ML071350523
| ML071350523 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee  |
| Issue date: | 03/12/2007 |
| From: | Entergy Nuclear Vermont Yankee |
| To: | Fish T H Operations Branch I |
| Sykes, Marvin D. | |
| Shared Package | |
| ML062050096 | List: |
| References | |
| ES-401, ES-401-1 | |
| Download: ML071350523 (22) | |
Text
ES-401 BWR Examination Outline Form ES-401-1 RO WA Category Points Tier Group K K K K K K A A A A G Tot 1234561234*
al 1. 1 344 42 3 20 Emergency Abnormal Evolutions
- 2. & 2 N/A N/A 'Iant Tier Totals 4 6 4 63 4 27 ----- 1 3233223123226 Systems TierTotals 3 4 I 3 14 3 12 4 15 2 I 38 Plant 2 1 1 111 12 0 12 2 0112 10 3. Generic Knowledge and 1 2 3 Abilities Categories 3 2 2 3 SRO-Only Points A2 G* Total 3 4 7 1 2 3 4 6 10 3 2 5 01 1 2 3 12347 4 4 8 1312 Ensure that at least two topics from every WA category are sampled within each tier of the RO outline (i.e., the "Tier Totals" in each WA category shall not be less than two). Refer to Section D.l .c for additional guidance regarding SRO sampling.
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 f 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. Select topics from many systems and evolutions; avoid selecting more than two WA topics from a given system or evolution unless they relate to plant-specific priorities.
Systems/evolutions within each group are identified on the associated outline. The shaded areas are not applicable to the categoryhier.
The generic (G)
WAS in Tiers 1 and 2 shall be selected from Section 2 of the WA Catalog, but the topics must be relevant to the applicable evolution or system. The SRO WAS must also be linked to 10 CFR 55.43 or an SRO-level learning objective.
On the following pages, enter the WA 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; summarize all the SRO-only knowledge and non-A2 ability categories in the columns labeled "K" and "A". Use duplicate pages for RO and SRO-only exams. Note: For Tier 3, enter the WA numbers, descriptions, importance ratings, and point totals on Form ES-401-3.
Refer to ES-401, Attachment 2, for guidance regarding the elimination of inappropriate WA statements.
- 1. 2. 3. 4. 5. 6.* 7. 8. NUREG-1021 1
ES-401 Emergency Procedures I Plan Knowledge of operational implications of EOP warnings/cautions/notes 2.4.20 Vermont Yankee NRC BWR SRO/RO Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 1 4,0 Form ES-401-1 Knowledge of the reasons for the following responses as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION
- Reactor power response.. . . .. . . . . . . . . . . . .. . . .. . . . . . . . . . Ability to operate and/or monitor the following as they apply to PARTIAL OR COMPLETE LOSS OF A.C. POWER : Emergency generators
..................................
3'7 4.2 76 295007 High Reactor Pressure
/ 3 X Emergency Procedures I Plan: Knowledge of the parameters and logic used to assess the status of safety functions including: 1 reactivity Control 2. Core Cooling and heat removal.
- 3. Reactor coolant system integrity
- 4. Containment conditions.
- 5. Radioactivity release control. (Hi secondarv containment area tern&. 1 3.8 295032 High Secondaty Containment Area Temp. I5 X 77 2.4.21 2.2.29 78 Equipment Control Knowledge of SRO fuel handling responsibilities.
Ability to determine and/or interpret the following as they apply to HIGH DRYWELL PRESSURE:
Suppression pool level .... . . . . . . . . . . . . . . . . . . . . .
.. . . __. 295023 Refueling Acc Cooling Mode I8 X 295024 High Drywell Pressure I5 X 1 EA2.03 79 Ability to determine and/or interpret the following as they water level apply to HIGH DRYWELL TEMPERATURE
- Reactor I 3.9 80 - 81 295028 High Drywell Temperature I5 295030 Low Suppression Pool Water Level / 5 + EA2.01 t 2.4.7 Ability to determine and/or interpret the following as they apply to LOW SUPPRESSION POOL WATER LEVEL : Sumression WOI level ................................
X - Emergency ProceduredPlan Knowledge of Event based I 3,8 EOP mitigation strategies 82 - 39 295031 Reactor Low Water Level I2 X 295001 Partial or Complete Loss of Forced Core Flow Circulation
/ 1 & 4 295003 Partial or Complete Loss of AC I6 --L Ix AK3.02 40 42 AA1.02 Ability to operate and/or monitor the following as they apply to PARTIAL OR COMPLETE LOSS OF D.C. POWER : Systems necessary to assure safe plant shutdown .........
..........
.... ... 3.8 295004 Partial or Total Loss of DC Pwr / 6 AA1.02 + II ~ 3.1 Ability to determine and/or interpret the following as they apply to MAIN TURBINE GENERATOR TRIP : Turbine valve position ................................
Emergency Procedures/Plan: Knowledge of Abnormal Condition Procedures AA2.03 295005 Main Turbine Generator Trip I3 295002 Loss of Main Condenser Vacuum / 8 2.4.1 1 3.4 43 - NU REG- 1 02 1 2 ES-401 295016 Control Room Abandonment
/ 7 29501 8 Partial or Total Loss of CCW / 8 295019 Partial or Total Loss of Inst. Air / 8 295021 Loss of Shutdown Cooling
/ 4 295023 Refueling Acc Cooling Mode / 8 295024 High Drywell Pressure
/ 5 295025 High Reactor Pressure / 3 295026 Suppression Pool High Water Temp. / 5 X X Vermont Yankee NRC BWR SRO/RO Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 1 X X X X Form ES-401-1 AK3.03 Knowledge of the reasons for the following responses as they apply to CONTROL ROOM ABANDONMENT
- Disabling Control Room Controls..
~ Knowledge of the operational implications of the following concepts as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER and the following: Effects on components/system operations Knowledge of the interrelations between PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR and the 1 following: Plant ventilation. 295031 Reactor Low Water Level
/ 2 AK2.03 AK3.02 Ill Knowledge of the interrelations between LOSS OF SHUTDOWN COOLING and the following: RHWshutdown cooling Knowledge of the reasons for the following responses as they apply to REFUELING ACCIDENTS
- Interlocks associated with fuel handling equipment
.... X X x X X 2.4.2 Emergency Procedures
/ Plan: Knowledge of system setpoints/interlocks and automatic actions associated with EOP entry condtions.
~ AK1.01 EK2.03 AK2.08 Knowledge of the interrelations between SUPPRESSION POOL HIGH WATER TEMPERATURE:
Suppression Chamber Pressure.
295028 High Drywell Temperature
/ 5 295030 Low Suppression Pool Water Level / 5 X Conduct of Operations:
Ability to explain and apply all svstem limits and Drecautions.
2.1.32 I Ability to operate and/or monitor the following as they apply to High Secondary Containment Sump/Area Water Level: Affected systems so as to isolate damaged temperature and the following: Drywell ventilation.
EA1.05 EA2.04 Ability to operate and/or monitor the following as they apply to LOW SUPPRESSION POOL WATER LEVEL : HPCl .........................
Ability to interpret and/or determine the following as they apply to REACTOR LOW WATER LEVEL Adequate Core Cooling. 3.5 3.5 2.8 3.6 3.4 3.9 3.4 3.2 3.5 3.6 3.5 4.6 44 46 47 49 - 50 NU R EG- 1 02 1 3 ES-401 Vermont Yankee NRC BWR SRO/RO Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 1 Form ES-401-1 295037 SCRAM Condition Present and Power Above APRM Downscale or Unknown / 1 I 1 295038 High Off-site Release Rate
/ 9 600000 Plant Fire On-site
/ 8 I WA Category Point Totals:
2/3 EK3.03 EK1.02 Knowledge of the reasons for the following responses as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN : Lowering Reactor Water Level Knowledge of the operational implications of the following concepts as they apply to HIGH OFF-SITE RELEASE 4.2 57 RATE : Protection of the general public
.............................
4.1 56 Knowledge of the operations applications of the following concepts as they apply to PLANT FIRE ON SITE: Fire Fighting 2.9 58 AK1.02 Group Point Total: I 20l7 NU REG- 1 02 1 4 ES-401 X Vermont-Yankee NRC BWR SRO/RO Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 2 X X 12 Form ES-401-1 0 X X X X 2 111 AA2.04 Ability to interpret or determine the following as they apply to COMPLETE OR PARTIAL LOSS OF AC POWER: Svstetn Lineum AK2.03 Knowledge of the interrelations between PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR and the following: Reactor Feedwater. Emergency Procedures I Plan Knowledge of the specific bases for EOPs 2.4.18 I 2,7 6o AA1.01 Ability to operate and/or monitor the following as they apply to INCOMPLETE SCRAM: CRD Hydraulics..
.................
Group Point Total:
713 295010 High Drywell Pressure I5 X - X I 3.4 I 83 I Emergency Procedures I Plan Knowledge of symptom based EOP mitiaatiin strateaies 2.4.6 I I 4.0 I I Conduct of Operations: Abillty to direct personnel activities inside the control room 2.1.9 I 295012 High Drywell Temperature
/ 5 295003 Partial or Complete loss of AC Power / 6 3.2 I 59 1 29501 9 Partial or Total Loss of Inst. Air I 8 X 295010 High Drywell Pressure 15 Ability to determine and/or interpret the following as they apply to INADVERTENT REACTIVITY ADDITION:
Violation of Fuel Thermal Limits. AA2.04 4.1 61 - 62 295014 Inadvertent Reactivity Addition I 1 295012 High Drywell Temperature 15 Ability to operate andor monitor the following as they apply to HIGH DRYWELL TEMPERATURE
- Drywell ventilation svstem ............................
AA1.01 3.5 3.8 1 63 1 295015 Incomplete SCRAM I 1 Knowledge of the operational implications of the following concepts as they apply to HIGH SECONDARY CONTAINMENT AREA RADIATION LEVELS : Radiation releases ....................................
Knowledge of the interrelations between SECONDARY CONTAINMENT HIGH DIFFERENTIAL PRESSURE and the following: Blowout Panels/Plant SDecific ..................
295033 High Secondary Containment Area Radiation Levels I9 EK1.03 3.9 295035 Secondary Containment High Differential Pressure I5 EK2.04 3.3 - 1 12 ~ WA Category Point Total:
I System #/Name ES-401 Vermont Yankee NRC BWR SRO/RO Written Examination Outline Plant Systems - Tier 2 Group 1 211OooSLC Form ES-401-1 206000 HPCl 21 8000 ADS 262001 AC Electrical Distribution
~~~ 204000 RWCU 203000 RHWLPCI: Injection Mode 205000 Shutdown Cooling 205000 Shutdown Cooling 206000 HPCl NUREG-1 021 6 206000 HPCl 209001 LPCS 223001 Primary Containment System and Auxiliaries 21 1000 SLC 212000 RPS 215003 IRM 21 5004 Source Range Monitor 215005 APRM I LPRM 217000 RClC 21 8000 ADS 223002 PCIS/Nuclear Steam Supply Shutoff X A3.03 PRESSURE COOLANT INJECTION SYSTEM: System 3.9 5 Lineup Knowledge of electrical power supplies to the following:
Initiation logic Knowledge of the purpose and function of major system Ability to (a) predict the impacts of the following on the STANDBY LIQUID CONTROL SYSTEM ; and (b) based or mitigate the consequences of those abnormal conditions or operations:
AC Power Failures Knowledge of the effect that a loss or malfunction of the SYSTEM: AC Electrical Distribution Ability to manually operate and/or monitor in the control room: Verification of proper functioning/ operability Knowledge of electrical power supplies to the following:
SRM channels/detectors Knowledge of AVERAGE POWER RANGE MONITOWLOCAL POWER RANGE MONITOR SYSTEM design feature(s) and/or interlocks which provide for the following:
Rod withdrawal blocks Ability to predict and/or monitor changes in parameters associated with operating the REACTOR CORE ISOLATION COOLING SYSTEM (RCIC) controls including:
RClC flow Knowledge of operational implications of the following DEPRESSURIZATION SYSTEM
- ADS Logic Operation Ability to monitor automatic operations of the PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR 2.9 6 X X 2.1 .28 components and controls. 3.2 7 X A2.03 on those predictions, use procedures to correct, control, 3.2 8 X K6.01 following will have on the REACTOR PROTECTION 3.6 9 3.6 10 A4,07 2.6 11 X 3.7 12 K4.01 X 3.7 13 Al ,ol X X K5.01 concepts as they apply to AUTOMATIC 3.8 14 NU REG- 1 02 1 7 239002 SRVs 259002 Reactor Water Level Control - Knowledge of the effect that a loss or malfunction of the Over pressurization Knowledge of the physical connections and/or cause-CONTROL and the following: Reactor water level X K3.02 reliefkafety valves will have on the following: Reactor 4.2 16 X K1.03 effect relationships between REACTOR WATER LEVEL 3.8 17 Control 261 000 SGTS 262001 AC Electrical Distribution 262002 UPS (ACIDC) 263000 DC Electrical Distribution 264000 EDGs 300000 Instrument Air 400000 Component Cooling Water 400000 Component Cooling Water KIA Category Point Totals:
NU REG- 102 1 recirculation pumps: Plant-Specific Emergency Procedures
/ Plan Knowledge symptom Knowledge of the physical connections and/or cause- effect relationships between A.C. ELECTRICAL 3.1 19 X 2'4'6 based EOP mitigation strategies.
3.1 20 X K1 '04 DISTRIBUTION and the following: Uninterruptible power supply Ability to manually operate and/or monitor in the control X A4.01 room:
Transfer from alternative source to preferred 2.8 21 source Knowledge of the physical connections and/or cause- DISTRIBUTION and the following: Ground detection Knowledge of EMERGENCY GENERATORS provide for the following: load shedding and sequencing Knowledge of the effect that a loss or malfunction of the Systems having pneumatic valves and controls Ability to manually operate and/or monitor in the control room: CCW indications and control Ability to predict and / or monitor changes in parameters X Al.04 associated with operating the CCWS controls including: 2.8 26 Surge Tank Level X K1.04 effect relationships between D.C. ELECTRICAL 2.6 22 X K4.05 (DIESEUJET) design feature(s) and/or interlocks which 3.2 23 X K3.02 INSTRUMENT AIR SYSTEM will have on the following:
3.3 24 3.1 25 A4,0, 2/23 2 3 3 2 2 31/32 3 Group Point Total: 2615 8 204000 RWCU 214000 RPlS 21 9000 RHWLPCI: Torus/Pool Cooling Mode 226001 RHWLPCI: CTMT Spray Mode 239001 Main and Reheat Steam 241 000 Reactornurbine Pressure Regulator 271 000 Off-gas IG I System #/Name I Knowledge of the operational implications of the CLEANUP SYSTEM : Heat exchanger operation Ability to (a) predict the impacts of the following on the ROD POSITION INFORMATION SYSTEM ; and (b) control, or mitigate the consequences of those abnormal conditions or operations: Reactor SCRAM Ability to monitor automatic operations of the X A3.01 RHWLPCI: Torus suppression pool cooling mode 3.3 31 including:
valve operation Ability to manually operate and /or monitor in the Knowledge of the effect that a loss or malfunction of the Electrical Power Ability to manually operate and/or monitor in the control Knowledge of the physical connections and/or cause the following:
Main Steam System.
X K5.04 following concepts as they apply to REACTOR WATER 2.7 29 X A2.02 based on those predictions, use procedures to correct, 3.6 30 A4.07 control room: valve logic reset/bypass/override 3.5 32 X K6.01 following will have on the MAIN AND REHEAT STEAM: 3.1 33 3.9 34 A4.06 room: Bypass valves operation X K1.06 effect relationships between OFFGAS SYSTEM and 2.8 35 202001 Recirculation 21 5005 APRMILPRM 256000 Reactor Condensate 201 006 RWM I 202001 Recirculation BWR SRO/RO Written Examination Outline Plant Systems - Tier 2 Group 2 I directives affect plant and system conditions.
Knowledge of the effect that a loss or malfunction of the ROD WORTH MINIMIZER SYSTEM (RWM) (PLANT SPECIFIC) will have on following:
Reactor manual control system: P-Spec(Not-BWR6)
Knowledge of electrical power supplies to the following: Recirc System Valves K3.01 3,2 27 X K2.03 28 2,7 X NU REG- 1 02 1 9 Facility: Vermont Yankee NRC Date of Exam: 412007 SRO-Only l7A-L-E Topic I I I I I I I 4.0 94 Ability to apply technical specifications for a 2.1.20 Ability to execute procedure steps. 4.3 66 2.1.3 Knowledge of shift turnover practices 3.0 67 3.8 68 2.1 .28 Knowledge of the purpose andlor function of major system components and controls.
- 1. Conduct of Operations I I Subtotal I 131 12 I 2.2.25 Knowledge of bases in TS for LCOs and safety Limits 3.7 95 96 97 I 2.2.24 I- Ability to analyze the effect of maintenance activities on LCO status. Knowledge of the refueling administrative requirements.
Ability to manipulate the console controls as required to operate the facility between shutdown and designated power levels.
Knowledge of new and spent fuel movement procedures.
3.8 3.7 2.2.26 t- 2.2.2 2. Equipment Control 4.0 69 2.6 70 2.2.28 Subtotal - - Ability to perform procedures to reduce excessive levels of radiation and guard against personnel exposure.
2 2 Knowledge of radiation exposure limits and contamination control1 including permissible levels in excess of those authorized.
Knowledge of facility ALARA program. 2.3.4 2.3.2 2.3.10 E Subtotal 3. Radiation Control 3.1 98 2.5 71 2.9 72 2 1 I 2.4.25 I Knowledge of fire protection procedures.
3.4 99 4.1 100 Knowledge of the emergency action level Procedures I Plan 4. Emergency
~- I 24.1 0 1 Knowledge Ofannunciafor response procedures.
3.0- I 75 Subtotal Tier 3 Point Total 2 7 - NUREG-1021 11 Reason for Rejection Tier / Randomly Group Selected WA 29501 9 K2.13 dant. (Q.#46) AK2.08, randomly selected - original selection did not apply to 111 295031 2 1 28 295028 2.4.30 (Q. #82) Impossible to meet KA Topic requirement at SRO level. Randomly reselected G2.1.20 for APE. G2.4.7 randomly selected. (Q. #53). EK2.04 randomly selected - original selection not applicable a1 RO level at this Dlant. Also e 2.5 ImP. 1:1 111 295005 K3.02 259002 K2.02 the plant. (Q.#3) K3.05 randomly selected - double jeopardy with Q.# 47 (Q. #17) K1.03 randomly selected, original selection does not apply to 211 211 11 3 I 2.2.3 I (Q.#69) 2.2.2 randomly selected, original selection is for multi unit plant 202001 K2.04 plant 2009002 K6.01 apply to the plant 29501 1 AA2.01 apply to plant. 207000 A3.02 amlv to Dlant ((2.28) K2.03 randomly selected, original selection does not apply to (Q.#7) 223001 2.2.28 randomly selected, original selection does not (Q.#61) 295014 AA2.04 randomly selected, original selection does not (Q.#5) 206000 A3.03 randomly selected, original selection does not 212 211 211 295027 EAl.01 plant. 29501 1 2.4.31 apply to plant. 206000 (Q.#52) 295032 randomly selected, original selection does not apply to (Q.#84) 29501 0 2.4.6 randomly selected, original selection does not (Q.# 88) 2.1.1 1 randomly selected, original selection was double 111 112 211 2.05 ieoDardv with Q.#41. (Q.#68) 2.1.28 randomly selected, original selection was not an RO level topic at VY. (Q.#92) 21 5005 G2.1.12 randomly reselected - an operationally valid G2.1 m8 21 5001 3 n in A2.08 295007 question at the SRO level could not be written for original selection. (Q.#76) 2.4.20 randomly reselected, original selection was an RO level L1 L 4 I4 2.4.50 topic 295020 AA2.02 295009 AK1.03 295006 II I (Q.#86) 295003 AA2.04 randomly reselected, an operationally valid question at the SRO level could not be written for original selection (Q.#59) 29501 9 AK2.03 randomly reselected due to double jeopardy with another reactor level topic KA. (Q.#43) 295002 2.4.1 1 randomly reselected, operationally valid 111 4 I1 2.1.28 discriminating question could not be written for original selection.
Ii I n ~n 290002 (Q.#38)
K4.01 randomly reselected.
An operationally valid, K4.02 G2'1 '4 G2'2*20 discrim-inating question-could not be written for original selection (Q.#95) 2.2.25 randomly reselected.
An operationally valid SRO level question could not be written for the original selection (Q.#96) 2.2.1 2 randomly reselected.
An operationally valid SRO level question could not be written for the original selection L1 L 3 3 NUREG-1021 12 ES-301 Administrative Topics Out1 i ne Form ES-301-1 Facility: Vermont Yankee Date of Examination:
4 130 107 Examination Level (circle one): RO / SRO Operating Test Number:
2007 NRC Type Code* D, S M, C Describe activity to be performed JPM: Perform Reactor Coolant Temperature Check WA: 2.1.7 (3.7) Ability to evaluate plant performance and make operational judgments based on operating characteristics
/ reactor behavior
/ and instrument interpretation. JPM: Perform Shutdown CRO Rounds WA: 2.1.33 (3.4) Ability to recognize indications for system operating parameters which are entry-level conditions for technical specifications. Prepare Switching and Tagging Order JPM: WA: 2.2.13 (3.6) Knowledge of tagging and clearance procedures. JPM: Locate and Determine Radiological Requirements for Inspection of RCU Valve V12-19A (CU-19A) WA: 2.3.1 (2.6) Knowledge of 10 CFR: 20 and related facility radiation control requirements.
NIA NUREG-1021, Revision 9
ES-301 Administrative Topics Outline Form ES-301- 1 *Type Codes
& Criteria: (C)ontrol room Class( R)oom (D)irect from bank (I 3 for ROs; I for SROs & RO retakes) (N)ew or (M)odified from bank
(> 1) (P)revious 2 exams (I 1 ; randomly selected) (S)imulator 2007 NRC Examination Summary Description of Admin Tasks A.l .a The candidate will perform reactor coolant temperature checks. This is a bank JPM. The candidate is required to recognize that the temperature difference is greater than 145 deg F and determine that the recirculation pump may not be started. This is a bank JPM. A.l .b The candidate will perform a portion of the Shutdown CRO Rounds. The candidate is required to recognize abnormal and out of spec conditions which are entry-level conditions for technical specifications. This is a new JPM. A.2 The candidate will prepare a switching and tagging order for change-out of the CRD pump suction filter with the computerized switching and tagging program unavailable. This is a bank JPM. A.3 The candidate will locate and determine radiological requirements for Inspection of RWCU valve V12-19A (CU-1 9A), including a calculation of stay time, determination of areas with the lowest dose, and determination of areas with the lowest contamination levels. This is a bank JPM. This JPM was used on the 2005 NRC exam; however, task conditions will be modified to result a different stay time and new areas of low dose and contamination levels.
NUREG-1021, Revision 9 ES-301 Administrative Topics Outline Form ES-301-1 Facility: Vermont Yankee Date of Examination:
4-30-0'1 Examination Level (circle one): RO / SRO Operating Test Number: 2007 NRC ~~ Administrative Topic Conduct of Operations Equipment Control Radiation Control Type Code* N D N N D Describe activity to be performed JPM: Evaluate an OD-3 printout following a Power Ascension to determine if Thermal Limits were violated.
WA: 2.1.7 (4.4) Ability to evaluate plant performance and make operational judgments based on operating characteristics
/ reactor behavior
/ and instrument interpretation.
JPM: Review Completed Surveillance and Take Action for Out of Spec Data WA: 2.1.12 (4.0) Ability to apply technical specifications for a system. JPM: Review Switching and Tagging Order WA: 2.2.13 (3.8) Knowledge of tagging and clearance procedures.
JPM: Review and Approve Primary Containment Purge cumulative hours log WA: 2.3.9 (3.4) Knowledge of the process for performing a containment purge JPM: WA: 2.4.44 (4.0) PAR Based on Plant Conditions (Shelter)
Knowledge of emergency plan protective action recommendations 4 items unless NUREG-1021, Revision 9 ES-301 Administrative Topics Outline Form ES-301-1 *Type Codes
& Criteria: (C)ontrol room (D)irect from bank (I 3 for ROs; 4 I for SROs & RO retakes) (N)ew or (M)odified from bank (> 1) (P)revious 2 exams (I 1 ; randomly selected) (S) i m ulator A.l .a A.l .b A.2 A.3 A.4 2007 NRC Examination Summary Description of Admin Tasks The candidate will. This is a new JPM. The candidate will review a completed RHR system surveillance and take action for out of spec data. This is a bank JPM. The candidate will review a switching and tagging order for 'A CRD pump, identify tagging errors, and determine that the tagout cannot be approved as written. This is a new JPM. The candidate will review the containment purge cumulative hours log in preparation for a containment purge.
The hour s log will be inaccurate and the candidate must determine that the purge can not be approved. This a new JPM. The candidate will make the initial PAR based during a LOCA event with a release in progress per OP 351 1. The candidate will determine that shelter is required. The task is time critical. This is a bank JPM. NUREG-1021, Revision 9
ES-301 Control Room/ln-Plant Systems Outline Form ES-301-2 4- 30 07 Faci I i t y: Vermont Yankee Date of Examination:
Exam Level (circle one): RO / SRO(I) / SRO (U) Operating Test No.: NRC 2007 Control Room Systems@ (8 for RO; 7 for SRO-I; 2 or 3 for SRO-U, including 1 ESF) Type Code*
System / JPM Title S-1 201 006 Source Range Monitor N, A, S, L Rx startup to criticality, high reactor period 5-2 261000 Standby Gas Treatment System p, s Secure Standby Gas Treatment S-3 241 000 Reactorflurbine Pressure Regulating System Transfer Pressure Control From MPR to EPR S-4 262001 A.C. Electrical Distribution System Energize Bus 4 From the Vernon Tie Line During a SBO S-5 217000 Reactor Core Isolation Cooling System Respond to Automatic RClC Auto Controller Failure S-6 209001 Core Spray System Perform Core Spray "A" Quarterly Full Flow Test S-7 223002 Primary Containment Isolation System PClS Group V isolation failure S-8 201 003 Control Rod and Drive Mechanism In-Plant Systems@ (3 for RO; 3 for SRO-I; 3 or 2 for SRO-U) Safety Function 7 9 3 6 2 4 5 1 NUREG-1021, Revision 9 ES-301 Control Room/ln-Plant Systems Outline Form ES-301-2 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 I (A)lternate path (C)ontrol room (D)irect from bank (E)mergency or abnormal in-plant (L)ow-Power I Shutdown (N)ew or (M)odified from bank including 1(A) (P)revious 2 exams I (WA s-1 s-2 s-3 s-4 s-5 S-6 s-7 Criteria for RO I SRO-I I SRO-U 4-6 I 4-6 12-3 ~11sais4 21 127 I21 21 I27 121 22122127 5 3 I s 3 15 2 (randomly selected) 2 1 I2 7 I2 7 2007 NRC Examination Summary Description of JPMs The candidate will continue a reactor startup pulling control rods while approaching criticality.
This is a new JPM in the Source Range Monitor System - Instrumentation Safety Function.
The alternate path requires that the candidate recognize indications of a sustained reactor period shorter than 30 seconds and take action to turn reactor period using the Emergency-In Switch IAW OP 0105. The candidate will secure SBGT, returning both SBG trains to a normal lineup. This is a bank JPM in the Standby Gas Treatment System - Radioactivity Release Safety Function. This JPM was used on the 2005 NRC exam.
The candidate will transfer Pressure Control From MPR to EPR - Safety Function. This is a modified bank JPM.
The candidate will energize Bus 4 from the Vernon Tie Line during a station blackout. This is a bank JPM in the A.C. Electrical Distribution System - Electrical Safety Function.
The alternate path requires that the candidate recognize indications of a failure of the bus tie breaker to close, requiring action to close the alternate bus tie breaker. The candidate will respond to an automatic RClC flow controller failure. This is a bank JPM in the Reactor Core Isolation Cooling System - Reactor Water Inventory Control Safety Function. The alternate path requires that the candidate recognize indications that RClC should have isolated, requiring action to manually trip and isolate RCIC. The candidate will perform the Core Spray "A" Quarterly Full Flow Test. This is a modified alternate path JPM in the Core Spray System - Heat Removal from The Core Safety Function.
The alternate path requires that the candidate recognize indications that the core spray pump minimum flow valve has failed to open when required, requiring action to trip the pump. This is a modified The candidate will backup a Group V isolation.
This is a bank JPM in the Primary Containment Isolation System - Containment Integrity Safety Function. The alternate requires the candidate to recognize the failure of RWCU to isolate and manually close the valves upon SLC initiation.
NUREG-1021, Revision 9 ES-301 Control Room/ln-Plant Systems Outline Form ES-301-2 S-8 The candidate will perform the weekly operable control rod check IAW OP 41 11. This is a modified JPM in the CRDM system - Reactivity Control Safety Function. This was used on the 2003 NRC exam with an alternate path.
No alternate path will be performed for this JPM. P-1 The candidate will shutdown the diesel generator locally. This is a bank JPM in the Emergency Generators System - Electrical Safety Function.
P-2 The candidate will lineup to operate SRV-71 A and B from the RClC room. This is a bank JPM in the Safety Relief Valves System - Pressure Control Safety Function.
The candidate will perform local firing of SBLC squib valves. This is a bank JPM in the Standby Liquid Control System - Reactivity Control Safety Function.
P-3 NUREG-1021 Revision 9 I AppendixD Scenario Outline Form ES-D-1 1 :acil i ty: VERMONT YANKEE Scenario No.:
1 Op Test No.: 2007 NRC :xaminers:
Operators:
nitial Conditions: -At 90% Dower for control rod Dattern adiustment.
a Power ascension required back to 100%. rurnover: Perform weekly remote testing of Turbine Oil pumps IAW OP 41 60. 2ritical Task: Malf. No. Event lType Event Description Event No. 1 NIA Weekly remote testing of Turbine Oil pumps OP 4160. N-ACRO N-CRS Power ascension IAW OP 0105. 2 NIA R-CRO N-CRS Loss of Bus 89B (TS). Loss of Circ Water Pump.
3 mfED-19B mfED-19B I-CRS C-ACRO C-CRS C-CRO C-CRS 4 5 mfRD-01 A CRD Pump A trips (ON). C-CRO C-CRS Control Rod 18-31 drifts outward (OT). 6 rnfRD-051831 100% 7 RC04 Inadvertent HPCl initiation (TS). I-ACRO I-CRS M-ALL 8 mfED-17 Loss of Offsite power 9 mfHP-04 0% C-ACRO C-CRS HPCl Flow Controller Failure.
10 M-ALL Recirc loop rupture (0.7% over 300 sec). HPCl trip. RPV-ED on low level. mf RR-0 1 A mf HP-0 1 mfCS-03A mf CS-03 B mfRH-07A 11 Preinsert P rei nsert C-CRS C-ACRO CS-12A and CS-12B failure to auto open. RHR 27A failure to auto open.
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor NUREG 1021 Revision 9 Appendix D I Appendix D Scenario Outline Form ES-D-1 I Facility: VERMONT YANKEE Scenario No.: 2 Op Test No.: NRC 2007 Examiners: Operators:
Initial Conditions:
0 100% power, preparing to chlorinate the Circ Water System 0 APRM C is bypassed due to inability to adjust gain - I&C troubleshooting is in progress RHR-39A Valve motor actuator is being repaired day LCO entered 1 day ago per TS 3.5.8.1 Turnover:
0 Place CW in Closed Cycle for chlorination.
0 Reduce reactor power in preparation for a control rod pattern adiustment.
Critical Task: Event I Malf. No. No. Event Type* Event Description I N'A N-CRS N-ACRO Place CW in Closed Cycle for chlorination.
Power Reduction IAW OP 01 05. R-CRO N-CRS C-CRS C-CRO 3 Feedwater regulating valve lockup (OT). APRM A fails downscale (TS) EPR Oscillations (OT). FW-O9A 4 mfNM-05A 0% I-CRO I-CRS 5 mfT C-04A C-ACRO C-CRS 6 mf E D-05C mfPC-11 A C-ALL Loss of 480V Bus 8 (TS), Failure of SBGT A to auto start. 7 mfAD-01 B C-ALL SRV-71 B leak (OT) leads to Rx scram (1 00% over 600 sec). mfRPOl B I-CRO Failure of manual scram. ARI required.
8 Preinsert 9 Preinsert
' 10 Preinsert I-CRS M-ALL 45% hydraulic ATWS (A). 55% hydraulic ATWS (B). mfRD-12A mfRD-12B mfSL-01 A mfSL-02B C-CRO C-CRS SLC pump A trips. B SLC squib valve fails to fire. (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor NUREG 1021 Revision 9 Appendix D Event Malf. No.
Event 1 NIA N-ALL No. Type* Event Description Perform Turbine Chest Warmup. 4 5 mfRD-11A C-CRO CRD Flow control Valve fails closed (ON).
C-CRS mfRD-02xxyy C-CRO Stuck Control Rod XX-YY (ON) C-CRS Appendix D Scenario Outline Form ES-D-1 I Faci I ity : VERMONT YANKEE Scenario No.: 3 Op Test No.: 2007 NRC Examiners:
Operators:
Initial Conditions:
0 Power is 2% with a reactor startup in progress.
Turnover:
0 OP 01 05 is complete thru Phase 2.C. ~ ~~~~~ 0 Perform Turbine Chest Warmup IAW OP 01 05 Phase 2.D. Step 1. 0 Continue Reactor Startup (60 to 80 degree heat up rate). Critical Task: I NIA R-CRO Pull rods to continue power ascension.
I N-CRS I 1 :-cis I IRM A faits upscale (TS). I NM04A ~ 6 Seismic event.
TBCCW "A" Pump Trip w/ TBCCW "B Pump Failure to auto start C-ACRO C-CRS C-ACRO C-CRS rfPP-06 mfSW-14A mfSW-21 B mfHP-11 mfHP-15 7 RClC steam leak (TS). RCIC fails to auto isolate. 8 rfPP-06 mfRR-18H M-ALL Seismic aftershock.
Group 1 isolation.
9 Auto scram failure. Manual scram required.
mfRP-01 C mfRP-OlA mfRP-08A mf RP-08B I-CRO I-CRS I-ACRO I-CRS 10 Preinseri PClS Group Ill failure. zC-lOpl M-ALL 1 Torus leak at "A" RHR suction (50% over 900 secs). I PRV-ED on low Torus level. * (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor NUREG 1021 Revision 9 Appendix D I Amendix D Scenario Outline Form ES-D-1 I Facility: VERMONT YANKEE Scenario No.: 4 Op Test No.: 2007 NRC Examiners:
Operators:
11 Initial Conditions:
0 90% power for control rod pattern adjustment which has been completed.
Turnover:
0 0 Return to 100% power. Perform Speed Load Changer Bypass Test IAW OP 41 60. Critical Task:
Event No. 2 3 4 5 6 7 8 Malf. No. NIA mfED-05Ca mfCD-01 A rfFW-04 mfRRQ1A mf MS-06 mfPC1-06 mfPCl-6B )ormal, (R)ei Event Type* N-ACRO N-CRS R-CRO N-CRS C-ALL C-CRS C-CRO I-CRO I-CRS C-CRS C-ACRO M-ALL C-CRS C-ACRO Event Description Speed Load Changer Bypass Test OP 41 60. Power ascension IAW OP-0105. LOSS of MCC-8A (TS). Condensate Pump A trips. Failure of RFP B to trip. (OT) (RP) Feedwater Pump B CD trip bypass switch in bypass. Steam flow summer fails upscale (OT) (1 00% over 60 sec). Recirc Leak (OT) (TS). MS line "A rupture in drywell. (10% over 1200 secs). AC-6 and AC-6B fail to auto close. ctivity, (I)nstrument, (C)omponent, (M)ajor Appendix D NUREG 1021 Revision 9