ML110110512

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Draft - Outlines (Facility Letter Dated 8/24/10) Folder 2)
ML110110512
Person / Time
Site: Vermont Yankee Entergy icon.png
Issue date: 08/24/2010
From: Stupak K, Bengtson J
Entergy Nuclear Operations
To: D'Antonio J
Operations Branch I
Hansell S
Shared Package
ML102070016 List:
References
TAC U01769
Download: ML110110512 (34)
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Text

  • Entergy

-::=.

Entergy Nuclear Operations. Inc.

Z i rnDnt Yanke' Jon A. Bengtson Training Manager August 24,2010 U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406-1415

Subject:

Vermont Yankee Initial License Operator NRC Examination Outline

Dear Mr. D'Antonio,

In accordance with NUREG 1021, "Operator Licensing Examination Standards for Power Reactors" and the agreed exam submittal time line with the NRC and Vermont Yankee, enclosed with this letter is the Initial License Operator examination outline submittal consisting of the following:

  • Examination Outline Quality Checklist (Form ES-201-2)
  • Copy of Examination Security Agreement (Form ES-201-3)
  • Administrative Topics Outline (Forms (2) ES-301-1 (RO and SRO>>
  • Control Room/In-Plant Systems Outline (Forms (3) ES-301-2 (RO, SRO-I, and SRO-U>>
  • Simulator Scenario Quality Checklist (Form ES-301-4 (partially filled out to support current examination outline material>>
  • Generic Knowledge and Abilities Outline (Form ES-401-3)
  • Simulator outline forms (4) ES-D-1 for scenarios 1 through 4
  • Operating examination schedule and candidate layout
  • Attachment 1: Systematic Sampling Methodology Used All materials associated with this examination outline must be withheld from pubic disclosure until after the examinations are complete.

We look forward to any comments you have concerning the examination outline submittal. Upon your review, if you have any comments or questions, we will provide whatever additional information you need upon request.

Sincerely,

~

Jon A. Bengtson

!4~

Kevin Stupak (Training Manager) (Facility Reviewer)

ES-401 BWR Examination Outline Form ES-401-1 Facility: VERMONT YANKEE Date of Exam: NOVEMBER 29, 2010 RO KIA Category Points SRO-Only Points Tier Group K K K K K K A A A A G A2 G* Total 1 2 3 4 5 6 1 ," 3 4

  • Total
1. 1 2 3 4 5 4 2 20 3 7 II Emergency &

I m

Abnormal Plant 2 N/A 1 1 2 7 2 1 3 Evolutions Tier Totals 6 5 4 27 L.j 4 10 1 3 1 1 3 3 4 2 26 3 2 5 2.

Plant i 2 2 1 !2 1 2 1 2 0 1 0 0 12 o I 1 2 3 Systems

~Totals 4 3 3 5 5 2 3 :3 4 4 2 38 4 8 II

3. Generic Knowledge and Abilities 1 2 3 4 10 7 Categories 2 3 Note: 1. Ensure that at least two topics from every applicable KIA category are sampled within each tier of the RO and SRO-only outlines (i.e., E~xcept for one category in Tier 3 of the SRO-only outline, the "Tier Totals" in each KIA 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/evolutions that are not included on the outline should be added.

Refer to Section D.1.b of ES-401 for guidance regarding the elimination of inappropriate KIA 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 KlAs 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) KlAs in Tiers 1 and 2 shall be selected from Section 2 of the KIA Catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.1.b of ES 401 for the applicable KlAs.

8. On the following pages, enter the KIA numbers, a brief description of each topic, the topics' importance ratings (IRs) 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 KIA catalog, and enter the KIA numbers, descriptions, IRs, and point totals (#) on Form ES-401-3. Limit SRO selections to KlAs that are linked to 10 CFR 55.43.

NUREG-1021 , Rev.g, Supplement 1 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 ElAPE # I Name Safety Function I G I 1-<.1 II-<.2TK3TA1 I A2 I Number I KIA Topic(s) limp. I Q# I 295001 Partial or Loss of Forced Core Flow alarm setpoints and operate 4.0 81 Circulation J 1 & 4 irlAntifiArl in the alarm response manual 295004 Partial or Total Loss of DC Pwr! 6 3.3 77 of each 295005 Main Turbine Generator Trip J 3 4.3 79 to interpret control room indications to status and operation of a and 295018 Pal1ial or Total Loss of CCW 18 4.4 82 operator actions and affect plant and system conditions Ability to determine and/or interpret the following as they 295031 Reactor Low Water Levell 2 4.2 76 apply to REACTOR LOW WATER LEVEL: Reactor power Ability to determine and/or interpret the following as they 295037 SCRAM Condition Present and Reactor Power apply to SCRAM CONDITION PRESENT AND 4.2 78 Above APRM Downscale or Unknown I i REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN: Reactor water level Ability to determine andlor interpret the followinq as they 295038 High Off-site Release Rate / 9 apply to HIGH OFF-SITE RELEASE RATE: Source of off 45 80 site release Ability to determine andlor interpret the following as they 295001 Partial or Complete Loss of Forced Core Flow apply to PARTIAL OR COMPLETE LOSS OF FORCED 3.1 20 Circulation 11 & 4 CORE FLOW CIRCULATION: Jet pump operability: Not BWR-1&2 295003 Partial or Complete Loss of AC 1 6 x Ability to interpret and execute Knowledge of the reasons for the following responses as 295004 Partial or Total Loss of DC Pwr! 6 x AK3.01 they apply to PARTIAL OR COMPLETE LOSS OF D.C. 3.4 15 POWER: load shedding: Plant specific Ability to operate andlor monitor the following as they 295005 Main Turbine Generator Trip / 3 AA1.04 apply to MAIN TURBINE GENERATOR TRIP: Main 2.7 18 generator controls Ability to operate andlor monitor the followino as 295006 SCRAM / 1 AA1.06 8 apply to SCRAM: CRD hydraulic system Ability to operate andlor monitor the following as they 295016 Control Room Abandonment / 7 x AA1.08 apply to CONTROL ROOM ABANDONMENT: Reactor 4.0 2 pressure NUREG-1021, Rev.g, Supplement 1 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 E/APE # / Name Safety Function [Gl K1 I K2 I K3 I A1 I A2 I Number I KIA Topic(s) limp. I Q# I Knowledge of the reasons for the following responses as they apply to PARTIAL OR COMPLETE LOSS OF 295018 Partial or Total Loss of CCW / 8 X AK3.03 3.1 5 COMPONENT COOLING WATER: Securing individual components (prevent equipment damage)

Knowledge of the interrelations between PARTIAL OR 295019 Partial or Total Loss of Inst. Air /8 X AK2.05 COMPLETE LOSS OF INSTRUMENT AIR and the 3.4 12 following: Main steam system Knowledge of the interrelations between LOSS OF 295021 Loss of Shutdown Cooling / 4 X AK2.07 SHUTDOWN COOLING and the following: Reactor 3.1 17 recirculation I Knowledge of the interrelations between REFUELING 295023 Refueling Acc / 8 X AK2.07 ACCIDENTS and the following: Standby gas 3.6 9 treatmenVFRVS Ability to interpret control room indications to verify the status and operation of a system, and understand how 295024 High Drywell Pressure / 5 X 2.2.44 4.2 19 operator actions and directives affect plant and system conditions Ability to operate and/or monitor the following as they 295025 High Reactor Pressure I 3 X EA1.03 apply to H!GH REACTOR PRESSURE: Safety/relief 4.4 10 valve: Plant specific Knowledge of the reasons for the following responses as 295026 Suppression Pool High Water Temp. / 5 X EK3.01 they apply to SUPPRESSION POOL HIGH WATER 3.8 1 TEMPERATURE: Emergency/normal depressurization Ability to operate and/or monitor the following as they 295028 High Drywell Temperature / 5 X EA1.03 apply to HIGH DRYWELL TEMPERATURE: Drywell 3.9 6 cooling system Knowledge of the reasons for the following responses as 295030 Low Suppression Pool Wtr Lvi / 5 X EK3.03 they apply to LOW SUPPRESSION POOL WATER 3.6 11 LEVEL: RCIC operation Ability to determine and/or interpret the following as they 295031 Reactor Low Water Level / 2 X EA2.04 apply to REACTOR LOW WATER LEVEL: Adequate core 4.6 13 cooling Knowledge of the operational implications of the following concepts as they apply to SCRAM CONDITION 295037 SCRAM Condition Present and Reactor Power X EK1.01 PRESENT AND REACTOR POWER ABOVE APRM 4.1 3 Above APRM Downscale or Unknown / 1 DOWNSCALE OR UNKNOWN: Reactor pressure effects on reactor power Knowledge of the operational implications of the following 295038 High Off-site Release Rate / 9 X EK1.02 concepts as they apply to HIGH OFF-SITE RELEASE 4.2 7 RATE: Protection of the general public NUREG-1021, Rev.g, Supplement 1 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 ElAPE # I Name Safety Function I G I K1 I K2 I K3 I A 1 I A2 I Number I KIA Topic(s) limp. I Q# I Ability to determine and interpret the following as they apply to PLANT FIRE ON SITE: Vital equipment and 600000 Plant Fire On Site / 8 X AA2.16 3.0 16 control systems to be maintained and operated during a fire Ability to determine and/or interpret the following as they 700000 Generator Voltage and Electric Grid apply to GENERATOR VOLTAGE AND ELECTRIC GRID X AA2.03 3.5 14 Disturbances / 6 DISTURBANCES: Generator current outside the capability curve KIA Category Point Totals: 213 2 3 4 5 4/4 Group Point Total: I 200 NUREG-1021, Rev.g, Supplement 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 ElAPE # / Name Safety Function [G] K1 I K2 I K3 I A 1 I A2 I Number I KIA Topic(s) limp. I Q# I Ability to determine and/or interpret the following as they 295008 High Reactor Water Levell 2 apply to HIGH REACTOR WATER LEVEL: Reactor water 3.9 I 85 level Ability to determine and/or interpret the following as they 295029 High Suppression Pool Wtr Lvi / 5 apply to HIGH SUPPRESSION POOL WATER LEVEL: 3.5 I 83 Drywellicontainment water level 295033 High Secondary Containment Area Radiation Knowledge of EOP mitigation strategies 4.7 84 Levels! 9 Knowledge of the operational implications of the following 295007 High Reactor Pressure / 3 X I AK1.03 I concepts as they apply to HIGH REACTOR PRESSURE: 3.8 27 Pressure effects on reactor power I Knowledge of the interrelations between HIGH DRYWELL 295012 High Drywell Temperature / 5 X I AK2.01 TEMPERATURE and the following: Drywell ventilation 3.4 I 24 Ability to locate control room switches, controls, and 295013 High Suppression Pool Temp. /5 X 2.1.31 I indications, and to determine that they correctly reflect the 4.6 I 22 desired plant lineup 295014 Inadvertent Reactivity Addition / 1 X 2.1.23 I Ability to perform specific system and integrated plant 4.3 26 procedures during all modes of plant operation 295015 Incomplete SCRAM / 1 X AA2.02 I Ability to determine and/or interpret the following as they 4.1 23 apply to INCOMPLETE SCRAM: Control rod position 295022 Loss of CRD Pumps / 1 X I AA1.02 I Ability to operate and/or monitor the following as they 3.6 21 apply to LOSS OF CRD PUMPS: RPS Knowledge of the reasons for the following responses as 295032 High Secondary Containment Area Temperature / 5 X I EK3.01 I they apply to HIGH SECONDARY CONTAINMENT AREA 3.5 25 TEMPERATURE: Emergency/normal depressurization KIA Category Point Total: 211 1t? I Group Point Total: I 7/3 NUREG-1021, Rev.g, Supplement 1 5

ES-401 Vermont Yankee NRC Form ES-401-1 BWR SRO/RO Written Examination Outline Plant Systems - Tier 2 Group 1 System #/Name K1 I K2 I K3 I K4 I K5 I K6 I A 1 KIA Topics Imp I Q#

Ability to explain and apply system limits and 206000 HPCI 4.0 90 precautions Ability to (a) predict the impacts of the following on the SOURCE RANGE MONITOR (SRM) SYSTEM; and (b) based on those predictions, use procedures to correct, 215004 Source Range Monitor 3.5 89 control, or mitigate the consequences of those abnormal conditions or operations: Faulty or erratic 218000 ADS 4.4 87 261000 SGTS 3.2 88 262002 UPS (AC/DC) and (b) based on those predictions, use procedures to I 2.8 I 86 correct, control, or mitigate the consequ abnormai conditions or 0 erations: Under 203000 RHA/LPCI: Injection Knowledge of electrical power supplies to the following:

3.5 49 Mode Pumps.

203000 RHA/LPCI: Injection Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits. 3.2 45 Mode Knowledge of the operational implications of the I following concepts as they apply to SHUTDOWN 205000 Shutdown Cooling I X I K5.02 COOLING SYSTEM (RHR SHUTDOWN COOLING I 2.8 I 32 MODE): Valve operation 205000 Shutdown Cooling X A4.07 I Ability to manually operate and/or monitor in the control I 37 48 room: Reactor temperatures (moderator, vessel, flange) .

X K2.01 I Knowledge of electrical power supplies to the following:

3.2 46 206000 HPCI System valves BWR-2,3,4 Ability to monitor automatic operations of the LOW 209001 LPCS I X I A3.06 I PRESSURE CORE SPRAY SYSTEM including: Lights 3.6 28 and alarms 211000 SLC X 2.4.11 Knowledge of abnormal condition procedures 4.0 30 Ability to monitor automatic operations of the 211000 SLC X A3.05 STANDBY LIQUID CONTROL SYSTEM including: 4.1 52 Flow indication.

NUREG-1021, Rev.g, Supplement 1 6

ES-401 Vermont Yankee NRC Form ES-401-1 BWR SROIRO Written Examination Outline Plant Systems - Tier 2 Group 1 System #/Name G K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 Number KIA Topics Imp Q#

Knowledge of REACTOR PROTECTION SYSTEM 212000 RPS X K4.10 design feature(s) and/or interlocks which provide for the 3.3 37 following: Individual rod SCRAM testing Knowledge of the operational implications of the 212000 RPS X K5.02 following concepts as they apply to REACTOR 3.3 44 PROTECTION SYSTEM: Specific logic arrangements.

Knowledge of INTERMEDIATE RANGE MONITOR 2150031RM X K4.01 (IRM) SYSTEM design feature(s) and/or interlocks 3.7 41 which provide for the following: Rod withdrawal blocks:

Ability to monitor automatic operations of the SOURCE 215004 Source Range Monitor X A3.02 RANGE MONITOR (SRM) SYSTEM including: 3.4 38 Annunicator and alarm signals.

Knowledge of the effect that a loss or malfunction of the following will have on the AVERAGE POWER RANGE 215005 APRM / LPRM X K6.01 3.7 40 MONITOR/LOCAL POWER RANGE MONITOR SYSTEM:RPS Ability to predict and/or monitor changes in parameters associated with operating the REACTOR CORE 217000 RCIC X A1.05 3.7 29 ISOLATION COOLING SYSTEM (RCIC) controls including: ACIC turbine speed Knowledge of the physical connections and/or cause/effect relationships between REACTOR CORE 217000 RCIC X K1.02 3.5 50 ISOLATION COOLING SYSTEM (RCIC) and the following: Nuclear boiler system.

Ability to manually operate and/or monitor in the control 218000 ADS X A4.02 4.2 42 room: ADS logic initiation Ability to (a) predict the impacts of the following on the PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF; and 223002 PCIS/Nuclear Steam X A2.01 (b) based on those predictions, use procedures to 3.2 47 Supply Shutoff correct, control, or mitigate the consequences of those abnormal conditions or operations: AC electrical distribution failures.

Knowledge of the effect that a loss or malfunction of the 239002 SRVs X K3.03 RELIEF/SAFETY VALVES will have on following: 4.3 51 Ability to rapidly depressurize the reactor.

Ability to manually operate and/or monitor in the control 259002 Reactor Water Level X A4.02 room: All individual component controllers in the 3.8 36 Control automatic mode NUREG-1021, Rev.g, Supplement 1 7

~~

I ES-401 Vermont Yankee NRC Form ES-401-1 BWR SRO/RO Written Examination Outline I Plant Systems - Tier 2 Group 1 System #/Name G K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 Number KIA Topics Imp Q#

Ability to (a) predict the impacts of the following on the STANDBY GAS TREATMENT SYSTEM; and (b) based on those predictions, use procedures to correct, 261000SGTS X A2.03 2.9 33 control, or mitigate the consequences of those abnormal conditions or operations: High train temfll!ature Ability to (a) predict the impacts of the following on the A.C. ELECTRICAL DISTRIBUTION; and (b) based on 262001 AC Electrical Distribution X A2.11 those predictions, use procedures to correct, control, or 3.2 34 mitigate the consequences of those abnormal conditions or operations: Degraded system voltages .

Knowledge of the physical connections and/or cause/effect relationships between 262002 UPS (AC/DC) X K1.01 UNINTERRUPTABLE POWER SUPPLY (A.C.lD.C.) 2.8 53 and the following: Feedwater level control: Plant specific Knowledge of the operational implications of the following concepts as they apply to D.C. ELECTRICAL 263000 DC Electrical Distribution X K5.01 2.6 39 DISTRIBUTION: Hydrogen generation during battery charging Knowledge of EMERGENCY GENEP"4.TORS 264000 EDGs X K4.08 (DIESEUJET) design feature(s) and/or interlocks which 3.8 43 provide for the following: Automatic startup Ability to manually operate and f or monitor in the 300000 Instrument Air X A4.01 2.6 35 control room: Pressure gages Knowledge of CCWS design feature(s) and or 400000 Component Cooling X K4.01 interlocks which provide for the following: Automatic 3.4 31 Water start of standby pump.

KIA Category POint Totals: 212 2 2 1 4 3 1 1 3/3 3 4 Group Point Total:

J ~~~~-

2615 NUREG-1021, Rev.g, Supplement 1 8

ES-401 Vermont Yankee NRC Form ES-401-1 BWR SRO/RO Written Examination Outline Plant Systems - Tier 2 Group 2 System #/Name G K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 Number KIA Topics Imp Q#

Ability to perform specific system and integrated plant 216000 Nuclear Boiler Ins!. 4.4 93

'~,

procedures during all modes of plant operation ffi 233000 Fuel Pool Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits, 4,2 91 Cooling/Cleanup Ability to (al predict the impacts of the following on the FIRE PROTECTION SYSTEM; and (b) based on those 286000 Fire Protection predictions, use procedures to correct, control, or 3.2 92 mitigate the consequences of those abnormal conditions or operations: Pump trips: Plant specific Knowledge of the effect that a loss or malfunction of 201001 CRD Hydraulic X K6.03 the following will have on the CONTROL ROD DRIVE 3.0 59 HYDRAULIC System: Plant air systems Ability to predict and/or monitor changes in parameters 201003 Control Rod and Drive associated with operating the CONTROL ROD AND X A1.02 2.8 60 Mechanism DRIVE MECHANISM controls including: CRD Drive Pressure.

Ability to predict and/or monitor changes in parameters associated with operating the ROD WORTH 201006 RWM X A1.02 MINIMIZER SYSTEM (RWM) (PLANT SPECIFIC) 3.4 65 controls Including: Status of control rod movement blocks: Plant specific (not BWR-6)

Knowledge of the effect that a loss or malfunction of the 202002 Recirculation Flow RECIRCULATION FLOW CONTROL SYSTEM will X K3.04 2.9 56 Control have on following: ReactorlTurbine pressure regulation system.

Ability to monitor automatic operations of the ROD 215002 RBM X A3.01 BLOCK MONITOR SYSTEM including: Four rod 3.1 55 display Knowledge of the physical connections and/or 219000 RHA/LPCI: Torus/Pool cause/effect relationships between RHA/LPCI:

X K1.01 3.9 57 Cooling Mode TORUS/SUPPRESSION POOL COOLING MODE and the following: Suppression Pool.

Knowledge of the operational implications of the following concepts as they apply to PRIMARY 223001 Primary CTMT and Aux. X K5.01 3.1 64 CONTAINMENT SYSTEM AND AUXILIARIES:

Vacuum breaker/relief valve operation Knowledge of the effect that a loss or malfunction of the 239001 Main and Reheat Steam X K3.15 MAIN AND REHEAT STEAM SYSTEM will have on 3.5 61 following: Reactor water level control Knowledge of electrical power supplies to the following:

259001 Reactor Feedwater X K2.01 3.3 63 Reactor feedwater pump(s): Motor-driven-only NUREG-1021, Rev.9, Supplement 1 9

ES-401 Vermont Yankee NRC Form ES-401-1 BWR SRO/RO Written Examination Outline Plant Systems - Tier 2 Group 2 System #/Name G K1 K2 K3 K4 K5 K6 A1 A2. A3 A4 Number KIA Topics Imp Q#

Knowledge of OFFGAS SYSTEM design feature(s) 271000 Offgas X K4.09 and/or interlocks which provide for the following: 2.8 54 Filtration of radioactive particulate Knowledge of the operational implications of the following concepts as they apply to RADIATION 272000 Radiation Monitoring X K5.01 3.2 62 MONITORING SYSTEM: Hydrogen injection operation's effect on process radiation indications Knowledge of the physical connections and/or 290001 Secondary CTMT X K1.09 cause/effect relationships between SECONDARY 2.9 58 CONTAINMENT and the following: Plant air systems KIA Category Point Totals: 0/2 2 1 2 1 2 1 2 0/1 1 0 Group Point Total: I 1213 I NUREG-1021, Rev.g, Supplement 1 10

ES-401 Generic Knowledge and Abilities Outline (Tier3) Form ES-401-3 Facility: VERMONT YANKEE Date of Exam: NOVEMBER 29, 2010

~

Category KJA# Topic RO I~

Knowledge of refueling administrative requirements.

2.1.40 3.9 Ability to use procedures to determine the effects on reactivity of 2.1.43 plant changes. such as reactor coolant system temperature, 4.3 1.

Conduct secondary plant, fuel depletion, etc.

~

of Operations Ability to perform specific system and integrated plant 2.1.23 procedures during all modes of plant operation 74 Ability to identify and interpret diverse indications to validate the

~

2.1.45 response of another indication

[},'.:

Subtotal 2 2.2.37 I Ability to determine operability and/or availability of safety related 4.6 96 Ability to apply Technical Specifications for a system.

2.2.40 4.7 97 2.

Equipment 2.2.21 Knowledge of pre- and post-maintenance operability 2.9 73 Control requirements 2.2.35 Ability to determine Technical Specification Mode of Operation 3.6 75 2.2.38 Knowledge of conditions and limitations in the facility license 3.6 67

~

Subtotal .'<'" 3 2 2.3.14 Knowledge of radiation or contamination hazards that may arise .,

during normal, abnormal. or emergency conditions or activities.

Knowledge of radiation exposure limits under normal or 2.3.4 3.2 66 emergency conditions

3. Ability to comply with radiation work permit requirements during Radiation 2.3.7 normal or abnormal conditions 3.5 Control Knowledge of radiological safety procedures pertaining to 2.3.13 licensed operator duties, such as response to radiation monitor 3.4 71 alarms, containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.

Subtotal  :\' 1 Knowledge of low power/shutdown implications in accident (e.g ..

2.4.9 loss 01 coolant accident or loss of residual heat removal) 99 mitigation strategies.

4. Ability to perform without reference to procedures those actions Emergency 2.4.49 t require immediate operation of system components and 4.4 100 Procedures I troiS.

Plan 2.4.31 Knowledge of annunciator alarms, indications, or response 4.2 70 procedures Knowledge of RO tasks performed outside the main control room 2.4.34 during an emerQency and the resultant operational effects 4.2 72

... ( .....

Subtotal 2 . 2 Tier 3 Point Total ..  ; 10 7 NUREG-1021 , Rev.9, Supplement 1 1

I ES-401 Record of Rejected KIAs I Form ES-401-4 I Tier 1 Randomly Reason for Rejection Group Selected KIA 1/2 295033/2.2.12 An operationally valid SRa question could not be written for original selection; re-selected KIA from same category (2.4.6) 2/2 233000/2 A.8 An operationally valid SRa question could not be written for original selection- there is no interrelation between the system and the EOPs; re-selected KIA from same category (2.2.25) 3 2.1.14 An operationally valid SRa question could not be written for original selection; re-selected KIA from same category (2.1.40) 3 2.2.7 An operationally valid SRa question could not be written for original selection; re-selected KIA from same category (2.2.37) 1/1 295026/EK3.03 KIA is not applicable to this facility. Randomly re-selected KIA from EK3 category (EK3.01) 211 205000/K5.04 KIA was less than 2.5 with no plant specific priority. K5 category randomly re-selected (K5.02) 2/1 211000/K4.07 One of two systems randomly selected to remove the K4 KIA due to the K6 and General categories in tier 2 not meeting the "at least two" sample frequency. KIA was randomly re-selected (2.4.11) 2/1 215005/K4.07 One of two systems randomly selected to remove the K4 KIA due to the K6 and General categories in tier 2 not meeting the "at least two" sample frequency. KIA was randomly re-selected (K6.01) 2/1 259002/A4.05 KIA is not applicable to this facility. Randomly re-selected KIA from A4 category (A4.02) 211 262002lK1.07 KIA is not applicable to this facility. Randomly re-selected KIA from K1 category (K1.01) 2/2 201 001 IK6.04 KIA was the same concept as the KIA drawn in Tier 1/Group 1 (295006 Scram as it relates to the CRD Hydraulic System). Randomly re-selected KIA from K6 category (K6.03) 2/2 201006/K2.01 The only KIA in category K2 was less than 2.5 with no plant specific priority. Randomly re-selected KIA (A1.02) 212 202002/K3.06 KIA is not applicable to this facility. Randomly re-selected KIA from K3 category (K3.04) 3 2.2.3 KIA not applicable- not a multi-facility site. KIA randomly re-selected (2.2.35) 3 2.2.5 KIA was less than 2.5 with no plant specific priority. General category randomly re-selected (2.2.38)

NUREG-1 021, Rev.9, Supplement 1 1

Attachment 1 Systematic Sampling Methodology Used 2010 VY NRC ILO Exam The random sample for the written exam outline (NUREG 1021 ES-401-1/ES-401-3) was conducted as follows:

  • In accordance with NUREG 1021, section ES-401, Attachment A, the written exam outline was developed using poker chips to randomly sample and systematically eliminate KIA statements for each Tier/Group as follows:

o RO portion:

  • Tier 1 Group 1 -7 295027 was removed to start based on not being applicable to VY (Mark III containment ONLY). This left 20 topics of which each was sampled. There was no need to randomly sample/eliminate further.
  • Tier 1 Group 2 -7 295011 was removed to start based on not being applicable to VY (Mark III containment ONLY). Seven topics were randomly selected using poker chips.
  • Tier 2 Group 1 -7 207000 and 209002 were removed to start due to being systems not applicable to the facility. The remaining 21 topics were sampled once. Five ElAPE numbers were randomly sampled using poker chips. Thus, five topics had two KlAs drawn.
  • Tier 2 Group 2 -7 239003 was removed to start due to being a system not applicable to the facility. Twelve were randomly selected from among the remaining 32 using poker chips.
  • Tier 3 -7 Was randomly selected using poker chips from the general KlAs from NUREG 1123, slction 2.

o SRO portion

  • Tier 1 Group 1 -7 295027 was removed to start based on not being applicable to VY (Mark III containment ONLY). Seven topics were randomly selected using poker chips.
  • Tier 1 Group 2 -7 295011 was removed to start based on not being applicable to VY (Mark III containment ONLY). Three topiCS were randomly selected using poker chips from the remaining ElAPE numbers once the selected RO topics were removed. This allowed for a broader sample of topics.
  • Tier 2 Group 1 -7 207000 and 209002 were removed to start due to being systems not applicable to the facility. Five topics were randomly sam pled using poker chips after the five RO topics which already had two were removed. This allowed for a broader sample of topiCS.
  • Tier 2 Group 2 -7 239003 was removed to start due to being a system not applicable to the facility. Once the 12 RO topics were determined and discarded for this draw, three were randomly selected using poker chips from the remaining 20. This allowed for a broader sample of topics.
  • Tier 3 -7 The RO Tier 3 topics were selected and subsequently discarded for the SRO draw. Additionally, any general KIA not linked to 10CFR 55.43(b)(1-7) was discarded. Seven topics were randomly selected using poker chips from the general KlAs from NUREG 1123, section 2.
  • Once the topics were selected two additional random draws were made using poker chips to determine the category (G, A1-A4,K1-K6) and then the specific KIA.

ES-301 Administrative Topics Outline Form ES-301-1 Facility: VERMONT YANKEE Date of Examination: 11129110 Examination Level: RO [&] SRO D Operating Test Number: VY 2010 Administrative Topic Type Describe activity to be performed (see Note) Code*

S,M 2.1.31 Ability to locate control room switches, controls, and indications and to determine that they correctly Conduct of Operations reflect the desired plant lineup (RO 4.6, SRO 4.3)

)i> Control Room Panel Walkdowns lAW OP 0150, Section "P' S,N 2.1.20 Ability to interpret and execute procedure steps (RO 4.6, SRO 4.6)

Conduct of Operations )i> OP 0105 Phase 1A, step 28 (record parameters and calculate stable period on VYOPF 0105.03) 2.2.41 Ability to obtain and interpret station electrical R,N and mechanical drawings (RO 3.5, SRO 3.9)

Equipment Control )i> Operator identifies the required components, position and recommended sequence needed to tagout the "A" RBCCW pump using P&IDs G 191159 Sheet 3 (mechanical) and G-191301 Sheet 1 (electrical)/CWD Sheet 442. EN-OP 102, "Protective and Caution Tagging",

Attachment 9.2 will be used as a guide.

Radiation Control Emergency Procedures/Plan 2.4.27 Knowledge of the "Fire in the Planf' procedures S,D (RO 3.4, SRO 3.9)

)i> Making the required plant announcements for a fire in the SWitchgear room lAW OP 3020, Figure 1.

NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when ailS are required.

  • Type Codes & Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank (S 3 for ROs; S 4 for SROs & RO retakes)

(N)ew or (M)odified from bank (<:: 1)

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

RO Administrative Job Performance Measure Summary

  • A1a: Control Room Panel Walkdowns o Operator performs the required daily panel walkdowns o Operator determines there are 4 abnormalities associated with the walkdowns and takes the required actions. These discrepancies are:
  • CRP 9-3, Danger tag hanging on CS pump "An with the switch in PTL
  • CRP 9-3, HPCI-17 closed (vice open) and HPCI 57 & 58 open (vice closed)
  • A1 b: Operator records critical data o lAW OP 0105 Phase 1A, step 28, record parameters and calculate stable period on VYOPF 0105.03
  • A2: Operator identifies the required components, position and sequence needed to tag out the "Al! RBCCW pump using P&IDs G-191159 Sheet 3 (mechanical) and G 191301 Sheet 1 (electrical)/CWD Sheet 442:.

o Operator determines the following components and sequence:

  • "A" RBCCW pump control switch on CRP 9-6 (SW-9-6-19) PTL
  • Bus 9, cubicle 5B (P59-1A pump breaker) OPEN
  • V70-94A (pump discharge valve) CLOSED
  • V70-96A (pump suction valve) CLOSED
  • V70-923 (PI-2A isolation valve on pump discharge line) OPEN
  • V70-600 (pump casing vent valve) OPEN
  • V70-924 (PI-2A drain valve on pump discharge line) OPEN
  • A4: Making the required plant announcements for a plant fire o The operator demonstrates familiarity with the site paging system o The operator makes the required announcements based on the location of the fire (the Switchgear room) lAW OP 8020, Figure 1.

NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when all 5 are required.

  • Type Codes & Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank (S 3 for ROs; S 4 for SROs & RO retakes)

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

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

SRO Administrative Job Performance Measure Summary

  • A 1 a: License Reactivation Requirements o Operator determines the license reactivation requirements for an individual returning to crew after supporting Training for the last year.

o From the initial conditions including a list of under instruction watches stood, the operator determines the requalification, under instruction, and documentation requirements associated with reactivating a license.

o Operator notices an error on required form that would not meet the acceptance criteria.

  • A1b: The SAGs have been entered. Evaluate H2/0 2 concentrations in the Torus and Drywell to determine what course of action is taken.

o Given an initial Drywell and Torus Oxygen and hydrogen concentrations, answer the following:

  • Can high radiation isolations be defeated?
  • What is/are the procedures(s) that can be used to vent containment?
  • What is the criterion for securing venting?
  • Can release rate limits be exceeded?

o To answer these questions, the Op4:lrator must be able to interpret the tables in the Severe Accident Guidelines.

o In addition to the SAGs, the Operator must be able to interpret the Emergency Action Level table to determine what Off Site release rate limits are for a General Emergency.

  • A2: Review a completed surveillance form and determine that there is an out of specification condition which has made that system inoperable (OPST-CS-4123-02A, Section 12.4: SRO Review).

o OPST -CS-4123-02A, Section 12.4: SRO Review

  • Upon performing this review, the SRO notices the following discrepancies:
  • not all data is entered (missing Torus water volume)
  • not all steps are initialed for (missing performer initials for a step)
  • there is an out of specification for CS-12A not documented
  • A3: lAW ON 3157, "Loss of Fuel Pool Level", the operator has been directed to reposition two Fuel Pool Cooling valves in order to restore Fuel Pool level (FPC-53 and FPC-28) ..

The sequence the operator must follow to c()mplete the JPM is:

o which valves are operated o the type of Radiologically controlled area the valves are located in o which RWP to enter under o the dose and rate alarm setpoints for the selected RWP

  • A4: Operator determines that there is an Emergency Action Level classification of a Site Area Emergency lAW AP 3125, Appendix "A" (Hot) based on FS 1.1 (Loss or potential loss of any two barriers (Table F-1):
  • Un-isolable primary system discharge outside PC resulting in Secondary Containment area radiation or temperature above any Maximum Safe Operating Limit (EOP-4, Table 0)
  • RPV-ED is required

ES-301 Control Roomlln-Plant Systems Outline Form ES-301-2 Facility: VERMONT YANKEE Date of Examination: 11/29/10 Exam Level: RO Il(] SRO-I D SRO-U [J Operating Test No.: VY 2010 Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

System I JPM Title TypeCode* Safety Function S-1 Average Power Range Monitor System (215005)1 N, L, S 7 Transfer Mode Switch to Run (A4.01 3.2/3.1)

S-2 ReactorlTurbine Pressure Regulating System M,A,S 3 (241000)/Respond to failed EPR and MPR (A4.19 3.5/3.4)

S-3 Primary Containment Isolation System (Main Steam) M,A,EN,S 5 2230021 MSIV Full Closure Timing Test (Valve fails its timing test) (A2.08 2.7/3.1) 8-4 Control Rod Drive Hydraulic System (201001)1 Respond to N,S 1 a trip of a CRD Pump (A2.01 3.2/3.3)

S-5 Main Turbine Generator and Auxiliary System (245000)1 D,A,S 4 Testing of the Emergency Governor (Governor fails to reset)

(A4.023.1/2.9)

S-6 Radiation Monitoring System (272000)/ Respond to Hi N,A,EN,S 9 Reactor Building Ventilation Radiation Alarm (failure of RB ventilation to isolate) (A2.11 3.4/3.7)

S-7 Reactor Feedwater System (259001)1 Transfer Feedwater D,A,S 2 Pumps at Power (A4.02 3.9/3.7)

S-8 AC Electrical Distribution (262001)/ Cross Tie Buses eight D,S 6 and nine (A4.01 3.4/3.7)

In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)

P-l Emergency Generators (Diesel/Jet) (295016)/Shutdown the D,R 6 "An Emergency Diesel Generator locally (AA 1.043.1/3.2)

P-2 Control Rod Drive Hydraulic System (295037) Isolate and D,E,R 1 Vent the Scram Air Header (EA1.05 3.9/4.0)

P-3 Reactor Protection System (212000)/ Reset RPS Power D,E 7 Protection Panel Trip (A2.02 3.7/3.9)

Facility: VERMONT YANKEE Date of Examination: 11/29/10 Exam Level: RO D SRO-llXl SRO-U [J Operating Test No.: VY 2010 Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

System I JPM Title TypeCode* Safety Function S-1 Average Power Range Monitor System (215005)1 N, L, S 7 Transfer Mode Switch to Run (A4.01 3.2/3.1)

S-2 ReactorlTurbine Pressure Regulating System M,A,S 3 (241000)/Respond to failed EPR and MPR (A4.19 3.5/3.4)

S-3 Primary Containment Isolation System (Main Steam) M,A,EN,S 5 2230021 MSIV Full Closure Timing Test (Valve fails its timing test) (A2.08 2.7/3.1)

S-4 Control Rod Drive Hydraulic System (201001)/ Respond to N,S 1 a trip of a CRD Pump (A2.01 3.2/3.3)

S-5 Main Turbine Generator and Auxiliary System (245000)/ D,A,S 4 Testing of the Emergency Governor (Governor fails to reset)

(A4.02 3.1/2.9)

S-6 Radiation Monitoring System (272000)1 Respond to Hi N,A,EN,S 9 Reactor Building Ventilation Radiation Alarm (failure of RB ventilation to isolate) (A2.11 3.4/3.7)

S-7 Reactor Feedwater System (259001)/ Transfer Feedwater D,A,S 2 Pumps at Power (A4.02 3.9/3.7)

In-PlantSystems@ (3 for RO); (3 for SRo-l); (3 or 2 for SRO-U)

P-1 Emergency Generators (Diesel/Jet) (295016)/Shutdown the D,R 6 "An Emergency Diesel Generator locally (AA1.04 3.1/3.2)

P-2 Control Rod Drive Hydraulic System (295037) Isolate and D,E,R 1 Vent the Scram Air Header (EA 1.05 3.9/4.0)

P-3 Reactor Protection System (212000)1 Reset RPS Power D,E 7 Protection Panel Trip (A2.02 3.7/3.9)

All AO and SAO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SAO-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)lternate path 4-6 (5) / 4-6 (5) I 2-3 (2)

(C)ontrol room (D)irect from bank  ::; 9 (6) / ::; 8 (5) I::; 4 (2)

(E)mergency or abnormal in-plant ~ 1 (2) / ~ 1 (2) I ~ 1 (2)

(EN)gineered safety feature - I - / ~1 (2) (control room system)

(L)ow-Power / Shutdown ~ 1 (1)/~ 1 (1) I~ 1 (1)

(N)ew or (M)odified from bank including 1(A) ~ 2 (5) I ~ 2(5) / ~ 1 (3)

(P)revious 2 exams  ::; 3 (1) I::; 3 (1) I::; 2 (0) (randomly selected)

(A)CA ~ 1 (2) / ~ 1(2) I ~ 1(1)

(S)imulator

JPM JPM Description S-1 A plant startup is in progress with reactor power just below the APRM downscale setpoint (2%). The Control Room Supervisor (CRS) has directed the Operator at the Controls (OATC) to transfer the Mode Switch to RUN lAW OP 0105, "Reactor Operations", phase 20, step 10a -7 10g. This involves withdrawing control rods to clear APRM downscales, transferring the mode switch to RUN, withdrawing IRM detectors, and switching recorders.

S-2 The crew has entered OT 3115, "Reactor Pressure Transients", due to a failed EPA. Direction has been given to perform OT 3115, step 2 to swap pressure regulation from the EPR to the MPA. Once the operator lowers pressure back to it's pre-transient pressure, the MPR will fail requiring the operator to take the immediate action of scramming the reactor due to a failed EPR AND MPA.

S-3 A plant startup is in progress. While operating at 57% RTP (OP 0105, "Reactor Operations", Phase 48, step 12), the operator is directed to perform MSIV Full Closure Timing and RPS Relay Actuation Functional Test lAW OP 4113, Section "A". Steps 1 through 5 are complete. This is an alternate path JPM as follows: the closure time for MS-86A will be UNSAT (step 9.a.3) requiring the operator to suspend further MSIV testing and close MS*80A.

S-4 While operating at rated power, a trip of the "A" CRD Pump has resulted in the crew entering OPON-3145*

01, "Loss of CRD Regulating Function". The operator has been directed to start the "B" CRD Pump to restore CRD Hydraulics lAW OPON-3145-01, step 3.5.

S-5 While operating at rated power, the operator is directed to perform the monthly Emergency Governor Test lAW OP 0150, "Conduct of Operations and Operator Rounds" VYOPF 0150.08 (Operations Department Monthly Task Performance Listing), and OP 41150, "Turbine Generator Surveillance", Section "G". This is an alternate path JPM as follows: the Emergency Governor will not reset requiring a second attempt at resetting it. After the second attempt is unsuccessful, the trip/test switch is left in the lockout position and Maintenance contacted.

S-6 While operating at rated power, the Control Room receives annunciator CRP 9-5*H*1, "RX BLDG/REFUEL FLR CH A RAD HI". The operator is directed to verify the validity of the alarm and if necessary confirm/initiate the automatic actions associated with the Alarm Response Sheet (ARS). These actions include verifying Standby Gas Treatment (SBGT) automatically started, a full Group III Primary Containment Isolation System (PCIS) isolation occurred, and Reactor Building (RB) Ventilation isolated.

This is an alternate path JPM as follows: Reactor Building ventilation failed to isolate and the operator is expected to take action lAW EN-OP-115, "Conduct of Operations", section 5.3[2].

S-7 A plant startup is in progress. While operating at 86% RTP, the operator is directed to start the "B" RFP lAW OP 0105, "Reactor Operations", Phase 4B, step 22. This is an alternate path JPM as follows: the discharge valve strokes closed following pump start. The operator is expected to trip the "B" RFP lAW OP 0105 Phase 4B caution for step 22 (If minimum flow path or normal flow cannot be established, the pump shall be immediately secured).

S-8 While the plant was operating at rated power, an electrical fault on Bus 3 resulted in the loss of Buses 3 and 8. The crew has entered ON 3171, "Loss of Bus 3". The operator is directed to cross tie buses 9 and 8 lAW ON 3171, operator action step 4 and OP 2143, "480 and Lower Voltage AC System", Appendix "C",

Energizing 480Vac Bus 8 (Dead Bus) From Bus 9".

P-l The Control Room has been abandoned and actions of OP 3126, "Shutdown Using Alternate Shutdown Methods" have been taken. The Control Room Supervisor has directed you to shutdown the "N Emergency Diesel Generator locally lAW OP 3126, Appendix 19, step 19 using OP 2126 "Emergency Diesel Generators", section "0".

P-2 While operating at rated power, an electrical ATWS occurred resulting in the failure of all control rods to insert. The crew has entered EOP-2, ""ATWS RPV Control". The operator has been directed to vent the scram air header lAW OE 3107, "EOP/SAG Appendices", Appendix "0", "Manual Isolation and Venting of the Scram Air Header".

P-3 While operating at rated power, spurious electrical trips have resulted in the trips of the Reactor Protection System (RPS) MG set output breaker and RPS Power Protection Panels PP-A-1 and PP-A-2. The operator has been directed to reset the RPS Power Protection Panel trips lAW OP 2134, "Reactor Protection System", section "F",

I Appendix D Scenario Outline Form ES-D-1 Facility: VERMONT YANKEE Scenario No.: 1 Op Test No.: VY 2010 Examiners: Operators: CRS OATC BOP-Initial Conditions: A reactor power reduction is in progress to support planned maintenance on the "C" Reactor Feedwater Pump seal and electrical grid maintenance.

Turnover: The crew is directed to continue with a power reduction to 80% RTP to support planned maintenance on the "c" Reactor Feedwater Pump seal. The seal has minor leakage and there are no operational restrictions on operating with the current leakage. The plant is expected to remain at 80% RTP for -48 hours to support maintenance activities. From there, maintain 80% RTP until the maintenance is complete and RFP restored to service.

Critical Tasks: 1. When PCIS Group 1, 2, 3, 5, or 6 fails to isolate with a leak present, initiate PCIS Group manually. STANDARD: Leak or release terminated within 10 minutes of receipt of the auto isolation signal.

2. When torus pressure exceeds the suppression chamber spray initiation pressure, initiate drywell containment spray while in the safe region of the drywell spray initiation limit. S1"ANDARD: spray the drywell within 10 minutes of exceeding a Torus pressure of 10 psig AND RPV level not an overriding priority.

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

1 N/A R-OATC With the plant at 90% RTP, continue a power reduction to 80%

N-CRS RTP to support planned maintenance on the "cn RFP seal.

N- BOP Remove the "cn Reactor Feedwater pump from service following power reduction.

2 mfRC_04 C-ALL Inadvertent Initiation of RCIC (positive reactivity addition) (OT)

TS-CRS (TS) 3 mfFW_14 1- OATC Steam Flow Summer failure low (OT) 1- CRS 4 Override/mf TS-CRS Respond to Annunciator 3-N-9 (RHR/CS B HPCI BUS/LOGIC TBD FAIL) (TS) 5 mfED_03A M-ALL Loss of Bus 1 resulting in the loss of Feedwater and Reactor Scram (OT) mfDG_05A C-BOP "A" EDG fails to start C-CRS

mfHP_01 C- OATC HPCI trip C-CRS 6 LL Large steam Leak in the drywell C- BOP PCIS Group 3 failure (AC-6 and AC-6B fail to close)

  • (N)ormal. (R)eactivity. (I)nstrument, (C)om ponent, (M)ajor

I Appendix 0 Scenario Outline Form ES-O-1 Vermont Yankee 2010 NRC Scenario #1 The crew takes the watch with the reactor operating at 90% RTP. They will continue a power reduction to 80% and remove the "c" Reactor Feedwater Pump (RFP) lAW OP-010S, "Reactor Operations". This will be done to perform corrective maintenance on the *c" RFP.

The crew will respond to an inadvertent initiation of Reactor Core Isolation Cooling lAW OT 3110, "Positive Reactivity Insertion". After verifying the requirements of EN-OP-11S, "Conduct of Operations", the crew can override the system by tripping RCIC. The CRS will determine that RCIC is INOPERABLE and enter TS LCO 3.S.G.2 (14 days). RCIC will remain AVAILABLE for the remainder of the scenario.

The crew will respond to a downscale failure of the Steam Flow Summer. The failure will result in RPV water level lowering and require the OATC to take manual control of the FWLC system to restore level lAW OT 3113, "Reactor Low Level" and transfer the FWLC System to Single Element.

The crew will respond to annunciator 3-N-9, "RHRlCS B HPCI BUS/LOGIC FAIL". When an operator is sent to investi~~ate, it will be determined that DC-1C circuit breaker #2 was found tripped, As a result, the CRS will enter TS LCO 3.S.A.6 (24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> cold shutdown).

The crew will respond to the loss of Bus 1 and the loss of ALL feedwater lAW OT 3169, "Loss of Bus 1", The "A" EDG will not start automatically or manually due to an air start solenoid failure. lAW EOP-1,"RPV Control", RPV level control will be shifted to HPCI which will trip once started. The crew will elect to restart the "c" RFP with the seal leakage or reset the trip on RCIC to restore water level.

Eventually, pressure control will have to be shifted to RCIC and/or the SRVs due to the loss of the MHC system/Bypass valve operations (Au x Oil pump powered from Bus 6).

Shortly after the immediate actions for the Reactor Scram are taken, a steam leak will develop in the drywell requiring entry into EOP-3, "Primary Containment Control", Two PCIS Group 3 valves will fail (AC-6 and AC-6B) requiring manual operation to shut them with a leak in containment (CRITICAL TASK). As Torus pressure rises to 10 psig, action will be taken to initiate Drywell Sprays (CRITICAL TASK). To do this power will have to be restored to Bus 4 from the Vernon Tie for the "A"rC" RHRSW pumps (since the "B" RHR Loop logic power is still INOPERABLE) OR manual operation of "B" RHR Loop valves.

I Appendix D Scenario Outline Form ES-D-1 Facility: VERMONT YANKEE Scenario No.: 2 Op Test No.: VY 2010 Examiners: Operators: CRS OATC BOP-Initial Conditions: The plant is operating middle of cycle at 100% RTP. RHR-39A (TORUS SPRAY/CLG RHR) valve motor actuator is being repaired (30-day LCO entered 1 day ago per TS 3.5.B.1).

Turnover: Maintain 100% RTP in conjunction with performing "OncelWeek Pump Performance Testing" lAW OP 4160, Turbine Generator Surveillance", section B.1.c 7 B.U.

Local testing is not desired. Sections B.1.a and B.1.b have been completed.

Critical Tasks: 1. With the reactor at power and a full auto scram signal, manually scram the reactor. STANDARD: Actuate the manual scram pushbuttons, place the mode switch in SHUTDOWN, or actuate the ARI/RPT pushbuttons within 1 minute of reaching the Limiting Safety System Setting (LSSS for APRM Hi-Hi).

2. During an ATWS with conditions met to perform power/level control TERMINATE AND PREVENT INJECTION into the RPV using appendix GG, until conditions are met to re-establish injection. STANDARD: completion of Terminate and prevent injection lAW OE 3107 Appendix GG within 5 minutes of loss of forced circulation.
3. With a reactor scram required and the reactor not shutdown, TAKE ACTION TO REDUCE POWER by injecting boron and/or inserting control rods, to prevent exceeding the primary containment design limits. STANDARD: Actions taken within 10 minutes of the scram failure to implement appropriate appendices and/or inject SLC. Only one method needs to be used. The method must result in successful control rod insertion or SLC injection.

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

1 N/A N-BOP Complete "OncelWeek Pump Performance Testing" lAW OP 4160, N-CRS Turbine Generator Surveillance", section B.1.c 7 B.1.f. Local testing is not desired.

2 mfED_06E TS-CRS Loss of DC-2AS 3 mfFW_10A ATC Failure of the 'A' feedwater control level signal (OT)

RS 4 mfED_05C C-ALL Loss of Bus 8 (ON) (TS)

TS-CRS mfPC_11A C-BOP Failure of SBGT train "A" fan to auto start

I Appendix D Scenario Outline Form ES-D-1 S mfMC_08 C- BOP Condenser air in-leakage/High Condenser backpressure (OT) due C-CRS to irA" Condenser casing failure N/A R-OATC Power reduction lAW OT 3120, "Condenser High Backpressure" 6 mfNM_OSD M-ALL APRM "D" Fails upscale with and Hydraulic ATWS mfRD_12A mfRD_12B mfRP_01A C-OATC Failure to auto scram; manual scram insertion results in partial rod mfRP_01B C-CRS insertion; ARI/RPT initiated mfRP_09A C-OATC Failure of RWCU to completely isolate on SLC initiation.

mfRP_09B mfSL_01A C-OATC Failure of running SLC pump

  • (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor

I Appendix D Scenario Outline Form ES-D-1 Vermont Yankee 2010 NRC Scenario #2 The crew will complete the "Once/Week Pump Performance Testing" lAW OP 4160, "Turbine Generator Surveillance", section B.1.c ~ B.1.f. Local testing is not desi red.

The crew will respond to a loss of DC-2AS. The CRS will take actions lAW ON 3163, "Loss of DC-2AS" and enter required TRM and TS LCOs (section 3.10)

Shortly after the "A" EDG control power has been transferred to its alternate source, a failure of the 'A' feed flow detector will occur. The crew will respond lAW OT 3114 (Reactor High Level) and place the "A" FRV Controller in manual control to block the auto signal failure.

Once level has been restored to its pre-transient value, the crew will respond to a loss of Bus 8. SBGT "A" fan will fail to auto-initiate upon receipt of the Group III isolation signal. The crew will backup the Group III isolation and initiate SBGT "A". Review of Tech Specs will reveal a 24-hour shutdown LCO due to inadequate RHR torus cooling/spray capability and inadequate LPCI (loss of emergency bus 8 will also get the plant into a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> LCO).

Once the 24-hour shutdown LCO has been determined, condenser air in-leakage will result in rising condenser backpressure and entry into OT 3120. While attempts are made to determine the cause, a power reduction will be ordered lAW OT 3120. The leak will be slow enough to perform a controlled power reduction in order to get a reactivity manipulation. During the power reduction, an upscale failure of the "0" APRM will occur.

When the APRM fails upscale, it will result in a trip of RPS Channel "B" which will fail and subsequently, an ATWS will result. The manual scram pushbuttons will only insert control rods partially (CRITICAL TASK) and ARI/RPT push buttons will be used unsuccessfully. The crew will be evaluated controlling and shutting down the plant in accordance with EOP-1 and EOP-2. lAW EOP-2, "ATWS RPV Control", actions are taken to insert control rods and/or initiate SLC (CRITICAL TASK), and terminate and prevent injection (CRITICAL TASK). When SLC is initiated, the Group V isolation will fail and CU-18 and 68 will fail to isolate (CU 15 lost power and tripped the RWCU pump during the loss of Bus 8). After three minutes of operation, the "A" SLC pump will trip resulting in no SLC injection.

I Appendix D Scenario Outline Form ES-D-1 Facility: VERMONT YANKEE Scenario No.: 3 Op Test No.: VY 2010 Examiners: Operators: CRS OATC BOP-Initial Conditions: The plant is operating at 100% RTP Turnover: A seven day LCO is in effect for the "B" train of Standby Gas Treatment (SBGT) being INOPERABLE (TS 3.7.B.3.a)

Critical Tasks: 1. With the reactor at power and a full auto scram signal. manually scram the reactor. STANDARD: actuate the manual scram pushbuttons, place the mode switch in SHUTDOWN, or actuate the ARI-RPT pushbuttons within 1 minute of reaching the Limiting Safety System Setting (LSSS MSIV closure).

2. With a Primary system discharging into Secondary Containment and area radiation/temperature/water levels exceed Maximum Safe Operating Levels in more than one area. STANDARD: Initiate RPV-ED within 5 minutes of area radiation/temperature/water levels exceeding Maximum Safe Operating Levels in more than one area.
3. When a leak is present, dispatch personnel to manually isolate associated PCIS valves that have failed to isolate automatically and manually from the Control Room. STANDARD: Direct I&C/Maintenance/AOs to manually isolate PCIS valves within 15 minutes of receipt of the isolation signal.

(RadiologicaVenvironmental considerations and power source may affect the time standard for this critical task).

Malf. No. Event Description 1 N/A N-BOP MonthlyTBCCW/RBCCW pump swaps lAW OP 0150, OP 2182, N-CRS and RP 4183 2 mtPC_2LR8 TS-CRS Respond to annunciator 4-M-3, "OWL SUMP VLV CLOSED" 294 (alarm due to blown fuses tor LRW-82 and LRW-94) (TS) 3 mfED_05B C- ALL Loss of Bus 7 (Off Normal event using OP 2143) mfEG_12A C-BOP Failure of the "A" Stator Water Cooling Pump to auto start (OT)

R-OATC Recirculation Pump Trip (OT) (TS)

TS-CRS 4 mfRX_01 C MSL High Radiation due to fuel element failure

I Appendix D Scenario Outline Form ES-D-1 mf 1- BOP Steam Packing Exhauster fan and discharge valve fail to trip/close TBD 5 mfMS_07 M-ALL Main Steam HC" Line Break in Steam Tunnel with Fuel Element Failure mfRP_01A C-OATC Failure to auto scram on MSIV closure C-CRS mfMS_01C C-BOP Failure of Main Steam Line HC" to isolate mfMS_02C C-CRS

  • (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor

I Appendix D Scenario Outline Form ES-D-1 Vermont Yankee 2010 NRC Scenario #3 The crew will be directed to perform the OP 0150, Section E, "Operations Department Weekly and Monthly Task Performance Listing", surveillance of swapping the TBCCW and RBCCW pumps lAW RP 2183 and OP 2182. VYOPF 0150.08 will be documented when the surveillances are completed. The swap of the TBCCW heat exchangers will be turned over as being completed with all temperatures stabilized.

The crew will respond to 4-M-3, "DWL SUMP VLV CLOSED". After it is reported the cause of the valves closing is a blown fuse, The CRS will enter TS LCOs 3.6.C.2 (7 days) and 3.7.D.2/4.7.D.2 (close and deactivate a valve in the line containill9 the INOPERABLE PCIS valve and verify the line is isolated every 31 days).

The crew will respond to a loss of Bus 7. The loss of Bus 7 will result in the trip of the "B" Recirculation Pump requiring entry into OT 3117, Recirculation Pump trip. The crew will take actions including a power reduction to 45% RTP. The CRS will address single loop operation Technical Specifications. Also on the loss of Bus 7 (MCC-7B), the crew will respond to a loss of Stator Water Cooling lAW OT 3167 due to the failure of the "A" Stator Water Coolillg Pump to automatically start. Action must be taken to manually start the pump within 1 minute to ensure the Main Turbine does not trip/Reactor Scram.

During the power reduction, a fuel element failure will result in rising Main Steam line (MSL) radiation levels. The crew will enter ON 3152, "MSL and Off Gas High Radiation",

and ON 3153, "Excessive Radiation Levels". Once MSL radiation levels reach >3x NFPB (annunciators 5-H-6/5-J-6), the Steam Packing Exhauster fan and discharge valve (AE 12B) will fail to trip/close requiring manual operator action.

As the crew addresses the rising MSL radiation levels, a Main Steam Line break in the steam tunnel will occur. The high steam flow isolation will shut MSIVs. The reactor will fail to auto scram and the crew will take action to manually scram the reactor within 1 minute of reaching the Limiting Safety System Setting (LSSS) for MSIV Closure (TS 2.1.G) (CRITICAL TASK). Additionally, the "C" Main Steam line will fail to automatically isolate on high steam flow. MS-80C and MS-86C will not be able to be manually closed as well. The crew will have to take action to contact support personnel to shut the valves locally within (TBD) minutes in an attempt to shut the PCIS valves with an automatic isolation signal failure (CRITICAL TASK).

Following the scram, RPV water level and pressure will be addressed by EOP-1, "RPV Control". Rising temperatures in the Reactor Building will result in entry in EOP-4, "Secondary Containment Control". Once temperatures reach the Maximum Safe Operating Limit in more than one area (RS 252' and 280' elevations), and RPV Emergency Depressurization will be performed lAW EOP-5, "RPV-ED" within 5 minutes (CRITICAL TASK).

I Appendix 0 Scenario Outline Form ES-O-1 Facility: VERMONT YANKEE Scenario No.: 4 Op Test No.: VY 2010 Examiners: Operators: CRS OATC BOP-Initial Conditions: Power is -2% with a reactor startup in progress.

Turnover: OP 0105, "Reactor Operations", is complete thru Phase 2.C. The crew will be directed to perform a Turbine Chest warm-up lAW OP 0105 Phase 2.0. Step 1 and continue Reactor Startup (60 to 80 degree/hour heat up rate).

Critical Tasks: 1. When torus level cannot be maintained above 7 ft, perform RPV emergency depressurization. STANDARD: Initiate RPV-ED such that RPV pressure is < 50 psig when Torus level reaches 5.5 ft.

2. When suppression pool level cannot be maintained above top elevation of the HPCI exhaust, if HPCI is running, trip and prevent HPCI operation irrespective of adequate core cooling. STANDARD: HPCI injection is terminated before Torus level falls below 6 feet if HPCI is running.
3. During an ATWS with Emergency Depressurization required, terminate and prevent injection into the RPV (using OE 3107, "EOP/SAG Appendices",

Appendix GG) until conditions are met to re-establish injection. STANDARD:

Terminate and prevent injection lAW Appendix GG such that no system other than SLC, CRD, and/or RCIC is/are injecting during the RPV-ED.

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

1 N/A N- BOP Perform Turbine Chest warm-up.

N-CRS 2 N/A R- OATC Withdraw control rods to continue power ascension.

2 mfNM_03F 1- OATC IRM "F" fails upsca~e (TS).

1- CRS TS-CRS 3 rfPP _06 C- BOP Seismic Event resulting in a leak in the SLC Tank (TS) (Off Normal C-CRS event using OP 3127)

TS-CRS mfMS_09 C-BOP Gland Seal Regulator Fails Closed mfSW_14A C--OATC Failure of the standby TBCCW pump to auto start after running mfSW_21B pump trips due to seismic event. (OT)

I Appendix D Scenario Outline Form ES-D-1 4 rfPP _06 M-ALL Seismic Event (after shock)/Loss of Normal Power mfED_17 mfRC_01 1- BOP RCIC turbine trip mfDG_09B C-BOP Failure of the "B" Emergency Diesel Generator breaker to auto close mfRP_01A C-OATC Failure of the automatic and manual scrams; 4 control rods stuck mfRP_01B C-CRS out mfRD_02XXYY mfRD_02XXYY mfRD_02XXYY mfRD_02XXYY 5 mfPC_10 M-ALL Leak in the Torus;

  • (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor

I Appendix D Scenario Outline Form ES-D-1 Vermont Yankee 2010 NRC Scenario #4 The crew will initiate Turbine Chest Warming and continue with the reactor startup, withdrawing control rods to continue with the power ascension. As the startup progresses, IRM "F" will fail upscale resulting in a rod withdrawal block and a half scram, requiring the crew to evaluate Tech Specs, and bypass the failed IRM.

The crew will be evaluated responding to a seismic event that causes a leak in the SLC tank and a trip of the running TBCCW/failure of the standby TBCCW pump to auto start. The actions of OP 3127, "Natural Phenomena" and EN-OP 115, "Conduct of Operations"IOT 3165, "Loss of TBCCW", will be taken to respond to the seismic event and failure of the standby pump to start. Technical Specifications will be consulted, revealing a 24-hour shutdown LCO (TS section 3.4). Also, the Gland Seal Regulator will fail closed requiring the crew to open the bypass valve lAW the ARS to maintain condenser backpressure.

A seismic aftershock will occur resulting in the Loss of Normal Power (LNP) and a break in the weld of the 'N RHR suction line to the torus. The crew will be evaluated responding to the seismic event (OP 3127), Loss of Normal Power (OT 3122, "Loss of Normal Power") and failure of the "B" Emergency Diesel Generator breaker to automatically close. The breaker will be able to be closed by the operator in the Control Room. A tailure of both automatic and manual scram capability exists. ARI/RPT initiation will result in successful rod insertion of all rods but 4 control rods which are stuck out.

A loss of high pressure injection from Feed and Condensate will result in direction to control level with EOP-2 Table "H" systems: Based on the low power history, the CRS may direct the use of ReIC. If RCIC is started, it will trip soon after. If HPCI is started, the crew will soon realize that the lowering Torus level will require HPCI to be inhibited as directed by EOP-3 (CRITICAL TASK- this mayor may not be evaluated based on what crew IJses for level control).

Maximizing CRD flow may also be used for level control.

During the RPV-ED, an alternate injection system (EOP-2, Table "J") may be required although maximizing CRD flow may be enough to support rapid depressurization based on low power histOlY.

Once the lowering torus level is noted, the crew will be evaluated on entry into and execution of EOP-3, "Primary Containment Control" and EOP-4, "Secondary Containment Control". The crew will also enter ON 3158, "Reactor Building High LevellTemperature", due to high RB water level. Because of the size of the leak, the crew will perform an RPV Emergency Depressurization (CRITICAL TASK).

With an RPV-ED required during an ATWS condition, the crew will be required to terminate and prevent injection prior to the RPV-ED (CRITICAL TASK).

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