ML071350523

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Draft - Exam Outlines (Folder 2)
ML071350523
Person / Time
Site: Vermont Yankee Entergy icon.png
Issue date: 03/12/2007
From:
Entergy Nuclear Vermont Yankee
To: Todd Fish
Operations Branch I
Sykes, Marvin D.
Shared Package
ML062050096 List:
References
ES-401, ES-401-1
Download: ML071350523 (22)
v • d • e


Text

ES-401 BWR Examination Outline Form ES-401-1 RO WA Category Points SRO-Only Points Tier Group K K K K K K A A A A G Tot A2 G* Total 1 2 3 4 5 6 1 2 3 4

  • al
1. 1 3 4 4 4 2 3 20 3 4 7 Emergency

& 2 N/A N/A 1 2 3 Abnormal Iant Tier Totals 4 6 4 6 3 4 27 4 6 10

-Evolutions

- - - -

1 3 2 3 3 2 2 3 1 2 3 2 2 6 3 2 5 2.

Plant 2 1 1 111 1 2 0 1 2 2 0112 0 1 1 2 3 Systems TierTotals 3 4 I 3 14 3 12 4 15 2 I 38 4 4 8

3. Generic Knowledge and 1 2 3 1 2 3 4 7 10 Abilities Categories 3 2 2 3 1 3 1 2 Note: 1. 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.

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 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.
3. 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.
4. Systems/evolutions within each group are identified on the associated outline.
5. The shaded areas are not applicable to the categoryhier.

6.* 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.

7. 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.
8. 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.

NUREG-1021 1

ES-401 Vermont Yankee NRC Form ES-401-1 BWR SRO/RO Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 1 1

Emergency Procedures IPlan Knowledge of operational 295007 High Reactor Pressure / 3 X 2.4.20 4,0 76 implications of EOP warnings/cautions/notes Emergency Procedures IPlan: Knowledge of the parameters and logic used to assess the status of safety 295032 High Secondaty Containment Area Temp. I 5 functions including: 1 reactivity Control 2. Core Cooling X 2.4.21 77 and heat removal. 3. Reactor coolant system integrity 4.

Containment conditions. 5. Radioactivity release control.

(Hi secondarv containment area tern&.

Equipment Control Knowledge of SRO fuel handling 295023 RefuelingAcc Cooling Mode I 8 X 2.2.29 78 1

responsibilities.

Ability to determine and/or interpret the following as they EA2.03

+--Lt 295024 High Drywell Pressure I 5 X apply to HIGH DRYWELL PRESSURE: Suppression pool 3.8 79 level.............................__.

295028 High Drywell Temperature I 5 Ability to determine and/or interpret the following as they apply to HIGH DRYWELL TEMPERATURE : Reactor water level I 3.9 80

-

Ability to determine and/or interpret the following as they 295030 Low Suppression Pool Water Level / 5 X EA2.01 apply to LOW SUPPRESSION POOL WATER LEVEL : 81 Sumression WOIlevel................................

-

295031 Reactor Low Water Level I 2 X 2.4.7 Emergency ProceduredPlan Knowledge of Event based EOP mitigation strategies I 3,8 82

-

Knowledge of the reasons for the following responses as 295001 Partial or Complete Loss of Forced Core Flow they apply to PARTIAL OR COMPLETE LOSS OF AK3.02 3'7 39 Circulation / 1 & 4 FORCED CORE FLOW CIRCULATION : Reactor power response................................

Ability to operate and/or monitor the following as they 295003 Partial or Complete Loss of AC I 6 Ix AA1.02 apply to PARTIAL OR COMPLETE LOSS OF A.C.

POWER : Emergency generators..................................

4.2 40

+II

-

Ability to operate and/or monitor the following as they apply to PARTIAL OR COMPLETE LOSS OF D.C.

295004 Partial or Total Loss of DC Pwr / 6 AA1.02 3.8 41 POWER : Systems necessary to assure safe plant shutdown......... .......... .......

~ -

Ability to determine and/or interpret the following as they 295005 Main Turbine Generator Trip I 3 AA2.03 apply to MAIN TURBINE GENERATOR TRIP : Turbine 3.1 42 valve position................................

Emergency Procedures/Plan: Knowledge of Abnormal 295002 Loss of Main Condenser Vacuum / 8 2.4.1 1 3.4 43 Condition Procedures

-

NUREG-1021 2

ES-401 Vermont Yankee NRC Form ES-401-1 BWR SRO/RO Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 1 Knowledge of the reasons for the following responses as 295016 Control Room Abandonment / 7 X AK3.03 they apply to CONTROL ROOM ABANDONMENT : 3.5 44 Disabling Control Room Controls..

~ ~ -

Knowledge of the operational implications of the following concepts as they apply to PARTIAL OR COMPLETE 295018 Partial or Total Loss of CCW / 8 X AK1.01 3.5 45 LOSS OF COMPONENT COOLING WATER and the following: Effects on components/system operations

-

Knowledge of the interrelations between PARTIAL OR 295019 Partial or Total Loss of Inst. Air / 8 X AK2.08 COMPLETE LOSS OF INSTRUMENT AIR and the 2.8 46 1 following: Plant ventilation.

Knowledge of the interrelations between LOSS OF 295021 Loss of Shutdown Cooling / 4 X AK2.03 SHUTDOWN COOLING and the following: 3.6 47 RHWshutdowncooling

-

Knowledge of the reasons for the following responses as 295023 Refueling Acc Cooling Mode / 8 X AK3.02 they apply to REFUELING ACCIDENTS : Interlocks 3.4 48 associated with fuel handling equipment....

-

Emergency Procedures/ Plan: Knowledge of system 295024 High Drywell Pressure / 5 X 2.4.2 setpoints/interlocks and automatic actions associated with 3.9 49 EOP entry condtions.

-

295025 High Reactor Pressure/ 3 X 2.1.32 I Conduct of Operations: Ability to explain and apply all svstem limits and Drecautions.

3.4 50 Knowledgeof the interrelations between SUPPRESSION 295026 Suppression Pool High Water Temp. / 5 X EK2.03 POOL HIGH WATER TEMPERATURE: Suppression 3.2 Chamber Pressure.

Ability to operate and/or monitor the following as they apply to High Secondary Containment Sump/Area Water x Level: Affected systems so as to isolate damaged 3.5 295028 High Drywell Temperature / 5 X 3.6 temperature and the following: Drywell ventilation.

Ability to operate and/or monitor the following as they 295030 Low Suppression Pool Water Level / 5 X EA1.05 apply to LOW SUPPRESSION POOL WATER LEVEL : 3.5 HPCl.........................

I l l Ability to interpret and/or determine the following as they 295031 Reactor Low Water Level / 2 X EA2.04 apply to REACTOR LOW WATER LEVEL Adequate Core 4.6 Cooling.

NUR EG-1021 3

ES-401 Vermont Yankee NRC Form ES-401-1 BWR SRO/RO Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 1 Knowledge of the reasons for the following responses as 295037 SCRAM Condition Present and Power Above they apply to SCRAM CONDITION PRESENT AND EK3.03 4.1 56 APRM Downscale or Unknown / 1 REACTOR POWER ABOVE APRM DOWNSCALEOR UNKNOWN : Lowering Reactor Water Level I

1 295038 High Off-site Release Rate / 9 EK1.02 Knowledge of the operational implications of the following concepts as they apply to HIGH OFF-SITE RELEASE RATE : Protection of the general public.............................

4.2 57 Knowledge of the operations applications of the following 600000 Plant Fire On-site/ 8 AK1.02 concepts as they apply to PLANT FIRE ON SITE: Fire 2.9 58 Fighting I WA Category Point Totals: 2/3 Group Point Total: I 20l7 NUREG-102 1 4

ES-401 Vermont-Yankee NRC Form ES-401-1 BWR SRO/RO Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 2 295010 High Drywell Pressure I5

-

X 2.4.6 I Emergency ProceduresI Plan Knowledge of symptom based EOP mitiaatiin strateaies I 3.4 I 83 I 295012 High DrywellTemperature / 5 X 2.1.9 I Conduct of Operations: Abillty to direct personnel activities insidethe control room I 4.0 I I Ability to interpret or determine the following as they apply 295003 Partial or Complete loss of AC Power / 6 X AA2.04 to COMPLETE OR PARTIAL LOSS OF AC POWER:

Svstetn Lineum 295019 Partial or Total Loss of Inst. Air I 8 X AK2.03 Knowledge of the interrelations between PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR and the following: Reactor Feedwater.

3.2 I 1 59 295010 High Drywell Pressure 1 5 X 2.4.18 I Emergency Procedures I Plan Knowledge of the specific bases for EOPs 2,7 6o Ability to determine and/or interpret the following as they 295014 Inadvertent Reactivity Addition I 1 X AA2.04 apply to INADVERTENT REACTIVITY ADDITION: 4.1 61 Violation of Fuel Thermal Limits.

-

Ability to operate andor monitor the following as they 295012 High DrywellTemperature 1 5 X AA1.01 apply to HIGH DRYWELL TEMPERATURE : Drywell 3.5 62 ventilation svstem............................

295015 Incomplete SCRAM I 1 X AA1.01 Ability to operate and/or monitor the following as they apply to INCOMPLETESCRAM: CRD Hydraulics...................

3.8 1 1 63 Knowledge of the operational implications of the following 295033 High Secondary ContainmentArea Radiation concepts as they apply to HIGH SECONDARY X EK1.03 3.9 Levels I 9 CONTAINMENT AREA RADIATION LEVELS : Radiation releases....................................

Knowledge of the interrelations between SECONDARY 295035 Secondary Containment High Differential X EK2.04 CONTAINMENT HIGH DIFFERENTIALPRESSURE and 3.3 Pressure I 5

- the following: Blowout Panels/Plant SDecific..................

WA Category Point Total: 112 1 2 0 2 111 Group Point Total: 713

~

NUREG-1021 5

ES-401 Vermont Yankee NRC Form ES-401-1 BWR SRO/RO Written Examination Outline Plant Systems - Tier 2 Group 1 I

System #/Name 211OooSLC 206000 HPCl 218000 ADS 262001 AC Electrical Distribution

~~~

204000 RWCU 203000 RHWLPCI: Injection Mode 205000 Shutdown Cooling 205000 Shutdown Cooling 206000 HPCl NUREG-1021 6

206000 HPCl X A3.03 PRESSURE COOLANT INJECTION SYSTEM: System 3.9 5 Lineup Knowledge of electrical power supplies to the following:

209001 LPCS X 2.9 6 Initiation logic 223001 Primary Containment Knowledge of the purpose and function of major system System and Auxiliaries X 2.1.28 components and controls. 3.2 7 Ability to (a) predict the impacts of the following on the STANDBY LIQUID CONTROL SYSTEM ; and (b) based 21 1000 SLC X A2.03 on those predictions, use procedures to correct, control, 3.2 8 or mitigate the consequencesof those abnormal conditions or operations: AC Power Failures Knowledge of the effect that a loss or malfunction of the 212000 RPS X K6.01 following will have on the REACTOR PROTECTION 3.6 9 SYSTEM: AC Electrical Distribution Ability to manually operate and/or monitor in the control 215003 IRM A4,07 3.6 10 room: Verification of proper functioning/ operability Knowledge of electrical power supplies to the following:

215004 Source Range Monitor X 2.6 11 SRM channels/detectors Knowledge of AVERAGE POWER RANGE MONITOWLOCAL POWER RANGE MONITOR 215005 APRM I LPRM X K4.01 SYSTEM design feature(s) and/or interlocks which 3.7 12 provide for the following: Rod withdrawal blocks Ability to predict and/or monitor changes in parameters associated with operating the REACTOR CORE 217000 RClC X A l o,l ISOLATION COOLING SYSTEM (RCIC) controls 3.7 13 including: RClC flow Knowledge of operational implications of the following 218000 ADS X K5.01 concepts as they apply to AUTOMATIC 3.8 14 DEPRESSURIZATION SYSTEM : ADS Logic Operation Ability to monitor automatic operations of the PRIMARY 223002 PCIS/Nuclear Steam CONTAINMENT ISOLATION SYSTEM/NUCLEAR Supply Shutoff -

Knowledge of the effect that a loss or malfunction of the 239002 SRVs X K3.02 reliefkafety valves will have on the following: Reactor 4.2 16 Over pressurization Knowledge of the physical connections and/or cause-259002 Reactor Water Level X K1.03 effect relationships between REACTOR WATER LEVEL 3.8 17 Control CONTROL and the following: Reactor water level NUREG-1021 7

Control recirculation pumps: Plant-Specific Emergency Procedures/ Plan Knowledge symptom 261000 SGTS 3.1 19 X 2'4'6 based EOP mitigation strategies.

Knowledgeof the physical connections and/or cause-effect relationships between A.C. ELECTRICAL 262001 AC Electrical Distribution 3.1 20 X K1'04 DISTRIBUTIONand the following: Uninterruptible power supply Ability to manually operate and/or monitor in the control 262002 UPS (ACIDC) X A4.01 room: Transfer from alternative source to preferred 2.8 21 source Knowledge of the physical connections and/or cause-263000 DC Electrical Distribution X K1.04 effect relationships between D.C. ELECTRICAL 2.6 22 DISTRIBUTIONand the following: Ground detection Knowledge of EMERGENCY GENERATORS 264000 EDGs X K4.05 (DIESEUJET)design feature(s) and/or interlocks which 3.2 23 provide for the following: load shedding and sequencing Knowledge of the effect that a loss or malfunction of the 300000 Instrument Air X K3.02 INSTRUMENTAIR SYSTEM will have on the following: 3.3 24 Systems having pneumatic valves and controls 400000 Component Cooling Ability to manually operate and/or monitor in the control A4,0, 3.1 25 Water room: CCW indications and control Ability to predict and / or monitor changes in parameters 400000 Component Cooling X Al.04 associated with operating the CCWS controls including: 2.8 26 Water Surge Tank Level KIA Category Point Totals: 2 / 2 3 2 3 3 2 2 3 1 / 3 2 3 Group Point Total: 2615 NUREG-1021 8

BWR SRO/RO Written Examination Outline Plant Systems - Tier 2 Group 2 I I System #/Name IG 202001 Recirculation 215005 APRMILPRM 256000 Reactor Condensate directives affect plant and system conditions.

Knowledge of the effect that a loss or malfunction of the ROD WORTH MINIMIZER SYSTEM (RWM) (PLANT 201006 RWM X K3.01 3,2 27 SPECIFIC) will have on following: Reactor manual control system: P-Spec(Not-BWR6)

I Knowledge of electrical power supplies to the following:

202001 Recirculation X K2.03 Recirc System Valves 2,7 28 I

Knowledge of the operational implications of the 204000 RWCU X K5.04 following concepts as they apply to REACTOR WATER 2.7 29 CLEANUP SYSTEM : Heat exchanger operation Ability to (a) predict the impacts of the following on the ROD POSITION INFORMATIONSYSTEM ; and (b) 214000 RPlS X A2.02 based on those predictions, use procedures to correct, 3.6 30 control, or mitigate the consequencesof those abnormal conditions or operations: Reactor SCRAM Ability to monitor automatic operations of the 219000 RHWLPCI: Torus/Pool X A3.01 RHWLPCI: Torus suppression pool cooling mode 3.3 31 Cooling Mode including: valve operation 226001 RHWLPCI: CTMT Spray Ability to manually operate and /or monitor in the Mode A4.07 control room: valve logic reset/bypass/override 3.5 32 Knowledge of the effect that a loss or malfunction of the 239001 Main and Reheat Steam X K6.01 following will have on the MAIN AND REHEAT STEAM: 3.1 33 Electrical Power 241000 Reactornurbine Pressure Ability to manually operate and/or monitor in the control A4.06 room: Bypass valves operation 3.9 34 Regulator Knowledge of the physical connections and/or cause 271000 Off-gas X K1.06 effect relationships between OFFGAS SYSTEM and 2.8 35 the following: Main Steam System.

NU REG- 1021 9

Facility: Vermont Yankee NRC Date of Exam: 412007 Topic l7A-L-E I I SRO-Only I

I I I I Ability to apply technical specifications for a 4.0 94 2.1.20 Ability to execute procedure steps. 4.3 66

1. 2.1.3 Knowledge of shift turnover practices 3.0 67 Conduct of Knowledge of the purpose andlor function of Operations 2.1 .28 3.8 68 major system components and controls.

I Subtotal I I

1 3 1 1 2 I

I 2.2.25 Knowledge of bases in TS for LCOs and safety Limits 3.7 95 I-2.2.24 Ability to analyze the effect of maintenance 3.8 96 t-activities on LCO status.

Knowledge of the refueling administrative 2.2.26 3.7 97

2. requirements.

Equipment Control Ability to manipulate the console controls as 2.2.2 required to operate the facility between 4.0 69 shutdown and designated power levels.

Knowledge of new and spent fuel movement 2.2.28 2.6 70

- procedures.

Subtotal

- Knowledge of radiation exposure limits and 2 2 E

2.3.4 contamination control1 including permissible 3.1 98 levels in excess of those authorized.

3. 2.3.2 Knowledge of facility ALARA program. 2.5 71 Radiation Control Ability to perform procedures to reduce 2.3.10 excessive levels of radiation and guard against 2.9 72 personnel exposure.

Subtotal 2 1 I 2.4.25 I Knowledge of fire protection procedures. 3.4 99 Knowledge of the emergency action level 4.1 100 4.

Emergency Procedures I Plan I 24.1 0 1 Knowledge Ofannunciafor response procedures.

~-

3.0- I 75 Subtotal 2 Tier 3 Point Total

- 7 NUREG-1021 11

Tier / Randomly Reason for Rejection Group Selected WA 295019 (Q.#46) AK2.08, randomly selected - original selection did not apply to 111 K2.13 dant.

295031 (Q. #82) Impossible to meet KA Topic requirement at SRO level.

1:1 2 1 28 Randomly reselected G2.1.20 for APE. G2.4.7 randomly selected.

295028 (Q. #53). EK2.04 randomly selected - original selection not applicable a1 111 2.4.30 RO level at this Dlant. Also e 2.5 ImP.

295005 211 (Q.#3) K3.05 randomly selected - double jeopardy with Q.# 47 K3.02 259002 (Q. #17) K1.03 randomly selected, original selection does not apply to 211 K2.02 the plant.

1 3 I 2.2.3 I (Q.#69) 2.2.2 randomly selected, original selection is for multi unit plant 202001 ((2.28) K2.03 randomly selected, original selection does not apply to 212 K2.04 plant 2009002 (Q.#7) 223001 2.2.28 randomly selected, original selection does not 211 K6.01 apply to the plant 29501 1 (Q.#61) 295014 AA2.04 randomly selected, original selection does not AA2.01 apply to plant.

207000 (Q.#5) 206000 A3.03 randomly selected, original selection does not 211 A3.02 amlv to Dlant 295027 (Q.#52) 295032 randomly selected, original selection does not apply to 111 EAl.01 plant.

29501 1 (Q.#84) 295010 2.4.6 randomly selected, original selection does not 112 2.4.31 apply to plant.

206000 (Q.# 88) 2.1.1 1 randomly selected, original selection was double 211 2.05 ieoDardv with Q.#41.

(Q.#68) 2.1.28 randomly selected, original selection was not an RO 3 G2.1 level topic at VY.

m8 n i n 215001 (Q.#92) 215005 G2.1.12 randomly reselected - an operationally valid L1 L A2.08 question at the SRO level could not be written for original selection.

4 I 4 295007 (Q.#76) 2.4.20 randomly reselected, original selection was an RO level I I I 2.4.50 topic 295020 (Q.#86) 295003 AA2.04 randomly reselected, an operationally valid 111 AA2.02 question at the SRO level could not be written for original selection 295009 (Q.#59) 295019 AK2.03 randomly reselected due to double jeopardy AK1.03 with another reactor level topic KA.

4 I 1 295006 (Q.#43) 295002 2.4.1 1 randomly reselected, operationally valid I i I 2.1.28 discriminating question could not be written for original selection.

n ~n 290002 (Q.#38) K4.01 randomly reselected. An operationally valid, L1 L K4.02 discrim-inatingquestion-could not be written for original selection (Q.#95)2.2.25 randomly reselected. An operationally valid SRO level 3 G2'1 '4 question could not be written for the original selection (Q.#96) 2.2.1 2 randomly reselected. An operationally valid SRO level 3 G2'2*20 question could not be written for the original selection NUREG-1021 12

ES-301 Administrative Topics Out1ine Form ES-301-1 Facility: Vermont Yankee Date of Examination: 4 130107 Examination Level (circle one): RO / SRO Operating Test Number: 2007 NRC Type Describe activity to be performed Code*

D, S 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.

JPM: Prepare Switching and Tagging Order WA: 2.2.13 (3.6)

Knowledge of tagging and clearance procedures.

M, C 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 (I3 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 radiologicalrequirements for Inspection of RWCU valve V12-19A (CU-19A), 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 Type Describe activity to be performed Code*

N 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.

D JPM: Review Completed Surveillance and Take Conduct of Operations Action for Out of Spec Data WA: 2.1.12 (4.0)

Ability to apply technical specifications for a system.

N JPM: Review Switching and Tagging Order Equipment Control WA: 2.2.13 (3.8)

Knowledge of tagging and clearance procedures.

N JPM: Review and Approve Primary Containment Radiation Control Purge cumulative hours log WA: 2.3.9 (3.4)

Knowledge of the process for performing a containment purge D JPM: PAR Based on Plant Conditions (Shelter)

WA: 2.4.44 (4.0)

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 ( 3Ifor ROs; 4 I for SROs & RO retakes)

(N)ew or (M)odified from bank (> 1)

(P)revious 2 exams ( 1I; randomly selected)

(S)imulator 2007 NRC Examination Summary Description of Admin Tasks A . l .a The candidate will. This is a new JPM.

A . l .b The candidate will review a completed RHR system surveillance and take action for out of spec data. This is a bank JPM.

A.2 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.

A.3 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.

A.4 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 FaciIit y: Vermont Yankee Date of Examination: 4- 3 0 07 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* Safety System / JPM Title Function S-1 201006 Source Range Monitor N, A, S, L 7 Rx startup to criticality, high reactor period 5-2 261000 Standby Gas Treatment System p, s 9 Secure Standby Gas Treatment S-3 241000 Reactorflurbine Pressure Regulating System 3 Transfer Pressure Control From MPR to EPR S-4 262001 A.C. Electrical DistributionSystem 6 Energize Bus 4 From the Vernon Tie Line During a SBO S-5 217000 Reactor Core Isolation Cooling System 2 Respond to Automatic RClC Auto Controller Failure S-6 209001 Core Spray System 4 Perform Core Spray "A" Quarterly Full Flow Test S-7 223002 Primary Containment Isolation System 5 PClS Group V isolation failure S-8 201003 Control Rod and Drive Mechanism 1 In-Plant Systems@(3 for RO; 3 for SRO-I; 3 or 2 for SRO-U)

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.

I

  • Type Codes Criteria for RO I SRO-I I SRO-U I

(A)lternate path 4-6 I 4-6 12-3 (C)ontrol room (D)irect from bank ~11sais4 (E)mergency or abnormal in-plant 2 1 127 I 2 1 (L)ow-Power I Shutdown 2 1 I 2 7 121 (N)ew or (M)odified from bank including 1(A) 22122127 (P)revious 2 exams 5 3 I s 3 1 5 2 (randomly selected)

(WA 2 1I 2 7 I 2 7 2007 NRC Examination Summary Description of JPMs s-1 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.

s-2 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.

s-3 The candidate will transfer Pressure Control From MPR to EPR - Safety Function. This is a modified bank JPM.

s-4 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 DistributionSystem - ElectricalSafety 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.

s-5 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.

S-6 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 s-7 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.

P-3 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.

NUREG-1021 Revision 9

I AppendixD Scenario Outline Form ES-D-1 1

acil ity: VERMONT YANKEE Scenario No.: 1 Op Test No.: 2007 NRC
xaminers: Operators:

nitial Conditions: -At90% Dower for control rod Dattern adiustment.

a Power ascension required back to 100%.

rurnover: Perform weekly remote testing of Turbine Oil pumps IAW OP 4160.

2ritical Task:

Event Malf. No. Event Event Description No.

1 lType NIA N-ACRO Weekly remote testing of Turbine Oil pumps OP 4160.

N-CRS 2 NIA R-CRO Power ascension IAW OP 0105.

N-CRS 3 mfED-19B I-CRS Loss of Bus 89B (TS).

4 mfED-19B C-ACRO Loss of Circ Water Pump.

C-CRS 5 mfRD-01 A C-CRO CRD Pump A trips (ON).

C-CRS 6 rnfRD-051831 C-CRO Control Rod 18-31 drifts outward (OT).

100% C-CRS 7 RC04 I-ACRO Inadvertent HPCl initiation (TS).

I-CRS 8 mfED-17 M-ALL Loss of Offsite power 9 mfHP-04 C-ACRO HPCl Flow Controller Failure.

0% C-CRS 10 mf RR-0 1A M-ALL Recirc loop rupture (0.7% over 300 sec). HPCl trip.

mf HP-0 1 RPV-ED on low level.

11 mfCS-03A C-CRS CS-12A and CS-12B failure to auto open.

Preinsert mf CS-03 B C-ACRO Preinsert mfRH-07A RHR 27A failure to auto open.

  • (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Appendix D NUREG 1021 Revision 9

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 chlorinatethe 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 30-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 No.

I Malf. No. Event Type*

Event Description I N'A N-CRS N-ACRO R-CRO Place CW in Closed Cycle for chlorination.

Power Reduction IAW OP 0105.

N-CRS 3 FW-O9A C-CRS Feedwater regulating valve lockup (OT).

C-CRO 4 mfNM-05A I-CRO APRM A fails downscale (TS) 0% I-CRS 5 mfTC-04A C-ACRO EPR Oscillations (OT).

C-CRS 6 mf E D-05C C-ALL Loss of 480V Bus 8 (TS), Failure of SBGT A to auto start.

mfPC-11 A 7 mfAD-01 B C-ALL SRV-71B leak (OT) leads to Rx scram (100% over 600 sec).

8 mfRPOl B I-CRO Failure of manual scram. ARI required.

Preinsert I-CRS 9 mfRD-12A M-ALL 45% hydraulic ATWS (A). 55% hydraulic ATWS (B).

Preinsert mfRD-12B

' 10 mfSL-01 A C-CRO SLC pump A trips.

Preinsert mfSL-02B C-CRS B SLC squib valve fails to fire.

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Appendix D NUREG 1021 Revision 9

Appendix D Scenario Outline Form ES-D-1 I FaciIity: 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 0105 is complete thru Phase 2.C.

~~~~~

0 Perform Turbine Chest Warmup IAW OP 0105 Phase 2.D. Step 1.

0 Continue Reactor Startup (60 to 80 degree heat up rate).

Critical Task:

Event Malf. No. Event Event Description No. Type*

1 NIA N-ALL Perform Turbine Chest Warmup.

I NA I I1 :-cis II R-CRO N-CRS Pull rods to continue power ascension.

4 I NM04A mfRD-11A C-CRO IRM A faits upscale (TS).

CRD Flow control Valve fails closed (ON).

C-CRS 5 mfRD-02xxyy C-CRO Stuck Control Rod XX-YY (ON)

C-CRS

~

6 rfPP-06 C-ACRO Seismic event.

mfSW-14A C-CRS TBCCW APump Trip w/ TBCCW B Pump Failure to auto start mfSW-21 B 7 mfHP-11 C-ACRO RClC steam leak (TS).

mfHP-15 C-CRS RCIC fails to auto isolate.

8 rfPP-06 M-ALL Seismic aftershock. Group 1 isolation.

mfRR-18H 9 mfRP-01 C I-CRO Auto scram failure. Manual scram required.

mfRP-OlA I-CRS 10 mfRP-08A I-ACRO PClS Group Ill failure.

Preinseri I

mf RP-08B zC-lOpl I-CRS M-ALL 1 Torus leak at A RHR suction (50% over 900 secs).

PRV-ED on low Torus level.

  • (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Appendix D NUREG 1021 Revision 9

I Amendix D Scenario Outline Form ES-D-1 I Facility: VERMONT YANKEE Scenario No.: 4 Op Test No.: 2007 NRC Examiners: Operators:

1 Initial Conditions: 0 90% power for control rod pattern adjustment which has been completed.

Turnover: 0 Perform Speed Load Changer Bypass Test IAW OP 41 60.

0 Return to 100% power.

Critical Task:

Event Malf. No. Event Event Description No. Type*

NIA N-ACRO Speed Load Changer Bypass Test OP 41 60.

N-CRS 2 R-CRO Power ascension IAW OP-0105.

N-CRS 3 mfED-05Ca C-ALL LOSSof MCC-8A (TS).

4 mfCD-01 A C-CRS Condensate Pump A trips. Failure of RFP B to trip. (OT)(RP) rfFW-04 C-CRO Feedwater Pump B CD trip bypass switch in bypass.

5 I-CRO Steam flow summer fails upscale (OT) (100% over 60 sec).

I-CRS 6 mfRRQ1A C-CRS Recirc Leak (OT) (TS).

C-ACRO 7 mf MS-06 M-ALL MS line A rupture in drywell. (10% over 1200 secs).

8 mfPC1-06 C-CRS AC-6 and AC-6B fail to auto close.

mfPCl-6B C-ACRO

)ormal, (R)ei ctivity, (I)nstrument, (C)omponent, (M)ajor Appendix D NUREG 1021 Revision 9

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