ML111010402
| ML111010402 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 01/24/2011 |
| From: | Entergy Nuclear Generation Co |
| To: | NRC Region 1 |
| Hansell S | |
| Shared Package | |
| ML102210114 | List: |
| References | |
| TAC U01833 | |
| Download: ML111010402 (48) | |
Text
ES-401 Written Examination Outline Form ES-401-1 Facility: Pilgrim NRC Exam Date of Exam: 01/21/11 RO KIA Category Points SRO-Only Points Tier Group K K K K K K A A A A G Total A2 G* Total i 1 2 3
- 1.
1 4 3 4 3 20 3 4 7 Emergency 2 1 1 1 1 7 1 2 3 Plant IEvaluations Tier 5 4 5 4 27 4 6 10 Totals 1 2 2 4 2 1 4 3 1 3 2 2 26 2 3 5 2.
2 1 1 2 1 1 1 1 1 1 1 12 0 2 1 3 Plant Systems Tier 3 3 5 4 2 5 4 2 4 3 3 38 4 4 8 Totals
- 3. Generic Knowledge & Abilities 2 3 4 10 1 2 3 4 7 3 2 2 a 1 2 2 2 Note 1. Ensure that at least two topicS from every applicable KJA category are sampled within each tier of the RO and SRO-only outlines (I.e., except for one category in Tier 3 of the SRO-only outline, the Tier Totals in each KJA category shall not be less than two).
- 2. The point total for each group and tier in the proposed outline must match that specified in the table.
The final point total for each group and tier may deviate by 1 from that specified in the table based on NRC revisions. The final RO exam must total 75 points and the SRO-only exam must total 25 points.
- 3. Systems/evolutions within each group are identified on the associated outline; systems or evolutions that do not apply at the facility should be deleted and justified; operationally important, site-specific systems that are not included on the outline should be added. Refer to section 0.1.b of ES-401, for guidance regarding elimination of inappropriate KJA statements.
- 4. Select topics from as many systems and evolutions as possible; sample every system or evolution in the group before selecting a second topic for any system or evolution.
- 5. Absent a plant specific priority, only those KAs having an importance rating (IR) of 2.5 or higher shall be selected. Use the RO and SRO ratings for the RO and SRO-only portions, respectively.
- 6. Select SRO topics for Tiers 1 and 2 from the shaded systems and KIA categories.
7.* The generic (G) KJAs in Tiers 1 and 2 shall be selected from Section 2 of the KJA Catalog, but the topiCS must be relevant to the applicable evolution or system. Refer to Section 0.1 ,b of ES-401 for the applicable KJA's
- 8. On the following pages, enter the KJA numbers, a brief description of each topic, the topics' importance ratings (IR) for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above. If fuel handling equipment is sampled in other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2 (Note #1 does not apply). Use duplicate pages for RO and SRO-only exams.
- 9. For Tier 3, select topics from Section 2 of the KJA Catalog, and enter the KJA numbers, descriptions, IRs, and point totals (#) on Form ES-401-3. Limit SRO selections 10 KlAs that are linked to 10CFR55,43
ES-401 Form ES-401-1 Pilgrim 7 Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 1 EAPE#/Name Safety Function KIA Topic(s)
EA2.01 - Ability to determine ancl/or interpret the following
&..;v,,,vv1 Reactor Low Water as they apply to REACTOR 4.6 76 Level / 2 LOW WATER LEVEL:
Reactor water level AA2.03 - Ability to determine and/or interpret the following I L...." . " ..... Partial or Total as they apply to PARTIAL 2.9 77
.Loss of DC Pwr / 6 OR COMPLETE LOSS OF D.C. POWER: Battery AA2.04 - Ability to determine and/or interpret the following 295003 Partial or Complete as they apply to PARTIAL 3.7 78
.Loss of AC / 6 OR COMPLETE LOSS OF A.C. POWER: System 2.4.41 - Emergency Procedures / Plan:
600000 Plant Fire On-site / Knowledge of the 4.6 79 emergency action level thresholds and classifications.
2.4.47 - Emergency Procedures / Plan: Ability to dia~~nose and recognize 295026 Suppression Pool trends in an accurate and 4.2 80 High Water Temp. /5 timely manner utilizing the appropriate control room refe*rence material.
2.4.4 - Emergency Procedures / Plan: Ability to recognize abnormal 295019 Partial or Total indications for system 4.7 81 Loss of Inst. Air / 8 operating parameters which are entry-level conditions for emmgency and abnormal rocedures.
ES-401 Form ES-401-1 Pilgrim 7 Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 1 EAPE#/Name Safety Function KIA Topic(s) 2.4.30 - Emergency Procedures / Plan; Knowledge of events related 700000 Generator Voltage to system operation / status and Electric Grid that must be reported to 4.1 82 Disturbances internal organizations or external agencies, such as the state, the NRC, or the transmission system EK1.03 - Knowledge of the operational implications of the following concepts as 295037 SCRAM Conditions they apply to SCRAM Present and Reactor Power X CONDITION PRESENT Above APRM Downscale or 4.2 39 AND REACTOR POWER Unknown /1 ABOVE APRM DOWNSCALE OR UNKNOWN: Boron effects on reactor .... ",.u*...*
AK1.03 - Knowledge of the operational implications of 295005 Main Turbine the following concepts as X they apply to MAIN 3.5 40 Generator Trip /3 TURBINE GENERATOR TRIP: Pressure effects on reactor level AK"l.01 - Knowledge of the operational implications of the following concepts as 295018 Partial or Total they apply to PARTIAL OR Loss of CCW / 8 X COMPLETE LOSS OF 3.5 41 COMPONENT COOLING WATER: Effects on component/system AK2.06 - Knowledge of the 295006 SCRAM /1 X interrelations between i 4.2 42 SCRAM and the following:
Reactor Power AK2.02 - Knowledge of the interrelations between 295016 Control Room CONTROL ROOM donment /7 X 4.0 i43 ABANDONMENT and the following: Local control stations: Plant-...:....'... I",t,I"
ES-401 Form ES-401-1 Pilgrim 7 Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 1 EAPE#/Name Safety Function KIA Topic(s) limp.' Q# I EK2.02 - Knowledge of the interrelations between HIGH 295028 High Drywell DRYWELLTEMPERATURE Temperature / 5 3.2 44 and the following:
Components internal to the AK3.02 - Knowledge of the reasons for the following responses as they apply to 700000 Generator Voltage GENERATOR VOLTAGE and Electric Grid AND ELECTRIC GRID 3.6 45 Disturbances DISTURBANCES: Actions contained in abnormal operating procedure for voltage and grid disturbances.
EK3.05 - Knowledge of the reasons for the following Suppression Pool responses as they apply to High Water Temp. /5 SUPPRESSION POOL 3.9 46 HIGH WATER TEMPERATURE: Reactor SCRAM EK~3.07 - Knowledge of the 295024 High Drywell reasons for the following Pressure /5 responses as they apply to 3.5 47 HIGH DRYWELL PRESSURE: ventin AA 1.04 - Ability to operate and/or monitor the following 1'-.:i'vVC,_v Refueling as they apply to Accidents / 8 X 3.4 48 REFUELING ACCIDENTS:
Radiation monitoring ent.
AA 1.06 - Ability to operate 600000 Plant Fire On-site / and / or monitor the following 8 3.0 49 as they apply to PLANT FI RE ON SITE: Fire alarm AA1.03 - Ability to operate and/or monitor the following 295003 Partial or Complete as they apply to PARTIAL Loss of AC /6 OR COMPLETE LOSS OF 4.4 50 A. G., POWER: Systems necessary to assure safe nt shutdown
ES-401 Form ES-401-1 Pilgrim 7 Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 1 EAPE#/Name Safety Function KIA Topic(s)
EA2.04 - Ability to determine 295038 High Off-site and/or interpret the following Release Rate / 9 as they apply to HIGH OFF 4.1 51 SITE RELEASE RATE:
Source of off-site release AA2.02 - Ability to determine and/or interpret the following as they apply to PARTIAL 295019 Partial or Total OR COMPLETE LOSS OF Loss of Inst. Air / 8 3.6 52 INSTRUMENT AIR: Status of safety-related instrument air system loads (see AK2.1
- AK2.1 Reactor Low Water Level /2 4.6 i53 core cool 2.4.. 18 - Emergency 295025 High Reactor Procedures / Plan:
Pressure /3 3.3 54 Knowledge of the specific bases for EOPs.
2.1.7 - Conduct of Operations: Ability to evaluate plant performance 295021 Loss of Shutdown and make operational Cooling /4 4.4 55 judgments based on operating characteristics, reactor behavior, and instrument inte AK3.03 - Knowledge of the reasons for the following Partial or Total responses as they apply to Loss of DC Pwr / 6 X 3.1 56 PARTIAL OR COMPLETE LOSS OF D.C. POWER:
Rea.ctor scram EK1.01 - Knowledge of the operational implications of 295030 Low Suppression the following concepts as X they apply to LOW 3.8 57 Pool Water Level / 5 SUPPRESSION POOL WATER LEVEL: Steam condensation.
ES-401 Form ES-401-1 Pilgrim 7 Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 1 EAPE#/Name Safety Function KIA Topic(s) limp. I Q# I 2.2~.37 - Equipment Control:
295001 Partial or Complete Ability to determine Loss of Forced Core Flow operability and/or availability 3.6 58 iCirculation / 1 & 4 of safety related I KIA CategoryTotals 4 3 4 3
ES-401 Form ES-401-1 Pilgrim 7 Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 2 EAPE#/Name Safety Function KIA Topic(s)
EA2.02 - Ability to determine and/or interpret the following 295032 High Secondary as they apply to HIGH Containment Area SECONDARY 3.5 83 Temperature /5 CONTAINMENT AREA TEMPERATURE:
rabi 2.4.6 - Emergency 500000 High CTMT Procedures/Plan: Knowledge 4.7 84 i Hydrogen Conc. / 5 of EOP m es 2.4.18 - Emergency 15 Incomplete SCRAM Procedures / Plan:
Knowledge of the specific 4.0 185 bases for EOPs.
EK1.01 - Knowledge of the operational implications of the following concepts as 295029 High Suppression iPool Water Level/5 x they apply to HIGH 3.4 59 SUPPRESSION POOL
\
WATER LEVEL:
Containment E~~.01 - Knowledge of the interrelations between HIGH
,......,,'"'v'",.... High Secondary
.X SECONDARY nment Area 3.5 60 CONTAINMENT AREA Temperature /5 TEMPERATURE and the followi : Area/room coolers AK3.01 - Knowledge of the reasons for the following 295015 Incomplete SCRAM responses as they apply to 1
x INCOMPLETE SCRAM:
3.4 61 Bypassing rod insertion blocks EA 1.08 - Ability to operate andlor monitor the following 295033 High Secondary as they apply to HIGH Containment Area SECONDARY i 3.6 \1. 62 Radiation Levels / 9 CONTAINMENT AREA RADIATION LEVELS:
Control Room ventilation
ES-401 Form ES-401-1 Pilgrim 7 Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 2 EAPE#/Name Safety Function KiA Topic(s)
AA2.01 Ability to determine and/or interpret the following as they apply to HIGH 295013 High Suppression SUPPRESSION POOL 3.8 63 Pool Temperature / 5 TEMPERATURE
- Suppression pool temperature 2.1 .20 - Conduct of 295009 Low Reactor Water Operations: Ability to 4.6 64 Level / 2 interpret and execute rocedure AA2.02 - Ability to determine and/or interpret the following 295022 Loss of CRD as they apply to LOSS OF 3.3 65 Pumps /1 CRD PUMPS: CRD system status KiA CategoryTotals Group Point Total: 7/3
ES-401 Form ES-401-1 Pilgrim 7 Written Examination Outline Plant Systems - Tier 2 Group 1 System #!Name KJA Topic(s) limp, I Q# I A2.08 - Ability to (a) predict the impacts of the following on the PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR 223002 STEAM SUPPLY PCIS/Nuclear SHUT-OFF; and (b)
Steam Supply 3.1 86 based on those Shutoff predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
Surveillance A2.02 - Ability to (a) predict the impacts of the following on the AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE 215005 APRM / MONITOR SYSTEM; LPRM and (b) based on 3.7 87 those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Upscale or downscale 2.2.42 - Equipment Control: Ability to 203000 RHR/LPCI: recognize system Injection Mode parameters that are 4.6 88 entry-level conditions for Technical ications.
2.1.20 - Conduct of 262002 UPS Operations: Ability to
'(AC/DC) 4.6 89 interpret and execute ure
ES-401 Form ES-401-1 Pilgrim 7 Written Examination Outline Plant Systems - Tier 2 Group 1 System #/Name KIA Topic(s) 259002 Reactor 2.1.19, Ability to use Water Level Control plant computers to 3.8 90 iSystem evaluate system or com ent status.
K1.03 - Knowledge of the physical connections and/or cause- effect 239002 SRVs X relationships between 3.5 1 RELIEF/SAFETY VALVES and the following: Nuclear boiler instrument m
K1.06 - Knowledge of the physical connections and/or cause- effect relationships between 2150031RM X INTERMEDIATE 3.9 2 RANGE MONITOR (I RM) SYSTEM and the following: APRM SCRAM signals:
Plant-S K2.02 - Knowledge of 215005 APRM / electrical power X supplies to the 2.6 3 LPRM following: APRM channels K2.01 - Knowledge of 215004 Source electrical power X supplies to the 2.6 4 Range Monitor following: SRM channels/detectors K3.02 - Knowledge of the effect that a loss or malfunction of the HIGH PRESSURE 206000 HPCI X COOLANT 3.8 5 INJECTION SYSTEM will have on following:
Reactor pressure control: BWR-2
ES-401 Form ES-401-1 Pilgrim 7 Written Examination Outline Plant Systems - Tier 2 Group 1 System #/Name KIA Topic(s) I Imp. I Q# I K3.01 - Knowledge of the effect that a loss or
. malfunction of the 205000 Shutdown SHUTDOWN Cooling COOLING SYSTEM 3.3 6 (RHR SHUTDOWN COOLING MODE) will have on following:
Reactor K4.06 - Knowledge REACTOR CORE ISOLATION COOLING SYSTEM (RCIC) 217000 RCIC X design feature(s) 3.5 7 and/or interlocks which provide for the following: Manual initiation K4.03 - Knowledge of (INSTRUMENT AIR SYSTEM) design 300000 Instrument feature(s) and or Air X interlocks which 2.8 8 provide for the following: Securing of lAS upon loss of cooli water K5.06 - Knowledge of the operational implications of the following concepts as 211000 SLC X they apply to 3.0 9 STANDBY LIQUID CONTROL SYSTEM:
Tank level measurement K3.02 - Knowledge of the effect that a loss or malfunction of the D.C.
263000 DC ELECTRICAL Electrical X DISTRIBUTION will 3.5 10 Distribution have on the following:
Components using DC .
control power (Le.
ES-401 Form ES-401-1 Pilgrim 7 Written Examination Outline Plant Systems - Tier 2 Group 1 System #/Name KIA Topic(s)
K6.08 - Knowledge of the effect that a loss or malfunction of the following will have on 261000 SGTS x the STANDBY GAS 3.1 11 TREATMENT SYSTEM: Reactor vessel level: Plant ic K6.08 - Knowledge of the effect that a loss or malfunction of the following will have on PCIS/Nuclear the PRIMARY Steam Supply CONTAINMENT 3.5 12 Shutoff ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF: Reactor A 1.08 - Ability to predict and/or monitor changes in parameters associated with 203000 RHR/LPCI: operating the Injection Mode RHA/LPCI: 3.7 13 INJECTION MODE (PLANT SPECIFIC) controls including:
Emergency generator loadi A1.04 - Ability to predict and/or monitor changes in parameters associated with 209001 LPCS operating the LOW 3.7 14 I PRESSURE CORE SPRAY SYSTEM
ES-401 Form ES-401-1 Pilgrim 7 Written Examination Outline Plant Systems - Tier 2 Group 1 System #/Name KIA Topic(s} limp, I Q# I K6.01 - Knowledge of the effect that a loss or malfunction of the 218000 ADS following will have on X 3.9 15 the AUTOMATIC DEPRESSURIZATION SYSTEM: RHR/LPCI A2.03 - Ability to (a) predict the impacts of the following on the CCWS and (b) based on those predictions, 400000 Component use procedures to Cooling Water 2.9 16 correct, control, or mitigate the consequences of those abnormal operation: High/low CCW re A3.03 - Ability to monitor automatic 262001 AC operations of the A.C.
Electrical ELECTRICAL 3.4 17 Distribution DISTRIBUTION including: Load sheddi A3.03 - Ability to monitor automatic operations of the EMERGENCY 264000 EDGs GENERATORS 3.4 18 (DIESEUJET) including: Indicating lights, meters, and recorders A4.01 - Ability to manually operate 262002 UPS and/or monitor in the (AC/DC) 2.8 i 19 control room: Transfer
'from alternative source!
to
ES-401 Form ES-401-1 Pilgrim 7 Written Examination Outline Plant Systems - Tier 2 Group 1 System #/Name KIA TopiC(S)
A4.04 - Ability to 259002 Reactor manually operate Water Level Control and/or monitor in the 3.7 20 control room: FWRV lockout reset controls 2.4.31 - Emergency Procedures / Plan:
212000 RPS Knowledge of 4.2 21 annunciator alarms, indications, or 2.2.40 - Equipment 206000 HPCI Control: Ability to apply 3.4 22 technical specifications for a A 1.08 - Ability to predict and/or monitor changes in parameters associated with 212000 RPS operating the 3.4 23 REACTOR PROTECTION SYSTEM controls including: Valve ition A3.02 - Ability to monitor automatic 209001 LPCS operations of the LOW 3.8 24 PRESSURE CORE SPRAY SYSTEM includ : Pu start K3.01 - Knowledge of the effect that a loss or malfunction of the AUTOMATIC DEPRESSURIZATION 218000 ADS x SYSTEM will have on 4.4 25 following: Restoration of reactor water level after a break that does .
not depressurize the reactor when uired
ES-401 Form ES-401-1 Pilgrim 7 Written Examination Outline Plant Systems - Tier 2 Group 1 System #/Name KIA Topic(s)
K6.01 - Knowledge of the effect that a loss or malfunction of the following will have on 217000 RCIC x the REACTOR CORE 3.4 26 ISOLATION COOLING SYSTEM (RCIC):
Electrical KIA Category Totals 2 2 421 4 3 2615
ES-401 Form ES-401-1 Pilgrim 7 Written Examination Outline Plant Systems - Tier 2 Group 2 System #/Name KIA Topic(s)
A2.02 - Ability to (a) predict the impacts of the following on the ROD POSITION INFORMATION SYSTEM; and (b) based 214000 RPIS on those predictions, use 3.7 91 procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Reactor SCRAM 2.4.9 - Emergency Procedures / Plan:
Knowledge of low power 233000 Fuel Pool / shutdown implications
.Cooling/Cleanup 4.2 92 I
in accident (e.g., loss of coolant accident or loss of residual heat removal) m A2.01 - Ability to (a) predict the impacts of the following on the FUEL HANDLING EQUIPMENT; and (b) 234000 Fuel based on those Handling Equipment predictions, use 3.7 93 procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Interlock failure K1.03 - Knowledge of the physical connections and/or cause- effect relationships between 215002 RBM X ROD BLOCK MONITOR 3.2 27 SYSTEM and the following: Reactor manual control: BWR
ES-401 Form ES-401-1 Pilgrim 7 Written Examination Outline Plant Systems - Tier 2 Group 2 System #lName KIA Topic(s)
A3.03, Ability to monitor automatic operations of 202002 the RECIRCULATION Recirculation Flow FLOW CONTROL 3.1 28 Control SYSTEM including:
Scoop tube operation:
BW 4 K3.06 - Knowledge of the effect that a loss or 259001 Reactor malfunction of the Feedwater REACTOR 3.1 29 FEEDWATER SYSTEM will have on following:
Core inlet subcooli K4.09 - Knowledge of NUCLEAR BOILER INSTRUMENTATION 216000 Nuclear design feature(s) and/or Boiler Inst. X interlocks which provide 3.3 30 for the following:
Protection against filling the main steam lines from the feed K5.10 - Knowledge of ROD WORTH MINIMIZER SYSTEM (RWM) (PLANT 201006 RWM X SPECIFIC) design 3.2 31 feature(s) and/or interlocks which provide for the following:
Withdraw error: P BW K6.08 - Knowledge of the effect that a loss or malfunction of the RWCU X following will have on the 3.5 32 REACTOR WATER CLEANUP SYSTEM:
PCIS/NSSSS
ES-401 Form ES-401-1 Pilgrim 7 Written Examination Outline Plant Systems - Tier 2 Group 2 System #/Name KIA Topic(s)
A1.01 - Ability to predict and/or monitor changes in parameters associated with operating the 272000 Radiation RADIATION Monitoring 3.2 33 MONITORING SYSTEM controls including:
Lights, alarms, and indications associated with normal ,."'..,....." *. ,1"1 A2.04 - Ability to (a) predict the impacts of the following on the PLANT VENTI LATIOI\I SYSTEMS ; and (b) 288000 Plant based on those Ventilation predictions, use 3.7 34 procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: High radiation: Plant-~ ....,,,,,,.. ,,..
K2.01, Knowledge of Main and electrical power supplies
- Reheat Steam x to the following: Main 3.2 35 steam isolation valve solenoids A4.02 - Ability to 245000 Main manually operate and/or Turbine Gen. / Aux. 3.1 36 monitor in the control room: Generator controls i 2.2.25 - Equipment Control: Knowledge of 001 CRD bases in technical
'Hydraulic 3.2 37 specifications for limiting conditions for operations and limits.
K4.01 - Knowledge of CONTROL ROOM HVAC design feature(s)
IL::7'UU\.I.::J Control and/or interlocks which iRoom HVAC 3.1 38 provide for the following:
System initiations/reconfiguration ific
ES-401 Form ES-401-1 Pilgrim 7 Written Examination Outline Plant Systems - Tier 2 Group 2 System #/Name KJA Topic(s) limp, I I Q#
KJA Category Totals Group Point Total:
ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Facility: Pilgrim 7 Date: i Category KA# Topic RO SRO-Only :
IR Q# IR Q# I Knowledge of operator responsibilities 2.1.2 4.1 66 I durin.-9 all modes of plant operation.
Knowledge of shift or short-term relief I 2.1.3 3.7 67 turnover practices.
Ability to explain and apply system 2.1.32 3.8 70 limits and precautions.
- 1. Conduct of Operations I 2.1.23 Ability to perform specific system and integrated plant procedures during all 4.4 94 I :modes of plant operation. !
i Subtotal "' 3 """ .. 1 Ability to interpret control room indications to verify the status and 2.2.44 operation of a system, and understand 4.2 68 I
how operator actions and directives I
affect plant and system conditions.
2.2.12 Knowledge of surveillance procedures. 3.7 69
- 2. Equipment Knowledge of maintenance work order 2.2.19 3.4 95 Control requirements.
Ability to perform pre-startup procedures for the facility, including 2.2.1 operating those controls associated 4.4 99 with plant equipment that could affect reactivity.
I ' ..
I Subtotal ; 2 " ".". 2
ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 2.3.11 Ability to control radiation releases. 3.8 71 Knowledge of Radiological Safety Principles pertaining to licensed operator duties, such as containment 2.3.12 3.2 74 entry requirements. fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.
- 3. Radiation 2.3.6 Ability to approve release permits. 3.8 96 Control Ability to use radiation monitoring I systems, such as fixed radiation 2.3.5 monitors and alarms, portable survey 2.9 98 instruments, personnel monitoring equipment, etc.
Subtotal 'i'i*..*'*.. 2 E 2 Knowledge of EOP entry conditions and 2.4.1 4.6 72 immediate action steps.
2.4.29 Knowledge of the emergency plan. 3.1 73 Ability to perform without reference to procedures those actions that require 2.4.49 4.6 75 immediate operation of system components and controls.
- 4. Emergency Procedures / Knowledge of abnormal condition 2.4.11 4.2 97 Plan procedures.
Ability to perform without reference to procedures those actions that require 2.4.49 4.4 100 immediate operation of system components and controls.
Subtotal 3 ", 2 Tier 3 Point Total: 1\ ... , 10 I i( 7
ES-401 Record of Rejected KIA's Form ES-401-4 Randomly Selected Tier / Group Reason for Rejection KA SRO (#76) - Ability to determine and/or interpret the following as they apply to REACTOR LOW WATER LEVEL: Reactor pressure. Originally proposed KIA would present a double 295031 / EA2.03 jeopardy situation for an SRO with RO question #53 (same KIA).
1 /1 replaced by 295031 I EA2.01 Randomly selected EA2.01, Ability to determine and/or interpret the following as they apply to REACTOR LOW WATER LEVEL:
Reactor water level (RO #42) - AK2.02 - Knowledge of the interrelations between 295006/ AK2.02 SCRAM and the following: Reactor water level control system.
1 /1 replaced by 295006 / Similar concept to #40 AK2.06 Randomlv selected AK2.06 - Reactor power RO (#43) - Knowledge of the interrelations between CONTROL ROOM ABANDONMENT and the following: Control room HVAC.
There are currently six HVAC questions on the NRC written, this 295016 / AK2.03 one is difficult to find a procedural reference for writing a 1 /1 replaced by 295016 / question.
AK2.02 Randomly selected AK2.02, Knowledge of the interrelations between CONTROL ROOM ABANDONMENT and the following:
Local control stations: Plant-Specific RO (#46) - Knowledge of the reasons for the following responses as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE: Suppression pool spray: Plant-Specific.
Pilgrim's mitigation strategy for high suppression pool 295026 / EK3.03 temperature does not use suppression chamber spray nor is it 1 /1 replaced by 295026 /
impacted by suppression chamber sprays.
EK3.05 Randomly selected EK3.05, Knowledge of the reasons for the following responses as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE: Reactor SCRAM RO (#48) - Ability to operate and/or monitor the following as they apply to REFUELING ACCIDENTS: Fuel transfer system: Plant-Specific. Pilgrim design does not include a Fuel Transfer 295023/ AA1.05 System.
1 /1 replaced by 295023 /
AA1.04 Random Iy selected AA 1.04, Ability to operate and/or monitor the following as they apply to REFUELING ACCIDENTS: Radiation monitoring equipment.
RO (#49) - Ability to operate and / or monitor the following as they apply to PLANT FIRE ON SITE: Respirator air pack. At Pilgrim, the ability to operate or monitor an air pack would not 600000/ AA1.01 discriminate as this ability is first obtained as a non licensed 1 /1 replaced by 600000 /
operator.
AA1.06 Randomly selected AA1.06, Ability to operate and lor monitor i i the following as they apply to PLANT FIRE ON SITE: Fire alarm I
ES-401 Record of Rejected KiA's Form ES-401-4 (RO #56) - Equipment Control: Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations. Assessing 295004/2.2.36 maintenance activities is an SRO function at Pilgrim.
1 11 replaced by 2950041 AK3.03 Randomly selected AK3.03 - Knowledge of the reasons for the following responses as they apply to PARTIAL OR COMPLETE LOSS OF D.C. POWER: Reactor scram (RO #58) - Conduct of Operations: Ability to locate control room switches, controls, and indications, and to determine that they correctly reflect the desired plant lineup. This is being tested in 295001 /2.1.31 the operating test portion of the exam.
1 11 replaced by 295001 12.2.37 Randomly selected 2.2.37 - Equipment Control: Ability to determine operability and/or availability of safety related equipment.
(RO #62) - EA 1.07 - Ability to operate and/or monitor the 2950331 EA 1.07 following as they apply to HIGH SECONDA~Y CONTAINMENT 1/2 replaced by 295033 1 AR~A RA~IA~I~N LEVELS: :ersonnel dosimetry. GET level EA1.08 tOPIC, not discriminatory for a license exam.
Randomly selected EA1.08 - Control Room ventilation (RO #63) - EA2.02 - Ability to determine and/or interpret the following as they apply to SECONDARY CONTAINMENT VENTILATION HIGH RADIATION: Cause of high radiation levels. Similar in concept to #51 and randomly resampled #62.
295034 1 EA2.02 Also #34 was similar 1/2 replaced by 295013/
AA2.01 Randomly selected 295013 AA2.01- Ability to determine and/or
.interpret the following as they apply to HIGH SUPPRESSION IPOOL TEMPERATURE :Suppression pool temperature (SRO #84) - 2.4.8 - Emergency Procedures / Plan:
Knowledge of how abnormal operating procedures are 500000/ 2.4.8 used in conjunction with EOP's. There are no abnormal 1/2 replaced by 500000 / procedures for this EAPE at Pilgrim 2.4.6 Randomly selected 2.4.6 - Emergency Procedures/Plan:
Knowledge of EOP mitigation strategies SRO (#87) - Ability to (a) predict the impacts of the following on the AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions Power supply degraded. Originally proposed KIA would present a double 215005/ A2.01 jeopardy situation for an SRO with RO question #3 (same 2/1 replaced by 215005/ system, same topic).
A2.02 Randomly selected A2.01, Ability to (a) predict the impacts of the following on the AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
I Upscale or downscale trips
ES-401 Record of Rejected KIA's Form ES-401-4 SRO (#90) 2.1.27 - Conduct of Operations: Knowledge of system purpose and / or function. Difficult to write an operationally valid 261000/2.1.27 SRO question on the purpose and/or function of SGTS.
2/1 replaced by 259002 /
2.1.19 Randomly replaced with 259002, Reactor Water Level Control i
System, 2.1.19, Ability to use plant computers to evaluate system or component status.
RO (#10) K5.01 - Knowledge of the operational implications of the following concepts as they apply to D.C. ELECTRICAL DISTRIBUTION: Hydrogen generation during battery charging.
263000 / K5.01 Could not find a suitable reference for this topic.
2/1 replaced by 263000 /
K3.02 Randomly replaced with K3.02, Knowledge of the effect that a I loss or malfunction of the D.C. ELECTRICAL DISTRIBUTION will have on the following: Components using DC control power (Le.
breakers)
RO (#15) - A2.01 - Ability to (a) predict the impacts of the following on the AUTOMATIC DEPRESSURIZATION SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal 218000/ K2.01 conditions or operations: Small steam line break LOCA.
2/1 replaced by K6.01 Oversample with concept in #25 Randomly selected K6.01 Knowledge of the effect that a loss or malfunction of the following will have on the AUTOMATIC DEPRESSURIZATION SYSTEM: RHR/LPCI system pressure RO (#20) A4.1 0 - Ability to manually operate and/or monitor in the control room: Setpoint setdown reset controls: Plant 259002 / A4.1 0 Specific. Pilgrim's Reactor Water Level Control System does not' 2/1 replaced by 259002 / include a Setpoint Setdowi1 feature.
A4.04 Randomly selected A4.04, Ability to manually operate and/or monitor in the control room: FWRV lockup reset controls SRO (#93) A2.02 - Ability to (a) predict the impacts of the following on the CONTROL ROOM HVAC; and (b) based on those predictions, use procedures to correct, control, or mitjgate the consequences of thOSE! abnormal conditions or operations:
Extreme environmental conditions. There are two tier 2, group 2 questions on CONTROL ROOM HVAC, (93 and 38, and another plant ventilation question #34). Procedure No. 2.1.42, 290003 / A2.02 OPERATION DURING SEVERE WEATHER has very little on 2/2 replaced by 234000 /
ventilation during severe weather. Deleting this KIA for better A2.01 coverage and a more appropriate SRO question.
Randomly replaced with A2.01 - Ability to (a) predict the impacts of the following on the FUEL HANDLING EQUIPMENT; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or I operations: Interlock failure.
ES-401 Record of Rejected KIA's Form ES-401-4 RO (#27) - Knowledge of the physical connections and/or cause-effect relationships between ROD BLOCK MONITOR SYSTEM and the following: Recirculation system: BWR-3,4,5. Following system modification there is no longer a cause - effect 215002/ K1.04 relationship between the ROD BLOCK MONITOR SYSTEM and 2/2 replaced by 215002 / the Recirculation system.
K1.03 Randomly selected K1.03, Knowledge of the physical connections and/or cause- effect relationships between ROD i BLOCK MONITOR SYSTEM and the following: Reactor manual control: BWR-3,4,5 RO (#28) - Knowledge of electrical power supplies to the following: Hydraulic power unit: Plant-Specific. Pilgrim does not utilize hydraulic power units as part of its Recirc Flow Control 202002 / K2.02 System.
2/2 replaced by 202002 /
A3.03 I Randomly selected A3.03, Ability to monitor automatic operations of the RECIRCULATION FLOW CONTROL SYSTEM including: Scoop tube operation: BWR-2,3,4 RO (#35) - Ability to monitor automatic operations of the MAIN AND REHEAT STEAM SYSTEM including: Opening and closing of drain valves as turbine load changes: Plant-Specific. Drain 239001 I A3.02 Valves associated with Pilgrim's Main Steam System do not 2/2 replaced by 239001 /
automatically respond to changes in Turbine Load.
K2.01 Randomly selected K2.01, Knowledge of electrical power supplies to the following: iVlain steam isolation valve solenoids RO (# 67) - Ability to manage the control room crew during plant transients. This is an SRO function at Pilgrim.
G1 /2.1.6 replaced Tier 3 by 2.1.3 Randomly selected 2.1.3 - Knowledge of shift or short-term relief !
turnoverpractices.
RO (#70) - Ability to comply with radiation work permit requirements during normal or abnormal conditions. This is being tested on the admin JPM portion of the exam.
G2.3.7 replaced by Tier 3 2.1.32 Randomly selected 2.1.28 - Knowledge of the purpose and function of major system components and controls. A discriminating question could not be written for system purpose I function. Reselected 2.1.32 RO (#75) - Knowledge of system set pOints, interlocks and automatic actions associated with EOP entry conditions. Same concept as #72.
G2.4.2 replaced by Tier 3 2.4.49 Randomly selected 2.4.49 - Ability to perform without reference to procedures those actions that require immediate operation of I system components and controls.
I i
i
ES-401 Record of Rejected KIA's Form ES-401-4 I
i I
ES-301 Administrative Topics Outline Form ES-301-1 Facility: PNPS NRC Date of Examination: 1/2011 Examination Level (circle one): RO / SRO Operating Test Number:
Administrative Topic Type I Describe activity to be performed (see Note) Code* i
! Short Form Heat Balance Conduct of Operations 1 P The candidate will perform a Short Form Heat Balance lAW PNPS 2.1.10 Att.4.
KIA: 2.1.7 (4.4)
Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument I interpretation.
Conduct of Operations 2 N Verification of License Requirements Given information related to maintenance of active license status for three operators, the candidate will determine which operator(s), if any, is (are) qualified to relieve the watch.
(the candidate will be given the status of 3 operators in regard to last medical exam, hours worked in last quarter, SCBA fit test latest date etc. AND given the procedure that describes the requirements, determine if anyone meets eligibility requirements)
KIA: 2.1.4 (3.3)
Knowledge of individual licensed operator responsibilities related to shift staffing, such as medical requirements, "no solo" operation, maintenance of active license status, 10CFR55, etc.
Identify the isolations required to tagout "E" RBCCW pump Equipment Control N for the shaft seal replacement.
The candidate will determine blocking pOints, tag types, and component position for a tagout on the "E" RBCCW pump.
KJA: 2.2.13 (4.1)
Knowledge of tagging and clearance procedures NUREG-1 021, Revision 9
ES-301 Administrative Topics Outline Form ES-301-1 Determine main stack release rates and subsequent Radiation Control - N actions The operator will determine that low main stack dilution f10wrates exist and high main stack release rates require an immediate power reduction.
KIA 2.3.11 (3.8)
Ability to control radiation releases.
Emergency Plan N/A NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when 5 are required.
{N)ew or (M)odified from bank (> 1)
(P)revious 2 exams (s 1; randomly selected)
(S)imulator NUREG-1021, Revision 9
ES-301 Administrative Topics Outlin_e_______F_o_rm_E_S_-_3_0_1-_1_
Facility: PNPS NRC Date of Examination: 1/2011 Examination Level (circle one): RO I SRO Operating Test Number:
Administrative Topic Type Code* Describe activity to be performed (see Note)
Perform & Review a Short Form Heat Balance Comparison Conduct of Operations P The candidate will perform a short form heat balance and determine any appropriate actions lAW PNPS 2.1.10 KIA: 2.1.7 (4.4)
Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior and instrument interpretation Review a portion of the Control Room Daily Logs Conduct of Operations N The candidate will rEiview a completed portion of the control room logs and identify OOS items and TS implications KIA: 2.1.18 (3.8)
Ability to make accurate, clear, and concise logs, records, status boards, and reports.
Analyze a Solomon case from 3D Monicore and determine Equipment Control N the appropriate action.
Following a dual Recirculation Pump runback the candidate will review a Solomon Case and determines that the Hot Channel Decay Ratio is unsat, then determine power must be lowered using the RPR array instruction sheet.
KIA: 2.2.38 (4.5)
Knowledge of conditions and limitations in the facility license.
Form -1 Determine the actions required when both channels of the Radiation Control I D Reactor Building Effluent Monitoring System become inoperable the ODCM With Reactor Building Effluent Monitoring System "A" RM 1705-32A out of service the control room will must determine the ODCM requirements when the "B" monitor becomes inoperable. This includes that grab samples are taken, that auxiliary sampling equipment is operable and flow rates are estimated.
KJA: 2.3.11 (4.3)
Ability to control radiation releases.
Emergency Classification Emergency Plan N The candidate will classify the event following performance in Scenario #2 OR #3 KJA: 2.4.29 (4.4)
Knowledge of the emergency plan NOTE: All items (5 total are required for SROs). RO applicants require only 4 items unless they are retaking only the administrative topics, when 5 are required.
(N)ew or (M)odified from bank (> 1)
(P)revious 2 exams 1; randomly selected)
(S)imulator
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: PILGRIM NRC Date of Examination: 2/2011 Exam Level (circle one): SRO(I) / SRO (U) Operating Test No:
Control Room Systems@ (8 for RO; 7 for SRO-I; 2 or 3 for SRO-U, including 1 ESF)
Type Code* Safety Function System / JPM Title S-1 Control Rod Exercising lAW 8.3.2 M, A, S 1 The reactor is at power. The weekly control rod Reactivity exercising in accordance with procedure 8.3.2 is Control required. When a coupling check is performed on a rod being withdrawn, the rod will go into an overtravel condition. The operator is expected to recouple the rod per off-normal procedure 2.4.11. The JPM will end when the rod is recoupled.
PNPS 8.3.2, 2.4.11 KIA 201002 A3.03 3.2/3.2 S-2 HPCI Swap-Over from Pressure Control to Injection M, L, A, EN, 2 S Reactor Water HPCI is operating in pressure control mode and must be Inventory swapped to injection mode. When HPCI is placed in Control injection mode and the candidate attempts to raise injection flow the HPCI Flow Controller FIC-2340-1 fails high, the operator must place the controller in manual to raise flow.
PNPS 2.2.21.5, Attachments 1 and 2 KIA 206000 A4.02 4.0/3.8 S-3 Re-Open MSIV's Following Closure D,S 3 Reactor An operator is directed to perform Reactor Manual Pressure Scram Test, Pf\lPS 8.M.1-23. The operator will start the Control test however when the channel B manual scram is inserted three control rods will drift into the core requiring the operator to manually scram the reactor.
KIA 239001 A2.03 4.0/4.2 NUREG-1021, Revision 9
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 S-4 S~nchronize Main TG to Grid D,S 4 Heat Removal A plant startup is progress. The Turbine Generator is From the Core ready to be synchronized to the grid. The TG will be I synched to the grid, the Turbine Bypass Valves closed.
KIA 245000 A4.09 3.1/2.9 S-5 Manuall~ Start SBGT and Vent the Torus D,A,S 9 Radioactivity The operator will align standby gas to vent the torus. Release After establishing the lineup, a reactor coolant pressure boundary leak develops in the drywell. The operator will secure the standby gas vent alignment lAW Section 7.10 of 2.2.70.
I KIA 261000 A4.09 2.7/2.7 S-6 B:iQass Diesel Generator Load Shed for Qlacing ,1! D, EN,S 6 CRD PumQ in Service Electrical A Reactor Scram has occurred due to a loss of offsite power and a small leak in containment has led to dies.el load shed. A Reactor low level condition requires placing two CRD pumps in emergency makeup. The candidate must assess Emergency Diesel Generator loading and then defeat the CRD Load shed logic.
I KIA 264000 K4.05 3.2/3.5 S-7 Perform Reactor Manual Scram Surveillance Test N,A,EN,S 7 PNPS 8.M.1-23. Instrumentation The operator is required to coordinate the transfer of RPS "A" to its alternate power supply. Following the transfer RPS will not reset due to APRM "C" failing upscale when re-energized. The operator will be required to diagnose the failure, bypass the APRM and continue on with resetting the RPS and other control I room instrumentation affected by the transfer.
PNPS B.M.1-23, 2.4.11 KIA 212000 A2.03 3.3/3.5 NUREG-1021, Revision 9
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2
- S-8 Isolate a Condenser Waterbox during Chloride D,S I 8 intrusion Plant Service Systems The operator will isolate Water Box 1-3 due to chloride intrusion lAW PNPS 2.4.33 Att.3.
Ii KIA 256000 A2.15, 2.8/3.1 i
! I.
I NUREG-1021 , Revision 9
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 In-Plant Systems@ (3 for RO; 3 for SRO-I; 3 or 2 for SRO-U)
P-1 OeQressurize Scram Volume Pressure Header O,E,A,R 1 Reactivity With the reactor having received a reactor SCRAM all Control rods did not insert due to an electrical malfunction in the RPS circuit. The control room has given the order to depressurize the SPVAH in the field per 5.3.23.
(preferred method will not work due to stuck valve.)
I KIA 295037 2.1.30 4.4/4.0 P-2 Install Backul2 N2 for Extended SRV Ol2eration 0, L, R 3 Reactor Following a seismic event with a subsequent loss of Pressure N2/air supply to the drywell. the Emergency Director Control requires backup N2 supplied to 'B' and 'C' SRVs for continued reactor pressure control.
I KIA 218000 A2.03 3.4/3.6 P-3 LineuQ Alternate Power to RHR Valves 0, L, R 6 Electrical During a refueling outage with shutdown cooling in service a loss of 480 Volt bus B20 has occurred, resulting in a loss of power to selected RHR valves. The operator will align alternate power to those RHR valves fed from B20 and which have failed as is.}
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KIA 295003 M1.01 3.7.3.8
@ All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.
- Type Codes Criteria for RO / SRO-I / SRO-U (A)lternate path 4-6 / 4-6 / 2-3 (C)ontrol room (D)irect from bank 9/..s 8 / 4 (E)mergency or abnormal in-plant 1/21/21 (EN)gineering Safeguards Feature - / - / 1 (control room)
(L}ow-Power I Shutdown 2 1 / 1/21 (N)ew or (M)odified from bank including 1(A) 22/22/21
{P)revious 2 exams ..s 3/,,::; 3 /..s 2 (randomly selected)
(R)CA 1/21/21 i (S)imulator NUREG-1021, Revision 9
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2
- Mark10 PILGRIM NRC Date of Examination: 2/2011 Facility:
Exam Level (circle one): RO (§@]) / SRO (U) Operating Test No:
Control Room Systems@ (8 for RO; 7 for SRO-I; 2 or 3 for SRO-U, including 1 ESF)
Type Code* Safety Function System / JPM Title S-1 Control Rod Exercising lAW 8.3.2 M, A, S 1 The reactor is at power. The weekly control rod Reactivity exercising in accordance with procedure 8.3.2 is Control required. When a coupling check is performed on a rod being withdrawn, the rod will go into an overtravel condition. The operator is expected to recouple the rod per off-normal procedure 2.4.11. The JPM will end when the rod is recoupled.
PNPS 8.3.2, 2.4.11 KIA 201002 A3.03 3.2/3.2 S-2 HPCI Swal2-0ver from Pressure Control to Injection M, L, A, EN, 2 S Reactor Water HPCI is operating in pressure control mode and must be Inventory swapped to injection mode. When HPCI is placed in Control injection mode and the candidate attempts to raise injection flow the HPCI Flow Controller FIC-2340-1 fails high, the operator must place the controller in manual to raise flow.
I!
PNPS 2.2.21.5, Attachments 1 and 2 KIA 206000 A4.02 4.0/3.8 S-3 Re-Ol2en MSIV's Following Closure D,S 3 Reactor The operator is required to reopen the outboard and Pressure inboard "0" MSIVs following MSIV closure lAW PNPS Control 2.2.92.
Ii KIA 239001 A2.03 4.0/4.2 NLiREG-1021, Revision 9
Control Room/In-Plant Systems Outline
- S-5 Manually Start SBGT and Vent the Torus D,A,S I 9 Radioactivity The operator will align standby gas to vent the torus. Release After establishing the lineup, a reactor coolant pressure boundary leak develops in the drywell. The operator will secure the standby gas vent alignment lAW Section ,'.10 of 2.2.70.
I KIA 261000 A4.09 2.7/2.7 S-6 BYQass Diesel Generator Load Shed for I2lacing a D,EN,S 6 CRD Pump in Service Electrical A Reactor Scram has occurred due to a loss of offsitS' power and a small leak in containment has led to diesel load shed. A Reactor low level condition requires placing two CRD pumps in emergency makeup. The candidate must assess Emergency Diesel Generator loading and then defeat the CRD Load shed logic.
I KIA 264000 K4.05 3.2/3.5 S-7 Perform Reactor Manual Scram Surveillance Test N,A,EN,S 7 PNPS 8.M.1-23. Instrumentation The operator is required to coordinate the transfer of RPS "A" to its alternate power supply. Following the transfer RPS will not reset due to APRM "C" failing upscale when re-energized. The operator will be required to diagnose the failure, bypass the APRM and continue on with resetting the RPS and other control room instrumentation affected by the transfer.
PNPS 8.M.1-23, 2.4.11 KIA 212000 A2.03 3.3/3.5 I
. S-8 Isolate a Condenser Waterbox during Chloride D,S 8 intrusion I Plant Service Systems The operator will isolate Water Box 1-3 due to chloride intrusion lAW PNPS 2.4.33 Att.3.
I KIA 256000 A2.15, 2.8/3.1 NUREG-1021 , Revision 9
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 In-Plant Systems@ (3 for RO; 3 for SRO-I; 3 or 2 for SRO-U)
P-1 Del2ressurize Scram Volume Pressure Header D,E,A,R 1 Reactivity With the reactor having received a reactor SCRAM all Control rods did not insert due to an electrical malfunction in the RPS circuit. The control room has given the order to depressurize the SPVAH in the field per 5.3.23.
(preferred method will not work due to stuck valve.)
I:
KIA 295037 2.1.30 4.4/4.0 P-2 Install Backul2 N2 for Extended SRV Ol2eration D,L,R 3 Reactor Following a seismic event with a subsequent loss of Pressure N2/air supply to the drywell, the Emergency Director Control requires backup N2 supplied to 'B' and 'C' SRVs for continued reactor pressure control.
I KIA 218000 A2.03 3.4/3.6
. P-3 Lineul2 Alternate Power to RHR Valves D,L,R 6 Electrical During a refueling outage with shutdown cooling in service a loss of 480 Volt bus B20 has occurred, resulting in a loss of power to selected RHR valves. The operator will align alternate power to those RHR valves fed from B20 and which have failed as is.)
I KIA 295003 AA1.01 3.7.3.8
@ All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.
- Type Codes Criteria for RO / SRO-I / SRO-U (A)lternate path 4-6/4-6/2-3 (C}ontrol room (D)irect from bank 9/<;8/<;4 (E)mergency or abnormal in-plant 1/~1/~1 (EN}gineering Safeguards Feature - /-/ 1 (control room)
(L)ow-Power / Shutdown ~ 1/ 1/~ 1 (N)ew or (M)odified from bank including 1(A) ~ 2/ 2/ ~ 1 (P)revious 2 exams <;3/<;3/ 2 (randomly selected)
(R)CA 1/~1/~1 (S)imulator NUREG-1021, Revision 9
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: PILGRIM NRC Date of Examination: 2/2011 Exam Level (circle one): RO / SRO(l) ~RO (~ Operating Test No: 1 Control Room Systems@ (8 for RO; 7 for SRO-I; 2 or 3 for SRO-U, including 1 ESF)
System / JPM Title Type Code* Tsafet y Function S-2 HPCI SwaQ-Over from Pressure Control to Injection M, L, A. EN. 2 S Reactor Water HPCI is operating in pressure control mode and must be Inventory swapped to injection mode. When HPCI is placed in Control injection mode and the candidate attempts to raise injection flow the HPCI Flow Controller FIC-2340-1 fails high. the operator must place the controller in manual to raise flow.
I PNPS 2.2.21.5. Attachments 1 and 2 KIA 206000 A4.02 4.0/3.8 S-7 Perform Reactor Manual Scram Surveillance Test !\I.A.EN,S 7 PNPS 8.M.1-23. Instrumentation The operator is required to coordinate the transfer of RPS "A" to its alternate power supply. Following the transfer RPS will not reset due to APRM "C" failing upscale when re-energized. The operator will be required to diagnose the failure. bypass the APRM and continue on with resetting the RPS and other control room instrumentation affected by the transfer.
PNPS 8.M.1-23, 2.4.11 KIA 212000 A2.03 3.3/3.5 S-8 Isolate a Condenser Waterbox during Chloride D,S 8 intrusion Plant Service Systems The operator will isolate Water Box 1-3 due to chloride intrusion lAW PNPS 2.4.33 Att.S.
I KIA 256000 A2.15, 2.8/S.1 NUREG-1021 , Revision 9
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 In-Plant Systems@ (3 for RO; 3 for SRO-I; 3 or 2 for SRO-U)
P-1 De~ressurize Scram Volume Pressure Header D,E,A,R 1 Reactivity With the reactor having received a reactor SCRAM all Control rods did not insert due to an electrical malfunction in the RPS circuit. The control room has given the order to depressurize the SPVAH in the field per 5.3.23.
(preferred method will not work due to stuck valve.)
I KIA 295037 2.1.30 4.4/4.0 P-3 Lineu~ Alternate Power to RHR Valves D, L, R 6 Electrical During a refueling outage with shutdown cooling in service a loss of 480 Volt bus 820 has occurred, resulting in a loss of power to selected RHR valves. The operator will align alternate power to those RHR valves fed from 820 and which have failed as is.}
I KIA 295003 AA1.01 3.7.3.8
@ All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.
- Type Codes Criteria for RO I SRO-II SRO-U (A) Iternate path 4-6/4-6/2-3 (C}ontrol room (D)irect from bank ::;9/::;8/..:;.4 (E)mergency or abnormal in-plant 1/:;:>1/::::1 (EN)gineering Safeguards Feature I:::: 1 (control room)
(L}ow-Power / Shutdown ::::1/::::1/ 1 (N)ew or (M)odified from bank including 1(A) :;:> 2/ 2/:;:> 1 (P}revious 2 exams :s; 3/ 3/ :s; 2 (randomly selected)
(R}CA :::: 1/ 1 I:::: 1
. (S)imulator NUREG-1 021, Revision 9
Scenario Event Description ES-D1 Pilgrim 2011 NRC Scenario 2 Facility: PILGRIM Scenario No.: 2 Op Test No.: 2011 NRC Examiners: Operators: SRO RO BOP Initial Conditions:
- Reactor Power: 90%
- Plant Status: Reactor power was reduced to 90% last shift for rod pattern adjustments.
- Core flow is 49 Mlbm/hr
- Current Rod Position: Sequence A 1, Step 87, rod 18-43
- Equipment Out of Service: uA" APRM has a faulty power supply and is OOS and bypassed. Tracking LCO initiated. All other APAMs are operable.
Turnover:
- Secure torus cooling and restore power to 100% lAW PNPS 2.1.14, Station Power Changes.
Critical Tasks: 1. During failure to scram conditions terminate and prevent injection from all sources (except CAD, RCIC, and SBLC) and lower level to prevent oscillations.
- 2. Inject SBLC before torus water temperature exceeds the BIIT or in response to core oscillations.
- 3. During failure to scram conditions, insert control rods using one or more methods contained within 5.3.23 and / or EOP-02 to achieve Rx. Shutdown under all conditions Event Malt. No. Event Event Description No. Type*
- 1. N/A N-BOP Secure Torus Cooling N-BOP
- 2. RD08 TS-SRO A control rod accumulator Trouble alarm is received. Local efforts to recharge the accumulator are ineffective. The SRO is expected to declare the accumulator inoperable and determine that the associated Control Rod should be declared "slow" or inoperable within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. TS 3.3.D.A.l
- 3. CW06 C-BOP "F" ABCCW pump trip. Loop uB" RBCCW becomes inoperable.
C-SRO TS 3.S.B.3 TS-SRO
- 4. AR13 & C-AII "B" Recirc Seal Failures (Both). Requires tripping and isolation of AR14. pump per PNPS 2.4.22 and 2.4.17
-1
Scenario Event Description ES-D1 Pilgrim 2011 NRC Scenario 2
- 5. N/A R-RO Inserts control rods to exit the Unanalyzed and Exclusion Regions R-SRO of the Power/Flow Map following the tripping of "B" Recirc Pump PNPS 2.1.14 Section 7.9
- 6. RR07 C-RO Second Recirc Pump Trip - Manual Reactor Scram Required.
C-SRO
- 7. RR27 and M-AII ATWS requiring SBLC. RPV injection will be terminated and RD29 at prevented and RPV level lowered to below the feedwater 99% spargers.
- 8. TC09 C-BOP All turbine bypass valves fail closed. BOP is required to take C-SRO manual control of pressure with SRVs
- 9. LP01 C-RO Standby Liquid Control pump fails to start. RO must recognize C-SRO failure and start the other system. The pump that does start will trip after 1 minute.
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor
-2
Pilgrim 2011 NRC Scenario #2 The plant is operating at 90% power following a rod pattern adjustment the previous shift. Torus cooling is in service following a HPCI surveillance also conducted last shift. Equipment out of service consists of APRM "A" and RBCCW Pump "D". Tracking LCOs have been initiated for both components. Directions to the shift are to secure Torus Cooling and then restore power to 100%.
After assuming the watch, the BOP operator will secure Torus Cooling and return RHR to a normal standby lineup lAW PNPS 2.2.19, RHR. After Torus Cooling is secured, an Accumulator Trouble Alarm will be received on a withdrawn control rod. The field operator will report that the accumulator cannot be recharged above 800 psig. The SRO is expected to declare the Accumulator inoperable and determine that the associated Rod is to be declared "sloW" (or inoperable) within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> lAW TS 3.3.D.A.1.
Next the "F" RBCCW pump will trip and the standby RBCCW pump will fail to auto start but can be started manually. All RBCCW flow will be lost in the "B" RBCCW loop until the BOP starts the one remaining pump. The loss of "F" RBCCW in conjunction with the inoperable "D" RBCCW will render the "B" loop inoperable and a 7 day LCO should be declared lAW Technical Specification 3.5.B.3.
The brief disruption in cooling flow to the "B" Recirc Pump seal cooler will result in a failure of the inboard seal followed by a subsequent failure of the outboard seal. Drywell pressure and temperature will begin to rise. The BOP operator is expected to trip and isolate the pump lAW Pt\lPS 2.4.22, Recirc Pump Seal Failure. The crew is also expected to execute PNPS 2.4.17, Recirc Pump Trip following the manual pump trip.
The reduction in core flow will result in the plant entering the Unanalyzed Region of the power to flow map. The RO is expected to insert steps of the RPR array to reduce power to exit both the Unanalyzed and Exclusion Regions.
After the crew has stabilized the plant in single loop, the "An Recirc Pump will trip placing the plant on natural circulation. The RO is expected to insert a manual scram lAW the immediate actions of PNPS 2.4.17, Recirc Pump Trip. Due to a hydraulic lock most of the control rods will fail to insert. The crew is expected to enter EOP02, Failure to Scram in response to the event. Due to the high power, low flow condition, core-wide oscillations will occur. Expected EOP actions include injecting Standby Liquid (Critical Task), terminating injection and lowering level in order to mitigate core oscillations (Critical Task), and defeating the MSIV low level isolation. The RO is expected to use the Reactor Manual Control System and repeated manual scrams to achieve a rod pattern that will ensure the reactor will remain shutdown under all conditions (Critical Task).
A failure of all main turbine bypass valves will require the crew to establish alternate means of pressure control. Additionally Standby Liquid will initially fail to inject due to a pump trip. The RO is expected to recognize the failure and start the redundant train. That train will then trip after 1 minute.
The scenario may be terminated when EOP02 is exited and EOP01 , RPV Control, is entered to restore RPV level and commence a plant cooldown.
Emergency Classification: Site Area Emergency EAL: 2.3.1.3, Reactor power> 3% and boron injection into the RPV intentionally initiated.
-3
Scenario Event Description ES-D1 Pilgrim 2011 NRC Scenario 3 Facility: PILGRIM Scenario No.: 3 Op Test No.: 2011 NRC Examiners: Operators: SRO RO BOP-Initial Conditions:
- Reactor Power: 81%
- Plant Status: Tech Spec required shutdown in progress following a catastrophic failure of MO-1 001-23A, RHR Loop A, Upper Drywell Spray Valve
- 1, which cannot be repaired within the specified LCO time due unavailability of replacement parts.
- Core flow: 57 Mlbm/hr
- Current rod position: Step 85, rod 18-43
- Equipment Out of Service: "Au APRM has a faulty power supply and is OOS and bypassed. Tracking LCO initiated. All other APRMs are operable.
- 'W' Containment Spray Loop Inoperable. Day 6 of 7 day LCO 3.5.B.2 Turnover:
- Continue the plant shutdown lAW PNPS 2.1.5, Section F. Step [3] is in progress.
Critical Tasks: 1. Scram the reactor before torus water temperature exceeds 110 degrees following SOSRV.
- 2. Initiate drywell sprays when torus bottom pressure exceeds 16 psig
- 3. Emergency Depressurize the RPV when torus bottom pressure cannot be maintained below the Pressure Suppression Pressure.
Event Malt. No. Event Event Description No. Type*
- 1. R-RO Reduces core flow to 43 Mlbm/hr and inserts rods as necessary;;=
R-SRO achieve 75%. PNPS 2.1.5 Section F, steps [3] (b) and (d).
- 2. N-BOP Plant shutdown actions for 75% power: Adjust Speed Load N-SRO Changer and secure a Reactor Feed Pump. PNPS 2.1.5 Section F, steps [3] (c) and (e).
- 3. o Crywolf TS-SRO DC control power fuse blows for EDG "A" generating alarm C3L Annunciat B1, "GENERATOR BKR TRIP/INOP". Tech Spec 3.5.F.1 requires or C3L-B1 upgrading to a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> cold shutdown LCO.
- 4. CU02 C-BOP RWCU Pump Trip. The BOP is expected to shutdown the RWCU C-SRO system lAW PNPS 2.4.27, REACTOR WATER CLEANUP SYSTEM MALFUNCTIONS.
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Scenario Event Description ES-D1 Pilgrim 2011 NRC Scenario 3
- 5. RR20, C-RO Recirc Pump "A" speed controller fails upscale. Requires Manual Recirc C-SRO Scoop Tube Lock. PNPS 2.4.20, REACTOR RECIRCULATION Flow TS-SRO SYSTEM SPEED OR FLOW CONTROL SYSTEM Controller MALFUNCTION. SRO will be required to evaluate Recirc Speed Fails mismatch per Tech Spec 3.6.F.1.
Upscale.
- 6. MS13(B) C-AII Following the power rise, SRV 3B begins to leak then fails open.
and When the SRV cannot be closed, a manual scram is required.
MS14(B) PNPS 2.4.29, STUCK OPEN SAFETY RELIEF VALVE
- 7. ED13, C-BOP 4KV Bus A-1 fails to auto transfer following main turbine trip. BOP 4KV bus C-SRO operator required to manually re-energize.
A-1 fails to Auto Transfer.
B. PC22 (B) M-AII SRV 3B Tail Pipe Leak causing direct pressurization of Torus Air Ramped Space. Drywell sprays will be required and Emergency Depress to 100% required when torus bottom pressure cannot be maintained below over 5 the PSP curve. Reactor mode switch failure will result in MSIVs minutes closing when pressure 10WE!rS to < 810 psig.
- 9. Remote C-BOP When drywell pressure exc,eeds 2.2 psig, RHR pumps "B" and "D" functions C-SRO will fail to auto start. The BOP is expected to start both pumps for DC following the failure. At least one pump will be required for the Knife subsequent spraying of the drywell.
Blades for RHR pumps B and D
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor
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Pilgrim 2011 NRC Scenario #3 The plant is operating at 90% power with a Tech Spec required shutdown in progress due to an inoperable containment cooling subsystem. APRM "A" and "D" RBCCW pump are also OOS.
The directions to the crew are to continue the plant shutdown lAW PNPS 2.1.5. Reactor Plant Shutdown.
The crew will lower reactor power to - 75% at which point they will perform procedurally directed actions to adjust the MHC Speed Load Changer and remove a Reactor Feed Pump (RFP) from service. After the RFP is secured. a fuse will blow in the control power circuit for the "A" EDG output breaker. The SRO is expected to declare a 24 hr cold SID LCO based on TS 3.5.F.1 due to the "A" EDG being inoperable along with an "A" side containment cooling subsystem. Next the RWCU pump will trip following a loss of seal cooling. The BOP is expected to shutdown the RWCU system lAW PNPS 2.4.27, REACTOR WATER CLEANUP SYSTEM MALFUNCTIONS.
Next, the "A" Recirc Flow Controller output will fail upscale causing an increase in "A" Recirc Pump speed and reactor power. The RO is expected to diagnose the failure and insert a manual scoop tube lock lAW PNPS 2.4.20 Reactor Recirculation System Speed or Flow Control System Malfunction. The malfunction will result in a mismatch between recirc pump speeds. The SRO is expected to evaluate the mismatch lAW with Tech Spec 3.6.F. The crew may direct local control of the "A" recirc MG set be established and lower the speed of the "A" recirc pump to reduce the mismatch.
The RPV pressure and power rise will result in SRV 3B beginning to leak and eventually fail open. The crew is expected to respond lAW PNPS 2.4.29. Stuck Open SRV. When efforts to close the valve are unsuccessful and before torus water temperature reaches 110 degrees the crew is expected to insert a manual scram (Critical Task). When the main turbine trips, 4KV bus A-1 will fail to fast transfer. causing a partial loss of feed and condensate. The BOP is expected to diagnose the failure and manually re-energize the bus. Additionally. switch contacts within the Reactor Mode Switch will fail resulting in MSIV closure when RPV pressure lowers to 810 psig via the stuck open SRV.
Following the scram the tail pipe of the stuck open SRV will fail. resulting in direct steam pressurization of the torus air space. Containment pressure and temperature will rise rapidly.
When drywell pressure exceeds 2.2 psig. the low pressure ECCS will initiate. but RHR pumps "B" and "0" will fail to auto start. The BOP is expected to start both pumps following the failure.
Drywell sprays will be required when torus bottom pressure exceeds 16 psig (Critical Task).
Only one loop will be available due to the inoperable containment cooling loop. Drywell sprays will be insufficient to maintain containment pressure. When pressure cannot be maintained below the Pressure Suppression Pressure of EOP-03. Primary Containment Control. an Emergency RPV Depressurization will be required (Critical Task).
The scenario can be terminated when the RPV is depressurized. RPV level is stable and containment pressure is lowering.
Emergency Classification: Site Area Emergency EAL 3.4.1.3: Torus bottom pressure cannot be maintained below the "Pressure Suppression Pressure" (PSP) EOP Figure 6.
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Scenario Event Description ES-D1 Pilgrim 2011 NRC Scenario 5 I Facility: PILGRIM Scenario No.: 5 Op Test No.: 2011 NRC
- Reactor Power: - 89%.
- Plant Status: Reactor power at 90% following control rod exercising
- Core flow is 57 Mlbm/hr
- Current rod position: Sequence A-1 , Step 85, Rod 18-43
- Equipment Out of Service: "A" APRM has a faulty power supply and is OOS and bypassed. Tracking LCO initiated. All other APRMs are operable.
Turnover:
- Un-isolate HPCIIAW 2.2.125.1, RESET OF PRIMARY AND SECONDARY CONTAINMENT ISOLATIONS
- Restore power to 100%.
Critical Tasks: 1. Initiate drywell sprays when torus bottom pressure exceeds 16 psig
- 2. Emergency Depressurize the RPV when RPV level cannot be restored and maintained above -150 inches.
Event Malf. No. Event Event Description No. Type*
1.
~6 TS-SRO HPCI will be un-isolated and placed in Standby line up.
- 2. RX 18 Rx Level Transmitter LT-263-120D Fails Upscale. SRO refers to Tech Spec 3.2.G and associated Table 3.2.G and determines that ATWS /ARI Division Two is inoperable and that a 14 day Hot Shutdown LCO is required.
- 3. MSIV Test C-BOP MSIV closure requires a power reduction to < 75% lAW PNPS PB C-SRO 2.4.30, MSIV Closure and the BOP aligning the Main Steam Override. System per 2.2.92, Main Steam Line Isolation and Turbine Bypass Valves," for continued operation with a MSIV closed.
- 4. R-RO Power reduction to less than 75% following an MSIV closure.
R-SRO
- 5. RD01 C-RO CRD Flow Control Valve Fails Open. RO will shift to the standby C-SRO FCV lAW PNPS 2.4.11.1, CRD System Malfunctions.
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Scenario Event Description ES-D1 Pilgrim 2011 NRC Scenario 5
- 7. FW34"A" M-AII Feedwater line break inside the drywell. Reactor Scram and Entry Ramped conditions to EOP-01 , RPV Control and EOP-03, Primary to 40% Containment Control. Location of leak will render RCIC and the PC01, feed system ineffective as injection sources. HPCI injection will be Ramped required to maintain level. Drywell Sprays will be required.
to 2000 GPM
- 8. ASP C-BOP The HPCI Steam Admission valve M02301-3 will fail to auto open Remote C-SRO when the HPCI injection is required. The BOP is expected to Function determine that HPCI is not injecting, diagnose that the valve failed for HPCI to open and manually open the valve for HPCI injection to occur.
AuxOil pump
- 9. HP02 C-ALL HPCI turbine trip due to an oil leak. ADS will be inhibited as level drops and MSIVs will close. Both RO and BOP are expected to align alternate injection sources (CRD, Standby Liquid Control).
When RPV level cannot be maintained above -150 inches, an Emergency Depressurization will be performed lAW EOP-17.
- 10. RH04 C-80P LPCllnjection Valve fails to Open. When pressure lowers to < 400 C-SRO psig, LPCI injection valve, MO-298 will fail to open. The BOP operator is expected to recognize the failure and manually open the valve.
- (N)ormal. (R)eactivity, (I) nstrument, (C)omponent, (M)ajor
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Pilgrim 2011 NRC Scenario #5 The plant is operating at 90% following control rod exercising the previous shift. HPCI is currently isolated due to an I&C error during surveillance testing last shift and is currently inoperable. "A" APRM and "0" RBCCW pump are also out of service. All other APRMs and RBCCW pumps are operable. Directions to the shift are to un-isolate HPCI, place in a standby lineup and then restore power to 100%.
After the crew assumes the watch the BOP will un-isolate HPCI lAW 2.2.125.1. RESET OF PRIMARY AND SECONDARY CONTAINMENT ISOLATIONS and verify the system has been restored to a standby lineup lAW PNPS 2.2.21, HPCI System. After HPCI is restored, Rx Level Transmitter LT -263-1200 fails upscale. This transmitter is one of two level transmitters inputting into Division Two of the ATWS system. The SRO is expected to determine the impact of the failure, refer to Tech Spec 3.2.G and associated Table 3.2.G, determine that Division Two is inoperable and that a 14 day Hot Shutdown LCO is required.
Following the Tech Spec assessment, the outboard "B" MSIV will close. The crew is expected to respond lAW PNPS 2.4.30, MSIV Closure and reduce power to < 75%. The BOP is expected to align the main steam system per 2.2.92, Main Steam Line Isolation and Turbine Bypass Valves, for continued operation with a MSIV closed. Next, the in service CRD Flow Control Valve (FCV) fails open. The RO is expected to diagnose the failure and shift to the standby FCV lAW PNPS 2.4.11.1, CRD System Malfunctions.
Next a loss of 4 KV safety related bus A5 will occur. The crew is expected to respond lAW PI\IPS 2.4.A.5, LOSS OF ELECTRICAL BUS A5, and will be required to cross connect RBCCW, respond to a Recirc Pump Trip and address the loss of multiple safety- related components. The trip of the Recirc pump will place the reactor in the Exclusion region of the power to flow map. The RO will be required to insert the RPR array to exit the region. Multiple Tech Spec 24 hr cold shutdown requirements will exist due to the loss of safety related equipment.
The scenario ending event commences with an "A" Feedline break inside the drywell. Rising drywell pressure will result in a scram if not previously scrammed and entry conditions to EOP-01, RPV Control and EOP-03, Primary Containment Control. Initially the feed system will be able to maintain level but as the size of the leak progresses, it will become ineffective in controlling level.
The location of leak will render RCIC ineffective as well. HPCI will be required for level control.
The HPCI Steam Admission valve, M02301-3, will fail to open when a HPCI initiation signal is received. The BOP is expected to determine that HPCI is not injecting and open the valve manually to permit HPCI injection. Containment parameters Will require the use of Drywell Sprays (Critical Task).
Finally, the HPCI turbine will trip due to an oil leak. ADS will be inhibited as level drops and the MSIVs will close. Both the RO and the BOP are expected to align alternate injection sources (CRD, Standby Liquid Control). When RPV level cannot be maintained above -150 inches, an Emergency Depressurization will be performed lAW EOP-17 (Critical Task). When pressure lowers to < 400 psig, LPCI injection valve, MO-29B will fail to open. The BOP operator is expected to recognize the failure and manually open the valve. The scenario will be terminated at the discretion of the Lead Examiner OR when the RPV has been depressurized, RPV level stabilized and containment parameters are lowering.
Emergency Classification: Alert EAL 3.4.1 .2: Primary containment pressure cannot be maintained < 2.2 psig
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