ML090420115

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Vermont Yankee Draft - Outlines (Folder 2)
ML090420115
Person / Time
Site: Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date: 11/23/2008
From: Romeo M A
Entergy Nuclear Vermont Yankee
To: Fish T H
Operations Branch I
Hansell S
Shared Package
ML082600332 List:
References
TAC U01743 50-271/09-301
Download: ML090420115 (25)
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ES-401 Written Examination Outline Form ES-401-1 ~ ~~ ~~ Facility: Vermont Yankee NRC Date of Exam: February 2009 Note: 1. 2. 3. 4. 5. 6. 7.* a. 9. Ensure that at least two topics from every applicable WA category are sampled within each tier of the RO and SRO-only outlines (Le., except for one category in Tier 3 of the SRO-only outline, the "Tier Totals" in each WA category shall not be less than two). 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 k1 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. Systemdevolutions within each group are identified on the associated outline; systems or evolutions that do not apply at the facility should be deleted and justified; operationally important, site-specific systems that are not included on the outline should be added. Refer to section D.l .b of ES-401, for guidance regarding elimination of inappropriate WA statements. 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.

Absent a plant specific priority, only those KAs having an importance rating (IR) of 2.5 or higher shall be selected. Use the RO and SRO ratings for the RO and SRO-only portions, respectively.

Select SRO topics for Tiers 1 and 2 from the shaded systems and WA categories. 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. Refer to Section D.l .b of ES-401 for the applicable WAS 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. 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. For Tier 3, select topics from Section 2 of the WA Catalog, and enter the WA numbers, descriptions, IRs, and point totals

(#) on Form ES-401-3.

Limit SRO selections to WAS that are ~ ~~~~ linked to 10CFR55.43 ES-401 2 Form ES-401-1 Vermont Yankee Written Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 1 29501 9 Partial or Total Loss of Inst. Air i 8 295005 Main Turbine Generator I Trip / 3 295004 Partid or Total Loss of UC: Pwr i 6 I 29!5006 SCRAM 1 1 295001 Partial or Corriplete Loas of Forced Core FIOW Circulation

/ 1 & 4 29!5025 High Reactor Pressure I 3 295028 High Drywell Temperature

/ 5 295005 Main Turbine Generator Trip

/ 3 Ix 295028 High Drywell Temperature

/ 5 295030 Low Suppression Pool Water Level / 5 295026 Suppression Pool High Water Temp.

/ 5 29501 9 Partial or Total Loss of Inst. Air / 8 600000 Plant Fire On-site

/ 8 295023 Refueling Accidents interpret the following as they apply to MAIN TURBINE GENERATOR TRIP PARTIAL OR COMPLETE LOSS OF 3.7 - 2.7 2.9 - 3.8 1 the significan&?

ot each anniinciator or I 4 3 interpret the ~ollo~in~

85 they apply to I I HIGH DRYWELL TEMPEffATURE

' : Pressure effects 76 - 77 78 - 79 80 81 - u2 39 41 42 44 45 ES-401 ~ ~~~ ~~ 295037 SCRAM Conditions Present and Reactor Power Above APRM Downscale or Unknown I1 295001 Partial or Complete Loss of Forced Core Flow 1 Circulation

/ 1 & 4 2 WA Category Totals: Form ES-401-1 3 Vermont Yankee Written Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 1 213501 8 Partial or Total Loss of ccw I 8 29501 6 Control Room Abandonment 17 700000 Generator Voltage and Electric Grid Disturbances 295004 Partial or Total Loss of DC Pwr I 6 295006 SCRAM I 1 295021 Loss of Shutdown Cooling I4 295024 High Drywell Pressure I I5 295038 High Off-site Release Rate I 9 I 295031 Reactor Low Water 295003 Partial or Complete Loss of AC I 6 295025 High Reactor Pressure 13 owing responses as - 3 WER : Manual and auto bus ES-401 3 Form ES-401-1 Vermont Yankee Written Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 2 295029 High Suppression Pool Water Level / 5 500000 High CTMT Hydrogen Conc. I5 ES-401 4 Vermont Yankee Written Written Examination Outline Plant Systems - Tier 2 Group 1 Form ES-401-1 20.3000 RHR'LPCI.

Injection Mcde 250002 Reactor Weter Level Cocitrol 400000 Component Cooling Water 21 1000 SLC 21 1000 SLC 400000 Component Cooling Water 263000 DC Electrical 205000 Shutdown Cooling REACTOR WATER LEVEL CONTROL SYSTEM and (b) based on those predictioris, use impacts of the following or1 the STANDBY I.IQUID CONTROL SYSTEM , and (b) based on those predictions, use procedures to correct, control.

or I 3.9 - 3.4 - 4.4 3.2 ri I itigate the consey uences of thase abrioimal condrtroi-rs 01 I I following: Radiation monitoring lies to the following:

ES-401 4 Vermont Yankee Written Written Examination Outline Plant Systems - Tier 2 Group 1 Form ES-401-1 3.1 supply K3.02 - Knowledge of the effect that a loss or malfunction of the TlON SYSTEM - 34 I I : ADS logic operation I K6.01 - Knowledge of the effect that a loss or malfunction of the following will have on the LOW PRESSURE CORE SPRAY ES-401 4 Vermont Yankee Written Written Examination Outline Plant Systems - Tier 2 Group 1 Form ES-401-1 INTERMEDIATE RANGE MONITOR (IRM) SYSTEM ; and (b) based on those predictions, use procedures to correct, control or mitigate the Up scale or down automatic operations of the PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF including: System indicating lights and Plan: Ability to recognize abnormal indications for system operating parameters which are entry-level conditions for 3.7 2.6 3.4 3.3 - 4.1 4.0 - 4.5 - 2.7 ES-401 4 Vermont Yankee Written Written Examination Outline Plant Systems - Tier 2 Group 1 Form ES-401-1 stem # / Name ES-401 5 Vermont Yankee Written Written Examination Outline Plant Systems - Tier 2 Group 2 Form ES-401-1 1 I System temperature ES-401 5 Vermont Yankee Written Written Examination Outline Plant Systems - Tier 2 Group 2 Form ES-401-1 202001 Recirculation I I 259001 Reactor Feedwater 234000 Fuel Handling I Equipment I 290001 Secondary CTMT 272000 Radiation Monitoring RECIRCULATION SYSTEM

3.1 34 3.4 35 3.4 36 3.6 37 4.2 38 12/3 ES-401 2.1.39 Generic Knowledge and Abilities Outline (Tier 3) Knowledge of conservative decision making Dractices.

Form ES-401-3 Facility: Vermont Yankee Written Date: 0711 7/08 Category KIA # Topic RO SRO-Only IR 4.2 1 100 Ability to interpret reference materials, such as graphs, curves, tables, etc. 2'1'25 - 3.4 2.1.8 I Ability to coordinate personnel activities 66 - 67 I 1. Condu of Operations I outside the control room. I Knowledge of individual licensed operator responsibilities related to shift staffing, such as medical requirements, "no-solo

operation, maintenance of active license status, 1 OCFR55. etc. 2.1.4 3.3 I Subtotal Ability to determine Technical Specification 22.35 1 Mode of Operation.

4.5 I 95 Kn~wledge of pre and p~~~~main~etiance opera~ili~y requirements.

2.2.21 T 2. Equipment Control 4.1 68 Knowledge of tagging and clearance procedures 2.2.1 3 I 69 75 - - Ability to apply technical specifications for a system. Ability to determine operability and/or availability of safety related ]equipment.

2.2.40 2.2.37 Subtotal 3 Ab~I~~y to control radiation releases.

2.3.1 t Knowledge of Radiological Safety Principles pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc. Knowledge of radiation exposure limits under normal or emergency conditions.

3. Radiation Control 2.3.12 3.2 - 3.2 71 - 2.3.4 Subtotal ES-401 Generic Knowledge and Abilities Outline (Tier
3) Form ES-401-3 Ability to perform without reference to procedures those actions that require immediate operation of system components 2'4'49 and controls.

Knowledge of operational implications of EOP warnings, cautions, and notes.

2.4.2Q 2'4'46 4.2 72 2.4.42 Knowledge of emergency response facilities.

2.6 73 2.4.6 Knowledae of EOP mitiaation strateaies.

3.7 74 Ability to verify that the alarms are consistent with the plant conditions.

71 - 4.4 4.3 97 98 2 7 ES-401 Record of Rejected WAS Form ES-401-4 Randomly Selected KIA 500000 I EA1.05 286000 I A2.02 204000 12.4.3 256000 I K3.07 2.2.4 201 006 I K1.08 259001 I A3.06 Reason for Rejection

(#62) EA1.05 - Ability to operate and monitor the following as they apply to iIGH CONTAINMENT HYDROGEN CONTROL:

Wetwell sprays Topic does not apply to VY. Randomly selected EA1.03, Ability to operate and monitor the following as they apply to HIGH CONTAINMENT HYDROGEN CONTROL: Containment atmosphere control system (#91) A2.02 - Ability to (a) predict the impacts of the following on the FIRE PROTECTION SYSTEM

and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or Dperations
D.C.

distribution failure: Plant-Specific Topic does not lend itself to a discriminating question (system function)

Randomly selected A2.08, Ability to (a) predict the impacts of the following on the FIRE PROTECTION SYSTEM

and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations
Failure to actuate when required (#93) 2.4.3 - Emergency Procedures

/ Plan: Ability to identify post-accident instrumentation. Topic does not apply to VY. Randomly selected 2.4.6, Emergency Procedures

/ Plan: Knowledge symptom based EOP mitigation strategies.

(#29) K3.07 - Knowledge of the effect that a loss or malfunction of the REACTOR CONDENSATE SYSTEM will have on following: Isolation condenser:

Plant-Specific Topic does not apply to VY. Randomly selected K3.04, Knowledge of the effect that a loss or malfunction of the REACTOR CONDENSATE SYSTEM will have on following: Reactor Feedwater System

(#68) 2.2.4 (multi-unit license)

Ability to explain the variations in control board layouts, systems, instrumentation and procedural actions between units at a facility.

Not a multi unit license. Randomly selected 2.2.1 3, Knowledge of tagging and clearance procedures

(#27) K1.08 - Knowledge of the physical connections and/or cause- effect relationships between ROD WORTH MINIMIZER SYSTEM (RWM) (PLANT SPECIFIC) and the following: Reactor power (turbine first stage pressure): P-Spec(Not-BWRG)

Does not apply to VY. Randomly selected K1.04, Knowledge of the physical connections and/or cause- effect relationships between ROD WORTH MINIMIZER SYSTEM (RWM) (PLANT SPECIFIC) and the following:

Steam flowheactor power VY has electric feed pumps could not write a specific question to match the KIA. Randomly selected A3.10, Ability to monitor automatic operations of the REACTOR FEEDWATER SYSTEM including:

Pump trips (#35) A3.06 - Ability to monitor automatic operations of the REACTOR FEEDWATER SYSTEM including:

Pump discharge pressure Record of Rejected WAS Form ES-401-4 ES-401 295023 I AK3.05 295025 12.4.41 29501 3 I 2.1.27 262002 I 2.1.28 2.2.43 295003 / AK3.04 2.2.43 (#45) AK3.05 - Knowledge of the reasons for the following responses as they apply to REFUELING ACCIDENTS

Initiation of SLC/shut-down cooling: Plant-Specific Does not apply to VY, randomly selected AK3.03 Knowledge of the reasons for the following responses as they apply to REFUELING ACCIDENTS:

Ventilation isolation.

(#81) 2.4.41 - Emergency Procedures I Plan: Knowledge of the emergency action level thresholds and classifications. (High Reactor Pressure) There is no E-Plan action level associated with High Reactor Pressure. Randomly selected 2.4.45 Ability to prioritize and interpret the significance of each annunciator or alarm. Low LOD for a discriminating SRO level question for this WA. Randomly selected 2.1.25 Ability to interpret reference materials, such as graphs, curves, tables, etc. Low LOD for a discriminating SRO level question for this K/A.Randomly selected 2.1.20 Ability to interpret and execute procedure steps.

Same WA as Common question

  1. 75. Randomly selected 2.2.21, Knowledge of pre and post-maintenance operability requirements. a discriminating question.

Very limited procedural guidance.

Randomly selected AK3.01, Manual and auto bus transfer Knowledge of the process used to track inoperable alarms. Used as a JPM on the Audit Exam.

Randomly selected 2.2.37, Ability to determine operability andlor availability of safety related lequipment.

(#85) 2.1.27 Knowledge of system purpose and/or function.

(#89) 2.1.28 - Conduct of Operations: Knowledge of the purpose and function of major system components and controls.

(#99) 2.2.43 - Knowledge of the process used to track inoperable alarms. Knowledge of the reasons for the following responses as they apply to PARTIAL OR COMPLETE LOSS OF A.C. POWER: Ground isolation. Could not Write ES-301 Administrative Topics Outline Form ES-301-1 Examination Level:

Conduct of Operations 1 I Equipment Control r Radiation Control Emergency I Procedures/Plan Date of Examination: 2/09 RO Operating Test Number:

N09-1 Type Code" N, S N, S M, R Describe activity to be performed 2.1.29 (4.1) Knowledge of how to conduct system lineups, such as valves, breakers, switches, etc. JPM: Perform the RHR System Valve Lineup 2.2.1 2 (3.7) JPM: Knowledge of Surveillance Procedures Perform a Drvwell Temperature Profile 2.3.1 1 (3.8) Ability to control radiation releases JPM: Determine Offgas Release Rate without ERFlS 2.4.43 (3.2)

Knowledge of Emergency Communications Systems and techniques JPM: Perform Control Room Emergency Communications Checks All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking *Type Codes

& Criteria: (C)ontrol room (0), (S)imulator (3), or Class(R)oom (1) (D)irect from bank (s 3 for ROs; 2 4 for SROs & RO retakes) (1) (N)ew or (M)odified from bank (2 1) (3) (P)revious 2 exams (2 1; randomly selected)

(0) NUREG-1 021, Revision 9 ES-301 Administrative Topics Outline Form ES-301-1 Examination Level:

Conduct of Operations Conduct of Operations Equipment Control Radiation Control Emergency Procedures/Plan Date of Examination:

2/09 SRO Operating Test Number:

N09-1 Type Code* N, S M, R N, S Describe activity to be performed 2.1.29 (4.0) Knowledge of how to conduct system lineups, such as valves, breakers, switches, etc. JPM: Perform the RHR System Valve Lineup 2.1.7 (4.7) Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation JPM: 2.2.12 (4.1)

Perform a Core Thermal Hydraulic Limits Evaluation Knowledge of Surveillance Procedures JPM: Review a Surveillance 2.3.1 1 (4.3) Ability to control radiation releases JPM: 2.4.44 (4.4) Determine Offgas Release Rate without ERFlS Knowledge of Emergency Plan Protective Action Recommendations JPM: Off-Site Protective Action Recommendations (evacuate)

All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking onlv the administrative tooics. when 5 are reauired.

  • Type Codes & Criteria: (C)ontrol room, (0) (S)imulator, (3) or Class(R)oom (2) (D)irect from bank (5 3 for ROs; I 4 for SROs & RO retakes) (1) (N)ew or (M)odified from bank (2 1) (4) (P)revious 2 exams (I 1; randomly selected)

(1) NUREG-1021, Revision 9 ES-301 Control Room/ln-Plant Systems Outline Form ES-301-2 Faci I i ty: Vermont Yankee Date of Examination: 2/2009 Exam Level (circle one): RO / SRO(I) / SRO (U) Operating Test No.: NRC-1 Control Room Systems@ (8 for RO; 7 for SRO-I; 2 or 3 for SRO-U, including 1 ESF) System / JPM Title S-1 Shift Rx Level Control From The Main Feed Reg Valve To The Auxiliary Feed Reg Valve 259002 RX Water Level Control System, WA: A4.01 3.8/3.6 Swap Pressure Regulators (EPR to MPR) 241 000 Reactornurbine Pressure Regulating System, WA: A4.19 3.513.4 -Ii: - S-2 S-3 Secure RHR from the Shutdown Cooling Mode 205000 Shutdown Cooling System, WA: A4.01 3.7/3.7 S-4 Line-up for Primary Containment Spray Using Fire System to RHR Loop "A" 226001 RHWContainment Spray Mode, WA: A4.02 3.1/3.1 Type Code* Safety M, A, s 2 Function D, s 3 N, L, s 4 l5 D, EN, S S-5 Swap RBCCW

& TBCCW Pumps 400000 Component Cooling Water System, WA: A2.01 3.3/3.4 Initiate SLC to the Vessel 21 1000 Standby Liquid Control System, WA: A4.02 4.2/4.2 Transfer Station Load from the Auxiliary Transformer to the Startup Transformer 262001 AC Electrical Distribution, WA: A4.04 3.6/3.7 S-6 S-7 S-8 Rx Startup to Criticality (RO) 215004 SRM System, WA: A4.01 3.9/3.8 In-Plant Systems@ (3 for RO; 3 for SRO-I; 3 or 2 for SRO-U) P-1 Place Charger BC-1-1 B in Service D 6 263000 DC Electrical Distribution, WA: A3.01 3.2/3.3 P-2 Boron Injection from the SLC Tank Using the CRD System D, E, R 1 APE 295037 Scram Condition Present and Reactor Power Above APRM Downscale or Unknown WA: EA1.10 3.7/3.9 D, A 8 P-3 Shifting Air Compressors t -'i 300000 Instrument Air System, WA: K4.01 2.8/2.9 1 D, A, EN, S _I NUREG-1021, Revision 9

I Appendix D Scenario Outline Form ES-D-1 1 Event No. 11 Facility:

VERMONT YANKEE Scenario No.: 1 Op Test No.: 2009 NRC Examiners: Operators:

SRO - RO - BOP - Malf. No. Event Event Description TvDe* I/ Initial Conditions:

At 100% power 1 2 DG "A has been operating for 30 minutes for Monthly Diesel Generator Slow Start Operability Test (Tech Spec) per OP 41 26, Sect B. The test is being performed following a diesel lube oil change. The test must be run for at least two hours at 2700 to 2750 kW and 1600 k 50 kVAR OUT. N/A N - BOP SRO Raise Main Generator output to heavy load schedule and maximum Lagging (OUT)

VAR Load IAW OP 2140, Sect. H. Turbine Vibration respond per ARS 7-F-2 (lower power stop test) mfTU-03A R - RO RHR Pump B is 00s for severe vibrations that occurred during surveillance testing and is tagged out for Maintenance investigation.

c 4 Turnover:

RHR Pump B is 00s for severe vibrations that occurred during surveillance testing and is tagged out for Maintenance investigation.

DG "A is in operation for the Monthly Diesel Generator Slow Start Operability Test per OP 4126, Sect B. This requires DG B being declared inop IAW T.S. 3.10.B.1 I/ 100% 120 I - SRO mfRR-11B I - RO 100% over I - SRO sec TS - SRO Critical Tasks: 1. With a reactor scram required and the reactor not shutdown, INHIBIT ADS to prevent an uncontrolled RPV depressurization to prevent causing a significant power excursion.

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.

60% over R - SRO I300sec. I I mfDG-O3A I - BOP Lower power OP 01 05, Reactor Operations (vibrations stop after stopping test and lowering generator load)

DG Voltage Regulator Malfunction requires removing DG "A from service. Declare DG "A inoperable and appropriate T.S. "B Recirc speed controller will fail and will begin to run away requiring the RO to take manual control and rebalance flows.

mfED 05D Loss of MCC-SA with a failure of the Group 3 Isolation (AC-6B will not auto close) Consult Tech Specs TS - SRO bendix D Scenario Outline Form ES-D-1 C - RO - SRO 6 SLC Squib Valve "A failure (the loss of 88 takes away SLC Pump B) Candidate must recognize that the Squib Valves must be fired locally using the battery. '7 mfED05C mfRP-02A mfRD-12A mfRD-128 (20/20%) m f SL-02A C - BOP C - SRO M- ALL 480V MCC-8B will trip causing a half scram (if RPS B was transferred to its alternate supply) also the loss of the bus will challenge DW pressure by the loss of power to RRUs 1A and 1 B, alternate RRUs must be started. (This power lost will affect ATWS recovery by preventing the use of cooling water flow to insert the control rods.) Loss of RPS MG Set A, Hydraulic ATWS with MSlV closure * (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor I Appendix D Scenario Outline Form ES-D-1 I Malf. No. NIA 11 Facility:

VERMONT YANKEE Scenario No.: 2 Op Test No.: 2009 NRC Event Event Description R-RO Type* Withdraw control rods to continue the startup R-SRO Exam iners: Operators:

SRO - RO - BOP - mfNM-O3C (1 00%) (1 Initial Conditions:

At -1% power, Startup in progress.

OP 0105, Phase 2.D Step 10 "A" IRM failed upscale during the startup and is bypassed.

RHR Pump "C" is 00s ~ I-RO I-SRO TS - SRO IRM "C" hop Failure, results in half scam.

Ibover : Indefinite LCO due to IRM "A" 00s (TS Table 3.1.1 and TRM 3.2.5) RHR Pump "C" is 00s for severe vibrations during surveillance testing. Tagged out on previous shift; estimated return to service is 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, 7-day LCO due to RHR Pump "C" 00s (TS 3.5.A.3) MSlV Isolation Testing is NOT required ll ~ mfHP-03 m f H P-04 Critical Task: 1. 2. Following a Loss of Normal Power diagnoses "B DG failed to auto-start and manually starts "B DG and places on 4KV Bus 4. With the reactor shutdown and reactor pressure greater than the shutoff head of the low pressure systems, initiate RPV-ED BEFORE RPV level reaches -1 9 inches Restore and maintain RPV level above TAF (+6 inches) I 3. C-BOP C-SRO HPCl inadvertently injects to the vessel with a controller failure (low - to prevent a reactor scram).

The crew will confirm that the HPCl injection is spurious and trip the HPCl turbine. Requires T.S. 3.5.E entrv TS-SRO Event No. 1 2 3 4 5 6 7 mfDG-05A mfDG-08B C - BOP C-SRO be manually started.

Both DGs fail to auto start, DG "A cannot be started, DG "6 can NIA Transfer Reactor Mode Switch to RUN and continue the startup. 1;:::: 1 MfRD-15 Failure of CRD Flow Controller Automatic Output Signal mfED-02A 1 M-ALL 1 Loss of the startup transformers which will result in a LNP and mf ED-026 reactor scram.

Eendix D Scenario Outline Form ES-D-1 I mfRR-01A OVRD ANN 8 M - ALL Core spray line "B break in the Drywell between the RPV and injection check valve resulting in a LOCA and loss of the remaining Core Spray system.

9 "B Loop RHR Pump "D' trips I C-ALL I RHO1 D * (N)orm al, (R)eactivity, (I)nstrument, (C)om ponent , (M)ajor I Appendix D Scenario Outline Form ES-D-1 I 1 2 3 Faci I i ty: VERMONT YANKEE Scenario No.: 3 Op Test No.: 2009 NRC Examiners: Operators:

SRO - RO - BOP - N/A N/A mfFW-28A 50% over 60 sec mfRH-O1B (if necessary)

Initial Conditions:

85% power IC-807 "A RHR 00s for severe vibration maintenance is investigating.

N - RO - SRO R - RO Turnover:

RHR Pump "A" is 00s while maintenance investigates high vibration.

It was tagged out on the previous shift; estimated return to service is 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, 7-day LCO due to RHR Pump "A" 00s (TS 3.5.A.3) Swap CRD Pumps, place CRD "B in service and remove CRD "A and from service to allow cooling for oil change scheduled for the next shift. Swap CRD Pumps, place CRD B in service and remove CRD A and from service Raise power using recirculation flow to 92% 1. When drywell temperature cannot be restored and maintained below 2808F, Critical Tasks: initiate RPV-ED (and/or anticipate ED and use bypass valves).

SRO - TS 2. IF Reactor water level cannot be determined, Enters EOP-6 , RPV Flooding, opens all SRVs and commences injection using Shutdown RPV Flooding Systems until the Main Steam lines are flooded.

OR Call from the Southeast RHR Corner Room that RHR Pump "B' lower motor bearing oil indicating sight glass is empty. Tech Specs Sections 3.5 and determine that a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> LCO, requiring a plant shutdown, has been entered. 3. Restores RPV water level and containment parameters with Condensate injecting directly into the RPV AND/OR aligns alternate injection systems such as RHR. Event Type* I Event Description R-SRO I ~ ~~ Feedwater flow transmitter slow failure upscale, causing the crew to take manual control of feedwater in order to recover and stabilize RPV level.

I Appendix D Scenario Outline Form ES-D-1 I IOR C - BOP RRlo042AS7B c - SRO IOR RRdi042AS7B IOR RRlo042AS7B IRF rfRR-12 mfEDO6A C-RO C-SRO TS - SRO mfTC-04A I - BOP I - SRO mf MS-06 M - CREW mfRN-08A C - BOP mf FW-08B c - SRO mf FW-08C mfCS-O3A I "B Recirc Pump discharge valve full open indication fails causing the "B Recirc Pump to runback to minimum flow.

Trip of CRD Pump B with a loss of 125 VDC Bus 1 (Inops 4 KV Bus 3 ECCS equipment). The crew will implement ON 3159, Loss of Bus DC 1. The SS/SRO will review Tech Specs Sections 3.10 and 3.5 and determine that a second 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> LCO, requiring a plant shutdown, has been entered. (STG 04)

EPR Oscillations OT 31 15, Reactor Pressure Transients - Place MPR in service.

II Main Steam Line Break in the Drywell Failure of the Reactor Feedwater Pumps and Core Spray Pump A Injection Valve require lining up Condensate Pumps to restore RPV water level.

I Appendix D Scenario Outline Form ES-D-1 I I Event No. 1 2 3 4 5 Faci I i ty: VERMONT YANKEE Scenario No.: 4 Op Test No.: 2009 NRC Examiners: Operators:

SRO - RO - BOP - Initial Conditions:

80% power (A and C RFP in service) "C RHR 00s for severe vibration maintenance is investigating.

c Turnover : RHR Pump "C" is 00s while maintenance investigates high vibration. Tagged out on previous shift; estimated return to service is 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, 7-day LCO due to RHR Pump "C" 00s (TS 3.5.A.3) Perform VYOPF 41 20.02, HCI Valve Tests for HPCl Steam Isolation Valves 15 and 16. HPCl Valve surveillance is due. 1. During an RPV Blowdown with the reactor not shutdown under all conditions perform power/level control TERMINATE AND PREVENT INJECTION into the RPV using appendix GG, until conditions are met to re-establish injection.

Critical Task: 2. During an ATWS with a steam leak in the Secondary Containment and two areas exceeding MAX SAFE Emergency Depressurize.

3. After Emergency Depressurization restore and maintain RPV water level +127 to +177 inches.

Malf. No. N/A N/A mf RD-021423 mfFW-16A mfRM-01 P (1 00%) Event Type* N - BOP N - SRO R-RO R - SRO C-RO C - SRO I - RO I - SRO TS - SRO I - BOP I - SRO TS - SRO Event Description Perform VYOPF 4120.02, Valve timing on HPCl Steam Isolation Valves 15 and 16. (TS) Lower power to remove the "A Condensate Pump from service While inserting control rods a control rod will stick in the core after moving a notch, CRD drive pressure must be raised to move rod (ON 2143). Failure of the controlling reactor water level instrument LI-6-94A, will result in a lowering reactor water level (OT 31 13) and a TS entry. Failure of the Refuel Floor High Radiation Monitor.

RP and I 8, C will immediately repair (After TS Call) The crew must restore the GP 3 isolation. (TS)

Appendix D NUREG 1021 Revision 9 Eendix D Scenario Outline Form ES-D-1 I mf MC-08 sec. 10% at 120 6 - Bop Respond to loss of condenser vacuum /Off-gas explosion C - SRO 7 mfOG-O3A mfHP-09 25% at 450 sec - CREW HPCl Steam Line Leak upstream of HPCI-14, this will result in Emergency Depressurization as Reactor Building temperatures in two areas are above Max Safe.

Crew may anticipate RPV-ED. 8 mfPC-1 HP15 mfPC-1 HP16 m f R D-0206 1 5 mfRD-020611 mfRD-021011 mfRD-021015 C-RO C - SRO HPCI-15 and 16 will fail to isolate and four control rods fail to insert during the scram. Entry into EOP-2 should have occurred earlier on the scram caused by the loss of vacuum. * (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Appendix D NUREG 1021 Revision 9

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