ML20135E392
| ML20135E392 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 08/22/1996 |
| From: | Baughman M ARIZONA PUBLIC SERVICE CO. (FORMERLY ARIZONA NUCLEAR |
| To: | Mccrory S NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| Shared Package | |
| ML20135E396 | List: |
| References | |
| NUDOCS 9703070027 | |
| Download: ML20135E392 (17) | |
Text
.
Palo Verde Nuclear Tel 601/39.3-6322 Mail Station Z834 Generating Station Fax
'.9 03-0164 P.O. Box 52034 e-r all-Phoenix, AZ 85072-2034 mbt Capsc. corn http://wwe apsc.com 054-02430-RAN/MDB August 22,1996 i
Nuclear Regulatory Commission - Region IV 611 Ryan Plaza Drive, Suite 400 Arlington, Texas 76011-8064 ATTN: Steve McCrory
SUBJECT:
Test Outimes Enclosed is the proposed test outlines for the Initial License Examination to be administered in November 1996 at Palo Verde Nuclear Generating Station.
If you have any questions please feel free to contact me at 602-393-6827.
Sincerel',
Mic ael D. Baugh an Section Leader: OPS Requal Training 1
MDB/cgc
)
i Enclosure i
i i
t 9703070027 960822 PDR ADOCK 05000528 PDR y
I Kr;owledge crd Abilities Rec:rd F rm PLANT-WIDE GENERIC RESPONSIBILITIES PWR - Senior Reactor Operator - 17%
Check if. -
-294001-Included K/A #
Statement Rating V
2.1.1 Knowledge of conduct of operations 3.8 requirements.
4 2.1.9 Ability to direct personnel activities inside the 4.0 control room.
V 2.1.10 Knowledge of conditions and limitations in the 3.9 facility license.
V 2.1.11 Knowledge ofless than one hour technical 3.8 specification action statements for systems.
V 2.1.20 Ability to execute procedure steps.
4.2 4
2.2.6 Knowledge of the process for making changes in 3.3 procedures as described in the safety analysis report.
V 2.2.13 Knowledge of tagging and clearance procedures.
3.8 4
2.2.23 Ability to track limiting conditions for operations.
3.8 4
2.2.30 Knowledge of new and spent fuel movement 3.5 procedures, j
V 2.3.1 Knowledge of 10 CFR 20 and related facility 3.0 radiation control requirements.
4 2.3.4 Know! edge of radiation exposure limits and 3.1 contam sation control, including permissible levels in exer, of those authorized.
4 2.3.9 Knowledge of the process for performing a 3.4 containment purge.
4 2.4.1 Knowledge of EOP entry conditions and 4.6 immediate avion steps.
4 2.4.38 Ability to tai actions called for in the facility 4.0 emergency plan, including (if required) supporting or acting as emergency coordinator.
4 2.4.41 Knowledge of the emergency action level 4.1 thresholds and classifications.
V 2.4.42 Knowledge of emergency response facilities.
3.7 4
2.4.48 Ability to interptet control room indications to 3.8 verify the status and operation of system, and understand how operator actions and directives affect plant and system conditions.
Examiners'llandbook 1
WAWORDNRC. DOC
1 i
Kn=dedge ced Abilities Rec:rd F rm
(
PLANT SYSTEMS PWR - Senior Reactor Operator-40%
Group I Plant systems.19%
h 001 Control Rod Drive System 026 Containmem Spray System 003 Reactor Coolant Pump System 056 Condensate System j
004 Chemical and Volume Contml System 059 Main Feedwater System 013 Engineered Safety Features Actuation 061 Auxiliary / Emergency Feedwater l
System System j
014 Rod Position Indication System 063 DC Electrical distribution System 015 Nuclear Instrumentation System 068 Liquid Radwaste System l
017 In-Core Te nperature Monitor System 071 Waste Gas Disposal System 022 Containment Cooling System 072 Area Radiation Monitoring System
- System #-
K/A #.
K/A Topic Rating i
001 K5.04 Operational implications of rod insertion limits.
4.7 001 A4.11 Determination of SDM.
4.1 i
003 A2.01 Problems with RCP seals, especially rates of seal 3.9 i
leak-off.
l 004 K3.07 Malfunction of CVCS effect on PZR level and 4.1 pressure.
j 013 Kl.01 Initiation signals for ESF circuit logic.
4.4 i
l 014 A2.04 Predict impacts ofmisaligned rod 3.9 1
i i
j 015 K3.01 NIS malfunction impact on RPS.
4.2 i
017 A3.01 Indications of normal, natural, and interrupted 3.8*
l circulation of RCS.
j j
022 K2.01 Power supplies to containment cooling fans.
3.1 026 A2.02 Failure of automatic recirculation transfer.
4.4 i
i 056 A2.04 Predict impacts of a loss of condensate pumps.
2.8*
?
059 K4.16 Automatic trips for MFW pumps.
3.2*
061 A3.04 Monitor automatic AFW isolation.
4.2 061 K4.02 AFW automatic start upon loss of MFW pump, 4.6 J
S/G level, blackout, or safety injection.
063 K3.01 Effect of a loss of DC on EDG.
4.1 068 K4.01 Safety and environmental precautions for handling 4.1 hot, acidic, and radioactive liquids.
071 A4.29 Monitor in control room sampling oxygen, 3.6*
hydrogen, and nitrogen concentrations in the WGDS decay tank, knowledge oflimits.
Examiners'ilandbook 2
Knxiledge and Abilities Rec:rd Fcrm PLANT SYSTEMS PWR - Senior Reactor Operator-40%
072 K4.02 ARM system interlocks for fuel building isolation.
3.4
- 004 A4.07 Manually operate boration/ dilution.
3.7 Group II Plant Systems - 17% -
002 Reactor Coolant System 035 Steam Generator System 006 Emergency Core Cooling System 039 Main and Reheat Steam System 010 Pressurizer Pressure Control System 055 Condenser Air Removal System 011 Pressurizer Level Control System 062 AC Electrical Distribution System 012 Reactor Protection System 064 Emergency Diesel Generating System 016 Non-Nuclear Instrumentation System 073 Process Radiation Monitoring System 028 Hydrogen Recombiner and Purge 075 Circulating Water System Control System 029 Containment Purge System 079 Station Air System 033 Spent fuel Pool Cooling System 086 Fire Protection System 034 Fuel Handling Equipment System 103 Containment System System #
. K/A #
K/A Topic Rating 002 A4.02 Indications necessary to verify natural circulation 4.5 from appropriate level, flow, and temperature indications and valve positions upon loss of forced circulation.
006 A4.07 Manually operate ECCS pumps and valves.
4.4 010 K4.03 PZR PCS design features providing over pressure 4.1 control.
011 A2.10 Predict impacts for failure of PZR level 3.6 instrument-high.
012 K4.02 Automatic reactor trip when RPS setpoints are 4.3 exceeded for each RPS function, bases for each.
012 A2.02 Predict impacts on RPS for loss ofinstrument 3.9 power.
012 A4.04 Operate / monitor bistable, trips, reset and test 3.3 switches.
016 A2.02 Predict impacts on NNIS for loss of power 3.2*
supply.
034 A1.02 Predict changes in parameters to prevent 3.7 exceeding design limits associated with the Fuel Handling System including water level in refueling canal.
Examiners'Ilandbook 3
Kn:wledge end Abilities Rec:rd Fcrm PLANT SYSTEMS PWR - Senior Reactor Operator-40%
035 K4.01 Design features of the S/GS which provide S/G 3.8 level control.
039 K4.05 Design features / interlocks which provide 3.7 automatic isolation of steam line.
039 A4.04 Manually operate and/or monitor from the control 3.9 room the emergency feedwater pump turbine.
062 K4.03 AC distribution system design features / interlocks 3.1 between automatic bus transfer and breakers.
062 K2.01 Bus power supplies to major system loads.
3.4 064 K4.02 ED/G design features / interlocks which provide 4.2 trips while operating (no,rmal or emergency).
086 K4.01 Design features for adequate supply of water for 3.7 Fire Protection System.
103 Kl.02 Relationship between containment and
- 4. l
- containment isolation / integrity.
1 Examiners'11andtxxA 4
l 1
Kn:;wledge c:d Abilities Rec::rd Fcrm PLANT SYSTEMS 4
PWR - Senior Reactor Operator-40%
j i
~
Group 111 Plant System - 4%
005 Residual Heat Removal System 045 Main Turbine Generator System 007 Pressurizer Relief Tank / Quench Tank 076 Senice Water System I'
System I
008 Component Cooling Water System 078 Instrument Air System 041 Steam Dump System / Turbine Bypass Control
' System #.
K/A #
K/A Topic Rating :
1
\\
l 005 K4.01 RHR design features / interlocks for overpressure 3.2
{
protection.
007 A4.10 Recognize leaking PZR code safety.
3.8 041 A2.03 Predict impacts of and mitigate consequences of a 3.1 loss ofInstrument air system on Steam Dump System.
045 K4.12 Design features / interlocks for MT/G automatic 3.6 turbine runback.
i Examiners'Ilandbook 5
i Kn:wledge c:d Abilities Rec:rd Fcrm EMERGENCY PLANT EVOLUTIONS PWR - Senior Reactor Operator-43%
l Group I Emergency _and Abnormal Plant Evaluations - 24%
a 001 Continuous Rod Withdrawal 051 Loss of Condenser Vacuum 003 Dropped Control Rod 055 Loss of Offsite and Onsite Power 005 Inoperable / Stuck Control Rod 057 Loss of Vital AC Electrical Instrument Buss l
011 Large Break LOCA 059 Accidental Liquid Radioactive-Waste Release 1
015 RCP Motor Malfunction 067 Plant Fire On Site l
024 Emergency Boration 068 Control Room Evacuation d
026 Loss of Component Cooling Water 069 Loss of Containment Integrity
)
029 Anticipated Transient Without 074 Inadequate Core Cooling j
Scram 040 Steam Line Rupture 076 High Reactor Coolant Activity 4
E/A #
K/A #
K/A Topic
. Rating i
001 AKl.02 SUR during Continuous Rod Withdrawal.
3.9 003 AK3.08 Knowledge of criteria for inoperable control rods.
4.2 005 AKl.06 Bases for power limit for rod misalignment.
3.8 011 EK3.14 RCP tripping requirement during LOCA 4.2 l
011 EA2.11 Determine or interpret the conditions for 4.3 throttling or stopping High Pressure Injection.
j 015 AA1.22 Ability to monitor for RCP seal 4.2 j
failure / malfunction.
024 AK3.01 Knowledge of when emergency boration is 4.4 3
j required.
029 EK2.06 Interrelationship between breakers, relays, and 3.l*
disconnects during ATWS.
040 AK3.04 Actions contained in EOPs for steam line rupture.
4.7 1
040 AA1.09 Ability to monitor setpoints of main steam safety 3.4
{
valves.
040 AA2.04 Determine / interpret conditions requiring ESFAS 4.7 initiation for steam line rupture.
051 AK3.01 Loss of steam dump capability upon loss of
- 3. l
- condeser vacuum.
055 EA2.02 Determine / interpret RCS core cooling through 4.6 1
natural circulation cooling to S/G cooling.
057 AK3.01 Actions contained in EOP for loss of vital AC 4.4 1
instrument bus.
Examiners'Ilandbmk 6
K=wledge c:d Abilities Rec:rd F:rm EMERGENCY PLANT EVOLUTIONS PWR - Senior Reactor Operator-43%
059 AKl.02 Biological effects on humans of various types of 3.1 radiation, exposure levels that are acceptable for nuclear power plant personnel, and units used for radiation intensity measurements and for exposure levels.
067 AA2.05 Determine ventilation alignment necessary to 3.6 secure affected area for Plant Fire on Site.
068 AA1.01 Operate S/G atmospheric dump valves following 4.5 i
Control Room Evacuation.
069 AA2.02 Verification of automatic and manual means of 4.4 restoring containment integrity.
074 EKl.01 Methods of calculating subcooled margin.
4.7 076 AIO.05 Corrective actions as a result of high fission-3.6 product radioactivity level in the RCS.
040/E05 EA2.1 Determine / interpret facility conditions and 4.0 selection of appropriate procedures for Excess Steam Demand.
026 AK3.02 Automatic actions within CCW/ nuclear cooling 3.9 water resulting from actuation of the ESFAS.
011 EK2.02 Interrelation between pumps and Large Break 2.7*
LOCA.
029 EK3.12 Actions contained in EOP for ATWS.
4.7 Group II Emergency and Abnormal Plant Evaluations - 16% :
007 Reactor Trip 037 Steam Generator Tube Leak 008 Pressurizer Vapor Space Accident 038 Steam Generator Tube Rupture 009 Small Break LOCA 054 Loss ofMain Feedwater 022 Loss ofReactor Coolant Makeup 058 Loss ofDC Power 025 Loss of Residual Heat Removal 060 Accidental Gaseous-Waste Release System 027 Pressurizer Pressure Control System 061 Area Radiation Monitoring System Malfunction Alarms 032 Loss of Source-Range Nuclear 065 Loss ofInstrument Air Instmmentation E/A #
K/A #
K/A Topic Rating 007 EKl.06 Relationship of emergency feedwater flow to S/G 4.1 and decay heat removal following reactor trip.
008 AA1.07 Reseating of PZR code safety valve.
4.2 009 EK3.20 Tech spec leakage limits.
4.3 Examiners' Handtak 7
Knnyledge c:d Abilities Rec rd Form EMERGENCY PLANT EVOLUTIONS PWR - Senior Reactor Operator - 43%
025 AK3.01 Shift to alternate flowpath on loss of RHR.
3.4 027 AA2.04 Determine / interpret Tech Spec limits for RCS 4.3 pressure.
037 AA2.02 Agreement / disagreement among redundant 3.9 radiation monitors.
038 EK2.02 Interrelations between sensors and detectors for a 2.5 SGTR.
038 EA2.15 Determine / interpret pressure at which to maintain 4.6 i
the RCS during S/G cooldown following a SGTR.
038 EK3.02 Prevention of secondary code safety valve cycling 4.5 during a SGTR.
054 AK3.04 Actions contained in EOPs for loss of Main 4.6 Feedwater.
054/E06 EA2.1 Facili y conditions and selection of appropriate 3.9 t
procedures during abnormal and emergency operations.
058 AA2.03 DC loads lost; impact on ability to operate and 3.9 monitor plant systems.
058/E09 EA2.1 Facility conditions and selection of appropriate 4.4 procedures for a loss of DC as applied to the FRP.
{
060 AK2.02 Interrelations between accidental gaseous 3.1 radwaste release and Auxiliary Building ventilation system.
061 AA2.06 Required actions if an Area Radiation Monitor 4.1 alarm channel is out of service.
065 AK3.04 Cross-over to backup air supplies upon loss of 3.2 instrument air.
Group III Emergency and Abnormal Plant Evaluations - 3%.
028 Pressurizer Level Malfunction 056 Loss of Offsite Power j
036 Fuel Handling Incident E/A #-
K/A # -
K/A Topic Rating 028 AKl.01 Operational implications of PZR reference leak 3.1 abnormalities.
036 AK3.03 Guidance contained in EOP for fuel handling 4.1 incident.
056 AA2.88 Determine / interpret necessary S/G water level for 4.2 natural circulation.
Examiners
- 11andbook g
i
. ~
i, j
Apoendix D Scenario Outline Form ES-D-1 Facility: PVNGS Scenario No.:
1 Op-Test No:
4 i
4 i
Examiners:
Operators:
1 1
t i,
Objectives:
h 3
)
Initial Conditions: 100% Power, Equilibrium Xenon, Middle of Cycle i
l Turnover: The following equipment is out of service:
Essential Auxiliary Feedwater pump 'A', to repair the governor linkage.
e Nuclear Cooling Water pump 'B' to replace a pump casing drain valve.
Stator Cooling pump 'A' to replace the motor.
Suspect slow N leak on Safety injection Tank 1A. After turnover repressurize Tank 1A 2
Containment entry in two hours to mvestigate.
Event Malf.
Event Event Description l
No.
No.
Type
- l 0
N Repressurize Safety injection Tank 1 A I
RD02A C/R Drop control rod 14 and recover.
2 TROI:
1 Steam Generator level transmitter LTil21 fail high to #2 SGNLTil21 Feedwater Control System.
3 MS07 M
Steam break in Turbine Building.
ED10A C
e Loss NBN-X03 transformer l
CP03:
C Essential auxiliary feedwater pump 'B' performance e
AFBP01 degrades j
i EG06A C
e After feed established with Non-essential auxiliary feedwater pump 'N', Diesel Generator 'A' trips on l
differential.
Transition to Functional Recovery Procedure to recover PBA-S03 from NBN-X04.
4 i
l
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Examination Standards Rev. 8, PUBLIC i
Appendix D Scenario Outline Form ES-D-1 Facility: PVNGS Scenario No.:
2 Op-Test No:
Examiners:
Operators:
Objectives:
Initial Conditions: 100% Power, Equilibrium Xenon, Middle of Cycle.
Turnover: Control Element Assembly operability checi ' vi 9 T-ISF01 are in progress. Continue the surveillance test after turnover. Instrument A. t ;< gressor 'A' is out of service for motor bearing replacement. Charging pump 'E' is out or.ervice for plunger repair.
Event Malf. No.
Event Event Description No.
Type
- 0 N
Continue 4IST-1SF01 surveillance 1
BS02:
1/C Nuclear Cooling Water flow switch FS613 fails low causing loss of NCNFSL613 letdown.
2 MC01A C/R Loss of Condenser Vacuum 3
RP06Al C
Inadvertent Main Steam Line Isolation CAE C
ATWS RV02:
M Steam Generator #1 safety valve opens SGEPSV572 I
RP078 1/C
'B' Train Sequencer failure.
CP03:
C
'A' High Pressure Safety Injection pump performance degrades.
SIAP02 Safety valve reseats at 500 psig.
- (N)ormal, (R)eactivity, (I)nstmment, (C)omponent, (M)ajor Examination Standards Rev. 8, PUBLIC
,1
Aopendix D Scenario Outline Form ES-D-1 Facility: PVNGS Scenario No.:
3 Op-Test No:
1 Examiners:
Operators:
Objectives:
Initial Conditions: 26% power and increasing for plant startup, Middle of Cycle.
i Turnover: Plant startup in progress. Pressurizer is in boron equalization. High pressure safety injection pump 'A' is out of service for pump inspection due to failure of its sunreillance test. After turnover continue the plant startup in 400P-9ZZ05.
Event Malf. No.
Event Event Description No.
Type
- 0 N/R Power increase 1
NIO2A I
Nuclear Instruement Control Channel fails low.
N102B l
2 RC02A C
Stuck open pressurizer spray valve, close instrument air, loss of letdown.
3 TH06B C
Steam Generator #2 tube leak.
TH06B M
Steam Generator #2 tube rupture.
ED02 C
Loss offsite power post trip BK02:
C
'B' Diesel Generator output breaker fails to auto close.
PBBSO4B CP05:
C
'B' Spray pond pump fails to auto start.
SPBP01
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Examination Standards Rev. 8, PUBLIC
l
\\
l Anoendix D Scenario Outline Form ES-D-1 l
3-J t
I Facility: PVNGS Scenario No.:
4 Op-Test No:
i l
Examiners:
Operators:
i i
Objectives:
i Initial Conditions: 100% power, Equilibrium Xenon, Middle of Cycle.
Turnover: Diesel Generator 'A' is out of service due to Field Flash circuit failure during its monthly surveillance test. Diesel Generator 'B' monthly operabil,ty test in progress in 41ST-IDG01.
One hour remaining on the run, at step 8.16.3. Turbine Cooling Water pump 'B' is out of service for inspection due to high vibration.
Event Malf.
Event Event Description No.
No.
Type
- 0 N
Diesel Generator 'B' test in progress.
1 EG06B C
Diesel Generator 'B' differential trip.
N/R Commence plant shutdown.
2 TROI:
I Pressurizer PI-100X fails high.
RCNPT100X j
3 FW15A C
Main feed pump 'A' high vibration, manual trip pump 'A' with no auto reactor power cutback.
EXO6A I
RX06B I
4 HX02:
M Reactor Coolant Pump 1 A High Pressure Seal cooler leak.
RCEE05A RD03A/F C
Two stuck control rods on reactor trip.
j BK05:
C Loss of Load Center LO9 and M15.
NGNLO9B2 1
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor i
I Examination Standards Rev. 8, PUBLIC
Appendix D Scenario Outline Form ES-D-1 Facility: PVNGS Scenario No.: Standbv Op-Test No.:
Examiners:
Operators:
l Objecti/es:
Initial Conditions: 50% power, Middle of Cycle,36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> post trip recovery.
Turnover: Essential Auxiliary Feedwater pump 'B' is out of service for motor bearing replacement.
Pressurizer level instrument Ll-110X is failed. Control selected to L1-110Y. After turnover place 'B' main feedwater pump in service and raise power level to 100% in accordance with 400P-9ZZ05.
1 Event Malf. No.
Event Event Description No.
Type
- i 0
N/R Place 'B' main feedwater pump in service, raise power level.
1 CVI fB 1
Pressurizer level LI-l 10Y fails low.
2 CC01A C
Loss of Nuclear Cooling Water, cross-tie Essential Cooling Water to Nuclear Cooling Water, loss ofletdown.
CP03:
C NCNP018 3
TCl1 I
Main Turbine EHC failure results in turbine runback.
CAE C
Reactor fails to automatically trip.
RV02:
M Pressurizer safety valve sticks open 10%
RCEPSV200 CP05:
C
'A' High Pressure safety injection pump fails to automatically start.
SIAP02 AV01:
C Steam Generator #1 Economizer sticks 5% open.
SGNFVill2
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Examination Standards Rev. 8, PUBLIC
~ -.
ES-301 Administrative Topics Outline Form ES-301-1 Facility: PVNGS Date of Examination:
Examination Level (Circle): RO /@
Operating Test Number:
Administrative Describe method of evaluation:
Topic / Subject
- 2. TWO Administrative Questions
]
A.1 Fuel liandling Explain actions to be taken for a loss of spent fuel pool level with fuel assembly movement m progress 1
(N) 033 A2.03, 3.5 Describe sources of makeup water to recover spent fuel pool level.
(N) 033 A4.06, 2.9 Reactor Startup Determine required actions based on anticipated critical position data.
(N) 2.1.7 4.4 l
l Requirements for Mode change checklist, Mode 3 to 2.
l (N) 2.1.23 4.0 A.2 Tagging and Demonstrate manual tagout/ clearance generation for ADV178.
Clearances (N) 2.2.13 3.8 Determine / explain retest requirements necessary to declare ADV178 operable.
(N) 2.2.21 3.5 A.3 Radiation Control Radiological Entry Permit discussion and containment entry at power.
(N) 2.3.43.1 liot particle contamination control requirements.
(N) 2.3.4 3.1 A.4 Emergency Plan JPM EP018 Classify an event and determine PARS.
l (N) 2.4.38 4.0 l
Examiners Standards R:v. 8, PUBLIC l
l
f ES-301 Individual Walk-Through Test Outline Form ES-301-2 Facility: PVNGS Date of Examination:
Examination Level (Circle): RO /@
Operating Test Number:
System / JPM Tile / Type Safety Planned Follow-up Question:
Codes Function K/A/G // Importance // Description
- l. AF EM007 Local / Man
- a. EK3.02, 4.6 start of non-essential Auxiliary HR Restrictions on running non-essential Feedwater Pump Auxiliary Feedwater pump during station blackout (N)
(A)
(P) b.061 A2.05, 3.4 Conditions that will trip the non-essential Auxiliary Feedwater pump breaker. (N)
- 2. DG DG007 Recover
- a. 064 A4.06, 3.9 i
4 Emergency DG during Station ELE Voltage and frequency adjustments while in Blackout
" Emergency Run"(N) j Emerg Action (P)
- b. 064 A2.05, 3.2 Diesel generator loading limits during station blackout crosstie operations. (D) l
- 3. SI S1036 Shift
- a. 005 A4.02, 3.1 Shutdown Cooling Purification HR Explain how shutdown cooling purification to "B" Train flowrate is controlled. (N) i (R)
(P)
- b. 005 A4.01, 3.4 Explain control room indications of the i
symptoms of air entrainment while on shutdown cooling.(D)
- 4. Cil CH003 Borate The
- a. 004 K4.12, 3.4 I
Reactor Coolant System RC Affects of a failure of volume control tank level instrument LI-227. (N)
(A)
(C)
- b. 004 A2.14, 3.9 Reason why CHA-UV-501 must be closed when borating through CHB-UV-536. (N) 5 Cil CH014 Place Reactor
- a. 003 Kl.03,3.6 Coolant Pump Seal injection in PS Importance of supplying Reactor Coolant service pump seal injection flow. (D)
(L)
(C)
- b. 003 K6.02, 3.1 Required actions for loss of Reactor Coolant Pump seal injection and Nuclear Cooling Water. (D)
- 6. SBCS SF027 Reconnect
- a. 016 K1.03, 3.2 the Steam Bypass Control INST.
Describe the requirements to put the Steam System Bypass Control System in " Test". (N)
(C)
- b. 016 A3.01, 2.9*
Conditions causing a Steam Bypass Control System " Quick Open Block". (N)
Examiners Standards Rev. 8, PUBLIC J
l i
ES-301 Individual Walk-Through Test Outline Form ES-301-2
- 7. EW EW001 Recover
- a. 008 A3.04, 3.2 from cross-tie Essential PS Operability 6f an Essential Cooling Water Cooling Water to Nuclear train cross-connected with Nuclear Cooling
{
Cooling Water Water. (D)
Time Crit (C)
- b. 008 K3.01, 3.5 Nuclear Cooling Water priority loads supplied by Essential Cooling Water in cross-a tied operations. (D) 8.EP EP003 Immediate
- a. AK3.1, 3.7 actions evacuate the control PC Natural circulation cooldown strategies to room due to fire keep RCS pressure high. (D)
Time Crit (C)
- b. AKl.2, 3.5 Reasons for closing main steam isolation valves before control room evacuation. (N) i 9.COLSS SF031 Actions for
- a. APE 003 AK3.08, 4.2 COLSS out of service INST.
Determining Technical Specification actions (DNBR, LIIR) and requirements for an immovable or
{
untrippable control rod. (N)
{
(N)
(C)
- b. APE 003 AKA3.04, 4.1 Determining the affects on the Core Operating Limits Supervisory System if a rod bottom contact is made up. (N)
- 10. SG SG018 Perform Main
- a. 039 K3.05, 3.7 Steam Isolation Valve test Cl Plant response if a Main Steam Isolation exercise stroke Valve stroked closed at power. (N)
(N)
(C)
- b. 039 K4.05, 3.7 Effects of solenoid failures on Main Steam Isolation Valve stroke tests. (N)
AF Auxiliary Feedwater LHR Linear Heat Rate DG Emergency Diesel Generator SBCS Steam Bypass Control System CH Chemical & Volume Control Sys.
EW Essential Cooling Water i
SI Safety Injection System EP Emergency Procedures COLSS Core Operating Limits Suprv. Sys.
SG Main Steam System DNBR Departure from Nucleate Boiling Ratio i
- - Typer Codes: (D)irect from bank, (M)odified from bank, (N)ew, (A)lternate path, (C)ontrol room, (S)imulator,(L)ow power, (P)lant, (R)CA Examiners btandards Rev. 8, PUBLIC