ML20141H929

Summary of 970717-18 Meeting W/Tue Representatives in Glen Rose,Texas Re Licensee Proposed risk-informed Inservice Testing Program.List of Attendees Encl

ML20141H929
Person / Time
Site:	Comanche Peak
Issue date:	07/30/1997
From:	Fischer D NRC (Affiliation Not Assigned)
To:	NRC (Affiliation Not Assigned)
References
NUDOCS 9708040091
Download: ML20141H929 (7)

Text

-- - - -

p, cee k UNITED STATES 3

j NUCLEAR REGULATORY COMMISSION So //yg WASHINGTON, D.C. 2066M201 l hQ*** l July 30, 1997 l LICENSEE: Texas Utilities Electric Company (TVE)

FACILITY: Comanche Peak Steam Electric Station (CPSES)

SUMMARY

OF PUBLIC MEETING BETWEEN THE U.S. NUCLEAR REGULATORY COMMISSION STAFF, ITS CONSULTANTS, AND TUE REPRESENTATIVES l

CONCERNIRI THE LICENSEE'S PROPOSED RISK-INFORMED INSERVICE TESTING (RI-IST) PROGRAM Between July 14 and 18, 1997, the NRC staff (M. Cheok, G. Parry, D. Fischer) and its contractors (P. Davis, and R. Youngblood) reviewed probabilistic risk assessment (FRA) models, backup calculations, and data at Comanche Peak Steam  !

Electric Strtion (CPSES), Glen Rose, Texas. The review was conducted as part of the staff evaluation of Texas Utilities Electric Company's (TVE's) proposed ,

RI-IST program and was aimed at determining whether the CPSES PRA meets the '

quality and scope guidelines in draft regulatory guide DG-1061, "An Approach

for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Current Licensing Basis."

On Friday, July 18, 1997, the staff met with representatives of TUE in a public meeting to describe the staff's review and to present preliminary observations regarding the adequacy of the licensee's PRA for the RI-IST (and other) applications. A list of meeting attendees is included as Attachment A.

The quality of the CPSES PRA was found to be generally adequate and backup calculations were well documented in most areas. Major review areas include:  ;

initiating event analysis; accident sequence analysis; mission success criteria; fault tree analysis; data analysis; dependent failure analysis

, including consideration of common cause failures; human reliability analysis; sequence quantification; internal flooding, fire , and tornado analysis; uncertainty analysis; and outage safety function guidelines. A list of documents that were made available to the staff during the review is included as Attachment B. While the review team identified some minor problems with the CPSES PRA for the RI-IST application (e.g., missing success paths, inadequate documentation of human error probabilities, optimistic recovery factors for equipment repair, plant-specific performance data not having already been incorporated into the PRA), the review team concluded that these issues could be addressed through the licensee's expert panel process. The staff also identified an area in the calculation of sequence success that would need further clarification. The calculated core damage frequency from 1 the licensee's base PRA will be on the order of IE-4 per year when external event initiators and shutdown operations are taken into account. Thus, h{$

v \

approval of the licensee's proposed RI-IST program, considering uncertainties, will be " subject to increased NRC technical and management review" as stated in DG-1061.

3 TUE stated that they would have a difficult time responding to the staff's

, March 12, 1997, and June 9, 1997, RAls by August 1997 (reference memorandum to

the Commission dated June 17,1997). TUE estimated that it would take one 97E8 0 1 970730 S) 2 PDR ADOCK 05000445 ""

P PDR

1 staff-year of effort to adequately respond to these RAls (i.e., to assess how their proposed RI-IST program comports with the staff's draft risk-informed regulatory guides).

David C. Fischer, Sr. Mechanical Engineer Mechanical Engineering Brani:h Office of Nuclear Reactor Ragulation DISTRIBUTION:

Hard Cony: E-Mail Coov:

File Central T. Martin (TTM)

PDR B. Sheron (BWS)

Docket File G. Lainas (GCL)

EMEB RF/CHRON R. Wessman (RHW)

PDIV-1 r/f G. Holahan (GMH)

OGC G. Parry (GWP)

ACRS J. Flack (JHF)

D. Ross (SLM3)

E-Mail Coov: A. Thadani (ACT)

J. Murphy (JAM 1) G. Werner (GEW)

S. Collins (SJC1) H. Freeman (HAF)

W. Hodges (MWH) F. Miraglia (FJM) l M. Cunningham (MAC3) J. Roe (JWR)

8. Hardin (WBH) E. Adensam (RGA1)

P. Baranowsky (PWB) W. Bateman (WHB)

S. Black (SCB) C. Hawes (CMH2)

R. Zimmerman (RPZ) T. P. Gwynn (TPG)

J. Clifford (JWC) J. Tapia (JIT)

T. Polich (TJP1) B. Henderson (BWH)

K. Thomas (KMT) C. Thomas (CRT)

M. Hammond (MFH2) PMNS OPA DOCUMENT NAME: G:\FISCHER\MEETSUM7.1B Nt N '/.n E uN N w N *'

OFFICE EMEB:DE g SC:EMEB: 6 SPSB:DSSA SC:SPSB BC:EMEB c QX DE _o :DSSA m :DE WE DCFischer DTerah MCheok e MRubY RHWedmh-DATE 7 /2f/97 -] /7 B/97 1 /2s/97 I7/ 7 7 F /97 0FFICIAL RECORD COPY I

ATTACHMENT 1 -

. tJRC Revie to of cpfes PR4 Jul.Y 18, /997 gggggg N#1E Come)V B/oal ji%jj C,1m 8.Conw 7v elecnetc - rs4-897 oiet - - !

!, !!, 9a 4 *"ju ~****-

2s4-re7 7541 1

!!!!!!i Mtukty R Gilyore 7k E/edr'6 254 N1 010J

) P'R. Da vis PRD Con.s ul tin 3 (3oi) c.r3-:S'17 6 g rtA V. key NdC/A'dA/p.CC8 (101) 4/C-IV49 h6 d4A)()&&oop SIMWEU(. lNr. ($> / ) 448- 4 t/Ls'~

3 A M F 6 c H tA. tJ RC J DE JEM9.G (3.d W -2 M B M\KE C.WE o K

, WA C.[ DssA[S PsS (3 en} q t 5 -9315D Tm Ol'O*rtl Wheftd/>H 4 L& 4tc- Isot Q9 ULrner MRC/koe- Resikt (Kv) - St1-ts co I

'v % u T tv w t a. 'Tu e c.ec r a i e (25Q M ,- 6o ? 4 I Ercats) h Kapaae 72t a=cecruc.

(es9) 277-2047 An /Yf9fJ 7~t/ Eleclitc 2 S IS - l'77- 68/(0 fi'*3.ip, itff T uWthe ga gy w m y 29 Y --et 4 -4a3

"~

7km Gews hesd & sue.% a ca, x&At (4bz) 313L s ts t

- . . - . - .. -- - . . .- - - - . _ -~. - . __ - _ .- - .-. . ----.-,

> "^  !

" L to Summert NRC "tVL of the PRA ATTACHE 1T 2 Page1 Note: All procedures, drawings, and references are IPE Baseline beoed on freeze date.

A. GeneralPlantInformation Mecaneneous procedures and task plans in Support of the IPE / IPEEE (al areas below)

4. 1. Operating Procedures I

. OperatonalProcedures-Teetng(OPTS) i

System Operaung Procedures (SOPS) l 1

t 2. AOPs '

I Abnormal Condbons Procedures (ABNs) {

3. EOPs 1

Emergency Operaung Procedures (EOPs) i CPSES Emergency Response Guidelines (ERGS) -

)

4. SR Procedures Master Survemence Test List (MSTL)

5. Tech Specs '

i

4. P&lDs Flow diagrams and I&C '

i M1-0200 & Mi-2200 as referred to in the IPE Notebooks  !

7.  !

Electrical schematics for the systems modeled

.j Electncal Drawings and Schema 6cs, E1 Drawings as rMerred to in the IPE Notebooks S. General PRA information i

1. Guidance Documents (if any) e.g. processes i

Original IPE Procedures (RXE) and updated ECE procedure and associated desktops j IPEEE Program Plan, January 1993 i

CPSES IPE Intemal Flooding Analysis Technical Procedures (RXE SY-IPE.7, Rev 0) l

2. Documentation of Plant Walk Downs CPSES IPE Intemal Flooding Analysis, Plant Walkdown information (including 3 books of photographs)  ;

CPSES IPEEE Sommic Walkdown Report & Photographs i CPSES IPEEE Tornado Walkdown Report & Notes & Photographs I C. Initiating Events l' Generic & Plant Specific date used (e.g., failure rates, etc. . may be in PRA airoedy)

CPSES IPE inibahng Events Analysis Support Systems initiators D. Accident Sequenos analysis and Success Critoria Peer independent review of the CPSES IPE by outside consultants & responses

1. Quantilled event treet CPSES IPE Accident Sequence Analysis

2. Suomess Criteria and References (e.g., FSAR, MAAP code, etc.)

Demon Basis Analyes of a Postulated Steam Generator Tube Rupture Event for CPSES, Unit 1 (RXE-88-101-P)

Jomt Wesunghouse Owners GroupM/es6nghouse Program: Assessment of Compliance with ATWS Rule Basis or Westinghouse PWRs(WCAP-11993)

Evaluation of Survoisance Frequencies and Out of Service Times for the Engineered Safety Features Actuation System (WCAP-10271 Suppioment 2)

Interfacing Systems LOCA: Pressurtred Water Reactors (NUREG/CR-5102)

MAAP Calculetons of Accident Basebnes Representing the CPSES IPE Plant Damage States (RXE-LA-003)

Main Report, Appendices A-C, D-G,

-i J

Materials to Summort NRC Review of the PRA July 24,1997 (1:45pm)

]

Page 2

, MAAP Parameter Ffe Development (RXE-LA-CP1/1-025) Volumes I & 11, Xerox box fut of MAAP runs

RXE-TA-CP1/0-023, Rev. 0,

• Responses to T-H Questons from the IPE Accident Soquence Model Development'

] RXE-SY-CP1/1-033, Rev. O, ' Success Crtteria Calculebons and Interferences

• 1

, RXE-TA-CP1/0-017, Rev.1,' Design Basis SGTR Analyse" i-

! SWEC Calculaton 16345-ME(B)-389, Rev.1, Cale Change Notco No.1, ' Refueling Water Storage Tank Setpoints, Volume Requirements and Time Depiebon Analysis' E. Systems Analysis l

CPSES iPE Systems Comment and Resoluton Notebook Peer independent review of the CPSES lPE by outside consultants & responses Review of the CPSES IPE by plant engineere & responses 1 i

! ' 1. Fault Trees l 2. system Notebooks

CPSES Stabon Blackout Study, Temperatures in Various Rooms

?- CPSES IPE System Model Notebooks

Auxihary Feedwater (AF)

Component Cochng Water (CC) '

CondensateFeedwater (CF)

Safety Chided Water (CH) instrument Air (Cl) l Containment Spray (CT)

Circuisbng Water (CW)

Electncal Power (EP)

Engineered Safeguards Features Actuabon (ES) i Main Stoem (MS)  !

Reactor Cooient(RC)

Reactor Heat Removal (RH) 1 Safety injection (SI) j Service Water (SW)

Mini Systems (CZ,DD,TW)

Chermcal Volume Control (CS)

CPSES IPE Systems Analysis Appendix A (AF.CC,CF; CH,Cl,CS;  !

CT,CW,CZ,DD; EP; ES Pt.1; ES pt.2; MS,RC,RH; St.SW,TW,VA)

CPSES IPE Intemal Flooding Analysis, Calculagon (RXE-SY CP1/1-021 Volumes 1 & 2)

CPSES IPE Support Systems interfaces RXE-SY-CP1/1-030, Rev. 0, *Probabety Analysis for Off-Site Power Non-Recovery Events' RXE-SY-CP1/1-034. Rev. O, ' internal Flooding Analysis Flow Propagabon"

3. Room heatup calculations HVAC design basis documents and calculabons RXE-SY-CP1/1-028, Rev. O,' Pump Failure Probabety Analysis Due to Loos of Room Cooling" RXE SY-CP1/1-027, Rev. O, ' Equipment Response to loss of Cooling"

4. EQ standards F. Data Analysis Peer independent rev6ew of the CPSES IPE by outside consultants & responses

1. Generic data and sources Losses of Off-Site Power at U.S. Nuclear Power Plants - Through 1995 (EPRI TR 106306)

CPSES iPE Data Analysis -IPE Genonc Data Base CPSES Plant SpeelRc Data Analysm

2. - Maintenance data and sources

' See Secton F,1 above for generic data IPE Proventagve Maintenance (PM) Date (Plant Specific)

G. E:;:M:-t Failure Analysis Peer independent review of the CPSES IPE by outside consultants & responses

l Materials to Buenort NRC Review of the PRA July 24,1997 (1:45pm)

,, Page 3

.6

Date Base used See Secton F above i .j j H. Human Reliability Analysis '

Peer independent review of the CPSES lPE by outside consultants & responses ,

1. Guidance document See Secton A above

! 2. Quentification of final HRA values used CPSES IPE Human Reliability Analysis (HRA) l

CPSES IPE HRA Calculation File l L Quellnusikn and Raoults

, Peer irdependent rewlew of the CPSES IPE by outside consultants & responses 4.

i

1. Sensitivity calcuktions & reports (if available)

' CPSES IPE Interfacing System LOCA Analysis CPSES IPE Accident Sequence Quantecation (Volumes 0,1 through 7, and System Modification Form Log Book)

I

2. Uncertainty calculations i

J. Shutdown PRA (Not yet done for CPSES?

Would like to see what is available.)

i

• Outage Risk Assessment and Management (ORAM) Software implementation at CPSES (ER-EA-011)

Outoge Safety Function (OSF) Guidelmas (Draft, Rev. 0)

Unit 1 refueling outage No. 5 Risk Profile (elides)

K. Seismic PRA l 1

i Equipment HCLPG (might or might not use)

CPSES IPEEE Seismec (ER-EA-001) (samme margins) l CPSES lPEEE Somme Waikdown Report L Maintenance and update Process (for IPE/PRA)

Guidelines 1 Procedure Embeddedin above documents M. Other Plant Documents TU's RI-IST sutettais including responses to RAls TU's IPEAPEEE submittais including responses to RAls IPE OverWwvugraphs IPE CPSES Volume I: Front-End Analysis (RXE-g2-01 A)

IPE CPSES Volume it Back-End Analysis (RXE-g2-01B)

Expert PanelGuidance Document GL 88-20 Submittais, NRC RAls and Responses.

IPEEE for Severe Accident Vulnersbeties (ER-EA-006)

CPSES IPEEE Tomado Risk Assessment (ER-EA-004)

CPSES IPEEE Tomado Waikdown Report -.

CPSES IPEEE Etemal Flood, Transportsbon & Nearby FacWty Accidents (ER-EA-005)

CPSES IPEEE Fire Evaluation (ER-EA-006, Rev 0)

N. Containment Analysis Level 11 CPSES IPE Containment Event Tree Analysis Plant Damage States for CPSES IPE (RXE-LA-CP1/D-002)

MAAP Calculations of Accident Baselines Representing the CPSES IPE Plant Damage States (RXE-LA-CP1/0-003)

i I

, Materials to Support NRC Review of the PRA July 24,1997 (1:45pm)

• Page 4 1

• Containment Failure Charactertration (RXE-LA-CP1&OO4)

The CPSES Containment Performance During Severo Accidents (RXE-LA-CP1/0-009)

RXE-SY-CP1/1-031, Rev. O, 'Probab6shc Analysis for Containment Recovery Due to Overpressurization' RXE-SY-CP1/1-029, Rev. 0,'Containtnent Rdahikty Analysis Due to Overpressurization*

i i