ML061800354

Final - SRO Written Examination with Answer Key (401-5 Format) (Folder 3)

ML061800354
Person / Time
Site:	FitzPatrick
Issue date:	06/14/2006
From:	Conte R Operations Branch I
To:	Ted Sullivan Entergy Nuclear Northeast
Sykes, Marvin D.
References
50-333/06-301 50-333/06-301
Download: ML061800354 (61)

Text

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet QUESTION # 1 Examination Outline Cross-reference:

Partial or Complete Loss of AC / 6 Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF A.C. POWER: (CFR: 41.10/ 43.5 /45.13) Level Tier # Group # WA # 295003 0 System lineups Proposed Question: Importance Rating The plant is at 100% power with the following conditions:

SRO 1 1 AA2.04 3.7 - FW pumps are in 3-element control on the Master Feedwater Controller. - 125 VDC Station Battery B is on equalize charge per OP-43A, 125 - UPS M-G set is on the DC Drive to "run-in" the DC motor brushes - EDG B and D are running unloaded per ST-9R, EDG System Quick- VDC Power System. under load per OP-466, 120 VAC Power System. Start Operability Test and Offsite Circuit Verification.

Subsequently, the following valid annunciator alarm is received:

09-8-4-18, L26 600 V SUPP FDR BKR 12602 TRIP All plant equipment responds per design. Which ONE of the following is the PROMPT action directed by the CRS as a result of these plant conditions?

a> b) c) d) Proposed Answer: Take manual control of one FW pump on the MSC per AOP-21, Loss of UPS. Lock the RWR Scoop Tubes to prevent a run back per AOP-21, Loss of u PS. Reduce Station Battery B loads per AOP-ISB, Loss of Switchgear L26. Shutdown EDG B and D per AOP-lgB, Loss of Switchgear L26. d) Shutdown EDG B and D per AOP-19B, Loss of Switchgear L26. Explanation (Optional):

Justification:

Power is lost to L26 resulting in a loss of ESW pump B which supplies cooling water to the EDGs that were running.

The UPS is still powered by the A DC battery system and the actions provided in the distracters A & 6 are not a prompt action to be taken for the plant conditions provided in the stem. Reducing DC loads is a subsequent action per A0 P- 1 9 B. 1 of 61 Rev3 ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet Comments:

QUESTION # 1 Continued v Plant Event 41987, JAF-CR-2005-3818, Momentary loss of UPS results in a RX Scram was used in part to supply the distracters for this question.

Distracters:

a) Take manual control of one FW pump on the MSC per AOP-21, Loss of UPS is incorrect::

Take manual control of one FW pump on the MSC per AOP-21, Loss of UPS. This action is part of the response to a momentary loss of the UPS, per the plant conditions provided in the stem, the UPS is still powered from the DC drive. When a subsequent step is taken to transfer the UPS to the alternate AC source, this would be part of the actions to take. b) Lock the RWR Scoop Tubes to prevent a run back per AOP-21 , Loss of UPS is incorrect:: Lock the RWR Scoop Tubes to prevent a run back per AOP-21, Loss of UPS. Per the stem, the UPS is still powered from the DC Drive and has not lost power. c) Reduce Station Battery B loads per AOP-ISB, Loss of Switchgear L26 is incorrect:

This is a subsequent action to the conditions provided in the stem. Technical Reference(s): AOP-19B1 AOP-21. (Attach if not previously provided)

Proposed references to be provided to applicants during examination:

NONE Learning Objective:

SDLP-71 E EO-I .09.C (As available) Question Source:

Bank # Modified Bank # (Note changes or attach parent) New X 2 of 61 Rev3 ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet QUESTION # 2 w Examination Outline Cross-reference:

Level Partial or Total Loss of DC Pwr / 6 Tier # Ability to determine andlor interpret the Group # WA # 295004 following as they apply to PARTIAL OR COMPLETE LOSS OF D.C. POWER: (CFR: 41.101 43.5 / 45.13) 0 Extent of partial or complete loss of importance Rating D.C. power Proposed Question:

a) b) c) d) Proposed Answer: SRO 1 1 AA2.02 3.9 The plant is at 100% power. Feedwater Pump A

& B are in 3 element control selected to RX WTR LVL COLUMN "A". There are NO evolutions in progress when the following are noted:

Time 0: - Annunciator 09-8-1-21, 125VDC BATT CHGR A DC GRD - 125VDC Bus A GND DET meter on Panel 09-8 indicates

+25 volts and Time 1 minute: - Annunciator 09-8-1-22,125VDC BATT CHGR 8 AC SUPP TROUBLE Time 4 minutes: - Annunciator 09-8-1-23, 125VDC BATT - 125VDC Bus 1 19VDC Which ONE of the following identifies the resultant equipment status - AND procedure used to respond to the above indications and alarms? AC POWER breaker at 71BC-1A 125V DC BATTERY CHARGER 4 tripped, AOP-45, Loss of DC Power System 71 BC-I B 125V DC BATTERY CHARGER 262-OAI, AOP-46, Loss of DC Power System RX VVTR LVL 06LI-946 indicates downscale, AOP-41 , Feedwater Malfunction (Rising Feedwater Flow-High RPV Level) RHR B initiation logic is inoperable, AOP-22, DC Power System 4 Ground Isolation 7lMCC-262-OA1, AOP-46. Loss of DC Power System B steady VOLT LO Output Voltage meter on Panel 09-8 indicates breaker tripped at 71MCC- b) 71 BC-1 B 125V DC BATTERY CHARGER B breaker tripped at Explanation (Optional):

Justification:

See ARP-09-811-22 causes & step 2, 2nd bullet, AOP-46, Loss of DC Power System B see symptom A-first bullets first dash- 09-8-1-22 is listed as 1 or more of the following annunciators in alarm. 3 of 61 Rev3 ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet QUESTION ## 2 Continued Distracters:

a) AC POWER breaker at 71 BC-IA 125V DC BATTERY CHARGER A tripped is incorrect there is only a small ground on DC Bus A see ARP- 09-8-1-21 this also means AOP-45, Loss of DC Power System A is an incorrect procedure to enter. c) RX VVTR LVL 06LI-94B indicates downscale is incorrect battery is supplying the bus at 119 VDC while this is a symptom of AOP-46, Loss of DC Power B, the voltage is still acceptable for the indicator to be normal. Conditions would NOT require entry into AOP-41, Feedwater Malfunction (Rising Feedwater Flow- High RPV Level) as the stem stipulates WTR Column "A" is selected.

d) RHR B initiation logic is unaffected by the loss of B DC. RHR A logic is powered by B DC. Technical Ref e re nce (s) : AOP-46, AOP-41 I AOP-45, (Attach if not previously provided) AOP-22, OP-46B - ARPs: 09-8-1-19, 09-8-1-21, 09-8-1 -22. 09-8-1 -23. Proposed references to be provided to applicants during examination:

Learning Objective: SDLP-71B EO- 1.10.A.I (As available) Question Source:

Bank # NONE Modified Bank # (Note changes or attach parent) New X Question History:

Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.) Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Assessment of Facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Comments:

4 of 61 Rev3 ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet QUESTION ## 3 4' Examination Outline Cross-reference: Main Turbine Generator Trip / 3 Knowledge of EOP terms and definitions. (CFR: 41.10/ 43.5 / 45.13) Level Tier # Group ## WA # 295005 Importance Rating SRO 1 1 G 2.4.17 3.8 Proposed Question: The plant is at 100% power. I&C is working on 06LT-52Cl Reactor Water Level Feedwater Control Level Transmitter when the following indications are noted on Panel 09- 5: - RX WTR LVL HI CHNL

'A' Amber Light is 'ON', - RX VVTR LVL HI CHNL 'B' Amber Light is 'E, - RX WTR LVL HI CHNL 'C' Amber Light is 'OJ. A correct automatic action occurs due to these indications. The CRS directs insertion of a manual scram and the SNO reports all rods in with the exception that: - - 3 rods are at position 02 1 rod is at position

48. The CRS initially directs entry into AOP-1 (Reactor Scram) and the applicable EO P( s) . The automatic action that occurred was a must direct actions in conditions without boron. Main Turbine, EOP-2 (RPV Control) because the reactor WILL HPCl Pump, EOP-2 (RPV Control) because the reactor WILL Main Turbine, EOP-3 (Failure to Scram) because the reactor will NOT HPCl Pump, EOP-3 (Failure to Scram) because the reactor will NOT a) Main Turbine, EOP-2 (RPV Control) because the reactor WILL Justification: Indications are directly part of Main Turbine 81 Nv Pump trip circuitry. Per EP-1, EOP Entry and Use, section 4.7.2, the reactor will remain shutdown under all conditions without boron with one rod at 48 if all other rods are at 02 (or inserted). trip. The CRS remain shutdown under all 4 b) c) d) Proposed Answer: Explanation (Optional):

5 of 61 Rev3 E S-40 I Sample Written Examination Form ES-401-5 Question Worksheet QUESTION # 3 CONTINUED Distracters:

b) HPCl Pump trip, EOP-2 (RPV Control) because the reactor WILL Justification: Indications are NOT part of HPCl trip circuitry.

The second part of the distractor is correct..

c) Main Turbine Generator trip, EOP-3 (Failure to Scram) because the Justification:

The first part is correct but the reactor will remain shutdown.

d) HPCl Pump trip, EOP-3 (Failure to Scram) because the reactor will NOT Justification:

The first part is incorrect, the second part is correct. reactor will NOT Technical Ref e rence( s) : EOP-2, EOP-3, EP-1 (Attach if not previously provided) ~ ~~ ~ Proposed references to be provided to applicants during examination:

EOP 2 and 3 Learning Objective:

MIT-301.11A EO- 1.04.i (As available) Question Source: Bank

1. 1. Modified Bank # (Note changes or attach parent) New X Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 5 Assessment of Facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. .J Comments:

6 of 61 Rev3 3 ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet - I QUESTION #4 / Examination Outline Cross-reference:

Refueling Acc / 8 Ability to apply technical specifications for a system. (I OCFR 55.43,2f4W?)

Level Tier # Group # WA # 295023 Importance Rating . . . . . . . . . . SRO 1 1 G 2.1.12 4.0 Proposed Question: The plant has been shutdown for 2 days and is being refueled.

An irradiated fuel bundle is being moved from the core to the spent fuel storage pool.

The bundle is over the core and being moved towards the fuel pool when level drops to 22 feet above the RPV flange and then continues to slowly decrease.

The shift declares a radiological emergency per RAP-7.1.04B, Refueling Procedure.

i. The Technical Specification bases for maintaining a minimum water level dver the flange is to insure ii. The Refuel Bridge SRO must direct the bundle to be placed in any empty location in the i. RHR Shutdown Cooling can maintain "Time to Boil" limitations ii. core or the Fuel Pool storage rack

i. RHR Shutdown Cooling can maintain "Time to Boil" limitations

i. iodine release from the design refueling accident is retained by the water and off site doses are maintained within limits, ii. core or the Fuel Pool storage rack i. iodine release from the design refueling accident is retained by the water and off site doses are maintained within limits, ii. Fuel Pool storage rack only a) b, ii. Fuel Pool storage rack only c) d) Proposed Answer: d) i. iodine release from the design refueling accident is retained by the water and off site doses are maintained within limits, ii. Fuel Pool storage rack only Justification:

In RAP 7.1.04B, section 7.3, the procedure allows, if a radiological emergency exists, for the bundle to be placed in an empty spent fuel rack location.

Without this emergency]

the bundle, if it can not be placed in its target location shall be returned to its prior location. Explanation (Optional):

7 of61 Rev4 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet QUESTION # 4 Continued Distracters:

a) i. RHR Shutdown Cooling can maintain "Time to Boil" limitations ii. core or the Fuel Pool storage rack b) i. RHR Shutdown Cooling can maintain "Time to Boil" limitations ii. Fuel Pool storage rack only c) i. iodine release from the design refueling accident is retained by the water and off site doses are maintained within limits, ii. core or the Fuel Pool storage rack Justification for incorrect answers: The incorrect portion of the distracters are "RHR Shutdown Cooling can maintain "Time to Boil" limitations" and the allowance to be able to store the fuel in any core location. The level over the flange affects the ability of the water to adsorb heat and the time to boil (plausible distractor). However, the actual reason for the level is for iodine releases. The fuel is allowed to be stored back into the core per RAP 7.1.04B but it must be returned to its prior location, not ANY location in the core. (Attach if not previously provided) c' Proposed references to be provided to applicants during examination:

NONE Learning Objective:

SDLP-08B EO- 1.17.a (As available) Question Source:

Bank # Modified Bank # (Note changes or attach parent) New X Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.) Question Cognitive Level:

Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 8 of 61 Rev 4 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet QUESTION # 4 Continued 55.43 2 2 -Facility operating limitations in the technical specifications and their bases. normal and abnormal situations, including maintenance activities and various contamination conditions.

6 -Procedures and limitations involved in initial core loading, alterations in core configuration, control rod programming and determination of various internal and external effects on core reactivity. 7-Fuel handling facilities and procedures.

4 -Radiation hazards that may arise during 4 6 7 Comments:

9 of61 Rev4 ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet QUESTION # 5 Examination Outline Cross-reference:

High Drywell Pressure

/ 5 Ability to determine and/or interpret the following as they apply to HIGH DRYWELL PRESSURE:

(1 OCFR 55.43.5) 0 Suppression pool temperature Level Tier # Group # WA # 295024 Importance Rating SRO 1 1 EA 2.06 4.1 Proposed Question:

The initial plant indications were: - Reactor Power 100% - Torus Water Temperature 83 OF - Torus pressure 0 psig - Drywell pressure 1.91 psig - Safety Relief Valve "A inadvertently opens.

IO Minutes later plant indications are: - Torus Water Temperature 83 OF - Torus pressure 11.40 psig - Drywell pressure 10.90 psig a) b) c) d) Proposed Answer: The above primary containment readings indicate that the suppression function is (1 1 and one of the procedures that the CRS is (1) working correctly, (2) AOP-36, Stuck Open Relief Valve(s). (I) NOT working correctly, (2) AOP-1, Reactor Scram. (I) working correctly, (2) AOP-39, Loss of Coolant. (I) NOT working correctly, (2) AOP-9, Loss of Primary Containment Integrity. to implement is (2) b) (I) NOT working correctly, (2) AOP-1, Reactor Scram.

Explanation (Optional):

Justification: With the opening of the SRV torus temperature remains constant but both torus pressure and pressure rise. Torus pressure is 0.5 psig higher than drywell pressure which indicates that the torus is pressurizing and lifting the torus to drywell vacuum breakers.

The high DW pressure caused a scram and AOP-1 entry. 10 of 61 Rev3 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet

'4 QUESTION # 5 Continued Distracters: a) (1) working correctly, (2) AOP-36, Stuck Open Relief Valve(s).

Justification:

The lack of a torus temperature rise with both a torus and DW press rise indicates bypass of torus pressure suppression function.

c) (1) working correctly, Justification:

The lack of a torus temperature rise with both a torus and DW press rise indicates bypass of torus pressure suppression function.

(2) AOP-9, Loss of Primary Containment Integrity.

(2) AOP-39, Loss of Coolant. d) (1) NOT working correctly, Justification: Although Primary Containment is not functioning properly, its integrity remains intact and the entry conditions are NOT met for AOP- 9. Technical Reference(s):

AOP-1, AOP-9, AOP-36, AOP- 39 (Attach if not previously provided) Proposed references to be provided to applicants during examination: Question Source:

Bank # NONE Learning Objective: SDLP-16A EO- 1.09.e & f (As available)

Modified Bank # (Note changes or attach parent) New X Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.) Question Cognitive Level:

Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 1 1 of 6 1 Rev3 ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet QUESTION # 5 Continued 55.43 5 - Assessment of Facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

5 Comments:

12 of61 Rev3 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet QUESTION # 6 -' Examination Outline Cross-reference: Level SRO High Reactor Pressure / 3 Tier # 1 Ability to determine and/or interpret the Group # 1 0 Suppression pool level Importance Rating 3.9 following as they apply to HIGH REACTOR PRESSURE: (IOCFR 55.43.5) Proposed Question:

wA # 295025 EA 2.04 The initial plant indications were: - Reactor Power 100% - RPV Pressure 1040 psig - RPV Water Level 201.5 inches - Torus Water Level 13.96 feet - Torus Water Temperature 83 OF - Drywell pressure 1.87 psig - Steam Tunnel Temperature 120 OF With NO operator actions 10 Minutes later plant indications are: - Reactor Mode Switch RUN - RPS A 8, B Scram groups lights ON - ARI Valves are OPEN - RPV Pressure - RPV Water Level - Torus Water Level - Torus Water Temperature - Drywell pressure - Steam Tunnel Temperature A low of 800 psig - slowly rising A low of 150" - slowly rising 14.12 feet - steady 95 OF - steady 1.87 psig - steady 140 OF - steady Which one of the following caused the increase in Suppression Pool level and which procedure must the CRS implement?

a) b, Feedwater Flow).

c) d) A small break LOCA inside the drywell, AOP-39, Loss of Coolant. A low vessel level, AOP-42, Feedwater Malfunction (Lowering A high RPV pressure, AOP-1, Reactor Scram.

A small main steam line break inside the steam tunnel, AOP 40, Main Steam Line Break.

Proposed Answer: c) A high RPV pressure, AOP-1, Reactor Scram.

13 of 61 Rev3 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet QUESTION # 6 Continued Explanation (Optional):

Justification:

high RPV pressure (determined by ARI valves indications open) resulted in SRV operation and subsequent ARI rod insertion.

RPV level was low enough to cause a scram signal at 177" but an AWS occurred as evidenced by the scram lights being on. This requires entry into AOP-1. A small break LOCA inside the drywell, AOP-39, LOSS Of Coolant. A low vessel level, AOP-42, Feedwater Malfunction (Lowering A small main steam line break inside the steam tunnel, AOP 40, Distracters:

a) b) Feedwater Flow). d) Main Steam Line Break. Justification:

a) There is no evidence of a DW leak. DW pressure remains normal 10 minutes into the event. b) A low vessel level and entry into AOP-42 is appropriate but would not cause a signal to be generated to lift the SRVs which in turn would cause the high torus level. d) A break in the steam tunnel could cause the MSlVs to go close and, with an ATWS, would cause SRV operation. However, there is no steam tunnel temperature isolation as evidenced by normal temperatures.

Technical Ref ere nce (s) : A0 p- 1 I AOP-36 AOP-39, (Attach if not previously provided)

AOP-40, Proposed references to be provided to applicants during examination:

Lea rn i ng 0 bject ive : SDLP- 29 EO- 1.09.b (As available)

Question Source: Bank # NONE Modified Bank # New X (Note changes or attach parent)

Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question .) Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 14 of61 Rev3 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet QUESTION # 6 Continued 10 CFR Part 55 Content: 55.41 55.43 5 - Assessment of Facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

5 Comments:

15 of 61 Rev3 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet QUESTION ## 7 ~-2 4 Examination Outline Cross-reference:

Reactor Low Water Level / 2 Knowledge of the process for performing a containment purge.

(10CFR 55.43.4) SRO Level Tier # Group # WA # 295031 Importance Rating Proposed Question:

The Plant is in Mode 4. To support a maintenance activity, a vent and purge of the drywell is being established per OP-37, Containment Atmosphere Dilution System. 1 1 G 2.3.9 3.4 To ensure a purge can be established the be reset and, to minimize off-site releases, the CRS must direct the Low RPV Level (177 inches), Standby Gas Treatment system, trip signal must be used for the purge.

a) High RPV Pressure (1080 psig), Standby Gas Treatment system, High RPV Level (222.5 inches), Drywell Ventilation and Cooling System, High Drywell pressure (2.7 psig), Drywell Ventilation and Cooling System, b) c) d) Proposed Answer: Explanation (Optional):

a) Low RPV Level (177 inches), Standby Gas Treatment system, Justification:

LOW RPV level isolates the purge valves and SGT has HEPA and charcoal filters to remove particulates and gaseous radioactive material.

16 of 61 Rev3 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet QUESTION # 7 Continued Distracters:

b) High RPV Pressure (1080 psig), Standby Gas Treatment system, c) High RPV Level (222.5 inches), Drywell Ventilation and Cooling System, d) High Drywell pressure (2.7 psig), Drywell Ventilation and Cooling System, JUSTIFICATION:

Hi DW pressure is the only signal in these 3 distracters that isolates Containment purge. Drywell Ventilation and Cooling does not limit or reduce any airborne activity.

AOP-15, o p-37 Techn ica I Ref e re n ce (s) : (Attach if not previously provided)

Proposed references to be provided to applicants during examination:

Learning Objective: SDLP- 16C EO- 1.09.~ (As available)

NONE Question Source: Bank # Modified Bank # (Note changes or attach parent)

New X Question History:

Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level:

Memory or Fundamental Knowledge X Comprehension or Analysis 55.41 55.43 4 Radiation hazards that may arise during 10 CFR Part 55 Content: normal and abnormal situations, including maintenance activities and various contamination conditions.

Comments:

17 of 61 Rev3 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet QUESTION # 8 Examination Outline Cross-reference:

High Reactor Pressure

/ 3 Ability to locate and operate components

/ including local controls.

(1 OCFR 55.43.5) Level Tier # Group ## WA # 295007 Importance Rating . . . . . . . . . . . . . . . . Proposed Question:

a) b) c) d) Proposed Answer: Given the following plant conditions:

Time 0: Reactor pressure - 1039 psig APRM power - 100% Recirc Pump 'A' speed Recirc Pump 'B' speed Load Limit Limiting Light - 86% - 87% - OFF #4 Turbine Control Valve Turbine Bypass Valves - closed - 40% open Time + 3 minutes: Reactor pressure - 1047 psig APRM power - 105% Recirc Pump 'A' speed Recirc Pump 'B' speed Load Limit Limiting Light -ON - 94% - 87% #4 Turbine Control Valve #I Turbine Bypass Valve - FULL open - 65% open SRO 1 2 G 2.1.30 3.4 Which one of the following actions must the CRS direct to exit ALL active LCOs and return Reactor pressure and power to normal? Run Load Limit up until the Turbine Bypass valves close.

Run Load Set up until the Turbine Bypass valves close. Reduce Recirc Pump 'A' speed locally at the scoop tube. Reduce Recirc Pump 'B' speed at panel 09-4.

c) Reduce Recirc Pump 'A' speed locally at the scoop tube.

18 of 6 1 Rev3 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet 4 QUESTION # 8 Continued Explanation (Optional):

Justification: Recirc pump

'A' runaway has resulted in Recirculation flow mismatch and entry into LCO 3.4.1. Correct answer c will restore Recirc pump 'A' speed locally and bring the recirculation mismatch within limits. Distracter d will result in making the mismatch greater. Distracter a & b will results in a raise in reactor pressure.

Distracters Just if ication : Technical Reference(s):

LCo 3.4.1, Op-27, AOP-32, (Attach if not previously provided)

RAP-7.3.16.

OP-9 Proposed references to be provided to applicants during examination: Learning Objective:

SDLP-021 EO- 1.13. Given the (As available) procedure, discuss the procedure steps, administrative limitations, precautions, or cautions for the following: c.- operate the scoop tube positioner using the hand crank (OP-27). NONE Question Source: Bank # Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level:

Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 5 Assessment of Facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Comments:

W' 19 of 61 Rev3 ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet QUESTION # 9 W' Examination Outline Cross-reference:

High Reactor Water Level

/ 2 Ability to determine and/or interpret the following as they apply to HIGH REACTOR WATER LEVEL: (IOCFR 55.43.5) 0 Heatup rate: Plant-Specific Reactor Scram from Proposed Question:

Level Tier # Group # WA # 295008 Importance Rating 100% power has just SRO 1 2 AA 2.04 3.3 occurred 5 minutes ago. Current post scram plant conditions are as follows: - RPV Water level dropped to 160 inches and was recovered rapidly to 220 inches - Feedwater/HPCI/RCIC injection is secured to RPV - RPV Pressure is 800 psig with a trend up at 10 psig per minute - EHC pressure set is at 970 . psig - Main Turbine Bypass Jack set is at 0% demand With no operator action, over the next 5 minutes, RPV Water level will entry and actions of . To address the above conditions, the CRS must direct 4 b) c) d) Proposed Answer: Lower due to cooldown, AOP-1 "Reactor Scram" Rise due to swell from an open Safety Relief Valve, OP-I "Main Steam System" Lower due to shrink from an open Main Turbine Bypass Valve, AOP-6 "Malfunction of EHC Pressure Regulator" Rise due to heatup, EOP-2 "RPV Control" d) Rise due to heatup, EOP-2 "RPV Control" Explanation (Optional):

Justification: The overfeeding upon the initial swam caused a cooldown and pressure reduction.

With feed terminated, decay heat is causing the water to heat and the reactor to pressurize. Since level had dropped to less than 177, entry and actions of EOP-2 apply. Distracters:

a) Lower due to cooldown, AOP-1 "Reactor Scram" b) Rise due to swell from an open Safety Relief Valve, OP-I "Main Steam System" c) Lower due to shrink from an open Main Turbine Bypass Valve, AOP-6 "Malfunction of EHC Pressure Regulator

'I Just if ication : a) With no feed or steam being drawn, decay heat will cause a heatup. b) The rate of pressure rise over the next 5 minutes would 50 psig with total RPV pressure being 850 psig, less than the lift setpoint of an SRV. c) Reactor pressure will be 850 psig in 5 minutes which is less than the 970 psig setpoint of EHC. Since 970 psig is the normal setpoint, there is no reason to believe that the controller has malfunctioned.

20 of 61 Rev3 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet QUESTION # 9 Continued W Technical Reference(s):

AOP-1, op-1 AOP-6, EOP-2 (Attach if not previously provided) Proposed references to be provided to applicants during examination:

Learning Objective: SDLP-06 EO- 1.09.~ (As available)

NONE Question Source: Bank # Modified Bank # (Note changes or attach parent)

New X Question History:

Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Assessment of Facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

v Comments:

21 of61 Rev3 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet QUESTION # 10 Examination Outline Cross-reference:

Inadvertent Reactivity Addition / 1 Ability to determine and/or interpret the following as they apply to INADVERTENT REACTIVITY ADDITION:

(10CFR 55.43.6) 0 Reactor period Proposed Question:

a) b) c> d) Proposed Answer: Explanation (Optional):

To complete the core into the top guide of tl Level Tier # Group # K/A # 29501 4 importance Rating refuel, the last new ?e core at location 1 bundle of fuel is 7-34. SRO 1 2 AA 2.02 3.9 being lowered The fuel bucdie experiences binding which results in a Slack Cable indication.

Subsequently the bundle frees itself from the obstruction and quickly slides partway into the core till the Hoist Loaded indication is met. The bundle stops approximately 2/3rds of the way into the core with its full weight on the hoist.

INITIAL SRM INDICATIONS:

SRM 'A' SRM 'B' SRM 'C' SRM 'D' 30 CPS & 90 sec period 20 CPS & Infinite period 20 CPS & Infinite period 15 CPS & infinite period SRM INDICATIONS DURING BUNDLE DROP: SRM 'A' 65 CPS & 20 sec period SRM 'B' SRM 'C' SRM 'D' 45 CPS & 90 sec period 25 CPS & 25 sec period 25 CPS & 120 sec period Which SRM indications during the bundle movement incident require the evolution to be immediately stopped and the Refuel Bridge SRO to be notified per RAP-7.1.04.8, Neutron Instrumentation Monitoring During ln-Core Fuel Handling?

SRM 'A' & 'B'. SRM 'B' & C'. SRM 'C' & 'D'. SRM 'D' & 'A'. a) SRM 'A' & 'B'. Justification:

Per RAP-7.1.04.C Step 8.6 If loading fuel or withdrawing a control rod not immediately adjacent to a SRMIFLC AND count rate DOUBLES, THEN perform the following:

a ) immediately stop the evolution.

b) Notify Refuel Bridge SRO and SM. This limitation was met with SRM 'A' & '6' counts. 22 of 61 Rev3 ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet e QUESTION # 10 Continued Distracters:

SRM 'B' & '(2'. c) SRM 'C'&'D'.

d) SRM 'D' & 'A'. JUSTIFICATION: Per RAP-7.1.04.C Step 8.6 if loading fuel or withdrawing a control rod not immediately adjacent to a SRM/FLC AND count rate DOUBLES, THEN perform the following:

a) immediately stop the evolution.

b) Notify Refuel Bridge SRO and SM. This limitation was met with SRM 'A' & 'B' counts. Technical Ref e ren ce (s) : RAP-7-1 -04.c (Attach if not previously provided) ~~ ~~ ~ ~

Proposed references to be provided to applicants during examination:

SDLP-076 Figure # 2 Learning Objective:

SDLP- 076 EO- 1.12.d (As available)

Question Source: Bank

1. Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 6 Procedures

& limitations involved in initial core loading, alterations in core configuration, control rod programming

& determination of Comments:

various internal & external effects on core reactivity.

23 of 61 Rev3 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet QUESTION # 10 Continued 24 of 6 1 Rev3 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet QUESTION ## 11 ~4 Examination Outline Cross-reference: Level HPCl Tier # Knowledge of EOP layout / symbols / and icons. (1 OCFR 55.43.5) Group # WA # 206000 Importance Rating Proposed Question:

Regarding the following EOP-2 Step: SRO 2 1 G 2.4.19 3.7 The instructions to manipulate the controls for 23MOV-15 are contained in a(n) to Major Decision Point, continue and complete the step whenever the "IF" condition is met Override, continue and complete the step whenever the "IF" condition is met Action Statement, stop at this step and wait for the "IF" condition to be met before continuing Hold Point, stop at this step and wait for the "IF" condition to be met before continuing b) Override, continue and complete the step whenever the "IF" condition is met and, when following the flowchart, the CRS is 25 of 61 Rev3 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet QUESTION # 11 Continued Explanation (Optional):

Justification:

Per AP-02.02, EOP & SAOG, Step 5.7 this is an Override that provides guidance on the HPCl system among others. An override must be continuously evaluated during the execution of a series of procedure steps. Distracters:

a) Major Decision Point, continue and complete the step whenever the "IF" condition is met c) Action Statement, stop at this step and wait for the "IF" condition to be met before continuing d) Hold Point, stop at this step and wait for the "IF" condition to be met before continuing Justification:

a). Major Decision Points are enclosed in diamonds.

c) Action statements are simple direct instructions enclosed in rectangles d) Hold points are enclosed in octagons.

Technical Reference(s): (Attach if not previously provided)

AP-O*.O*l EOP-2, EOP-3, EOP- 4 & EOP-7 ~~ ~- ~ ~~~ Proposed references to be provided to applicants during examination:

EOP-2 Learning Objective:

MIT-301 .I IA, EO- 1.03.h Question Source: Bank

1. (As available)

Modified Bank # New X (Note changes or attach parent) Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question .) Question Cognitive Level:

Memory or Fundamental Knowledge X Comprehension or Analysis 55.41 55.43 5 Assessment of Facility conditions and 10 CFR Part 55 Content: selection of appropriate procedures during normal, abnormal, and emergency situations.

Comments:

26 of 61 Rev3 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet QUESTION # 12 'd' Examination Outline Cross-reference:

Level SRO SLC Tier # 2 Ability to analyze the effect of maintenance Group # 1 activities on LCO status.

(1 OCFR 55.43.2) wA 21 ., 000 G 2.2.24 Importance Rating 3.8 Proposed Question: The Plant is in Mode-I at 30% power during a startup.

Due to indications of bus overheating, L16 Bus was de-energized in preparation for corrective maintenance. Compensatory actions have been taken per AOP-19AI Loss of Switchgear L16. The following items supplied by this bus are being evaluated for Technical Specification LCO actions: - 1lP-25 6 SLC Pump 125FN-1B Standby Gas Treatment Filter Train B Fan Motor - 13MOV-15 RClC Steam Supply tnbd lsol Valve Which of the following is the Technical Specification required action to be taken regarding the evaluation of these three items? Restore SLC B subsystem in 7 days Restore SLC B subsystem in 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> a) b) c) Enter LCO 3.03 Immediately d) Be in Mode 3 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> with steam dome pressure e 150 psig in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> Proposed Answer: a) Restore SLC B subsystem in 7 days Explanation (Optional):

Justification:

Refer to TS 3.1.7 Action A Distracters:

b) Restore SLC B subsystem in 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> (Refer to TS 3.1.7 Action B for 2 SLC Inop- only 1 is inop for evaluation)

The loss of L16 also causes the loss of tank heater and heat tracing. However, temperatures are Tech. Spec limits and not the heaters.

c) Enter LCO 3.03 Immediately (Refer to TS 3.6.4.3 Action D for 2 SGTs inop- only one SGT is inop for evaluation) d) Be in Mode 3 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> with steam dome pressure < 150 psig in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> (Refer to TS 3.5.3 action B incorrect RClC LCO allows 14 days the Containment lsol Valve is inop as it is failed open, TS 3.6.1.3 Action A.l requires RClC Steam Line to be isolated in 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> which would then make RClC inoperable NOTE: this action was NOT provided as part of the distractor) Technical Reference(s): . AOP-19At TS-3.1-7 Action A B, TS- 3.6.4.3 Action D, TS- (Attach if not previously provided) 3.5.3 action A & B, TS- 3.6.1.3 Action A.l Proposed references to be provided to applicants during examination:

Tech Specs- No bases Learning Objective:

SDLP-11, EO- 1.16 (As available) 27 of 61 Rev3 Form ES-401-5 ES-40 1 Sample Written Examination Question Worksheet Question Source:

Bank # l./, Modified Bank # (Note changes or attach parent)

QUESTION # 12 Continued New X Question History:

Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level:

Memory or Fundamental Knowledge Comprehension or Analysis X IO CFR Part 55 Content: 55.41 55.43 2 Facility operating limitations in the tech specifications

& their bases. Comments:

28 of 61 Rev3 d ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet QUESTION # 13 Examination Outline Cross-reference: Level SRO RClC Tier

1. 2 Ability to (a) predict the impacts of the Group # 1 following on the REACTOR CORE WA # 217000 A 2.11 ISOLATION COOLING SYSTEM (RCIC);

and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (IOCFR 55.43.5) 0 Inadequate system flow Importance Rating 3.2 Proposed Question:

The Plant was at 100% when a Scram occurred. Reactor level is 120 and slowly decreasing.

RClC is injecting and has been running for 5 minutes with the following indications:

Proposed Answer: - RClC Flow CNTRL 13FIC RClC Room Temperature - TURB STM SUPP VLV 13MOV-131 - INJ VLV 13MOV MIN FLOW VLV 13MOV OIL CLR WTR SUPP 13MOV-132 - TEST VLV TO CST 13MOV INBOARD STEAM SUPPLY VLV 13MOV OUTBOARD STEAM SUPPLY VLV 13MOV 375 gpm - 100 deg. F - Open - Open - Open - Running - Open - Closed - Open - Open - VAC PMP 13P-3 With these indications it has been determined that RClC is NOT operating normally.

Which one of the following OP-19 "Reactor Core Isolation Cooling System" procedural sections must the CRS direct to correct this situ at ion? a) Isolation Verification and Recovery b) Man. Startup for RPV Pressure Control c) Auto-Initiation Verification and Subsequent Actions d) Manual Initiation Using Test Pot (Injection into RPV) c) Auto-Initiation Verification and Subsequent Actions. Explanation (Optional):

Justification:

RClC flow is < 410 gpm, the Only valve out Of position is the Min Flow Valve 13MOV-27 which is Open and must be shut. The auto-initiation procedure has the SNO verify the valves are in the correct position, The SNO will report the mispositioned valve and the CRS will direct its closure. 29 of 61 Rev3 ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet QUESTION ## 13 Continued Distracterst a) Isolation Verification and Recovery b) d) Man. Startup for RPV Pressure Control Manual Initiation Using Test Pot (Injection into RPV) Justification:

Choices 8 and D require the auto initiation signal of 126.5" to be clear before proceeding. Choice A does not correct the open MOW 27 valve and is not appropriate because RClC room temp. is less that the isolation setpoint (-133 deg. F) Technical Reference(s):

AP-02.01 , op-19. Drawing FM- 22A (Attach if not previously provided)

Proposed references to be provided to applicants during examination:

NONE Learning Objective:

SDLP-13, EO- 1.12.b (As available) Question Source:

Bank # Modified Bank # (Note changes or attach parent)

W New X Question History:

Last NRC Exam (Optional - Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.) Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Assessment of Facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Comments:

30 of 61 Rev3

,-z , ES-40 I Sample Written Examination Form ES-401-5 Question Worksheet

'Lj QUESTION # 14 Examination Out I i ne Cross-reference: Level SRVs Tier

1. Ability to (a) predict the impacts of the following on the RELIEF / SAFETY VALVES; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

(1 OCFR 55.43.5) 0 Stuck open vacuum breakers Importance Rating Proposed Question: While performing ST-ZZB, Manual Safety Valve Monitoring System Functional Test conditions were noted:

Group # WA # 239002 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . SRO 2 I A 2.01 3.3 Reiief Valve Operations and (IST), the following Plant - RHR is in full Torus Cooling - Suppression Pool Temperature is 90°F trending down - Annunciator 09-4-2-6 SRV Sonic Mon Alarm Hi is alarmed - SRV Sonic Mon Channel

'A' meter is just in the RED region - SRV 02RV-71A White Light is 'ON' on Panel 09 SRV 02RV-71A Control Switch is in 'Auto' on Panel 09 Torus Water Level is 13.9 feet and steady - Torus Pressure is 0.03 psig and steady - Drywell Pressure is 3.0 psig trending up - Drywell Temperature is 97OF trending up - Main Turbine Bypass Valves initially cycled closed about 10%. - Main Turbine Bypass Valves reopened approximately 7% when SRV when SRV 02RV-71A Control Switch was placed in 'Open'. 02RV-71A Control Switch was returned to 'Auto' from

'Open'. The crew has entered AOP-36 Stuck Open Relief Valve(s). Besides addressing the stuck open SRV, what other failure has occurred and what is the correct procedure to use? a) b, Containment Integrity c) d) SRV 02RV-71A vacuum breaker failed, EOP-4 Primary Containment Control SRV 02RV-71A vacuum breaker failed, AOP-9 Loss of Primary Turbine Bypass Valves failed, AOP-6 Malfunction of EHC Pressure Regulator Turbine Bypass Valves failed, EOP-2 RPV Control Proposed Answer:

a) SRV 02RV-71A vacuum breaker failed, EOP-4 Primary Containment Control 31 of61 Rev4

- - ES-401 Sample Written Examination Form ES-401-5 Question Worksheet QUESTION # 14 Continued L W Explanation (Optional):

Justification The SRV is stuck partially open, it is discharging directly into the DW through the SRV vacuum breakers as noted by DW press & temp increases

& it is NOT going into the TORUS as noted by Torus temp & pressure. The Main Turbine Bypass valves (BPV's) have responded to the change in SRV position, initially closing about 10% when the SRV was open & would be expected to re-open 10%

if the SRV went full shut, in this case they went open only 7% due to the SRV being partly open so they are responding correctly. EOP-4 is entered to mitigate containment challenges from direct pressurization due to the SRV Vacuum breaker being open with the SRV partially open. EOP-2 is a correct procedure to enter but when tied with the BPV's failure it is the wrong choice. AOP-6 is wrong as the EHC Pressure regulator has NOT malfunctioned.

AOP- 9 Loss of Primary Containment Integrity Entry conditions are NOT met. From SDLP-02J, A failure of the vacuum Breakers to close would admit steam to the DW air space, resulting in rising DW press & temp, upon subsequent SRV opening. Distracters:

b) SRV 02RV-71A vacuum breaker failed, AOP-9 LOSS of Primary c) Turbine Bypass Valves failed, AOP-6 Malfunction of EHC Pressure d) Turbine Bypass Valves failed, EOP-2 RPV Control Justification:

SRV is stuck partially open, it is discharging directly into the DW through the SRV vacuum breakers as noted by DW press & temp increases

& it is NOT going into the TORUS as noted by Torus temp

& pressure.

The Main Turbine Bypass valves (BPV's) have responded to the change in SRV position, initially closing about 10% when the SRV was open & would be expected to re-open 10% if the SRV went full shut, in this case they went open only 7% due to the SRV being partly open so they are responding correctly.

EOP-4 is entered to mitigate containment challenges from direct pressurization due to the SRV Vacuum breaker being open with the SRV partially open.

EOP-2 is a correct procedure to enter but when tied with the BPV's failure it is the wrong choice. AOP-6 is wrong as the EHC Pressure regulator has NOT malfunctioned. AOP-9 Loss of Primary Containment Integrity Entry conditions are met. From SDLP-02J, A failure of the vacuum Breakers to close would admit steam to the DW air space, resulting in rising DW press & temp, upon Subsequent SRV oDenina. Containment lnteg rity Regulator Technical Reference(s):

sT-22B1 Aop-6, Aep-91 AOP- (Attach if not previously provided) 36, EOP-2, EOP-4 Proposed references to be provided to applicants during examination: Learning Objective: SDLP-02J, EO- 1.09.f (As available)

None Question Source:

Bank # Modified Bank # (Note changes or attach parent)

New X 32 of 61 Rev 4

.- ES-401 Sample Written Examination Form ES-401-5 Question Worksheet QUESTION # 14 Continued Question History:

Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level:

Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

Assessment of Facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

33 of 61 Rev 4 ES-401 Sample Written Examination Form ES-40 1 -5 Question Worksheet QUESTION ## 15 c/ Examination Outline Cross-reference: Reactor Water Level Control Knowledge of the process for controlling temporary changes. (IOCFR 55.43.3) Level Tier # Group # WA # 259002 Importance Rating W d Proposed Question:

a) b) c) d) Proposed Answer: Explanation (Optional):

SRO 2 1 G 2.2.11 3.4 There is a tagout expected to be in place for greater than 90 days that tags out the "A" level column. In preparation for this tagout the

'B' level column is to be selected for Feedwater Level Control. The tagout is to support a proposed change to Technical Specifications to move TS 3.3.2.2 Feedwater and Main Turbine High Water Trip Instrumentation to the Technical Requirements Manual (TRM). The selection of the '6' Level Column and the associated tagout requires a High Water Trip Instrumentation to the TRM is controlled with Temporary Change to OP-2A Feedwater System, AP-02.04 Control of Procedures Temporary Change to OP-2A Feedwater System, AP-01.02 License and Technical Specification Administration 50.59 Screen per ST-IX Protected Tags and Temporary Alterations Audit, AP-02.01 Procedure Writers Manual 50.59 Screen per ST-1X Protected Tags and Temporary Alterations Audit, AP-20.06 Final Safety Analysis Report (FSAR) Amendment Preparation and Control . The addition of the Feedwater and Main Turbine d) 50.59 Screen per ST-1X Protected Tags and Temporary Alterations Audit, AP-20.06 Final Safety Analysis Report (FSAR) Amendment Preparation and Control 50.59 Screen is required for tagouts expected to be in place >30 days per ST-IX Protected Tags and Temporary Alterations Audit, AP-20.06 Final Safety Analysis Report (FSAR) Amendment Preparation and Control controls changes to the TRM. Temporary Change to OP-2A is - NOT required as it has a section (3.29 for swapping from Water Column 'A' to 'B'. Distracters:

a) Temporary Change to OP-2A Feedwater System, AP-02.04 Control of Procedures b) Temporary Change to OP-2A Feedwater System, AP-01.02 License and Technical Specification Administration c) 50.59 Screen per ST-IX Protected Tags and Temporary Alterations Audit, AP-20.06 Final Safety Analysis Report (FSAR) Amendment Preparation and Control Justification:

50.59 Screen is required for tagouts expected to be in place >30 days per ST-1X Protected Tags and Temporary Alterations Audit, AP-20.06 Final Safety Analysis Report (FSAR) Amendment Preparation and Control controls changes to the TRM. Temporary Change to OP-2A is NOT required as it has a section G.29 for swapping from Water Column

'A' to 'B'. 34 of 61 Rev3 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s):

AP-20.6, ST-Ix, AP-02.01 (Attach if not previously provided) AP-01.02, AP-02.04, TRM, TS-3.3.2.2 QUESTION ## 15 Continued Proposed references to be provided to applicants during examination:

NONE Learning Objective:

LP AP, EO- 1.01 Question Source:

Bank # ~ ~~ (As available)

Modified Bank # (Note changes or attach parent) New X Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.) Question Cognitive Level:

Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 3 Facility licensee procedures required to obtain authority for design and operating changes in the facility.

Comments:

35 of61 Rev3

'd ES-401 Sample Written Examination Form ES-401-5 Question Worksheet QUESTION # 16 Examination Outline Cross-reference:

Level Control Rod and Drive Mechanism Tier

1. Ability to (a) predict the impacts of the Group # following on the CONTROL ROD AND WA # 201 003 DRIVE MECHANISM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

(1 OCFR 55.43.5) Stuckrod Importance Rating Proposed Question: A plant Startup and heatup is in progress SRO 2 2 A 2.01 3.6 with'RPV pressure at 600 psig. The following conditions were noted when the ROD MOVEMENT CNTRL switch was taken to "Out Notch" to move the selected rod to position 12: 09-5-2-1 RWM ROD BLOCK RPlS INOP - clear 09-5-2-2 ROD WITHDRAWAL BLOCK - clear 09-5-2-3 ROD DRIFT - clear 09-5-2-4 ROD OVER TRAVEL - clear Rod 22-39 indicating light on Full Core Display - "ON" - "ON" Rod In Green light - cycled "ON" and "OFF' Rod Out Red light - cycled "ON" and "OFF" Rod Settle Amber light - cycled "ON" and "OFF" IRMs - all mid scale Range 8 and steady with no change Panel 09-5 Indications: 03PDI-302 CHG VVTR Press 03PDI-303 DRV WR Diff Press 03PDI-304 CLG WTR Diff Press 03PDI-306 CLG WTR Flow Local Indications:

03FI-216 Stab Valves A & B Outlet Flow Ind - 6 gpm Rod Out Perm light Rod 22-39 position 1500 psig - 650 psid - 21 psid - 60 gpm 03PDI-305 DRV VVTR Flow -0gpm The impact to the plant and equipment is enter and the CRS is to a) damage to the drive mechanism seals, AOP-24 Stuck Control Rod b) over-heating the drive mechanism seals, AOP-24 Stuck Control Rod c) excessive control rod drive speeds, AOP-25 Uncoupled Control Rod d) excessive reactivity addition rate, AOP-25 Uncoupled Control Rod Proposed Answer:

a) damage to the drive mechanism seals, AOP-24 Stuck Control Rod 36 of 61 Rev3 ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet QUESTION # 16 Continued Explanation (Optional): Indications are rod did NOT move, both AOPS have symptoms of lack of NI response to rod movement while AOP-24 includes RPlS failure to indicate rod motion. The candidate must also determine Drive D/P is excessive, > 600 psid with RPV pressure < 650 psig & per procedure caution, this condition could damage the drive mechanism seals. Overheating the seals would only apply if the cooling water was isolated but this was NOT done per the stem. The distracters for AOP- 25 are part of a caution in regards to individual Scram to re-couple.

b) over-heating the drive mechanism seals, AOP-24 Stuck Control Rod c) excessive control rod drive speeds, AOP-25 Uncoupled Control Rod d) excessive reactivity addition rate, AOP-25 Uncoupled Control Rod Justification:

Indications are rod did symptoms of lack of NI response to rod movement while AOP-24 includes RPlS failure to indicate rod motion. The candidate must also determine Drive D/P is excessive, > 600 psid with RPV pressure < 650 psig & per procedure caution, this condition could damage the drive mechanism seals. Overheating the seals would only apply if the cooling water was isolated but this was NOT done per the stem. The distracters for AOP-25 are conditions that could occur an uncoupled rod and a stuck control rod. A successful recoupling and rod movement could cause excessive reactivity addition.

Also a drive that does not have the weight of the blade on it would move at a faster rate than normal. Distracters:

move, both AOPs have Technical Reference(s): AOP-24, AOP-25 (Attach if not previously provided) Proposed references to be provided to applicants during examination:

NONE. Learning Objective:

LP AOP, EO- 1.04 ~ (As available) Question Source:

Bank # Modified Bank # (Note changes or attach parent) New X Question History:

Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.) Question Cognitive Level:

Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Assessment of Facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Com m e n ts: 37 of 61 Rev3 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet QUESTION # 17 . . . . . . . . . . . d' Examination Outline Cross-reference: Level SRO Nuclear Boiler Inst. Tier # 2 Ability to (a) predict the impacts of the 2 following on the NUCLEAR BOILER INSTRUMENTATION; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (I OCFR 55.43.5) Group # WA # 216000 A 2.03 0 Instrument line leakage Importance Rating 3.1 Proposed Question: The plant is at 100%. Feedwater i-evel Control is in %element controi and is selected to RPV Water level Column 'B'. A report is received from a Radiation Protection Technician that the Reactor Building 344 ft ARM is in ALARM and steam and water is leaking into Reactor Building 300'. Coincident with the above the Control Room has the following indications: 5-1-28 RX VVTR LVL ALARM HI OR LO 5-2-29 FDVVTR CNTRL A OR B OR C HI RX LVL TRIP - EPIC Pt #92- RFP HI WTR LVL A TRIP - EPIC Pt #93- RFP HI WTR LVL B TRIP - EPIC Pt #94- RFP HI WR LVL C TRIP "ON" -"ON" - "NORMAL" - "TRIPPED" - "NORMAL" WITHOUT any operator actions the plant will respond by crew response, PRIOR to receiving any other alarms or indication changes shall be per Scramming, AOP-41 FEEDWATER MALFUNCTION (RISING FEEDWATER FLOW - HIGH RPV WATER LEVEL)and AOP-39 LOSS OF COOLANT Scramming, AOP-42 FEEDWATER MALFUNCTION (LOWERING b) FEEDWATER FLOW) and EOP-5 SECONDARY CONTAINMENT CONTROL c) Continuing operation at a higher RPV level, AOP-41 FEEDWATER LEVEL) and EOP-5 SECONDARY CONTAINMENT CONTROL Continuing operation at a lower RPV level, AOP-42 FEEDWATER MALFUNCTION (LOWERING FEEDWATER FLOW) and AOP-39 LOSS OF COOLANT b) Scramming, AOP-42 FEEDWATER MALFUNCTION (LOWERING FEEDWATER FLOW) and EOP-5 SECONDARY CONTAINMENT CONTROL , The a) MALFUNCTION (RISING FEEDWATER FLOW - HIGH RPV WATER d) Proposed Answer:

38 of 61 Rev3 ES-40 I Sample Written Examination Form ES-401-5 Question Worksheet

'd Explanation (Optional):

Distracters:

a) QUESTION # 17 Continued Stem symptoms indicate 06LT-52B RPV VVTR LEVEL X-mitter has a reference side instrument line break, DIP went to 0 resulting in a indicated high level as confirmed by annunciators

& EPIC point with a confirmation provided by the RP Tech that line break is outside the CNMT in the RB. Report stipulates ARM alarm which would be an EOP-5 entry

& exit from AOP-9. Without operator action, RPV level would lower to the low SCRAM setpoint as FW backs down due to "8 level x-mitter in control & it would be the dominant control signal over Steam & Feed flow. Crew response prior to receiving further alarms is to enter AOP-42 based on indications

& EOP-5 based on RP Tech report. 39. entry symptoms are present for AOP-41, AOP-1 or AOP- Scramming, AOP-41 FEEDWATER MALFUNCTION (RISING FEEDWATER FLOW - HIGH RPV WATER LEVEL)and AOP-39 LOSS OF COOLANT c) Continuing operation at a hiqher RPV level, AOP-41 FEEDWATER LEVEL) and EOP-5 SECONDARY CONTAINMENT CONTROL d) Continuing operation at a lower RPV level, AOP-42 FEEDWATER MALFUNCTION (LOWERING FEEDWATER FLOW) and AOP-39 LOSS OF COOLANT Justification:

Stem symptoms indicate 06LT-52B RPV WTR LEVEL X- mitter has a reference side instrument line break, D/P went to 0 resulting in a indicated high level as confirmed by annunciators

& EPIC point with a confirmation provided by the RP Tech that line break is outside the CNMT in the RB. Report stipulates ARM alarm which would be an EOP-5 entry

& exit from AOP-9. Without operator action, RPV level would lower to the low SCRAM setpoint as FW backs down due to "B" level x-mitter in control & it would be the dominant control signal over Steam

& Feed flow. Crew response prior to receiving further alarms is to enter AOP-42 based on indications

& EOP-5 based on RP Tech report.

NO entry symptoms are present for AOP-41, AOP-I or AOP-39. MALFUNCTION (RISING FEEDWATER FLOW - HIGH RPV WATER Technical Reference(s):

AOP-41 I AOP-42, AOP-9, (Attach if not previously provided)

AOP-1, EOP-5 Proposed references to be provided to applicants during examination: Learning Objective:

SDLP-06, EO- 1 .I 0.d (As available)

EOP 5 Question Source: Bank

1. Modified Bank # (Note changes or attach parent)

New X Question History:

Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.) Question Cognitive Level: Memory or Fundamental Knowledge 39 of61 Rev3 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet QUESTION ## 17 Continued Comprehension or Analysis X IO CFR Part 55 Content: 55.41 55.43 5 Assessment of Facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Comments:

40 of 61 Rev3 ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet QUESTION # 18 SRO '4 Examination Outline Cross-reference:

Level RHR/LPCI: Torus/Pool Cooling Mode Tier ## 2 Ability to (a) predict the impacts of the Group # K/A # 21 9000 2 A 2.12 following on the RHR/LPCI:

TORUS / SUPPRESSION POOL COOLING MODE; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

(1 OCFR 55.43.5) Valve logic failure: Plant-Specific Importance Rating 3.1 Proposed Question: The plant is operating at 100% power. The SNO is operating RHR for surveillance and reports that IOMOV-66A "RHR HEAT EXCH A BYPASS VLV will not close. The CRS must declare the implement procedure a) Torus cooling, AP-10.01 Work Order Processing b) Torus cooling, AP-20.13 10CFR21 Reporting c) LPCl flow, AP-05.13 Maintenance During LCOs d) LPCl flow, AP-12.08 LCO Tracking and Safety Function Determination mode of RHR inoperable and u Program Proposed Answer: Explanation (Optional):

a) Torus cooling, AP-10.01 Work Order Processing With RHR HEAT EXCH A BYPASS VLV failing to remain shut due to the valve logic, it will RHR loop. Procedures that apply are AP-10.01 (initiate a Work request), AP-12.08 & AP-05.13 are applicable procedures to be used but are NOT correct choices due to RHR will provide the required LPCl flow. AP-20.13 is NOT correct as a simple valve failure does not meet the reporting requirements. meet required heat removal capacity in A Distracters:

b) Torus cooling, AP-20.13 1 OCFR21 Reporting c) LPCl flow, AP-05.13 Maintenance During LCOs d) LPCl flow, AP-12.08 LCO Tracking and Safety Function Determination Program Justification:

AP-12.08 & AP-05.13 are applicable procedures to be used but are NOT correct choices due to RHR will provide the required LPCl flow. AP-20.13 is NOT correct as a simple valve failure does not meet the reporting requirements Technical Reference(s):

op-13, op-1 3Bl TS-B3.5.1 I (Attach if not previously provided) B3.6.2.3, AP-10.01, AP-01.02, AP-05.13, AP-12.08 Proposed references to be provided to applicants during examination:

NONE Learning Objective:

SDLP-10, EO- 1.09.d.2 (As available) 41 of 61 Rev3 ES-40 1 Sample Written Examination Form ES-40T-5 Question Worksheet QUESTION # 18 Continued d Question Source: Bank

1. 1. Modified Bank # (Note changes or attach parent)

New X Question History:

Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.) Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X IO CFR Part 55 Content: 55.41 55.43 5 Assessment of Facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Comments:

42 of 61 Rev3 ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet QUESTION ## I9 Examination Outline Cross-reference: Level Tier # Group # wA # 2.1 Knowledge of less than one hour technical specification action statements for systems. (1 OCFR 55.43.2) Importance Rating Proposed Question:

The Plant is at 100%

Power on 4/16/06. Proposed Answer: SRO 3 1 G 2.1.11 3.8 - At 19:OO on 4/16/06 a loss of Lighthouse Hill-Fitzpatrick Line # 3 occurs. - ST-9W Electrical Lineup and Power Verification was last performed at 17:OO on 4/16/06 per regularly scheduled surveillance frequency. Applicable portions of ST-9R EDG System Quick-Start Operability Test and Offsite Circuit Verification must be performed next by a) 17:OO on 4/23/06 b) 1 I :00 on 4/25/06 c) 0l:OO on 4/17/06 d) 20:00 On 4/16/06 d) 20:OO on 4/16/06 Explanation (Optional):

Stem provided indications that one offsite power source was lost and thus requires entry into LCO 3.8.1 Action A requiring completion of SR 3.8.1.1 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> & once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> there-after to ensure that offsite power is available.

LCO entry time is 19:OO making SR due by 20:OO. The 17:OO on 4/23/06 is the normal 7 day frequency based upon last completing the SR at 17:OO on 4/16/06. The choice of 11:OO on 4/25/06 allows for 1.25 extension of the normal 7 day frequency based upon SR 3.02. The choice of 01:OO on 4/17/06 is an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> time based upon last completing the SR at 17:OO on 4/16/06. SR 3.02 is - NOT allowed for the, perform within 1 hour- see example 1.4-2 in TS. Distracters:

a) 17:OO On 4/23/06 b) 1 I :00 on 4/25/06 c) 01:OO on 4/17/06 Justification: Stem provided indications that one offsite power source was lost and thus requires entry into LCO 3.8.1 Action A requiring completion of SR 3.8.1.1 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> & once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> there-after to ensure that offsite power is available. LCO entry time is 19:OO making SR due by 2O:OO. The 17:OO on 4/23/06 is the normal 7 day frequency based upon last completing the SR at 17:OO on 4/16/06. The choice of 11:OO on 4/25/06 allows for 1.25 extension of the normal 7 day frequency based upon SR 3.02. The choice of 01:OO on 4/17/06 is an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> time based upon last completing the SR at 17:OO on 4/16/06. SR 3.02 is NOT allowed for the, perform within 1 hour- see example 1.4-2 in TS. 43 of 61 Rev3

'L' ES-401 Sample Written Examination Form ES-401-5 Question Worksheet QUESTION # 19 Continued L' Technical Reference(s):

AOP-72, ST-9W1 ST-9R1 TS-SR (Attach if not previously provided) 3.02, TS-3.8.1, SR-3.8.1.1, TS-B3.8.1 Proposed references to be provided to applicants during examination:

Technical Specs- No Bases Learning Objective:

SDLP-710, EO- 1.16 (As available)

Question Source: Bank

1. Modified Bank ## (Note changes or attach parent) New X Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.) Question Cognitive Level:

Memory or Fundamental Knowledge 55.41 55.43 2 Facility operating limitations in the technical Comprehension or Analysis X IO CFR Part 55 Content: specifications and their bases. Comments:

44 of 61 Rev3 ES-40 I Sample Written Examination Form ES-401-5 Question Worksheet QUESTION # 20 ~d . . . . . . . Examination Outline Cross-reference:

Level SRO Tier # 3 Ability to supervise and assume a Group

1. 1 management role during plant transients wA # 2.1 G 2.1.6 and upset conditions.

(10CFR 55.43.5) Importance Rating 4.3 Proposed Question:

The Plant is at 70% power to allow removing 'A' Feedwater Pump from service for maintenance. - All 3 Circ Water Pumps are in service. - Tempering is in progress to maintain Cond Demin inlet temperature - Lake level went from 245.5 ft to 242.5 ft in the last 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> as noted on EPIC Log 1. - The Outside NPO reports that there is traveling screens or intake structure. - Traveling Screens are 'Continuous Run' with indications of debris in the Fish Basket. 95-1 OOOF. ice formation on the 'J In accordance with AOP-64 Loss of Intake Water Level, which of the following actions will be the next required action to perform? Raise tempering flow per OP-4 Circulating Water System Reduce Reactor Power to less than 65% per RAP-7.3.16 Plant Power Stop Circ Wtr Pump C 36P-IC per OP-4 Circulating Water System Manually Scram the Plant per AOP-1 Reactor Scram b) Reduce Reactor Power to less than 65% per RAP-7.3.16 Plant Indications are provided that the loss of intake level is due to other than ice formation. AOP-64 requires power reduced to 65% prior to stopping CW Pump C if lake level has lowered

> 2 ft in the last 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> as stipulated in the stem. The plant is manually scrammed if lake level is < 240 ft which was NOT given in the stem.

Raise tempering flow per OP-4 section G if Ice formation is the cause of the lowering level, this was NOT given in the stem, in fact indications reported locally of ice formation in the intake structure was provided in the stem. a) b, Changes c) d) Proposed Answer: Explanation (Optional):

Power Changes 45 of 61 Rev3 ES-40 I Sample Written Examination Form ES-401-5 Question Worksheet -u' QUESTION # 20 Continued Distracters:

a) Raise tempering flow per OP-4 Circulating Water System c) Stop Circ Wtr Pump C 36P-IC per OP-4 Circulating Water System d) Manually Scram the Plant per AOP-1 Reactor Scram Justification:

Indications are provided that the loss of intake level is due to other than ice formation. AOP-64 requires power reduced to c 65% prior to stopping CW Pump C if lake level has lowered

> 2 ft in the last 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> as stipulated in the stem. The plant is manually scrammed if lake level is < 240 ft which was NOT given in the stem. Raise tempering flow per OP-4 section G if Ice formation is the cause of the lowering level, this was NOT given in the stem, in fact indications reported locally of ice formation in the intake structure was provided in the stem. Technical Reference(s): AOP-64 (Attach if not previously provided)

'4 Proposed references to be provided to applicants during examination: Learning Objective: LPAOP, EO- 1.03.a (As available) Question Source: Bank

1. NONE Modified Bank # (Note changes or attach parent) New X Question History:

Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.) Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41 55.43 5 Assessment of Facility conditions and Comprehension or Analysis X selection of appropriate procedures during normal, abnormal, and emergency situations.

Comments:

46 of 61 Rev3 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet QUESTION # 21 SRO '-1 Examination Outline Cross-reference:

Level Tier # 3 Knowledge of the process for managing Group

1. 2 G 2.2.17 WA# 2.2 maintenance activities during power operations. (IOCFR 55.43.5) Importance Rating 3.5 Proposed Question: The Plant is at 100% power on a Sunday night. The Operating Shift has determined that the Plant will be de-rated to 65% power to support emergent repair work. Per procedure AP-12.13, "345/115 KV Transmission Line Operations and Interface", the NYPA Energy Control Center is required to be notified by the a) Shift Manager b) Reactor Engineer Operations Manager d) Field Support Supervisor Proposed Answer:

a) Shift Manager Explanation (Optional): AP-12.13 provides guidance for NYPA ECC interface for this situation while OF-65 has a step to notify ECC of a shutdown schedule, the plant is NOT being shutdown but is being de-rated.

Per AP-12.13, Operations Manager is responsible for overall implementation of this procedure.

Reactor Engineer (RE) is responsible to coordinate generation scheduling with ENN Power Marketing.

Advanced Scheduling: All generation scheduling will be done between ENN Power Marketing

& RE Dept. RE must notify ENN On-Call Scheduler at least 7 days in advance of any planned power changes. The Shift Manager (SM) is responsible for authorizing access to JAF Switch yard, communicating with transmission operator for resolving emergent issues.

SM is responsible for changes to unit power scheduling with NYISO. Work Control Center Supervisor (WCCS)is responsible for ensuring that 115/345KV work with potential to affect operation of JAF, are scheduled on the weekly work schedule per AP- 10.02 "12 Week Rolling Schedule", & coordinated by the JAF 115/345 KV Coordinator.

JAF 115KW345KV Coordinator is responsible to interface with Power Control & Regional Central Control to review, coordinate 8, schedule line outages & work that has potential for causing an unplanned line outage. WCCS is the 115/345 KV coordinator.

Real-Time Operations:

For unplanned down powers or delayed power restorations, JAF Ops is required to contact NYPA ECC for a "derate", the term "derate" must be used & give plant status. FOR the stem conditions, the only member of Operations present at Sunday night would be the SM as the OM is off. 47 of 61 Rev3 ES-40 Sample Written Examination Form ES-401-5 Question Worksheet QUESTION # 21 Continued Distracters: b) Reactor Engineer c) Operations Manager d) Field Support Supervisor Justification:

Per AP-12.13, Operations Manager is responsible for overall implementation of this procedure. Reactor Engineer (RE) is responsible to coordinate generation scheduling with ENN Power Marketing.

Advanced Scheduling:

All generation scheduling will be done between ENN Power Marketing

& RE Dept. RE must notify ENN On-Call Scheduler at least 7 days in advance of any planned power changes. The Shift Manager (SM) is responsible for authorizing access to JAF Switchyard, communicating with transmission operator for resolving emergent issues.

SM is responsible for changes to unit power scheduling with NYISO. Real-Time Operations:

For unplanned down powers or delayed power restorations, JAF Ops is required to contact NYPA ECC for a "derate", the term "derate" must be used & give plant status. FOR the stem conditions, the only member of Operations present at Sunday night would be the SM as the OM is off, The Field Support Supervisor is part of a normal crew on Sunday but has not responsibility in AP-12.13.

Technical Reference(s):

AP-12-13, EN-OP-11%

RAP- (Attach if not previously provided) 7.3.16 Proposed references to be provided to applicants during examination:

Learning Objective:

LPAP, EO- 20.02 (As available)

Question Source: Bank # Modified Bank # New X NONE (Note changes or attach parent)

Question Historv: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question .) Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 48 of 61 Rev3 ES-40 1 Sample Written Examination Question Worksheet Form ES-401-5 Comments:

QUESTION # 21 Continued 55.43 5 Assessment of Facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

49 of 6 1 Rev3 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet QUESTION # 22 Examination Outline Cross-reference:

Knowledge of the refueling process. (1 OCFR 55.43.6) Proposed Question:

a) b) c) d) Proposed Answer: The Plant is in ar Level Tier # Group # WA# 2.2 Importance Ratii *efuel outaae and fuel '9 mo vement is under SRO 3 2 G 2.2.27 3.5 way. The next'move per SNM move sheet 06-058, step 275, is fuel bundle (YJX 230) which is being moved into core location 43-48 (Clip NE). A Refuel Error per RAP-7.I.04Bl Refueling Procedure, would occur if the bundle has its nose cone partially inserted into 43-48 (Clip SE) then is changed to 43-48 (Clip NE) prior to the start of the next move. is inserted 30 inches into location 03-32 (Clip NE) and is subsequently removed and inserted into location 43-48 (Clip NE). Is moved from core location 39-42 (Clip SE) and is returned to location 39-42 (Clip SE) due to poor visibility in location 43-48. is fully inserted into 43-48 (Clip SE) with the grapple disengaged and then is changed to 43-48 (Clip NE) prior to the start of the next move. subsequently removed and inserted into location 43-48 (Clip NE). b) is inserted 30 inches into location 03-32 (Clip NE) and is Explanation (Optional):

Per RAP-.l .I .04B, 5.9 Refuel Error, the only choice that constitutes.

a refuel error is 'B' as the nose cone is partially inserted into the wrong location. Distractor

'A' & 'D' are wrong orientation that is corrected prior to start of next move. Distractor

'C is NOT a refuel error per step 5.9.3. 5.9.1 A fuel bundle fully or partially placed (i.e., past the nose cone) in 5.9.2 A mis-orientated fuel bundle is not considered a Refuel Error if an incorrect location is a Refuel Error. the miss-orientation is corrected immediately.

It is a Refuel Error if the mis-orientated bundle is identified after the start of the next move. reason and the bundle is returned to the starting location.

5.9.3 It is not a Refuel Error if a move cannot be completed for any 50 of 61 Rev3 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet LJ QUESTION # 22 Continued Distracters:

a) has its nose cone partially inserted into 43-48 (Clip SE) then is changed to 43-48 (Clip NE) prior to the start of the next move. c) Is moved from core location 39-42 (Clip SE) and is returned to location 39-42 (Clip SE) due to poor visibility in location 43-48. d) is fully inserted into 43-48 (Clip SE) with the grapple disengaged and then is changed to 43-48 (Clip NE) prior to the start of the next move. Justification: Per RAP-1.1.048, 5.9 Refuel Error, the only choice that constitutes a refuel error is 'B' as the nose cone is partially inserted into the wrong location. Distractor 'A'

& 'D are wrong orientation that is corrected prior to start of next move. Distractor

'C' is NOT a refuel error per step 5.9.3. 5.9.1 A fuel bundle fully or partially placed (Le., past the nose cone) in an 5.9.2 A mis-orientated fuel bundle is not considered a Refuel Error if the incorrect location is a Refuel Error.

miss-orientation is corrected immediately.

It is a Refuel Error if the mis-orientated bundle is identified after the start of the next move. 5.9.3 It is not a Refuel Error if a move cannot be completed for any reason and the bundle is returned to the starting location.

Technical Reference(s):

RAP-7.1.04B (Attach if not previously provided) Proposed references to be provided to applicants during examination:

NONE Learning Objective:

LPAP, EO- 73.04 (As available) Question Source:

Bank # Modified Bank # (Note changes or attach parent) New X Question History:

Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level:

Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 Comments:

55.43 6 Procedures

& limitations involved in initial core loading, alterations in core configuration, control rod programming

& determination of various internal & external effects on core reactivity.

51 of61 Rev3 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet QUESTION # 23 v' Examination Outline Cross-reference: Level SRO Tier # 3 Knowledge of the requirements for Group # 3 G 2.3.6 reviewing and approving release permits.

wA 2m3 Importance Rating 3.1 The plant is in cold shutdown and all equipment is operable.

A liquid (1 OCFR 55.43.4) Proposed Question:

Proposed Answer: Expl an ati on v Radwaste discharge to the canal is about to occur. An independent review of the Canal Discharge Worksheet (attached) by the FSS shows that the discharge can NOT take place. The reason for this is that: The Chemistry Superintendent's signature is required An Independent Analysis signature is required The radiation monitor, 17RM-350, alarm and isolation setpoints are to be set lower The radiation monitor, 17RM-350, alarm and isolation setpoints are to be set higher d) The radiation monitor, 17RM-350, alarm and isolation setpoints are to be set higher (Optional): , The canal discharge activity level is obtained from the discharge permit. The permit activity is larger than the number recorded on the worksheet.

A larger activity number, if used on the worksheet, would calculate to a higher monitor setpoint. Thus, since a lower activity number was used, a lower alarm and setpoint were used.

Choice A is wrong because the Chemistry Superintendents signature is required if the minimum CW pumps (1) are not operating.

Choice B is wrong because an independent analysis is required if the radiation monitor is inoperable.

Choice C is wrong because the actual activity number, 3.8xE-4 is larger than the number on the worksheet of 2.8xE-5. The question is higher order in that the candidate must analyze the calculations and required signatures to see what mistake was made. With the mistake determined to be an incorrect transcribe activity number he must analyze and determine how this affects the setpoint settings without having the formula to actually calculate the setpoint.

52 of 61 Rev3 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet QUESTION # 23 Continued Distracters:

Technical Reference(s):

op-49 Liquid Radioactive waste Svstem (Attach if not previously provided) Proposed references to be provided to applicants during examination: Provide calculator and filled in discharge permit and worksheet. Learning Objective: SDLP-20, EO 1.13 (As available)

Question Source: Bank

1. Modified Bank # New X (Note changes or attach parent)

Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question .) Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content:

55.41 55.43 4 Radiation hazards that may arise during normal & abnormal situations, including maintenance activities

& various contamination conditions.

Comments:

4' 53 of 61 Rev3 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISCHARGE DATA Dale lime Dilution Water Pumps Level Rate OperaBng Circ Service Start Pumpart End Pumpout v operator Initials Durlnq Discharae:

Rad Wr Reading 17RR-337 cps Flw Rcdr Reading 20FR-441 Discham valve line-up returned to normal in accordance with canal discharae shutdown lineup for aDplicable tank: Tank: Aw. Operator (PrinVSlgn):

Date Time FORWARD DISCHARGE PERMIT TO CHEMISTRY IMMEDIATELY FOLLOWING DISCHARGE ATTACHMENT 54 of 61 Rev3 d ES-401 Sample Written Examination Form ES-401-5 Question Worksheet v DISCHARGE WORKS- Page 1 of 2 DATA 1. Number of running circulating water pumps (36P-:A/B/C)

I 2. Number of running service water pumps (46P-lA/B/C) 8 v-5 3. Tank Discharge Flow Rate (maximum)

TDFR mm 4. Tank Activity (ACT) Je pCi/ml (from discharge permit) 5. Required Dilution Factor (DF) 100 (from discharge permit) 6. Liquid rad monitor (17P.M-350) reading. CP = {EPIC-A-1209)

NOTE 1: Items 7 and 8 are obtained at panel 09-14 NOTE 2: Background should be maintained LESS THAN 1000 cps. It is recommended that the detector canister be flushed to levels below this prior to discharge.

k' - d*l- 7 7. Liquid rad monitor t17RM-350) background t f cps 8. Liquid rad monitor (17RM-350)

K-factor 2 *Gq% IU pCi/ml/cps Pi 9. Tempering gate/*low fJ % (EPIC-A-3547)

CALCULATIONS

10. CFR = I (#l X 120,000)+(#2 x 18,000)1 x \(1 - #9/100)1 = f56, occl 11. Calculate Canal Dilution Factor (CDF): gpm CFR Y10 I NOTE x F,= Fraction of allowed dilution (dimensionless, must be 12. Calculate F,: less than 1.0 for discharge).

F, = CDF X DF = #11 X #5 = acli 13. Calculate Background Correction Activity (BCA) in uCi/rnl: COMPLETED FORMS ARE ATTACHED TO THE DISCHARGE PERMIT LIQUID RADIOACTIVE 55 of 61 Rev3 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet .J' CANAL DISC HARGE WORKS HEET Page 2 of 2 14. Calculate Hi/Hi setpoint in pCi/rnl: Hi/Hi = (ACT) = #4 + t13

• 2 X F, 2 X #12 15. Calculate Hi setpoint in w3./nd.: Hi = (ACT) = #4 + t13 = vCi /ml 4 X F, 4 X t12 16. Obtain 17RM-350 potentiometer setting for HI-Hi setpoint from Chemi strv 17. Obtain l7RM-350 potentiometer setting for Hi setpoint from Chemistry.

Hi 18. Enter potentiometer settings for Hi and Hi-Hi setpoints on Discharge Permit Section B and attach this worksheet to the discharge permit. Independent Verification Print/Sign/Date COMPLETED FORMS ARE ATTACBXD TO TXE DISCHARGE PERMIT 56 of 61 Rev3

. . - . - -c r ES-40 I Sample Written Examination Form ES-401-5 Question Worksheet QUESTION # 24 SRO ..-,. Examination Outline Cross-reference:

Level Tier # 3 Knowledge of how event-based Group # 4 emergency/abnormal operating procedures wA # 2.4 G 2.4.8 are used in conjunction with the symptom based EOPs. (7 OCFR 55.43.5) Importance Rating 3.7 Proposed Question:

The plant is at 100% power with two Service Water pumps running and the third one in "Standby" when a large unisolable Service Water rupture in the reactor building occurs. The Crew enters AOP-10 "Loss of Service Water Cooling" and EOP-5 "Secondary Containment Control". These procedures direct the following actions:

AOP Ensure standby service water pump(s) start, manually EOP isolate all systems that are discharging into the area, scram the reactor shutdown the reactor The CRS must direct that the reactor be service water pumps must be and that all a) shutdown normally, started b) shutdown normally, tripped C) scrammed, started d) scrammed, tripped Proposed Answer:

d) scrammed, tripped Explanation (Optional): EP-1 states that other procedures may be used with EOPs but shall not contradict nor subvert actions specified in the EOPs. If SW was allowed to continue its operation, it would subvert the intent of "isolating all systems that are discharging into the area". Per the EOP bases, the requirement to shutdown does not preclude a scram. With a loss of all SW, a scram is required.

57 of 61 Rev 4

.I .-.' ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Distracters:

a) shutdown normally, started b) shutdown normally, tripped c) scrammed, started Justification: Choices A and C allow SW to continue running.

In EOP-5 a reactor shutdown is required when both crescent area water levels are 18 or greater. Per the EOP bases, "a direct threat exists relative to secondary containment integrity, to equipment located in the reactor building and to continued safe operation of the plant." This, along with the EOP requirement to isolate the leak, requires that SW be tripped. If SW is completely lost, then ESW automatically aligns to supply some ventilation cooling loads and can be aligned to supply cooling to essential RX Bldg loads. However, it does not cool loads that are required for power production. If choice B is selected, and the SW pumps are tripped, a normal shutdown would be impossible. Additionally, AOP-10 requires a reactor scram for a complete loss of SW. Technical Reference(s):

AOP-10, EOP-5, EP-1 I MIT- 301 .I 1 F (Attach if not previously provided)

Proposed references to be provided to applicants during examination:

NONE Learning Objective:

SDLP-46A, EO- 1.14.a Question Source:

Bank # (As available)

Modified Bank # X (Note changes or attach parent) New Question History: Last NRC Exam X Modified from 2005 Vermont Yankees SRO exam (question

1. 99, attached) (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Assessment of Facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations..

Comments:

58 of 61 Rev 4 ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet QUESTION # 25 .. .. SRO 'd Examination Outline Cross-reference:

Level Tier # 3 Ability to perform without reference to Group # 4 procedure those actions that require WA# 2.4 G 2.4.49 immediate operation of system components and controls.

(10CFR 55.43.2) Importance Rating 4.0 Proposed Question: The Plant is at 100% power.

The following event and indications occur subsequently: - Feedwater Pump A trips off - RWR Pump A trips off - RWR Pump B - speed 30% - Annunciator 09-5-2-44 APRM UPSCALE - ON - All APRM recorders - Cycling 65% to 77% every 2 seconds - All SRM Period meters - Cycling minus 80 to plus 30 seconds every 1 - Various LPRM Upscale Alarms - Alarming and clearing every 2 % seconds seconds Which one of the following is immediately required?

v a) Trip RWR Pump B b) Insert CRAM Groups c) Manually Scram the Reactor d) Raise RWR Pump B speed and flow c) Manually Scram the Reactor.

AOP-8 LOSS or Reduction of Reactor Coolant Flow requires immediate action to manually SCRAM if indications of thermal hydraulic instabilities (THI) are observed.

Refer to Attachment 1 of AOP-8 for Indications of THI. Distracters B and D are possible AOP-8 actions but not in this situation. Tripping of the B RWR pump would make THI worse (high power with low flow). a) Trip RWR Pump B. b) Insert CRAM Groups d) Raise RWR Pump B speed and flow Justification:

AOP-8 Loss or Reduction of Reactor Coolant Flow requires immediate action to manually SCRAM if indications of thermal hydraulic instabilities (THI) are observed.

Refer to Attachment I of AOP-8 for Indications of THI. Distracters B and D are possible AOP-8 actions but not in this situation. Tripping of the B RWR pump would make THI worse (high power with low flow). Proposed Answer: Explanation (Optional):

Distracters:

Technical Reference(s):

AOP-8, TS- Bases 3.4.1 (Attach if not previously provided)

CR-JAF-2000-06312 (SER 7-00, Reference Only BWR Core Power Oscillations) 59 of 61 Rev3 ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet Proposed references to be provided to applicants during examination:

NONE -.-/ QUESTION # 25 Continued Learning Objective:

LPAOP, EO- 1.03.a (As available)

QUESTION # 25 Continued Question Source: Bank

1. Modified Bank # New X (Note changes or attach parent) Question History:

Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X IO CFR Part 55 Content: 55.41 55.43 5 Assessment of Facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Comments:

60 of 61 Rev3 ES-40 I Sample Written Examination Form ES-401-5 Question Worksheet QUESTION # 24 Attachment Vermont Yankees 2005 SRO Question # 99 d Select the correct answer:

While operating at power, a service water rupture in the reactor building has occurred and it can not be isolated. During implementation of procedures, the following directions conflict:

OP 2181 - secure all SW pumps ON 3 148 - manually scram the reactor, reduce SW pumps operating to two EOP complete Reactor Shutdown per OP 0 105 ARS (6-A-5) SERV WTR HDR PRESS LO - start all SW pumps, perform Reactor Shutdown What action must be implemented first? Why?

a) b) c) d) Implement OP 2181; preventing pump damage is critical Implement ON 3148; reactor scram is required to reduce heat loads Implement EOP-4; EOP actions override low tier procedures Implement ARS (6-A-5); controlled restoration of SW and plant shutdown is required 'J Answer: B 61 of 61 Rev3