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| document type = CONTRACTED REPORT - RTA,QUICK LOOK,ETC. (PERIODIC, TEXT-PROCUREMENT & CONTRACTS | | document type = CONTRACTED REPORT - RTA,QUICK LOOK,ETC. (PERIODIC, TEXT-PROCUREMENT & CONTRACTS | ||
| page count = 126 | | page count = 126 | ||
| project = TAC:M66591 | |||
| stage = Draft Other | |||
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Latest revision as of 16:43, 20 March 2021
| ML20235J229 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 09/30/1987 |
| From: | Fresco A, Usher J BROOKHAVEN NATIONAL LABORATORY |
| To: | NRC |
| Shared Package | |
| ML20235J225 | List: |
| References | |
| CON-FIN-A-3453 A-3453-87-5, A-3453-87-5-DRFT-R, A-3453-87-5-DRFT-R00, TAC-M66591, NUDOCS 8710010359 | |
| Download: ML20235J229 (126) | |
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1 A A-3453-87-5, Rev. 0 - GRAND GULF NUCLEAR STATION UNIT 1 PROBABILISTIC RISK ASSESSMENT-BASED. SYSTEM INSPECTION PLANS 'I Prepared by: l l J. Usher and A~. Fresco l Engineering Technology Division September 1987 l I l
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'i Prepared for: U.S. Nuclear Regulatory Cornission (Region I) ;I Contract.No.' DE-AC02-76CIl00016 NRC FIN A-3453 l 8710010359 PDR ADOCK O B7 g g l pg i P
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, BNL Technical Report A-3453-87-5, Rev. O g
i l l j j GRAND CULF NUCLEAR STATION UNIT 1 PROBABILISTIC RISK ASSESSMENT-BASED j SYSTEM INSPECTION PLANS l Prepared by: J. Usher and A. Fresco Engineering Technology Division Department of Nuclear Energy Brookhaven National Laboratory Upton, New York 11973 September 1987 i Prepared for: U.S. Nuclear Regulatory Coamission ) under FIN A-3453 1 1 l u ;
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CONTENTS Section Title Page
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' DISCUSSION............................................ 1
- 2. SYSTEM PRIORITY LIST................................. 1
- 3. DOMINANT ACCIDENT SEQUENCES.......................... 1
- 4. COMMON CAUSE FAILURES..........................~...... 4
- 5. IMPORTANT HUMAN ERR 0RS............................... 4-
- 6. SYSTEM INSPECTION TABLES............................. 5 i
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" -iv-TABLES Table No.- Title Page1 1 System Priority Ranking.............................. 3 LA1 Emergency Electric ~ Power System (EPS)'............... A-1 Al-1 Importance Basis and Failure Mode Identification..... A-1 Al-2 ISE Inspection Procedures for System Operation....... A-4' Al Modified System Wa1kdown............................. . A-5 Al-4 Proposed Inspection Plan for Diesel' Generators at Nuclear.P1 ants.................................... A-7 A2 High Pressure Core-SprayJ('HPCS)' System............... A-ll-A2-1 Importance Basis'and Failure Mode Identification..... A -A2-2 I&E Inspection Procedures for.Systen Operation.......'
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A-14 A2-3 Modified System Wa1kdown..................... ....... A-15 A3 Reactor Core Isolation _ Cooling (RCIC) System.'... ....
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. A-20' A3-1 Importance Basis and Failure Mode Identification..... A-20 A3-2 I&E Inspection Procedures for System Operation....... A-23 A3-3 Modified System Wa1kdown............................. A-24' A4 Reactor Protection System -(RPS)...................... A-27 A4-1 Generalized Inspection P1an......'............ ....... A-27 A5 Standby service Water (SSW) System.................,.. A-31 AS-1 Importance Basis cnd Failure Mode Identification..... A-31 A5-2 I&E Inspection Procedures for System Operation....... A-34 AS-3 Modified System Wa1kdown............................. ,A-35 A6 Standby Liquid Control (SLC) System.................. .A-48 A6-1 Importance Basis and Failure Mode Identification..... sA-48 A6-2 I&E Inspection Procedures for System Operation....... A-50 A6-3 Modified System Wa1kdown............................. A-51 B1 Automatic Depressurization System (ADS).............. B-1 B1-1 Importance Basis and Failure. Mode Identification....~. B-1 B1-2 I&E Inspection Procedures for System Operation....... B-2 B1-3 ' Modified System Wa1kdown..............................- B-3 B2 Condensate System (CDS)............<................. B-6
'Importance Basis and Failure Mode Identification.....
B2-1 B-6 B2-2 I&E Inspection Procedures for.Systeu Operation....... B-8f B2-3 Modified System Wa1kdown............................. B-9
t > t a _v. TABLES (Cont'd) Table No. Title Page B3 Containment Venting System (CVS)..................... B-ll B3-1 Importance Basis and' Failure Mode Identification..... B-ll B3-2 I&E Inspection Procedures for System'0peration....... B-13 B3-3 Modified Systen Ua1kdown............................. B-14 B4 Control Rod Drive (CRD) System....................... -B-16 B4-1 Importance Basis and Failure Mode Identification..... B-16 B4 I&E inspection Procedures for Systeu Operation....... B-18 B4-3 Modified System Walkdown............................. B-19 B5 Low Pressure Core Spray (LPCS) System................ B-21 B5-1 Importance Basis and Failure Mode Identification..... B-21 B5-2 I&E Inspection Procedures for System Operation....... B-24 ) B5-3 Modified System Walkdown............................. B-25 i B6 Residual. Heat Removal (RHR) 5ystem................... B-28 B6-1 Importance Basis and Failure Mode Identification..... B-28 B6-2 I&E Inspection Procedures for System Operation....... B-36 B6-3 Modified System Walkdovn............................. 3-37 B7 Suppression Pool Makeup (3PMU) System................ B-44 B7-1 Importance Basis and Failure Mode Identification..... B-44 B7-2 I&E Inspection Procedures for System Operation....... B-46 B7-3 Modified System Walkdown............................. j B-47 ; a C1 Plant Operations Ir.spection Guidance................. C-1 C2 Surevillance and Calibration Inspection Guidance..... C-3 C3 Maintenance Inspection Guidance...................... C-9 ] C4 Containment and Drywell Walkdown..................... C-15 l l l L.
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~' Figure No.- Title Page- d Al Emergency. Electric Power System Schematic......... A-9 Al-2 Emergency Electric Power System Dependenc Diagram..................................y .........
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p A2-1 High' Pressure More Spray. System. Schematic......... A-16 A2-2 High Pressure Core Spray System. Dependency Diagram....................'....................... A-17 i l
'A2-3 '
System Actuation Dependency Diagram...............' A A3-1 Reactor Core Isolation Cooling System Schematic......................................... A-25. A3-2 Reactor Core' Isolation Cooling System-
, Dependency Diagram................................ A-26' A4-1 Basic Scram System.........'....................... A-29 A4-2 RPS Bus Power Distribution........................' -A-30 A5-1 . Standby Service Water System. Schematic............ .A-33' A5-2 Standby Service Water Cross-tie System 1 Schematic.. A-40 AS-3 Standby Service Water System...................... A-41' AS-4A- SSW Rasins and Pumps..............................-
A-42 AS-4B SSW Rasins and Pumps.............................. A-43 AS-5 SSW Loop A.............'...........................- A-44 ] A5-6 SSW Loop B........................................ A-45 j AS-7 HPCS Service. Water................................ 'A-46 i A5-8 System Interfaces........-......................... A-47, I A6-1 Standby Liquid. Control System Schematic........... A-52 a
- i. B1-1 Reactor Ucpressurization. System Valves and i B1-2 Safety Relief Valves Schematic.................... B-4 Reactor Depressurization System Dependency D1agram....................'.......................' B-5 B2-1 Condensate System:Schecatic....................... B-10 B3-1 Containment Venting System Schematic.............. -l B4-1 B-15 i Control Rod Drive-System, Schematic................ B-20 l B5-1 Low Pressure Core Spray System Schematic.......... '
B-26 l B5-2 Low Pressure Core Spray System Dependency; Diagram........................................... B-27 B6-1 j Low Pressure Coolant Injection System Schematic... IF39 j B6-2 Low Pressure Coolant Injection System Dependency Diagram................................ 3-40. l B6-3 Residual 11 eat Removal' System: Containment B6-4 Spray Mode Schematic.............................. B-41 Residual Heat Removal System:. Shutdown' Cooling' '
. Mode Schematic....................................
B-42 ; B6-5 Residual Heat Removal System: Suppression Pool Cooling Mode Schematic............................ ' &-4 3 B7 Suppression Pool Makeup System' Schematic.......... B-48 L 4 l l
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'CRAND GULF-NUCLEAR STATION,. UNIT 1 PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLANS
- 1. DISCUSSION The tables and; paragraphs'in this inspection plan have been prepared to pro-vide Inspection guidance . based on review of the NUREG-1150 / ' Grand Gulf Probabilistic Risk " Assessment -(PRA), (Section 7, Reference 1).; The. guidance should be used to aid in~ the selection of areas to inspect and is not intended either to. replace current-I&E inspection'gui' dance or to constitute an additional
-set of ' inspection ' requirements. In using this information one should realize that is based almost entirely on the : Grand . Gulf PRA.. Hence, recent: system experience, f ailures, and -modifications should be considered when reviewing
'these tables.- :Since- plant modifications; are normally an ' ongoing continual '
process, it is recommended that relevant changes be catalogued 'so that these inspection plans can be periodically revised as required. j j
- 2. SYSTEM PRIORITY LIST !
The Grand Gulf core melt prevention systems'have been ranked or grouped in j Table 1 according to their importance in, preventing core acit. They have- been } arranged into two groups of high importance (A) and lower inportance (B). Other- j plant systems not appearing in the list are generally .of lesser importance- than j those included here. . Within each group, system importance is considered ' to be ; quite similar, hence, the systems are listed alphabetically. The systeras were 'I ranked or grouped by reviewing 'importance calculations perf ormed in the PRA or by the SARA computer code developed by INEL. These calculations were ' performed based on system appearances . in the dominant accident sequences of - the event trees. The final grouping was determined . af ter a review of-both the PRA and I SARA importances, na well'as generic BWR importance considerations.
- 3. DOMINANT ACCIDENT SEQUENCES The Grand Gulf PRA has a number of different accident ' sequences that t
contribute significantly to overall core . melt risk. The three types 'of. j sequences that dominate core melt' risk at Grand Gulf are: ' Station Blackout Transients with loss of core cooling L -
' Anticipated Transients Without Scram (ATWS).
There - are five station blackout L sequences, which contribute approximately 99% of the total core melt risk: 5 I 1 i i i 0 N:
e , SB-1: Following'the loss of offsite power, successful scram occurred. The SPVs properly cycled, AC power to Div. 1 & 2 has . failed but AC power to Div. 3 (HPCS) is available. Core cooling is provided by HPCS but eventually f alls when suppression pool temperature reaches the 200*C to 280*C range. Core damage occurs and' containment fails f rom over- i pressurization. ' SB-2: Similar to S3-1 except HPCS fails immediately (random failures, main-tenance,: etc.). RCIC provides initial core cooling but eventually fails on high turbine exhaust pressure. SB-3: Same as SB-2 except RCIC fails immediately. SB-4: Same as SB-2 except HPCS fails because of loss of AC power to Div. 3. SB-5: Same as SB-3 except itPCS fails because of loss of AC. power to Div. 3. There are two sequences contributing predominantly to transients with PCS initially available. l Af ter reactor scram, no high pressure or (subsequent to reactor f depressurization) low pressure core cooling is provided, resulting in j l l early core damage. Containment cooling is successful, but it is j assumed that 'non-condensible f ormation ultimately results in the need for containment venting. Containment venting is not successful; con-tainment failure subsequently occurs. SPtfJ is not available and all j j modes of RHR fail. (Sequence Il 84) ! j ! l Same as above except that the reactor. is not depressurized following j failure of the high pressure injection systems (IIPCS, RCIC, CRD). 1 (Sequence # 100) There are three dominant ATWS sequences contributing to core-melt risk: 1 A transient occurs but the reactor fails to trip (RPS mechanical fail- I l ure). MSIVs are closed, recirculation pump trip ( RPT) , and one SLC l pump actuated. The reactor is initially at high pressure ( ADS f ail-ure), SRVs cycling open/close, and HPCS actuated. The operator fails to depressurize the reactor when the suppression pool temperature i (200*C to 220*C) causes HPCS failure. Core uncovers and core damage results. (Sequence # 7) MSIVs open and both automatic and manual actions to scram the reactor f have failed. RPT has occurred and power level equilibrates at 17 to 40% depending on reactor water level. MSIVs remain open and FW con-tinues to provide coolant makeup. Due to.RPS mechanical. failure reac-tor suberiticality is not achieved. The suppression pool slowly heats up . thus slowly pressurizing containment. Since coolant makeup is being provided by FW, only the containment is vulnerable (containment failure does not result in deformation of injection lines and thus a subsequenc loss of inj ec tion when reactor power transferred to the suppression pool is greater than 5% (i.e., operator is maintaining a high level in the reactor). (Sequence # 34)
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.. J Transient has occurred but the reactor. failed .to. trip. MSIVs j initially open but then close (FW fails), RPT has occurred. 'SLC 'has b failed. ADS has succeeded. Low pressure systems inj ect. Reactor suberiticality is not. achieved. As the suppression pool. heats up, the, j containment is pressurized from steam generation. Containment failure El eventually occurs. (Sequence # 57).
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l Table,1 - System Priority Ranking Top Croup: Emergency Electric Power (EPS) ( Emergency Ventilation System (EHV) High Pressure Core Spray (HPCS); ] Reactor Core Isolation Cooling (RCIC) 'f Reactor Protection (RPS) { Standby Service Uater (ESW) ; Standby Liquid Control.(SLC) ' Lower Group: Automatic Depressurization ( ADS) Condensate (CDS) f l l Containment Spray (CSS)* Containment Ven'ing t (CVS) l Control Rod Drive (CRD) Instrument Air Systea (IAS) Low Pressure Coolant Injection (LPCl)* Low Pressure Core Spray.(LPCS) Shutdown Cooling (SDC)* ) Suppression Pool Cooling (SPC)*' Suppression Pool Makeup (SPMU)
- These systens have been combined under the Residual' Heat Removal (RH in the inspection plan
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. 4 COMMON CAUSE FAILURES to risk.
The failure of nultiple items from some common cause can be very significant I The Grand Gulf PRA are particularly important: has identified several common cause failures that Loss of offsite-power, I Common node failure of diesel generators A (11) and B (12), Common ~ mode failure of the DC batteries, Common mode failure of the standby service water system trains A and B (1, 2), 1 l ADS valves fail because of a common cause failure. j 1 l Other common .cause failures, not considered to be as important as .those l l above, are identified in the failure mode tables which follow. )j
- 5. IMPORTANT HUMAN ERRORS l
Human errors can be very significant to overall plant risk. .The Grand l Gulf risk: PRA has identified several human errors as important contributors to l
. 0perator f ails to nanually initiate the diesel generatora, l Operator f ails to manually initiate !!PCS, 1
Operator f ails to manually initiate RCIC, Operatar fails to open suppression pool suction valve F031 for RCIC suction switchover, Failure to restore valves -in SLC test line, flow diverted from the reactor, Failure to restore SLC valve F031 af ter. suction . test, flow diverted from reactor Operator fails to initiate SLC, I< Operator inhibits ADS, Operator fails to inhibit ADS. Other human errors are also identified in the failu.re node tables. i v sbpoi e
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, 6. SYSTEM INSPECTION TABLES Three table's.have generally.been prepared:forceach system to provide:inspec-tion guidance. These tables are described below:
Table X-l'- Failure Modes-s Those components. or! licensee activities which play a dominant role in' con-i tributing . to system importance care presented, :along with .a brief , description of-why these items are important. . Inspection focussed onf these items 'will address - approximately ' 95%; of the. risk significant . areas. For. experienced ' inspectors this table is probably ' sufficient.: A . simplified system diagram. extracted from the PRA is included ' giving. the valvel numbers y f or each system. :No ' separate - tables _ are included for :ESF, EVS or IAS since. these. systems are modeled in the PRA only.as contributo's r to the other systems.- Table X I&E Procedures-For those who prefer additional guidance,.this table ' identifies those11&E inspection procedures which can be usedrco. assure the availability of.the items shown in Table X-1. The inspection procedures l were identified based oni the f ailure modes presented :and .an understanding . of 1&E procedures. 'The procedures selected are .those which provide routine guidance on the ' principal plant ' pro-gramnatic activities such as operations, maintenance, instrumentation / control, and surveillance testing. There are many other inspection procedures which-could ference. also be used depending on.the inspection criteria or'the' inspector's pre-- However, the procedures ' selected will- generally ' provide ' adequate inspection coverage of the doulnant failure modes. Table X Hodified Systeu Walkdown This table provides an abbreviated version of the licensee's system' check-list, where available, but includes only those items . which are related to the-dominant failure' modes. It is generally auch less' than' the . normal checklist. Caution should be observed when using ;the checklists, since they - are based on certain versions.of the licensee's system operating instructions. Valve numbers used are those identified in the licensee system checklists.
- 7. REFERENCES
- 1. M.T. Drovin, et al.,
" Analysis of Core Damage Frequency from' Internal .
- j. Events: Grand Gulf, Unit 1, (Main Report and Appendices)," _ NUREC/CR-4550r l
SAND 86-2084,. Volume 6, Sandia National Laboratories (November 1986). (
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. A-1 GRAND GULF NUCLEAR STATION, UNIT 1 PROBABILISTIC RISK' ASSESSMENT-BASED INSPECTION PLAN Energency Electric Pouer System.(EPS)
TABLE Al-1 IMPORTANCE BASIS AND FAILURE MODE IDENTIFICATION CONDITIONS TilAT CAN LEAD TO FAI' URE General Guidance 1 Surveillance of the licensee's periodic calibration, testing and/or pre- ] ventive or unscheduled maintenance activities, procedures and training and/or i normal and emergency operating procedures, training and check-off lists in ac- l cordance with the Technical Specifications and relevant NRC bulletins and in-formation notices should reduce the' probability of failure for the conditions listed below. The most relevant aspects are designated for each condition as follows: 1 FC - Periodic calibration activities, procedures and training. PT - Periodic testing activities, procedures and training. l MT - Preventive or unscheduled maintenance activities, procedures and ; 2 training. OP - Normal and etae rge ncy operating. procedures, check-off lists, training, etc. _j - Mission Success Criteria The EPS consists of the AC and DC power divisions required by the Engi-neered Safety Features (ESP) to safely shut down the plant. Both AC and DC power are divided into three separate divisions: two of the divisions (1 and. ;
- 2) are for the majority of the ESF and the third (3) is dedicated to the HPCS ;
system and its required support systems. The ESF AC divisions normally ' receive power from one of three offsite sources; two offsite sources are 500kV which are stepped down to 34.5kV and then converted to 4.16kV, the third off- j site source is 115kV which is stepped down directly through a ll5kV/4.16kV transformer. In addition to the normal supply from the ESF transformers, each H ESF 4.16 kV bus has a standby diesel generator (DC) which is available to sup-ply bus loads upon a loss of normal AC. The ESF 125 V DC system includes three divisions, each consisting of two battery chargers.which normally supply the load and a bank of batteries which function as a backup. - 1 1
- 1. DG 11 Fails to Start or Run or is Out for Maintenance j l
In the event of loss of normal AC power, DG 11 powers Division 1 of the i EPS (ESF systems requiring 4160 V bus 15AA). (MT, PT) -
- 2. DG 12 Fails to Start or Run or is Out for Maintenance In the event of loss of normal AC power, DG 12 powers Division 2 of the EPS (ESF systees requiring 4160 V bus 16AB). (MT, PT) j l
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4- , ,.=*; , A-2 TABLE Al-1 (Cont'd) CONDITIONS TilAT CAN LEAD TO FAILURE i
- 3. OG 13 Fails to Start or Run or is Out for Maintenance j l
In the event of loss . of normal AC power, DG 13 powers Division 3 of the ! EPS (llPCS - 4160'V bus 17AC). (MT, PT) i 1
- 4. Common Mode Failure of Diesel Generators .
1 A common mode failure of the' diesel generators for AC Div. 1 and 2 is - included in the fault tree model. Common mode failure of the HPCS DG was not
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modeled since it is smaller than the other diesel generators. .
- 5. 4160 V AC Bus 16AB Failure ESF systems reliant upon AC Div. 2 power are rendered unavailable by fail-ure of 4160 V AC bus 16AB. (MT, PT)
- 6. 4160 V AC Bus 15AA Failure :
ESF systems reliant upon AC Div. I power are rendered-unavailable by fail-ure of 4160 V AC bus 15AA. (MT, PT)
- 7. 4160 V AC Bus Failure Due to Loss of HVAC <
Room cooling is required for the emergency electrical switchgear rooms. Ohmic heat must be removed from the switchgear rooms to prevent switchgear failure. Loss of cooling to the switchgear rooms renders .the AC power. divi- ] sions inoperable. (MT, PT)
- 8. Common Mode Failure of Battery Divsions The ESF AC divisions require DC power f rom the associated DC buses for circuit breaker control power, diesel generator field flashing, and the diesel fuel oil booster pump. (MT, PT) 9.125 V DC Bus Failure Due to Loss of IIVAC l Room cooling is required for the battery rooms, the _ major purpose of which is to remove hydrogen generated during battery charging. See Item '8 above.
(MT, PT)
- 10. Battery Bank 1B3 Failure Failure of DC battery bank 1B3 renders AC Div. 2 unavailable. (MT, PT)
- 11. Battery Bank 1A3 Failure Failure of DC battery bank 1A3 renders AC Div. I unavailable. (MT, PT)
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A-3 TABLE Al-1 (Cont'd) j i CONDITIONS THAT CAN LEAD TO FAILURE I l
- 12. 125 V DC Bus 11DB Failure
-l Failure of DC bus 11DB rendets EPS Div. 2 unavailable. (MT, PT) 13.'125 V DC Bus 11DA Failure Failure of DC bus' 11DA renders' EPS Div. 1 unavailable. (MT, PT)
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- 14. 125 V DC Bus 11DC Failure .i i
Failure of DC bus 11DC renders EPS Div. 3 unavailable. (MT, PT) i i I l l l 1 1 l . 'W 1 1 ? l i t a
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A-4 GRAND CULF NUCLEAR STATION, UNIT 1. PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN { Energency Electric Power System (EPS) TABLE Al-2 I6E ' INSPECTION PROCEDURES FOR SYSTEM 0PERATION PROCEDU RE FAILURE NUMBER TITLE . COMPONENTS MODES 62702 Maintenance (Refueling) DCs 11, 12, 13;. 1-14 62703 Monthly Maintenance 4160 V AC Busses; l Obs erva tion Battery Banks; l 72700 Startup Testing-Refueling 125 V DC Busses; Switchgear/ Battery Room Coolers 61725 Surveillance and Calibration DGs, Busses, 1-14
' P rog ram Batteries, Fan l
,61726 Monthly Surveillance Coolers j Observation R 71707 Operational Safety l Verification 71710 ESF System Walkdown l
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. A-5 GRAND CULF NUCLEAR STATION, UNIT 1 PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN Emergency Electric Power System (EPS)
TABLE Al-3 !!0DIFIED SYSTEM UALKDOWN Desired Act. Pow.Sup. Required Act. Description ID No Location- Position Pos. Breaker # Location Position Pos. ESF Div. I DG 11 P400/401 Operate Diesel Gen. Mode ESF Div. II DG 12 P400/401 Operate Diesel Cen. Mode 4 i
!!PCS Diesel DG 13 P400/401 Operate i Generator Mode 4160V AC Bus 16AB P-lDB1 Closed SUG Control 72-11011 i
Voltage ) Feed From 152-1601 Racked-ESF 11
, in Feed Fron 152-1611 Racked-ESF 12 in Feed From 152-1614 Racked-ESF 21 in l
4160V AC Bus 15AA P-IDAl Closed Swvr Control 72-IIAll VoTtage Feed From 152-1514 Racked-ESF 11 in Feed From 152-1511 Racked-ESF 12 in Feed From 152-1501 Racked-ESF 21 in 125V DC Bus llDA 11DA Closed Btry Otpt Bkr 72-IIA 01 125V DC Bus 11DB 11DB Closed Btry Otpt Bkr 72-llB01 Battery Bank 1 A3 Battery Bank 1B3 I
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A-6 TABLE Al-3 (Cont'd) Reference Doc,pnents ID No. Title ~ Rev. Date'- t
- 1. N/ A Normal' AC Power System R27/20 1 '
2/25/83
System Description
- 2. 04-1-01-L11 System Operating Instructions: 14. ' 3/.7/8 5 -'
~ Plant DC Systems 4
- 3. 04-1-01-R21 System Operating Instructions: 1; 12/9/86 4.16/6.9kV System
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A-7 I GRAND GULF NUCLEAR STATION, UNIT 1 ' PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN Emergency Electric Power System (EEPS) TABLE Al-4 PROPOSED INSPECTION PLAN FOR DIESEL GENERATORS AT NUCLEAR PLANTS A. Objective To review and evaluate Diesel Generator design operation, and mainte-nance at NPPs to ensure that the DGs will be available when needed to power safety systems. B. Details
- 1. The inspection of the following items should focus on DG auxiliary systems as follows: Fuel Injection System, Turbocharger, Starting System, Speed / Load Control, Jacket Water, Cooling Water, Lube Oil, Fuel Oil, Control and Monitoring Systems, and Generator.
- 2. Using the LER, 50.55e, and Part 21 systems computer printout, select 3 recent failures (within 2 years) for followup at the NPP. When at the plant select an additional 2 fdilures f rom the internal systems.
Evaluate the licensee's response to these failures for proper failure analysis, corrective action, notification of vendor, Part 21 evaluation and documentation.
- 3. Maintenance: Refer to IE I.P.s 62700 and 62702, as they apply to DG maintenance. Additionally, does the NPP have, and have they imple-mented the DG vendors' maintenance recommendations (especially those recommendations unique to nuclear service DGs such as Colt's d e-scribed in NSAC-79)? Are maintenance personnel specially trained on DGs? Is failure information fed back into maintenance program? Itas the NPP implemented recommendations of various studies referenced in Section 4.above.
- 4. Design Change Control: Select two DG modifications and verify prop-er implementation. Utilizing information from DG ' vendor inspection ou modifications recommended, verify that NPP is receiving all pertinent information in this area from the vendor. (Reference IE I.P. 37700).
- 5. Spare Parts and Procurement: Review how spare parts and services are purchased and parts stored, both f rom DG vendor and direct from sub-vendor. Verify adequate Part 21 and QA, particularly when vendors are only supplying commercial grade parts and services (e.g.,' Wood-ward Governor and Stewart and Stevenson). Verify ASME code specified where appropriate. Tour spare parts stor'gea area. (Reference IE 1.F. 38701B).
m - q , c . , - e s- ., ,.. l A-8 i TABLE Al-4 (Cont'd)
- 6. . Training:
Ensure appropriate DC specific training given to mainte-nance, operations, QA, and mananment personnel. Are there adequate l j
' documents to describe ' DG operation onsite (both main engine and aux-iliary system)? (Reference'IE I.P. 41700). 5
- 7. Observe DGs in operation. Ensure they run smoothly and are operated -
j per procedure.- Look for abnormal vibration and leaks (air, ~ fuel oil, or. lube oil). Check that readings are within specified limits. Are : limits per . DG vendor. recommendations? (Are recommendations clearly specified?) Is air quality in DG room satisfactory without excessive dust? Are control cabinets properly gasketed? Are instruments cali-brated? Is trending of operating data performed to detect degrada- ) tion early? l
- 8. Is NPP receiving all appropriate service information from vendor: '
design, maintenance, operational, etc? This is especially important for General Motors DC owners (verify they receive '" Power Pointers" )' from GM).
- 9. Review site practices to limit DG cold fast starts. d
- 10. Reliability records and calculations: Check logs, procedures, and 1 calculations versus Reg. Guide 1.108 criteria. I j
- 11. Ensure that pertinent studies on DG performance have been reviewed and recommendations implemented as appropriate (e.g., NUREC/CR-0660 l and NSAC-79).
- 12. Torq uing:
Ensure plant has adequate specifications for all torquing. 3 ! Ensure it is documented and done with calibrated equipment. Observe re-torquing if in progress. l 1 Source ' I J.C. Higgins and M. Subudhi, "A Review of Emergency Diesel Generator Performance at Nuclear Power Plants," NUREG/CR-4440, Brookhaven National Laboratory, November 1985. , References
- 1. NSAC-79, "A - Limited Performance Review of Fairbanks Morse and General s Motors Diesel Generators at Nuclear Plants," Nuclear Safety Analysis '
Center, Electric Power Research Institute, April 1984. 2. G. Boner and H. Hanners, " Enhancement of Onsite Emergency Diesel Generator' - Reliability," NUREG/CR-0660, University.of Dayton, February 1979. : j s ., e
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. A-ll GRAND GULF NUCLEAR STATION, UNIT 1 PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN liigh Pressure Core Spray (HPCS) System TABLE A2-1 IMPORTANCE BASIS AND FAILURE !! ODE IDENTIFICATION CONDITIONS TilAT CAN LEAD TO FAILURE General Guidance Surveillance of the licensee's periodic calibration, testing and/or pre-ventive or unscheduled maintenance activities, procedures and training and/or normal and emergency operating procedures, training and check-off lists in ac-cordance with the Technical Specifications and relevant NRC bulletins and in-formation notices should reduce the probability of f ailure for the conditions listed below. The most relevant aspects are designated for each condition as follows:
PC - Periodic calibration activities, procedures and training. PT - Periodic testing activities, procedures and training. MT - Preventive or unscheduled maintenance activities, procedures and training. OP - Normal and emergency operating procedures, check-off lists, train-ing, etc. Mission Success Criteria The liigh Pressure Core Spray (HPCS) System provides coolant to the reactor vessel at rated flow during accidents in which pressure remains high. HPCS consists of a single train with motor-operated valves and a motor-driven puup. Suction sion pool. is taken from either the Condensate Storage Tank (CST) or the suppres-Injec. tion to the reactor vessel is via a spray ring mounted inside the core shroud. HPCS is automatically initiated and controlled, lloweve r , operator intervention is required to_ throttle flow to prevent the 11PCS injec-tion valve from opening and closing in response to the reactor pressure level. The operator may also be required to manually start the system if an automatic start failure occurs.
- 1. Standby Service Water Failures:
- a. SSW Pump MDP2C Fails to Start or Run or Is Unavailable Due to Maintenance 1
SSW Train C is dedicated to llPCS. The train provides llPCS pump room cool-ing as well as llPCS diesel generator jacket cooling. IIPCS pump roomeooling failure is assumed to fail operating the llPCS pump in 12 hours. (MT, PT)
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TABLE A2-1 (Cont'd) CONDITIONS THAT.CAN. LEAD TO FAlLURE. l b. SSW Valve MVil -Fails to Open or Is Unavailable Due to Maintenance l Sec.ltem la above. Valve IN11 provides for return flow from the HPCS die-i sel generator and. pump room cooler. (MT, PT)
- 2. HPCS Pump Fails to Start or Continue to Run or Is Unavailable Due to Maintenance
.. Motor-driven pump MDP1 is the only pump in HPCS; its failure renders the I system unavailable. .(MT, PT)-
- 3. Hardware Failure in' Feed Line PS-8: Valve F004 Fails to Open or Is Out for Maintenance Motor-operated valve MV4 (F004) . fails to 'open to provide sufficient flow to the spray ring, or is unavailable due to maintenance. (MT, PT) i
- 4. Minimum Flow Valve F012 Fails to Open or Is Out for Maintenance The unavailabilityfof motor-operated valve MV12 '(F012) is assumed to fail the HPCS pump. (MT, PT)
- 5. Suppression Pool Suction Line Hardware Fault:-F015 Fails to Open Motor-operated valve MV15 (F015) is located in suction lineL PS-2 from the suppression pool. Failure of this line prevents flow f rom the suppression i pool to pump MDil. (MT, PT)
- 6. HPCS Pump Room Cooler Failure or Out for Maintenance-1 Unavailability ;of , the HPCS pump room cooler is assumed to fail operating l the HPCS pump in 12 hours. (MT, PT)
- 7. HPCS Actuation Logic / Circuit Failure The HPCS system is automatically initiated and controlled. All. motor-operated valves receive signals to realign on actuation. (MT,.PT)
- 8. Minimum Flow Valve (F012) Control Failure See Item 4 above. (MT, PT)
- 9. Common Mode Miscalibration of CST Level Instrument,s Miscalibration of CST level instrumentation would disable the automatic initiation of suppression pool suction normally based on receipt of'a low CST 1cvel signal. (PC).
___-_x_-____________-________ _ _ _ _
a ~. A .. TABLE A2-1 (Cont'd)- CONDITIONS THAT CAN LEAD TO FAILURE
- 10. ' Common Mode Miscalibration of Reactor Vessel Level Sensors The !!PCS system is automatically initiated on the receipt of a' low reactor water level (-42 inches or. Level 2) signal. (PC) 1
- 11. Operator Fails to Manually Open Valve F015'Given Auto Failure See Item 5 above (OP)
- 12. Operator Fails to Manually Initiate HPCS Given Auto Failure E
, The operator is required to manually start the system if an. automatic l start failure occurs. (OP) l l l l I l l , i l i
._____-__--_s
8 4 _1 A-14 ' GRAND CULF NUCLEAR STATION, UNIT 1 PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN 1-High Pressure Core Spray (IIPCS) System TABLE A2-21&E INSPECTION PROCEDURES FOR SYSTE!! OPERATION l PROCEDURE FAILURE ! NUMBER TITLE ' COMPONENTS. MODES 41700 Training HPCS Operation 11,12 41701 Requalification Training 52051 Instrument Components and CST Level Sensors, 9,10 Systems-Procedure Review Reactor Vessel 52053 Instrument Components and Level Sensors , Systems-Work Observation I 52055 Instrume.nt Components and Systems-Record Review 56700 Calibration 62702 Maintenance (Refueling) HPCS Pump, SSW 1-6 62703 Monthly Maintenance Pump C, MOVs, Observation llPCS Pump Room 72700 Startup Testing-Refueling Cooler. 61725 Surveillance and Calibration Sensors, Puups, 1-10' ! Frogram MOVs, Fan Coolers' ! 61726 Monthly Surveillance { Observation l l 71707 Operational Safety Pumps, MOVs. 1-8 Verification Controllers, Logic l 71710 ESF System Walkdown Circuitry
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l 1 A-15 l CRAND GULF NUCLEAR STATION, UNIT 1 PROBABILISTIC RISK' ASSESSMENT-BASED INSPECTION-PLAN I Iligh Pressure Core Spray (IIPCS) System i TABLE A2-3 MODIFIED SYSTEM WALKDOUN l l Desired Act. Pow.Sup. Required Act. Description ID No Location Position Pos. Breaker # Location Position Pos. j l HPCS Pump F015 CR Panel Sw: AUTO 52- 17B01 Closed Suction'from P601-16C Closed 170110 Supp. Pool I HPCS Minimum F012 CR Panel AUTO 52- 17B01 Closed . Flow to P601-16C Closed 170109 i Supp. Pool HPCS Inject. F004 CR Panel AUTO 52- 17B01 Closed Shutoff V1v. P601-16C Closed 170101 IIPCS Testable F005 CR Panel N/A Check V1v. P601-16C Closed l llPCS Pump C001 CR Panel AUTO 152- 17AC Racked-P601-16C 1702 In HPCS Pump Rm T51- 52- 17B01 Closed Cooler Fan B001C 170117 Reference Documents ID No. Title Rev. Date
- 1. 04-1-01-E22-1 Systen Operating Instruction: 22 1/8/87 f High Pressure Core Spray System Safety-Related
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I A-20 l CRAND CULF NUCLEAR STATION, UNIT 1 H PROBABILISTIC RISK ASSESSMENT-BASED I INSPECTION PLAN Reactor Core Isolation Cooling (RCIC) System i TABLE A3-1 IMPORTANCE BASIS AND FAILURE MODE IDENTIFICATION j
]
1 CONDITIONS THAT CAN LEAD TO FAILURE General Guidance Surveillance of the licensee's periodic calibration, testing and/or pre- j ventive or unscheduled maintenance activities,. procedures and training and/or I normal and emergency operating procedures.. training and check-off lists in ac-cordance with the Technical Specifications and relevant NRC bulletins and in-I formation notices should reduce the probability of failure for the conditions i listed below. The uost relevant aspects are designated'for each condition as follows: PC - Periodic calibration' activities, procedures and training. PT - Periodic testing activities, procedures and training. MT - Preventive or unscheduled maintenance activities,- procedures 'and - training. OP - Normal and emergency operating procedures, check-off lists, train-ind, etc. 1 Mission Success Criteria 1 l The Reactor Core Isolation Cooling (RCIC) System is tU provide coolant to' the reactor vessel at rated flow during accidents in which system pressure remains high. RCIC consists of a single train with motor-operated valves and i a turbine-driven pump. Suction is taken f rom either the Condensate. Storage - ' Tank (CST)-or the, suppression pool. The RCIC puup discharges the water to the
~
"B" Residual Heat Removal (RHR) system header which discharges to the "B" feedwater line. RCIC is automatically . initiated and controlled. However, operator intervention is required as follows: (a) to prevent either vessel overfill or continuous system trip / restart cycles, (b) to manually start the system given an auto-start failure, and (c) to set up the system for contin-uous operation when battery depletion is imminent.
- 1. Standby Service Water Failures:
l
- a. SSW Pump A Fails or Is Out for Maintenance Train A of the SSW system provides RCIC puup roou- cooling as well as ;
jacket cooling for DC'A. (MT, PT) ; i
- b. SSW Valve MVIA Fails to Open or Is Out for Maintenance Valve MVIA enables flow from the SSW Train A-pump to.the cooling-systems.
(MT , PT) 1 - - - - - - - _ - _ - _ _ - _ _ - - _ _ - _ _ _ _ - _ _ _ = _ - _
A-21 TABLE A3-1 (Cont'd) CONDITIONS THAT CAN LEAD TO FAILURE
- c. SSW Valve MV5A Fails 'to Open or Is Out for Maintenance
.)
Valve >NSA provides recirculation flow f rom SSW Train A to SSW basin A. (HT, PT) d
- 2. RCIC Pump Fails to Start or Is Out for Maintenance Turbine-driven pump TDP1 is the only pump,in RCIC; its failure renders the
' system unavailable.- (MT, PT)
- 3. High Reactor Water Level Signal Hardware Failure: Actuation Logic Failure, LT-N695A Failure, LT-N695B Failure RCIC is assumed to fail in a non-recoverable state if it fails to trip on -
high reactor water level since water would enter and damage the turbine. (MT, PT)
- 4. Motor-Operated Valve F013 Fails to Open or Is Out for Maintenance The failure of motor-operated valve MV13 (F013) to open or its unavailability due to maintenance disables flow from the RCIC pump to the-feedwater line. (MT, PT)
- 5. MOV F046 Fails to Open or Is Out for Maintenance Line PS-8 contains the lube oil cooler.
gpa) is diverted to this line from line PS-3. A Lube small amount of flow (16-25 oil cooling is required for turbine bearing cooling. (MT, PT)
- 6. MOV F019 Fails to Open or Is Out for Maintenance Failure of the minimum flow line to open during initial startup is assumed to fall the RCIC pump. (MT, PT)
- 7. Motor-Operated Valve F045 Fails to Open or Is Out for Maintenance <
6 Line PS-9 provides steam flow from the reactor to the RCIC turbine. " Failure of MV45 (F045) disables this flow. (MT, PT)
- 8. RCIC Pump Room Cooler Fails or Is Out for Maintenance Room cooling failure is assumed to fail the RCIC pump in 12 hours. (MT, PT)
- 9. Switchover to Suppression Pool Actuation Faildre The CST is the initial suction source for RCIC. Suction is automatically I 1
switched to the suppression pool on either low CST level or high suppression
f . . A-22 1 TABLE A3-1 (Cont'd) ' CONDITIONS TIIAT CAN LEAD TO FAILURE
.I pool level.
Line PS-2 is the suction line f rom the suppression pool.- The 1 ' line contains motor-operated valve MV31 which must be opened. (MT, PT)
- 10. Minimum Flow Valve' F019 Controller Failure See Item 6 above. (MT, PT)
- 11. RCIC Actuat, ion Logic Failure Failure of actuation logic disables the RCIC system. (MT, PT)
- 12. Operator F5ils to Manually initiat'e'RCIC The operator must manually start the system given an auto-start failure.
(OP) i
- 13. Operator Fails to Open Valve F031 i
i Motor-operated valve MV31 (F031) enables suction from the suppression ; pool. (0P) I
- 14. Common Mode Miscalibration of CST Level Instruments The CST is the initial suction source for RCIC. Suction is automatically switched to the suppression pool on either low CST level or high suppression-pool level. (PC)
- 15. Common Mode Miscalibration of Reactor Level Sensors-The RCIC system is automatically initiated on the receipt of.a low water level signal of -42 inches (Level 2). (PC) l 4
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- _ _ _ _ _ _ _ _ - _ _ - _ _ _ _ _ _ _ _ _-__________-_-_-______________-_____: ._ ._ .- s
pl * * , A-23 GRAND GULF NUCLEAR. STATION, UNIT 1 PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN Reactor Core Isolation Cooling (RCIC) System TABLE A3-2 I&E INSPECTION PROCEDURES FOR SYSTEM OPERATION PROCEDURE FAILURE { NUMBER TITLE COMPONENT S MODES 41700 Training RCIC Operation 12, 13 41701 Requalification Training 9 52051 Instrument Components and CST Level Sensors, 14, 15 Systems-Procedure Review Reactor Vessel- j 52053 Instrument Components and Level Sensors i Systems-Work Observation j 52055 Instrument Components and. ] Systems-Record Review 56700 Calibration
]
1 l' 62702 Maintenance (Refueling) RCIC Punp/ Turbine, 1,2,4-8 62703 Monthly Maintenance SSW Pump A, MOVs, Observation RCIC Pump Room I 72700 Startup Testing-Refueling Cooler l 61725 Surveillance and Calibration Sensors, Pumps, 1-11,14, Program MOVs, Fan Coolers 15 61726 Monthly Surveillance Observation 71707 Operational Safety Pumps, MOVs,' 1-11 Verification Controllers, Logic 71710 ESF System Walkdown Circuitry i
. A-24 CRAND GULF NUCLEAR. STATION, UNIT I PROBABILISTIC RISK ASSESSffENT-BASED INSPECTION PLAN Reactor Core Isolation Cooling (RCIC) System TABLE A3-3 !!0DIFIED SYSTEM WALKDOUN Desired Act. Pow.Sup. Require Act.
Description ID No. Location Position Pos. Breaker # Location Position Pos. RCIC Pump C001 RCIC Turbine 1 1 RCIC Inject F013 CR Panel Sw: AUTO 72- IDA2 Closed Shutoff Viv P601-21C Closed ll A53 RCIC Turb F015 Open ___ Oil Cooler PRV RCIC Water to F046 CR Panel AUTO 72- IDA2 Closed Turb Lube P601-21C Closed 11A60 011 Cooler i RCIC Min Flo F019 CR Panel AUTO 72- 1DA2 Closed to Supp Pool P601-21C Closed llA56 RCIC Min Flo F021 Aux Bldg Locked ; to Supp Pool Area 8 Open { Stopcheck. El 93 ft RCIC Stm Sply F045 CR Panel AUTO 72- IDAl Closed to RCIC Turb P601-21C Closed llA16 l RCIC Pump Sue F031 CR Panel AUTO 72- IDA2 Closed frm Supp Pool P601-21C Closed llA58 RCIC Rm Fan E51- 52- 15B11 Closed Coil Unit B006 151130 l 3 RCIC Turbine C002 l q
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Reference Documents ID No. Title Rev. Date
- 1. 04-1-01-C71 System Operating Instruction: 27 12/9/86 Reactor Core Isolation Cooling System Safety Related
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Reactor Protection-System (RPS), f l l
-TABLE'A4-1 GENERALIZED INSPECTION' PLAN !
{ Discussion The Grand Gulf PRA .does lnot ' model the Reactor Protection System. (RPS) .inL 1 any detail.' RPS electrical f ailure . and mechanical' failure on t demand . were 3 assigned values of 2.0E-5. and,1.0E-5, ' respectively, i.e.,' the system was1 simply-treated as a data v'alue. No dominant failure modes are. determined; JA generic inspection' plan _is.' adapted and discussed below. 1 System Description i The RPS s monitors critical parameters ;during all modes of ' reactor operation to protect'against conditions that could. threaten the fuel containment barriers and reactor coolant pressure boundary. RPS consists.of' sensors and logic'trainsL that monitor key plant operating parameters. When these. critical 1 parameters exceed preselected setpoints, . the RPS generates signals ; to ' automatically , shut-down. the < reactor by rapidly inserting. all control rods - (scram). The system is- ! divided into.two separate but redundant trip: systems with each; trip system fur - l ther divided into' two trip '_ logic divisions. . These.fourEdivisions or-. channels ' are identical in the parameters theyimonitor, which provides relia _bilitynin'sys-tem operation. The four channels are . termed. A, B,' C, and D (which correspond to divisions 1, 2, 3 and _4) with Channels A1 and C. comprising Trip System A', . and ! channels' B and D comprising Trip . System B. -For a scram to occur one' channel for-I Trip System A must trip, and' one channel f rom = Trip System B must trip concur-rently.o The RPS normally receives power f rom two high inertia motor-generator sets.
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Alternato power is available to either RPS bus. _An: interlock prevents parallel-. ! ing of 'a motor-generator with the alternate supply. ' The 480 V^ AC system sup-plies power to the RPS MG- sets and thel alternate power 1 supplies through trans-l formers IXY77 and:1XY78. If' power is lost to just one MG set while at power, a. l half scram occurs. The - alternate . power supplies are then. manually selected and , the scram reset. If power to both MG sets is lost, . a f ull scram occurs. ! Loss of power to the backup. supplies poses no problem during normal conditions since. I. the backup supplies are not normally used.' l t
.There are two scram valves and one pilot scram valve for each control, rod.
l The backup scram' valves are energized by the 125 V 'DC system,l a highly. reliable - i- source of power which is not considered interruptable. ;Eachl pilot scram _ valve i ' is' operated by' dual solenoids with the solenoids normally energized. 'The pilot j
- valve controls the air supply to the scram valves for each control rod. With <
-either pilot scram valve solenoid energized, air pressure holdsSthe scram valves closed. On a. loss f of instrument air, the scram l valves open causing a .' full scram.> Tripping 'only one of ' the - trip . systems deenergizes half _ of ' the . pilot -
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, A-28 TABLE A4-1 (Cont'd) scram valve solenoids (one for each control rod), but causes no control rod l j
motion. This action is.ternre.1 a half scram. Uhen both trip systems are trip-1 ped, air is vented from the scram valves.and lines up control rod drive water 1 under high pressure in the scram accumulators to the insert side of the drive pistons and all control rods' are rapidly inserted.- q
'l Inspection Areas 1
- 1. Review'and witness RPS function surveillance tests and preventive mainte-nonce.
4 Include witness of partial manual scram test, single rod scram,. tests i of individual RPS channels, and RPS circuit breaker and motor generator. set preventive maintenance. References include: R.G. 1.22, " Periodic Testing of Protection Systen ; Actuation Function," for RPS and RRCS; R. G . - 1.118, " Periodic Testing
- of Electric Power and protection Systems," which endorses IEEE Std
.333-1977, " Criteria for Periodic Testing of Nuclear Power Generating j Station Safety Systems," for RRCS only. j Detailed guidance for review of LPRM and APRH calibration is contained l
in IE Inspection Procedures 61703 and 61704.. l
- 2. Inspect sensing instrument racks for correct valve configuration, labelling, and separation.
f
- 3. Ensure no abnormal RPS alarms in the control roca, and. verify bypass i conditions are properly logged and justified.
- 4. Check RPS panels f or jumpers and lif ted leads. Documentation of same with the appropriate review and approval!is required.
- 5. Review post work testing of RPS maintenance tasks.
- 6. Review calibration records of RPS sensors and compare results to Grand. Gulf i
technical specifications. Observe trends. f 7. Review qualifications and training for technicians performing testing and/or maintenance on the system. ,
- 8. Review control rod drive mechanism maintenance inspection procedure andL i results. Ensure trending of detected wear is per. formed.
- 9. Review preventive maintenance practices for solenoid operated valves located in the instrument air header and at the HCU scram inlet and outlet valves.
Reference Docuuents ID No. Title Rev. Date-
- 1. N/A. MP&L-GGNS-Reactor Protection System- 1
- 2. 04-1-01-C71-1 11/24/82 !
System Operation Instruction: 19 12/17/86 I Reactor Protection System i
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CRAND GULF NUCLEAR STATION, UNIT 1 PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION. PLAN Standby Service Uater (SSW) Sjstem TABLE A5-1 IlfPORTANCE BASIS AND FAILURE. MODE . IDENTIFICATION ' CONDITIONS THAT CAN LEAD.TO FAILURE General Guidance Surveillance of the licensee's periodic calibration, testing and/or pre-ventive or unscheduled maintenance activities, procedures and training and/or normal and emergency operating procedures, training and check-off lists in ac-cordance with the Technical Specifications 'and relevant. NRC bulletins ar2d in-formation notices should reduce the probability of f ailure for the conditions listed below. The most relevant aspects - are designated for each condition as follows: PC - Periodic calibration activities, procedures and training. PT - Periodic testing activities, procedures and training. MT - Preventive or unscheduled maintenance activities, procedures and training. . OP - Normal and emergency operating procedures, check-off lists, train-ing, etc. Mission Success Criteria The Standby Service Water (SSW) Systea is designed' to provide heat removal from plant auxiliaries that require cooling water during an emergency shutdown ; j of the plant. SSW is made up' of three independent' trains: A, B, and C. E.ach train consists of a motor-driven pump, motor-operated valves 'and heat exchan-gers. Only those SSW loads which were identified by the PRA as important in mitigating accidents and, hence, have high risk significance are considered in .i this analysis. Train A is composed of the Residual Heat Removal (RHR) System Loop. A heat exchanger coolers, pump coolers and pump - room coolers; Low Pres-sure Core Spray (LPCS) and Reactor Core Isolation Cooling (RCIC) room coolers; i and Standby Diesel Generator 11 jacket cooler. Train B consists of RHR Loops B and C pump coolers and pump room coolers; R3R Loop B heat exchanger coolers;
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and Standby Diesel Generator 12 jacket cooler.. Train C is dedicated to .the ; High Pressure Core Spray (HPCS) System. Train C'contains the HPCS Diesel Gen- ! erator jacket coolers and the HPCS pump room cooler. SSW is automatically initiated and controlled; however,- operator intervention is required to uan- f 1 ually start the system given an auto-start failure. a SSU was modeled using separate fault trees for each of its functions; for examples LPCS pump. room cooling, ERRR Pump A cooling, etc. The failure modes listed below are culled from other system fault trees wherein SSW faults impact, a
k e .. A-32
. TABLE AS-1(Cont'd)
CONDITIONS THAT CAN LEAD TO FAILURE' .! i
- 1. SSW Valve MVIA/B (F001A/B) Fails to Open or is Out for Maintenance-Unavailability of motor-operated valve F001A or F001B, SSW Pump A/B
' discharge valve, renders SSW Train A/B coolers inoperative. (MT,.PT).
- 2. SSW Valve MVSA/B (F005A/B) Fails to Open or is Out for Maintenance Unavailability of motor-operatedEvalve'F005A or F005B, SSU Loop A/B return to cooling tower, renders SSW Train A/B coolers inoperative. (MT,'PT)
- 3. SSW Valve MVil (F011C) Fails to Open or is Out for Maintenance Unavailability of motor-operated valve F011C, SSW Loop C return to cooling tower, renders ,$SW Train C (HPCS) coolers inoperative. (MT, PT)
- 4. SSW Valve XV13 (F013) Is Plugged
'l Unavailability of manual valve F013, SSW Train C common injection valve, j renders SSW Train C (HPCS) coolers inoperative. (MT, PT) '
- 5. SSW Pump MDPl A/B Fails to Start' or Continue to Run or is Out for Main- !
'tenance '
Unavailability of motor-driven pump C001A or C001B renders SSW Train A/B coolers. inoperative. (MT, PT) l
- 6. SSW Pump MDP2C Fails to Start or Continue to Run or is Out for Mainte- .j
_n_o n c e Unavailability of motor-driven pump C002C renders SSW Train . C (HPCS) coolers inoperative. (MT, PT) l 7. SSW Valve XVl99A/B (F199A/B) is Plugged Failure of manually-operated valve F199A i or F199B, SSW Pump A/B notor bearing oil cooler inlet, due to plugging renders Train A/B unavailable. (MT, PT) j 8. SSW Valve MV14A/B (F014A/B) Falls to Open or Is Out for Maintenance
's Failure of motor-operated inlet valve F014A or F014B to open or unavail-ability'due to maintenance fails RHR A/B heat exchanger coolers. (MT, PT) 9.t SSW Valve MV68A/B (F068A/B) Fails to Open or In'Out for Maintenance Failure of motor-operated outlet valve F068A or F068B to open or.unavail-
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ability duc to maintenance fails RHR A/B heat exchanger coolers. (MT, PT) u4 + A%0bn _ _ _ - _ . . _ _ . _ . _ . - _ _ _ _ - _ - _ - _ - M
m; - A-33 TABLE AS-1-(Cont'd)- l CONDITIONS THAT CAN LEAD TO FAILURE l
- 10. SSW Valve XV37/38 (FG37/38) Is Plugged Fallure of manual valve. F037 or F038, LPCS pump room cooler inlet or outlet, renders tha..LPCS pump unavailable. (MT, PT)
- 11. SSW Valve XV105/106 (F105/106) Is Plugged Failure of manual valve F105 or F106, RCIC pump room cooler inlet or; l- outlet, renders the RCIC pump unavailable. (MT, PT) .
- 12. SSW Valve XV47/48 (F047/048) Is' Plugged Failure of manual valve F047 ' or F048 RHR C pump room cooler inlet or ,
outlet, renders the RHR C pump unavailable._ (MT, PT) ! J
- 13. SSW Valve XV51/52 (F051/052) Is Plugged .
]
Failure of manual valve F051 or F052, RHR C pump seal cooler inlet or outlet, renders the RHR C pump unavailable. (MT, PT) ,1 i
- 14. SSW Valve MV18A/B (F018A/B) Fails to Open q Failure of motor-operated valve F018A or F018B, inlet to Diesel Generator I
11/12 jacket cooler, renders the respective.DG unavailable. (MT, PT)' -j l
- 15. SSW Valve !!V23A/B (F023A/B) Fails to Open j l
Failure . of motor-operated valve F023A or F023B, outlet from Diesel l Generator 11/12 jacket cooler, renders the respective DG unavailable. (MT, l PT). 'I ' i i
- 16. DG 11/12 Jacket Cooler (HX4A/B) Plugged or Out for tialntenance Unavailability of jacket cooler for DO 11 or DG 12 renders the respective i DG inoperative. (MT, PT) l l
N 1 l r > __._.__.___m._m_ _ - - _ _ _ . _
A .3 4 ' GP,AND CULF Ni CLEAR STATION, UNIT 1 , PROBABILISTIC' RISK ASSESSMENT-BASED I INSPECTION PLAN Standby Service Water (SSW) System TABLE AS-2 I&E INSPECTION PROCEDURES FOR SYSTEM OPERATION
' PROCEDURE FAILURE NUMBER TITLE- COMPONENTS MODES 62702 Maintenance (Refueling) SSW Pumps A, B, C;. 'l-16 62703 Monthly Maintenance MOVs, Manual Valves, Observation DG Jacket Coolers 61725 Surveillance.and Calibration Program 61726 Monthly Surveillance Observation 71707 Operational' Safety Verification 71710 ESF System Walkdown 72700 Startup Testing-Refueling ,
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. .A-35 l
CRAND GULF NUCLEAR STATION, UNIT 1 PROBABILISTIC RISK ASSESSMENT-BASED l INSPECTION PLAN Standby Service Water (SSW) System L TABLE A5-3 MODIFIED SYSTEM WALKDOWN l i Desired Act. Pow.Sup. Reguired Act. Description ID No Location Position Pos. Breaker # Location Position Pos. SSW Inlet to F014A CR Panel Sw: AUTO 52- 15B31 Closed RiiR HX A P870-lC Closed ___ _ 153147 I SSW Outlet fm F068A CR Panel Sw: AUTO 52- 15B31 Closed RilR RX A PS70-1C Closed 153145 SSW Inlet to F014B CR Panel Sw AUTO 52- 1653.1 Closed RilR llX B PS70-7C Closed 163145 SSW Outlet fm F068B CR Panel Sw AUTO 52- 16U31 Closed RIIR flX B P370-7C Closed 163142 . SSW Pump A F001A P870-lc AUTO 52- 15B51- Closed f Disch Valve Closed 155107 SSW Loop A F005A P870-lC AUTO 52- 15B51 Closed Rtn to C1g Closed 155112 ( Tower .) SSW Pump B F001B P870-7C AUTO 52- 16B51 Closed Disch Valve Closed 165109 1 SSW Loop B F005B PS70-7C AUTO 52- 16B51 Closed Ren to C1g Closed 165112 Tower SSW Pump A C001A SSW Bs A AUTO 152- 15AA- Racked-. j CR P870 Off 1503 In . i SSW Pump B C001B SSW Bs B AUTO 152- 16AB Racked-CR P870 Off 1616 in l SSW Pump A F199A SSW Pump Open Motor Bearing flouse I 011 Clr In1 ' F
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_ _ _ _ _ -_ ._--_.-------u--------
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- 1. l
. A-36 TABLE A5-3 (Cont'd)
Desired Act. Pow.Sup. Reguired Act. Description ID No. Location Position Pos. Breaker # Location Position Pos. SSW Puap B F199B SSW Pump Open Mctor Benring House 011 Clr In1 i RHR C Roan F047 Aux Bldg Open i Clr Inlet Area 10 93 ft RHR C Room F048 Aux Bld5 Throt-Clr Outlet Area 10' tied
- 93 ft t
RHR Pump C F051 Aux Bldg Open Seal Clr Area 10 In1(Loop D) 93 ft RHR Pump C F052 Aux Bldg Throt-Seal Clr Area 10 tied
- Out1(Loop B) 93 ft HPCS SSW C002C SSW Bs A AUTO 52- 17D01 Closed Pump (Loop C) CR P870 Off 170124 l
SSW Loop C F011C P870-5C AUTO 52- 17B11 Closed Rtn to C1g Closed 171106 Twr A i i RCIC Room F105 Aux Bldg Open Clr In1 (A) RCIC Rm 108 ft RCIC Room F106 Aux Bldg Open Clr Otit (A) RCIC Rm 108 ft LPCS Room F038 Aux Bldg Open Clr Ini'(A) Area 9 l 93 ft
*The position of throttled valves may be investigated by checking the proce-dures that determine proper valve position, and the method used.to verify that the valve is maintained in that position.
t
. A-37 TABLE AS-3 (Cont'd)
Desired Act. Pow.Sup. Required Act. Description 1D No. Location Position Pos. Breaker # Location Position Pos. LPCS Room F037 Aux Bldg Throt-Clr'Otlt (A) Area 9 tied
- 93 ft HPCS SSW Pump F013 SSW Locked Discharge Pump Open (Loop C) House SSW Inlet to F018A CR AUTO 52- 15B11 Closed DG11 Wtr Clr P870-1C Closed 151115 1
SSW Inlet to F018B CR AUTO 52- 16B11 Closed DC12 Utr Clr P870-7C Closed 161108 DGil Cooler F023A Aux Bldg Throt-Outlet Area 12 tied
- 133 ft DG12 Cooler F023B Aux Bldg Throt-Outlet Area 12 tied
- 133 ft DG11 Jacket ilX4A Water Cooler DC12 Jacket HX4B Water Cooler
- The position of throttled valves may be investigated by checking the proce-dures that deterdine proper valve position, and the method used to verify that-the valve is maintained in that position.
Reference Documents ID No. Title Rev. Date
- 1. N/A Standby Service Water System - 1 N/A P41 System Description
, 2. 04-1-01-P41-l' System Operating Instruction: 28 12/9/86 Standby Service Water System l
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,l A-48 GRAND GULF NUCLEAR STATION, UNIT 1 PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN I
Standby Liquid Control (SLC) System j 1 TABLE A6-1 IMPORTANCE BASIS AND FAILURE MODE IDENTIFICATION I l CONDITIONS THAT CAN LEAD TO FAILURE General Guidance lj brveillance of the licensee's periodic calibration, testing and/or pre-vencive or unscheduled maintenance activities, procedures and training and/or i normal and emergency operating procedures, training and check-off lists in ac- l cordance with the Technical Specifications and relevant NRC bulletins:and in- 1 fornation notices should reduce the - probability of f ailure for the conditions { listed below. The most relevant aspects are designated for each condition as j follows l PC - Periodic calibration activities, procedures and training. PT - Periodic testing activities, procedures and training. MT - Preventive or unscheduled maintenance activities, procedures and training.
.)
OP - Normal and emergency operating procedures, check-off lists, { l training, etc. Mission Success Criteria i i The Standby Liquid Control (SLC) System provides a backup method, redun- I dant but independent of control rods, to establish and maintain the reactor ) suberitical. The system takes suction from the SLC tank containing sodiun pentaborate in solution with demineralized water and ' injects through one of ] two parallel, positive displacement puups, motor-operated valves and one of l j two explosive val'ves into the the reactor vessel lower plenum. The pump suc- J tions are cross-tied to ensure suction to both pumps if one of the suction ! valves fails closed. Two parallel explosive valves are downstream of the pump discharges. A cross-tie line upstream of the explosive valves is present to ! ensure that flow f rom either pump can be discharged through either explosive - valve. Downstream of the explosive valves, the system combines in a comnon discharge line.
- 1. Failure to Restore Two Valves in Discharge Test Line after Test i
l The system has a common test return line. This piping originates at the pump outlet cross-tie line. If this line is not isolated following a test, pump discharge in the event of system actuation would pref erentially flow to the test tank. The test line is isolated by manual closure of valves XV16 (F016), XV17 (F017). (MT, PT) l l
s, . j i A-49 I TABLE A6-1 (Cont'd). CONDITIONS THAT CAN LEAD TO FAILURE
- 2. Failure to Reclose Manual Test Valve F031 after Suction Test
-l Flow from the test tank to the motor-driven pump (s) is through manually operated valve XV31 (F031 - normally locked closed). (MT, PT)
- 3. One of Two Check Valves in Injection Line Fails .
J Common discharge line PS1 located downstream of the explosive valves discharges into the reactor vessel lower plenum. ' PSI incorporates in series motor-operated testable check valve TCV6 (F006) and check valve CV7 (F007). Line also contains locked-open manual valve XV8 (F008). (MT, PT) i l
- 4. Operator Fails to Initiate SLC f
I The operator manually actuates SLC with two keylock switches (one for each train) on control room panel 1H13-P601. (0P) ~ l
- 5. Pump Suction Inlet Valves F001A and F001B Fail to Open'or Are Out For l Maintenance Motor-operated inlet valves FNlA and MV1B are closed when in the standby.
mode. (MT, PT) l
- 6. Motor-operated Valves G33-F001 and G33-F004 Fail to Close I j
These valves provide for isolation of the Reactor Water Cleanup (RWCU) l System. If they are open, boron concentration in the reactor may not be sufficient for reactivity control. (MT, PT)
- 7. Motor-operated Valves G33-F250 and G33-F251 Fail to Close See Item 6 above. (MT, PT) 1
- 8. SLC Train A and B Pumps Fail to Start or' Continue to Run or Are Out For '
Maintenance This event requi. a. unavailability in each train of motor-driven pumps MDPA and MDPB. (MT, PT)
, A-50' GRAND GULF NUCLEAR STATION', UNIT 1 PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN Standby Liquid Control _ (SLC) System -
TABLE.A6-2 I&E INSPECTION PROCEDURES FOR SYSTEM OPERATION PROCEDURE FAILURE-NUMBER TITLE COMPONENTS MODES 41700 Training SLC Operation ~ 4 41701 Requalification Training 62702 Maintenance (Refueling) SLC Pumps, MOVs, 1-3, 5-8 62703 Monthly Maintenance Check Valves Observation 72700 Startup Testing-Refueling 61725 Surveillance and Calibration SLC Pumps, MOVs,- 1-3, 5-8 Program Check Valves-61726 Monthly' Surveillance Observation 71707 Operational Safety Verification 71710 ESF System Walkdown l l 1 I l
]
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j
- A-51 GRAND CULF NUCLEAR STATION, UNIT 1 PROBABILISTIC RISK ASSESSMENT-BASED
-{
INSPECTION PLAN 1 j Standby Liquid Control (SLC) System TABLE A6-3 MODIFIED SYSTEM WALKDOWN 1 Desired Act. Pow.Sup. Reguired Act. Description ID No. Location Position Pos. Breaker # Location Position Pos. l
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SLC Recirc to F016 Contain Closed l Test Tank Iso 184 ft L SLC Recirc to F017 Contain Closed 2 Test Tank Iso 184 ft l SLC Test Tank F031 Contain Locked Outlet 184 ft Closed SLC Outboard- F006 Contain Locked Stopcheck 161 ft Open llandwhe el \,1 SLC Inboard F007 Drywell Locked-Stopcheck 100 ft Open Handwheel i SLC Inboard F008 Drywell Locked-J Isolation i 100 ft Open SLC Storage F001A CR Panel Closed 52- 15B21 Closed Tank Outlet H22-P011 152115 SLC Storage F001B CR Panel Closed 52- 16B31 Closed I Tank Outlet H22-P011 163135 SLC Pump A C001A Contain Norual 52- 15B21 Closed 185 ft 152114 P601-19B SLC Pump B C001B Contain Norm'al 52- 15B21 Closed 185 ft 163129 P601-18B Reference Docuraents l ID No. Title- Rev. Date 1-04-1-01-C41-1 System Operating Instruction: 21 12/16/86-Standby Liquid Control System Safety Related l l C
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, B-1 GRAND CULF NUCLEAR STATION, UNIT 1 i PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN i
Automatic Depressurization System (ADS) ]
'1 i
TABLC B1-1 ItiPORTANCE BASIS AND FAILURE MODE IDENTIFICATION CONDITIONS THAT CAN LEAD TO FAILURE General Guidance Surveillance of the licensee's periodic calibration, testing and/or. pre-ventive or unscheduled maintenance activities, procedures and training and/or i normal and emergency operating procedures, training and check-off lists in ac- { cordance with the Technical Specifications and relevant NRC bulletins and in-formation notices should reduce the' probability of failure for the conditions listed below. The most relevant aspects are designated for each condition as follows: l l PC.- Periodic calibration activities, procedures and training. l PT - Periodic testing activities, procedures and' training. l MT - Preventive or unscheduled maintenance activities, procedures and training. , OP - Normal and emergency operating procedures, check-off lists, l training, etc. ! Mission Success Criteria The Automatic Depressurization System (ADS) depressurizes the reactor ves-set to a pressure at which the low pressure injection systems can inject cool- l ant to the reactor vessel. The system consists of eight relief valves which ' are capable of being manually opened. The' system is automatically initiated. Three of eight valves opening to depressurize the reactor vessel constitutes system success.
- 1. Common Mode ADS Valve Failure The common mode failure cust f ail six of the eight ADS relief valve's to fail the system. All other safety relief valves are still available as a pos-sible blowdown path. (MT, PT)
'l 1
.- ., . q B-2 0 1 GRAND GULF NUCLEAR STATION, UNIT 1 PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN Automatic Depressurization System (ADS)- TABLE B1-2 I&E INSPECTION PROCEDURES FdR SYSTEM OPERATION l l PROCEDURE FAILURE NUl!BER TITLE COMPONENTS MODES ! i 41700 Training ADS Valve Operation 41701 Requalification Training 62702 Maintenance (Refueling) ' ADS Valves, 1 62703 Monthly Maintenance Actuation Logic i Observation ; 61725 Surveillance and Calibration I Program 61726 Monthly Surveillance Observation 71707 Operational Safety I Verification 71710 ESF System Walkdown 72700 Startup Testing-Refueling i 1
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B-3 GRAND GULF NUCLEAR STATION, UNIT 1 PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN Automatic Depressurization System (ADS) TABLE B1-3 MODIFIED SYSTEM WALKDOWN Desired Act.' Pow.Sup. Reguired Act. , Description ID No. Location Position Pos. Breaker # Location Position Pos. t MSL B.SRV. F041K P601 SW: Auto (ADS)' _ . _ . _ .. . . . Closet MSL B SRV F041F P601 Auto (ADS) Closed MSL B SRV F051B P601 Auto (ADS) Closed l MSL D SRV F041D P601 Auto ' (ADS) Closed MSL A SRV F047A P601 Auto (ADS) Closed MSL A SRV F051A P601 Auto Closed (ADS) 1 MSL C SRV F047L P601 Auto (ADS) Closed MSL C SRV F051C P601 Auto (ADS) Closed Reference Documents ID No. Title Rev.- Date t
- 1. 04-1-01-B21-1 Systera Operating Instruction 25 12/12/86 Nuclear Boiler Systen Safety Related i
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. i CRAND CULF. NUCLEAR STATION, UNIT 1 PROBABILISTIC RISK ASSESSMENT-BASED i INSPECTION PLAN Condensate System (CDS)
TABLE B2-1 IMPORTAMCE~ BASIS AND FAILURE MODE IDENTIFICATION CONDITIONS THAT CAN LEAD TO FAILURE - 1 General Guidance Surveillance of the licensee's periodic calibration, testing and/or' pre- ! ventive or unscheduled maintenance activities, procedures and training and/or ] normal and emergency operating procedures, training and check-off' lists in ac- 'l cordance with the Technical Specifications and relevant NRC bulletins and in- k formation notices should reduce the probability of failure for the conditions j listed below. The most relevant aspects are designated for each condition as follows: {
.j l
PC - Phriodic calibration activities, procedures and training. j PT - Periodic testing activities, procedures and training. MT - Preventive or unscheduled maintenance activities, procedures and l j training.
]
OP - Normal and emergency operating procedures, check-off lists, training, etc. > Mission Success Criteria The Condensate System (CDS) is utilized as a backup lou pressure injec-tion system. The system consists of three main condenser units, three conden-sate pumps, three condensate booster pumps, three strings of four low pressure heaters, a condensate drain tank, and associated valves, piping, instruments - tion and controls.to supply reactor' feed pumps with heated feedwater at the necessary net positive suction head. The condensate system also supplies water to the reactor vessel during. low pressure conditions such as startups,- shutdowns, and, in emergency situations, through feedwater startup valve AV513. Flow from any of the six condensate or condensate booster pumps will result in successful reactor vessel cooling.
- 1. Operator Fails to Align Valve AV513 for Reactor Vessel Injection Reactor vessel supply is through air-operated valve AV513 (feedwater startup valve). (0P)
- 2. Operator Fails to Restore Instrument Air Instrument air is isolated by either a LOCA signal or a containment isolation signal. The operator must restore instrument air in order to supply air to the condenser makeup valve and also to open the feedwater startup valve. (0P)
,. . .- 3 I
B-7
~
Table B2-1 (Cont'd) CONDITIONS THAT CAN LEAD TO FAILURE
- 3. Loss of Offsite Power Condensate pumps are powered by non-safety 4.16 kV busses. Loss of I offsite power fails these busses. I l
i
- 4. Failure 'o f Turbine Building Cooling- Water (TBCW) System with ]
Offsite Power Available l l Failure of TBCU cause's loss of cooling .to air compressors which f ails. instrument air. (MT, PT) i l l l
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.I GRAND GULF NUCLEAR STATION, UNIT 1 i PROBABILISTIC RISK ASSESSMENT-BASED i INSPECTION PLAN I Condensate System (CDS) l TABLE B2-2 I&E INSPECTION PROCEDURES FOR SYSTEM OPERATION
]
i PROCEDURE i FAILURE. j NUMBER TITLE COMPONENTS MODES J 41700 Training- .AV513, 1,2 i 41701 -{ Requalification Training IAS Operation ) 6 62702 Maintenance (Refueling) AV513,.Non-safety 1,3,4 62703 Monthly Maintenance '4160 V AC Busses, Observation TBCW 1 l 61725 Surveillance and Calibration i Program i 61726 Monthly Surveillance i Observation 71707 Operational Safety If l Verification 71710 ESF System Walkdown 72700 Startup Testing-Refueling i I; l t l l l l l l 1
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[ ... l l t GRAND GULF NUCLEAR STATION, UNIT 1 1 PROBABILISTIC RISK ASSESSMENT-BASED INSPSCIION PLAN Condensate System (CDS) l TABLE B2-3 MODIFIED SYSTEM WALKDOWN The Condensate System is lined 'up correctly for : normal operation. The i dominant f ailure mode for the CDS, as a backup to LPCI, is failure of the feedwater startup flow control valve to operate as required, and this failure j mode is addressed via Table B2-2 procedures, j l 1 J j 1 1 i a J i e ____-_-___-_A
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q E-J1 l { GRAND GULF NUCLEAR STATION, UNIT 1 PROBA'BILISTIC RISK ASSESSMENT-BASED ' INSPECTION PLAN Containment Venting System (CVS) TABLE B3-1 IMPORTANCE BASIS AND FAILURE MODE IDENTIFICATION CONDITIONS THAT CAN LEAD TO FAILURE Ceneral Guidance ! Surveillance of the licensee's periodic calibration, testing and/or pre-ventive or unscheduled maintenance activities, procedures and training and/or normal and emergency operating procedures,. training and check-off lists in ac-cordance with the Technical Specifications and relevant NRC bulletins and in-formation notices should reduce the probability of failure for the conditions listed below. The most relevant aspects are designated for each condition as' ' follows: PC - Periodic calibration activities, procedures and training. i PT - Periodic testing activities, procedures and training. I MT - Preventive or unscheduled maintenance activities, procedures and training. f OP - Normal and emergency operating procedures, check-off lists, , training, etc. Mission Success Criteria The Containment venting System (CVS) is used to prevent the primary con-tainment pressure limit from being exceeded when suppression pool cooling and ! containment sprays have failed to reduce primary containment pressure. The _j vent path used is a 20 inch diameter purge exhaust line which is part of the containment ventilation and filtration system. This line includes four 1 air-operated dampers which are normally closed. All four fail closed on loss of air. Two of the dampers are closed by a containment isolation signal. The other two are closed by standby gas treatment system initiation. The CVS , discharges to the roof of the auxiliary building.
- 1. Operator Fails to Vent Containment i
i The venting procedure requires containment venting when the pressure ; exceeds 17.25 psig. Venting requires that the operator jumper the isolation ' relays for each damper and then open them. (0P) !
- 2. _ Operator Fails to Restore Instrument Air i
Instrument air is. disconnected by a containment isolation signal or a LOCA signal (high drywell signal). The operator must restore instrument air to successfully vent. (0P) l
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i! B-12 i
-TABLE B3-1 (Cont'd)-
CONDITIONS THAT CAN LEAD.TO. FAILURE
- 3. Vent Path Hardware Failure: 1 of 4 Dampers Fails to Open or Remain Open- ,
The four CVS dampers are aligned in series, f ailure of one damper fails CVS. (MT, PT) l l l l l U-_- --_.----.-_-._-_._._-___-.__l__-____-..-.-__., _ _ - _ _ _ . _ . _ _ . _ _ . _ _ _ _ _ _ _ . = _ _ . _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ . _ _ _ _ _ . _ _ _ __ _ _ _ _ _ _ _ ._ __.
B-13 GRAND GULF NUCLEAR STATION, UNIT 1 PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN Containment Venting Systen (CVS) TABLE B3-2 I&E INSPECTION PROCEDURES FOR SYSTEM OPERATION PROCEDURE FAILURE NUMBER TITLE COMPONENTS MODES 41700 Training CVS Operation, 1, 2 l 41701 Requalification Training. IAS Operation i 62702- Maintenance (Refueling) CVS Dampers 3 62703 Monthly Maintenance Observation 61725 Surveillance and Calibration Program 61726 Monthly Surveillance Observation 71707 Operational Safety Verification 71710 ESF System Walkdown 72700 Startup Testing-Refueling l l l l
l B-14 1
) . .
CRAND GULF NUCLEAR STATION, UNIT 1 PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN Containment Venting System (CVS) ) TABLE B3-3 MODIFIED SYSTEM WALKDOWN j I Desired Act. Pow.Sup. Required .Act. 1 Description ID No. Location Position Pos. Breaker # Location Position Pos. ! 1 CVS Air-oper CR Panel l Dampers AV-F034 P870-90 Closed AV-F035 P870-3C Closed AV-F036 P870-2C Closed i AV-F037 P870-8C Closed l l l l l l 1 I I i l l l l l
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B-16 CRAND GULF NUCLEAR STATION, UNIT 1 PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN Control Rod Drive (CRD)' System Table B4-1 Importance Basis and Failure Mode Identification CONDITIONS THAT CAN LEAD TO FAILURE -l t I Ceneral Cuidance ? 1 Surveillance of the licensee's periodic calibration, testing and/or pre- [ ventive or unscheduled maintenance activities, procedures and training and/or . j normal and emergency operating procedures, training and check-off lists in ac- ! cordance with the Technical Specifications and. relevant NRC' bulletins and in-formation notices should reduce the probability of failure for the conditions ] j listed below. The most relevant aspects are designated f or each condition as 1 follows: J l PC - Periodic calibration activities, procedures and training. PT - Periodic testing activities, procedures and training. MT - Preventive or unscheduled maintenance activities, procedures and training. OP - Normal and emergency operating procedures, check-off lists, 1 training, etc. 1 Mission Success Criteria ! The Control Rod Drive (CRD) Hydraulic System is designed to reposition j control rods in the reactor core. On command, the CRD system rapidly shuts down the reactor in emergency situations by simultaneously inserting all the control rods into the core. For purposes of the PRA analysis, the CRD system is used to provide a backup source of high pressure injection cooling in lieu l of HPCS / RCIC. The flow available f rom CRD is only ef fective against very small leaks. The CRD system consists of two motor-driven pumps, i motor-operated valves and filters. The CRD pumps take suction ' f rom the condenser hotwell makeup /rej ect line. Makeup to the condenser hotwell is provided by the CST. Excess condensate f rom the condenser is rejected to the CST by the condensate system. From the condenser hotwell makeup / reject lines, water flows to the CRD pumps through a pump backwash suction filter and one of two pump suction filters. To provide sufficient cooling flow both pumps must be running and the dydraulic Control Units (HCU) cooling header ~ discharge path must be available.
- 1. LOCA Signal Is Present for More Than One Hour The CRD pumps are cooled by the Component' Cooling Water (CCW) system. The CCW systen normally transfers heat via heat exchangers to the Plant Service Water (PSW) systen. Portions of the PSW system (including cooling of the CCW heat exchangers) are interf aced with the Standby Service Water (SSW) system-l
{
-l B-17 Table B4-1 (Cont'd)
CONDITIONS THAT CAN LEAD TO FAILURE for cooling when loss of offsite power occurs. Cooling ~ f rom both the PSW and SSW is inhibited if a Loss of Coolant Accident (LOCA) is present.
- 2. Containment Isolation Signal Present and Operator Fails to Restart the IAS Instrument air to the containment is disconnected by a containment isolation signal. The Instrument Air System (IAS) is required for the operation of the flow control valves. Loss of instrument air results in closure of the flow path to the HCU cooling headers. .(OP)'
- 3. Operator Fails to Initiate Standby CRD Pump l Normally one CRD pump is running with the suction and discharge valves to the standby pump being open. The CRD pumps receive no automatic initiation signals. (0P)
- 4. Failure to Restore Manual Valve F0217B/A Following Pump Maintenance Manual valve XV217B/A (F0217B/A) is located at the standby pump outlet.
Testing of the pump following maintenance would not require reopening of this valve since the pump minimum flow line is upstream of the valve. (MT) h
- 5. Standby Pump C001B/A Fails to Start or Continue to Run l l
Both pumps are required for sufficient cooling flow. (MT, PT)
- 6. Normally Operating CRD Pump C001A/B Fails to Continue to Run Both pumps are required for sufficient cooling flow. (HT, PT) l 1
l l l 1 l l l 3 ' __ n
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.i B-18 f GRAND GULF NUCLEAR STATION, UNIT 1 -
PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN i Control Rod Drive (CRD) System TABLE B4-2 I&E INSPECTION PROCEDURES FOR SYSTEM OPERATION i l PROCEDURE FAILURE NUMBER TITLE { COMPONENTS MODES j j l 41700 Training CRD Pumps, IAS 2, 3 41701 Requalification Training Realignment 62702 Maintenance (Refueling) CRD Pumps, Manual .4-6 62703 Monthly Maintenance Valves ! Observation j 72700 Startup Testing-Refueling ' 61725 Surveillance and Calibration Pumps 5, 6 I i Program 61726 Monthly Surveillance Observation 71707 Operational. Safety Pumps, Manual 4-6 Verification Valves 71710 ESF System Walkdown l l
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- i
,, B-39 j I
GRAND GULF NUCLEAR STATION, UNIT 1 PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN Control Rod Drive (CRD) System TABLE B4-3 MODIFIED SYSTEM WALKDOWN Desired Act. Pow.Sup. Required Act. i Description ID No. Locaf. ion Position Pos. Breaker # Location Position Pos. I l CRD Water C001A CR Panel 152- IAS Racked- - Pump'A P601 1505 In CRD Water C001B CR Panel 152- 1A6 Racked-Pump B P601 1605 In ! CRD Water ! F217A Aux Bldg Open Pump A 1 Area 10 j Discharge Vlv El 93 ft 1 CRD Water F217B Aux Bldg Opon Pump B Area 10 Discharge Viv 'i El 93 ft
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I J Reference Documents
}
ID No. Title Rev. Date I i
- 1. 04-1-01-C11-1 System Operating Instructions- 94 12/23/86 i Control Rod Drive Hydraulic System I l Safty Related l
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.B-21 GRAND GULF NUCLEAR STATION, UNIT 1 PROBABILISTIC. RISK ASSESSMENT-BASED INSPECTION. PLAN ,
~ Low Pressure Core Spray-(LPCS) System TABLEB5-1-IMPORTANCEBASISANDFAILURE-MODE' IDENTIFICATION
' CONDITIONS THAT CAN LEAD TO FAILURE General. Guidance Surveillanceof[thelicenseesperiodiccalibration,-testing'and/orpre-ventive or unscheduled maintenance 1 activities, procedures'and training and/or normal and emergency operating procedures, training:and. check-off lists.in ac--
cordance with the Technical Specifications and relevant NRC bulletins andlin-formation notices should reduce the. probability of failure for'the conditions-listed ~below. 'The most relevant aspects areLdesignated for each condition sa
'follows:
PC -' Periodic calibration activities, procedures and training. PT.- Periodic testing activities, procedures and training. MT - Preventive or unscheduled maintenance activities, procedures and training.. . l OP - Normal.and emergency operating procedures, check-offilists, , training, etc. l Mission Success Criteria 1 The purpose of the Low Pressure Core Spray (LPCS)l System'is tofprovide coolant to the reactor vessel during accidents in which vessel' pressure'isc j l low. ADS can be used in conjunction with LPCS to attain aflow enough' system . { L pressure for injection to occur.. 'LPCS is a single train. system consisting ofE j l motor-operated and manual valves and a motor-driven pump.3 'The LPCS pump l takes water from Lthe suppression' pool through' strainers located 10- feet above the suppression pool floor. LPCS is automatically initiated and controlled. The-l operator may be required to manually' start the system if an: automatic actua-tion failure occurs.
- 1. LPCS Pump MDP1 Fails to Start or Continue to Run Due to Hardware Faults or Is Out for Maintenance LPCS is a one-train system containing a single flowLpath' to the reactor 1
ve s s el.'. Failure of the single pump fails the system. (MT, PT)
- 2. Standby Service Water Failures:
a) SSW Pump MDPI A Fails Due to -Hardware Fa.ults or Is' 0at for Mainte' ~ - i L -nance < t
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)
l _._______ni
q , B-22' TABLE B5-1 (Cont'd) CONDITIONS THAT CAN LEAD TO FAILURE
.i Failure of flow in SSW Train A fails LPCS pump room cooling.' Failure of pump room cooling is assumed to fail the LPCS pump in four hours. (MT, PT) b) SSW Valve MV5A Fails to Open or Is Unavailable Due to Maintenance See Item 2a above. (MT, PT)
{ c) SSW Valve MVIA Fails to Open or Is Unavailable Due to Maintenance. q See Item 2a above. (MT, PT)
- 3. Motor-operated Valve MVS Fails to Open -,
LPCS is a one-train system containing a single flow path to the reactor vessel. Failure of any of the three valves in this segment fails. the system. (MT, PT)
- 4. Minimum Flow Valve MV11 Fails to Close due to Hardware Faults The minimum flow for the pump is considered as a potential diversion path because of its size. Flow would be diverted to the suppression pool. (MT, PT) .
- 5. LPCS Pump Room Cooler Fails or Is Out for Maintenance Failure of room cooling is assumed to fail the LPCS pump in four hours.
(MT, PT)
- 6. Minimum Flow Valve Actuation Circuitry Fails due to Hardware ;
Faults I i See Item 4 above. (MT, PT) 1'
- 7. Automatic LPCS Actuation Circuitry Hardware Failure LFCS is automatically initiated and controlled. (MT, PT)
- 8. Operator Fails to Manually Close MV11 See Item 4 above. (0P)
- 9. Operator Fails to Manually Actuate LPCS I
The operator may be required to manually start the system if an automatic actuation failure occurs. (OP)
B-23' TABLE B5-1 (Cont'd) CONDITIONS THAT CAN LEAD TO FAILURE
- 10. Test Line Valve MV12 Out for Maintenance Upon receipt of a LPCS injection ' signal the test return valve MV12 is demanded to close. The test line represents a flow diversion path to the suppression pool. (MT) 3 i
- 11. LPCS Suction Valve MV1 Out for Maintenance There is only one suction line from the suppression pool to the LPCS pump. Alternate suction from the spent fuel pool must be manually valved in and, therefore, is not included in the PRA model. (MT)
.i l
l
B-24 GRAND GULF NUCLEAR STATION, UNIT 1 PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN Low Pressure Core Spray (LPCS) System TABLE B5-2 I&E INSPECTION PROCEDURES FOR SYSTEM OPERATION j l PROCEDURE FAILURE NUMBER TITLE
)
COMPONENTS MODES i l 41700 Training LPCS Pump, Valves 41701 8, 9 ) Requalification Training l 1 52051 Instrument Components and LPCS Actuation 6, 7 j Systems-Procedure Review
~
Circuitry. 52053 Instrument components and Systems-Work Observation 52055 Instrument Components and Systems-Record Review 56700 Calibration 62702 Maintenance (Refueling) LPCS Pump, SSW 1-5, 10, j 62703 Monthly Maintenance . Pump A, MOVs, LPCS- 11 j Observation Pump Room Cooler . I 72700 Startup Testing-Refueling )
- 61725 Surveillance and Calibration Pumps, MOVs, Fan. 1-5, 10 1
Program Coolers. , 11 l 61726 Monthly Surveillance
]
Observation l 71707 Operational Safety Pumps, MOVs, 1-7, 10, 1 Verification Controllers, Logic 11 ! 71710 ESF System Walkdown Circuitry J
j gj I
, , B-25 i
GRAND GULF NUCLEAR STATION, UNIT 1 i PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN , i Low Pressure Core Spray (LPCS) Systen l TABLE B5-3 MODIFIED SYSTEM WALKDOWN 1 Desired Act. Pow.Sup. Required Act. l Description ID No. Location Position Pos. Breaker # Location Position Pos. . 1 LPCS Inject F007 Drywell Locked - Isolation Vlv Open
]
LPCS Testable F006 CR Panel SW N/A Check Valve P601/H13 V/V !' Closed LPCS Inject F005 CR Panel AUTO 52- 15B11 Closed Shutoff Valve P601/H13 Closed 151114 j LPCS Min Flo F011 CR Panel AUTO 52- 15B11 Closed to Supp Pool P601/H13 Open 151134 LPCS Test Ret F012 52-CR Panel AUTO 15B11 Closed to Supp Pool P601/H13 Closed 151113 LPCS Supp F001 CR Panel Open 52- 15B11 Closed Pool Suction P601/H13 Open 151109 1 1 LPCS Puup C001 CR Panel AUTO 152- 15AA Racked P601-21C 1506 In LPCS Pump Rm T51- CR Panel AUTO 52- 15B41 Closed ! Fan Coil Unit B0002 P870-1C 154127 Reference Documents ID No. Title Rev. Date
- 1. 04-1-01-E21-1 System Operating Instruction: 21 12/18/86 -
Low Pressure Core Spray System ! Safety Related 1 . . . s
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PROBABILISTIC RISK ASSESSMENT-BASED i INSPECTION PLAN Residual Heat Removal (RHR) System TABLE B6-1 IMPORTANCE BASIS AND FAILURE MODE IDENTIFICATION l CONDITIONS THAT CAN LEAD TO FAILURE ) General Guidance ) Surveillance of the licensee's periodic calibration, testing and/or pre- i ventive or unscheduled maintenance activities, procedures and training and/or. l normal and emergency operating procedures, training and check-off lists in ac . I cordance with the Technical Specifications,and relevant NRC bulletins and in-formation notices should reduce the probability of failure for the conditions i listed below. The most ralevant aspects are designated for each condition as ! follows: ' l PC - Periodic calibration activities, procedures and training. PT - Periodic testing activities, procedures and training. I MT - Preventive or unscheduled maintenance activities, procedures and training. I OP - Normal and emergency operating procedures, check-off lists, I training, etc. Mission Success Criteria l The RHR system is composed of several modes: LPCI, CSS, SPC and SDC shar-ing certain components. These commonalities are as follows: (a) the RHR pumps are common to the CS, LPCI, SPC and SDC modes; (b) the suppression pool suction valve for each pump train is common to the CS, SPC and LPCI modes; and (c) heat exchanger cooling is common to the CS, SDC and SPC modes. Low Pressure Core Injection (LPCI) System The purpose of the Low Pressure Core Injection (LPCI) System is to provide coolant to the reactor vessel during accidents in which vessel pressure is low. ADS can be used in conjunction with LPCI to attain a low enough system pressure for injection to occur. LPCI is a three-train system: each train consisting of a motor-driven pump and motor-operated valves. LPCI pumps take water from the suppression pool. LPCI is automatically initiated and control- i led. The operator may be required to manually start..the system if an automa-tic actuation failure occurs, and also to control flow during an ATWS if i ' required. The success criterion for the LPCI is injection of flow from any one pump to the reactor vessel. Also considered in the PRA is a Standby Service Water cross-tic system used to provide a coolant makeup source to the reactor vessel during accidents in which normal sources of emergency injection have f ailed. The SSW cross-tie system is comprised of Train B of the SSW system and Train B of the LPCI sys-tem. SXT usen SSW Pump B to inject water into the reactor via the LPCI system
B-29 TABLE B6-1 (Cont'd). CONDITIONS THAT CAN LEAD TO FAILURE Train B injection lines. The RHR heat exchangers are'not needed for this system. This backup function is not considered further in this report. i
- 1. Presence of Containment Spray Actuation Signal The presence of a containment spray (CS) actuation signal fails LPCI Trains A and B because of flow diversion to the sprays.
- 2. Standby Service Water (SSW) Failures: -!
a) SSW Train B Fails Due to Failure of or Maintenance on SSW B Pump
~
i Room HVAC l Failure of SSW Train B causes RHR pumps MDP2B/C to be unavailable due to f pump room cooling failure. (MT, PT) i b) SSW Pump Train A/B Fails or Unavailable Due to Maintenance SSW Pump A/B f ails to start or continue to run or check valve CV8A/B has a hardware fault, or pump or valve is out for maintenance. (MT, PT) c) SSW Train A/B Unavailable Due to Failure / Maintenance of'SSW , l Valve MVI A/B l I RHR pump A/B is unavailable due to pump room cooler unavailability: SSW l valve IWL A f ails closed, or is out for maintenance. (MT, PT)
- 3. Failure of Flow through Pipe Segment LP-2 Due to Hardware Faults Injection line LP-2 (Train C) into the reactor vessel fails due to.a valve }
hardware fault: normally-open manual valve XV239 is plugged, or testable ' check valve TCV241 fails due to hardware fault, or motor-operated valve MV242 falls to open. (MT, PT)
- 4. Failure of Flow through Pipe Segment LP-4 Due to Valve Faults Train C check valve CV31C fails due to a hardware fault or normally-open manual valve XV29C is plugged. (MT, PT)
- 5. RHR-C Pump Room Cooler Hardware Failure RHR Pump Room C fan coil unit hardware failure renders pump C unavailable in four hours. (MT, PT)
- 6. LPCI Pumps MDP2C/A Failure or Out for Maintenance LPCI Train C/A is rendered unavailable by failure or maintenance of pump. (MT, PT)
, . , B-30 TABLE B6-1 (Cont'd) 'h CONDITIONS THAT CAN LEAD TO FAILURE y 1
- 7. Minimum Flow Line Unavailable Due to Valve Faults I LPCI Train C is unavailable.due to valve faults in minimum flow line:
motor-operated valve MV64C is plugged, or manual valve XVi8C is plugged, or check valve CV46C fails due to a hardware fault. .(MT, PT)
- 8. Failure of Flow through Pipe Segment LP-3 Due to Hardware Faults ,
t injection line LP-3 (Train A) into the reactor vessel f ails due to a l l valve hardware fault: normally-open manual valve'XV39A is plugged, or-l testable check valve TCV41A fails due to hardware fault,.or motor-operated valve MV42A fails to open. (MT, PT)
- 9. LPCS Actuation Circuitry Hardware Fails, and Operator Fails to Manually Actuate LPCS Autouatic .and manual actuation of LPCS fail. RHR Train A shares actuation OP, PT) sensors with LPCS: failure of LPCS actuation fails RHR Train A. (MT, 1 Containment Spray System (CSS)
The Containment Spray (CS) Mode of the Residual Heat Renoval (RHR) System is to suppress pressure in containment du ring accidents; an additional function is to reduce radionuclides concentration in containment after an ] accident. CSS is a two loop system: c'ach loop consisting of motor-operated )i valves and one motor-driven pump. There are two heat exchangers in series per loop. Cooling water flow to the heat exchangers is required for CS mode.'The . CS suction source is the suppression pool. CSS is automatically initiated and' controlled. The operator may be required to raanually start the system'if an automatic actuation f ailure occurs. The success criterion for the CSS is j injection of flow from any one pump train through both heat exchangers to the i spray ring in containment. l
- 1. Standby Service Uater (SSW) Failures:
a) SSW Train B Fails Due to Failure of or Maintenance on SSW B Pump Room HVAC Failure of SSW Train B causes RHR pump MDP2B to be unavailable due.to pump room cooling failure. (MT, PT) b) RHR Train A/B Heat Exchanger Coolers Fail or Are Unavailable Due to Maintenance Heat exchanger cooler unavailability rende rs RHR Train A/B unavailable. (MT, PT)
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__ __ _ _ _ _ A
B-31
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Table B6-1 (Cont'd) a CONDITIONS THAT CAN LEAD TO FAILURE -) 1 c) SSW Pump Train A/B Fails or Unavailable Due to Maintenance SSW Pump A/B fails to start or continue to run or check valve CV8A/B has a hardware fault, or pump or valve is out for maintenance. (MT, PT) -; d) SSW Train A/B Unavailable Due to Failure / Maintenance of SSW Valve FNI A/B RHR pump A/B is unavailable due to pump room cooler unavailability: SSW valve MVIA fails closed, or is out for maintenance. (MT, PT) I i e) SSW Train A/B Unavailable Due to Failure / Maintenance of SSW Valve MVSA/B l RHR pump A/B le unavailable due to pump room' cooler unavailability: SSW valve MV5A/B f ails closed, or is out for maintenance. (MT, PT)
- 2. RHR HX A/B Bypass Valve MV48A/B Fails Open or Is Out for Maintenance :
l Motor-operated valve FN48A/B, normally open, f ails open thus diverting ! flow away from the heat exchangers, or valve is out'for maintenance. (MT, PT) )
- 3. RHR Train A/B Unavailable Due to Hardware Fault in or Maintenance on LP-12/21 I RHR pump MDP2A/B fails to start or continue to run, or check valve CV31A/B has hardware failure or manual valve XV29A/B is plugged, or component is out for maintenance. (MT, PT)
- 4. RHR-A/B Pump Room Cooler Hardware Failure RHR pump Room A/B fan coil unit hardware failure renders pump'A/B unavailable in four hours. (MT, PT)
- 5. CSS Train A/B Actuation Failure Due to Hardware Fault CSS Train A is not actuated due to hardware failure. (MT, PT)
- 6. LPCS Actuation Circuitry Hardware Fails, and Operator Fails to Manually Actuate LPCS Automatic and manual actuation of LPCS fail. RHR Train A shares actuation sensors with LPCS: failure of'LPCS actuation fails RHR Train A. (MT, OP, PT) l.
1 B-32
. TABLE B6-1 (Cont'd) J i
CONDITIONS THAT CAN LEAD TO FAILURE ,{ l Shutdown Cooling (SDC) System l j
.k The purpose of the Shutdown Cooling (SDC) Mode of the Residual Heat Re- !
moval (RHR) System is to remove decay heat during accidents in which reactor. vessel integrity is maintained. SDC is a two-loop system: each loop consist-
)
j ing of motor-operated valves and one motor-driven pump. There are.two heat exchangers in series per loop. Cooling water flow to the heat exchangers is J j required for SDC mode. The SDC suction source is recirculation pump B's { suction line. SDC is manually initiated and controlled. The operator is I required to align the system and to start the pumps. The success criterion j l for the SDC is injection of flow from any one pump train through both heat i exchangers to the reactor vessel. 1
- 1. Jperator Fails to Properly Align for Shutdown Cooling There is no automatic actuation of the SDC mode of RHR; manual actuation is required. (OP)
- 2. Standby Service Water (SSW) Failures:
i a) SSW Train B Fails Due to Failure of or Maintenance on SSU Pump Room HVAC . l l Failure of SSW Train B causes RHR pump MDP2B to be unavailable due to pump room cooling failure. (UT, PT) b) RHR Train A/B Heat Exchanger Coolers Fail or Are Unavailable Due to Maintenance l Heat exchanger cooler unavailability renders RHR Train A/B unavailable. (MT, PT) c) SSW Pump Train A/B Fails or Unavailable Due to Maintenance ! SSW Pump A/B fails to start or continue to run or check valve CV8A/B has a hardware fault, or pump or valve is out for maintenance. (MT, PT) d) SSW Train A/B Unavailable Due to Failure / Maintenance of SSW Valve FNI A/B RHR pump A/B is unavailable due to pump room cooler unavailability: SSW valve MVIA fails closed, or is out for maintenance. (MT, PT) e) SSW Train A/B Unavailable Due to Failure / Maintenance of SSW Valve MV5A/B 1 RHR pump A/B is unavailable due to pump room cooler unavailability: SSW valve MV5A/B fails closed, or is out for maintenance. (MT, PT)
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B-33 TABLE B6-1 (Cont'd) i 1 CONDITIONS THAT CAN LEAD TO FAILURE
- 3. RHR HX A/B Bypass Valve MV48A/B Fails Open or Is Out for Maintenance
~
Motor-operated _alve v MV48A/B, normally open, fails open thus diverting flow away from the heat exchangers, or valve is out for maintenance. (MT, PT) J
- 4. RHR Train A/B Unavailable Due to Hardware Fault in or Maintenance on-l j
LP-12/21 l J RHR pump MDP2A/B fails to start or continue to run, or check valve CV31A/B has hardware failure or manual valve XV29A/B is plugged, or component is out for maintenance. (MT, PT)
- 5. RHR-A/B Pump Room Cooler Hardware Failure j
RHR pump Roon A/B fan coil unit hardwa re failure renders . pump A/B unavailable in four hours. (MT, PT) J
- 6. SDC Train A/B Injection Line Fails Due to Hardware Fault or is Out for ~
l Maintenance j jq SDC Train A/B inj ection fails due to va lve fault: motor-operated valve ] MV53A/B fails to open, or check valve CV50A/B fails due to hardware fault or j maintenance on MV53A/B renders line unavailable. (MT, PT) Suppression Pool Cooling (SPC) System The purpose of the Suppression Pool Cooling (SPC) Mode of the . Residual , Heat Removal (RRR) System is to remove decay heat f rom the suppression pool I during accidents. SPC is a two loop system: each loop consisting of motor-operated valves and one motor-driven l pump.. There are two heat exchangers in series per loop. Cooling water flow to the heat exchangers is required for SPC mode. The SPC suction source is the suppression pool. SPC is manually initiated and controlled. The operator is required to align the system and start the pumps. The success criterion for the SPC is injection of flow f rom any one pump train through both heat exchangers to the suppression pool.
- 1. Operator Fails to Properly Align for Suppression Pool Cooling i
There is no automatic actuation of the SPC mode of RHR; manual actuation is required . The operator must open valves MV24A/B. (0P) C____ . _ _ _ _ _ _ _ _ _ _ _ _ ____m_____m_ _ _ _ . _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _
B-34. TABLE B6-1 (Cont'd)~ I CONDITIONS THAT CAN LEAD TO FAILURE
- 2. Standby-Service Water (SSW) Failures:
J l a) -SSW Train B Falls Due to Failure of or Maintenance on SSW Pump. } Room HVAC I Failure of SSW Train B causes RHR pump MDP2B to be unavailable due to pump room cooling failure. (MT, PT) {
'i b) RHR Train A/B Heat Exchanger Coolers Fall or Are Unavailable Due l to Maintenance {
q Heat exchanger cooler. unavailability renders RHR Train A/B tinavailable. I l (MT, PT) l' ) l c) SSW Pump Train A/B Fails or Unavailable Due to Maintenance i i SSW Pump A/B f ails to start or continue to run or check valve CV8A/B has a hardware fault, or pump or valve is out for maintenance. (MT, PT) d) SSW Train A/B Unavailable Due to Failure / Maintenance of SSW Valve l MVIA/B . RHR pump A/B is unavailable due to pump room cooler unavailability: valve MVIA fails closed, or is out for maintenance. (MT, PT) SSW
]
l e) SSU Train A/B Unavailable Due to Failure / Maintenance of.SSW Valve l MVSA/B l RHR pump A/B is unavailable due to pump room cooler unavailability: SSW ; valve MV5A/B fails closed, or is out for maintenance. (MT, PT) d
- 3. RHR HX A/B Bypass Valve MV48A/B Fails Open or Is Out for Maintenance I Motor-operated valve MV48A/B, normally open, fails open thus diverting flow away from the heat exchangers, or valve is out for maintenance. (MT, i PT) '
- 4. RHR Train A/B Unavailable Due to Hardware Fault in or Maintenance on LP-12/21 7 RHR pump MDP2A/B fails to start or continue to run, or check valve CV31A/B has hardware failure or manual valve XV29A/B is plugged, or component is out for maintenance. (MT, PT) l l
l i
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, , B TABLE B6-1 (Cont'd)
CONDITIONS THAT CAN LEAD TO FAILURE
- 5. RHR-A/B Pump Room Cooler Hardware Failure
-RHR pump Room A/B fan coil unit hardware failure renders put.p A/B unavailable in four hours. (MT, PT) i
- 6. SPC Valve MV24A/B Fails to Open or Is Out~for Maintenance l Motor-operated test return valve MV24A/B fails to open or is out for j maintenance. (MT, PT) d 1
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B-36 I GRAND CULF NUCLEAR STATION, UNIT 1 l PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN ; Residual Heat Removal (RHR) System TABLE B6-2 I&E INSPECTION PROCEDURES FOR SYSTEM OPERATION ! PROCEDURE FAILURE NUMBER TITLE '! ODES COMPONENTS ( 41700 Training LPCI Operation CSS.6 LPCI.9 i 41701 Requalification Training CSS' Operation SDC.1 , LPCI/LPCS Operation SPC.1 SPC/SDC Alignment 52051 Instrument Components e-d CSS Actuation LPCI.1,9
- Systems-Procedure Review Logic, '1SS.6
~l l 52053 Instrument Components and LPCI/LPCS Actuation l
Systems-Work Observation Logic 52055 Instrument Components and Systems-Record Review I 56700 Calibration l 62702 Maintenance (Refueling) RHR Pumps A/B/C,SSW LPC1 2-8 . l 62703 Monthly tlaintenance Pumps A/B, MOVs, CCS.1-4 Observation ] RHR A/B/C Pum? Rm SDC.2-6 j l 72700 Startup Testing-Refueling Coolers, SPC.2-6 { ' RHR A/B HX Coolers ! 61725 Surveillance and Calibration Pumps, MOVs, Fan LPCI.2 Program Coolers, HX, Coolers CCS.1-4 : 1 61726 Monthly Surveillance -SDC.2-6 Observation SPC.2-6 71707 Oper'ational Safety Pumps, MOVs, LPCI.2k Verification Logic Circuitry CCS.1-b 71710 ESP System Walkdown SDC.2-6 SPC.2-6
e , GRAND CULF NUCLEAR STATION, UNIT 1 PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN Residual Heat Removal (R11R) System TABLE B6-3 MODIFI2D SYSTEM WALKDOWN Desired Act. Pow.Sup. Required Act. Description ID No. Location Position Pos. Breaker # Location Position Pos. i RHR Pump C F029C Aux Bldg Locked i Discharge Area.10 Open 93 ft RilR Loop C F039C Contain Locked Isolation 135 ft Open 17C .j RRR C Test- F041C Drywell Sw:NA ; i able Check P601- Closed 1 Valve 17C j RilR C Inject F042C P601- Sw: AUTO 52- 16B11 Closed Shutoff Valve 17C Closed 161124 I ' l RilR C Min Flo F064C P601- AUTO 52- 16B11 Closed to Supp Pool 17C open 161123 RilR Pump C F013C AB 10 Locked { Min Flo Isol 93 ft Open j I RHR PUMP C C002C AB 93 ft AUTO , 152- 16AB1 Racked- ( P601-17C 1606 in 1 RHR Room C T51- 52- 16B11 Closed Coil Unit B005 161136 A Loop LPCI F039A Contain Locked Isolation 135 ft Open i RHR A Test- F041A Drywell NA able Check P601-20C Closed Valve RHR A Inject F042A P601-20C AUTO 52- 15B31 Closed l Shutoff Valve Closed 153136 CSS A Sparger F028A P601-20C SW: AUTO 52-' 15B31 Closed l Inlet Valve Closed 153150 CSS B Sparger F023B P603-17C SW: AUTO 5 2-' 16B31 Closed Inlet Valve Closed 163133
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B-30.: ? I Table 36-3 (Cont'd) Desired- Act. Pow.Sup. Req. Act. Description 13 No. Location Position Pos. Breaker // Location Pos. Pos. RilR A Shutdwn F053A CR Sw:AUTb 52- 15B31 Closed C1g Rtn to FW P601- Closed IS3141 20C RHR B Shutdwn F053B CR Sw: AUTO 52- 16B31 Closed Cig Rtn to FW P601- Closed 163107 17C RRR Hx A F048A CR Sw: AUTO- 52- 15B31 Closed Bypass Valve P601 - Open 153128 20C RHR lix B F048B CR Sw: AUTO 52- 16B31 Closed Bypass Valve P601- Open 163103 i 17C l RilR Pump A C002A Aux Bldg AUTO 152- 15AA Racked-l Room A 1509 In P601-l 20C ! RllR Pump B C002B Aux Bldg AUTO 152- 16AB Racked- . , _ _ _ Room B 1606 In P601-17C RilR Ru A Fan T51- AB Rm A 52- 15B11 Closed-Coil Unit B003 93 ft 151129 I ~~~ RHR Rm B Fan T51- AB Rn B 52- 16B3) Closed-B004 ' l Coil Unit 93 ft 163119 1 1 , l RHR A test Rtn F024A P601-20C SW: AUTO 52 15 B3't Closet { l to Supp Pool Closed 153122 RHR B Test Rtn F024B P601-17C SW: AUTO 52 16B31 Closed to Supp Pool Closed 163132 l l l kq l l \ Reference Documents j l ID No. Title Rev. Date
- 1. 04-1-01-E12-1 System operating Instruction: 35 12/8/86 Residual Heat Removal Systen Safety j Related.
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GRAND GULF NUCLEAR STATION,. UNIT 1 :l
' PROBABILISTIC RISK ASSESSMENT-BASED ' !
INSPECTION. PLAN. Suppression Pool Makeup (SPMU) Systen-TABLE B7-1 IMPORTANCELBASIS AND FAILUkE MODE IDENTIFICATION CONDITIONS THAT CAN' LEAD TO FAILURE-1-. General Guidance Surveillance of the licensee's periodic calibration, testing and/or. preventive' or unscheduled maintenance activities, procedures.and training . . and/or normal'and emergency operating procedures, training and check-off lists in accordance with.the Technical Specifications and relevant NRC bulletins'and j information. notices should reduce the' probability of failure.for the -] conditions listed below. The most relevant aspects are designated for each- j condition as follows: j PC - Periodic calibration activities,' procedures and training.- PT - Periodic testing activities,. procedures and. training.- MT - Preventive or unscheduled maintenance activities, procedures and training. OP - Normal and emergency operating procedures, check-off lists, " training, etc. Mission Success Criteria-The Suppression Pool Makeup (SPMU) System provides water from the upper containment pool to the suppression pool following a LOCA.- Water which - gravity flows f rom the upper containment pool to the suppression poo1~ is of-sufficient quantity to keep the uppermost drywell vents covered for all conceivable accidents. SPMU consists of two lines which penetrate the sidewalls in the separator storage area of the upper containment pool. These lines are' routed down to the suppression pool on either' side of the steamc tunnel. Each makeup line has two normally closed, motor-operated butterfly: valves in series. Valves are powered from onsite emergency power sources maintaining divisional separation and redundancy. .The; upper pool-is dumped-when the valves receive a divisionally separate but simultaneous signal to open.
- 1. Mode Selector Switches HS-M603A and HS-M603B Not 'in AUTO Mode selector handswitches must be in the AUTO position in; order'to
~
actuate each SPMU valve either automatically or. manually. (OP)
- 2. Train A/B. Hardware Fault: 1 of 2 Valves Fails to Open Normally-closed motor-operated valve F001A/B'or F002A/B f ails to open due to hardware failure. '(MT, PT) g 1,
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~B-45 TABLE B7-1 (Cont'd)
CONDITIONS THAT CAN LEAD TO FAILURE 2, Train A/B Hardware Fault: 1 of 2 Valves Fails to Open Normally-closed motor-operated valve F001A/B or F002A/B fails to open due to hardware failure. (MT, PT)
- 3. Train A/B Actuation Logic Failure Train A/B actuation logic fails. (MT, PT)
I I 1 o
e, . B-46 I GRAND CULF NUCLEAR STATION, UNIT 1 PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN , 1 i Suppression Pool Makeup (SPMU) System TABLE B7-2 I&E INSPECTION PROCEDURES FOR SYSTEM OPERATION PROCEDURE FAILURE ) NUMBER TITLE. COMPONENTS ~ MODES 41700 Training Mode Selectors 1
)
41701 Requalification Training 'HS-M603A/s 'j 62703 Maintenance (Refueling) MOVs 2 62703 Monthly Maintenance l Observation 72700 Startup Testing-Refueling 61725 Surveillance and Calibration MOVs, Actuation 2,3 ) Program Logic 61726 Monthly Surveillance
)
j Observation 71707 Operational Safety l Verification 71710 ESF System Walkdown i l l l l ___._.___.___-___._J
.,t .. . . a B-47 CRAND GULF NUCLEAR STATION, UNIT 1 PROBABILISTIC RISK ASSESS!!ENT-BASED IllSPECTION PLAN Suppression Pool Makeup (SPMU) System TABLE B7-3 MODIFIED SYSTEM WALKD0ilN Desired Act. Pow.Sup. Required Act. Description ID No. Location Position Pos. Breaker # Location Position Pos. SPMU Div 1 F001A CR Sw: AUTO 52- 15B21 Closed Inbd Dump .P870-4C Closed 152144 Viv SPMU Div 2 F001B CR Sw: AUTO 52- 16D41 Closed Inbd Dump P870-100 Closed 164135 Viv SPMU Div 1 F002A CR Sw: AUTO 52- 15B21 Closed Otbd Dump PS70-4C Closed 152107 Viv SPMU Div 2 F002B CR Sw: AUTO 52- 16B41 Closed Otbd Ducp P870-10C Closed 164136 Viv 1 Reference Documents ID No. Title Rev. Date
- 1. 04-1-01-E30-1 System Operating Instruction: 15 6/26/86 Suppression Pool Makeup System Safety Related l
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CRAND GULF NUCLEAR STATION, UNIT 1 j PROBABILISTIC RISK ASSESSMENT-BASED l INSPECTION PLAN TABLE C1 - PLANT OPERATIONS INSPECTION CUIDANCE I 4 Recognizing that the normal system lineup is important for any given ] standby safety system, the following human errors are identified in the ] PRA as important to risk. SYSTEM FAILURE DISCUSSION l 4 High Pressure Operator Fails to Manually 0 pen Valve Table A2-1, Item 11 Core Spray F015 Given Auto Failure ; (HPCS) . j Operator Fails to Manually Initiate HPCS Table A2-1, Item 12 4 Given Auto Failure f 1 Reactor Core Operator Fails to Manually Initiate RCIC Table A3-1, Item 12 Isolation { Cooling (RCIC) Operator Fails to Open Valve F031 Table A3-1, Item 13 I Standby Liquid Operator Fails to Initiate SLC Table A6-1, Item 4 ! Control (SLC)
)
Condensate (CDS) Operator Fails to Align Valve AV513 for Table B2-1 Item 1 Reactor Vessel Inj ection Operator Fails to Restore Instrument Air Table B2-1, Item 2 l Containment Operator Fails to Vent Containment Table '13-1, Item 1 , Venting (CVS) Operator Fails to Restore Instrument Air Table B3-1, Item 2 Control Rod Operator Fails to Restore Instrument Air Table B4-1, Item 2 Drive (CRD) Operator Fails to Initiate. Standby CRD Table B4-1, Item 3 Pump Low Pressure Operator Fails to Manually Close MV11 Table B5-1, Item 8 Core Spray (LPCS) Operator Fails to Manually Actuate'LPCS Table B5-1, Item 9 Residual Heat -Operator Fails to Properly Align for Table B6-1, Removal: Shutdown Cooling Item SDC 1 Shutdown Cooling (SDC) l 1
a .: I C-2
, a TABLE C1 (Cont'd)
SYSTEM FAILURE DISCUSSION
]
Residual Heat LPCS Actuation Circuitry Hardware Fails, Table 36-1, i
. Removal: and Operator Fails to Manually Actuate Item CSS 6 j Containment LPCS Spray (CSS) Mode i
Residual Heat LPCS Actuation Circuitry Hardware Fails, Table B6-1, ! Removal: Low and Operator Fails to Manually Actuate Item LPCI'9 Pressure Core LPCS
. Injection (LPCI) j l
Residual Heat. Operator Fails to Properly Align for Table B6-1,
- Removal: Suppression Pool Cooling Item SPC 1 l
, Suppression Pool l Cooling (SPC) ] { Suppression Pool Mode Selector Switches HS-M603A-and Tabl'c B7-1, Ites 1 Makeup (SPMU) HS-M603B Not in AUTO i I f 1
? ... .
'C-3 4 5 GRAND GULF NUCLEAR STATION, UNIT 1 PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN TABLE C2 - SURVEILLANCE AND CALIBRATION INSPECTION GUIDANCE The listed components are the risk significant components for which j surveillance and/or calibration should minimize failure.
- . SYSTEM FAILURE DISCUSSION Emergency DG 11 Fails to Start or Run or is Out Table Al-1, Item 1 Electric Power for Maintenance l (EPS) l DG 12 Fails to Start or Run or is Out Table Al-1, Item 2 for Maintenance l
DG 13 Fails to Start or Run or is Out Table Al-1, Item 3 for Maintenance 4160 V AC Bus 16AB Failure Table Al-1, Item 5 4160 V AC Bus 15AA Failure Table Al-1, Item 6 4160 V AC Bus Failure Due to Loss of Table Al-1, Item 7 HVAC Connon Mode Failure of Battery Divisions Table Al-1, Item 8 125 V DC Bus Failure Due to Loss of HVAC Table Al-1, Item 9 Battery Bank 1B3 Failure Table Al-1, Item 10
~ Battery Bank 1 A3 Fallure Table Al-1, Item 11 l
125 V DC Bus 11DB Failure , Table Al-1, Item 12 125 V DC Bus 11DA Failure Table Al-1, Item 13 l 125 V DC Bus 11DC Failure Table Al-1, Item 14 1 l High Pressure HPCS Pump Fails to Start or Continue to Table A2-1, Item 2 Core Spray Run or Is Unavailable Due to Maintenance (HPCS) Hardware Failure in Feed Line PS-8: Table A2-l', Item 3 Valve F004 Fails to Open or Is Out. for Maintenance
i
.. C-4 TABLE C2 (Cont'd)
SYSTEM FAILURE DISCUSSION High Pressure Minimun Flow Valve F012 Fails to Open or Table A2-1, Item 4 I Core Spray Is Out for Maintenance l
-(HPCS)
(Cont'd) Suppression Pool Suction Line Hardware Table A2-1, Item 5 Fault: F015 Fails to Open ] HPCS Pump Room Cooler Failure or Out for Table A2-1, Item 6' Maintenance HPCS Actuation Logic / Circuit Failure- Table'A2-1, Item 7 Minimum Flow Valve (F012) Control Failure Table A2-1, Item 8
.I I
Common Mode Miscalibration of CST Level -Table A2-1, Item 9 ') Instruments i Common Mode Miscalibration of Reactor Table A2-1, Item 10 Vessel Level Sensors Reactor Core RCIC Pump Fails to Start or Is Out for Table A3-1, Item 2 l . Isolation Maintenance l Cooling (RCIC) High Reactor Vater Level Signal Hardware Table A3-1, Item 3 Failure: Actuation Logic Failure, I LT-N695A Failure, LT-N695B Failure ' Motor-Operated Valve F013 Fails to Open Table A3-1, Item 4 or Is Out for Maintenance MOV F046 Fails to Open or Is Out for Table A3-1, Item 5 Maintenance MOV F019 Fails to Open or Is Out for Table A3-1, Item 6 Maintenance Motor-Operated Valve F045 Fails to Open Table A3-1, Item 7 or Is Out for Maintenance l RCIC Pump Room Cooler Fails or Is Out Table A3-1, Item 8 for Maintenance ' Switchover to Suppression Pool Actuation Table A3-1,. Item 9 Failure Minimuu Flow Valve F010 Controller Table A3-1, Item 10 Failure
1
. .. ] ,? ., ,
l C-5 - 4 TABLE C2 (Cont'd) SYSTEM FAILURE DISCUSSION Reactor Core RCIC Actuation Logic Failure Table A3-1, Item 11 Isolation Cooling (RCIC) Common Mode Miscalibration of CST Level' Table A3-1, Item 14 (Cont'd) Instruments Comuon Mode Miscalibration of Reactor Table A3-1, Item 15 l Level Sensors l
}
Standby Service SSW Valve MVIA/B'(F001A/B) Fails'to Open Table A5-1, Item 1 Water (SSW) or is Out for Maintenance . SSW Valve MV5A/B (F005A/B) Fails to Open Tatle A5-1, Iten 2 or is Out for Maintenance , 1 SSW Valve MV11 (F011C) Fails to Open or Table AS-1, Item 3 is Out for Maintenance SSW Valve XV13 (F013) Is Plugged Table AS-1, Item 4 l SSW Pump MDP1A/B Fails to Start or Table A5-1, Iten 5 l Continue to Run or is Out for Maintenance ! l SSW Pump !@P2C Fails to Start or Table A5-1, Iten 6 Continue to Run or is Out for Maintenance I SSW Valve XV199A/B (F199A/B) is Plugged Table A5-1, Item 7 ' SSW Valve IW16A/B (F014A/B) Fails to Table A5-1, Item 8 Open or Is Out for Maintenance SSW Valve MV68A/B (F068A/B) Fails to Table AS-1, Item 9 Open or Is Out for Maintenance SSW Valve XV37/38 (F037/38) Is Plugged Table A5-1, Item 10 SSW Valve XV105/106 (F105/106) Is Table A5-1, Item 11 Plugged SSW Valve XV47/48 (F047/048) Is Plugged Table A5-1, Item 12 SSW Valve XV51/52 (F051/052)- Is Plugged Table A5-1, Item 13 SSW Valve IW18A/B (F018A/B) Fails to Table A5-1, Item 14 Open SSW Valve MV23A/B (F023A/B) Fails to Table A5-1, Item 15 Open
. .. j
') ., .
> C-6
+ m TABLE C2 (Cont'd)
SYSTEM FAILURE DISCUSSION i 3 1 Standby Service DC 11/12 Jacket Cooler (HX4A/B) Plugged Table A5-1, Item 16 ; Water (SSW) or Out for Maintenance (Cont'd) ; J Standby Liquid Failure to Restore Two Valves in Table A6-1, Item 1 Control (SLC) Discharge Test Line after Test Failure to Reclose Manual Test. Valve F031 Table A6-1, Item '2 after Suction Test ! One of Two Check Valves in Injection Table A6-1, Item 3 q Line Fails l Pump Suction Inlet Valves F001A and Table A6-1, Item 5 l F001B Fail to Open or Are Out for ; Maintenance Motor-operated Valves F001 and F004 Fail Table A6 Item 6 to Close 1 l Motor-operated Valves F250 and F251 Fail Table A6-1, Item 7 l to Close SLC Train A and B Pumps Fail to Start or Table A6-1, Item 8 Continue to Run or Are Out for Maintenance Automatic Common Mode ADS Valve Failure Table B1-1, Item 1 Depressurization (ADS) Condensato (CDS) Failure of Turbine Building Cooling Table B2-1, Item 4 Water (TBCW) System with Offsite Power Available Containment Vent Path Hardware Failure: 1 of 4 Table B3-1,. Item 3 Venting (CVS) Dampers Fails to Open or Remain Open Control Rod Standby Pump C001B/A Fails to Start or ' Table B4-1, Item 5 Drive (CRD) Continue to Run Normally Operating CRD Pump C001A/B Table B4-1, Iteu 6 Fails to. Continue to Run Low Pressure LPCS Pump MDP1 Fails to Start or Table 35-1, Item 1 l Core Spray Continue to Run Due to Hardware Fa'ults (LPCS) or Is Out for Maintenance Motor-operated Valve MVS Fails to Open Table B5-1, Item 3 L_.______________.
1 ., . i ,) C-7 TABLE C2 (Cont'd) SYSTSH FAILURE DISCUSSION j-Low Pressure Minimum Flow Valve MVil Fails to Close Table B5-1, Itea 4 Core Spray Due to Hardware Faults (LPCS) _
]
(Cont'd) LPCS Pump Room Cooler Fails or Is Out Table B5-1, Iten 5 for Maintenance i Minimum Flow Valve Actuation Circuitry Table B5-1,~ Item 6 Fails due to Hardware Faults Automatic LPCS Actuation Circuitry Table B5-1, Iten 7 I Hardware Failure Residual Heat Failure of Flow through Pipe Segment LP-2 Table B6-1, Removal: Low Due to Hardware Faults Item LPCI 3 Pressure Core Injection (LPCI) Failure of Flow through Pipe Segment LP-4 Table B6-1, i Due to Valve Faults Item LPCI 4 i RHR-C Pump Room Cooler Hardware Failure Table B6-1, Item LPCI 5 LPCI Pumps MDP2C/A Failure or Out for Table B6-1, Maintenance Iten LPCI 6 i Minimuu Flow Line Unavailable Due to Table B6-1, j Valve Faults Item LPCI 7 ) Failure of Fluw through Pipe Segment Table B6-1, LP-3 Due to Hardware Faults Item LPCI 8 LPCS Actuation Circuitry Hardware Fails Table B6-1, Item LPCI 9 Residual Heat RHR HX A/B Bypass Valve FN48A/B Fails Table B6-1, Removal: Open or Is out for Maintenance Item CSS 2 Containment Spray (CSS) Mode RHR Train A/B Unavailable Due to Table.B6-1, Hardware Fault in or Maintenance on Item CSS 3 LP-12/21 RRR-A/3 Pump Room Cooler Hardware Failure Table B6-1, Item CSS 4 CSS Train A/B Actuation Failure Due to -Table B6-1, ! Hardware Fault Item CSS 5 LPCS Actuation Circuitry Hardware Fails Table B6-1, Item CSS 6 l 1
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TABLE C2 (Cont'd) l SYSTEM FAILURE DISCUSSION { 1 Residual Heat RilR HX A/B Bypass Valve MV48A/B Fails Table B6-1, l Removal: Open or Is Out for Maintenance Item SDC 3 j Shutdown Cooling i (SDC) RHR Train A/B Unavailable.Due to Table B6-1, I Hardware Fault in or Maintenance on Item SDC 4 . LP-12/21 J RHR-A/B Pump Room Cooler Hardware Failure Table B6-1, Item SDC 5 , l SDC Train A/B Injection Line Fails Due Table B6-1, ) to Hardware Fault or Is Out for Item SDC 6 ] Maintenance 1 li Residual Heat 'RHR HX A/B Bypass Valve MV48A'B Fails Table B6-1, Removal: Open or Is Out for Maintenance- Item SPC 3 Suppression Pool ) Cooling (SPC) RHR Train A/B Unavailable Due to Table B6-1, . liardware Fault in or Maintenance on Item SPC 4 q LP-12/21 j
\
RilR-A/B Pump Room Cooler Hardware Failure Table B6-1, Item SPC 5 SPC Valve MV24A/B Fails to Open or Is Table B6-1, . Out for Maintenance Item SPC 6
]
Suppression Train A/B Hardware Fault: 1 of 2 Valves Table B7-1, Fool Makeup Fails to Open Item 2 (SPMU) . l Train A/B Actuation Logic Failure Table B7-1, i Item 3 l l l e M.t
l 4 [f ., .-
.Q C-9 1 1
GRAND GULF UUCLEAR STATION, UNIT 1 PROBABILISTIC RISK ASSESSMENT-BASED , IM3PECTION PLAN j TABLE C3 - MAINTENANCE INSPECTION GUIDANCE
- )
j The components listed here are significant to risk because of unavailability I for maintenance. The dominant contributors are usually frequency and duration of maintenance, with some contribution due to improperly performed maintenance. SYSTEM FAILURE DISCUSSION l l Emergency DG 11 Fails to Start or Run or is Out Table Al-1, Item 1. Electric Power for Maintenance ( EPS) DG 12 Fails to Start or Run or is out Table Al-1, Item 2 ) for Maintenance I
.I DG 13 Fails to Start or Run or is Out Table Al-1. Item 3 -
for Maintenance 4160 V AC Bus 16AB Failure Table Al-1, Item 5 l 4160 V AC Bus 15AA Failure Table Al-1, Item 6 4160 V AC Bus Failure Due to Loss of Table Al-1, Item 7 HVAC Common Mode Failure of Battery Divisions Table Al-1, Item 8 125 V DC Bus Failure Due to Loss of HVAC Table Al-1, Iteu 9 , I Battery Bank 1B3 Failure Table Al-1, Item 10 l 3attery Bank 1 A3 Failure Table Al-1, Item 11 125 V DC Bus 11DB Failure Table Al-1, Item 12 125 V DC Bus 11DA Failure Table Al-1, Item 13 125 V DC Bus 11DC Failure Table Al -1, Item 14 High Pressure HPCS Pump Fails to Start or Continue to Table A2-1,. Item 2 Core Spray- Run or.Is Unavailable Due to Maintenance (HPCS) j Hardware Failure in Feed Line PS-8: Table A2-1, Item 3 ' Valve F004 Fails to Open or Is Out;for { Maintenance f _ )
V ... .
.,. C-10
' TABLE C3 (Cont'd)
SYSTEM FAILURE DISCUSSION High Pressure Minimum Flow Valve F012 Fails to Open or' . Table A2-1, Item 4 Core Spray Is Out for Maintenance (HPCS) (Cont'd) Suppression Pool Suction Line Hardware- Table A2-1, Item 5 Fault: F015 Fails to Open HPCS Pump Room Cooler Failure or Out for Table A2-1, Iteu 6 Maintenance l HPCS Actuation Logic / Circuit Failure Table A2-1, Item 7-Minimum Flow Valve (F012) Control Failure Table A2-1, Item 8 l Reactor Core RCIC Pump Fails to Start or Is Out for Table A3-1, Item 2 Isolation Maintenance l Cooling (RCIC) ! High Reactor Water Level Signal Hardware Table A3-1, Item 3 , Failure: Actuation Logic Failure, i LT-N695A Failure, LT-N695B Failure 1 i l Motor-Operated Valve F013 Fails to Open Table A3-1, Item 4 l l or Is Out for Maintenance MOV F046 Fails to Open or Is Out for Table A3-1, Item 5 ) Maintenance MOV F019 Fails to Open or Is Out for Table A3-1, Item 6 ! Maintenance ) l tiotor-Operated Valve F045 Fails to Open Table A3-1, Item 7
'or Is Out for Maintenance RCIC Pump Room Cooler Fails or Is Out Table A3-1, Item 8 for Maintenance Switchover to Suppression Pool Actuation . Table A3-1, Item 9 Failure Minimum Flow Valve F019 Controller Table A3-1, Item 10 Failure RCIC Actuation Logic Failure Table A3-1, Item 11 ,
'b e, .
i
.ijy C-11 TABLE C3 (Cont'd)
SYSTEM FAILURE DISCUSSION Standby Service SSW Valve IWl A/B (F001 A/B) Fails to Open Table AS-1, Iten 1 Water (SSW) or is Out for Maintenance-SSW Valve MV5A/B (F005A/B) Fails to Open Table A5-1, Item 2 or is Out for Maintenance 3 SSti Valve MV11 (F0 llc) Fails to Open or Table A5-1, Iten 3 is Out for Maintenance SSW Valve XV13 (F013) Is Plugged Table AS-1, Item 4 SSW Puap MDPlA/B Fails to Start or Table AS-1,. Item 5 , Continue to Run or is Out for Maintenance ! SSW Pump MDP2C Fails to Start or Table AS-1, Iten 6 Continue to Run or is Out for Maintenance - SSW Valve XVl99A/B (F199A/B) is Plugged Table AS-1, Item 7 l SSU Valve IN14A/B (F014A/B) Falls to Table AS-1, Iten 8 j Open or Is Out for !!aintenance i SSW Valve MV68A/B (F068A/B) Fails to Table A5-1, Item 9 i Open or Is Out for Maintenance j l SSW valve XV37/38 (F037/38) Is Plugged Table A5-1, Item 10 SSW Valve XV105/106 (F105/106) Is Table A5-1, Item 11 Plugged SSW Valve XV47/43 (F047/048) Is Plugged Table A5-1, Item 12 SSW Valve XV51/52 (F051/052) Is Plugged Table AS-1, Item 13 SSW Valve 11V18A/B (F018A/B) Fails to Table A5-1,. Item 14' Open L SSW Valve MV23A/B (F023A/B) Fails to Table AS-l', Item 15 Open l DG 11/12 Jacket Cooler (IlX4A/B) Plugged Table A5-1. Item 16 l or Out for Maintenance Standby Liquid Failure to Restore Two Valves in Table A6-1, Item 1 Control (SLC) Discharge Test Line After Test Failure to Reclose Manual Test Valve F031 Table A6-1, Item 2 after Suction. Test
.o ,
t
,? 'df' ,
! C-12 l TABLE C3 (Cont'd)
SYSTEtt FAILURE DISCUSSION Standby Liquid One of Two Check Valves in Injection Table A6-1, Item 3 Control (SLC) Line Fails (Cont'd) Pump Suction Inlet Valves F001A and Table A6-1, Item 5 F001B Fall to Open or Are Out for-Maintenance Motor-operated Valves F001 and F004 Fail Table A6-1, Item 6 to Close Motor-operated Valves F250 and F251 Fail Table A6-1, Item 7 to Close 1 l SLC Train A and B Pumps Fail ~to Start or Table A6-1, Item 8 ! Continue to Run or Are Out for Maintenance Automatic Common Mode ADS Valve Failure Table B1-1, Item 1 Depressurization (ADS) i j Condensate (CDS) Failure of Turbine Building Cooling Table B2-1, Item 4 1 Water (TBCW) System with Offsite Power Available j Containment
)
Vent Path Hardware Failure: 1 of 4 Table B3-1, Item 3 j Venting (CVS) Dampers Fails to Open or Remain Open l Control Rod Failure to Restore Manual Valve F0217B/A Table B4-1, Item 4 Drive (CRD) Following Pump Maintenance Standby Pump C001B/A Fails to Start or Table B4-1 Item 5 Continue to Run Normally Operating CRD Pump C001A/B Table B4-1,-Item 6 Fails to Continue to Run Low Pressure LPCS Pump MDP1 Fails to Start or Table B5-1, Item 1 Core Spray Continue to Run Due to Hardware Faults (LPCS) or Is Out for Maintenance Motor-operated valve MVS Fails to open Table B5-1, Item 3 Minimum Flow Valve MV11 Fails to Close Table B5-1, Item'4 Due to Hardware Faults
t ., .
-. /!
4 C-13 -! TABLE C3 (Cont'd) SYSTEM FAILURE- DISCUSSION Low Pressure LPCS Pump Room Cooler Fails or Is Out Table B5-1, Core. Spray for Maintenance Itcu 5 '? (LPCS) (Cont'd) Minimum. Flow Valve Actuation Circuitry . Table B5-1, Fails due to Hardware Faults Item 6 Automatic LPCS Actuation' Circuitry Table B5-1, Hardware Failure Item 7 l Test Line Valve MV12 Out for Maintenance -Table B5-1, Item 10 LPCS Suction Valve IN1 Out for Table B5-1, Maintenance Item 11 Residual Heat Failure of Flow through Pipe Segment LP-2 Table B6-1, Removal: Low Due to Hardware Faults Item LPCI 3 Pressure Coolant Injection (LPCI) Failure of Flow through Pipe Segment LP-4 Table B6-1, I Due to Valve Faults Item LPCI 4 RHR-C Pump Room Cooler Hardware Failure Table 36-1, Iten LPCl 5 j i LPCI Pumps MDP2C/A Failure or out for Table B6-1, Maintenance ] Item LPCI 6 i Minimum Flow Line Unavailable Due to Table B6-1, Valve Faults Iten LPCI 7 Fa.ilure of Flow through Pipe Segment Table B6-1, LP-3 Due to Hardware Faul ts Item LPCI 3 LPCS Actuation Circuitry Hardware Fails Table B6-1, Iteu LPCI 9 Residual Heat RUR RX A/B Bypass Valve tN48A/B Fails Table B6-1, Removal: Open or Is Out for Maintenance Item CSS 2 ( Containment Spray (CSS) Mode RHR Train A/B Unavailable Due to Table B6-1, Hardware Fault in or Maintenance on Item CSS 3 LP-12/21 RHR-A/B Pump Room Cooler Hardware Failure Table B6-1, Item CSS 4 CSS Train A/B Actuation Failure Due to Table B6-1, Hardware Fault Item CSS 5 LPCS Actuation Circuitry Hardware Fails Table B6-1, Item CSS 6 l 4db
._._____._._____W"
.? . ,' ,
C-14 TABLE C3 (Cont'd) SYSTEM FAILURE DISCUSSION Residual Heat RHR RX A/B Bypass Valve MV48A/B Fails Table B6-1, Removal: Open or Is Out for Maintenance Item SDC 3 Shutdown Cooling (SDC) RHR Train A/B Unavailable Due to Table B6-1, I Hardware Fault in or Maintenance on Item SDC 4 LP-12/21 , i RHR-A/B Pump Room Cooler Hardware Failure Table B6-1, Item SDC 5 4 SDC Train A/B Injection Line Fails Due Table B6-1, to Hardware Fault or Is Out for Item SDC 6 i Maintenance Residual Heat RHR RX A/B Bypass Valve MV48A/B Fails Table B6-1, Removal: Open or Is Out for Maintenance Item SPC 3 Suppression Pool Couling (SPC) RHR Train A/B Unavailable Due to Table B6-1 Hardware Fault in or Maintenance on Item SPC 4 j LP-12/21
)
l RHR-A/B Pump Room Cooler Hardware Failure Table B6-1 Item SPC 5 SPC Valve FN24A/B Fails to Open or Is Table B6-1 Out for Maintenance Item SPC 6 Suppression Train A/B Hardware Fault: 1 of 2 Valves Table B7-1, Pool Makeup Fails to open Item 2 (SPMU) Train A/B Actuation Logic Failure Table B7-1, Item 3 a
'b $6 s
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'{s~ / C-15 CRAND CULF NUCLEAR STATION, UNIT J PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN TABLE C4 - CONTAINMENT AND DRYWELL UALKDOWN Discussion Since the drywell is generally inaccessible during normal plant operation, those components listed in the preceding tables which are located either within the drywell or otherwise in the containment are listed below: I I
i l Desired Act. l l Description ID No. Location Position Pos. l J A
- 1. SLC Recire. to Test Tank Isol. F016 Contain. 184' Closed Valve.
I l l 2. SLC Recire. to Test Tank Isol. F017 Contain. 184' Closed ! ! Valve
- 3. SLC Test Tank Outlet F031 Contain. 184' Locked /
I Open
- 4. SLC Outboard Stopcheck F006 Contain. 161' Locked / I Handwheel Open i
- 5. SLC Inboard Stopcheck F007 Contain. 100' Locked /
Handwheel Open i
- 6. SLC Inboard Isolation F008 Drywell 100' Locked /
Open 4 l , 7. LPCI Loop A Isolation F039A Contain. 135' Locked / l Open
- 8. RHR Loop C Isolation F039C Contain 135' Locked / I Open
- 9. LPCS Injection Isolation Valve F007 Drywell 147' Locked /
Open l l t l}}