ML18150A024
| ML18150A024 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 04/07/1987 |
| From: | Reyes L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | Stewart W VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| References | |
| NUDOCS 8704140436 | |
| Download: ML18150A024 (25) | |
Text
e AP~ 0 7.1987 Yi'!:ginia Electric and Power Company
~11N:
Mr. W. L. Stewart, Vice President, Nuclear Operations*
P. o.* Box 26666 Richmond, VA 23261 Gentlemen:
e
SUBJECT:
MEETING
SUMMARY
- SURRY, DOCKET NOS. 50-280 AND 50-281 This refers to the meeting conducted at our request in Richmond, Virginia on March 25, 1987.
This meeting was held to discuss use of probabilistic risk assessment ( PRA) results for NRC inspections.
A list of attendees at the meeting is shown in Enclosure 1.
Some of the details of the meeting are provided in Enclosure 2 and Enclosure 3 contains the meeting handouts.
It is our opinion that this meeting was beneficial in that it enabled us to better understand your concerns related to the use of PRA information.
In accordance with Section 2.790 of NRC 1s 11Rules of Practice, 11 Part 2, Title 10, Code of Federal Regulations, a copy of this letter and its enclosure will be placed in the NRC Public Document Room.
Should you have any questions concerning this letter, we will be pleased to discuss them.
Sincerely, *
- .. fl¥l
~~uis A. Reyes, Director jv ~ivision of Reactor Projects
Enclosures:
- 1. Meeting Attendees
- 2. Meeting Summary
- 3. Meeting Handouts
£c w/encls:
/~. F. Saunders, Station Manager
~- J. Hardwick, Manager - Nuclear L Programs and Licensing
.JI<, Wright, Idaho National Engineering
- Laboratory (bee w/encls:
See page 2) 8704140436 870407 PDR ADOCK 05000280 G
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e ENCLOSURE 1 MEETING ATTENDEES VIRGINIA POWER/NRC MEETING MARCH 25, 1987 SUB~ECT:
USE OF PRA RESULTS IN THE NRC INSPECTION PROGRAM Name R. P. Croteau J. L. Ca 1 dwe 11 L. E. Nicholson S. Tingen D. Sommers R. Berryman R. Calder D. VandeWalle R. J. Hardwick S. A. Ahmed N. E. Clark E. S. Grecheck M. Smith J. Ahlados W. E. Holland F. Jape R. Wright Affiliation NRC NRC NRC NRC VA PWR VA PWR VA PWR VA PWR VA PWR VA PWR VA PWR VA PWR VA PWR VA PWR NRC NRC Idaho Nat. Engineer Lab.
ENCLOSURE 2 MEETING
SUMMARY
Licensee:
Virginia Electric and Power Company (VEPCO)
Facility: Surry Docket Nos~
50-280 and 50-281
- suBJECT:
USE OF PRA RESULTS FOR NRC INSPECTIONS The topics discussed included PRA applications for NRC inspection of Surry power station and future NRC inspection plans.
The opening remarks were made by Frank Jape followed by a presentation on the PRA applications program for inspection of Surry Power Station given by Ron Wright of the Idaho National Engineering Laboratory.
Frank Jape ended the meeting with a discussion of the future NRC inspection plans. contains more details on the areas of discussion.
e ENCLOSURE 3 AGENDA VIRGINIA ELECTRIC AND POWER COMPANY MARCH 25, 1987 INTRODUCTION PRA APPLICATIONS PROGRAM FOR INSPECTION OF SURRY POWER STATION FUTURE INSPECTION PLANS FINAL DISCUSSIONS AND QUESTIONS ALAN R. HERDT RON WRIGHT FRANK JAPE ALL
e USE OF PRA INSIGHTS FOR NRC INSPECTIONS I
PRESENTED TO VIRGINIA ELECTRIC AND POWER COMPANY FOR SURRY POWER STATION MARCH 25, 1987
Idaho National Engineering Laboratory PRA APPLICATIONS PROGRAM FOR INSPECTION OF NUCLEAR POWER PLANTS RON WRIGHT e
e
/
)
INTRODUCTION 0
PROGRAM HISTORY AND CURRENT STATUS e
o PROGRAM REQUIREMENTS AND PRODUCTS o
REGION II PLANTS e
PROGRAM HISTORY o
PURPOSE - THIS PROGRAM INTEGRATES PRA INSIGHTS INTO THE NUCLEAR POWER PLANT INSPECTION PROCESS o
- PROGRAM ATTRIBUTES
~ IDENTIFIES IMPORTANT PlANT SYSTEMS.
IDENTIFIES IMPORTANT COMPONENTS FOR RISK SIGNIFICANT SYSTEMS IDENTIFIES COMPONENTS FAILURE MODES IDENTIFIES COMPONENTS TO SPECIFIC INSPECTION MODULES PROVIDES IMPORTANT COMPONENT PLANT SPECIFIC CHECKOFF LIST
ADVANTAGES TO NRC
- 1.
THE PROGRAM IS A DIRECT RESPONSE TO THE COMMISSION'S POLICY AND PLANNING GUIDANCE WHICH CALLS FOR THE USE OF PRA IN SETTING INSPECTION PRIORITIES.
- 2.
INSPECTIONS WILL BE BETTER FOCUSED ON EQUIPMENT WHOSE FAILURE HAS THE GREATEST IMPACT ON PUBLIC RISK.
- 3.
INSPECTORS CAN MANAGE THEIR INSPECTION TIME BASED ON THE IMPORTANCE OF SYSTEMS AND THEIR COMPONENTS.
THIS SHOULD RESULT IN THE MORE EFFICIENT USE OF INSPECTION TIME.
__J
IMPORTANT FEATURES OF THE PROGRAM.
I
- 1.
IT IS CONSISTENT WITH AND CONSTRUCTED AROUND THE.
CURRENT IE MODULES.
- 2.
IT IS STRUCTURED FOR USE BY RESIDENT AS WELL AS REGION BASED INSPECTORS.
- 3.
IT COVERS ALL OF THE COMPONENTS AND ACTIVITIES THAT CONTRIBUTE SIGNIFICANTLY TO PUBLIC RISK.
- 4.
THE PROGRAM CAN BE APPLIED TO ANY FACILITY FOR WHICH A PRA HAS BEEN DEVELOPED.
- 5.
THE PROGRAM CAN B~ USED BY NRC INSPECTORS WITHOUT.
THE *NEED TO CONDUCT A DETAILED REVIEW OF THE PRA.
- 6.
A CLEAR DEFINITION OF WHAT CONTRIBUTES TO RISK IS PROVIDED WITHOUT REQUIRING PRA EXPERTISE.
PRA APPLICATIONS PROGRAM FUNCTIONAL AREAS o
PRA BASED INSPECTION GUIDANCE o
GENERIC BASED INSPECTION GUIDANCE
PROGRAM STATUS o
INEL INDIAN POINT 2 SEABROOK ZION HADDAM NECK (CV), GENERIC o
BNL LIMERICK INDIAN POINT 3 SHOREHAM MILLSTONE GRAND GULF o
PNL OCONEE I
REQUIRED INPUTS o
PROBABILISTIC RISK ASSESSMENT o
SYSTEM DESCRIPTIONS e*
o TECHNICAL SPECIFICATIONS o
TESTING PROGRAMS (LIST OF TITLES) o MAINTENANCE PROGRAMS (LIST OF TITLES) o EMERGENCY OPERATING PROCEDURES o
CHECK OFF LISTS
BASIC PRA PLANT PROGRAM DESCRIPTION o
INPUT PRA SEQUENCES o
DETERMINE EVENT IMPORTANCES o
SYNTHESIZE SYSTEM IMPORTANCE o
RANK SYSTEMS o
FOR TOP SYSTEMS, PROVIDE COMPONENT AND FUNCTION ANALYSIS e
TABLE 16.
MOST IMPORTANT SYSTEMS Importance for Importance for:
Codea Name Public Health Plant Damage RHR Residual heat removal
.83
.59 EP Electric power
.11
.07 cs Containment spray
.08
.07
.16 RC Reactor coolant
.01
.07 RP Reactor protection
.01
.13 SI Safety injection
.00
.08 cc Component cooling
.00
.06 e
MS Main steam
.00
.03
- a. These codes agree with those in Tables 14 and 15.
TABLE lA.
RESIDUAL HEAT REMOVAL SYSTEM FAILURE MOOE IDENTIFICATION The residual heat removal system is important fo~ long term recirculation cooling of the reactor following successful safety injection. The most important system failure is the V sequence, which consists of a loss of coolant accident (LOCA) via an interfacing system. This LOCA thus bypasses reactor containment.
Other system failures result in a loss of reactor long term recirculation cooling due to multiple RHR system failures.
Conditions That Lead to Failure
- 1. Reactor Coolant System to RHR Pumps Isolation Valves 1MOV-RH8701 and 1MOV-RH8702 Fail Open These valves line up RHR suction from the reactor coolant system.
Failure of these valves would expose the RHR piping to RCS pressure, thus creating a leak path which bypasses containment.
This accident scenario is s~veral times more important *to public health risk than all the other failures in Tables 1 through 9 combined.
Maintenance and surveillance of these valves should be observed or reviewed to minimize these failures.
- 2.
Operator Fails to Initiate Switchover from Injection to Recirculation Mode or Fails to Stop Pumc at RWST Low Level This is the dominant failure for the low head recirculation mode and is significant for the high head recirculation mode.
It involves the proper interpretation of plant status and proper initiation and completion of full switchover to recirculation.* Operator awareness of the criteria for switchover and adherence to emergency procedures are imoortant.
TABLE lB.
IE MODULES FOR RESIDUAL HEAT REMOVAL SYSTEM INSPECTION Failurea Module Title Components Mode 61701 Surveillance(CompJex)
RHR Pumps A, B 6
e 61726 Monthly Surveillance MOV-RH8701, 8702 1
Observation MOV-RH8700A, B 3
RHR Pumps A, B 6
62700 Maintenance RHR Pumps A, B 4
62703 Monthly Maintenance RHR Pumps A, B 4
Observation 71707 Operatirinal Safety MOV-RH8701, 8702 1
Verification MOV-RH8700A, B 2,3 Containment Sump 5
e RHR Pumps A, B 6
71710 ESF System Walkdown MOV-RH8701, 8702 1
MOV-RH6700A, B 3
Containment Sump 5
RHR Pumps A, 8 6
- a.
See Table lA for failure identification.
TABLE lC.
MODIFIED RESIDUAL HEAT REMOVAL SYSTEM WALKDOWN Component Number Pump lA Pump lB Noun Name Electrical RHR Pump lA RHR Pump lA DC and Ckt Bkr Spring Charging Motor Switch RHR Pump 18 RHR Pump 18 DC and Ckt Bkr Spring Charging Motor Switch 1MOV-RH8701 RCS to RHR Pumps Isolation 1MOV-RH8702 RCS to RHR Pumps Isolation Required Actual Location Position Position Bus149 G34 Racked In On Bus148 G33 Racked In On MCC1391-83 On MCC1381-B6 On 1MOV-RH8700A RHR Pump lA Suction Isolation MCC1393C-T5 On 1MOV-RH87008 RHR Pump 18 Suction Isolation MCC1383A-A4 On Valve Lineup 1MOV-RH8700A RHR Pump IA Suction Isolation 542 1 L22 Open 1MOV-RH8700B RHR Pump 18 Suction Isolation 542 1 M22 Open 1MOV-RH8701 RCS to RHR Pumps Isolation
.568 1 Z30 Open e
Initiator V
ETl ET2 ETl 1B41 ET3 ET2 ETl ET2 ETl ET4 ETllA ET12A ET7 ETl FIRE FIRE FIRE ET4 Notes:
Initiator V
ETl ET2 ET3 ET4 ET7 ETllA ETl 1841 ET12A TABLE 5.
INDIAN POINT 2 MOST IMPORTANT SEQUENCES Faulted S~stems Involved Public Health Events Importance A
.701 H
.120 H
.120 F
.003 A R2 RHR
.003 E
.002 E
.002 A OP41 SL2 PZR, PCS
.001 A LPl ACC, RHR
.001 A FZlA FP
.001 A FZ14 FP
.001 A FZ32A FP
.001 A Sll PCS
.001 Description Interfacing System Loss of Coolant Accident Large Loss of Coolant Accident Medium Loss of Coolant Accident Small Loss of Coolant Accident Steam Generator Tube Rupture Sass of Main Feedwater Turbine Trip Loss of Offsite Power Spurious Safety Injection 17
PRA Applications to *inspection Current PRA i'
IP2, IP3, i'
Westinghouse f'
Generic Plant Application
- Zion,
. non-PRA Plant Application Plan V Seabrook V
Application Plan V
Trial Plant e
~
1' Generic Generic Plan Application I"
V Trial Plant V. Evaluation e
CV& 201S
IMPORTANT SYSTEMS FOR EACH PLANT IP2 IP3 SEABROOK ZION RHR SW RHR RHR
-.e SW EP SSPS EP ccw ccw RWST cs CF RPS PCCW AFW EP HPI ESFAS RCS cs MS SW RPS SI RECIRC EFW SI ACC RCS EP ccw
'?
PR AFW MS AFW LPI SAS ACC cs CF e
GENERIC METHODOLOGY o
SYSTEM IDENTIFICATION FROM GENERIC DATA AND PLANT FUNCTIONAL REQUIREMENTS o
IMPORTANT COMPONENTS GENERIC DATA GENERIC FAULT TREE RESULTS. (ASEP)
-SYSTEM FAULT TREE ANALYSIS ENGINEERING JUDGEMENT
REGION II PLANTS o
SURRY (NUREG/1150) e o
McGUIRE, GENERIC
J, FUT~RE INSPECTION PLANS
- ANNOUNCED INSPECTION TEAM*OF FOUR OR FIVE INSPECTORS,-PLUS TEAM LEAD£~
INSPECTION REQUIRES ABOUT Two WEEKS ON-SITE REQUIRES COOPERATION FROM UTILITY:
OPERATORS SUPERVISORS CRAFTSMEN