ML17284A390
| ML17284A390 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 06/30/1987 |
| From: | Fresco A, Taylor J, Villaran M BROOKHAVEN NATIONAL LABORATORY |
| To: | |
| Shared Package | |
| ML17279B125 | List: |
| References | |
| CON-FIN-A-3453 A-3453-87-6, TAC-M66591, NUDOCS 8806240154 | |
| Download: ML17284A390 (188) | |
Text
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) fp DRAFT TECHNICAL REPORT A-3453-87-6 WASHINGTON NUCLEAR PLANT NO>>
2 GENERIC PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN Prepared by:
J. Taylor, A. Fresco, M. Villaran, W. Gunther,
& R. Lewis Engineering Technology Division Department of Nuclear Energy Brookhaven National Laboratory
- Upton, N.Y 11973 June 1987 Prepared for the U.S. Nuclear Regulatory Commission r
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WASHINGTON NUCLEAR PLANT NO.
2 GENERIC PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN 1 ~
INTRODUCTION The tables and paragraphs in this inspection plan have been prepared to provide inspection guidance and have been based on a review ef the following:
1) previously prepared PRA based system inspection plans (1-4),
2) prior generic studies (+6),
3)
ASEP (7),
and GESSAR (8).
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 guidance or to consti-tute an additional set of inspection requirements.
In using this information one should realize that it is based primarily on prior PRA'ork.
- Hence, recent system experience,
- failures, and modifications should be considered when using 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.
2.
SYSTEM PRIORITY LIST The grouping on WNP-2 systems into three groups is based on the impor-tance of these systems as demonstrated in 9 BWR PRA's, and as modified for the WNP-2 specific plant design.
Systems were considered of high importance if they were one of the dominant contributors to core melt or high offsite conse-quences.
Within each group, no relative ranking has been done.
Systems that
'o not appear in any of the three groups were deemed of lesser importance than those that were ranked.
3 ~
DOMINANT ACCIDENT INITIATORS AND SEQUENCES The following dominant accident initiators and sequences are ranked in order of importance:
Initiators:
1.
2.
Major transients in power conversion system (PCS),
such as turbine trip, loss of FW, MSIV closure, etc.
Sequence5:
1.
Station blackout.
2 ~
ATWS 3.
Transient, coupled with failure of PCS, loss of HPCS and RCIC, ADS failure.
4.
Transient plus failure of long term RHR.
- 5..Interfacing systems LOCA.
4.
SYSTEM INSPECTION TABLES Three tables have generally been prepared for each system to provide inspection guidance.
These tables are described below (see Table 1).
Table X-1 Failure Modes Those components or licensee activities which play a dominant role in contributing to system importance are presented, along with a brief descriptioq of why these items are important.
Inspection focus on these items vbb1ShoAA.TvPicallf address
>75% of the risk significant areas.
For experienced inspectors, this table is probably sufficient.
A simplified system diagram extracted from the WNP-2 Training Manuals is included for each system, which gives the valve num-bers used.
Table X-2 I&K Procedures For those who prefer additional guidance, this table identifies those I&E inspection procedures which can be used to assure the availability of the items shown in Table 2.
The inspection procedures were identified based on the failure modes presented and an understanding of I&E procedures.
The pro-cedures selected are those which provide routine guidance on the principal plant programmatic activities such as operations, maintenance, instrumenta-tion/control and surveillance testing.
There are many other inspection proce-dures which could also be used depending on the inspection criteria or the inspector's preference.
- However, the procedures selected will generally pro-vide adequate inspection coverage of the dominant failure modes.
Table X-3 Modified S stem Walkdown This table provides an abbreviated version of the licensee's system checklist, but includes only those items which+e related to the dominant failure modes.
It is generally less than one third (sometimes only 1/10) of the normal checklist.
Caution should be observed when using the checklists, since they are based on certain versions of the licensee's checklist.
The revision date of the licensee's checklist that was used is indicated at the end of Table X-3.
0
S stems Im ortance Rankin Hi h Im ortance Electric Power (AC & DC)
Reactor Protection System (RPS)
Primary Containment (Includes Supp.
- Pool, DW Spray)
High Pressure Core Spray (HPCS)
Reactor Coolant Isolation Cooling (RCIC)
Automatic Depressurization System (ADS)
Residual Heat Removal (RHR) System Medium Importance Low Pressure Core Spray (LPCS)
Power Conversion System (PCS)
Standby Gas Treatment (Including RX Bldg I Main Steam Isolation Valves (MSIVs) g ntegrity)
Recirculation Pump Trip (RPT)
Low Importance Standby Liquid Control (SLC)
Reactor Closed Cooling (RCC) 4 Low Pressure Coolant Injection (LPCI)
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REFERENCES 1.
A. Fresco, et al., "Limerick Generating Station Unit 1, Probabilistic Risk Assessment-Based System Inspection Plans,"
Brookhaven National Laboratory Technical Report A-3453-87-2, Rev. 0, May 1987.
2.
A Fresco, et al.,
"Shoreham Nuclear Power Station Probabilistic Risk Assessment-Based System Inspection Plans,"
Brookhaven National Laboratory Technical Report A-3453-87-3, Rev. 0, May 1987.
3.
J. Higgins, R. Fullwood, A.'oppola, "DRAFT Millstone Nuclear Power Station Unit 1 Probabilistic Risk Assessment-Based System Inspection Plans,"
Brookhaven National Laboratory Technical Report A-3453-3-87, Rev. 0, March 1987 J. Usher, A. Fresco, "Draft Grand Gulf Nuclear Station Unit 1
Probabilistic Risk Assessment-Based System Inspection Plans,"
Brookhaven National Laboratory Technical Report A-3453-4 Draft, May 1987.
5 ~
N.A. Hanan, et al., "A Review of BMR/6 Standard Plant (GESSAR)
Probabilistic Risk Assessment:
Vol.
1 Inter/pl Events, Core Damage Frequency,"
NUREG/CR-4135P, Volume 1, Brookhaven National Laboratory, May 1985.
6.
J.C Higgins, "Generic PRA-Based BWR Insights," Brookhaven National Laboratory Technical Report A-3453-9-86, September 1986.
7.
"ASEP Methodology Guideline for the RebaAining of the NRC Reference
- Plants, Table 4 BWR Accident Sequence Insights," Draft, September 1985.
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WASHINGTON NUCLEAR PLANT N0.2 GENERIC PROBABILISTIC RISK ASSESSMENT BASED INSPECTION PLAN Electrical Power Distribution System TABLE EP-1 IMPORTANCE BASIS AND FAILURE MODE IDENTIFICATION CONDITIONS THAT CAN LEAD TO FAILURE General Guidance Surveillance of the licensee's periodic testing and/or preventive or unscheduled maintenance activities and procedures and/or normal and emergency operating procedures,
- training, and check-off lists in accordance with the Technical Specifications and relevant NRC bulletins and information notices should reduce the pro'bability of failure for the conditions listed below.
The most relevant aspects are designated for each condition as follows:
PC-PT MT OP-TS-Periodic calibration activities, procedures and training.
Periodic testing activities, procedures and training.
preventive or unscheduled maintenance activities, procedures and training.
Normal and emergency operating procedures, check-off lists, training, etc.
Technical specifications.
Mission Success Criteria Failure Conditions A.
Alternatin Current (AC) Power Distribution S stem (Includes Standb AC Power S stem 5
120V/240V Uninterru tible Power S
l S stem Cnf The dominant AC syst: em failures are Item 1 combined with>one item from Items 2-5.
1.
Loss of normal power (TR-N1, TR-N2) and failure to transfer to startup (TR-S) or backup (TR-B) source (failure of main generator lockouts or bus'ransfer breakers) or isolation of 6900V and 4160V AC buses from grid (system or switchyard problem) requiring emergency diesel generators (EDGs) to be started (PT,OP);
AND
TABLE EP-1 (Cont'd)
CONDITIONS THAT CAN LEAD TO FAILURE 2.
a) Failure to start or run, or b) Failure or degradation of room ventilation, or c) Loss of service water coo1.ing, or d) Failure to restore components after test or maintenance.
OR 3.
Failure of undervoltage relays that sense Loss of offsite power and send start signal to EDGs (PT).
OR 4.
Various circuit breaker faults (PT):
a) Failure of circuit breakers to close that tie emergency buses to sources of emergency power b) Failure of circuit breakers to open that isolate failed sources of power OR 5.
Failure of operators to take appropriate recovery actions (OP):
a) Failure to recover de-energized buses b) Failure to manually start EDGs 6.
Failure of 12OV vital Ac inverters and failure of automatic transfer via static swi ch from inverter to alternate source of AC power.
B.
Direct Current (DC) Power Distribution S stem 1.
Failure of battery charger or in maintenance:
125V DC, chargers Cl-l, C1-2, Cl-HPCS b) 25QV DC, Charger C2-1 2 ~
Failure of batteries due to: (PT,MT) a) Insufficient charge b) Loss of battery room ventilation 3 ~
Operational Test or Maintenance Error Resulting in: (PT,MT,OP) a) Deenergizing, or cascade failure of DC power supplies b) Failure to properly restore batteries or charger after maintenance or testing.
WASNINGTON NUCLEAR PLANT N0.2 GENERIC PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN Electrical Power Distribution System TABLE EP-2 I&E INSPECTION PROCEDURES FOR SYSTEM OPERATION PROCEDURE NUMBER TITLE FAILURE COMPONENTSP MODES 41700 Training Offsite power sources, All.
diesel-generators, switchgear, trans formers, battery sets,
- chargers, breakers 61725 Surveillance Testing and Calibration Program Offsite power sources, A:1-4,6 diesel-generators B:1-3 switchgear, transformers battery sets,
- chargers, breakers 61726 Monthly Surveillance Observation 71707 71710 62702 62703 Diesel-generators, A:2,6 switchgear, battery B:1-3
- sets, chargers Monthly Maintenance Observation
~
'perational Safety Verification ESF System Walkdown Maintenance (refueling)
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WASHINGTON NUCLEAR PLANT NO 2 GENERIC PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN Electrical Power Distribution System TABLE EP-3 MODIFIED SYSTEM WALKDOWN Control Room Verify that power is available to or from the following:
From the off-site power sources 25kV through TR-N1 from the ASHE substation 230kV through TR-S from the ASHE substation 115kV through Tr-B from the Benton substation 2 ~
To the three 4KV busses SM-4 SM-7 S.i>>8 3 ~
To the five 48OV busses SL-71 SL-73 SL-81 SL-83 MC-4A 4 ~
To the three 125V DC busses 5 ~
S 1-1 81-2 125V DC HPCS distribution panel Verify that all diesel generator alarms are cleared Diesel Generator Rooms Verify the following:
1 ~
2 ~
3 ~
4 ~
5.
6.
7 ~
DG local/remote selector switch is in remote position.
Ventilation fan switches are in Auto or Standby.
Governor oil level is satisfactory.
Air receiver tank pressure is psig.
Fuel oil day tank level is satisfactory gallons or feet.
Electric immersion heater is operating correctly to maintain the engine jacket cooling water at 125'F when the engine is not running.
AC motor-driven water circulation pump, for moving the water through the jacket cooling water system when the engine is not running, is operating properly.
TABLE EP-3 (Cont'd) l Critical 120V AC and Uninterruptible Power Su ply S stems Verify power is available from the following sources:
IN-1 1.
Normal AC MC-7A 2.
Normal DC DP-52-1 3.
Alternate AC MC-7F 4.
Bypass AC MC-7A IN-2 1.
Normal DC DP-51-2 2.
Alternate AC MC-8A IN-3 1.
Normal DC DP-51-1 2.
Alternate AC MC-7A Required Position 125V/250V Direct Current (DC) Power Distribution S stem Description Ac ual Position MC-7A/BKR-4A MC-7A/BKR-18 MC-8A/BKR-1C MC-4/BKR 480V Supply to C2-1 480V Supply to C1-1 480V Supply to Cl-2 480V Supply to Cl-HPCS AC Input BKR C2-1 AC Input BKR Cl-1 AC Input BKR C1-2 AC Input BKR Cl-HPCS DC Output BXR C2-I DC Output BKR Cl-1 DC Output BKR Cl-2 DC Output BKR Cl-HPCS Battery Output Disconnect Bl-1 Battery Output Disconnect Bl-2 Battery Output Disconnect Bl Battery Output Disconnect B2-1 Closed Closed Closed Closed Closed Closed Closed Closed Closed Closed Closed Closed Closed Closed Closed Closed
WAStiINGTON NUCLEAR POWER PLANT NOe 2
GENERIC PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN
)
I v
Elec rical Power Distribution System TABLE EP-4 PROPOSED INSPECTION PLAN FOR DIESEL GENERATORS AT NUCLEAR PLANTS A. ~Ob'ective 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 auxlli.ary systems as follows:
Fuel Injection System, Turbocharger, Star-ing
- System, Speed/Load Control, Jacket Water, Cooling Water, Lube Oil, Fuel Oil, Control and Monitoring Systems, and Genera-or.
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 failures from the internal systems.
Evaluate the licensee's response to these failures for proper failure
- analysis, corrective action, notification of vendor, Par" 2 1 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'aintenance recommendations (especially those recommendations unique to nuclear service DGs such as Col 's de-scribed in NSAC-79)?
Are maintenance personnel specially trained on DGs?
Is failure information fed back into maintenance program?
Has 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 on 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 from 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 storage area.
(Reference IE I.P.
387013)
~
TABLE EP-4 (Cont'd) r 6.
Training:
Ensure appropriate DG specific training given to mainte-
- nance, operations, QA, and management personnel.
Are there adequate documents to describe DG operation onsite (both main engine and aux-iliary system)?
(Reference IE I.P. 41700).
7.
Observe DGs in operation.
Ensure they run smoothly and are operated 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?
8.
Is NPP receiving all appropriate service information from vendor:
- design, maintenance, operational, etc?
This is especially important for General Motors DG owners (verify they receive "Power Pointers" from GM) ~
9.
Review site practices to limit DG cold fast starts.
10.
Reliability records and calculations:
Check logs, procedures, and calculations versus Reg.
Guide 1.108 criteria.
11.
Ensure that pertinent studies on DG performance have been reviewed and recommendations implemented as appropriate (e.g.,
NUREG/CR-0660 and NSAC-79) ~
12.
Torquing:
Ensure plant has adequate specifications for all torquing.
Ensure it is documented and done with calibrated equipment.
Observe re-torquing if in progress.
Source 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 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'eliability,"
NUREG/CR-0660, University of Dayton, February 1979.
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WASHINGTON NUCLEAR 'POD@ PLANT NO.
2 GENERIC PROBABZLISTIC RISK ASSESSMENT-BASED INSPECTION PLAN Electrical Power Distribution System TABLE EP-5 PROPOSED INSPECTION PLAN FOR CRITICAL 120V AC 6 UNINTERRUPTIBLE POWER SUPPLY SYSTEMS 1.
Ensure that periodic verification of the system line-up of normal AC and DC power supplying the inverters and the availability of alternate power sources occurs.
In particular, check that.the "IN SYNCH LIGHTS/ON" are powered.
2.
Verify that there is a periodic operability check (during refuelings) of the auto transfer switch capability.
3.
Verify that procedures for the start of the GE computer specify switchover to an alternate power source when placing the computer on line and restoration to normal when the computer is on line.
4 Review the alarm response procedures for the UPS.
5.
Verify the adequacy of the procedure for pre-charging of the inverters.
~
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.NAEHINGTON NUCLEAR ~EDliUUAIPLANE NO.
2 GENERIC PROBABILISTIC RISK ASSESSMENT BASED INSPECTION PLAN Elec"rical Power Distribution System TABLE EP-6 PROPOSED INSPECTION PLAN FOR DIRECT CURRENT (DC)
POWER DISTRIBUTION SYSTEM General 1.
Review the alarm response procedures to ensure that timeliness of response is emphasized.
2.
Verify that all appropriate actions are specified when a charger or battery is removed from service.
3 ~
Ensure that there are no ties between DC divisions.
In particular, verify that procedures are clearly defined for placing the spare charger for the 125V DC system into service without cross-tying between divisions.
Batteries 1.
Review periodic surveillance and testing of:
~
Battery capacity (2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> minimum)
~
Electrolyte level
=
" 0' Specify gravity 2.
Check the blown fuse indicator between the battery and the distribution panel.
3.
Ensure that all seismic supports are in place.
~Char era 1.
Review periodic capacity test to ensure that it reflects expected loads and recharging of batteries within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
2.
Verify calibration and functional testing of charger alarms.
3 Review the alarm response procedures for emphasis on timeliness in restoring chargers to service, actions to take on low DC voltage, etc.
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) 4-Dl -3
) 4-2 4160V SH-4
) 4-41 TR 4-41 CRIT MC 4A VT X
1 FIGURE +
GRITIGALPOV/ER SUPPLY OIVISION 3 840305.19A JAN 1986
~
4
~ A
CURRENT TRANSFCRMER evs ACB TO LOAD GNO OET. SEAL IN 8 RESKT ON ITK EOUIP ONLY RESET SOLID STATE LOGIC UNGROUNDED CIRCUIT CVRRENTS BALANCE ANO CANCEL EACH OTHER.
NO CURRENT FLOWS THROUGH C.T.
TO GROUND DETECTOR
- CIRCUITS, NO LOGIC OUTPUT 8 RELAY IS OEENERGIZED.
50 GX RELAY W
r W
50eA 63 50eA X/63 GROUND LAMPS ARE
- LIT, ALARMS ARE CLEAR.
50BAX 63 COMPUTER CONT.
Rhf.
ANNVN NOTE:
ITE GROUND DETECTORS SEAL IN ANO REQUIRE MANUAL RESET Ml-2 GROUND DETECTORS ARK SELF-RESETTING Bl 'S PACB 5A cA SOA CURRENT TRANSFORMER TO I.OAO 20A
- GRCUND GNO OKT. SEAL IN 6 RESET ON ITK KOUIP ONLY GROUND LAMPS OUT INOICA > E CIRCVIT GROUND SOLID STATE LOGIC r
EI.A'Y GR UN0
'T CURRENT TRANSFORMER SEES 20A UNBALANCE AS 20A CURRENT.
C.T.
SECONDARY OUTPUT TO SOLID STATE LOGIC CAUSES LOGIC OUTPUT TO 50GX RELAY.
50GX ENERGIZING OPENS CONTACT TO GROUND LAMP CIRCUIT ANO OPERATES ALARM TYPER ANO CONTROL ROOM ANNUNCIATOR.
50BAX/63 50BAX/63 cllCA 0
COMPUTFR CONT. RM.
ANNUN FIGURE 5. GROUND DETECTION CIRCUIT 840305.24 A MARCH I984
TR-N2 1
N2-5
)
RRC-P-lA t
TR-S 6900V BUS SH-5
) S-5 CT-A )
C.T. BUS 51 C.T.
BUS 52 480Y BUS 53 I
l
>l 1
IO ICI ICI I
O O
I I
I-I-
I CD tl Cy C
IO I
O o
O I
I I
O O
O CD IO I
I O
O 4
R IA IA oI o ~ MC-58-A DIVISION A
DIVISION 8
TR-N2 N2 6 I) 6900V BUS SH-6 RQ.S CT-8)
RRC -P-IB t
TR-S
) S-6 480V BUS 63
~
C.T. BUS 6l C.T BUS 62 llew
~)l O
C0 C0 o
o
'MC68A LD CCI I
O O
I-I I
O O
O CL EOIO I-CD 0
C CO CO I
I O
Q I-I FlGURE 6. 6900V NON-CRITlCALD1STR1BUTlON 840505.25A MAACHI984
I - I I 4IGOY BUS SM - I I
CI I
ClZo CD CVI CL I
Cl O
I CL I
O O
LLI LLI 4.
TR I-I I 480V BUS I I I
CL I
CL:
m O
n Lu I
I O
O O
O O
2I-I I~ BUS 2(
FIGURE 7A. NON-CRITICALBUS SM-1 4160Y & 480Y DISTRIBUTION 840305.2I A MARCH I984
)))
4I BOV BUS SM 2 m
m m i
CL Q
0 g
z z+
C)
C)
V O TR 2-2l 2l-I l 48OV BUS 2l
)))) i)
Bl-2l musca.
m cu I
CD ~
C3 a-m Gl CL 04 C3 FIGURE 7B. 'NON-CRITICALBUS SM-2 4160V & 480V DISTRIBUTION 84Q3Q5.22A MARCH 1984
Nl-3 )
4l6OV BUS SM-3
) 3-3I
)S-I
) 3.8 0
CL d.
I l
I z
z 0
O CD I
C9 L
TR 3-3l 480V BUS 3t 3I-2I tOID CD tO ID O
tO IO Cl tO tD I)
IA tO D
FIGURE 7C.
NON-CRITICALBUS SM-3 4160Y 8c 480Y DISTRIBUTION 840305.23A MARCH iee4
WASHINGTON NUCLEAR POWER PLANT NO.
2 E XPERIENCE-BASED INSPECTION PLAN Reactor Protection System (RPS)
TABLE RPS-1 GENERALIZED INSPECTION PLAN Discussion PRAs for other BWRs do not provide a detailed system model of the RPS, but rather utilize historic failure data to establish a system failure probability.
For completeness, a non PRA-based inspection plan for the WNP-2 RPS has been developed from the following documents:
WNP-2 FSAR, Section 7
WNP-2 Systems Training Manual, Volume 2, Section 7
Recent I&E Information Notices
& Bulletins (1985 and 1986 to date)
NUREG-0460, Anticipated Transients Without Scrams (ATWS) for Light Water Reactors GE Standard Technical Specifications I&E Manual and Inspection Procedures Selected Regulatory Guides
& Standards S stem Description The WNP-2 RPS includes the standard BWR-RPS arrangement.
The RPS receives power from two MG sets, one for each of the two Trip Systems, A&B.
Each MG set is electrically isolated from its RPS bus by-two Electrical Protection Assemblies (EPA) arranged in series.
These EPAs provide addi" ional
~
protection by monitoring and tripping their respective EPA breaker on overvoltage, undervoltage, or underfrequency.
Each trip system has two trip
- logics, Al&A2 and Bl&B2.
They are arranged in a one out, of two taken twice logic.
There are 12 separate signals that can generate a scram.
When the scram system actuates, power is cut from the scram pilot valve solenoids, bleeding air from the scram header and opening the scram inlet and outlet valves.
This allows water pressure to insert the control rods.
Outlet water is sent to the scram discharge volume and scram instrument volume.
Inspection Areas 1 ~
Review and witness RPS function surveillance tests for RPS.
Include witness of Partial Manual Scram Test, single rod scram, tests of individual RPS Channels, and RPS EPA Circuit Breakers.
- References include:
R.G.
1.22, "Periodic Testing of Protection System Actuation Function," for RPS; R.G.
1.118, "Periodic Testing of Electric Power and Protection Systems,."
which endorses IEEE Std 338-1977, "Criteria for Periodic Testing of Nuclear Power Generating Station Safety Systems," for RRCS only.
TABLE RPS-1 (Cont'd)
Detailed guidance for review of LPRM and APRM calibration is con-tained in IE Inspection Procedures 61703 and 61704.
2.
Tour and inspec" control rod drive hydraulic control unit (HCU) area.
3.
Review calibration records for system inputs versus-WNP-2 Technical Specifications.
Particular attention should be paid to the Reactor Vessel level sensors.
4.
Ensure no abnormal RPS alarms in control room.
5.
Check RPS and RRCS panels for jumpers.
Ensure any existing jumpers are warranted and proper.
6 ~
The Control Structure Chilled Water System and associated cooler units are necessary to assure RPS operates in compliance with design requirements Ensure it is available and in good condition 7 ~
Verify that selected Technical Specifications are met:
Control Rods Section 3/4 1.3 RPS Instrumentation Section 3/4 3.1 RPS Setpoints Section 2.2 8.
Inspect instrument sensing racks for correct valve configuration, labelling, and separation 9 ~
Review past work testing of RPS maintenance tasks.
- 10. Review qualifications and training for technicians performing testing and/or maintenance on the system
~ ~',J' wq
~ ~
~t
%kg\\I a
CRIT. BUS SM-7 BUS SH-6 CRIT. BUS SM-8 4160Y 4ppy 69oov ppy 4160V 40py SL-73 SI -63 SL-83 MC-7A asoV
)
G C72-5003A )
C72-50030 )
120V MC-6B 480V 3-ff~ 'l20Y
!) C72-50035
!) C72-50035.
) ca.l ALTERNATE SUPPLY BUS
~ INTERLOCK~
MC-8A
!)
t
) C72-50038
) CT2-5003D 120V OTHER CRIT.
LOADS RPS BUS "A" RPS BUS "B" OTHER CRIT.
GP.
GP.
GP.
1A 2A 3A 4A SCRAM PILOT VLVS.
(V-117) SOLENOIDS SCRANI DISCH. VOL.
ISOL. VLVS.
PILOT SOL.
TRIP CHAN. "B" GP.
GP.
GP.
GP.
18 28 38 48 SCRAM PILOT VLVS.
(V-118) SOLENOIDS
'IGURE 1. RPS POWER DISTRIBUTION 820941.1LT OCT 1966 OPS
CIST II (IDTII4 545100551 Vl4 M 50ttlt Cll tCTI Cal t(Tlll t(Villa Cal IVtltat 0(0 I
)
IIDM
~5 I
SM IOII~
1 alla I
I I
5(aasl TatIll
)
Ill Ill
~I tlat ISI 5
51 caDTIS
(
10 cool(via Cao Tllll
~5'.W IIDMala(104 MtaollcIMI51511M Cao 5 110 I Dllllll 10 tao St IIII (SOT III SCaall tall TILT(5 1101451 III I
Io(0 totII SI Vial 10 0lal0 0(45 M
M Ik WS4
~I IIOMa(04 IVt.Ol t
~lit III(
III
'())sl 10 all (lta taolltlav 515ll55 St!5M 055(aaatl 1451a.vctSM(
4 10 alltl44 TIOMCIMI 115 11M al II
'10 a(a(Ill tao 1lCIMO allIIM 01 I I II
/
/
10 Ioo WlVDISMI t
0(ota 55 II 5
55 I
IS I
I5 II 15 55 I)
I I 1 i
C C
SII.CI 5
~us IDI 511 ~
I al (
I ~ I y
y 10 al a(i oa taolICIKIV 515IIM (14vlsla 5(aaM 455(aaaCt 14510. ID(4M('0 SOOII ato X
\\
6 V
IIDM 4(4'5 Cao 1 II I I TITS'I cao T Ilia 1
cao 1 I ca4111 CSNalisl CIDVlll IDO (IDTIIS I
5
~
'l.'
(
HVMs I
L llo'MalaCIII tao 1l ( IIo0 SISIIM I Ilaaosr I
I(ID1 151 I
Icaovl10 I
Itg IIIM ala (I0 1 ta4lltlga 5I5I(M FIGURE 3A; RPS INTERFACE WITH THE CRD SYSTEM 655576.9Ly
~
Ocy'986 APS
WASHINGTON NUCLEAR PLANT N0.2 GENERIC PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN Service Water Systems (SWS)
TABLE SW-1 IMPORTANCE BASIS AND FAILURE MODE IDENTIFICATION CONDITIONS THAT CAN LEAD TO FAILURE General Guidance Surveillance of the licensee's periodic testing and/or preventive or unscheduled maintenance activities and procedures and/or normal and emergency operating procedures,
- training, and check-off lists in accordance with the Technical Specifications and relevant NRC bulletins and information 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.
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.
TS Technical specifications.
Mission Success Criteria LATER Failure Conditions 1 ~
Failure of pumps SW-P-1A, P-1B, and HPCS P-2 tostart and run on demand due to:
a) maintenance outage or pump failure.
b) logic circuitry failure.
(PC,PT) c) plugging of pump suction in sump.
(PT) d) failure to open pump discharge valves V-2A, V-2B, V-29 or PCV 38 A,B.
TABLE SW-1 (Cont'd)
CONDITIONS THAT CAN LEAD TO FAILURE 2.
Failure of heat exchanger valves to operate on demand:
a) V-68 A,B RHR heat exchanger.
(PT,MT) b) V-4 A,B diesel generator cooling.
(PT,MT) c) V-24 A,B,C RHR pumps 2 A,B,C and room coolers.
(PT,MT) d) V-44 LPCS pump and room coolers.
Loss of water supply in panel A(B) due to:
a) failure to equalize level from panel B(A) ~
(OP) b) failure to manually provide makeup.
(OP) c) failure to manually maintain temperature above freezing.
(OP) 4.
Failure to restore system after test or maintenance.
RHR P 2 A~B~C V 22 A~B~C~ V 23 A~B~C~ V 17 A~B~C RHR heat exchanger 1A,B, V-14 A,B LPCS P-l, V-37, V-48, V-49 Syphon vent valves V-168 A,B, V-169 A,B Diesel generators V-3 A,B, V-10 A,B, V-40 A,B V-5 A,B, V-ll A,B, V-214/215/216/217 Spray pond return V-12 A,B, V-170 A,B 5.
Piping system rupture due to loss of heat tracing.
(OP)
Notes:
1.
For modified check-off sheets the following criteria was used:
a) vents and drains deleted b) certain room coolers were deleted along with associated instrumentation because they were not a dominant failure mode, sufficient time would be available to identify and correct situation, or they were not essential to system operation.
2.
Certain valves may show on two different system lists, i.e.,
RHR heat exchanger SW inlet/outlet valves could also show in both SW and RHR lists.
WASHINGTON NUCLEAR PLANT NO+2 GENERIC PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN Service Water Systems (SWS)
TABLE SW-2 I&E INSPECTION PROCEDURES FOR SYSTEM OPERATION PROCEDURE NUMBER TITLE COMPONENTS*
FAILURE MODES 56700 61725 61726 Calibration Surveillance Testing Calibration Program Monthly Surveillance Observation SW pumps, valves Logic circuitry 1>2>4 62702 Maintenance SW pumps, valves, 1,2,4 screens 62703 71707 Monthly Maintenance Observation Operational Safety Verification SW pumps, valves 1,2,4 screens Logic circuitry 71710 41700 ESF System Walkdown Training SW pumps',
valves 3,4,5 spray pond siphon 41701 Requalification Training
- Refers only to components identified in Tables SW-1 and SW-3.
WASHINGTON NUCLEAR PLANT N0.2 GENERIC PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN Standby Service Water System (SSWS)
TABLE SW-3 MODIFIED SYSTEM WALKDOWN DRAWING NO.
Ao VALVE CHECKLIST VALVE NUMBER LOC.
BLDG.
ELEVE DESCRIPTION REQUIRED POSITION ACTUAL POSITION COMMENTS SW-V-703 A-441 PI-3 and PS-40B Root Valve Open SW-V-29 A-44)
SW-V-707 A-441 HPCS-P-2 Discharge Valve (MO)
PT-40, PI-40, PS-40A Root Closed Open SW-V-702A A-441 P X-1A Root SW-V-701A A-441 PS-1API-1AG Root Open Open SW-V-71A A-441 SW-P-1A and 1B Pond Crossover Closed SW-V-71B A-441 SW-P-1A and 1B Pond Crossover Closed SW-V-2A A-441 SW-P-1A Discharge Valve (MO)
Closed SW-V-706A A-441 PT-32A PI-32A G Root Open SW<<V-65A8 A-441 PRA-CC-1A Inlet Isolation Open SW-V-791A A-441 FIS-42A Root SW-V-791B A-441 FIS-42A Root Open Open SW-V-66AC A>>441 PRA-CC-lA Outlet Isolation Throt.f SW-V-705A A-441 PX-42A Root Closed 8 Requires independent verification of valve position by licensee personnel.
f Sealed after flow s'et per PPM 7.4.7-1.1.1
TABLE SW-3 (Cont'd)
DRAWING NO.
A.
VALVE CHECKLIST VALVE NUMBER SW-V-12A LOC.
BLDG.
ELEVE A-4 31 DESCRIPTION REQUIRED POSITION A Loop to B Pond (MO) Locked Open ACTUAL POSITION COMMENTS TMU-LVC-1A A-43 1 A Pond Makeup Closed Tl'm-V-10A SW-V-69A A-431 A-431 A Loop Discharge to Cooling Towers (MO)
Locked Closed TMU-LCV-1A Isolation Closed SW-V-70A SW-V-820A SW-V-821A SW-PCV-38A A-431 A-44 1 A-44 1 A-431 A Loop Discharge to Cooling Towers (MO)
FT-8A Root Valve FT-8A Root Valve A Loop Pressure Controller Locked Closed Open Open Closed SW-V-165B Outside A Spray Header Bypass Closed Open in Cold Weather SW-V-170B Outside A Spray Header Isola-Open lation Open in Cold Weather SW-V-171B SW-V-168A Outside Outside A Spray Header Vent Open A Pond Siphon Vent Closed Check if pond level is below value SW-V-169A Outside A Pond Siphon Vent Closed Check if pond level is below value SW-V-165A Outside B Spray Header Bypass Closed Open in Cold Weather SW-V-170A Outside B Spray Header Isola-Open lation Closed in Cold Weather
TABLE SW-3 (Cont'd)
DRAWING NO+
Ao VALVE CHECKLIST VALVE NUMBER LOG ~
BLDG ELEV.
DESCRIPTION REQUIRED ACTUAL POSITION POSITION COMMENTS SW-V-171A Outside I
SW-V-168B Outside B Spray Header Vent B Pond Siphon Vent Open Closed Check if pond level is below value SW-V-169B Outside B Pond Siphon Vent Closed Check if pond level is below value SW-V-72A B-44 1 SW-P-1B to A Pond Crossover Closed'W-V-72B B-441 SW-P-1B to A Pond Crossover Closed SW-V-2B B-441 SW-V-65B(t B-441 SW-P-1B Discharge(MO)
PRA-CC-lB Inlet Iso-lation Closed Open SW-V-66BQ B-441 SW-705B B-441-SW-V-792A B-441 SW-V-792B B-441 SW-V-12B B-431 TMU-V-, 10B B-431 TMU-V-10B B-431 PRA-CC-1B Outlet Isolation PK-42B Root PIS-42B Root FIS-42B Root B Loop to A Spray Pond (MO)
B Pond Makeup TMU-LCV-1B Isolation Throt.t Closed Open Open Locked Open Closed Closed 8 Requires independent verification of valve position by licensee personnel.
Sealed after flow set per PPM 7.4.7.1.1.2.
~ TABLE SW-3 (Cont'd)
DRAWING NO.
A.
VALVE CHECKLIST VALVE NUMBER SW-V-69B SW-V-70B LOC BLDG.
ELEV.
B-431 B-431 DESCRIPTION B Loop to Cooling Towers (MO)
B Loop to Cooling Towers (MO)
REQUIRED POSITION Locked Closed Locked Closed ACTUAL POSITION COMMENTS SW-V-820B SW-V-821B B-441 FT-8B Root Valve B-441 FT-8B Root Valve Open Open SW-PCV-38B B-431 B Loop Pressure Control Closed SW-V-105D Q
SW-V-107D Q
RW-525 R1A-CC-51-Al Supply RW-525 Wma-CC-51-Al Supply Open Open SW-V-737G SW-V-104D g
RW-525 PX 51A Root RW-525 WMA-CC-51-..A1 Out le t Closed Throt.
I'W-V-106D g
RW-525 WMA-CC-51-Al Outlet Valve Throt.f'W-V-781A SW-V-781B RW-525 SW-FIS-35 Root RW-525 SW-FIS-35 Root Open Open SW-V-822A Q
RW-525 Iso, %$.-CC-51A-1 Supply Open M775 SW-V-823A g
RW-525
- Iso, WMA-CC-51A-1 Outlet Open M775 SW-V-225A Q
RW-525 WM-CC-51A-1 Bypass Closed Closed if chiller is idle.
9 Requires independent verification of valve position by licensee personnel.
f Throt le each valve to obtain 60 gpm per valve for a total flow of 120 + 6 gpm (see 7.4.7.1.1.1).
TABLE SW-3 (Cont'd)
DRAWING NO.
Ae VALVE CHECKLIST VALVE NUMBER LOC.
BLDG.
ELEV.
DESCRIPTION REQUIRED ACTUAL POSITION POSITION COMMENTS SW-V-224A RW-525 Q
- Iso, Supply to CCH-CR-1A Closed M?75 SW-V-826A RW-525 SW-V-227A RW-525 Q
- Iso, CCH-CR-lA
- Iso, CCH-P-1A Suction Open Closed M775 SW-V-898 RW-525 SW-V-89?
RW-525 SW-V-896A RW-525 SW-V-899A RW-525 SW-V-59A RW-525
- Vent, CCH-P-1A Suction Hdr Vent, CCH-EV-lA Discharge Hdr SW-FX-lA Root SW-V X-1A Root WMA-CC-52-Al Supply Isolation Closed Closed Closed Closed Open M?75 M775 M775 M775 SW-V-60AQ RW-525 SW-V-109A RW-525 Q
WMA-CC-52-A1 Return Isolation WMA-CC-53-Al Supply Isolation Throt.
Open SW-V-110A RW-525 'MA-CC-53-Al Return Line Isolation Open SW-V-63AQ RW-525 WMA-CC-53-Al Return Line Isolation Throt.f SW-V-109B RW-525 Q
WMA-CC-53-Bl Supply Isolation Open SW-V-111B RW-525 Q
WMA-CC-53-Bl Supply Isolation Open Q Requires independent verification of valve position by licensee personnel.
t Sealed after flow set per PPM 7'4.7.1.1-1.
TABLE SW-3 (Cont'd)
DRAWING NO.
Ao VALVE CHECKLIST VALVE NUMBER LOCo BLDG.
ELEV.
DESCRIPTION REQUIRED POSITION ACTUAL POSITION COMMENTS SW-V-108B RW-525 Q
tPiiA-CC-53-Bl Return Isolation Open SW-V-110B RW-525 Q
SW-V-63BQ RW-525 SW-V-59BQ RW-525 SW-V-60BQ RW-525 SW-V-105C RW-525 Q
SW-V-107C RW-525 Q
SW-V-104C RW-525 Q
SW-V-106C RW-525 Q
S'W-V-822B RW-525 Q
SW-V-823B RW-525 Q
WMA-CC-53-Bl Return Isolation WNA-CC-53-Bl Return Isolation WMA-CC-52-Bl Supply Isolation WMA-CC-52-Bl Return Isolation WMA-CC-51-Bl Supply Isolation WMA-CC-51-Bl Supply Isolation WMA-CC-51-Bl Return Isolation WMA-CC-51-Bl Return Isolation Iso>> WMA-CC-51B-1 Supply
- Iso, WMA-CC-51B-1 Outlet Open Throt.
Open Throt. t Open Open Throt.t Throt.
f'losed Closed M775 M775 Q Requires independent verification of valve position by licensee personnel.
Sealed after flow set per PPM 7.4.7.1.1.2.
f Throttle each valve to obtain 60 gpm per valve for a total flow of 120 + 6 gpm (see 7.4.7.1.1.2)
~
TABLE SW-3 (Cont'd)
DRAWING NO.
Ao VALVE CHECKLIST VALVE NUMBER SW-V-826B SW-V-224B Q
LOC.
BLDG.
ELEVE RW-525 RW-525 DESCRIPTION
- Iso, CH-CR-1B Outlet Header Iso, Supply to CCH-CR-1B REQUIRED POSITION Open Open ACTUAL POSITION COMMENTS M775 M775 SW-V-227B Q
RW-525
- Iso, CCH-P-13 Suction Open M775 SW-V-830B RW-525 SW-FX-2B Low Side Root Closed M775 SW-V-8313 RW-525 SW-FX-2B High Side Rooi Closed M775 SW-V-832B SW-V-852B SW-V-8963 RW-525 SW-FX-1B High Side Iso RW-525 SW-P I-42B Root RW-525 SW-PI-413 Root Open Open Closed M77 5 M775 M775 SW-V-899B SW-V-737F SW-V-3BI SW-V-4038 RW-525 SW-P X-513 Root DG RMlB Supply Isolation DG RMlB DCW-HX-1B1 1B2 Supply Isolation Closed Open Open RW-525 SW-FX-1B Low Side Iso Closed M775 M775 SW-V-736 B DG RM13 PX-15A Root Closed SW-V-220B Q
t DG RM1B DCW-TK-1Bl, 1B2 Iso Open SW-V-221B 18 SW-V-2213 26 DG RM1B DCW-TK-1B1 Iso DG RM1B DCW-TK-1B2 Iso Closed Closed 8 Requires independent verification of valve position by licensee personnel.
~
'ABLE SW-3 (Cont'd)
DRAWING NO+
A.
VALVE CHECKLIST VALVE NUMBER LOC ~
BLDG'LEV.
DESCRIPTION REQUIRED POSITION ACTUAL POSITION COMMENTS SW-V-43 DG RM13 DG 1B Return Iso (MO)
SW-V-539 DB RM1B DCW-HX-131 and 1B2 Return Isolation Open Throt.
SW-V-714B DG RM13 PX-15B Root SW-V-748 DG RM13 FIS-15 Root SW-V-749 DG RH13 FIS-15 Root SW-V-217 DG RM1B FIS-15 Root SW-V-217 DG RMlB DCW-HX-132 Inlet (AO)
SW-V-712B DG RH1B PX-14A Root Isolation Closed Open Open Open Closed Closed SW-V-769A DG RM1B FIS-14A Root Isola-tion SW-V-769B DG RH1B FIS-14A Root Isola-tion Open Open SW-V-1139 DG RMlB DMA-CC-21 Upper Out-let Isolation Throt.
f'W-V-104B DG RM13 DMA-CC-21 Lower Out-
/
let Isolation Open SW-V-106B DG RM1B DHA-CC-21 Lower Out-
/
let Isolation Open 9 Requires independent verification of valve position by licensee personnel.
Sealed after flow set per PPM 7.4.7.1.1.2
~
TABLE SW-3 (Cont'd)
DRAWING NO+
Ai VALVE CHECKLIST VALVE NUMBER LOC.
BLDG.
ELEV.
DESCRIPTION REQUIRED ACTUAL POSITION POSITION COMMENTS SW-V-105B Q
DG RMlB DMA-CC-21 Inlet Isolation Open SW-V-107B g
DG RM1B DMA-CC-21 Inlet Isolation Open SW-V-737C SW-V-713B SW-V-10B8 SW-V-770A SW-V-770B SW-V-8B8 SW-V-715 DG RM1B DG RM1B DG RM1B DG RM1B DG RM1B DG RM1B DG Cable RM PX-47A Root SW-P X-14B PMA-CC-22 Return FIS-14B Root FIS-14B Root Closed Closed Throt.
Open Open PX-38 Root Closed DMA-CC-22 Supply Iso Open SW-V-938 DG Cable RM DMA-CC-51 Return Isolation Throt.
f S W-V-787A DG Cable RM FIS-38 Root Open SW-V-787B DG Cable RM FIS-38 Root Open SW-V-90 DG Cable RM DMA-CC-51 Supply Isolation (MO)
Closed 9 Requires independent verification of valve position by licensee personnel.
Sealed after flow set per PPM 7.4.7.1.1.2.
'TABLE SW-3 (Cont'd)
DRAWING NO.
A.
VALVE CHECKLIST VALVE NUMBER SW-V-4C LOC ~
BLDG+
ELEVE DESCRIPTION REQUIRED POSITION HPCS DG Return Isolation (MO) Open RM ACTUAL POSITION COMMENTS SW-V-744 HPCS DG FIS-9 Root Isolation RM Open SW-V-745 HPCS DG FIS-9 Root Isolation Open RM SW-V-709 HPCS DG PX-9B Root RM Closed SW-V-82l3 HPCS DG DCW-HX-1C Return RM Isolation Throt.
SW-V-708 HPCS DG PX-9A Root RM Closed SW-V-899 HPCS DG DCW-HX-1C Supply Open RM Isolation to HPCS DG Cooler SW-V-2228 HPCS DG DG-ENG-1C Supply RM Closed SW-V-809 HPCS DG DCW-HX-1C Supply RM Isolation to HPCS DG Cooler Locked Open SW-V-858 HPCS DG DMA-CC-32 Supply Iso RM Open SW-V-766A HPCS DG FIS-8B Root Isolation Open RM SW-V-766B HPCS DG FIS-8B Root Isolwfl4e Open RM 8 Requires independent verification of valve posit'ion by licensee personnel.
Sealed after flov set per PPM 7.4.7.1.1.2.
TABLE SW-3 (Cont'd)
DRAWING NO+
A.
VALVE CHECKLIST VALVE NUMBER LOC.
BLDG.
ELEVE DESCRIPTION REQUIRED POSITION ACTUAL POS ITION COM".KNTS SW-V-878 HPCS DG DMA-CC-32 Return RM Isolation Throt.
SW-V-711 HPCS DG PX-8B Root RM Closed SW-V-737A HPCS DG PX-45A Isolation RM Closed SW-V-1019 HPCS DG DMA-CC-31 Supply RM Isolation Open SW-V-1038 HPCS DG DMA-CC-31 Supply RM Isolation Open SW-V-1008 HPCS DG DMA-CC-31 Return RM Isolation Open SW-V-1028 HPCS DG DMA-CC-31 Return RM Isolation Open SW-V-889 HPCS DG DMA-CC-31 Return RM Isolation Throt.
SW-V-710 HPCS DG PX-8A Root RM Closed SW-V-765A HPCS DG FIS-8A 'Root RM Open SW-V-765B HPCS DG FIS-8A Root RM Open 9 Requires independent verification of valve position by licensee personnel.
t Sealed after flow set per PPM 7.4.7.1.1.2.
~
~
TtBLE SW-3 (Cont'd)
DRAWING NO.
~
VALVE CHECKLIST VALVE NUMBER LOC.
BLDG.
ELEV.
DESCRIPTION REQUIRED ACTUAL POSITION POSITION COMMENTS SW-V-4A DG RM1A SW-V-746 DG RM1A SW-V-747 DG RM1A SW-V-714 DG RMlA SW-V-5'G RMIA Return Isolation (MO)
FIS-12 Root FIS-12 Root P X-12B Isolation DCW Hx's Return Isolation Open Open Open Closed Throt.
SW-V-214 DG RM1A DCW-HX-1A1 Supply Isolation (AO)
Closed SW-V-220A DG RM1A Q
DCW-TK-1A1, 1A2 Iso Open SW-V-221 A19 DG RM1A DCW-TK-lA1 Iso Closed SW-V-221 A29 DG RM1A DCW-TK-1A2 Iso Closed SW-V-736A DG RMlA SW-V-40AQ DG RMlA PX-12A Root DCW Hx's Supply Isol.ation Closed Open SW-V-3'G RM1A SW-V-215 DG RM1A DG-1A Supply DCW-HX-1A2 Supply Isolation (AO)
Open Closed 9 Requires independent verification of valve position by licensee personnel.
Sealed after flow set per PPM 7.4.7.1.1-2.
TABLE SW-3 (Cont'd)
DRAWING NO+
A.
VALVE CHECKLIST VALVE NUMBER LOC.
BLDG.
ELEV.
DESCRIPTION REQUIRED POSITION ACTUAL POSITION COMMENTS S W-V-7 1 2 A SW-V-767A SW-V-767B SW-V-llA8 DG RM1A PX-11A Root DG RMlA FIS-11A Root DG RM1A FIS-11A Root DG RM1A DMA-CC-11 Return Closed Open Open Throt.
f'W-V-105A Q
SW-V-107A DG RM1A DMA-CC-11 Inlet Isol Open DG RM1A DMA-CC-11 Inlet Isol Open SW-V-106A Q
DG RM1A DMA-CC-11 Lower Outlet Isol Open SW-V-737B DG RM1A DMA-CC-11 Inlet PX Isol Closed SW-V-104A Q
DG RM1A DMA-CC-11 Upper Ou let Isolation
, Open SW-V-713A SW-V-10AQ DG RM1A PX 11B Root DG RM1A DMA-CC-12 Peturn Isol.ation Closed Throt.
SW-V-768A SW-V-768B SW-V-8A9 DG RM1A FIS-llB Root Isol DG RM1A FIS-11B Root Isol Open Open DG RM1A DMA-CC-12 Supply Isol Open 9 Requires independent verification of valve position by licensee personnel.
Sealed after flow set per PPM 7.4.7.1.1.2.
TABLE SW-3 (Cont'd)
DRAWING NO.
A.
VALVE CHECKLIST VALVE NUMBER SW-V-1139 LOC BLDG.
ELEV.
RB-572 DESCRIPTION RRA-CC-14 Outlet Isolation REQUIRED POSITION Throt.
t ACTUAL POSITION COMMENTS SW-V-730 SW-V-796A SW-V-796B S W-V-1128 RB-572 PX-61 Root RB-572 FIS-61 Root RB-572 FIS-61 Root Closed Open Open RB-572 RRA-CC-14 Supply Iso Open SW-V-128A Q
RB-572 CAC-1A Outlet Iso Throt. f SW-V-126A Q
RB-572 CAC 1A Aftercooler Inlet Isolation Open SW-V-127A g
RB-572 CAC 1A Scrubber Inlet Open Isolation Sw-V-1158 SW-V-1148 RB-572 RRA-CC-13 Return Iso Throt.
f RB-572 RRA-C&13 Supply Iso Open SW-V-1278 8
SW-V-126B Q
RB-572 RB-572 CAC-1B Scrubber Supply Isolation CAC-1B Aftercooler Isolation Open Open SW-V-128B Q
RB-572 CAC-1B Return Iso Throt.
9 Requires independent verification of valve position by licensee personnel.
Sealed after flov set per PPM 7.4.7.1-1.2.
TABLE SW-3 (Cont'd)
DRAWING NO+
1
~ g Ao VALVE CHECKLIST VALVE NUMBER LOC.
BLDG+
ELEVE DESCRIPTION REQUIRED ACTUAL POSITION POSITION COUNTS SW-V-798A RB-572 SW-V-798B RB-572 S'W-V-762B RB-572 SW-V-763B RB-572 SW-V-762A RB-572 SW-V-763A RB-572 SW-V-75A RB-522 SW-V-75B RB-522 SW-V-187B RB-548 SW-V-188B RB-548 SW-V-187A RB-548 SW-V-188A RB-548 FIS-69B Root Isol FIS-69B Root Isol SW-SR-43 Return Isol SW-SR-43 Supply Isol SW-SR-42 Supply Isol SW-SR-42 Return Isol Emergency Makeup to Fuel Pool (MO)
Emergency Makeup to Fuel Pool (MO)
FPC-HX-1B Inlet Iso (MO)
FPC-HX-1B Outlet Iso (MO)
FPC-HX-1A Inlet Iso (MO)
FPC-HX-lA Outlet Iso (MO)
Open Open Open Open Open Open Closed Closed Closed Closed Closed Closed SW-V-1908 SW-V-1898 RB-548 RB-548 RRA-CC-19 Outlet Iso RR-CC-19 Inlet Iso Open Open 9 Requires independent verification of valve position by licensee personnel.
Sealed after flow set per PPM 7-4.7.1.1.2.
\\
TABLE SW-3 (Cont'd)
DRAWING NO+
A.
VALVE CHECKLIST VALVE NUMBER LOG ~
BLDG+
ELEVE DESCRIPTION REQUIRED ACTUAL POSITION POSITION COMMENTS SW-V-184B RB-548 Q
RRA-CC-19 Outlet Throt.. f SW-V-1858 RB-548 SW-V-1919 RB-548 SW-V-1929 RB-548 SW-V-1848 RB-548 SW-V-1938 RB-548 SW-V-1338 RB-548 SW-V-1329 RB-548 SW-V-137 RB-548 SW-V-138 RB-548 RRA-CC-19 Inlet Iso RRA-CC-20 Inlet Iso RRA-CC-20 Outlet Iso RRA-CC-20 Outlet Iso RRA-CC-20 Inlet Iso RRA-CC-15 Return Iso RRA-CC-15 Supply Iso IR-13
& IR-20 Supply Isol. Deactivated IR-13
& IR-20 Return Isol. Deactivated Open Open Open Throt.
f'pen Throt. f Open Closed Closed SW-V-1358 RB-548 SW-V-1349 RB-548 SW-V-140 RB-548 RRA-CC-17 Return Iso RRA-CC-17 Supply Iso SR-14
& 21 Return Isol. Deactivated Throt.
Open Closed SW-V-139 RB-548 SR 14
& 21 Supply Isol. Deactivated Closed 8 Requires independent verification of valve position by licensee personnel.
Sealed after flow set per PPM 7.4.7.1.1.2.
TABLE SW-3 (Cont'd)
DRAWING NO.
A.
VALVE CHECKLIST VALVE NUMBER LOG ~
BLDG.
ELEVE DESCRIPTION REQUIRED ACTUAL POSITION POSITION COKfENTS SW-V-1758 RB-548 Demineralized Water Isolation Closed RHR-V-14B RB-560 Q
RHR-HX-1B Serv Water Inlet Throt.
RHR-V-68B RB-551 SW-V-733A RB-548 SW-V-732B RB-548 SW-V-731B RB-548 SW-V-735A RB-548 SW-V-734A RB-548 SW-V-756B RB-548 SW-V-731A RB-548 SW-V-732A RB-548 SW-V-733B RB-548 SW-V-175A RB-548 RHR-HX-1B Serv Water Outlet (MO)
PX-5B Root FT 7B Root FT 7B Root PT-38B Root PT-6B Root SP-63B Isolation FT 7A Root FT 7A Root PX-5A Root Demineralized Water Isolation Open Closed Open Open Open Open Open Open Open Closed Closed RHR-V-14A RB-558 Q
RHR-HX-1A Serv Water Inlet Throt.
RHR-V-68A RB-553 RHR-HX-1A Serv Water Outlet (MO)
Open 9 Requires independent verification of valve position by licensee personnel.
Sealed after flov set per PPM 7.4-7.1.1.2.
STABLE SW-3 (Cont'd)
DRAWING NO.
A.
VALVE CHECKLIST VALVE NUMBER SW-V-735B SW-V-734B SW-V-756A SW-V-75BB 8
LOC BLDG.
ELEV.
DESCRIPTION RB-530 Loop B Manual Isol.
to FPC RB-548 PT-38A Root RB-548 PT-6A Root RB-548 SP-63A Isolation REQUIRED POSITION Open Open Closed Closed ACTUAL POSITION
~t COMMENTS SW-V-75AA 8
RB-530 Loop A Manual Isol.
to FPC Closed SW-V-958 SW-V-968 SW-V-918 sw-v-948 SW-V-684 SW-V-17C8 SW-V-777A SW-V-777B SW-V-719C SW-V-16C8 SW-V-348 RB-522 RRA-CC-11 Supply Iso RB-522 RRA-CC-11 Return Iso'B-522 RRA-CC-10 Inlet Iso RB-522 RRA-CC-10 Return Iso R-480 PSR-SR-47 Isol RB-446 RRA-CC-1 Return Iso RB-446 FIS-23C Root Valve RB-446 FIS-23C Root Valve RB-446 PX-23C Isolation
'B-446 RRA-CC-1 Supply Iso RB-446 RRA-CC-6 Return Iso Open Throt.
f'pen Throt.
Open Throt.
f Open Open Closed Open Open 8 Requires independent verification of valve position by licensee personnel.
Sealed after flow set per PPM 7.4.7.1.1.2.
TABLE SW-3 (Cont'd)
DRAWING NO.
Ao VALVE CHECKLIST VALVE NUMBER LOC.
BLDG+
ELEVE SW-V-278 RB-446 SW-V-479 RB-446 SW-V-780A RB-446 SW-V-780B RB-446 SW-V-718 RB-446 SW-V-268 RB-446 SW-V-548 RB-446 DESCRIPTION RRA-CC-6 Return Iso (MO)
FIS-29 Root Valve FIS-29 Root Valve PX-29 Isolation RRA-CC-6 Supply Iso RRA-CC-4 Cooler Outlet Isolation RRA-CC-4 Cooler Outlet Isol.
(MO)
REQUIRED ACTUAL POSITION POSITION COMMENTS Throt.. t Open Open Closed Open Open Throt. f SW-V-779A RB-446 SW-V-779B RB-446 SW-V-716 RB-446 SW-V-468 RB-446 SW-V-778A RB-446 SW-V-778B RB-446 SW-V-717 RB-446 SW-V-36 RB-446 FIS-27 Root Valve FIS-27 Root Valve PX-27 Root Valve RRA-CC-4 Inlet Iso FIS-25 Roo" Valve FIS-25 Root Valve PX-25 Root Valve RRA-CC-5 Inlet Iso Open Open Closed Open Open Open Closed Open 9 Requires independent verification of valve position by licensee personnel.
Sealed after flow set per PPM 7.4-7.1.1.2.
'ZABLE SW-3 (Cont'd)
DRAWING NO.
A.
VALVE CHECKLIST VALVE NUMBER LOG ~
BLDG.
ELEVe SW-V-37/
RB-446 SW-V-22C8 RB-421 SW-V-721C RB-421 SW-V-773A RB-421 SW-V-773B RB-421 SW-V-17A8 RB-446 SW-V-775A RB-446 SW-V-775B RB-446 SW-V-719A RB-446 SW-V-719B RB-446 Q
SW-V-16A8 RB-446 SW-V-17B RB-446 SW-V-776A RB-446 SW-V-776B RB-446 DESCRIPTION RRA-CC-5 Return Iso C RHR Pump Motor Cooler Supply Isol.
PX-17C Roo" Valve FIS-17C2 Robt Valve FIS-17C2 Root Valve RRA-CC-2 Return Isol FIS-23A Root Valve FIS-23A Root Valve PX-23A Root Valve RRA-CC-2 Supply Isol RRA-CC-3 Return Isol FIS-23B Root Valve FIS-23B Root Valve PX-238 Roo" Valve REQUIRED ACTUAL POSITION POSITION COMlKNTS Throt.
Open Closed Open Open Throt.
Open Open Closed Open Throt.
Open Open Closed SW-V-16B RB-446 SW-V-24AQ RB-421 SW-V-'721A RB-421 SW-V-722 RB-421 RRA-CC-3 Supply Isol A RHR Room Return Iso PX-17A Root Valve PX-39 Root Open Open Closed Closed 8 Requires independent verification of valve position by licensee personnel.
Sealed after flow set per PPM 7 4.7 ~ 1 ~ 1 2.
TABLE SW-3 (Cont'd)
DRAWING NO+
A.
VALVE CHECKLIST VALVE NUMBER LOC ~
BLDG.
ELEVE SW-V-23A RB-421 DESCRIPTION A RHR Pump Motor Return Isolation REQUIRED POSITION Throt I'CTUAL POSITION COMMENTS SW-V-771A RB-421 SW-V-771B RB-421 SW-V-22A8 RB-421 SW-V-24BQ RB-421 SW-V-721B RB-421 SW-V-2389 RB-421 A RHR Pump Motor Supply Isolation Open B RHR Room Return Iso Open PX-17B Root B RHR Pump Motor Cooler Return Isol Closed Throt.
FIS-17A2 Roo" Valve Open FXS-17A2 Root Valve Open SW-V-772A RB-421 SW-V-772B RB-421 SW-V-22B9 RB-421 B RHR Pump Motor Supply Isolation Open FIS-17B2 Root Valve Open FIS-17B2 Root Valve Open SW-V-488 RB-421 SW-V-499 RB-421 LPCS Pump Motor Cooler Supply Isol LPCS Pump Motor Cooler Return Isol.
Open Throt.
SW-V-774A RB-421 SW-V-774B RB-421 FIS-19 Root Valve FIS-19 Root Valve Open Open 9 Requ'ires independent verification of valve position by licensee personnel.
1'ealed after flow se" per PPM 7.4 7.1.1 2.
TABLE SW-3 (Cont'd)
DRAWING NO.
Ao VALVE CHECKLIST LOG ~
VALVE NUMBER BLDG+
ELEVE DESCRIPTION REQUIRED POSITION ACTUAL POSITION COMMENTS SW-V-720 RB-421 PX-19 Root Valve Closed SW-V-448 RB-421 LPCS Pump Room Return Isolation Open SW-V-24'B-421 SW-V-23C8 RB-421 C
RHR Room Return Isolation C
RHR Pump Motor Cooler Return Isol Open Throt. t SW-V-979 RX Bldg RRA-CC-12 Inlet Iso 522 Open SW-V-848 R-47 1 Iso from PSR-SR-47 Open SW-V-155R R-522 Q
PSR-SR-47 Return Iso RRA-CC-12 Outlet Open SW-V-989 R-485 RRA-CC-12 Re turn SW-V-847 R-471 Supply to PSR-SR-47 SW-V-688 R-471 Supply to PSR-SR-47 Throt.
Open Open SW-V-840 R-47 1 SW-Y-842 R-47 1 SW-V-844 R-47 1 SW-V-846 R-47 1 PSR Cooling Supply from Loop A (SO)
PSR Cooling Supply from Loop B (SO)
PSR Cooling Supply from Loop A (SO)
PSR Cooling Supply from Loop B (SO)
Closed Closed Closed Closed Controlled from RX 487'SR Rm Controlled from RX 487'SR Rmo Controlled from RX 487'SR Rm.
Controlled from RX 487'SR Rm.
8 Requires independent verification of valve position by licensee personnel.
Sealed after flow set per PPM 7.4.7.1.1
~2.
TABLE SW-3 (Cont'd)
DRAWING NO.
B.
SYSTEM POWER SUPPLY CHECKLIST POWER SUPPLY SM-7 LOG ~
BLDG+
ELEVE W-467 DESCRIPTION SW-P-1A REQUIRED ACTUAL POSITION POSITION Racked In CO>MENTS SM-8 W-467 SW-P-1B Racked In MC-7A MC-7A MC-7A MC-7A MC-BA MC-8A HC-8A MC-8A MC-8A-A CKTSE MC-4A MC-4A MC-4A MC-4A W-467 W-467 W-467 W-467 W-467 W-467 W-467 W-467 D-441 D-441 D-441 D-441 D-441 SW-V-2A SW-V-12A SW-V-69A SW-V-70B C
SW-V-2B SW-V-12B SW-V-69B SW-V-70A SW-V-90 POA-FN-2A SW-V-4C SW-V-29 SW-V-54 Closed Open +
Open +
Open +
Closed Open +
Open +
Open +
Closed Closed Closed Closed Closed
+ Thes'e breakers are to be locked open and tagged.
TABLE SW-3 (Cont'd)
DRAWING NO.
'OWER SUPPLY LOC.
BLDG.
ELEVE DESCRIPTION SYSTEM POWER SUPPLY CHECKLIST REqUIRED POSITION ACTUAL POSITION COMMENTS MC-4A D-441 MC-7B MC-7B R-522 R-522 MC-7A-A D-441 MC-7A-A D-441 MC-7A-A D-441 MC-7A-A D-441 PP-7A-B84 A-441 MC-7A-A D-441 MC-7A-A D-441 MC-8A-A D-441 MC-8A-A D-441 MC-8A-A D-441 MC-8A-A D-441 MC-8A-A D-441 PP-8A-B84 B-441 HPCS-P-2 PRA-FN-lA PRA-EUH-1A PRA-EUH-2A PRA-EUH-3A PRA-EUH-4A SW-V-4A SW-V-4B PRA-FN-1B POA-FN-2B PRA-EUH-1B PRA-EUH-2B PRA-EUH-3B PRA-EUH-4B SW-V-24A SW-V-44 Closed Closed Closed Closed Closed Closed Closed Closed Closed Closed Closed Closed Closed Closed Closed Closed
TABLE SW-3 (Cont'd)
DRAWING NO+
B.
SYSTEM POWER SUPPLY CHECKLIST POWER SUPPLY LOC BLDG'LEV.
DESCRIPTION REQUIRED POSITION ACTUAL POSITION COMMENTS PP-8A-E CKT 10E R-471 SW-V-34 Normal Supply Closed PP-8A-F CKT 10 RW-467 SW-V-34 Emergency Supply Closed MC-8B-A MC-8B-A MC-7B-B MC-8B-B MC-7B-A MC-7B-A MC-7B-B MC-8B-A MC-8B-A MC-8B-B CS-RSIP-1 R-522 SW-V-24B R-522 SW-V-24C R-572 SW-V-68A R-572 SW-V-68B R-522 SW-V-75A R-572 SW-V-187A R-522 SW-V-188A
'-522 SW-V-753 R-522 SW-V-187B R-572 SW>>V-188B RW-467 Instrument Cont Pwr
'FER SW Closed Closed Closed Closed Closed Closed Closed Closed Closed Closed
'ff FRTS-3 RW-467 Iso SW for SW-V-4B Norm
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622 I57 10A
WASHINGTON NUCLEAR PLANT NO.
2 GENERIC PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN Primary Containment Systems TABLE PRCT-1 IMPORTANCE 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 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.
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 main containment systems considered here are the Primary Containment, Drywell Sprays, the Suppression Pool, the Secondary Containment and the Standby Gas Treatment System (SGTS).
The Primary Containment serves to contain fission products post-accident and consists of the drywell and suppression pool walls and the containment isolation system.
From a risk standpoint, small leaks are not a ma)or concern, but large leaks, on the order of the tech spec limit (La) or larger can be very significant to risk.
Drywell Sprays are important to ensure H2 mixing, pressure reduction and removal of fission products from the atmosphere.
The Suppression Pool serves to limit containment pressure, scrub fission products, and provide a source of water for the ECCS-The Secondary Containment and the SGTS work together to help ensure any leakage from the Primary Containment is either captured or released through an elevated, filtered vent path.
The SGTS also has area coolers to limit temperatures in the various ECCS equipment areas.
Since there are no detailed fault trees or minimal cutsets in most PRAs for these containment
- systems, this inspection plan does not have the detail of the front end systems.
Important Aspects of Containment S stems l.
Emergency procedures and training established to vent containment wetwell (Suppression Pool) during containment overpressure scenarios.
bA,
~" i1
~ ~
TABLE PRCT-1 (Cont 'd)
CONDITIONS THAT CAN LEAD TO FAILURE 2.
Maintaining functionability of SP due to its multiple uses of condensing
- steam, scrubbing fission products, and providing the ECCS with water.
3 ~
Sequences that bypass containment or result in early containmen" failure are important to offsite risk.
Important areas associated with bypass of containment are containment isolation failure due to valve failures (such as Reactor Building to wetwell vacuum breakers) or gross leakage through containment.
4.
One important area associated with early containment overpressure is SP bypass.
(The SP bypass leak test addresses this.)
Failure modes leading to SP bypass include leakage of the drywell to wetwell vacuum breakers, and rupture ox leaks in the downcomer pipes.
The drywell to SP differen-tial pressure monitoring instrumentation is important in monitoring the integrity of these boundaries.
5.
Suppression Pool Mater Unavailable Due to Rupture:
Loss of the Suppres-sion Pool water inventory prevents the reactor vessel from being cooled by the ECCS systems.
To minimize this occurrence, the results of the containment integrated leak rate tests can be reviewed to determine if any leaks in the Suppression Pool had been identified.
Also, plant records could be reviewed to detect any abnormal nitrogen usage which might indicate leakage of the pool in the inerted containment and/or a
visual inspection of the pool could be conducted.
6.
Severe accident research has shown that a very important function of dry-well sprays is the scrubbing of fission products during the core melt process.
This limits the inventory available for release.
Through pro-
- cedures, test results, and system reviews, inspectors should ensure the post-accident availability of drywell sprays.
WASHINGTON NUCLEAR PLANT NO 2
GENERIC PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN Primary Containment Systems TABLE PRCT-1 IMPORTANCE 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 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.
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 main containment systems considered here are the(Primary Containment,lil","dpi
',i'".'I.-:
'~o The Primary Containment serves to contain fission products post-accident and consists of the drywell and suppression pool walls and the containment isola-tion system.
From a risk standpoint, small leaks are not a major concern, bu" large leaks, on the order of the tech spec limit (La) or larger can be very significant to risk.
Drywell Sprays are important to ensure Q mixing, pres-sure reduction and removal of fission products from the atmosphere.
The Sup-pression Pool serves to limit containment
- pressure, scrub fission products, and provide a source of water for the ECCS.
The Secondary Containment and the SGTS work together to help ensure any leakage from the Primary Containment is either captured or released through an elevated, filtered vent path.
The SGTS also has area coolers to limi" temperatures in the various ECCS equipment areas.
Since there are no detailed fault trees or minimal cutsets in most PRAs for these containment
- systems, this inspection plan does not have the detail of the front end systems.
Im ortant Aspects of Containment S stems l.
Emergency procedures and training established to vent containment wetwell (Suppression Pool) during containment overpressure scenarios.
hfdf r '
TABLE PRCT-1 (Cont'd)
CONDITIONS THAT CAN LEAD TO FAILURE 2 ~
Haintaining functionability of SP due to its multiple uses of condensing
- steam, scrubbing fission products, and providing the ECCS with water.
3 ~
Sequences tha" bypass containment or result in early containment failure are important to offsite risk.
Important areas associated with bypass of containment are con ainment isolation failure due to valve failures (such as Reactor Building to wetwell vacuum breakers) or gross leakage through containment.
4.
One important area associated with early containment overpressure is SP bypass.
(The SP bypass leak test addresses this.)
Failure modes leading to SP bypass include leakage of the drywell to wetwell vacuum breakers, and rupture or leaks in the downcomer pipes.
The drywell to SP differen-tial pressure@nitoring instrumentation is important in monitoring the integrity of these boundaries.
5 ~
Suppression Pool Water Unavailable Due to Rupture:
Loss of the Suppres-sion Pool water inventory prevents the reactor vessel from being cooled by the ECCS systems.
To minimize this occurrence, the results of the containment integrated leak rate tests can be. reviewed to determine if any leaks in the Suppression Pool had been identified.
Also, plant records could be reviewed to detect any abnormal nitrogen usage which migh indicate leakage of the pool in the inerted containment and/or a
visual inspection of the pool could be conducted.
6.
Severe accident research has shown that a very important function of dry-well sprays is the scrubbing of fission products during the core melt process.
This limits the inventory available for release.
Through pro-
- cedures, test results, and system reviews, inspectors should ensure the post-accident availability of drywell sprays.
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RELIEF DRYWELL DRYWELL
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OUTSIDE WEST WALL OUTSIDE SOUTH WALL OUTSIDE NORTH WALL OUTSIDE EAST WALL HP LP OPT HP LP DPT HP LP DPT HP LP DPT AH DPR r
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LYS LAG DPIC 18 cIPS 28 AL TO REA-FN-18 AND ROA-FN-1 8 18 18 18 TO REA-EP-1 8 TO SGT-FN-18-2 TO SGT-FN-1A-2 FlCGcRE 7
- SECONDARY CONTAtNMENTPRESSURE CONTROL SYSTEM "8" 851264.2A DEC 1885
OUTSIDE WEST WALL OUTSIDE SOUTH WALL OUTSIDE NORTH WALl.
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- SECONDARY CONTAINMENTPRESSURE CONTROL SYSTEM "A" 851264.3A DEC 1985
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. (SECTION VIEW - FACING EAST) 00 1 2028
t
WASHIiVGTON NUCLEAR PLANT NO.
2 GENERIC PROBABILZSTIC RISK ASSESSMENT-BASED INSPECTION PLAN High Pressure Core Spray (HPCS)
System TABLE HPCS-1 IMPORTANCE BASIS AND FAILURE MODE IDENTIFICATION CONDZTIONS THAT CAN LEAD TO FAILURE 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 information 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.
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 LATER Failure Conditions 1.
HPCS Pump P-1 or Diesel (DG-3) Fails to Start and Run or Fails on Restart.
See Table EP-4 for proposed inspection plan of diesel-generators.
TABLE HPCS-1 (Cont'd)
CONDITIONS THAT CAN LEAD TO FAILURE 2.
HPCS in Maintenance.
(MT,TS) 3.
Failure to Maintain HPCS Discharge Fill System Pump HPCS-P-3 in Operation Causing Piping Damage Upon Start. of HPCS Pump P-1 (OP) 4.
Loss of Service Water to HPCS Diesel.
The HPCS diesel is designed to be capable of operatin without service water to the engine heat exchangars for a minimum of one
( f) minu-e, fully loaded.
Loss of SW can be caused by failure of:
HPCS Service Water Pump HPCS-P-2 HPCS Water Leg Pump SW-MOV-4C Fails to Open 5.
HPCS Pump P-1 Discharge Line MOV-V-4 Fails to Open or No Output from Contx'ol Circuit. (PC,PT,MT) 6.
Lube Oil Cooler Heat Exchanger Leakage or Oil Pumps Fail to Supply Starting Lube Oil. (PT,MT)
Engine Driven lube-oil pumps for:
Main Lubricating Oil System (P-7)
Piston Cooling System (P-8)
Scavenging Oil System (P-9)
An independent AC circulating pump and standby DC motor pump provide pre-lubrication of the tuxbocharger before and run continuously during shutdown while an immersion lube oil cooler water side maintains oil temperature at shutdown.
driven soak back engine start-up heater in the 135'F during 7.
Loss of HPCS Pump Motor or HPCS Diesel Room Cooling. (PT)
Service Water Valve MOV-54 failing to open prevents cooling water to HPCS Pump Room (RRA-C-4) ~
8.
Initiation and Isolation Logic Testing Outage (PT)
Initiation signals:
1.65 psig drywel.l pressuxe Reactor water level 2 (-50")
Manual pushbut on Undervoltage on SM-4 (HPCS diesel)
HPCS suction auto shift to suppression pool on high pool level 466'-8" or CST level at 448'-3" HPCS diesel trip functions HPCS pump/motor room HVAC/HiRad isolation functi'ons (?)
ms ~,
4
'f' ~
~
m ss
~
TABLE HPCS-1 (Cont'd)
CONDITIONS THAT CAN LEAD TO FAILURE 9.
False Signals or Miscalibration of Sensors.
(PC,PT)
Applies to same function>as in 8 above.
10.
Failure of Relay Logic for Suppression Pool or CST Level. (PC,PT)
See 8 above.
11.
HPCS No" Reset for Auto Operation.
(OP) 12.
Plugging of Spargers/Clogged Suppression Pool Strainers on HPCS Pump Suction Line.
(PT,MT,OP) 13'ailure to Properly Restore Components After Test or Maintenance.
HPCS-V-51, in)ection line isolation valve (inside drywall)
HPCS-V-5 testable check valve (inside drywel.l)
HPCS-V-1 suction MOV from CST COND-V-9A,9B condensate supply (from CST) COND-P-3.4.5 14.
Loss of HPCS Safety Function by Failure to Follow Procedures Properly or Inadequacy of Procedures.
(OP,TS)
WASHINGTON NUCLEAR PLANT NO..'Q GENERIC PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN High Pressure Core Spray (HPCS)
System TABLE HPCS-2 ISE INSPECTION PROCEDURES FOR SYSTEM OPERATION PROCEDURE NUMBER 61225 61726 71707 71710 TITLE Surveillance Testing and Calibration Program Monthly Surveillance Observation Operational Safety Verification ESF System Walkdown COMPONENTS HPCS pump/motor, HPCS
- diesel, MOUs, HPCS-U-4, SW-V-4, SW-U-4C, SW-V-54 HPCS water leg pump, HPCS discharge fillpump, CST valves V-9A, 9B; HPCS-V-1, HPCS-V-51, V-5 FAILURE MODES 1,4-10$
12, 13 62702 Maintenance 62703
'onthly Maintenance Observation 53051 Instrument Components and Systems-Procedure Review Level, pressure temperature sensors 5,7-10 53053 53055 56700 41700 Instrument Components and Systems-Work Observation Instrument Components and Systems-Record Review
, Calibration Training All 9-14
~
~
I WASHINGTON NUCLEAR PLANT N0.2 GENERIC PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN Reactor Core Isolation Cooling (RCXC) System TABLE RCIC-1 IMPORTANCE BASIS AND FAILURE MODE IDENTIFICATION CONDITIONS THAT CAN LEAD TO FAILURE General Guidance Surveillance of the licensee's periodic testing and/or preventive or unscheduled maintenance activities and procedures and/or normal and emergency operating procedures,
- training, and check-off lists in accordance with the Technical Specifications and relevant NRC bulletins and information 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.
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.
TS Technical specifications.
Mission Success Criteria LATER Failure Conditions 1.
RCIC Pump P-1 or RCIC Turbine DT-1 Fails to Start and Run or Pails on Re-start.
RCIC in Maintenance.
(MT,TS) 3.
Failure to Maintain RCIC Water Leg Pump RCIC-P-3 in Operation Causing Damage Upon Start of RCIC Pump P-l.
(OP)
Can also be caused by valves failing to remain open.
~
V-60 Water Leg Pump Suction Valve
~
V-62 Water Leg Pump Discharge Valve
~
V-62 Water Leg Pump Recirculation Valve
4 Leakage or Rupture of Turbine Exhaust Line Rupture Disks. (ISI,PT)
L q ~
g r 5.
RCIC'ump Suction /Discharge Line MOV Fails to Open or No Output from Control Circuit. (PT,MT)
~
V-31 RCIC-P-1 Suppression Pool Suction Isolation MOV
~
V-13, Head Spray Line Isolation MOV
~
V-65 Testable (AO) Check Valve to Reactor Head Spray 6.
RCIC Steam Supply Isolation MOV Fails to Open or No Output from Control Circuit. (PT,MT)
V-45, Steam to RCIC-P-1 Turbine Isolation MOV 7.
Lube Oil Cooler HX Leakage or Oil Pump Fails to Supply Starting Lub Oil.
Lube Oil Cooling Mater Isolation MOV, V-46~ Fails to Open or No Output from Control Circuit. (PT,MT)
~
V-46 Auxiliary Cooling Water Supply Isolation MOV 9.
Initiation and Isolation Logic Testing Outage.
(PT,TS)
Initiation:
Reactor Water Level 2 (-50")
Manual pushbutton Isolation:
RCIC Equipment Area and/or Pipe Routing Area High Temperature (130'F)
RCIC Equipment Area High Differential Temperature (40'F)
Low Steam Supply Pressure (62 psig)
Exhaust Diaphragm High'Pressure (10 psig)
RCIC High Steam Flow or Instrument Line Break (283%)
Combined RCIC and RHR High Steam Flow (118%)
~
Manual pushbutton on P-601
~
Overspeed mechanical (125%)
~
Overspeed electrical (125%)
~
Turbine high exhaust pressure (25 psig)
~Low pump suction pressure (20" VAC)
~
Local manual trip
- 10. False Signals or Miscalibration of Sensors.
(PC,PT)
Applies to same functions as in 9 above.
~ I
~
I
\\ g \\,
~
ll. Failure of Relay Logic for Suppression Pool or CST Level.
(PC,PT)
Suppression pool suction valve V-31 opens upon CST level of 448'-3".
CSQ suction valve RCIC-V-10 and CST flow test return valves RCIC-V-22 and V-99 to CST auto close when V-31 is open.
- 12. Loss of Pump/Turbine Room Cooling.
Service Water Valve MOV-34 failing to open causes loss of RCIC pump room cooling coil RRA-CC-6.
13.
HPCI Not Reset for Auto Operation.
(OP)
- 14. Plugging of Spargers/Clogged Suppression Pool Strainers on HPCS Pump Suc'tion Line (PT yMT y OP)
- 15. Failure to Properly Restore Components After Test or Maintenance.
~
RCIC pump discharge/suction line valves V-66, V-65 (testable check valves)
V-12, V-101 (locked open manual valves)
~
RCIC turbine steam supply/discharge line valves V-63, V-8, V-68 (MOVs normally open) 0~ cern@ em~ Qs s4AR anioh4, Roc-9i 16.
Loss of RCIC Safety Function by Failure to Follow Procedur+Properly or Inadequacy of Procedures.
(OP,TS)
- 17. Operator Failure to Extend Battery Life, Water Sources, for RCIC Operatin and Room Cooling Under Extended Station Blackout.
(OP)
- 18. Environmental Effects Due to Steam Line Break. (ISI)
WASHINGTON NUCLEAR PLANT N0.2 GENERIC PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN 5
,/
Reactor Core Isolation Cooling (RCIC) System TABLE RCIC-2 I&E INSPECTION PROCEDURES FOR SYSTEM OPERATION PROCEDURE NUMBER TITLE COMPONENTS*
FAILURE MODES 41700 Training RCIC pump
& turbine 3,13,14, temperature, pressure 16,17
- level,
& speed sensors area temperature
- sensors, lube oil cooler isolation valve.
53051 53053 53055 56700 Instrument Component and.
Systems-Procedure Review Instrument Components and Systems-Work Observation Instrument Components and Systems-Record Review Calibration Temperature, pressure 5,6,9-ll
- level, speed sensors.
Area temperture sensors.
61725 61726 72700 71707 71710 Surveillance and
'alibration Program Monthly Surveillance Observation Startup Testing-Refueling Operational Safety Verification ESF System Walkdown RCIC pump
& turbine 1-2,4-12 temperature, pressure 14-15,18 level
& speed
- sensors, area temperture sensors RCIC discharge valve V-13, lube oil cooling water valve V-46, lube oil cooling water PCV-15, turbine stop valve V-l, turbine control valve, V-2, steamline isolation valves V-45, V-36.
- Refers only to components identified in Tables RCIC-1 and RCIC-3
WASHINGTON NUCLEAR PLANT N0.2 GENERIC PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN Reactor Core Isolation Cooling (RCIC) System DRAWING NO:
M-519 A.
VALVE CHECKLIST TABLE RCIC-3 MODIFIED SYSTEM WALKDOWN VALVE NUMBER LOC.
BLDG.
ELEV.
DESCRIPTION REQUIRED POSITION ACTUAL POSITION COGENT S RCIC-V-60 RB-422 g
Water leg Pump Suction Valve Open RCIC-V-62 RB-422 Q
RCIC-V-67 RB-422 Q
Water Leg Pump Discharge Valve Water Leg Pump Recirc. Valve Open Open RCIC-V-16 RB-422 RCIC P-1 Suction Iso Q
Locked/
Open RCIC-V-701 RB-422 RCIC-P-5 Root Iso Open RCIC-V-704 RB-422 RCIC-P-1 Suction Pressure PS-6 Root Isolation Open RCIC-V-619 RCIC-V-1018 RCIC-V-728 RB-422 RB-422 RB-422 Vent on Suction on Water Leg Pump RCIC-P-1 Discharge Isolation RCIC-P-1 Discharge Inst.
Header Isol Closed Locked/
Open Open RCIC-V-729 RB-422 Root Isolation RCXC Flow Trans RCIC-FT-3 Open RCIC-V-730 RB-422 RCIC-FT-3 Root Isol Open RCIC-V-12 RB-422 Q
Supply to Reactor Head Manual Isol Locked/
Open 8 Requires independent verification of valve position by licensee personnel.
TABLE RCIC-3 (Cont'd)
DRAWING NO:
M-519 A.
VALVE CHECKLIST VALVE NUMBER LOC.
BLDG.
ELEV.
DESCRIPTION REQUIRED ACTUAL POSITION POSITION COMMENTS RCIC-V-46 RB-422 'ux. Cooling Water Supply Isol (MO)
Closed RCIC-V-722 RB-422 RCIC-PCV-15 Sensing Isolation Open RCIC-V-725 RB-422 RCIC-P S-l Roo t Isol (Cooling Supply)
Open RCIC-V-10 RB-422 Q
Suction from Cond.
Storage Tanks (MO).
Open RCIC-V-721 RB-422 RCIC-PS-13 Root Isol (Vacuum Tank Press)
Open RCIC-V-49 RB-422 RCIC-P-4 Condensate Pump Discharge Throt.
Valve Open RCIC-V-1 RB-422 RCIC-P-1 Trip Throt.
Valve (MO)
Open Mechanical Trip Latched RCIC-V-2 RB-422 RCIC-P-1 Governor Valve (HO)
Open RCIC-V-45 RB-422 Steam to RCIC-P-1 Turbine Isol (MO)
Closed RCIC-V-739 RB-422 RCIC-PT-7 Root Isol Open RCIC-V-751A RB-422 RCIC-LS-10 Root Isol Open RCIC-V-751B RB-422 RCIC-LS-10 Root Isol Open RCIC-V-38 RB-422 Q
RCIC-T-3 Upstream Isolation Open 9 Requires independent verification of valve position by licensee personnel.
~
~
TABLE RCIC-3 (Cont'd)
DRAWING NO:
M-519
~,
VALVE CHECKLIST VALVE NUMBER RCIC-V-39 Q
LOG ~
BLDG.
ELEVE RB-422 DESCRIPTION RCIC-T-3 Downstream isolation REQUIRED POSITION Open ACTUAL POSITION COMMENTS RCIC-V-25 RCIC-V-26 RCIC-V-24 RCIC-V-29 RB-422 RB-422 RB-422 RB-422 Supply Steam Drip Pot Outlet Drain Valve (AO)
Supply Steam Drip Pot Outlet Drain Valve (AO)
RCIC Turbine Gland Exhaust Isolation RCIC Turbine Gland Exhaust Isolation Open Open Open Open RCIC-V-712 RCIC-V-707 NCxc-V-7lo RCIC-V-711 RCIC-V-20 Q
RCIC-V-27 Q
RB-422 RB-422 RB-422 RB-422 RB-422 RCIC-PT-3 Root Iso (Gland Exhaust)
RCIC-PT-8 Root Iso (Turbine Exhaust)
(SAme.~ 7il),
RCIC-LS-3 Root Iso (Turbine Exhaust Drip Pot)
RCIC-T-4 Upstream Iso RCIC-T-4 Downs tream Isolation Open Open Open Open Open RCIC-V-708 RB-422 RCIC-PS-9A Root Isolation (Turbine Exhaust)
Open 9 Requires independent verification of valve position by licensee personnel.
TABLE RCIC-3 (Cont'd)
DRAWING NO:
N-519 Ae VALVE CHECKLIST VALVE NUMBER LOC.
BLDGe ELEVe DESCRIPTION REQUIRED ACTUAL POSITION POSITION COMMENTS RCIC-V-709 RCIC-V-713 RCIC-V-714 RCIC-V-715 RCIC-V-716 RCIC-V-127 8
RB-422 RB-4+~
RB-+
RB-4 RB-4@
RB-Q$
RCIC-PS-9B Root Iso (Turbine Exhaust)
RCIC-PS-12A Root Iso (Turbine Exhaust)
RCIC-PS-12C Root, Iso (Turbine Exhaust)
RCIC-PS-12B Root Iso (Zurbine Exhaust)
RCIC-PS-12C Root Iso (Iurbine Exhaust)
RCIC-P-1 Minimum Flow Containment Iso Open Open Open Open Open Locked/
RCIC-V-19 RB-444 RCIC-P-1 Discharge Minimum Flow (MO)
Closed RCIC-V-31 RB-444 RCIC-V-69 RB-444 RCIC-P-1 Suppression Pool Suction Iso (MO)
RCIC-P-2 Discharge to Suppression Pool (MO)
Closed Open RCIC-V-59 RB-444 RCIC-V-191 9
RB-444 RCIC-V-74 RB-444 Q
RCIC-V-68 RB-444 RCIC-V-22 RB-444 CST Flow Test Valve (MO)
CST Flow Test Valve (MO)
High Head Loss Orifice RCIC-RO-12 Bypass Auxiliary Steam Supply RCIC Turbine Exhaust Isolation (MO)
Closed Closed Locked/
Open Locked Closed Open 8 Requires independent verification of valve position by licensee personnel.
TABLE RCIC-3 (Cont'd)
DRAWING NO:
M-519 VALVE NUMBER LOC.
BLDG.
ELEV.
~
VALVE CHECKLIST DESCRIPTION REQUIRED ACTUAL POSITION POSITION COMMENTS RCIC-V-110 RCIC-V-113 RB-471 RB-471 RCIC Turbine Exhaust Open Vacuum Breaker (80)
(MO)
RCIC Turbine Exhaust Open Vacuum Breaker (ee)
(MO)
'54 RHR-V-102 RHR-V-26A RHR-V-26B RB-471 RB-471 RB-47 1 RCIC/RHR System Vacuum Relief Isol RHR-Loop A Cond.
Return (MO)
RHR-Loop B Cond.
Return (MO)
Open Locked/
Closed +
Locked/
Closed
+
RCIC-V-8 RCIC-V-64 RCIC-V-65 RB-511 RB-548 RB-548 RCIC Turbine Steam Open Supply Isolation (MO)
Steam Supply to RHR Locked/
Drain Line High Point Closed +
Testable Check to EK Closed Head Spray (AO)
RCIC-V-13 RCIC-V-23 RCIC-V-740 RCIC-V 184 RB-548 RB-548 RB-536 RB-536 Head Spray Line Iso (Mo)
RHR Head Spray Isolation (MO)
Root Stop RCIC-V-66 Diaphragm Oper.
(AZ 315')
Root Stop RCIC-V-66 Diaphragm Oper.
(AZ 315')
Closed Closed Locked/
Closed Locked/
Closed 8 Requires independent verification of valve position by licensee personnel.
+ Handwheel locked with breaker tagged open.
TABLE RCIC-3 (Cont'd)
DRAWING NO:
M-519 A.
VALVE CHECKLIST VALVE NUMBER LOC.
BLDG.
ELEV.
DESCRIPTION REQUIRED POSITION ACTUAL POSITION COMMENTS RCIC-V-742 RB-548 SP-19B Root Isolation Locked/-
Closed RCIC-V-19B RB-548 SP-19B to SR-6 Iso (AO)
Open RCIC-V-66 C-596 RCIC-V-63 C-551 Testable Check Valve Head Spray Line (AO)
Stm. Iso. to RCIC Turb.
Closed Open RCIC-V-76 C-551 RCIC-V-63 Bypass (MO) Closed COND-V-9A CST Q
Area COND-V-9B CST (3
Area CST "A" Supply Valve CST "B" Supply Valve Locked/
Open Locked/
Open 9 Requires independent verification of valve position by licensee personnel.
TABLE RCIC-3 (Cont'd)
~
DRAWING NO:
M-519 B.
SYSTEM POWER SUPPLY CHECKLIST POWER SUPPLY MC-Sl-1D9 LOC.
BLDG.
ELEV0 RW-467 DESCRIPTION RCIC-V-46 Turbine Cooling Water Supply 6A REQUIRED ACTUAL POSITION POSITION Closed COMMENTS MC-S2-lA9
-471 RCIC-V-1 RCIC Trip Throttle Valve 2B Closed MC-Sl-1D9 RW-467 RCIC-V-68 RCIC Steam Exhaust 7A Closed MC-S2-lA9 MC-S2-1A9 RB-471 RB-471 RCIC-V-13 RCIC Pump Discharge 5B RCIC-V-19 RCIC Min. ~
Flow 5C Closed Closed MC-S2-1A MC-S2-1A MC-S2-1Ag MC-S2-1Ag MC-S2-1A9 MC-S2-1Ag MC-S2-1AQ RB-471 RB-471 RB-471 RB-471 RB-471 RB-471 RB-471 RCIC-V-22 RCIC Test Bypass 6B RCIC-V-59 RCIC Test Bypass 7A RCIC-V-45 RCIC Steam Supply Valve 8A RCIC-V-64 RCIC Steam to RHR 8B RCIC-P-2 RCIC Vacuum Pump 9B RCIC-V-4 RCIC Cond.
Pump 9C RCIC-V-69 RCIC Vacuum Pump Disch.10C Closed Closed Clos Locked/
Open Closed Closed Closed MC-S2-1A RB-47 1 RCIC-V-23 RHR Held Spray 3B Closed 9 Requires independent verification of valve position by licensee personnel.
TABLE RCIC-3 (Cont'd)
DRAWING NO:
M-519 B.
SYSTEM POWER SUPPLY CHECKLIST POWER SUPPLY MC-Sl-2D9 MC-Sl-1D8 LOC.
BLDG.
ELEV.
RW-467 RW-467 DESCRIPTION RCIC-V-113 (MO-86)
RCIC Vacuum Bkr.
5B RCIC-V-110 (MO-80)
RCIC Vacuum Bkr. 6B REQUIRED ACTUAL POSITION POSITION Closed Closed COMMENTS MC-Sl-1D8 RW-467 RCIC-V-8 RCIC Steam Isolation Valve 6C Closed MC-S1-1D8 RW-467 RCIC-V-10 RCIC Closed Suction Valve CST
.3C MC-Sl-1D9 RW-467 RCIC-V-31 RCIC Suction Valve Suppr.
Pool 3B Closed MC-7B-A MC-8B-A RB-522 RB-522 RHR-V-26A Loop-A Cond Tagged Return 6B Open RHR-V-26B Loop-B Cond Tagged Return 6A Open MC-8B-AQ MC-8B-Ag RB-522 RB-522 RCIC-V-63, RCIC Steam Supply 9D RCIC-V-76 BPV Around RCIC-V-63 6B Closed Closed MC-7B 8 RB-522 RCIC Water Leg Pump RCIC-P-3 6C Closed Requires indepen ent verification of v lve posi ion by l censee personnel.
WASHINGTON NUCLEAR PLANT NO-2 GENERIC PROBABILISTZC RISK ASSESSMENT-BASED INSPECTION PLAN Automatic Depressurization System (ADS)
TABLE ADS-I IMPORTANCE BASIS AND FAILURE MODE IDENTIFICATION CONDITIONS THAT CAN LEAD TO FAILURE General Guidance Surveillance of the licensee's periodic calibration and testing and/or preventive or unscheduled maintenance activities and procedures, and/or normal and emergency operating procedures; training and check-off lists in accordance with the Technical Specifications and relevant NRC bulletins and information notices should reduce the probability of failure for the conditions listed below.
The mos" 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, training, etc.
Mission Success Criteria LATER Failure Conditions System fails to auto initiate due to
- 1) miscalibration of one of three signals: reactor low low water level, high drywell pressure, CS/LPCZ pumps operating, or 2) logic circuitry failure such as faulty contacts on actuation channel, sticky relay coil, and 3) operator fails to recover.
(PC,PT,OP)
TABLE ADS-1(Cont'd)
CONDITIONS THAT CAN LEAD TO FAILURE 2.
ADS valves (i1S-RV-3D, 4A, 4B, 4C, 4D, 5B, 5C) fail to operate in auto or manual due to 1) common mode test or maintenance
- error,
- 2) faulty pilot
- valves, 3)
SOV failure.
Total loss of 125V DC power normal power supply trip (fuse failure) due to transient and backup bus unavailable or bus transfer fails.
(PT,NT,OP) 4.
Failure to properly restore system after test or maintenance, such as failure to reopen root valves to Hi Drywell pressure sensor.
(PT,~1T)
WASHINGTON NUCLEAR PLANT NO 2
GENERIC PROBABILISTIC'ISK ASSESSMENT-BASED INSPECTION PLAN Automatic Depressurization System (ADS)
TABLE ADS-2 I&E INSPECTION PROCEDURES FOR SYSTEM OPERATION PROCEDURE NUMBER TITLE COMPONENTS FAILURE MODES 56700 Calibration Reactor Low Level Sensors High Drywell Pressure Sensors 1,4 62702 52051 Maintenance (Refueling)
Instrument Components and Systems-Procedure Review ADS Valves Accumulator Systems
- Sensors, Switches, Control Circuitry 2,3,4 1 <<2)4) 52053 52055 61725 Instrument Components and Systems-Work Observation Instrument Components and Systems-Record Review Surveillance 6 Calibration Control Program
- Sensors, Switches, Control Ci.rcuits
- Sensors, Switches, Control Circuits Control Circuitry 172,4, 1,2,4) 1 -4 7 1707 Operational Safety Verification ADS Valves, Accumulator Systems 1-4 41700 41701 Training Requallficatlon Training ADS Operation ADS Operation 1-4 v
+vs
~ '4aa
~ ~
~
WASHINGTON NUCLEAR PLANT NO.
2 GENERIC PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN Automatic Depressurization System (ADS)
TABLE ADS-3 MODIFIED SYSTEM WALKDOVN Since most of the failures associated with ADS are miscalibration errors, logic failures, and operator errors, these are not many specific items listed in this walkdown.
Observation of the ADS calibration procedures and/or func-tional test should provide the inspector with some assurance that these fail-ure modes are averted.
DESCRIPTION ID NO.
DESIRED LOCATION POSITION ACTUAL POSITION Control Room 1.
ADS Annunciation Panel PNL P601-A2 MCR No windows PNL P601-A3 illuminated 2.
Q supply pressure 3.
Air supply pressure 4.
ADS DIV 1
& 2 Inhibit PNL 601 Switches MCR Normal (Div.l)
Normal,(Div.2) 5.
Solenoid Control Switches 6
SRV Control Switches Div 1
125V DC Div 2 125V DC PNL 628 PNL 631 PNL 601 ADS "A" Logic ADS "B" Logic MCR MCR Au"o( 7)
( Div. 1)
Auto(7) (Div.2)
Auto (18)
BKR Closed (Sl-l)
BKR Closed (S1-2)
s REC 1 I
1
)
III I
)
)
I VACUUM BREAKER CIA SUPPLY
'I I
))I I
I I '
I I
III I I
)
I'I I I ACCIIM PS
<C 8
I ND l) o $r pro TYP RMS RMS r iC OF 601 SOP 3
2 I
I I
I TYPICALOF ELEVEN NON AOS VALVES MAIN STEAM LINE CtA CIA SUPPLY t SUPPLY+
c:J-)
(Ai l.')
TYP OF 2
REACTOR VESSEL I
ACCIIM ACCUM EXM.
C B
A, AO AOS rl i ro PS -+-WWI r iC) r RMS RMS RMS 601 628 63'1 BREAKER TYPICALOF SEVEN AOS VALVES SUPPRESSION POOL FIGURE 5. SAFETY/RELIEF YALYECONTROLS JUNE 1985
P
WASHINGTON NUCLEAR PLANT NO.2 GEiVERIC PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN Residual Heat Removal System (RHRS)
Low Pressure Coolant In)ection (LPCI) System TABLE LPCI-1 IMPORTANCE BASIS AND FAILURE MODE IDENTIFICATION CONDITIONS THAT CAN LEAD TO FAILURE General Guidance Surveillance of the licensee's periodic testing and/or preventive or unscheduled maintenance activities.and procedures and/or normal. and emergency operating procedures,
- training, and check-off lists in accordance with the Technical Specifications and relevant NRC bulletins and information 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.
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.
TS Technical specifications.'ission Success Criteria LATER Failure Conditions 1.
Failure of System to Auto Initiate Mhen Required due to:
a) Miscalibraton of Sensors (PC,PT)
~ Hi drywell pressure
~ Lo RPV level b) Failure of Logic Circuitry (PT) 6 ~6'
TABLE LPCX-1 (Cont'd}
CONDITIONS THAT CAN LEAD TO FAILURE 2.
Failure of Valves to Change Position for LPCI Mode and Failure to Manually Recover (PT, MT, OP):
~ Injection Valves V-42A,B,C Fail to Open
~ Heat Exchanger Bypass Valves V-48A,B Fail to Open
~ Suppression Pool Spray Valves V-27A,B Fail to Close
~ Suppression Pool Return Valves V-24A,B Fail to Close
~
Loop C Tes" Return Valve V-21 Fails to Close
~
Hea-Exchanger Level Control Valve V-65A,B Fails to close 3.
RHR Pumps 2A, B and C Unavailable Due to Maintenance and/or Failure of Other Pump to Start (Similarly for SW-P-1A,B)
I 4
RHR Hea-Fxchangers Unavailable Due to Plugging, Maintenance, Rupture, Failure of In1.et or Outlet Valves (PT,MT,ISI) 5.
Failure to Restore Components to Proper Position After Test or Maintenance (PT,MT,OP):
~ Injection Isolation Valves V-111A,B,C Left Closed
~
Pump Suction MOVs V-4A,B,C Left Closed
~
Pump Discharge Valves V-110A,B Left Closed
~
Pump Suction (Shutdown Cooling Mode) V-6A,B, V-67 Left Open
~ Minimum Flow Isolation Valve V-18,A,B,C Left Closed 6.
Human Error-Failure to Take Corrective Action:
a) LPCI Pumps Manually Shu" Off On High Level During Accident and Operator Fails to Recover (OP) b) Failure to Start System When Auto Initiation Fails (OP) 7 ~
Failure of Minimum Flow Valves to Control/Close When Required V-64A,B,C (PT) 8.
Failure to Maintain Water Leg in Loops (OP,MT):
RHR-P-3 For Loops B,C: V-82, V-85B,C LPCS-P-2 For Loop A: V-85B 9.
Loss of Pump Seal Cooling (PT,MT,OP)
NASHIiXGTON NUCLEAR PLANT N0.2 GENERIC PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN Residual Heat Removal System (RHRS)/Low Pressure Coolant Injection (LPCI) System TABLE LPCI-2 IGE INSPECTION PROCEDURES FOR SYSTEÃ OPERATION PROCEDURE NUMBER TITLE COMPONENTS+
FAILURE MODES E4 1700 52051 52053 52055 56700 Training Instrument Component and Systems-Procedure Review Instrument Components and Systems-Hoxk Observation Instrument Components and Systems-Record Review Calibration All Reactor Pressure and Level Sensors 2,5,6,8,9 61725 61726 Surveillance and Calibration Program Monthly Surveillance Observation Reactor Pressure and Level Sensors 1-5,7,9
- Pumps, MOVs check
- valves, manual valves, heat exchangers 72700 71707 Startup Testing>>Refueling Operational Safety Verification 71710 62702 62703 ESF System Malkdown Maintenance (Refueling)
I Monthly Maintenance Observation Pumps MOVs 2-518-9
- Refers only to components identified in Tables A3-1 and A3-3.
WASHINGTON NUCLEAR PLANT N0.2 GENERIC PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN Residual Hea" Removal (RHR)/Low Pressure Coolant Injection (LPCI)
DRAMIÃG NO:
8-521 A.
VALVE CHECKLIST TABLE LPCI-3 MODIFIED SYSTEM WALKDOWN-VALVE NUMBER LOC.
BLDG.
ELEV.
DESCRIPTION REQUIRED POSITION ACTUAL POSITION COMMENTS RHR-V-12B RHR-V-13B RHR-V-85B RHR-V-110BQ RHR-V-18B Q
RHR-V-67 Q
RHR-V-12A RHR-V-13A 422'22'44'44'44'22'22'22I UPPER LEVEL Open Open OPEN Locked Open Locked/
Open Min. Flow Isolation RHR "A" PQ Loop "C" Cross Tie and Cond Fill Pump Seal Leakoff Pump Seal Leakoff 1P ROOM Locked Cl.osed Open Open RHR "B" PUMP ROOM Pump Seal Leakoff Pump Seal Leakoff Water Leg Isolation Disch. Isolation Overhead Overhead RHR-V-85A RHR-V-110AQ RHR-V-18A Q
RHR-V-121 RHR-V-120 444'44'44'44'HR "A" PUMP ROOM LPCS Cross-Tie Water Leg Isolation Disch. Isolation Min. Flow Isolation FDR Isolation FDR Isolation UPPER LEVEL Open Locked/
Open Locked/
Open Locked/
Closed Locked/
Closed Overhead Overhead RHR-V-'109 RHR-V-12C RHR-V-13 C RHR-V-82 RHR-V-210 p
422'22'22'221 422' ROOM Locked/
Closed Open Open Open Open Pump Seal leakoff Pump Seal leakoff Water Leg Suction Water Leg Min. Flow Isolation RHR "C" PUM Condensate Supply 9 Requires independent verification of valve position by licensee personnel.
TABLE LPCI-3 (Cont'd)
DRAWING NO N-521 As VALVE CHECKLIST VALVE NUMBER LOC.
BLDG.
ELEV.
DESCRIPTION REQUIRED ACTUAL POSITION POSITION COMMENTS RHR-V-85C RHR-V-110C8 RHR-V-17 4 Q
RHR-V-18 C (3
RHR-V-172 AQ RHR-V-172 B
RC 444'44'44'44'71'71'PPER LEVEL Open Locked/
Open Locked/
Open Locked/
Open Locked/
Open Locked/
Open Flow Test Isolation Min. Flow Isolation.
Flow Test Isolation Flow Test Isolation RHR "C" PUMP ROOM Water Leg Isolation Disch Isolation RHR-V-14 B Q
R?iR-V-104 I(t'HR-V-14A 8
548'48'48'HR HX-1B SW Inlet RHR to FPC RHR-HX-1A SW Inlet Locked/
Open Locked/
Closed Locked/
Open RHR-V-1 13 Q
RHR-V-112 8
RHR-V-112 B9 RHR-V-111 Ba RHR-V-111 501'14'14'61'RYWEL Shutdown Cooling Suction Isolation Shutdown Cooling Return Isolation Shutdown Cooling return Isolation LPCI Injection Isolation Locked/
Open Locked/,
Open Locked/
Open Locked/
Open RHR-V-111 AQ 561 561'PCI Injection Isolation LPCI Injection Isolation Locked/
Open Locked/
Open RHR-V-4C g
P-601 REMOTE OPERATED VALVES-CONTROL ROOM Suppression Pool.
Open Suction (S Requftes independent vetificlllon oi valve position by licensee petsonnel.
TABLE LPCI-3 (Cont'd)
DRAWING NO..I-521 Ao VALVE CHECKLIST VALVE NUMBER RHR-FCU-64C RHR-V-4B Q
RHR-U-6B RHR-V-68B RHR-V-16B RHR-V-27B RHR-V-23 RHR-V-17B RHR-V-47 B Q
RHR-FCV-64B RHR-V-42C RHR-V-48B RHR-V-3B9 RHR-V-523 goal RHR-V-53B RHR-V-42B RHR-U-21 RHR-V-116 RHR-V-115 LOC.
BLDG ELEV.
P-601 P-601 P-601 P-601 P-601 P-601 P-601 P-601 P-601 P-601 P-601 P-601 P-601 P-601 P-601 P-601 P-601 P-601 P-601 DESCRIPTION Min Flow Suppression Pool Suction Pump Suction X-Tie RHR-HX-1B SW Outlet Lower Drywell Spray Suppression Pool Spray RHR to Head Spray Lower Drywell Spray RHR-HX-1B Inlet Min Flow LPCI Isolation RHR-HX-1B Bypass RHR-HX-1B Outlet Steam Condensing Mode Steam Inlet SCM Shutdown Cooling Return LPCI Isolation Flow Test Standby Service Water to Containment Flooding Standby Service Water to Containment Flooding REQUIRED ACTUAL POSITION POSITION Closed Open Closed Open Closed Closed Closed Closed Open Closed Closed Open Open Locked/
Closed Closed Closed Closed Closed Closed COKKNTS RHR-V-111B RHR-V-112B P-601 P-601 Shutdown Cooling Manual Isolation Open LPCI Manual Isolation Open RHR-V-'11C RHR-V-49 RHR-U-9 P-601 P-601 P-601 LPCI Manual Isolation RHR Drain to RW Shutdown Cooling Inboard Suction Open Closed Closed 8 Requires independent verification of valve position by licensee personnel.
'gp>>
p
~>,,~
~t i ~ 4 A
~ 'k+4
TABLE LPCI-3 (Cont'd)
DRA'ICING NO.
M-521 Ao VALVE CHECKLIST VALVE NUMBER LOC.
BLDG.
ELEV.
DESCRIPTION REQUIRED ACTUAL POSITION POSITION COMMENTS RHR-V-40 RHR-V-8 RHR-V-4A g
RHR-V-6A RHR-V-68A RHR-V-16A RHR-V-27A RHR-V-17A RHR-V-4? A Q
RHR-FCV-64A P-601 P-601 P-601 P-601 P-601 P-601 P-601 P-601 P-601 P-601 RHR Drain to RW Shutdown Cooling Supply Suppression Pool Suction
'Pump Suction X-Tie RHR-HX-1A SW Outlet Upper Drywell Spray Suppression Pool Spray Upper Drywell Spray RHR-HX-lA Inlet Min Flow Closed Closed Open Closed Open Closed Closed Closed Open Closed P-601 P-601 P-601 P-60 1.
RHR-V-48A RHR-V-24A RHR-V-3A9 RHR-V-52A Q
RHR-V-53A P-601 RHR-V-42A P-601 RHR-HX-1A Bypass Test Line Isolation RHR-HX-1A Outlet Steam Condensing Mode Steam Inlet SCM Shutdown Cooling Return LPCI Isolation Open Closed Open Locked/
Closed Closed Closed 8 Requires independent verification of valve position by licensee personnel.
TABLE LPCI-3 (Cont'd)
DRAWING NO. if-521 B.
SYSTEM POWER SUPPLY CHECKLIST POWER SUPPLY Siif-7 SM-8 SM-8 MC-78-A 2D MC-78-A 3A MC-78-A-38 MC-78-A-4D MC-78-A-5A MC-78-A-5C MC-78-A 6A MC-78-A 78 MC-78-A 7C MC-78-A 7D MC-88-A 2A MC-88-A 28 MC-88-A 2C MC-88-A 2D LOC BLDG'LEV.
DESCRIPTION RW-46 7'HR-P-2A RW-467'HR-P-28 RW-467'HR-P-2C RB-522'HR-V-11A RB-522'HR-V-6A RB-522'HR-V-4A RB-522'HR-V-24A RB 522'HR-V-42A RB-522'HR-V-27A RB-522'HR-V-26A RB-522'HR-FCV-64A RB-522'HR-V-124A RB-522'HR-V-1248 RB-522'HR-V-9 RB-522'HR-V-68 RB-522 'HR-V-4B RB-522'HR-V-4C REQUIRED POSITION Racked In Racked in Racked in Locked/
Open Closed Closed Closed Closed Closed Locked/
Open CLOSED Locked Open Locked Open Closed Closed CLosed Closed ACTUAL POSITION COMMENTS Not to be racked in until system is filled and vented.
Not to be racked in until system is filled and vented Not to be ra1;jed in until system is filled and vented
TABLE LPCI-3 (Cont'd)
DRAWING NO. N-521 BE SYSTEM POWER SUPPLY CHECKLIST POWER SUPPLY LOC.
BLDG.
ELEV.
DESCRIPTION REQUIRED POSITION ACTUAL POSITION COMMENTS MC-8B-A 3B MC-8B-A 3C MC-8B-A 3D MC-8B-A 4A MC-8B-A 4B MC-8B-A 4C MC-8B-A 4D MC-8B-A 5D MC-8B>>A 6A MC-8B-A 7A MC-8B-A 7B MC-8B-7A MC-7B-B-5A MC-7B-B-5B MC-7B-B-6C MC-8BgB-4D MC-8B-B-5A MC-8B-B-5B MC-8B-B-7C FRTS/8-81 Power XFER RB-522 'HR-V-125A RB-522 'HR-V-125 B RB-522 'HR-FCV-64B RB-522'HR-FCV-64C RB-522'HR-V-24B RB-522'HR-V-42B RB-522'HR-V-42C RB-522R RHR-V-27B RB-522 'HR-V-26B RB-522'HR-V-21 RB-522'HR-V-11B RB-522'HR-P-3 RB-572'HR-V-48A RB-572'HR-V-3A RB 5722 RHR V 52A RB-57 2'HR-V-47 B RB-572'HR-V-48B RB-572'HR-V-3B RB-572'HR-V-52B RW-467 'B8-81 Local Cont rol RM-467'HR-V-24B &
RHR-V-123A Power Tran Locked/
Open Locked/
Open Closed Closed Closed Closed Closed Closed Locked/
Open Closed Locked/
Open Closed Closed Closed Locked/
Open Closed Closed Closed Locked/
Open Normal Normal 8-81 BKR CUB C62-P001
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WASHINGTON NUCLEAR PLANT N0.2 GENERIC PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN Low Pressure Core Spray (LPCS)
System TABLE LPCS-1 IMPORTANCE BASIS AND FAILURE MODE IDENTIFICATION CONDITIONS THAT CAN LEAD TO FAILURE General Guidance Surveillance of the licensee's periodic testing and/or preventive or unscheduled maintenance activities and procedures and/or normal and emergency operating procedures,
- training, and check-off lists in accordance with the Technical Specifications and relevant NRC bulletins and information 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.
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.
TS - Technical specifications.
Mission Success Criteria LATER Failure Conditions 1.
LPCS pump P-1 does not start or run, or in maintenance.
Failure of auto initiation due to: (PC,PT) a). miscalibration of reactor water level/drywell pressure sensors.
b) failure of logic circuitry, i.e., relays.
3.
Failure of pump seals on motor due to loss of standby service water cooling.
TABLE LPCS-1 (Cont'd)
CONDITIONS THAT CAN LEAD TO FAILURE 4.
Failure to restore system after test or maintenance:
LPCS V-51 (injection valve) left closed manual valves b)
LPCS V-52 (min. flow) left closed manual valves c)
LPCS MOV V-1 (sustain valve) left closed 5.
Failure of following valves to change position after initiation signal due to valve failure, control failure,,permissive
- sensor, miscalibration:
(PC,PT) a) injection valve (MOV) LPCS V-5 fails to open permissive reactor pressure (470 gpm b) minimum flow valve (MOV) LPCS V-11 fails to control permissive flow
)721 gpm c) Test valve (MOV) LPCS V-12 fails to close 6.
Failure of, or maintenance on, room cooler pump room 8
Operators fall to recognize following problems:
(OP) a)
no auto initiation, and failure to manually initiate.
b) pump auto starts, but injection does not take place.
Prolonged pump operation through min flow line causes pump damage.
8.
Failure to maintain water leg, pump P-2, valves V-32, V-34.
(OP)
WASHINGTON NUCLEAR POWER PLANT NO.
2 GENERIC PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN Low Pressure Core Spray (LPCS} System TABLE LPCS-2 ISE INSPECTION PROCEDURES FOR SYSTEM OPERATION PROCEDURE NUJMBER TITLE COMPONENTS*
FAILURE MODES 41700 52051 52053 52055 56700 Training Instrument Components and Systems-Procedure Review Instrument Components and Systems Work Observation Instrument Components and Systems Record Review Calibration All Reactor Low Level
- Sensors, Drywell High Pressure
- Sensors, Reactor Vessel Pressure Sensors 4,7,8 2,5 61725 61726 Surveillance and Calibration Control Program Monthly Surveillance Observation
- Sensors, Pumps, Strainers, Check Valves Manual Valves, Discharge MOVs Suppression Pool 1-6 72700 Startup Testing-Refueling 62702 62703 Maintenance, (Refueling)
Monthly Maintenance Observation
- Pumps, Discharge MOVs, Strainers 1,3,4,6 71707 Operational Safety Verification
- Pumps, Strainers, MOVs 1-6.
Suppression
- Pool, Electric Power System 71710 ESF System Walkdown
- Refers only to components identified in Tables LPCS-1 and LPCS-3 ~
WASHINGTON NUCLEAR PLANT N0.2 GENERIC PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN Low Pressure Core Spray (LPCS)
System TABLE LPCS-3 MODIFIED SYSTEM WALKDOWN DRAWING NO. M-520 Ao VALVE CHECKLIST VALVE NUMBER LPCS-V-32 LOC.
BLDG.
ELEV.
RB-422 DESCRIPTION LPCS-P-2 Suction Valve REQUIRED ACTUAL POSITION POSITION Open COMMENTS LPCS-V-34 RB-422 LPCS-P-2 Discharge Isolation Valve Open 1/4 Turn SW-V-48 LPCS-V-49 SW-V-49 LPCS-V-41 RB-422 RB-422 RB-422 RB-422 Motor Cooling Inlet Open LPCS-P-1 Pump Vent Closed LPCS-P-1 Seal Vent Isolation Valve Locked Open Motor Cooling Outlet Throt.
4 GPM LPCS-FCV
-11 LPCS-V-52 Q
RB-422 RB-422 Minimum Flow Control Valve (MO)
Minimum Flow Line Isolation Closed Locked Open LPCS-V-19 LPCS-V-707 RB-441 RB-441 Upstream Isolation FE-2,FT-3,FIS-4 Open Suction Isolation(MO) Open LPCS-V-708 RB-441 Downstream Isolation FE-2, FT-3, FIS-4 Open LPCS-V-709 RB-441 PS-5, PI-2 Isolation Open 9 Requires independent verification of valve position by licensee personnel.
4
~'ABLE LPCS-3 (Cont'd)
DRAWING NO. M-520 A
VALVE CHECKLIST VALVE NUMBER LOG ~
BLDG ELEV.
DESCRIPTION REQUIRED ACTUAL POSITION POSITION COMMENTS LPCS-V-12 LPCS-V-60 Q
LPCS-V-66 RB-441 RB-441 RB-501 Test Line Control Valve (MO)
Test Line Isolation Control Air Isolation for AO-6 Closed Locked Opens Locked Closed LPCS-V-67 RB-501 Control Air Isolation for AO-6 Locked Closed LPCS-V-68 LPCS-V-X78b RB-501 RB-501 Control Air Vent Contaimment Instru-ment Iso. Detection D/P Iso.
Closed Open LPCS-V-5 LPCS-V-6 LPCS-V-51 Q
LPCS-V-78 RB-522 C132 DW-547 C123 DW-554 RB-422 Injection Line Control Valve (MO)
Injection Line Test-able Check Valve (MO)
Injection Line Iso-lation Valve P-2 Min. Flow Line Isolation Closed Closed Locked Open Open 9 Requires independent verification of valve position by licensee personnel.
t Throttled and locked to prevent pump run-out (run-our occurs at flows greater than 6400 gpm).
B ~
SYSTEM POWER SUPPLY CHECKLIST POWER SUPPLY SM-7 LOC.
BLDG.
ELEV.
RW-467 DESCRIPTION LPCS-P-1 LPCS Pump E-CB-LPCS REQUIRED ACTUAL POSITION POSITION Racked In COMMENTS MC-7B-CUB
-6B MC-7B-A-CUB-2A RB-522 RB-522 LPCS-P-2 Water Leg Pump 6B LPCS-V-1 (MO) 2A Closed Closed MC-7B-A-CUB-2B RB-522 LPCS-V-5 (MO) 28 Closed MC-7B-A-CUB-4B RB-522 LPCS-V-11 (MO) 4B Closed MC-78-A-CUB-2C RB-522 LPCS-V-12 (MO) 2C Closed DP-S I-1A 89 PP7AA 81 PP7AA 84 PP7AE 829 MCR MCR MCR RB-471 Control Logic Power (LPCS/RHR/RCIC)
LPCS V-6 Power LPCS FT-3 Power Controller/Indicator for LPCS V-6 Closed Closed Closed Closed PP7AE 85 RB-471 LPCS P-1 Motor Space Heaters Closed
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WASHINGTON NUCLEAR PLANT NO.
2 GENERIC PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN Main Steam Isolation Valves (MSIVs)
TABLE MSIV-1 IMPORTANCE BASIS AND FAILURE MODE IDENTIFICATION CONDITIONS THAT CAN LEAD TO FAILURE General Guidance Surveillance of the licensee's periodic calibration and testing and/or preventive or unscheduled maintenance activities and procedures, and/or normal and emergency operating procedures, training and check-off lists in accordance with the Technical Specifications and.relevant NRC bulletins and information 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.
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.
TS Technical Specifications Mission Success Criteria There are two redundant MSIVs for each of the four main steam lines, for a total of 8 MSIVs.
One set of valves, MS-V-22-A-D, are within the drywell and the others, MS-V-28 A-D, are located just outside the drywell, in the steam tunnel.
The safety objectives of the MSIV's are:
C a)
To prevent damage to the fuel barrier by limiting the loss of reactor coolant in the event of a major steam line break outside the primary containment.
A high steam Slow condition will cause an MSIV isolation.
b)
By rapid isolation of MSIV's the release of radioactive materials to the environs will be limited.
Should the fuel cladding fail, fission p'roducts will be released into the coolant.
The steam produced will be highly radioactive.
Radiation monitors placed close to each line will sense MSL radiation and when this signal represents three times normal background radiation, the MSIV's will be given the signal to isolate.
TABLE MSIV-1 (Cont'd)
CONDITIONS THAT CAN LEAD TO FAILURE c)
In the event of a line break within the drywell, release of radioactive materials to the environs is curtailed, again by MSIV isolation.
The signals which are used to indicate a line break are main steam line high flow and/or reactor low water level.
d)
The valves must remain open or be re-opened if the turbine bypass valves are to be used to remove decay heat via the main condenser and the Condensate and Feedwater Systems.
Failure of the operator to re-open the valves after closure is considered a system failure.
Normal fast closure time for the MSIVs is 3 to 5 seconds.
Two AC powered, solenoid operated, pilot actuating valves route air to desired ports, thereby positioning the MSIV.
The electrical supplies for the valves come from two separate
- sources, RPS buses A and B.
Each MSIV has 2 pilot operating solenoids.
One solenoid is powered from RPS bus A
and one solenoid is powered from RPS bus B.
A loss of both RPS buses is required to cause the valves to close.
An accumulator, mounted on the MSIV, provides backup pneumatic pressure to close the valve when both solenoids are de-energized or pneumatic supply pressure to the valve operator fails.
The air supply for the outboard MSIVs is from the control air system.
The inboard MSIVs are supplied from the containment instrument air/N>
inerting system.
The supply air/N>, accumulator, or the spring pressure, is capable of isolating the valve independently with the reactor at full pressure.
In the event of a 'failure in any two systems the third will close the valve, provided the under-the-piston area of the air cylinder is vented off.
The MSIVs will automatically close on any of the following signals:
a.
Reactor low water level (-50", level 2) b.
Main steam line high radiation (3X normal) c.
Main steam line high steam flow (104 psig, 140% steam flow) d.
Main steam line low pressure (831 psig, with mode switch in run) e.
Main steam line tunnel high temperature or, high ventilation system differential temperature.
f.
Main condenser low vacuum (7" Hg, may be bypassed by manual switch).
The MSIVs can also be manually closed by their associated control switches on P601 or by arming and depressing the four NS" pushbuttons on P601 (any combination of "A" or "C", and "B" or "D" pushbuttons).
TABLE MSIV-1 (Cont'd)
CONDITIONS THAT CAN LEAD TO FAILURE 1 ~
MSIUs Fail to Remain 0 en or 0 erator Fails to Re-Open Valves l3uring a transient the MSIUs may be spuriously closed or they may close automatically by the signals described above.
The MSIVs must be open for the Power Conversion System to be used for heat removal from the reactor and the containment.
If the MSIVs have closed and they are needed for a heat removal
- path, the operator must reopen them remote manually from the control room.
MSIVs Fail to Close or I.eak Zxcessivel If containment isolation is required, then the MSIVs must close and provide a leak tight seal.
Failure to close can be caused by miscalibration of sensors or failure of logic circuitry. (PC,PT,MT)
~
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WASHINGTON NUCLEAR PLANT NO.
2 GENERIC PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN Feedwater, Condensate, and Power Conversion Systems (FW/CD/PCS)
TABLE PCS-1 IMPORTANCE 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 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-PT MT-OP>>
IS-TS-Periodic calibration activities, procedures and training.
Periodic testing activities, procedures and training.
Preventive or unscheduled maintenance activities, procedures and training.
Normal and emergency operating procedures, check-off lists, training, etc.
Inservice inspection activities, procedures and training.
Technical specifications.
Mission Success Criteria The Power Conversion System (PCS) consists of the Main Steam
- System, the Main Turbine Generator, the Turbine Bypass Valves (4 valves, 25X capacity),
the Main Condenser, and the Circulating System. 't functions to convert reac-tor power to electricity, or to merely reject heat to the ultimate heat sink via the turbine bypass valves, the condenser, and the circulating water system.
The feedwater and condensate systems transfer the condensate from the main condenser hotwells to the reactor feedwater
- pumps, pre-heat the feedwater and return it to the reactor pressure vessel to be converted to steam.
These systems are not initiated automatically but are normally operating, non-safety systems which automatically control reactor level within a predetermined range.
Manual/automatic startup, operation and shutdown take place from the con-trol room.
Loss of offsite power automatically trips the main turbine, feed-
- water, and condensate systems.
Automatic shutdown of the feedwater system also occurs upon isolation of the MSIVs or a FW turbine trip.
TABLE PCS-1 (Cont'd)
CONDITIONS THAT CAN LEAD TO FAILURE There are two steam turbine-driven FH pumps, three condensate booster
- pumps, and three condensate pumps, all divided into two partially separated flowpaths A and B interspersed with the High and Low Pressure Heaters.
At least one flow path containing one FW pump, one condensate booster
- pump, and one condensate pump is required to maintain reactor water level following a trip.
A.
Feedwater S stem 1.
Failure of Long Term Operator Actions to Control Feedwater During Cool-down (e.g., Failure to Provide Long-Term Makeup to the Condenser)
Operator actions are an essential aspect of long-term feedwater control during cooldown.
Loss of feedwater can occur by such actions as failure to provide makeup to the condenser to maintain proper condenser level (OP).
2.
Miscalibration of Sensors Causes False FW System Trip Miscalibration of sensors such as hi h reactor water level can cause FW isolation due to a false high water level signal or failure of the FW pump turbines due to gross moisture carryover (PC,PT).
Miscalibration of the low reactor ressure sensors can cause FW pump trip due to a low turbine steam supply pressure, although condensate flow would continue (PC,PT).
Miscalibration of or false signals from the low condenser vacuum sensors, paths A and B, can also cause an automatic trip of the FW system on low con-denser
- vacuum, although condensate flow would continue in this case as well (PC,PT).
3 Common Mode Failure of All Offsite Power Sources Common mode failure of all offsite power sources causes an automatic trip of both the feedwater and condensate systems.
Refer to Electric Power Distri-bution System Table EP-1, for specific inspection information.
4.
FW System Trips Due to Failure in FW Control Logic, Failure of FW Turbine Control Causing Loss of H.P.
- Steam, Spurious MSIV Closure, or Genuine Isolation Signal Causing MSIV Closure A, FW system trip causes loss of the system.
Such trips can be caused.by several instrumentation and controls-related problems such as failure in the FW control or FW turbine control logic components or a spurious closure of the MSIVs.
A genuine containment isolation signal also causes MSIV closure lead-ing to FW trip.
The former failures can be related to periodic calibration
p
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TABLE PCS-1 (Cont'd)
CONDITIONS THAT CAN LEAD TO FAILURE and testing activities while the latter is part of the normal system response to a containment isolation and, in fact, the inspection focus should be on ensuring this response (PC,PT).
5.
Transient Condition Leads to FW Trip
- Sudden, sharp variations in operating conditions such as reactor water level and other process variables can cause FW trip.
Such transients can be caused by operator errors (OP).
6.
Steam Jet Air K)ector Legs Unavailable and Mechanical Vacuum Pumps Unavailable Cause Loss of Condenser Vacuum The loss of the capability of the steam jet air ejectors to maintain condenser vacuum combined with unavailability of the mechanical vacuum pumps will lead to a loss of condenser vacuum which in turn leads to loss of FW.
a)
Steam Jet Air Ejectors (SJAE) Legs Unavailable This can be caused by:
i)
Loss of Main Air Ejectors (OP,MT) ii)
Test or Maintenance of SJAE (PT,MT}
iii}
Off Gas System Failure/Trips (OP,MT) b)
Mechanical Vacuum Pump Unavailable This can be caused by:
i)
Operator Fails to Start Pump (OP) ii)
Pump Leg in Test or Maintenance (PT,MT) iii)
Pump Fails to Start and Run (OP,PT) iv)
Condenser Vacuum Limitation (OP)
.v)
Operational Limit Prohibits Pump Operation When Reactor Pressure
)125 psig.
Feedwater Minimum Flow Control Valve RFW-FCV-15 or Start-up Flow Control Valve FCV-10 Fails to Remain Open Failure of either of these valves can cause loss of FW flow under certain conditions.
(OP,MT)
TABLE PCS-1 (Cont'd)
CONDITIONS THAT CAN LEAD TO FAILURE B.
Condensate S stem 1.
Failure of Long Term Operator Actions to Control Condensate During Cooldown (e.g., Failure to Provide Long-Term Make-up to the Condenser)
Operator actions are an essential aspect of long-term condensate control during cooldown.
Loss of condensate can occur by such actions as failure to provide make-up to the condenser to maintain proper condenser level (OP).
2.
Common Mode Failure of All Offsite Power Sources Common mode failure of all offsite power sources causes an automatic trip of both the feedwater and condensate systems.
Refer to Emergency Electric Power System, Table A5-1, for specific inspection information.
3.
Flow Control Instruments Fail to Supply Signal or Supply False Signal to Trains A, B and C
Failure of the flow control instruments to supply signals or by supplying false signals to the FM/CD flow control devices can cause a
CD system trip (PC,PT)
~
4.
Rupture of Piping or Heat Exchangers Rupture of piping or of the various heat exchangers within the CD system can cause unavailability of the CD system (IS).
5.
Condensate or Condensate Booster Pumps A,
B and C Fail to Continue Running Failure of either the Condensate Pump or Condensate Booster Pump (CBP) in Train A combined with failure of either pump in Trains B and C will cause loss of the CD system.
Failure can also be caused by:
CBP Minimum Flow Valves
- 15A, B or C failing to open.
(OP,MT,TS) 6.
Condensate Demineralizer System (CDS) Failure Since the CDS processes all CD flow from both condensate
- pumps, any failure in the CDS can cause total loss of the CD system (OP,MT).
7.
Rupture of Condenser Hotwell.
A rupture of the, condenser hotwell tubes would allow contaminated circulating water to enter the CD system or cause loss of the hotwell inventory, thereby preventing CD system flow (IQ,OP).
TABLE PCS-1 (Cont'd)
CONDITIONS THAT CAN LEAD TO FAILURE C.
Power Conversion S stem (PCS) 1.
Failure of Turbine Bypass Valves Mechanically or Electronically The 4 Turbine Bypass Valves are automatically and sequentially controlled.
Post-accident they must be used since the main turbine is not available.
(OP,MT,PC) 2.
Loss of Vacuum Vacuum can be lost by a leak in the condenser or by loss of the circulating water system.
(IS,OP,MT)
WASHINGTON NUCLEAR PLANT NO.
2 GENERIC PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN Feedwater, Condensate, and Power Conversion Systems (FW/CD/PCS)
TABLE PCS-2 I&E INSPECTION PROCEDURES FOR SYSTEM OPERATION PROCEDURE NUMBER TITLE COMPONENTS FAILURE MODES 41700 Training FW control system ':1, 5,6
- SJAE, Mech.
vacuum
- pumps, offgas system condensate
& booster
- pumps, CD demineralizer
- system, condenser.
52051 52053 Instrument Components and Systems-Frocedure Review Instrument Components and Systems-Work Observation Reactor. High level, A:2,4 Reactor low pressure, B:3 Condenser low vacuum C:1
- sensors, FW control logic 52055 Instrument Components and Systems-Record Review 56700 Calibration 61725 61726 Surveillance and Calibration Program Monthly Surveillance Observation A:2,4,6 C:1 FW control system FW turbine control, Isolation signal
- SJAE, Mech.
Vacuum
- pumps, CD flow control
- system, heat exchangers,
- Piping, Condenser Hotwell.
72700 Startup Testing-Refueling
iABLE PCS-2 I&E INSPECTION PROCEDURES FOR SYSTEM OPERATION (Cont'd)
PROCEDURE NUMBER TITLE COMPONENTS FAILURE MODES 71707 71710 62702 62703 Operational Safety Verification ESF System Walkdown Maintenance (Refueling)
Monthly Maintenance Observation SJAE,'ech.
Vacuum A:6
- Pump, CD flow control instruments, B:5>6,8 condensate
& booster C:1,2 pumps, CD Demineralizer System.
WASHINGTON NUCLEAR PLANT NO 2
GENERIC PROBABILISTXC RISK ASSESSMENT-BASED INSPECTION PLAN Feedwater, Condensate, and Power Conversion Systems (FW/CD/PCS)
TABLE PCS-3 MODIFIED SYSTEM WALKDOWN LIST A.
Feedwater S stem Since the large majority of system failure modes involves operator errors, false instrument signals, control logic failures or miscalibration of
- sensors, a system walkdown will reveal little tangible information.
- However, the probability of failures leading to loss of condenser vacuum through fail-ures of the steam )et air e)ectors and the mechanical vacuum pump can be reduced by system walkdown for those components described in the next section for the Condensate System.
(Loss of condenser vacuum causes loss of the Feedwater System but not the Condensate System.)
B.
Condensate S stem The failure modes for the Condensate System consist, for the most part, of operator errors, failure of the flow control instruments, loss of off-site power, failure of the Condensate and Booster Pumps to continue running, etc.
- Again, a system walkdown will yield very little tangible information with the exception of the following:
C.
Power Conversion S stem (PCS)
The main failure mode of the PCS is failure of the turbine bypass valves either mechanically or electrically.
MSIVs are covered in MSIV-1 to 3.
1.
For FW System Failure Caused by Loss of Condenser Vacuum Description
- a. Gland Steam Cond.
Bypass (7 PSID)
I.D. No.
COND-P CV-5 Location CRBd A Desired Position Auto Actual Position b.
SJAE Bypass (7 PSID)
- c. Ejector Cond A Inlet
- d. Ejector Cond A Outlet
- e. Ejector Cond B Inlet
- f. Effector Cond B Outlet COND-V-lllA COND-V-114B COND-V-111B Same Same Same COND-PCV-7 Same COND-V-114A Same Auto Open Open Open Open
TABLE PCS-3 (Cont'd) 1.
Condensate (Cont'd)
Description I.D.
Now Location Desired Actual Condition Condition
- g. Mechanical Vacuum Pumps (Should be off during power operation.)
- h. Offgas System St"a tus Radwaste Bldg.
Status 2.
Condensate Demineralizer System 3.
Walkdown of High Rupture Risk Components Description I.D. No.
Location Desired Actual Condition Condition
- a. Piping
- b. Heat Exchangers c.
Condenser Hot Well Status Status Status
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WASHINGTON NUCLEAR PLANT NO.
2 GENERIC PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN Standby Gas Treatment System TABLE SGT-1 IMPORTANCE BASIS AND FAILURE MODE IDENTIFICATION CONDITIONS THAT CAN LEAD TO FAILURE General Guidance Surveillance of the licensee's periodic calibration and testing and/or preventive or unscheduled maintenance activities and procedures, and/or normal and emergency operating procedures, training and check-off lists in accordance with the Technical Specifications and relevant NRC bulletins and information 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.
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.
S stem Descri tion Mission Success Criteria The SET has two 100/ capacity trains of filters and each train has two 100% capacity fans whose primary purpose is to prevent exfiltration from the Ractor Bu9.ding by maintaining the secondary containment at 0.25 inches water gauge
- vacuum, and by exhausting air through the filter trains.
The secondary function is to filter the purge exhaust of the Primary Containment when radiation levels in the PC preclude direct exhaust.
1 ~
The most likely mode of failure is one train out of service for maintenance and failure of the redundant train.
Note:
The remaining failure modes focus on events that cause the unavailability of one train.
2.
Fire in charcoal beds of train A(B) caused by inadequate/uneven air flow, or loss of both fans FN 1A-1/1A-2 (1B-l, 1B-2), or operator error in removing a train from service.
Failure to provide adequate air flow through 'train A(B) caused by the following. (PC,PT,MT) 1)
Failure of FN1A-1 (1B-1) to start or run and FNlA-2 (1B-2) fails to auto start due to mechanical/instrumentation problems.
Failure of fan automatic vortex damper due to mechanical problems and a failure or miscalibration of secondary containment pressure control system.
3)
Any of the following values fail to go to indicated position following a FA2 signal:
SGT-V-1A(1B)
PC Purge Close SGT-V-2A(2B) RB Intake Open SGT-3A1,3A2(3B1,3B2)
SGT Inlet Open SFT-V-4A1,4A2(4B1,4B2)
SGT Outlet to RB Close SGT-V-5A1(5B2)
SGT Outlet to Outside Open SGT-V-5A2(51)
SGT Outlet to Outside Close 4.
Failure to maintain vacuum in secondary containment due to:
Air infiltration ) 2240 CFM 2)
SGT V-1A(1B) fail open 3)
Failure of SGT trains A and B as described in failure modes 1-4.
Note:
To assure Reactor Building integrity, it is suggested that the licensee's periodic negative pressure test at -0.25in.
M.C.
be witnessed or otherwise verified.
WASHINGTON NUCLEAR PLANT NO.
2 GENERIC PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN Standby Gas Treatment System (SGTS)
TABLE SGT-2 I&E INSPECTION PROCEDURES FOR SYSTEM OPERATION PROCEDURE NUMBER TITLE COMPONENTS FAILURE MODES 56700 Calibration FA2 signal initiators 3~4>
62702 52051 52053 52055 61725 Maintenance (Refueling)
Instrument Components and Systems-Procedure Review Instrument Components and Systems-Work Observation Instrument Components and Systems-Record Review Surveillance
& Calibration Control Program
- Valves, Fans,
- Dampers, Charcoal Beds
- Sensors, Switches, Control Circuitry
- Sensors, Switches, Control Circuits
- Sensors, Switches, Control Circuits Control Circuitry 1-4 61726 71707 71710 41700 41701 Monthly Surveillance Observation Operational Safety Verification ESF System Walkdown Training Requalif ication Training Valves, Fans, Dampers 1-4 Charcoal Beds Valves, Fans,
- Dampers, Charcoal Beds
WASHINGTON NUCLEAR PLANT NO.
1 GENERIC PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN Standby Gas Treatment TABLE SGT-3 MODIFIED SYSTEM WALKDOWN DRAWING NO: M-544 A.
VALVE CHECKLIST VALVE NUMBER FP-V-72 LOC.
BLDG.
ELEV.
R-572 DESCRIPTION Carbon Filter 1A-1 Deluge Isolation REQUIRED POSITION Locked/
Open ACTUAL POSITION COMMENTS NE Control Against Wall SGT-V-702A R-572 SGT-PI-6A-1 Root Open NE Control Against Wall CAS-V-178 R-572 Air Supply to SGT-V-F26 Operator Open NE Control Against Wall SGT-V-F21 R-572 SGT-PCV-F26 Inlet Iso Open NE Control Against Wall SGT-V-F24 R-572 SGT-PI-6A-2 Root Open NE Control Against Wall FP-V-71 R-572 Prefilter 1A Deluge Isolation Locked/
Open NE Control Against Wall SGT-V-701A SGT-V-F14 R-572 SGT-PI-8A-1 Root R-572 SGT-PI-8A-2 Root Open Open NE Control Against Wall NE Control Against Wall SGT-V-F11 R-572 SGT-V-F16 Inlet Iso Locked/
Open NE Control Against Wall CAS-V-177 FP>>V-73 R-572 R-572 Air Supply to SGT-V-F16 Operator Carbon Filter lA Deluge Isolation Open Locked/
NE Control Against Wall NE Control Against Wall SGT-V-703A R-572 SGT-PI-7A-1 Root Open NE Control Against Wall
,TABLE SGT-3 (Cont'd)
DRAWING NO: M-544 Ao VALVE CHECKLIST VALVE NUMBER LOC.
BLDG.
ELEV.
DESCRIPTION REQUIRED POSITION ACTUAL POSITION COMMENTS SGT-V-F34 R-572 SGT-PI-7A-2 Root SGT-V-F31 R-572 SGT-V-F36 Inlet Iso Open Locked/
Open NE Control Against Mall NE Control Against Mall CAS-V-100>>5 R-572 Air Supply to SGT Iso Open NE Control Against Wall CAS-V-179 R-572 FP-V-74 R-572 Air Supply to SGT-V-F36 Operator Pressure 1B Deluge Isolation Open Locked/
Open NE Control Against Wall NE Control Against Wall SGT-V-701B R-572 SGT-PI-8B-1-Root Open NE Control Against Wall SGT-V-F44 R-572 SGT-PI-8B-2 Root SGT-V-F41 R-572 SGT-V-F46 Inlet Iso Open Locked/
Open NE Control Against Wall NE Control Against Wall CAS-V-180 R-572 Air Supply to SGT-V-F46 Operator Open NE Control Against Wall SGT-V-F42 R-572 SGT-V-F46 Inlet Ret.
Closed NE Control Against Wall FP-V-76 R-572 Carbon Filter 1B-2 Deluge Isolation Locked/
Open NE Control Against Wall SGT-V-703B R-572 SGT-PI-7B-1 Root Open NE Control Against Mall
TABLE SGT-3 (Cont'd)
DRAWING NO: M-544 A.
VALVE CHECKLIST VALVE NUMBER LOC.
BLDG ELEV.
DESCRIPTION REQUIRED POSITION ACTUAL POSITION COMMENTS SGT-V-F64 R-,572 SGT-PI-7B-1 Root Open NE Control Against Wall SGT-V-F61 R-572 SGT-V-F66 Inlet Iso Locked/
Open NE Control Against Wall SGT-V-F63 R-572 SGT-V-F66 Outlet Iso Closed NE Control Against Wall CAS-V-182 R-572 FP-V-75 R-572 Air Supply to SGT-V-66 Operator Carbon Filter 1B-1 Deluge Isolation Open Locked/
NE Control Against Wall NE Control Agains t Wall SGT-V-702B R-572 SGT-PI-6B-1 Root Open NE Control Against Wall SGT-V-54 R-572 SGT-V-F51 R-572 SGT-V-F53 R-572 CAS-V-181 R-572 SGT-PI-6B-2 Root SGT-V-F56 Inlet Isolation SGT-V-F56 Outlet Drain Air Supply to SGT-V-F56 Operator Open Locked/
Open Closed Open NE Control Against Wall NE Control Agains t Wall NE Control Against Wall NE Control Against Wall SGT-V-710A SGT-V-711A R-572 R-572 SGT-DPIS-4A Root (Carbon Filter)
SGT-DPIS-4A Test Open Closed On SGT-Train A On SGT-Train A
TABLE SGT-3 (Cont'd)
DRAWING NO: M-544 I
A.
VALVE CHECKLIST VALVE NUMBER LOC.
BLDG~
ELEV.
DESCRIPTION REQUIRED POSITION ACTUAL POSITION COMMENTS SGT-V-720 R-572 SGT-V-721 R-572 SGT-V-722 R-572 SGT-V-723 R-572 SGT-V-727 R-572 SGT-FT-1B-1 Root (Fan Discharge)
SGT-FT-1B-2 Root SGT-FT-1B-2 Root (Fan Discharge)
SGT-FT-18-2 Root SGT-FT-2B Root (Train Discharge to Stack)
Open Open Open Open Closed On SGT-Train A
On SGT-Train A
On SGT-Trai'n A On SGT-Train A On SGT-Train A SGT-V-728 R-572 SGT-V-729 R-572 SGT-FIS-2A2 Root SGT-FIS-2Bl Root Open Open B.
SYSTEM POWER SUPPLY CHECKLIST POWER SUPPLY MC-7B-B MC-7B-B LOC ~
BLDG ELEV~
R-572 Col M6 R-572 Col M6 DESCRIPTION REQUIRED POSITION SGT-FN-1A-1 BKR 2C Closed SGT-FN-1B-1 BKR 2D Closed ACTUAL POSITION COMiMENTS MC-7B-B R-572 Col M6 SGT-V-lA BKR 3A Closed MC-7B-B R-572 SGT-V-3A-1 Col M6 BKR 3B Closed MC-7B-B R-572 Col M6 SGT-V-3B-1 BKR 3C Closed
TABLE SGT-3 (Cont'd)
DRAWING NO: M-544 A.
VALVE CHECKLIST VALVE NUMBER LOG ~
BLDG.
ELEV.
DESCRIPTION REQUIRED POSITION ACTUAL POSITION COMMENTS SGT-V-720 R-572 SGT-V-721 R-572 SGT-V-722 R-572 SGT-V-723 R-572 SGT-V-727 R-572 SGT-V-728 R-572 SGT-V-729 R-572 SGT-FT-1B-1 Root (Fan Discharge)
SGT-FT-1B-2 Root SGT-FT-1B-2 Root (Fan Discharge)
~
SGT-FT-1B-2 Root SGT-FT-2B Root (Train Discharge to Stack)
SGT-FIS-2A2 Root SGT-FIS-2B1 Root Open Open Open Open Closed Open Open On SGT-Train A On SGT-Train A
On SGT-Train A On SGT-Train A On SGT-Train A B.
SYSTEM POWER SUPPLY CHECKLIST POWER SUPPLY MC-7B-B LOG ~
BLDG.
ELEV.
R-572 Col M6 DESCRIPTION SGT-FN-1A-1
,BKR 2C REQUIRED ACTUAL POSITION POSITION Closed COMMENTS MC-7B-B R-572 Col M6 SGT-FN-1B-1 BKR 2D Closed MC-7B-B MC-7B-B MC-7B-B R-572 Col M6 R-572 Col M6 R-572 Col M6 SGT-V-1A SGT-V-3A-1 SGT-V-3B-1 BKR 3A BKR 3B BKR 3C Closed Closed Closed
gABLE SGT-3 (Cont'd)
DRAWING NO: M-544 B.
SYSTEM POWER SUPPLY CHECKLIST POWER SUPPLY LOCe BLDG+
ELEVE DESCRIPTION REQUIRED ACTUAL POSITION POSITION COMMENTS MC-7B-B R-572 Col M6 SGT-V-4A-1 BKR 3D Closed MC-7B-B MC-7B-B MC-7B-B MC-8B-B R-57 2 Col M6 R-572 Col M6 R-572 Col M6 R-572 Col N8 SGT-V-4B-1 SGT-V-5A-1 BKR 4B SGT-V-5B-1 BKR 4C SGT-FN-1B-2 BKR 1F Closed Closed Closed Closed MC-8B-B R-572 Col N8 SGT-FN-1A-2 BKR 2B Closed MC-8B-B MC-8B-B MC-SB-B MC-8B-B MC-SB-B MC-8B-B MC-SB-B MC-8B-B R-572 Col N8 R-572 Col N8 R-572 Col N8 R-572 Col N8 R-572 Col N8 R-572 Col N8 R-572 Col N8 R-572 Col N4 SGT-V-1B BKR 3A 8 GT-V-3A-2 BKR 3B SGT-V-3B-2 BKR 3C SGT-V-4A-2 BKR 39 SGT-V-5A-2 BKR 4A SGT-V-4B-2 BKR 4B SGT-V-5B-2 BKR 4C SGT-ESH-1B BKR 2BL Closed Closed Closed Closed Closed Closed Closed Closed
TABLE SGT-3 (Cont'd)
DRAWING NO: M-544 BE SYSTEM POWER SUPPLY CHECKLIST POWER SUPPLY LOC.
BLDG.
ELEV DESCRIPTION REQUIRED ACTUAL POSITION POSITION
.. COMMENTS MC-8B R-572 Col N4 SGT-ESH-2B BKR 2BR Closed MC-7B R-572 Col N4 SGT-ESH-1A BKR 1BL Closed MC-7B PP-7A-E R-572 Col N4 R-471 SGT-ESH-2A BKR lBL SGT-FN-lA-1 Heater CKT 17 Closed Closed PP-8A-E R-471 SGT-FN-1A-2 Heater CKT 15 Closed PP-7A-E R-471 SGT-FN-1B-1 Heater CKT 18 Closed PP-8A-E R-471 SGT-FN-1B-2 Heater CKT 16 Closed PP-7A-E R-471 SGT-ESH-1A Control CKT 7 Closed PP-8A-E PP-7A-E R-471 R-471 SGT-ESH-1B Control CKT 5 SGT-ESH-2A Control CKT 8 Closed Closed PP-8A-E R-471 SGT-ESH-2B Control CKT 6 Closed PP-7A-E PP-8A-E PP-7A-E R-471 R-471 R-471 SGT-V-2A CKT 4 SGT-V-2B CKT 3 Pressure Gauges on SGT-FU-1A CKT 1
Closed Closed Closed
gABLE SGT-3 (Cont')
DRAWING NO: M-544 B.
SYSTEM POWER SUPPLY CHECKLIST POWER SUPPLY LOC.
BLDG.
ELEVE DESCRIPTION REQUIRED POSITION ACTUAL POSITION COMMENTS PP-7A-E R-471 Humidity Transmitter on SGT-FU-lA CKT 6 Closed P P-7A-E R-47 1 Heat Detector Water Closed Spray SGT-FU-1A CKT 9 PP<<8A-E R-471 Pressure Gauges on SGT-FU-1B CKT 1
Closed PP-8A-E R-471 Humidity Transmitter Closed on SGT-FU-1B CKT 4 PP-8A-E R-471 Heat Detector Water Closed Spray SGT-FU-1B CKT 7
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WASHINGTON NUCLEAR PLANT NO.
2 GENERIC PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN Hain Steam Isolation Valves (MSIVs)
TABLE MSIV-1 IMPORTANCE BASIS AND FAILURE MODE IDENTIFICATION CONDITIONS THAT CAN LEAD TO FAILURE General Guidance Surveillance of the licensee's periodic calibration and testing and/or preventive or unscheduled maintenance activities and'rocedures, and/or normal and emergency operating procedures, training and check-off lists in accordance with the Technical Specifications and.relevant NRC bulletins and information 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.
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.
TS Technical Specifications Mission Success Criteria There are two redundant MSIVs for each of the four main steam, lines, for a total of 8 MSIVs.
One set of valves, MS-V-22-A-D, are within the drywell and the others, MS-V-28 A-D, are located gust outside the drywell, in the steam tunnel.
The safety objectives of the MSIV's are:
a)
To prevent damage to the fuel barrier by limiting the loss of reactor coolant in the event of a major steam line break outside the primary containment.
A high steam flow condition will cause an MSIV isolation.
\\
b)
By rapid isolation of MSIV's the release of radioactive materials to the environs will be limited.
Should the fuel cladding fail, fission products will be released into the coolant.
The steam produced will be highly radioactive.
Radiation monitors placed close to each line will sense MSL radiation and when this signal represents three times normal background radiation, the MSIV's will be given the signal to isolate.
TABLE MSIV-1 (Cont'd)
CONDITIONS THAT CAN LEAD TO FAILURE c)
In the event of a line break within the drywell, release of radioactive materials to the environs is curtailed, again by MSIV isolation.
The signals which are used to indicate a line break are main steam line high flow and/or reactor low water level.
d)
The valves must remain open or be re-opened if the turbine bypass valves are to be used to remove decay heat via the main condenser and the Condensate and Feedwater Systems.
Failure of the operator to re-open the valves after closure is considered a system failure.
Normal fast closure time for the MSIVs is 3 to S seconds.
Two AC powered, solenoid operated, pilot actuating valves route air to desired ports, thereby positioning the MSIV.
The electrical supplies for the valves come from two separate
- sources, RPS buses A and B.
Each MSIV has 2 pilot operating solenoids.
One solenoid is powered from RPS bus A
and one solenoid is powered from RPS bus B.
A loss of both RPS buses is required to cause the valves to close.
An accumulator, mounted on the MSIV, provides backup pneumatic pressure to close the valve when both solenoids are de-energized or pneumatic supply pressure to the valve operator fails.
The air supply for the outboard MSIVs is from the control air system.
The inboard MSIVs are supplied from the containment instrument air/N inerting system.
The supply air/N2, accumulator, or the spring pressure, is capable of isolating the valve independently with the reactor at full pressure.
In the event of a failure in any two systems the third will close the valve, provided the under-the-piston area of the air cylinder is vented off.
The MSIVs will automatically close on any of the following signals:
a.
Reactor low water level (-50", level 2) b.
Main sted'm line high radiation (3X normal) c.
Main steam line high steam flow (104 psig, 140% steam flow) d.
Main steam line low pressure (831 psig, with mode switch in run) e.
Main steam line tunnel high temperature or, high ventilation system differential temperature.
f.
Main condenser low vacuum (7" Hg, may be bypassed by manual switch).
The MSIVs can also be manually closed by their associated control switches on P601 or by arming and depressing the four NS" pushbuttons on P601 (any combination of "A" or "C", and "B" or "D" pushbuttons).
~
'f 0 ~ ~
TABLE MSIV-1 (Cont'd)
CONDITIONS THAT CAN LEAD TO FAILURE 1 ~
MSIVs Fail to Remain 0 en or 0 erator Fails to Re-0 en Valves During a transient the MSIVs may be spuriously closed or they may close automatically by the signals described above.
The MSIVs must be open for the Power Conversion System to be used for heat removal from the reactor and the containment.
If the MSIVs have closed and they are needed for a heat removal
- path, the operator must reopen them remote manually from the control room.
MSIVs Fail to Close or Leak Excessivel If containment isolation is required, then the MSIVs must close and provide a leak tight seal.
Failure to close can be caused by miscalibration of sensors or failure of logic circuitry. (PC,PT,MT)
WASHINGTON NUCLEAR PLANT NO.
2 GENERIC PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN Main Steaa Isolation Valves (MSIVs)
TABLE MSIV-2 INSPECTION PROCEDURES FOR SYSTEM OPERATION PROCEDURE NUMBER TITLE COMPONENTS FAILURE MODES 41700 52051 0
52/53 52055 61725 61726 Training Instrument Components and Systems Procedure Review Instrument Components and Systems - Work Observation Ins trument Components and Systems Record Review Surveillance and Calibration Control Program Monthly Surveillance Observation MSIVs Reactor Low Level, MS line high rad high steam flow, low pressure, MS tunnel high temp.
or high AT.
MSIVs Logic functions described above 61720 Containment Local Leak rate Testing 62702 62703 Maintenance Monthly Maintenance Observation MSIUs 1,2
WASHINGTON NUCLEAR PLANT N0.2 GENERIC PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN Main Steam Isolation Valves (MSIVs)
TABLE MSIV-3 MODIFIED SYSTEM WALKDOWN DRAWING NO:
M 502,M506 M529,M557 A.
VALVE CHECKLIST VALVE NUMBER LOC.
BLDG.
ELEV.
DESCRIPTION REQUIRED ACTUAL POSITION POSITION COMMENTS MS-V-22A F601 (AO) 8 MS-V-22B F601 (AO) 8 MS-V-22C P601 (AO) 8 MSIV-22D F602 (Ao)
MSIV-28A F602 (AO) 8 MSIV-28B F602 (AO) 8 MSIV-28C P602 (AO) 8 MSIV-28D P602 (AO) 8 MS-V-705A Htr Bay Q
T-471 MS-V-705B Htr Bay Q
T-471 "A" Inboard MSIV "B" Inboard MSIV "C" Inboard MSIV "D" Inboard MSIV "A" Outboard MSIV "B" Outboard MSIV "C" Outboard MSIV "D" Outboard MSIV MS-PT-1A Root Iso Valve MS-PT-1B Root Iso Valve Auto/
Open Auto/
Open Auto/
Open Auto/
Open Auto(
Open Auto/
Open Auto/
Open Auto/
Open Open Open Top of Stm Ln Above MS-V-146 Top of Stm Ln Above MS-V-146 MS-V-705C Htr Bay Q
T-471 MS-V-705D Htr Bay (t
T-471 MS-PT-1C Root Iso Valve MS-PT-1D Root Iso Valve Open Open Top of Stm Ln Above MS-V-146 Gage removed
TABLE MSIV-3 (Cont'd)
DRAWING NO:
M-502,M506,M529,M557 A.
VALVE CHECKLIST VALVE NUMBER LOC ~
BLDG.
ELEVE DESCRIPTION REQUIRED ACTUAL POSITION POSITION COMMENTS MS-V-706A Htr Bay Q
T-471 MS-V-706B Htr Bay Q
T-471 MS-V-706C Htr Bay Q
T-471 MS-V-706D Htr Bay Q
T-471 Stm Line "A" to Ave.
Manifold Iso Valve Stm Line "B" to Ave.
Manifold Iso Valve Stm Line "C" to Ave.
Manifold Iso Valve Stm Line "D" to Ave.
Manifold Iso Valve Open Open Open Open Top of Stm Ln Above MS-V-146 Top of Stm Ln Above MS-V-146 Top of Stm Ln Above MS-V-146 Top of Stm Ln Above MS-V-146 MS-V-708A Htr Bay Q
T-471 MS-DPIS-13A Root Iso Manifold Iso Valve Open By Bypass Valves MS-V-709A Htr Bay Q
T-471 MS-DPIS-13A Root Iso Manifold Iso Valve Open By Bypass Valves MS-V-709B Htr Bay Q
T-47 1 MS-DPIS-13B Root Iso Manifold Iso Valve Open By Bypass Valves MS-V-709C Htr Bay Q
T-471 MS-DPIS-13C Root Iso Manifold Iso Valve Open By Bypass Valves MS-V-709D Htr Bay Q
T-471 MS-V-707A T-471 Q
MS-V-707B T-471 Q
MS-V-707C T-471 Q
MS-V-707D T-471 Q
MS-DPIS-13D Root Iso Manifold Iso Valve MS-PS-15A Root Iso Valve MS-PS-15B Root Iso Valve MS-PS-15C Root Iso Valve MS-PS-15D Root Iso Valve Open Open Open Open Open By Bypass Valves Under HP Turbine Under HP Turbine Under HP Turbine Under HP Turbine Q Requires independent verification of valve position by licensee personnel.
4 TO TURBINE CONTROL SYSTEM
~
~
REACTOR
~It~
~III.II$
II p hIS V 28 RfACTOR RUtIOIKC I'+ 1
~
~
H Il 4
II v -Q~
TO HP TURBINE r I~ ~
H~M 1I I Me
~4C AIW4ll
~ TO HP TURBINE
(
Illlllll YtCIIIIII 5tfttlt5 Et MS V 22
~
IN I IIIT III TO EDR to kQT ICC IIITIS Stttl tUC 5ISSII UIC DSIIT TO MAINCONDENSER W
~C Q
~ aI IIIT III TO SEAL FTEAM EYAPORATORS TO Rff'URBINES TO FIISR'S TMD STAGE REHEATERS TO MAIMCOMOEMSER TO OFF GAS PREHEATERS TO SJAE'S DATIVELL EDR FOR I I I
I I I I I I
I I I I I S.
LJ 5IISTAfCSIOK YOSII4lf FIGUAE s.
MAIN STEAM SYSTEM
~X@IISt SAULT ISSS
AIR AIR SUPPLY EXH AIR PILOT VALVE S.
CLOSING SPRING RING LOADED AIR OPER.
SLAVE CYL. SPRING AND/OR
~AIR WILL CLOSE VAL~VE AC TEST SOLENOID AIR CYLINDER PILOT ACTUATING SOLENOID AC PILOT ACTUATING SOLENOID AC OPEN INTER CLOSE LIMIT SW.
TO VALVE MAIN STEAM'LINEVLV. CONT. DIAG.
TYPICAL OP e VAI.VES FIGURE BB 841128.1A MAY 1985
AIA AIA u
SUttlT ETH AUI PILOT VALVE aOSWG SP NING AN ANI SUPPLV ETH S
00 AUL PILOT VALVE CLOSING SPNING S
AN ANI u
SUPtLT ETH u
AIR Pa 01 VALVE CLOSWG SPANIG AC IEST SOLENOID AUI CTLWDEA AC lEST SOlfNOID ANI CTLWDEA AC I EST SOLENOID AN CTLWDEA tN.OT ACIUAIWG SOLEHQO AC
~ILOT ACIUA'TING SOLENOID AC PILOI ACIUAIWG 4DlfHOEI
- C PAOT ACIUATWG SDLEHTNO AC Ot EN IHIEA GLOM I
'IO VALVED LNIIT SW PACT ACTUATWG A*'OLENOID CLOM 10 VALVE tlLOT AC'IUATWG SOLENOID A-OPEN
/
WIT A CLOSE
/y I
10 VALVE QSIV IH CLOSED POSITION IISIT W OPEN tOS IIION IISIVIH 'IfST TA Ill%
liu IWS FIGURE 4. MSIV POSITIOII COIITROL
AIR CYLINDER (CLOSING PISTON INSIDE)
MSIV SPEED CONTROL VALVE HYDRAUt.lC DASH POT SPRING GU IDE ACTUATOR SUPPORT AND SPRING GUIDF SHAFT CLOSING SPRING SPRING SEAT MEMBER STEM STEM PACKING LEAK OFF CONNECTION BONNET BOLTS
~soever.
PISTON RING PILOT SPRING PILOT PILOT SEAT BALANCING ORIFICE MAINVALVE SEAT BODY FlGVRE 7. MAlNSTEAM ISOLATlON VALVE
WASHINGTON NUCLEAR PLANT NO.
2 GENERIC PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN Recirculation Pump Trip TABLE RPT-1 IMPORTANCE BASIS AND FAILURE MODE IDENTIFICATION CONDITIONS THAT CAN LEAD TO FAILURE General Guidance Surveillance of the licensee's periodic calibration and testing and/or preventive or unscheduled maintenance activities and procedures, and/or normal and emergency operating procedures, training and check-off lists in accordance with the Technical Specifications and relevant NRC bulletins and information notices should reduce the probability of failure for the conditions listed below.
The most relevant aspects are designated for each condition as fol-lows:
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.
TS - Technical specifications.
S stem Descri tion Mission Success Criteria In the event that there is an 'Anticipated Transient Without Scram (ATWS) at WNP-2 there are several systems designed to mitigate this by reducing reac-tor power, through the insertion of negative reactivity.
One system which was previously described is the Reactor Protection System (RPS).
This system con-sists of a separate set of ATWS scram valves used to remove instrument air from the scram header that keeps the normal scram inlet and outlet valves closed.
- open, then the control rods should insert, shutting down the reactor.
Another ATWS initiation systems is the Recircula-tion Pump Trip (RPT) system which is covered here.
The Standby Liquid Control (SLC) system, which is addressed in a later table, also may be manually initi-ated to mitigate ATWS effects through injection of a neutron absorber solution for reactivity control.
The RPT system provides a mechanism for rapidly reducing core power in the everiCthe normal reactor protection system scram function fails following an event requiring shutdown.
RPT may be accomplished by either one of two in-dependent methods:
~
ATWS panel automatic trip of both recirculation pump motor-generator set field breakers on indication of low-low reactor water level and/
or high reactor pressure (greater than or equal to 1150 psig), or
TABLE RPT-1 (Cont'd)
CONDITIONS THAT CAN LEAD TO FAILURE
~
manual tri.p of both recirculatin pump motor-generator output breakers.
Success is trip of the recirculation pumps either automatically or manually.
Dominant Failure Modes 1.
Generator Field Breaker Fails to Open If the generator field breakers for both Recirculation Pump Motor-Generator (MG) sets fail to open in response to an automatic trip signal then the recirculation pumps will both continue to run and reactor power will not decrease.
0 erator Fails to Tri Recirculation Pum s If the automatic trips do not function, the opertor should manually trip both recirculation pumps.
(OP) 3 Failure of Auto Tri Si nal Due to Miscalibration of Initiation Sensors or Failure of Rela s in Tri Circuitr
~
Drive Motor Breaker Failure If the drive motor breaker for the recirculation MG set fails to open, then flow will continue in the main reactor recirculation loops.
WASHINGTON NUCLEAR PLANT NO.
2 GENERIC PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN Recirculation Pump Trip TABLE RPT-2 I&E INSPECTION PROCEDURES FOR SYSTEM OPERATION PROCEDURE NUMBER TITLE COMPONENTS FAILURE MODES 41200 61725 Training Surveillance Testing Calibration Program Recirc.
pump Recirc.
MG set breakers 1,4 Recirc.
pump control 3
switch 62702 Maintenance (refueling)
Recirc.
MG set breakers 1,4
WASHINGTON NUCLEAR PLANT NO+
2 GENERIC PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN
~
Recirculation Pump Trip TABLE RPT-3 MODIFIED SYSTEM WALKDOWN Verify that there are no alarms in main control panel P602-Verify that the following switches are in the "normal" position:
ATWS Test Switches
~'l W
4' Fr al..
jl i<
01 PW 4IL gt
)fLf1L
'8
~ P P
3.P611 LL La I
TURBINE GOVERNOR VALVEOIL L PRESSURE LOW~
II I
- KSD (PS-SD)
I I
I KBC (PS.SC)
C 1
r I
TURBINE I
THROTTLE VALVE L
) J I
K10D
. (TV-4)
I K10G (TV-3)
I 1
I AUTO BYPASS iIBELOW30%
REACTOR PWR L
J I
I I
I I
I I
I I
I I
I I
I I
I 125 VDC BUS B K9C I
I H13-F609 I
K9D S13D BYPASS SWITCH rr-n
<<e I ~
I LL H13-P 611 RECIRC PUMP 4A TRIP COlL LOCAL K30D (15Hz-INITIATIONI RECIRC PUMP se I
TRIP COIL LOCAL FIGURE 23.
RECIRC P RIP SYSTEM "B" 840108 SEP
't
H13 P609 TURBINE GOVERNOR VALVE L
I I
I I
"1 r
TURBINE THROTTLE VALVE L--------1" r
1 I
AUTO BYPASS BELOW 30%
REACTOR PWR L
J 125 VDC BUS A rr-I.
lL, 4 I
I I
I KBA (PS-5A)
I I
I KBB (PS-5B)
K9B K9A I
K10A (TV-1)
I I
I I
K10F (TV-2)
I I
I I
I I
I I
I I
I I
I I
I I
I I
I S13A BYPASS SWITCH rr-n
~~I ~
I LL H134 609 RECIRC PUMP CB3A TRIP COIL LOCAL K30A (15Hz.INITIATION)
RECIRC PUMP as I
TRIP COIL LOCAL 840 I08,4LT SEPT I886 ABC FIGURE 22.
RECIRC PUMP TRIP SYSTEM "A"
e e'I
WASHINGTON NUCLEAR PLANT NO.
2 GENERIC PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN Standby Liquid Control (SLC) System TABLE SLC-1 IMPORTANCE 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 accordance with the Technical Specifications and relevant NRC bulletins and information notices should reduce the probability of failure for the condi-tions listed below.
The most relevant aspects are designated for each condi-tion as follows:
PC-PT MT OP TS-Periodic calibration activities, procedures and training.
Periodic testing activities, procedures and training.
Preventive or unscheduled maintenance activities, procedures and training.
Normal and emergency operating procedures,,
check-off lists, training, etc.
Technical specifications.
Mission Success Criteria The Standby Liquid Control (SLC) System is a redundant, independent, backup system for the control rod drive hydraulic control system.
It consists of a storage tank, test tank, two 100% capacity
- pumps, two explosive squib valves, and associated
- valves, check valves, piping, instrumentation, and controls necessary to prepare and inject a neutron absorbing solution (sodium penta-borate) into the reactor for backup reactivity control.
The quantity of neu-tron absorber to be injected is based upon the reactivity difference required to bring the reactor from steady state power operation at the most reactive condition anytime in core life to cold shutdown condition with allowances for shutdown margin, imperfect mixing, temperature, and dilution of recirculation loops and reactor water with RHR water for decay heat removal.
The system may be initiated manually from the control room via a three-posi-tive keylock switch under the following conditions:
reactor power unknown or greater than 5% per AP'fN's AND, reactor cannot be shutdown before suppression, pool temperature reaches 110F.
Switching from the OFF position to either "System A" or "System B" will: fire hath explosive squib valves to the open
- position, open both SLC storage tank outlet valves, start the selected
- pump, and isolate SLC from the Reactor Water Cleanup System.
The entire contents of the SLC storage tank are injected into the vessel over the next two hours unless directed otherwise by Emergency Operating Procedures.
Successful injection is verified by observing decreasing reactor power level (approxi-mately 1% per minute from 100% power) and decreasing SLC storage tank level.
TABLE SLC-1 (Cont'd)
CONDITIONS THAT CAN LEAD TO FAILURE Failure Conditions Operator fails to initiate SLC from system keylock control switch due to:
a)
Inadequate assessment of situation (OP) b)
Human reluctance factor (OP) c)
Inadequate keylock switch procedural controls (OP) 2 ~
One of the two pumps SLC-P-1A or SLC-P-1B fails to start and the other out of service for maintenance.
Failure to properly restore system ater testing and maintenance.
b) c)
d) e)f) g)
h)i)j)
& 1B.
SLC pumps suction isolation valves SLC-V-2A & 2B.
SLC pumps discharge isolation valves SLC-V-3A & 3B.
Explosive squib valves SLC-V-4A & 4B.
Test tank inlet
& outlet valves SLC-V-17
& 31.
SLC circulation test valve SLC-V-16.
Demin. supply to SLC pumps suction SLC-V-14.
Storage tank outlet valves SLC-V-1A & 1B.
Inlet to,RPV SLC-V-8 Vent andmain valves.
4 ~
Insufficient boron concentration due to:
a) b)
c) d)
e) improper chemical analysis,(TS) excessive dilution, (OP)
Demin. supply to SLC pump suction SLV-V-14 RWCU system isolation valve RWCV-V-4 inadvertent tank draining, (OP) insufficient mixing in reactor vessel, (OP) inadequate heating of storage tank results in boron precipitation.
(OP,TS)
Loss of flow paths due to:
b) c)
line plugged between tank and pumps, (MT,OP) valves/check valves fail to open, (MT,PT,OP,TS) explosive squib valves SLC-V-4A & 4B fail to fire (monitor licensee controls on explosive valve charges:
procurement,
- storage, surveillance
- testing, shelf life),
storage tank outlet valves SLC-V-1A & 1B fail to open, pump discharge check valves SLC-V-33A & 33B fail to open, outboard injection check valve SLC-V-6 fails to open, inboard injection check valve SLC-V-7 fails to open, tank/line rupture SLC-RV-29A or 29B failure, (MT) system improperly vented/filled.(OP)
WASHINGTON NUCLEAR PLANT No.
2 GENERIC PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN Standby Liquid Control (SLC) System TABLE SLC-2 I&E INSPECTION PROCEDURES FOR SYSTEM OPERATION P ROCEDURE NUMBER TITLE COMPONENTS FAILURE MODES 41700 61725 61726 Training Surveillance and Calibration Control Program Monthly Surveillance Observation Manual initiation switches Storage
- tank, Pumps, Manual Valves Check Valves 2-5 71707 Operational Safety Verification 71710 62702 62703 ESF System Walkdown Maintenance Monthly Maintenance Observation Manual valves, Pumps 2-5
WASHINGTON NUCLEAR PLANT No.
2 GENERIC PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN Standby Liquid Control (SLC) System TABLE SLC-3 MODIFIED SYSTEM WALKDOWN DRAWING NO: H-522 A.
VALVE CHECKLIST VALVE NUMBER LOC.
BLDG.
ELEVE DESCRIPTION REQUIRED ACTUAL POSITION POSITION COMMENTS SLC-V-1A (MO)
Q RX-548 SLC Storage Tank Outlet Valve Closed SLC-V-1B (MO)
Q SLC-V-2AQ SLC-V-2BQ RX-548 RX-548 RX-548 SLC Storage Tank Outlet Valve SLC Pump 1A Suction Valve SLC Pump 1B Suction Valve Closed Locked/
Open Locked/
Open SLC-V-3AQ SLC-V-3BQ RX-548 RX-548 SLC Pump 1A Discharge Valve SLC Pump 1B Discharge Vale Locked/
Open Closed SLC-V-25A Q
SLC-V-25B Q
SLC-V-24A Q
RX-548 RX-548 RX-548 SLC Pumps Discharge Line Drain Valve SLC Pumps Discharge Line Drain Valve SLC Pumps Discharge Line Drain Valve Closed Closed Closed SLC-V-24 B Q
RX-548 SLC Pumps Discharge Line Drain Valve Closed SLC-V-15Q RX-548 SLC Pumps Suction Line Drain Valve Closed SLC-V-31Q RX-548 SLC Test Tank Outlet Valve Locked/
Closed Q Requires independent verification of valve position by licensee personnel.
,TABLE SLC-3 (Cont'd)
'DRAWING NO: M-522 A.
VALVE CHECKLIST VALVE NUMBER LOC.
BLDG.
ELEV.
DESCRIPTION REQUIRED ACTUAL POSITION POSITION COMMENTS SLC-V-189 RX-548 SLC Test Tank Drain Valve Closed SLC-V-14Q RX-548 SLC-V-169 RX-548 DW Addition to SLC Pumps Suction SLC System Circula-tion Test Valve Locked/
Closed locked/
SLV-V-179 RX-548 Q
SLC-V-479 RX-548 SLC Test Tank Inlet Valve SLC Test Line Vent Valve Closed Closed SLC-V-219 RX-548 SLC Test Line Vent Valve Closed SLC-V-109 RX-548 SLC-V-269 RX-522 DW Addition to SLC Storage Tank Vent
& Test Connec-tion Valve at Con-tainment Locked/
Closed Closed SLC-V-279 RX-522 Vent
& Test Connec-tion Valve at Con-tainament Closed SLC-V-409 PC Vent
& Test Connec-tion Valve at RPV Closed SLC-V-419 PC SLC-V-89 PC Vent
& Test Connec-tion Valve at RPV Inlet to PZV Closed Locked/
Open DW-V-159 Q
Rx-548 Demin Water Supply to SLC System Open Drawing M-517 SLC-V-601 RX-548 9
SLC-V-603 RX-548 Q
Test Line Connection Test Line Connection Closed Closed 9 Requires independent verification of valve position by licensee personnel.
TABLE SLC-3 (Cont'd)
DRAWING NO: M-522 A.
VALVE CHECKLIST VALVE NUMBER LOC.
BLDG.
ELEV.
DESCRIPTION REQUIRED POSITION ACTUAL POSITION
.. COMMENTS SLC-V-602 Q
RX-522 Test Line Connection Closed SLC-V-604 (3
RX-522 Test Line Connection Closed SLC-V-459 SLC-V-469 SLC-V-499 PC PC RX-522 Drain Test Connection Closed Drain Test Connection Closed Drain Test Connection Closed (v)
SLC-V-606 Q
RX-522 Drain Test Connection Closed (v)
SLC-V-488 SLC-V-605 Q
RX-548 Test Connection (V)
DW RX-522 2" Manual Valve Closed Closed RWCU-V-4 Later RWCU System Iso Valve Open 9 Requires independent verification of valve position by licensee personnel.
I AWg) ~,
~
gABLE SLC-3 (Cont'd)
DRAWING NO: N-522 B.
POWER SUPPLY POWER SUPPLY MC-8B-CUB
-8A 9 LOC.
BLDG~
ELEVE 522 DESCRIPTION SLC-P-1B SLC Pump 1B and SLC-V-4B REQUIRED POSITION Closed ACTUAL POSITION COMMENTS MC-8B-CUB
-8B 8 522 SLC-EHC-3 Mixing Htr Closed MC-8B-CUB
-8C 9 522 SLC-EHC-2 Maintaining Closed Heater MC-8B-CUB
-9A 0 522 SLC-V-1B SLC Storage Tank Outlet Valve Closed MC-7B-CUB
-8C 8 522 SLC-P-1A SLC Pump 1A and SLC-V-4A Closed MC-7B-CUB
-7D 8 522 SLC-V-lA SLC Storage Tank Outlet Valve Closed MC-later later RWCU-V-4 RWCU System Isolation Valve Closed 8 Requires independent verification of circuit breaker position by licensee personnel.
SLCY 5 ON I 4C V II SA Pe SLCY II U
~4
~LCV I~
4CY445 5LCV4I SlC VII SLC.Y II IIU UVU tAAIAAT TANUCUT SlCY TTI 4CV.IT SlCV.II 5LCV I~
l.x't 0
'LC SCI YJI 0
E C4 WINO UCATCA
<LLLLLLLLLLLLL" SLCCUCI StAACCA 5LC STOIIACC 'IAI>>l SlC IK I Ott AAINO UCIItIT 4C CHCI SlC V40 4CVJ I U
n 5LCYTT t
SLCV.TI a
SLC V4$
SLCV40 5LCY40 SLCV40 SLCV400 4CV40 SLCV40IIOT SLCV401TOT SLCV405 4CV40l e
o IT 0
tt
~
I AUSCII AUSCA 4ICT 5LCVA0 5lCVJA III I eo I
IIII I
I NC tvw NC 4>>A SLCVJIA tVUt A IU>>05TOt
~ UI>>U SLCV4IA SOU..O TT~
tVUt"0 AVWSIC>>
SLCAVII~
SLC.VJA I 0
II IACTLOCC g I
L TO ClOSC OVT SOAAO ANCV5T5TCII ISOLATKWVALYC III SLCtltlt SLC>> 10 II III III II
'N SLCV I~
SLCV IA I
5lCYIIS SLCVJS SLCV.IIS SLCV-II FIGURE 1. STANDSY LIQUIDCONTROL SYSTEM Y
CVASCO AACA 6306STOL'I OCC 1956 SLC
WASHINGTON NUCLEAR PLANT NO.
1 GENERIC PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN High Pressure Core Spray (HPCS)
System TABLE HPCS-3 MODIFIED SYSTEM WALKDOWN DRAWING NO: M-520 I
HPCS PUMP/.iOTOR A.
VALVE CHECKLIST VALVE NUMBER LOC.
BLDG.
ELEV DESCRIPTION REQUIRED POSITION ACTUAL POSITION COMMENTS HPCS-V-1 RB-422 Suction from CST MOV Open HPCS-V-701 RB-422 Root Valve for PS-3 Open HPCS-V-703 RB-422 Root Valve for PX-1 Closed HPCS-V-704 RB-422 Root Valve for PX-1 Closed HPCS-V-80 RB-422 HPCS-P-1 Seal Drain Open HPCS-V-81 RB-422 g
HPCS-P-1 Seal Drain Open HPCS-V-705 RB-422 Roo" Valve for PX-3 Closed HPCS-V-12 RB-422 HPCS-P-1 Minimum Plow MOV Closed HPCS-V-53 RB-422 Q
Minimum Flow Line Iso Locked/
Open HPCS-V-34 RB-422 HPCS-P-3 Suction Iso HPCS-V-77 RB-422 HPCS-P-3 Minimum Flow Open Open HPCS-V-707 RB-422 Root Valve for PI-1G Open HPCS-V-6 RB-422 Root Valve for PIS-13 Open 9 Requires independent verification of valve position by licensee personnel..
~ ~
~
i ~
4 i t ta
%449)
I 4,.
% h'
TABLE HPCS-3 (Con 'd)
DRAWING NO: M-520 Ao VALVE CHECKLIST VALVE NUMBER HPCS-V-708Q LOC.
BLDG.
ELEV.
DESCRIPTION RB-422 Roo" Valve for PIS-13 REQUIRED POSITION Open ACTUAL POSITION COMMENTS HPCS-V-7099 HPCS-V-7108 RB-444 RB-444 Root Valve for FI-5, FIS-6, 6 FI-603 Roo Valve for FI-5, FIS-6 5 FI-603 Open Open HPCS-V-15 RB-444 Suppression Pool Suction MOV Closed HPCS-V-19 RB-444 HPCS Suction Tie to RHR Locked/
Closed HPCS-V-10 RB-444 HPCS-P-1 Test to CST MOV Closed HPCS-V-712 RB-444 Roo" Valve for PI-2 Open HPCS-V-11 RB-444 HPCS-P-1 Test to CST MOV Closed HPCS-V-23 RB-444 HPCS-V-64 RB-444 HPCS-V-65 RB-501 HPCS-V-68 RB-501 HPCS-P-1 Test to Suppression Pool MOV HPCS-P-1 Test to Pool Isolation Control Air Isol for V-5 Control Air Isol for V-5 Closed Locked/
Open Locked/
Closed Locked/
Closed HPCS-V-69 RB-501 Control Air Bleed Vlv Closed HPCS-PI-VX-732 RB-536 HPCS-DPIS-9 (45')
Open 8 Requires independent verification of valve position by licensee personnel.
'ABLE HPCS-3
( Con"')
DRAWING NO: M-520 Ao VALVE CHECKLIST VALVE NUMBER LOC.
BLDG'LEV
~
DESCRIPTION REQUIRED ACTUAL POSITION POSITION COMMENTS HPCS-V-5 HPCS-V-76 Q
DW-547 DW-547 Testable Check Valve Closed Future SLC Connection Closed +
Isolation HPCS-V-37 Q
HPCS-V-38 Q
HPCS-V-51 HPCS-V-102 HPCS-V-3 HPCS-V-31 HPCS-V-4 COND-V-9A 8
COND-V-9B Q
DW-547 DW-547 DW-547 CONT-518 RB-522 RB-522 RB-522 CST Area CST Area Test Connection Iso (55',236')
Test Connection Iso (551',
236
)
Injection Line Iso Valve (551',240')
High Point Vent for DPIS 9
Condensate Flushing Supply Isolation Condensate Flushing Supply isolation Injection Line MOV COND Supply COND-P-3.4.5 COND Supply COND-P-3.4.5 Closed Closed Locked/
Open Closed
+
Locked/
Closed Closed Closed Locked/
Open Locked/
Open COND-V-13A9 COND-V-13B8
'ST ARea CST Area Tes-Return to COND-TK-1A Tes" Return to COND-TK-1B Locked/
Open Locked/
Open 8 Requires independent verification of valve position by licensee personnel.
+ Valve capped
TABLE HPCS-3 (Cont'd)
DRAWING NO: M-520 BE POWER SUPPLY CHECKLIST POWER SUPPLY LOC.
BLDG+
ELEVE DESCRIPTION REQUIRED POSITION ACTUAL POSITION COMMENTS SM-4 MC-4A DG Room High Pressure Core Spray (P-1)
DG Room HPCS Water Leg Pump (P"3) '(CUB 1C)
Racked In Closed See Note MC-4A MC-4A DG Room DG Room HPCS Suction From CST (V-1)(CUB 2D)
HPCS Injection Valve (V-4) (CUB 5B)
Closed Closed See Note MC-4A DG Room HPCS Inboard Return CST (V-10)(CUB 2E)
Closed MC-4A MC-4A DG Room DG Room HPCS Outboard Return to CST (V-ll)(CUB 3A)
HPCS Minimum Flow Valve (V-12)
(CUB 3B)
Closed Closed See Note MC-4A DG Room HPCS Suppression Pool Closed Suction (V-15)(CUB3C)
See Note MC-4A PP-4A DG Room HPCS Test: Valve (V-23) (CUB 3D)
DG Room HPCS-V-5, HPCS-V-51 Ckt 9 Closed Closed PP-4A PP-4A 125 VDC HPCS Dist
'G Room HPCS-V-5 Ckt ll DG Room PT-4, FT-5 Ckt 8 DG Room HPCS Logic Ckt D-7 Closed Closed Closed 125 VDC HPCS Dist DG Room HPCS-V-10 HPCS-V-11 Position Indication Ckt D-9 Closed NOTE:
Licensee personnel are instructed that if these breakers are open, not to close them until directed hy the Control Room Operator, and to see filling and ven ing instruct.'ions.
F
,TABLE HPCS-3 (Cont'd)
~f'I.
HPCS DIESEL GENERATOR Refer to Table EP-4, "Proposed Inspection Plan for Diesel Generators at Nuclear Power Plants."
~
~
~ ~