplified External Events Analysis (EEA) on Davis-Besse or Catawba similar to that performed for NUREG-ll50 ISLOCA sequences under the leadership of Sandia National Laboratories (SNL). Obtain pertinent references from the SNL NUREG-ll50 EEA PM. In conducting the EEA, utilize as much existing information as possible including documents produced under the Seismic Safety Margins Research Program (SSMRP) (also contact SNL).

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August 1990 Pf 0!OCl Tit:0

' 0CA RESEARCH PROGRAM cot preposing organaation IDAHO OPERATIONS OFFICE (10)

PROJECT oEsCRIPTION <Procce ea"atne cescnctions of the reowred tepics in %mencat orcer a am nem is not a:pocacie, 0111114 493 50 5t410 )

(3) WORK TO BE PERFORMED AND EXPECTED RESULTS (Continued)

a. Work Peauired (Continued)

FY-1991:

Tasks 3, 4, 6, 8, 9, and 10 for PWRs will be completed in FY-1991 as the analyses of the Catawba and Waterford Plants (Plants 2 and 3) are com-pleted. The following additional tasks relating to the PWR analyses will be initiated and completed in FY-1991.

Task 7: Develoo ISLOC A Evaluation Procedures The integrated list of critical plant characteristics and operational issues which dominates the risk of ISLOCA will be updated a final time.

Thase critical plant design characteristics and operational issues, along with the information and insights from the pilot audits, will then be used to produce a final ISLOCA Evaluation Procedures Document guidelines that were originally draf ted in Task 5 and upyated in Task 6.

I 1 This document will identify such things as combinations of poor design character 4 tics, surveillance procedures, control room indication, oper-ating procedures, emergency procedures, and feasibility of potential recov-ery actions, which significantly influence ISLOCA frequency and risk.

I included will be a checklist of practices or other items useful in main-taining low risk from ISLOCA.

l Inside Containment ISLOCA Analysis

! Task 11:

For one Pressurized-Water Reactor (PWR) (Davis-Besse or Catawba), review and screen by pipe size all systems inside containment with ISLOCA potential. Of those not screened out, failure scenarios will be postu-lated and quantified. Given an overpressure failure of the interfacing system, probabilities and likely locations will be determined and the likelihood of water draining into the containment sump will be assessed.

Scenarios that include water reaching the containment sump ans' are within the plant design basis will not be examined further. Specifically, multi-pie independent failures of the Emergency Core Cooling System (ECCS) will not be included. However, inside containment ISLOCA events which are coupled in a common-cause or cascade manner to ECCS failures will be analyzed.

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' ProieCl let'e 15LOCA RESEARCH PROGRAM OCE Preocim; O';anaation ID AHO OPER ATIONS OFFICE (10)

PhoJECT DE s '.R:PfloN:(Pro, ice estrative ce stions et tee rewres tcp:cs in neericai orcer. It am nem is eot acpiicatie.

i,it niie eno no staie i (3) kOPK TO BE PERF0DPED AND EXPECTED RESULTS (Continued)

a. Wrk Recuired (Continued)

FY 1991: (Continued)

Task 12: Cost Benefit Analysis in support of rem iatory analysis requirra for resolution of GI 105, a study will be A cted to identify and to assess the cost of ISLOCA risk reduc' alter dives. Work on this Lask will be in accord with the contet.-t f NURl /CR 005B, " Regulatory Analysis Guidelines of the U.S.

Nuclea r,gulatory Comn.ission," Rev.1,1984, and NUREG/CR 3568, "A Handbook for Value impact Assessment," PNL, 0: Ember 1983. Other useful docur.1ents might be NUREG/CR 3971 and NUREG/CR 4568. Work under this task will proceed on one PWR only until the overall Br.lling Water Reactor (dWR) portion of the program is formulated. This work will consist of the followi g subtasks:

lyftask 12.1: Identifv ISLOCA Chances 11tntify potential changes (hardware and nonhardware) from ISLOCA research. These changes will include. consideration of, but not be limited to, the five items listed on page 17 of the Task Action Plan for GI 105,

" Interfacing Systems LOCA in Light Water Reactors," Rev 4, February 1990.

Subtask 12.2: Collection of Data on Costs follect data and other information as required to assess the costs associ-

+ted with the changes identified in Subtask 12.1 Included in this suotask ut visits to one each PWR and BWR to realistically assess all aspects dfecting costs nf changes.

Subtask 12.3: Cost Benefit Calculations Calculate costs and benefits from information emanating from Subtasks 12.1 and 12.2. Include sensitivity analyses. Include as benefits, costs, and

'a savings from averted precursors (exposures and cleanup).

Subtask 12.4: Letter Recort Document the results of Task 12 in : letter report.

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.. e.e2 B%99 o,i, PAOJECT AND BUDGET PROPOSA1. FOR NRC WORK August 1990 Pecioci t te

!$LOCA RESEARCH PROGRAM oct Preposing organint oa IDAHO OPER ATIONS OFFICE (10)

Paostet otscniptioN.tPro, ice canative cowerms of Ime roowdec tcpics en n umental c'cer H en Hem is not appcape, test titre 460 to state 1 (3) V0PK TO BE PERf0PFED AND EXPECTED RESE IS (Continued)

a. Work Recuired (Continued)

FY 1991: (Continued)

Finalization of the BWR tasks to be performed are also expected to be com-pleted by the end of the first quarter fY.1991, contingent on identifi-cation of scope by the middle of the fourth quarter of FY 1990. Final documents will be produced on the PWR analyses and the 15LOCA Evaluation Procedures.

b. Meetinas and Travel FY-1990:

Upon approval of the NRC PM, one initial visit will be made for each of the plants with two or three people participating. One data collection visit will also be made for each of the plants with & total of three people participating on each visit. The locations to be visited will depend on the plants selected which will be a joint decision of NRR and project personnel.

Upon approval of the NRC PM, one trip for three persons to Rockville, Maryland (MD) to discuss the results of the initial plant visit and two trips fnr one person to Rockville, MD to coordinate the program activities will be required at a later time.

Upon approwsi of the NRC PM, three trips will be made to Mission Viejo, California (CA) to monitor and discuss results from the subcontract with IMPELL Corporation for Task 3.

One visit to a PWR by two people l' *equired to obtain hardware and proce-dure changes cost and precursor t.seanup cost information.

FY 1991:

One visit to a BWR by two peopl is required to obtain hardware and proce-dure changes cost and precursor cleanup cost information. Six trips will be required for program review and program management discussions. Travel to Water Reactor Safety Information Meeting (WRSIM) is authorized for one IMPELL Corporation employee for presentation of a paper on the engineering analysis work done on Davis Besse.

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not ime ans so siste i (3) WOPK TO BE PERFORMED AND EXPECTED RESULTS (Continued)

c. NRC Furnished Materials A list of lERs possibly relevant to ISLOCA (Neil Thomasson memorandum to Samy Diab, May 12, 1989) has been provided.

Draft NRR planning material on overall 1SLOCA program and ISLOCA pilot inspection programs have been provided.

For each plant undergoing detailed analysis, the following information will be re';uired:

. Updated FSAR.

. Plant Technical Specifications (15s).

. Plant elevation and layout drawings for interfacing systems.

. Sample PlV maintenance procedures.

. Test and surveillance procedures for P!Vs. '

. Emergency operation procedures for an ISLOCA event.

. Interfacing system descriptions (e.g., from training manuals).

The NRC will provide for access, by contractor personnel, to the plants selected for the detailed analyses. The contractor will be identifying the plants to be analyzed by September 29,19?9 to ensure no programatic delay.

NRR's ISLOCA Team Audit inspection plans as they become available.

(4) DESCRIPTION OF ANY FOLLOW ON EFFORTji FY-1992. FY-1993. and Beyond:

(

! No follow on efforts are anticipated beyond FY-1991.

l (5) AELATIONSHIP TO OTHER PP0]ECTS i

This work is closely related to that performed by BNL on GI 105 (funding iden-l tification Number (FIN): A3829) and the Argonne National Laboratory (ANL) technical assistance being provided to NRR under FIN: L1231. This work will l

be coordinated with Task I of flN: 1.6854, as modified 7/17/39.

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Paoatet or scRIPTION- (Pf evice narrelive cescriptions of t*e reawi<ec icpics in evrnerical orcer. If an Hem es not appocacio, osi noe ano so siais i (6) REPORTING REOUIREMENTS AND SCHEDUL[

a. Technical ReDorts FY 1990:

A letter report documenting the results of the review of the initial or pilot plant will be completed in February 1990.

A draft NUREG/CR report on Davis Besse will be provided to the NRC by the end of September 1990.

A draft NUREG of the IMPELL Davis-Besse report will be provided to the NRC by the end of September 1990.

A paper describing the engineering analysis work on Davis-Besse will be prepared by IMPELL Corporation for pretentattun at the WRSIM (cf. Subtask 9.8).

FV-1991:

Draft NUREG/CR reports on Catawba and Waterford will be completed by the end of November and by mid March, respectively. The Waterford draft NUREG will include the developed ISLOCA Evaluation Procedures. Letter reports will be provided on the EEA. the cost benefit analysis, and the results of the 15LOCA inside containment analysis. Documentation of the BWR portion of the program is anticipated in FY-1991.

b. Monthly letter Status Rerort EG&G Idaho, Inc. will prepare a monthly business letter status report in accordance with the current NRC Manual Chapter 1102.

c. Publication of Research Pesults The INEL and the NRC FM will discuss whether the information resulting from this research warrants publication in refereed scientific and eng-ineering journals. Such rublication will focus on advances in science and technology, and minitaize conclusions and/or recom.mendations that may have The NRC PM will make the final determination regulatory whether theimplications.

publication of research results may be pursued as an allow-able cost under this project and will notify the laboratory of the de-cision. If the decision is made to publish, the laboratory will actively pursue the publication in accordance with the provisions of NRC Manual Chapter 3206. The notification wi'l be provided to the laboratory in p,;, pg o : w .w ...

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+*e 6 6a PROJECT AND BUDGET PROPOSAL FOR NRC WORK ooie August 1990 Prosect 1 st's l

ISLOCA RESEARCH PROGRAM DCE Pecocsing Organitation IDAHO OPERATIONS OFFICE (10)

PROJECT DEsCRIPfl0N: IProvice narrative coscriptions et tr's reawirec ico ca in numerical creer.11 an item is not applicable, not ime ano so siete >

(6) REPORTING FEOUIREMENTS AND SCHEDULE (Continued)

c. Publication of Research Pesults (Continued) writing and the laboratory will not incur costs associated with such publication until after receipt by the laboratory of notification that publication should be pursued. The laboratory will coordinate all such puolications with, and transmit a copy of the proposed article to, the NRC PM prior to publication.

(7) SUECONTRACTOR INFORMATION A subcontract has been given, through competitive bid process, to IMPELL Corporation to perform the engineering analysis described in Task 3. The cost of this effort is $120K for IMPELL and $11K for material handling charges. The subcontract will not result in a real or apparent conflict of interest situation. If the NRC orogram office requires additional specific subcontractor information or limitations, those requirements shall be stated on the NRC Form 173.

(8) L]ST NEW CAPITAL E0VIPMENT RE0VIDED FY-1990/FY 1991:

Not applicable.

(9) DESCRISE SPECIAL FACILITIES PE0VIPED Not applicable.

(10) CONFLICT OF INTEREST INFORMATION All current and previous work conducted by EC&G Idaho, Inc., and where applicable, subcontractor personnel assigned to this project, .1 ave been re-viewed and no known real or apparent conflict of interest exists.

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,n. *u B5699 PROJECT AND BUDOET PROPOSAL FOR NRC WORK o,ie August 1990 Pro,ecs t.ve ISLOCA RESEARCH PROGRAM Oct Ptetosing Organdauon IDAHO OPERATIONS OFFICE (10)

PROJf 07 ot$CRIPTION. (Provice nattative descriphons of the required top'C8 in numte: Cal orcer,il an item se not applicab'e, fill title and to State 1 (11) EXPECTED CLASilflCATION OR SENSITIVITY If proprietary or company confidential data is provided to the INEL by the government in connection with this contract, the INEL agrees to safeguard such information in accordance with 10 CFR 2.790 and agrees not to release such information to any person not directly involved in the performance of work under this contract unless such release is authorized in writing by the Contracting Officer. Upon completion or termination of this contract, all copies of any such proprietary or company confidential data will be returned.

(12) OTHER 10ENTiflED C ATEGORIES None, t

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