Text

2012-10 Draft Written Exam
	ML12297A439
Person / Time
Site:	Cooper
Issue date:	10/30/2012
From:	Vincent Gaddy; Operations Branch IV
To:	; Nebraska Public Power District (NPPD)
	laura hurley
References
	Download: ML12297A439 (445)
	v • d • e

ES-401 Site-Specific RO Written Examination Form ES-401-7 Cover Sheet U.S. Nuclear Regulatory Commission Site-Specific RO Written Examination Applicant Information Name:

Date: 11-05-2012 Facility/Unit: Cooper Nuclear Station Region: I II III IV Reactor Type: W CE BW GE Start Time: Finish Time:

Instructions Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. To pass the examination, you must achieve a final grade of at least 80.00 percent. Examination papers will be collected 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after the examination begins.

Applicant Certification All work done on this examination is my own. I have neither given nor received aid.

Applicants Signature Results Examination Value ____75____ Points Applicants Score __________ Points Applicants Grade __________ Percent

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 203000 K3.01 Importance Rating 4.3 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): USAR Table IV-8-1 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR Part 55 Content: 55.41 8 55.43 Comments:

Question more appropriately addresses 55.41.8 vice 55.41.7.

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 1 New 00 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 2 1 4 Multiple Choice Topic Area Description Systems COR0022302001030O Residual Heat Removal System Related Lessons COR0022302 RESIDUAL HEAT REMOVAL Related Objectives COR0022302001030O Describe RHR System design feature(s) and/or interlocks which provide for the following: Reduction in vessel injection flow during accident conditions COR0022302001060B Given an RHR control manipulation, predict and explain changes in the following: RHR pump/system flow and pressure Related References NONE Related Skills (K/A) 203000 K3.01 Knowledge of the effect that a loss or malfunction of the RHR/LPCI:

INJECTION MODE (PLANT SPECIFIC) will have on following: (CFR: 41.7 /

45.4) Reactor Water Level (4.3/4.4)

QUESTION: 1 A LOCA is in progress with B Loop of RHR injecting into the RPV. RPV level is steady at 30 inches on the narrow range instruments.

RPV pressure subsequently rises from 200 psig to 320 psig.

What effect does this have on RPV level and why?

Reactor water level

a. lowers due to system pressure rising above pump shutoff head.

b. lowers due to automatic injection valve closure.

c. rises at a steady rate due to the rise in system pressure.

d. rises at a rapid rate due to the RHR pump design rating.

ANSWER: 1

a. lowers due to system pressure rising above pump shutoff head.

Explanation:

NOTE: See attached partial of USAR Table IV-8-1.

Per USAR Table IV-8-1, the shutoff head of the RHR pumps is 294 psi. With RPV pressure rising, the pressure exceeds pump shutoff head pressure and injection stops. The injection check valve closes as RPV pressure exceeds pump shutoff head. RPV level lowers due to the LOCA.

Distractors:

b. The injection valve open on a LOCA signal but they do not reclose due to system pressure.

c. The injection rate will go down as RPV pressure rises, but when shutoff head is reached RPV level will lower.

d. The injection rate will go down as RPV pressure rises, but when shutoff head is reached RPV level will lower.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 205000 K4.02 Importance Rating 3.7 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): SP 6.RHR.306 (Rev 14) Reactor High Pressure (Attach if not previously provided) Channel Calibration (including version/revision number) USAR Chapter IV Section 8.0 SOP 2.2.69 (Rev 90) RHR System Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank # 24818 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 2 24818 00 07/18/2009 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 2 1 3 Multiple Choice Topic Area Description Systems Ability to predict RHR SDC pump trips and automatic system line up changes.

Related Lessons COR0022302 RESIDUAL HEAT REMOVAL Related Objectives COR0022302001030J Interpret RHR System design feature(s) and/or interlocks which provide for the following: High pressure isolation Related References NONE Related Skills (K/A) 205000 K4.02 Knowledge of SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) design feature(s) and/or interlocks which provide for the following: (CFR: 41.7) High pressure isolation: Plant-Specific (3.7)

QUESTION: 2 24818 RHR Loop B is operating in the Shutdown Cooling (SDC) Mode, when RPV pressure rises to 85 psig.

Which one of the following describes the expected system response?

(Assume all systems being operated from the control room.)

RHR Pump B trips and

a. RHR pump suction valve RHR-MO-15B auto closes, and Inboard Injection valve RHR-MO-25B auto closes.

b. Shutdown Cooling suction valves RHR-MO-17 and RHR-MO-18 auto close, and Inboard Injection valve RHR-MO-25B auto closes.

c. Shutdown Cooling suction valves RHR-MO-17 and RHR-MO-18 auto close, RHR pump suction valve RHR-MO-15B auto closes.

d. Min flow control valve RHR-MO-16B auto opens, and Inboard Injection valve RHR-MO-25B auto closes.

ANSWER: 2 24818

b. Shutdown Cooling suction valves RHR-MO-17 and RHR-MO-18 auto close, and Inboard Injection valve RHR-MO-25B auto closes.

Explanation:

NOTE: See attached Procedure 2.2.69 partial.

Surveillance Procedure 6.RHR.306 (Rev 14), System Operating Procedure 2.2.69, Residual Heat Removal System (Rev 90) Information Section, and USAR Chapter IV Section 8.0 When in SDC a high RPV pressure <72# PCIS Group 2 isolation occurs. This isolation causes RHR-MO-17, RHR-MO-18, and RHR-MO-25 to shut. RHR pumps are interlocked with RHR-MO-17/18 to trip if either of these valves is not full open when that RHR pump is in SDC (associated RHR-MO-15.also open).

Three isolation valves are used for each RHR pump, to isolate the pump suction from the recirculation system. Two of these isolation valves (RHR-MO-17, and 18) are common to all four pumps, with the third (RHR-MO-15A, B, C, or D) being associated with its companion pump.

Distractors:

a. RHR-MO-15B does not receive a close signal on high RPV pressure. RHR-MO-15B does input into the RHR pump trip when in the SDC mode. A loss of suction path in the RHR system will cause a pump trip so it is reasonable to choose this answer based on valve/pump interlocks.

c. RHR-MO-15B does not receive a close signal on high RPV pressure. RHR-MO-15B does input into the RHR pump trip when in the SDC mode. A loss of suction path in the RHR system will cause a pump trip so it is reasonable to choose this answer based on valve/pump interlocks.

d. RHR-MO-16B is de-energized in the closed position when operating in the SDC mode.

This is a procedural requirement and the minimum valve should normal position is open.

Procedure 2.2.69 partial ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 205000 K6.08 Importance Rating 3.5 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): COR002-27-02 (Rev 34)

(Attach if not previously provided) SOP 2.2.70 (Rev 67)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 3 New 00 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 3 1 0 Multiple Choice Topic Area Description Systems COR0022302 Residual Heat Removal System Related Lessons COR0022302 RESIDUAL HEAT REMOVAL Related Objectives COR0022302001040G Describe the interrelationship between the RHR system and the following: RHR Service Water COR0022302001080R Predict the consequences a malfunction of the following will have on the RHR system: RHR Service Water Related References NONE Related Skills (K/A) 205000 K6.08 Knowledge of the effect that a loss or malfunction of the following will have on the SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE): (CFR: 41.7 / 45.7) RHR Service Water: Plant-Specific (3.5)

QUESTION: 3 The following conditions exist:

The plant is in MODE 5.

RHRSW Booster Pump A is operating in support of RHR SDC mode.

RPV coolant temperature is 140°F and rising 1°F every 15 minutes.

RHRSW Loop A flow on SW-FI-132A, SW FLOW, is indicating 2500 gpm.

The CRS directs the BOP operator to lower the RPV coolant temperature to 110°F.

The BOP operator places control switch HX-A SW DISCH VLV 89A to the OPEN position.

Simultaneously MCC-K de-energizes.

What is required to lower the RPV coolant temperature to 110°F?

a. Start RHRSW Booster Pump B.

b. Start RHRSW Booster Pump C.

c. Coordinate with an NLO to manually open SW-MO-89A locally.

d. Continue holding the control switch until SW-FI-132A is indicating the desired flow.

ANSWER: 3

c. Coordinate with an NLO to manually open SW-MO-89A locally.

Explanation:

The RHR Heat Exchanger Service Water outlet valve (SW-MO-89A) is powered from MCC-Q and MCC-Q is powered from MCC-K. MCC-K loss causes an alarm on Control Room vertical board C-2 but MCC-Q de-energized causes no alarms. The operator must realize the connection hierarchy of power supplies between these MCCs. The valve is de-energized so operation from the control room is not available. Local operation of the valve must take place in order to raise cooling to the heat exchanger.

COR002-27-02 (Rev34) lists the power supplies to the RHRSW components and pump arrangements. Station Operating Procedure 2.2.70 (Rev 67) Precautions and Limitations section states Simultaneous operation of both SWBPs in a subsystem for periods of > 1 minute is prohibited at all times except when required by EOPs which prohibits two pump operation.

Distractors:

a. RHR pumps A and B are powered from the same 4160V bus. It would be reasonable to assume starting a similar arrangement with the RHRSW Booster pumps. However, Booster pump B is in SDC loop B so starting this pump would not provide more cooling.

b. Starting another Booster pump in loop A would provide more cooling but it is procedurally forbidden to operate both pumps in one loop unless injecting into the RPV for emergency injection as directed by the EOPs.

d. Continuing to hold the control switch would not provide more cooling as the valve is de-energized. If the electrical alignment is not recognized, then holding the control switch would seem like the correct action to take.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 206000 K1.12 Importance Rating 3.4 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): SOP 2.2.33 (Rev 69)

(Attach if not previously provided) USAR Chapter VI Section 4 (including version/revision number) GE Drawing 791E271 Sheet 3 (Rev N23)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 4 New 00 4/02/2012 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 3 1 5 Multiple Choice Topic Area Description Systems COR0021102 HPCI System interrelations with NBI Related Lessons COR0021102 High Pressure Coolant Injection Related Objectives COR0021102001050K Describe the interrelationship between HPCI and the following:

Nuclear Boiler Instrumentation COR0021102001100E Predict the consequences of the following on the HPCI system:

Nuclear Boiler Instrumentation failures Related References NONE Related Skills (K/A) 206000 K1.12 Knowledge of the physical connections and/or cause-effect relationships between HIGH PRESSURE COOLANT INJECTION SYSTEM and the following: (CFR: 41.2 to 41.9 / 45.7 to 45.8) Nuclear boiler instrumentation:

BWR-2,3,4 (3.4/3.4)

QUESTION 4 The plant is operating at power and Barton Narrow Range instrument NBI-LIS-101B fails downscale. Subsequently a reactor scram occurs due to a PCIS Group 1 and HPCI starts and initiates in response to RPV level lowering. No operator actions are taken with the HPCI controls.

How does HPCI respond to these conditions?

a. HPCI continues to inject as RPV level goes above 59 inches.

b. HPCI speed lowers to minimum regardless of RPV level.

c. HPCI turbine trips and HPCI-MO-14, STM TO TURB VLV closes.

d. HPCI-MO-16, STM SUPP OUTBD ISOL VLV closes and HPCI speed ramps down to zero.

ANSWER: 4

a. HPCI continues to inject as RPV level goes above 59 inches.

Explanation:

NOTE: See attached GE Drawing partials for relay operation causing HPCI trip.

Station Operating Procedure 2.2.33 (Rev 69) provides information on HPCI trip signals. USAR Chapter VI Section 4 provides information on HPCI trip signals. GE Drawing 791E271 Sheet 3 (Rev N23) and Sheet 4 (Rev N24) provides information on LIS-101B and D required to cause trip signal on high RPV level.

Narrow Range Barton instruments (LIS-101B and D) input to the HPCI high RPV level trip circuitry. HPCI receives a trip signal when RPV level is 54.0 inches. The logic is a two out of two logic and requires both channels to trip. With LIS101B failing low, HPCI cannot automatically trip from a high RPV level signal so HPCI continues to inject. The control room operator has to manually trip HPCI to stop injection.

Distractors:

b. If HPCI would have tripped on high RPV level, it would automatically start as the high level signal auto resets on a low level initiation signal.

c. It requires both LIS-101 B and D to trip. HPCI-MO-14 requires both LIS-101 B and D contacts closing to close this valve as well.

d. The valves do not close on high RPV level but does close on a PCIS containment isolation signal.

LIS-101B LIS-101D See next page for 23A-K11 contact closure and HPCI trip

23A-K11 relay energizes and closes contact 1-2 which causes 23A-K12 to energize and trip HPCI

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 209001 A1.05 Importance Rating 3.5 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): B&R P&ID drawing 2022 Sheet 1 (Rev N78)

(Attach if not previously provided) B&R P&ID drawing 2045 Sheet 1 (Rev N58)

(including version/revision number) GE drawing 117C3303 Sheet 6 (Rev N08)

System Operating Procedure 2.2.9 (Rev 74)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank # 5630 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR Part 55 Content: 55.41 5 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 5 5630 00 04/02/2012 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 2 1 4 Multiple Choice Topic Area Description Systems COR0020602, Core Spray flow effect on torus level indication Related Lessons COR0020602 OPS CORE SPRAY SYSTEM Related Objectives COR0020602001080E Given a Core Spray component manipulation, predict and explain the changes in the following: Suppression Pool water level Related References NONE Related Skills (K/A) 209001 A1.05 Ability to predict and/or monitor changes in parameters associated with operating the LOW PRESSURE CORE SPRAY SYSTEM controls including:

(CFR: 41.5 / 45.5) Torus/suppression pool water level (3.5/3.6)

QUESTION: 5 5630 Core Spray pump A is delivering 2400 gpm as indicated on CS-FI-50A, PUMP FLOW, while operating in the test lineup for surveillance testing. The control room operator raises system flow to 4000 gpm.

How does Core Spray pump operation affect Wide Range Torus level indication (PC-LRPR-1A),

CNTMNT/TORUS PRESS & LVL RCDR, and why?

Wide Range Torus water level indicates . . .

a. higher than actual water level due to higher pressure in the discharge piping of the operating pump.

b. higher than actual water level due to reduced pressure in the suction piping of the operating pump.

c. lower than actual water level due to higher pressure in the discharge piping of the operating pump.

d. lower than actual water level due to reduced pressure in the suction piping of the operating pump.

ANSWER: 5 5630

d. lower than actual water level due to reduced pressure in the suction piping of the operating pump.

Explanation:

NOTE: See attached drawings 2022, 2054 and 117C3303 B&R P&ID drawing 2022 Sheet 1 (Rev N78) shows the low pressure connection with PC-DPT-3A1. B&R P&ID drawing 2045 Sheet 1 (Rev N58) shows the high pressure connection with PC-DPT-3A1.

GE drawing 117C3303 Sheet 6 (Rev N08) shows PC-DPT-3A1 where the high and low connections can be determined. System Operating Procedure 2.2.9 (Rev 74) contains the limitation warning the operator of CS pump operation effects on PC-LRPR-1A.

The main control room has torus water level indicator PC-LRPR-1A which receives its input from PC-DPT-3A1. This dP instrument connects to CS suction piping for its high side tap and its low side tap is connected to the torus air space vent piping. Operating CS at a low 4800 gpm causes indicated torus level on PC-LRPR-1A to indicate lower because of the lower pressure in the CS suction piping due to pump operation causes a lower dP sensed on PC-DPT-3A1. The higher the system flow, the lower PC-LRPR-1A will indicate up to a flow of 4800 gpm.

Distractors:

a. The high pressure side of DPT-3A1 becomes lower when Core Spray suction piping lowers on rising flow and a lower dP lowers level indication on PC-LRPR-1A. If the high pressure tap were on the discharge piping, then indicated torus level would be higher.

b. The high pressure side of DPT-3A1 becomes lower when Core Spray suction piping lowers on rising flow and a lower dP lowers level indication on PC-LRPR-1A. If the low pressure tap of DPT-3A1 were on the suction piping then, indicated torus level would be higher.

c. The high pressure side of DPT-3A1 becomes lower when Core Spray suction piping lowers on rising flow and a lower dP lowers level indication on PC-LRPR-1A. If the low pressure tap were on the discharge piping, then indicated torus level would be lower.

B&R P&ID 2022 B&R P&ID 2045

GE drawing 117C3303 Sheet 6 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 211000 K3.02 Importance Rating 3.0 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): COR002-29-02 (Rev 18)

(Attach if not previously provided) COR002-06-02 (Rev 21)

(including version/revision number) USAR Section VII-4.5.4 Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 6 New 00 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 4 2 1 4 Multiple Choice Topic Area Description Systems SLC system/Core Spray leak detection relationship Related Lessons COR0022902 Standby Liquid Control Related Objectives COR0022902001120A Briefly describe the relationships that exist between the SLC system and the following: Core Spray line leak detection Related References NONE Related Skills (K/A) 211000 K3.02 Knowledge of the effect that a loss or malfunction of the STANDBY LIQUID CONTROL SYSTEM will have on following: (CFR: 41.7 / 45.4) Core spray line break detection system: Plant-Specific (3.0*/3.2*)

QUESTION: 6 The plant is operating at rated conditions and the Standby Liquid Control above core plate inner pipe has separated from the bottom core plate. The pipe is now measuring below core plate pressure.

What effect does this condition have on Core Spray line break detection system?

a. CS-DPIS-43A & B indicate approximately -10 psid for BOTH systems.

b. CS-DPIS-43A & B indicate approximately zero for BOTH systems.

c. Core Spray break detection alarms annunciate for BOTH systems.

d. Core Spray break detection alarm annunciates in ONLY ONE Core Spray system.

ANSWER: 6

c. Core Spray break detection alarms annunciate for BOTH systems.

Explanation:

COR002-29-02 (Rev 18) describes the above core plate (inner tube) tap of the SLC system being an input to the Core Spray leak detection system. COR002-06-02 (Rev 21) describes the break detection normal reading being negative at operating conditions. USAR Section VII-4.5.4 describes the CS leak detection instrumentation. DPIS43A and B are calibrated by SP 14.1CS.305 and 14.2CS.305.

Operator rounds indicate full power DPIS-43A & B indicating -3.1 psig and core plate dP indicating 17.1 psig. With the inner pipe (high pressure tap of CS-DPIS-43A and B) now exposed to below core plate dp, the line will have 17.1 psig more pressure in that leg. The low pressure side of CS-DPIS-43A and B remains unchanged so they should be pegged high and above their alarm setting of 0.5 psid.

Distractors:

a. CS-DPIS-43A & B downscale is -10 psid. The indicators will both be positive and above the 0.5 psid alarm setting.

b. The indicators will both be positive near full scale of 15 psid. An equalizing valve leak would cause dp to approach zero.

d. Both systems are effected equally and will alarm. It is unusual to have one tap service both divisions.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 211000 A1.09 Importance Rating 4.0 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): SOP 2.2.74 (Rev 46)

(Attach if not previously provided) B&R P&ID drawing 2045 (Rev N21) Sheet 2 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank # 2194 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR Part 55 Content: 55.41 5 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 7 2194 01 08/07/2011 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 2 1 3 Multiple Choice Topic Area Description Systems Is SLC being injected into the RPV?

Related Lessons COR0022902 OPS STANDBY LIQUID CONTROL Related Objectives COR0022902001050G Describe the SLC design features and/or interlocks that provide for the following: System initiation upon operation of SLC control switch COR0022902001060D Briefly describe the following concepts as they apply to SLC system:

Squib valve operation COR0022902001100D Predict the consequences a malfunction of the following would have on the SLC system: Pump trip COR0022902001100E Predict the consequences a malfunction of the following would have on the SLC system: Squib valves Related References NONE Related Skills (K/A) 211000 A1.09 Ability to predict and/or monitor changes in parameters associated with operating the STANDBY LIQUID CONTROL SYSTEM controls including:

(CFR: 41.5 / 45.5) SBLC system lineup (4.0)

QUESTION: 7 2194 During a failure to scram condition, the CRS orders both SLC Pumps started. Both SLC Pump control switches are taken to START and the following conditions result:

SLC Pump A trips after running 10 seconds.

Squib Valve 14B fails to fire.

What is the condition of the SLC system?

a. SLC is NOT being injected into the RPV into the RPV due to Squib Valve 14B failing to fire.

b. SLC can be injected into the RPV by SLC Pump B only by manual valve manipulations.

c. SLC is NOT being injected into the RPV because Squib Valves 14A and 14B are in series.

d. SLC is being injected into the RPV because SLC Pump B is discharging to the RPV via Squib Valve 14A.

ANSWER: 7 2194

d. SLC is being injected into the RPV because SLC Pump B is discharging to the RPV via Squib Valve 14A.

Explanation:

Station Operating Procedure 2.2.74 (Rev 46), Information Section describes the system response to starting a SLC pump including firing the Squib Valve. Burns & Roe P&ID drawing 2045 (Rev N21) Sheet 2 depicts the pump discharge flowpath showing both Squib Valves in parallel downstream of the common pump discharge piping.

Placing the SLC Pump control switches to start closes the circuit to fire the its respective Squib Valve regardless of pump operation. Both Squib Valves are connected in parallel on the pump discharge common header so either Squib Valve firing will allow either or both pumps to inject into the RPV.

Distractors:

a. The firing circuit for the 14B SQUIB valve is associated with the pump control switch position. The pump will inject because of the common discharge header upstream of the Squib Valves.

b. There are manual valves in the pump discharge piping but only to allow flow to the test tank. The pump does inject through the 14A Squib Valve.

c. The Squib Valves are in parallel as described above.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 212000 A2.12 Importance Rating 4.0 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): GE Drawing 791E256 Sheet 9 (Rev N21)

(Attach if not previously provided) GE Drawing 791E256 Sheet 11 (Rev N13)

(including version/revision number) Procedure 2.1.5 (Rev 67)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR Part 55 Content: 55.41 5 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 8 New 00 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 2 1 4 Multiple Choice Topic Area Description Systems COR0022102001100C Reactor Protection System Related Lessons COR0022102001100C Reactor Protection System Related Objectives COR0022102001040M Describe the RPS design features and/or interlocks that provide for the following: Related system inputs to RPS COR0022102001100F Describe the interrelationship between the RPS and the following:

Main Turbine Generator Related References NONE Related Skills (K/A) 212000 A2.12 Ability to (a) predict the impacts of the following on the REACTOR PROTECTION SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5 / 45.6) Main turbine stop control valve closure (4.0)

QUESTION: 8 The plant is operating at 70% power making preparations for surveillance testing. BOTH Main Turbine Stop Valve 1 limit switches fail open causing the Stop Valve to indicate closed but actual Stop Valve position remains unchanged.

The limit switches are now repaired and the contacts are closed.

How does the Reactor Protection System (RPS) respond to both the limit switches opening?

How does the operator restore the Reactor Protection System according to Procedure 2.1.5, Reactor Scram?

The RPS system

a. Channel A de-energizes.

The operator places the REACTOR SCRAM RESET switch to Group 1 and 4, then back to NORM.

b. Channel A de-energizes.

The operator places the REACTOR SCRAM RESET switch to Group 1 and 4, Group 2 and 3, then back to NORM.

c. Channel B de-energizes.

The operator places the REACTOR SCRAM RESET switch to Group 2 and 3, then back to NORM.

d. Channel B de-energizes.

The operator places the REACTOR SCRAM RESET switch to Group 1 and 4, Group 2 and 3, then back to NORM.

ANSWER: 8

b. Channel A de-energizes.

The operator places the REACTOR SCRAM RESET switch to Group 1 and 4, Group 2 and 3, then back to NORM.

Explanation:

NOTE: See attached for further information.

General Electric drawing 795E256, Sheet 9 (Rev N21) depicts the Stop Valve limit switches (SVOS 1(1) and SVOS 1(2) de-energizing relays 5A-K10A, 5A-K10C, 5A-K10E, and 5A-K10G.

These relays input to RPS Channels A1 and A2. Procedure 2.1.5 (Rev 67) Step 4.1 provides guidance for re-setting a half scram.

The Main Turbine has two Stop Valves (SVOS1 and SVOS2) and each Stop Valve has two limit switches. Each limit switch has two contacts. Limit switch 1 contact 1 (SVOS 1(1)) inputs to relays 5A-K10A and E which operate in parallel in RPS Channel A1. It takes both relays actuating to de-energize channel A1. Limit switch 2 contact 2 (SVOS 1(2))inputs to relays 5A-K10C and G which operate in parallel in RPS Channel A2. It takes both relays actuating to de-energize channel A2. Either Channel A1 OR Channel A2 de-energizing causes a half scram.

General Electric drawing 795E256, Sheet 9 depicts the Stop Valve limit switches (SVOS 1(1) and SVOS 1(2) de-energizing relays 5A-K10A, 5A-K10C, 5A-K10E, and 5A-K10G. These relays input to RPS Channels A1 and A2 which cause RPS Channel A to de-energize and cause a half scram. Since actual valve position did not change the reactor would not have scrammed. Procedure 2.1.5 provides guidance for re-setting a half scram by placing the reset switch in the 1-4 then the 2-3 position and releasing to NORM.

Distractors:

a. RPS Channel A de-energizes but Procedure 2.1.5 Procedure 2.1.5 requires placing the REACTOR SCRAM RESET switch to Group 1 and 4, Group 2 and 3, then back to NORM. Only taking the switch to Group 1 and 4 would reset RPS but Procedure 2.1.5 does not allow it.

c. RPS Channel B does not trip and placing REACTOR SCRAM RESET switch to Group 2 and 3 would not reset RPS.

d. RPS Channel B does not trip but movement of REACTOR SCRAM RESET switch would reset RPS Channel A.

Logic string showing contact 1 (SVOS 1 (1)) serving 5A-K10A and E. Contacts are shown de-energized and open but are normally energized and closed when NOT in tripped condition.

RPS A1 logic string showing 5A-K10A and E in parallel path. Contacts shown open but are closed when NOT in a trip condition.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 215003 K6.02 Importance Rating 3.6 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): Instrument Operating Procedure 4.1.2A (Rev 0)

(Attach if not previously provided) USAR Chapter VII Section 5.7.3.2 and Figure VII-5-24 (including version/revision number) Student Text COR002-12-02 (Rev13)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank # 24791 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 9 24791 00 07/15/2009 06/15/2005 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 1 1 3 Multiple Choice Topic Area Description Systems Loss of 24/48 V DC division 1 to IRMs Related Lessons COR0021202 INTERMEDIATE RANGE MONITOR Related Objectives COR0021202001070B Predict the consequences of a loss or malfunction of the following would have on the IRM system: 24/48 VDC Related References NONE Related Skills (K/A) 215003 K6.02 Knowledge of the effect that a loss or malfunction of the following will have on the INTERMEDIATE RANGE MONITOR (IRM) SYSTEM : (CFR: 41.7 /

45.7) 24/48 volt D.C. power: Plant-Specific (3.6)

QUESTION: 9 24791 A reactor startup is in progress, when a fire completely de-energizes 24/48 VDC Div I Bus.

Which IRMs will still be available for monitoring Reactor power?

a. A, C, E, G

b. B, D, F, H

c. A, B, C, D

d. E, F, G, H ANSWER: 9 24791

b. B, D, F, H Explanation:

NOTE: See attached USAR figure VII-5-24 on IRM power supply arrangements.

The detector and all the electronics in the drawer are powered from the 24/48V batteries.

Channels A,C,E,G are powered from 24/48V DIV 1. Channels B,D,F,H are powered from 24/48V Div 2.

Distractors:

a. These IRMs are powered from Div 1 24/48V system and are not available for use.

c. B and D are available but A and C are de-energized. It is reasonable to believe these IRMs share the same power supply.

d. F and H are available but E and G are de-energized. It is reasonable to believe these IRMs share the same power supply.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 215003 A3.02 Importance Rating 3.3 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): Student Text COR002-12-02 (Rev13)

(Attach if not previously provided) USAR VII-5.7.3.5 Reference APED-5706 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank # 21774 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR Part 55 Content: 55.41 2 55.43 Comments:

Question more appropriately addresses 55.41.2 vice 55.41.7.

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 10 21774 01 08/03/2011 06/15/2005 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 2 1 4 Multiple Choice Topic Area Description Systems COR0021202, IRM, one IRM drifts up and then back down, what is the status of the indicating lights?

Related Lessons COR0021202 INTERMEDIATE RANGE MONITOR Related Objectives COR0021202001050F Describe the IRM system design features and/or interlocks that provide the following: Alarm seal-in Related References COR0021202 Intermediate Range Monitor Related Skills (K/A) 215003 A3.02 Ability to monitor automatic operations of the INTERMEDIATE RANGE MONITOR (IRM) SYSTEM including: (CFR: 41.7 / 45.7) Annunciator and alarm signals (3.3)

QUESTION: 10 21774 During a Reactor Startup, the following indications were observed;

IRM F indication rose from 50/125 to 122/125 on range 6.

All other IRMs remained at approximately 50/125 on range 6.

All automatic actions took place at their Tech Spec Values.

Following the observations above, the following events occurred;

IRM F indication returned back to 50/125 from 122/125 on range 6.

All other IRMs remained at approximately 50/125 on range 6.

On panel 9-5 the control room operator performed the following actions:

bypassed IRM F.

reset all automatic actions that resulted from the actions of IRM F.

depressed the Annunciator acknowledge and reset pushbuttons as appropriate.

For this event, what is the status of the Alarms and Alarm Indicating Lights, both on Panel 9-5 and on Panel 9-12? (Exclude the IRM bypass lights)

The alarms on Panel 9-5 are

a. clear; the alarm indicating lights on Panels 9-5 are reset and are NOT illuminated but the ones on Panel 9-12 are NOT reset and still illuminated.

b. clear; the alarm indicating lights on BOTH Panels 9-5 and 9-12 are reset and are NOT illuminated.

c. NOT clear; the alarm indicating lights on Panels 9-5 are reset and are NOT illuminated, but the ones on Panel 9-12 are NOT reset and are still illuminated.

d. NOT clear; the alarm indicating lights on NEITHER Panel 9-5 or 9-12 are reset and are still illuminated.

ANSWER: 10 21774

a. clear; the alarm indicating lights on Panels 9-5 are reset and are NOT illuminated but the ones on Panel 9-12 are NOT reset and still illuminated.

Explanation:

NOTE: See attached figure from USAR Reference APED-5706 Student Text COR002-12-02 (Rev 13)Section II.L.2.e describes the IRM trip units and the auto reset/Seal-in indications.

The 9-5 Panel alarm indicating lights automatically reset when the trip condition clears. The 9-12 Panel lights are seal-in and must be manually reset on the IRM drawer face.

Distractors:

b. Back panel indicating lights are still illuminated.

c. Front panel indicating lights are out.

d. Alarms on 9-5 are clear and indicating lights on 9-5 are reset.

Taken from USAR VII-5.7.3.5 Reference APED-5706. Note TRIP RESET block for Trip Units which output ot LOCAL LAMP which is located on Panel 9-12. The local lamps must be manually reset.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 215004 G2.1.27 Importance Rating 3.9 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): Instrument Operating Procedure 4.1.1 (Rev 21)

(Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 11 New 00 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 1 1 2 Multiple Choice Topic Area Description Systems COR00230020010900 Source Range Monitor Subsystem Related Lessons COR0023002 SOURCE RANGE MONITOR Related Objectives COR0023002001010A State the purpose of the following items related to the SRM system:

SRM System Related References NONE Related Skills (K/A) 215004 Source Range Monitor (SRM) System G 2.1.27 Knowledge of system purpose and/or function (CFR: 41.7) (3.9).

QUESTION: 11 What is the purpose of the Source Range Monitor (SRM) system?

The SRM system

a. monitors local thermal neutron flux and provides input to the Rod Block Monitor (RBM) system.

b. provides a continuous indication and permanent record of the core bulk thermal power in the power range.

c. monitors the core neutron flux levels and their rate of change during shutdown, refueling, and startup.

d. automatically detects and blocks control rod withdrawal that could violate MCPR limits.

ANSWER: 11

c. monitors the core neutron flux levels and their rate of change during shutdown, refueling, and startup.

Explanation:

From Instrument Operating Procedure 4.1.1 (Rev 21), Attachment 2 Strep 1.1.1:

SRMs provide neutron flux information during reactor startup, low flux level operation, and while reactor is shutdown.

Distractors:

a. This is the purpose of the Local Power Range Monitor (LPRM) system.

b. This is one purpose of the Average Power Range Monitor (APRM) system.

d. This is the purpose of the Rod Block Monitor (RBM) system.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 215005 A4.04 Importance Rating 3.2 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): USAR VII-5.8 (Attach if not previously provided) Student Text COR002-13-02 (Rev 15)

(including version/revision number) Procedure 4.1.3 (Rev 25)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 12 New 00 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 3 1 4 Multiple Choice Topic Area Description Systems COR0021302 LPRM System downscale trip indication Related Lessons COR0021302 OPS LOCAL POWER RANGE MONITOR Related Objectives COR00213020010900 Predict the changes in indication associated with operating a given control in the LPRM system.

Related References NONE Related Skills (K/A) 215005 A4.04 Ability to manually operate and/or monitor in the control room: (CFR: 41.7 /

45.5 to 45.8) LPRM back panel switches, meters and indicating lights (3.2)

QUESTION: 12 What position must the associated meter function switches be placed when the control room operator reads the LPRM output for the above illuminated indicating light?

What does the output on the meter face indicate?

One switch pointing to

a. 1D position and one switch pointing to the COUNT position.

The meter indicates upscale.

b. 1D position and one switch pointing to the COUNT position.

The meter indicates downscale.

c. the 1 position and one switch pointing to the D position.

The meter indicates upscale.

d. the 1 position and one switch pointing to the D position.

The meter indicates downscale.

ANSWER: 12

d. the 1 position and one switch pointing to the D position.

The meter indicates downscale.

Explanation:

Student Text COR002-13-02 (Rev 15) explains the function switches. Instrument Procedure 4.1.3 (Rev 25) describes the number of LPRMs per APRM/APRM association and the downscale trip setting. USAR VII-5.8 describes the indicating lights but not the order displayed on the top of Panel 9-14.

NOTE: See attached figures for depiction of the LPRM indicating lights and meter function switches.

The illuminated light is associated with LPRM string 28-45, detector D. The illuminated LPRM light indicates the LPRM output is downscale (<3 watts/cm2 ). The order of the lights for each of the LPRM detectors is: (left to right) BYPASS, DOWNSCALE, AND UPSCALE. In order for the operator to read the LPRM output, one meter function switch (S3) must be in the 1 position and the other (S2) in the D position. With the meter function switches in these positions, the output on the associated meter will be downscale.

Distractors:

a. The COUNT position will display 5% per LPRM string not bypassed and providing an input to the associated meter. The meter would indicate approximately 70% (14 LPRM inputs times 5).

b. The COUNT position will display 5% per LPRM string not bypassed and providing an input to the associated meter. The meter would indicate approximately 70% (14 LPRM inputs times 5).

c. The function switches are in the correct position, but the meter output will be indicating downscale.

BYPASS DOWNSCALE UPSCALE

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 217000 K1.02 Importance Rating 3.5 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): Burns & Roe P&ID drawing 2041 (Rev N85)

(Attach if not previously provided) Burns & Roe P&ID drawing 2043 (Rev N54)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR Part 55 Content: 55.41 3 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 13 New 00 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 1 1 4 Multiple Choice Topic Area Description Systems COR0021802 RCIC System Related Lessons COR0021802 OPS Reactor Core Isolation Cooling System Related Objectives COR0021802001050B Describe the interrelationship between RCIC system and the following: Nuclear Boiler Related References NONE Related Skills (K/A) 217000 K1.02 Knowledge of the physical connections and/or cause-effect relationships between REACTOR CORE ISOLATION COOLING SYSTEM (RCIC) and the following: (CFR: 41.2 to 41.9 / 45.7 to 45.8) Nuclear boiler system (3.5)

QUESTION: 13 Where does RCIC receive its steam supply and what feedwater (FW) line does it inject?

MS Line FW Line

a. A B

b. C A

c. C B

d. A A ANSWER: 13

b. C A Explanation:

NOTE: See attached B&R P&ID drawings 2041 and 2043 B&R P&ID drawings 2041 (Rev N80) and 2043 (Rev N54) show Main Steam Line C as the steam supply and Feedwater Line A as the injection flow path.

Distractors:

a. RCIC does not receive steam supply from MSL A. HPCI injects into FW Line B.

c. RCIC receives steam from MS Line C but HPCI injects into FW Line B.

d. RCIC does not receive steam supply from MSL A but injects into FW Line A.

B&R P&ID 2043 RCIC injection line Feedwater line A

B&R P&ID 2041 Steam Line C

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 218000 A1.04 Importance Rating 4.1 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): Tech Specs Table 3.3.5.1-1 Function 4b (Attach if not previously provided) GE Drawing 791E253 (Rev N28) Sheet 2 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR Part 55 Content: 55.41 5 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 14 New 00 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 3 1 3 Multiple Choice N Topic Area Description Systems COR0021602001050B, COR0021602001060A Nuclear Pressure Relief Related Lessons COR0021602 OPS NUCLEAR PRESSURE RELIEF Related Objectives COR0021602001040C Given a Nuclear Pressure Relief system component manipulation, predict and explain the changes in the following parameters:

Reactor Pressure Related References NONE Related Skills (K/A) 218000.A1.04 Ability to predict and/or monitor changes in parameters associated with operating the AUTOMATIC DEPRESSURIZATION SYSTEM controls including:

(CFR: 41.5 / 45.5) Reactor pressure (4.1)

QUESTION: 14 An Automatic Depressurization System (ADS) initiation has occurred, and RPV blowdown is in progress. The following conditions exist:

The initiation signals are still present.

RPV Pressure is 950 psig and lowering.

The control room operator presses and releases the ADS LOGIC A TMR and ADS LOGIC B TMR RESET pushbuttons.

What is the RPV pressure response over the next five minutes?

a. RPV pressure stops lowering while the RESET pushbuttons are depressed and immediately resumes lowering when the RESET pushbuttons are released.

b. RPV pressure stops lowering and continues rising until the safety valves open.

c. RPV pressure stops lowering for 109 seconds, and then resumes lowering.

d. RPV pressure continues lowering without pause.

ANSWER: 14

c. RPV pressure stops lowering for 109 seconds, and then resumes lowering.

Explanation:

NOTE: See attached figures for ADS Logic A and ADS valve A.

Tech Specs Table 3.3.5.1-1 Function 4b, ALLOWABLE VALUE for ADS System Initiation Timer is 109 seconds. General Electric Electrical Drawing 791E253 (Rev N28) Sheet 2 depicts ADS Valve control logic.

The ADS valves (A, B, C, E, G, and H) solenoid being energized (SRV open) when relays 2E-K6A and 2E-K7A and/or 2E-K6B and 2E-K7B are energized and their related contacts are closed. Relays 2E-K5A for Logic A and 2E-K5B have time delay pickups of 109 seconds and then they re-energize after the RESET pushbuttons are released. Relays 2E-K7A and 2E-K7B immediately re-energize when the pushbuttons are released and 109 seconds later relays 2EK-6A and 2E-K6B are re-energized. At this time the SRV solenoids are re-energized and the ADS valves re-open and remain open.

Distractors:

a. Once the Timer RESET PBs are released, the timer restarts timing from 109 seconds.

b. Given the plant conditions, the ADS actuates on low RPV level with a low pressure ECCS pump running. A full PCIS Group 1 (MSIV) is present so RPV pressure will begin rising from 950 psig. The safety valves open at 1240 +/- 37.2 psig (TS SR3.4.3.1) and this pressure would not be reached prior to the 109 second timer timing out. Also, the SRVs safety function lift setpoints are 1080, 1090 and 1100 psig and they would open prior to prevent the safety valves from opening. The student may confuse the ADS INHIBIT switch with the RESET PBs because using this switch would prevent SRVs from re-opening.

d. The valves close while the PBs are depressed and then re-open in 109 seconds.

Closed when Core RESET PUSHBUTTONS Spray or LPCI Pump running Closes in 109 sec TIMER Immediately re-energizes 2E-K6A energizes when when PB released.

timer times out.

2E-K6A & 2E-K7A contact closing will energize solenoid.

Same with B channel.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 223002 K6.08 Importance Rating 3.5 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): Procedure 2.1.22 (Rev 55)

(Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank # 626 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 15 626 2 01/26/1999 06/15/2005 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 1 1 1 Multiple Choice Topic Area Description Systems COR0022102 Reactor Protection System Related Lessons COR0022102 REACTOR PROTECTION SYSTEM Related Objectives COR0022102001080C Given a specific RPS malfunction, determine the effect on any of the following: PCIS Related References NONE Related Skills (K/A) 223002 K6.08 Knowledge of the effect that a loss or malfunction of the following will have on the PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF : (CFR: 41.7 / 45.7) Reactor protection system (3.5)

QUESTION: 15 626 What effect does the tripping of RPS MG Set A EPAs have on the Primary Containment Isolation System?

a. Half Group 1, 2, 3, 6 and 7 isolations.

b. Full Group 1, 2, 3, 6 and 7 isolations.

c. Half Group 1 and 2 isolations.

Full Group 3, 6 and 7 isolations.

d. Half Group 1, 2 and 7 isolations.

Full Group 3 and 6 isolations.

ANSWER: 15 626

d. Half Group 1, 2 and 7 isolations.

Full Group 3 and 6 isolations.

Explanation:

General Operating Procedure 2.1.22 (Rev 55) describes half isolations and contains Notes describing a loss of RPS causing the half isolation.

The Reactor Protection System (RPS) supplies divisional power to the PCIS Group 1, 2, 3, 6 and 7 isolation logics. PCIS is a one-out-of-two-taken-twice logic for those that receive a half group isolation on a loss of logic power. The PCIS logic is fail-safe in that the relays must de-energize to actuate the Group isolations. RPS Bus A causes a half Division 1 Group 1, Group 2 (the inboard valves close) and Group 7 (inboard valve closes). PCIS Groups 3 is cross-divisional on the loss of RPS A side only as the high temperature for the non-regenative heat exchanger outlet temperature switch relaying so in this case a full Group 3 occurs. PCIS Group 6 is cross-divisional always.

a. A full Group 3 and 6 occur.

b. Group 1, 2 and 7 are only half PCIS isolations although half the valves (inboard) close.

c. Group 7 is only a half group.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 239002 K3.03 Importance Rating 4.3 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): USAR IV-4.6 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 16 New 00 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 3 1 4 Multiple Choice Topic Area Description Systems COR0021602 OPS NUCLEAR PRESSURE RELIEF Related Lessons COR0021602 OPS NUCLEAR PRESSURE RELIEF Related Objectives COR0021602001070B Given a specific NPR malfunction, determine the effect on any of the following: Ability to rapidly depressurize the reactor Related References NONE Related Skills (K/A) 239002 K3.03 Knowledge of the effect that a loss or malfunction of the RELIEF/SAFETY VALVES will have on following: (CFR: 41.7 / 45.4) Ability to rapidly depressurize the reactor (4.3)

QUESTION: 16 Due to ongoing events, the following are displayed in the control room:

Alarm 9-3-1/C-2, DRYWELL PNEUMATIC HDR LOW PRESSURE VID Display (1029) DRYWELL PNEUMATIC HDR PRESSURE LOW

Alarm 9-3-1/D-2, RELIEF VALVE ACCUMULATOR LOW PRESSURE VID Display (1030) RELIEF VALVE A ACCUMULATOR PRESSURE LOW VID Display (1031) RELIEF VALVE B ACCUMULATOR PRESSURE LOW VID Display (1032) RELIEF VALVE C ACCUMULATOR PRESSURE LOW VID Display (1033) RELIEF VALVE D ACCUMULATOR PRESSURE LOW VID Display (1034) RELIEF VALVE E ACCUMULATOR PRESSURE LOW VID Display (1035) RELIEF VALVE F ACCUMULATOR PRESSURE LOW VID Display (1036) RELIEF VALVE G ACCUMULATOR PRESSURE LOW

The MSIVs are closed.

The reactor is shutdown with pressure at rated conditions. How can the control room operator lower RPV pressure so the Condensate Booster Pumps can inject into the RPV?

a. Close IA-SOV-SPV21, DRYWELL IA SUPPLY VLV from Panel 9-3, and SRV valves are opened with their control switches as needed.

b. Open SRV VALVES 71E, 71F, 71G by placing their Isolation switches to ISOL in the ASD room.

c. Open the main turbine bypass valves in MANUAL mode from an HMI.

d. Open SRV VALVE 71H with its control switch.

ANSWER: 16

d. Open SRV VALVE 71H with its control switch.

Explanation:

NOTE: See attached figure of B&R P&ID drawing 2010 (Rev N95) depicting Instrument Air and Nitrogen supply to the SRV solenoids.

USAR IV-4.6 states: Six ADS accumulators are tested to ensure that they will provide sufficient motive force to actuate the main steam relief valves at least five times at atmospheric drywell pressure after being isolated from the nitrogen supply for one hour. The five actuations at atmospheric drywell pressure are equivalent to the two actuations at 70% drywell design pressure.

The SRVs utilize pneumatic to operate in manual. The SRV accumulators have enough stored pneumatics to operate the SRV for five operations with the drywell at atmospheric conditions.

The conditions above indicate the SRV air and nitrogen supplies are low and the accumulators have lost their pressure to manually operate the valves. Only SRV H is available for use. The control room operator can lower RPV pressure from 1000 psig to 550 psig without exceeding RPV cooldown rate limits so the SRV will be opened one time. SRV H (ADS associated valve) accumulator pressure is enough to open the valve and depressurize the RPV to ~550 psig so the Condensate Booster Pumps can be utilized to inject into the RPV.

Distractors:

a. This valve has to be opened to allow instrument air to be delivered to the drywell and re-pressurize the SRV accumulators.

b. Operating these SRVs from the ASD room would not work. Also, not only must the ISOLATION switch be placed in ISOL, but each valve must be opened with their control switch. Just placing the ISOLATION switch in ISOL will not open the valve.

c. The MSIVs are closed due to pneumatics being lost. Opening the bypass valves will not depressurize the RPV.

The SRVs require either nitrogen or instrument air to operate utilizing a control switch.

SPV21 is normally closed at power but would be open at this power level as primary containment does not have to be inerted within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of operating at 15% power (Tech Spec LCO 3.6.3.1).

Instrument Air supply Nitrogen supply Typical SRV solenoid

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 239002 K5.04 Importance Rating 3.3 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): Station Operating Procedure 2.2.1 (Rev 37)

(Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR Part 55 Content: 55.41 5 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 17 New 0 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 2 1 1 Multiple Choice Topic Area Description Systems COR0021602 Nuclear Pressure Relief Related Lessons COR0021602 OPS NUCLEAR PRESSURE RELIEF Related Objectives COR0021602001060D Briefly describe the following concepts as they apply to NPR: Tail pipe temperature monitoring Related References NONE Related Skills (K/A) 239002 K5.04 Knowledge of the operational implications of the following concepts as they apply to RELIEF/SAFETY VALVES : (CFR: 41.5 / 45.3) Tail pipe temperature monitoring (3.3)

QUESTION: 17 The plant is starting up and in MODE 1. SRV testing has been completed satisfactorily and two hours have lapsed. MS-TR-166, MAIN STEAM RELIEF AND SAFETY VALVE LEAK TEMP RECORDER, Channel 5 for Relief/Safety Valve B is indicating 195°F and steady. PMIS Point T143, MS RELIEF VALVE B is indicating 195°F and steady. No control room alarms associated with safety relief valves are present.

What is the condition of SRV B?

a. The valve is leaking.

b. The valve is full open.

c. The tailpipe thermocouple is failed open.

d. The valve is closed and indicating normally.

ANSWER: 17

a. The valve is leaking.

Explanation:

System Operating Procedure 2.2.1 (Rev 37) Section 4, STANDBY OPERATION, contains a NOTE stating the valve is closed if tailpipe temperatures are ~140°F to 158°F. Section 6 MANUAL OPERATION, states to check for alarm 9-3-1/C-1, SAFETY/RELIEF VALVE LEAKING comes in when a valve is opened. This alarm is received if a tailpipe temperature reaches 300oF.

The normal temperature range for a closed SRV is ~140°F to 158°F. The valve is NOT fully closed so it is leaking. A full open SRV tailpipe temperature is ~ 285-300oF.

Distractors:

b. A full open valve tailpipe temperature ~ 285-300oF. A valve leaking alarm would be present if the valve were open.

c. A failed open thermocouple causes a valve leaking alarm. A failed thermocouple (RTD) for this system makes the temperature recorder go full upscale or downscale. The temperature indication is midscale.

d. A closed valve normal indication is ~140°F to 158°F.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 259002 A4.01 Importance Rating 3.8 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): Station Operating Procedure 2.2.28 (Rev 95).

(Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 18 New 00 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 2 1 4 Multiple Choice Topic Area Description Systems COR0023202 Loss of 3 speed probes when in MDEM Related Lessons COR0023202 OPS REACTOR VESSEL LEVEL CONTROL Related Objectives COR0023202001020K Describe the interrelationship between RVLC and the following: HMIs COR0023202001050A Briefly describe the following concepts as they apply to the RVLC system: Limiting RFP speed for two RFP operation COR0023202001060B Predict the consequences of the following on the RVLC system:

Speed Feedback Signal Failure Related References NONE Related Skills (K/A) 259002 A4.01 Ability to manually operate and/or monitor in the control room: (CFR: 41.7 /

45.5 to 45.8) All individual component controllers in the manual mode (3.8)

QUESTION: 18 The plant is starting up and the second Reactor Feed Pump has been placed in service one minute ago. RFPT-1A and RFPT-1B are being operated in MDEM mode. The three speed probes for RFPT-1A turn INVALID.

How are both RFPTs controlled in this condition?

a. RFPT A speed is changed utilizing the HMI UP/DOWN arrows in MDVP.

RFPT B speed is changed utilizing the HMI UP/DOWN arrows in MDVP.

b. RFPT A speed cannot be changed due to the loss of three speed probe inputs.

RFPT B speed is changed utilizing the HMI UP/DOWN arrows in MDEM.

c. RFPT A speed is changed utilizing the HMI UP/DOWN arrows in MDEM.

RFPT B speed is changed utilizing the HMI UP/DOWN arrows in MDEM.

d. RFPT A speed is changed utilizing the HMI UP/DOWN arrows in MDVP.

RFPT B speed is changed utilizing the HMI UP/DOWN arrows in MDEM.

ANSWER: 18

d. RFPT A speed is changed utilizing the HMI UP/DOWN arrows in MDVP.

RFPT B speed is changed utilizing the HMI UP/DOWN arrows in MDEM.

Explanation:

NOTE: See attached Procedure 2.2.28 partial Station Operating Procedure 2.2.28 (Rev 95) Attachment 2 (Step 1.2.6.2) describes the different control modes and the loss of all three speed probes.

With both RFPTs operating in MDEM (Manual Demand), each controller utilizes three speed probes for feedback to set target speed settings for control of the RFPT. If alI three probes are invalid, control of that RFPT transfers to MDVP control and the speed of the RFPT is controlled by direct controller output to the nozzle block. The control room operator can only change controller out put to adjust speed.

Distractors:

a. RFPT A speed can be changed in MDVP only. RFPT B remains in MDEM and does not transfer to MDVP control.

b. RFPT A speed can be changed in MDVP mode. RFPT B remains in MDEM and does not transfer to MDVP control.

c. RFPT A is controlled in MDEM but RFPT B control transfers to MDVP.

Procedure 2.2.28 partial ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 261000 A2.11 Importance Rating 3.2 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): General Operating Procedure 2.1.22 (Rev 55)

(Attach if not previously provided) System Operating Procedure 2.2.73 (Rev 50)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR Part 55 Content: 55.41 5 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 19 New 00 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 3 1 4 Multiple Choice Topic Area Description Systems COR0022802, Auto start of standby train Related Lessons COR0022802 OPS STANDBY GAS TREATMENT Related Objectives COR0022802001050B Describe the interrelationships between SGT and the following:

Primary Containment COR0022802001080A Describe the Standby Gas Treatment design features and/or interlocks that provide for the following: Automatic system initiation Related References NONE Related Skills (K/A) 261000 A2.11 Ability to (a) predict the impacts of the following on the STANDBY GAS TREATMENT SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5 / 45.6) High containment pressure (3.2)

QUESTION: 19 With the plant at power, the following timeline of events/actions occur:

0800 Drywell pressure starts rising.

0805 The control room operator is aligning for venting.

0807 The reactor automatically scrams on high primary containment pressure.

0830 Reactor building pressure is indicating -0.30 inches wg.

At 0845 the control room operator is directed to respond to primary containment isolations per Procedure 2.1.22, RECOVERING FROM A GROUP ISOLATION.

Considering the SGT system, what does procedure require to be performed within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months of the high primary containment trip signal?

a. Both SGT fan control switches must be placed in OFF and then to STANDBY.

b. Both SGT fan control switches must be placed in RUN.

c. The preferred SGT fan control switch must be placed in RUN and the other SGT fan control switch must be placed in OFF and then to STANDBY.

d. The preferred SGT fan control switch must be placed in OFF and then to STANDBY and the other SGT fan control switch must be placed in RUN.

ANSWER: 19

c. The preferred SGT fan control switch must be placed in RUN and the other SGT fan control switch must be placed in OFF and then to STANDBY.

Explanation:

NOTE: See attached Procedures 2.1.22 and 2.2.73 partials.

General Operating Procedure 2.1.22 (Rev 55) Step 9.4 requires the SGT trains aligned per its system operating procedure 2.2.73 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . This is a commitment to the NRC (Commitment NLS2000029-01). System Operating Procedure 2.2.73 (Rev 50) Section 4.2 provides guidance to place preferred SGT train to RUN and the other train in a standby configuration. Note the PREFERRED train is identified by a red tag placed on the SGT fan switch.

Distractors:

a. The preferred train must be placed in RUN and the other train in STANDBY.

b. The non-preferred train must be placed in STANDBY.

d. The preferred train is not placed in STANDBY, the other train is.

Procedure 2.1.22 partial 9.4 Align SGT per Procedure 2 .2 .73 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months of receiving Group 6 .

Procedure 2.2.73 partial

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 262001 G2.2.36 Importance Rating 3.1 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): Technical Specifications LCOs 3.8.1 and 3.8.2 (Attach if not previously provided) Technical Specifications LCOs 3.8.7 and 3.8.8.

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # 23132 (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 20 23132 01 02/25/2009 06/15/2005 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 3 1 5 Multiple Choice Topic Area Description Technical Specifications, INT0070509, What specification is entered Related Lessons INT0070509 OPS Tech. Spec. 3.8, Electrical Power Systems Related Objectives INT00705090010100 Given a set of plant conditions, recognize non-compliance with a Section 3.8 LCO.

Related References NONE Related Skills (K/A) 262001 A.C. Electrical Distribution G2.2.36 Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations (CFR: 41.10 / 43.2 /

45.13) (3.1)

QUESTION: 20 23132 The plant is operating at rated conditions. The Electrical Maintenance department is troubleshooting a relay on 4160VAC critical bus 1F. Troubleshooting activities cause the bus to lockout. The control room operators respond according to alarm procedure guidance.

What condition and action statements of Technical Specifications require entry?

a. TS 3.8.1 AC Sources - Operating only.

b. TS 3.8.2 AC Sources - Shutdown only.

c. TS 3.8.1 AC Sources - Operating and TS 3.8.7 Distribution Systems - Operating.

d. TS 3.8.2 AC Sources - Shutdown and TS 3.8.8 Distribution Systems - Shutdown.

ANSWER: 20 23132

c. TS 3.8.1 AC Sources - Operating and TS 3.8.7 Distribution Systems - Operating.

Explanation:

NOTE: See attached TS LCOs 3.8.1 and 3.8.7 Due to the loss of Critical Bus 1F, the operators are required to scram the reactor. The plant is in Mode 3 post scram. Critical bus 1F is de-energized and locked out. With the bus de-energized, TS 3.8.7 becomes applicable as TS 3.8.7 Basis requires the bus to be energized.

Because the bus is de-energized it also makes the off-site power source to the bus inoperable.

TS 3.8.1 Basis requires the off-site power source to be able to power the critical bus in order to be operable. TS 3.8.1 and 3.8.7 apply in this condition.

Distractors:

a. TS 3.8.1 applies but so does TS 3.8.7.

b. TS 3.8.2 only applies in Modes 4 and 5.

d. TS 3.8.2 and 3.8.8 only apply in Modes 4 and 5.

ORIGINAL QUESTION 23132 The plant is in Mode 3 following an inadvertent scram. A loss of all offsite AC power occurs.

Both Diesel Generators start and load their respective busses.

What condition and action statements of Technical Specifications require entry?

a. TS 3.8.1 AC Sources - Operating only.

b. TS 3.8.2 AC Sources - Shutdown only.

c. TS 3.8.1 AC Sources - Operating and TS 3.8.7 Distribution Systems - Operating.

d. TS 3.8.2 AC Sources - Shutdown and TS 3.8.8 Distribution Systems - Shutdown.

ANSWER: 41 23132

a. TS 3.8.1 AC Sources - Operating only.

Explanation:

TS 3.8.1 is applicable in modes 1, 2 and 3. And the plant is in mode 3 with offsite power unavailable.

Distractors:

b. is incorrect because this spec is applicable in modes 4 and 5.

c. is incorrect because 3.8.7 does not require entry.

d. is incorrect because 3.8.8 and 3.8.2 do not require entry.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 262002 A3.01 Importance Rating 2.8 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): System Operating Procedure 2.2.22 (Rev 63)

(Attach if not previously provided) B&R Electrical Drawing Sheet 16 (Rev N25)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank # 25667 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam Yes Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 21 25667 00 05/25/2011 06/15/2005 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 2 1 5 Multiple Choice Topic Area Description Systems MCC-R provides power to NBPP via step down transformer and inverter cabinet switch.

Related Lessons COR0010102 AC Electrical Distribution Related Objectives COR0010102001090G Describe the AC Electrical Distribution System design feature(s) and/or interlock(s) that provide for the following: Transfer from preferred power to alternate power supplies Related References NONE Related Skills (K/A) 262002 A3.01 Ability to monitor automatic operations of the UNINTERRUPTABLE POWER SUPPLY (A.C./D.C.) including: (CFR: 41.7 / 45.7) Transfer from preferred to alternate source (2.8)

QUESTION: 21 25667 The plant is in a normal full power electrical lineup. The following alarm is received:

C-4/E-7 NO BREAK SYSTEM INVERTER 1A VOLTAGE FAILURE.

The electrical system responds as designed. What is the source of power to the NBPP?

a. MCC-L via a step down transformer and bypassing the inverter cabinet static switch.

b. MCC-L via a step down transformer and the inverter cabinet static switch.

c. MCC-R via a step down transformer and bypassing the inverter cabinet static switch.

d. MCC-R via a step down transformer and the inverter cabinet static switch.

ANSWER: 21 25667

d. MCC-R via a step down transformer and the inverter cabinet static switch.

EXPLANATION:

NOTE: See attached power supply scheme and simplified drawing of Inverter Cabinet.

System Operating Procedure 2.2.22 (Rev 63) describes NBPP automatic transfer to MCC-R.

See Attachment 1 Steps 1.2.5 and 2.3.

The inverter failure alarm indicates that the power into or out of the inverter is failed causing the NBPP to transfer to MCC-R. MCC-R power goes from the MCC to a step down (115V AC) transformer to the static switch in the inverter cabinet. From the static switch, power goes to the NBPP.

Distractors:

a. MCC-L powers the PMIS inverter as an alternate supply. MCC-LX supplies power to the 250VDC Bus 1A battery charger.

b. MCC-L powers the PMIS inverter as an alternate supply. MCC-LX which supplies power to the 250VDC Bus 1A battery charger.

c. MCC-R automatically powers NBPP through the Static switch. To feed the NBPP directly requires a MANUAL BYPASS SWITCH to be utilized at the inverter cabinet.

Normal Supply 250VDC Bus 1A Emergency Supply MCC-R Burns & Roe Electrical drawing 3010 Sheet 16 (rev N25)

FROM 250 VDC EMERGENY AC BUS 1A FROM 480V CRITICAL MCC-R NO-BREAK SYSTEM INVERTER 1A VOLTAGE 200 A FAILURE 80 A C

27 F V V A

A INVERTER STATIC SWITCH 1A 27 C NO-BREAK SYSTEM F EMERGENCY AC FAILURE 27 V

V LEGEND:

A A BYPASS XFMR/INV A - AMMETER SELECTOR SWITCH F - FREQUENCY METER C INVERTER CABINET V C LOSS OF V - VOLTMETER AC POWER 27 27 - UNDERVLOTAGE RELAY VS C C - PANEL C (CONTROL ROOM)

VS - VOLTMETER SELECTOR SWITCH NO-BREAK POWER SYSTEM Figure 7, Rev. 9 f:\home\jyknapp\figures\cxa06278\cor00101.fig\fig7.r09 COR001-01 CXA06278

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 263000 K2.01 Importance Rating 3.1 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): System Operating Procedure 2.2.77A (Rev 7)

(Attach if not previously provided) B&R Drawing 3058 (Rev N56)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 22 New 00 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 1 1 3 Multiple Choice Topic Area Description Electrical COR0020702, DC Electrical Distribution Related Lessons COR0020702 OPS DC ELECTRICAL DISTRIBUTION Related Objectives COR0020702001060C Describe the interrelationship between the DC Electrical Distribution System and the following: Battery charger and battery COR00207020010300 State the design bases for the DC Electrical Distribution System as described in the associated Student Text.

COR0020702001080D Given a specific DC Electrical Distribution system malfunction, determine the effect on any of the following: Battery chargers Related References NONE Related Skills (K/A) 263000 K2.01 Knowledge of electrical power supplies to the following: (CFR: 41.7) Major D.C. loads (3.1)

QUESTION: 22 What is the normal power supply to the Main Turbine Emergency Bearing Oil Pump?

a. 125VDC Distribution Panel 1A via 125VDC SWGR 1A.

b. 125VDC Distribution Panel 1B via 125VDC SWGR 1B.

c. 250VDC Turbine Building Starter Rack via 250VDC SWGR 1A.

d. 250VDC Turbine Building Starter Rack via 250VDC SWGR 1B.

ANSWER: 22

d. 250VDC Turbine Building Starter Rack via 250VDC SWGR 1B.

Explanation:

NOTE: See attached B&R Electrical drawing section.

System Operating Procedure 2.2.77A (Rev 7) Attachment 2 lists the MT Emergency Bearing Oil Pump power supply as the 250VDC Turbine Building Starter Rack. B&R Drawing 3058 (Rev N56) depicts the 250VDC Turbine Building Starter Rack power supply and loads.

Distractors:

a. The MT Emergency Bearing Oil Pump is powered from 250VDC.

b. The MT Emergency Bearing Oil Pump is powered from 250VDC.

c. The MT Emergency Bearing Oil Pump is powered from SWGR 1B.

Turbine Building Starter Rack being powered from 250VDC SWGR 1B ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 263000 K5.01 Importance Rating 2.6 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): Student Text ACD002-31-11 (Rev 02)

(Attach if not previously provided) Student Text COR0020702 (Rev 29)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR Part 55 Content: 55.41 5 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 23 New 00 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 1 1 4 Multiple Choice Topic Area Description Systems COR0020702 OPS DC Distribution-hydrogen production during battery charging.

Related Lessons COR0020702 OPS DC Electrical Distribution Related Objectives COR0020702001060D Describe the interrelationship between the DC Electricial Distribution System and the following:: Battery ventilation COR0020702001100A Briefly describe the following concepts as they apply to DC Electrical Distribution System: Hydrogen generation during battery charging Related References NONE Related Skills (K/A) 263000 K5.01 Knowledge of the operational implications of the following concepts as they apply to D.C. ELECTRICAL DISTRIBUTION : (CFR: 41.5 / 45.3) Hydrogen generation during battery charging (2.6)

QUESTION: 23 The 125VDC Div 1 batteries are being charged after the charger had been removed from service for maintenance. Which of the following by-products could lead to serious damage to the 125VDC distribution system, if allowed to accumulate?

a. Potassium hydroxide

b. Hydroxide

c. Hydrogen

d. Arsine ANSWER: 23

c. Hydrogen Explanation:

Re-charging the batteries can produce excess hydrogen and oxygen if the batteries are overcharged. A combination of hydrogen and oxygen with an ignition source could cause a fire to damage the 125VDC system.

Distracters:

a. This is produced if an alkaline battery is charged.

b. This is produced on an nickel-cadmium battery overcharge.

d. Arsine (arsenic hydride) is a highly toxic by-product of some lead acid batteries, but this type is not used in the CNS 125VDC battery system.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 264000 A4.05 Importance Rating 3.6 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): Station Operating Procedure 2.2.18 (Rev 147)

(Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 24 New 00 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 2 1 4 Multiple Choice Topic Area Description Systems COR0020802 OPS DIESEL GENERATORS Transfer load from DG to critical 4160 Bus Related Lessons COR0020802 OPS DIESEL GENERATORS Related Objectives COR0020802001060B Describe the interrelationship between Diesel Generators and the following: AC Electrical Distribution COR0020802001100E Briefly describe the following concepts as they apply to the Diesel Generator system: Paralleling A.C. Power Sources Related References NONE Related Skills (K/A) 264000 A4.05 Ability to manually operate and/or monitor in the control room: (CFR: 41.7 /

45.5 to 45.8) Transfer of emergency generator (with load) to grid (3.6)

QUESTION: 24 Diesel Generator 1 automatically starts on a valid signal and is powering 4160V Bus 1F. The following exist:

All DG automatic initiation signals are clear.

4160V Bus 1A is energized.

DG 1 EMERGENCY TO NORMAL RESET is pushed and released.

DG 1 DROOP PARALLEL switch is in PARALLEL.

SYNCH SWITCH is in 1FA.

What indications are procedurally required to be present for the control room operator to transfer 4160V Bus 1F to 4160V Bus 1A?

The synchroscope has to be rotating slowly in the

a. counter-clockwise (SLOW) direction and DG voltage is slightly LOWER than Bus 1A voltage.

b. counter-clockwise (SLOW) direction and DG voltage is slightly HIGHER than Bus 1A voltage.

c. clockwise (FAST) direction and DG voltage is slightly LOWER than Bus 1A voltage.

d. clockwise (FAST) direction and DG voltage is slightly HIGHER than Bus 1A voltage.

ANSWER: 24

a. counter-clockwise (SLOW) direction and DG voltage is slightly LOWER than Bus 1A voltage.

Explanation:

Station Operating Procedure 2.2.18 (Rev 147) Section 24 provides guidance for transferring 4160V Bus 1F from DG1 to Bus 1A.

The DG speed, voltage and synchroscope parameters, and adjustments, are the opposite of the usual indication and control when transferring from the DG to a parallel with the grid. 4160V Bus 1A is energized from the Station Startup Transformer which is powered from off-site power.

The synchroscope has to rotate slowly in the counter-clockwise direction. The DG voltage must be lower than Bus 1A voltage or Bus 1A will not pick up any load and the DG will trip on reverse power.

Distracters:

b. If DG voltage is higher than 4160V Bus 1A voltage, Bus 1A will not pick up any load.

With the indication and controls operating reverse than normal, the student may confuse the voltage requirement.

c. If DG speed is too high, Bus 1A will not pick up load. With the indication and controls operating reverse than normal, the student may confuse the speed requirement

d. If DG speed or voltage is too high, Bus 1A will not pick up load. With the indication and controls operating reverse than normal, the student may confuse the speed and voltage requirement

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 300000 K4.02 Importance Rating 3.0 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): System Operating Procedure 2.2.59 (Rev 72)

(Attach if not previously provided) B&R P&ID Drawing 2010 (Rev N103) Sheet 1 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank # 7057 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 25 7057 00 08/31/2006 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 1 1 4 Multiple Choice N Topic Area Description Systems COR0011702 OPS Plant Air-SA Isolation from IA.

Related Lessons COR0011702 OPS Plant Air Related Objectives COR0011702001050D Describe the Plant Air system design features and/or interlocks that provide for the following: Service Air isolation COR0011701001060C Describe the operation of the interlocks associated with the following components in the Plant Air System: Plant Air Pressure Control Valve (PCV-609)

Related References NONE Related Skills (K/A) 300000 K4.02 Knowledge of (INSTRUMENT AIR SYSTEM) design feature(s) and or interlocks which provide for the following: (CFR: 41.7) Cross-over to other air systems (3.0)

QUESTION: 25 7057 What automatic interlock functions on a service air leak to prevent losing all air to the critical components in the Instrument Air (IA) System?

a. IA-SOV-SPV21, DRYWELL IA SUPPLY, automatically opens to supply the reliable air header.

b. IA-SOV-SPV21, DRYWELL IA SUPPLY, automatically closes to supply the reliable air header.

c. SA-PCV-609, SERVICE AIR SYSTEM ISOLATION, automatically closes to isolate the SA System from the IA System.

d. SA-PCV-609, SERVICE AIR SYSTEM ISOLATION, automatically opens to supply additional air to the IA System.

ANSWER: 25 7057

c. SA-PCV-609, SERVICE AIR SYSTEM ISOLATION, automatically closes to isolate the SA System from the IA System.

Explanation:

NOTE: See attached B&R 2010 Sheet 1 (Rev N103) P&ID partial drawing of SA-IA system.

System Operating Procedure 2.2.59 (Rev 72) Attachment 3, Steps 1.2.8 and 2.5 describe PCV-609 interlocks.

The outlet of the plant Air Receivers spits into two separate feeds. One feeds unfiltered Service Air throughout the plant via PCV-609. The other goes on to the Air Dryers and Filters for Instrument Air throughout the plant. The Service Air is isolated (PCV-609) on low Air Receiver pressure of 77 psig.

Distractors:

a. IA-SOV-SPV21 must be manually opened.

b. IA-SOV-SPV21 fails closed on loss of air but it does not automatically close.

d. SA-PCV-609 closes to separate the systems. The valve automatically opens when air pressure is restored and a reset pushbutton is depressed on the control room panel.

B&R P&ID 2010 From other Air Receiver To Air Dryers and Instrument Air Header From Air Receivers outlet crosstie header to Service Air PCV-609 isolates at 77 psig.

Service Air Header

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 400000 K2.01 Importance Rating 2.9 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): B&R Electrical Drawing 3006 (Rev N76) Sheet 5 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 26 New 00 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 1 1 2 Multiple Choice Topic Area Description Systems COR0021902001020B Reactor Equipment Cooling Related Lessons COR0021902 REACTOR EQUIPMENT COOLING Related Objectives COR0021902001020B State the electrical power supplies to the following REC components:

Pump Motors Related References NONE Related Skills (K/A) 400000 K2.01 Knowledge of electrical power supplies to the following: (CFR: 41.7) CCW pumps (2.9)

QUESTION: 26 What is the power supply for each REC pump?

a. 1A and 1B Pumps are powered from MCC-Q.

1C and 1D Pumps are powered form MCC-Y.

b. 1A and 1B Pumps are powered from MCC-K.

1C and 1D Pumps are powered form MCC-S.

c. 1A and 1C Pumps are powered from MCC-K.

1B and 1D Pumps are powered form MCC-S.

d. 1A and 1C Pumps are powered from MCC-Q.

1B and 1D Pumps are powered form MCC-Y.

ANSWER: 26

b. 1A and 1B Pumps are powered from MCC-K.

1C and 1D Pumps are powered form MCC-S.

Explanation:

See Attached: B&R Electrical Drawing 3006 (Rev N76) Sheet 5 depicts the MCC breaker drops for the REC Pumps.

Distractors:

a. MCC Q and Y supply power to several of the REC system valves. These MCCs are powered from MCC K and R, respectfully.

c. Normal CNS Divisional separation labeling is A and C in Division 1 and B and D in Division 2. The REC pumps are coupled A and B in Division 1 and C and D in Division 2 which is out of normal.

d. MCC Q and Y supply power to several of the REC system valves. These MCCs are powered from MCC K and R, respectfully.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 201003 A1.03 Importance Rating 2.9 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): System Operating Procedure 2.2.8 (Rev 85)

(Attach if not previously provided) Attachment 3 Step 1.2.4 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR Part 55 Content: 55.41 6 55.43 Comments:

Question more appropriately satisfies 55.41.6 vice 55.41.5.

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 27 New 00 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 1 1 4 Multiple Choice Topic Area Description Systems COR0020502 Amount of drive water flow on rod insertion Related Lessons COR0020502 CONTROL ROD DRIVE MECHANISM Related Objectives COR0020502001100C Briefly describe the following concepts as they apply to CRDMs:

Hydraulics COR0020502001120A Determine the interrelationships between the CRDMs and the following: Control Rod Drive Hydraulic System Related References NONE Related Skills (K/A) 201003 A1.03 Ability to predict and/or monitor changes in parameters associated with operating the CONTROL ROD AND DRIVE MECHANISM controls including:

(CFR: 41.5 / 45.5) CRD drive water flow (2.9)

QUESTION: 27 Control rod 26-27 is inserted from notch 16 to 12.

What does the drive water flow meter, CRD-FI-305 on Panel 9-5, indicate during the rod movement?

a. 2 gpm

b. 4 gpm

c. 6 gpm

d. 8 gpm ANSWER: 27

b. 4 gpm Explanation:

NOTE: See attached simplified drawing of drive water flowpath.

System Operating Procedure 2.2.8 (Rev 85) Attachment 3 Step 1.2.4 states: Drive water pressure ~ 265 psi above reactor pressure is required at a flow rate of ~ 4 gpm to insert a control rod.

For control rod insertion, CRD valves are sequenced as follows:

Valves 121 and 123 open.

Valves 121 and 123 close.

After a set time, valve 120 opens to settle the drive.

Then valve 120 closes.

Distractors:

a. This is the amount of flow for control rod withdrawal.

c. This is the total flow through the stabilizing valves when there is no control rod movement.

d. This is the drive flow momentarily observed when a control rod begins inward movement.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 202001 K4.04 Importance Rating 3.0 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): Station Operating Procedure 2.2.68 (Rev 76)

(Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 28 New 00 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 1 1 4 Multiple Choice Topic Area Description Systems COR0022202 REACTOR RECIRCULATION seal flow design Related Lessons COR0022202 REACTOR RECIRCULATION Related Objectives COR0022202001100E Describe the Reactor Recirculation system and/or Recirculation Flow Control system design features and/or interlocks that provide for the following: Controlled Seal Flow Related References NONE Related Skills (K/A) 202001 K4.04 Knowledge of RECIRCULATION System design feature(s) and/or interlocks which provide for the following: (CFR: 41.7) Controlled seal flow (3.0)

QUESTION: 28 How is the Reactor Recirculation seal flow rate controlled?

a. Setting a manual flow regulator between 0.25 and 0.5 gpm.

b. Setting a manual flow regulator between 1.6 and 1.8 gpm.

c. Seal internal orifices regulate flow between 0.25 and 0.5 gpm.

d. Seal internal orifices regulate flow between 1.6 and 1.8 gpm.

ANSWER: 28

b. Setting a manual flow regulator between 1.6 and 1.8 gpm.

Explanation:

Station Operating Procedure 2.2.68 (Rev 76) Step 4.11 provides guidance for setting RR pump seal flow.

1.1 If flow on CRD-FI-45A is not between 1.60 and 1.80 gpm, adjust CRD-FREG-46A as follows:

1.1.1 Loosen locknut on flow regulator stem.

CAUTION - Adjust stem by hand only. Using a wrench to adjust stem may provide too much torque and damage regulator.

1.1.2 (Independent Verification) Adjust CRD-FREG-46A to obtain between 1.60 and 1.80 gpm on CRD-FI-45A.

Performed By:

Verified By:

1.1.3 While holding stem in position, retighten locknut.

Distractors:

a. 0.25 gpm is the outer seal leak alarm setting and 0.5 gpm is the seal staging flow low alarm setting.

c. Seal flow is set by the manual flow regulator. 0.25 gpm is the outer seal leak alarm setting and 0.5 gpm is the seal staging flow low alarm setting.

d. Seal flow is set by the manual flow regulator.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 202002 K1.02 Importance Rating 4.2 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): System Operating Procedure 2.2.68.1 (Rev 67)

(Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank # 5060 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR Part 55 Content: 55.41 6 55.43 Comments:

Reactivity control of the jog circuit requires this question to be tied to 55.41.6 vice 55.41.5.

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 29 5060 00 10/28/1999 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 1 1 4 Multiple Choice Topic Area Description Systems COR0022202 Reactor Recirculation System Related Lessons COR0022202 REACTOR RECIRCULATION Related Objectives SKL012422200A030G Given plant conditions, predict changes in the following Reactor Recirculation System components/parameters: Reactor power COR0022202001040T Describe the interrelationships between Reactor Recirculation system or the Recirculation Flow Control system and the following: Reactor power.

Related References 2.2.68.1 Reactor Recirculation System Operations Related Skills (K/A) 202002 K1.02 Knowledge of the physical connections and/or cause-effect relationships between RECIRCULATION FLOW CONTROL SYSTEM and the following:

(CFR: 41.2 to 41.9 / 45.7 to 45.8) Reactor power (4.2*/4.2)

QUESTION: 29 5060 What is the reason for ensuring the Reactor Recirculation Pump Discharge Valve Jog Circuit is used when starting a RR Pump above 30% power?

The Discharge Valve Jog Circuit ensures . . .

a. the resultant shrink in Reactor level is within the capability of the Reactor Level Control System.

b. the resultant swell in Reactor level is within the capability of the Reactor Level Control System.

c. the subsequent rise in Reactor power does not exceed the flow biased High Flux scram setpoint.

d. any flow induced vibrations will not result in damage to the Jet Pumps.

ANSWER: 29 5060

c. the subsequent rise in Reactor power does not exceed the flow biased High Flux scram setpoint.

Explanation:

NOTE: See attached Procedure 2.2.68.1 (Rev 67) Section, PUMP START PRECAUTIONS (Step 2.2.5)

Distractors:

a. The circulation of cooler downcomer water through the core will cause some shrink, but it is well within the capability of the RFP response, but it is not the reason for the jog circuit.

b. The rise in downcomer level due to a rise in RFP flow does not occur. It might be expected to rise but the RR pump start takes inventory from the downcomer.

d. If the RR pump were started with core flow and power in the wrong region of the power to flow map, then vibration would be a concern.

Procedure 2.2.68.1 partial ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 204000 G2.4.41 Importance Rating 2.9 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): Procedure 5.7.1 (Rev 45) Attachment 3 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: Category F and Table F-1 of the EAL WALL CHART Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 30 New 00 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 3 1 4 Multiple Choice Topic Area Description Systems Related Lessons COR0012002 Reactor Water Cleanup System Related Objectives COR0012002001080E Predict the consequences a malfunction of the following would have on the RWCU system: PCIS COR0012002001090F Describe the RWCU design features and/or interlocks that provide for the following: System isolation Related References NONE Related Skills (K/A) 204000 Reactor Water Cleanup System G2.4.41 Knowledge of the emergency action level thresholds and classifications. (CFR:

41.10 / 43.5 / 45.11) (2.9)

QUESTION: 30 The plant is operating at rated conditions. The flow in the RWCU system piping downstream of RWCU-MO-18, OUTBD ISOL VLV, has reached 195% of rated conditions. No RWCU valve actuations have automatically occurred. All control room attempts to close RWCU valves have failed. Secondary Containment temperatures are rising.

Using the provided Wall Chart, what is the HIGHEST Emergency Action Level (EAL) reached for this condition?

a. An ALERT due to the loss of the Primary Containment Fission Product Barrier.

b. An ALERT due to the loss of the Secondary Containment Fission Product Barrier.

c. A Site Area Emergency due to the loss of the Reactor Coolant System Barrier and Primary Containment Barrier.

d. A Site Area Emergency due to the loss of the Fuel Clad Barrier and Reactor Coolant System Barrier.

ANSWER: 30

c. A Site Area Emergency due to the loss of the Reactor Coolant System Barrier and Primary Containment Barrier.

Explanation:

NOTE: See attached Table F-1 Fission Product Barrier Matrix and Category.

Procedure 5.7.1 (Rev 45) Attachment 3 states: The threshold is met if the breach is not isolable from the Control Room or an attempt for isolation from the Control Room has been made and was unsuccessful. An attempt for isolation from the Control Room should be made prior to the emergency classification. If Operator actions from the Control Room are successful, this threshold is not applicable. Credit is not given for Operator actions taken in-plant (outside the Control Room) to isolate the breach.

With high flow in the RWCU piping, a leak exists. At 191% of rated flow, the RWCU isolation valves (MO-15 and MO-18) receive a PCIS close signal. The valves failed to automatically close and control room attempts to close the valves failed. No credit is taken for in-plant actions. With the leak downstream of MO-18 (Outboard isolation valve) the leak is outside Primary containment so a loss of the RCS (Reactor Coolant System) fission product barrier is lost. Failure of the isolation valves to close results in the Primary Containment Barrier to be lost.

The loss of two barriers is a Site Area Emergency.

Distractors:

a. A loss of the Primary Containment Fission barrier results in a NOUE classification.

b. An ALERT on the loss of RCS exists.

d. The Fuel Clad Barrier is lost if RPV level or high radiation conditions exist.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 214000 K5.01 Importance Rating 2.7 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): Instrument Operating Procedure 4.3 (Rev 27)

(Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # 1223 (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge H Comprehension or Analysis 10 CFR Part 55 Content: 55.41 5 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 31 1223 00 5/10/2012 Licensed RO: Y Modified Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 2 1 3 Multiple Choice Topic Area Description Systems COR0022002001050L Reactor Manual Control System Related Lessons COR0022002 OPS REACTOR MANUAL CONTROL SYSTEM Related Objectives COR0022002001050L Predict the consequences the following would have on the RMCS and/or RPIS: Reactor scram COR0022002001040B Describe the RMCS design features and/or interlocks that provide for the following: Detection of drifting control rods COR00220020011400 List the RPIS reed switch locations.

Related References 4.3 Reactor Manual Control System and Rod Position Information System Related Skills (K/A) 214000 K5.01 Knowledge of the operational implications of the following concepts as they apply to ROD POSITION INFORMATION SYSTEM : (CFR: 41.5 / 45.3)

Reed switches (2.7)

QUESTION: 31 1223 A full reactor scram from rated conditions has been initiated. What will be the indication of a control rod that has moved past the full-in position 10 seconds after the scram (assume no operator action is taken)?

Full In Light Four Rod Display

a. ON Blank

b. ON 00

c. OFF Blank

d. OFF 00 ANSWER: 31 1223

a. ON Blank Explanation:

Instrument Operating Procedure 4.3 (Rev 27) describes the RPIS Reed Switch indication.

Section 11. OVERTRAVEL IN CHECK 11.1 With selected control rod at Position 00, hold ROD MOVEMENT CONTROL switch to IN for 3 to 5 seconds.

11.2 Check green FULL-IN light remains on.

11.3 Release switch and check rod settles to Position 00. Step 1.1.6:The green "full-in" lights on the full core display are provided from the outputs of reed Switch S00, S51, or S52.

With a scram present the reed switch that lights the green full in light is lit and the reed switch that provides the 00 indication on the four rod display is open as the rod is inserted slightly beyond this position.

Distractors:

b. 4-rod display indicates 00 only when the control rod is at position 00 and not overtravel in.

c. The full in light is off only when the control rod is not at position 00 or overtravel in.

d. The full in light is off only when the control rod is not at position 00 or overtravel in.

MODIFIED QUESTION: Bank question that was modified - 1223 A full reactor scram from rated conditions has been initiated. What will be the indication of a control rod that has moved past the full-in position 10 seconds after the scram (assume no operator action is taken)?

Four Rod Display Full In Light Drift Light Position Indication

a. ON ON Blank

b. ON OFF 99

c. OFF ON 00

d. OFF OFF Blank ANSWER: 31 1223

a. ON ON Blank

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 215002 A4.02 Importance Rating 2.9 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): Instrument Operating Procedure 4.1.5 (Rev 21)

(Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 32 New 00 Licensed RO: Y Operator SRO: Y Question NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 1 1 2 Multiple Choice Topic Area Description Systems COR002240, ROD BLOCK MONITOR (2002 NRC Exam)

Related Lessons COR0022402 OPS ROD BLOCK MONITOR Related Objectives COR0022402001040B Describe the RBM design features and/or interlocks that provide for the following: Initiation setpoint COR0022402001050B Describe each of the following concepts as they relate to the function or operation of the RBM: Null sequence control.

Related References 4.1.5 Rod Block Monitor System Related Skills (K/A) 215002 A4.02 Ability to manually operate and/or monitor in the control room: (CFR: 41.7 /

45.5 to 45.8) RBM back panel switches, meters and indicating lights: BWR-3,4,5 (2.9)

QUESTION: 32 Reactor power is 54% and control rod 34-27 is selected for withdrawal. The following occur:

Control rod 34-27 is notched from position 18 to position 22.

Control rod 26-37 is selected.

After control rod withdrawal and the subsequent selection of control rod 26-37, what is the response of Rod Block Monitor A as observed on RBM A meter at Panel 9-14?

AFTER 34-27 AFTER 26-37 MOVEMENT SELECTED

a. 56-58% 54%

b. 100% 54%

c. 100% 100%

d. 102-103% 100%

ANSWER: 32

d. 102-103% 100%

Explanation:

NOTE: See attached Procedure 4.1.5 partial.

Instrument Operating Procedure 4.1.5 (Rev 21) describes the RBM nulling:

Step 1.2.2.1 RBM System is designed to automatically amplify average of the input LPRMs to read the same as a reference source signal by increasing gain of the LPRM averaging amplifier in RBM System. This comparison and gain change is accomplished during null sequence that occurs immediately after a new rod is selected. Once gain is changed and RBM average nulls out with the constant reference, RBM is set and will remain fixed until another rod is selected.

Step 1.2.4.3 Magnitude of each RBM channel output is normalized to a constant reference value (100/125) whenever a control rod is selected. A signal from one APRM channel assigned to each Reactor Protection System trips the system supplied by this reference signal for the RBM channel on that same trip system. This gain setting is held constant during movement of that particular control rod; thus, providing indication of the change in relative local power level.

Distractors:

a. Reactor power is 54% but the RBM is normalized to 100% so the RBM indicates slightly above the normalized reference depending upon control rod worth.

b. Reactor power is 54% but the RBM is normalized to 100% so the RBM indicates slightly above the normalized reference depending upon control rod worth.

c. Reactor power is 54% but the RBM is normalized to 100% so the RBM indicates slightly above the normalized reference depending upon control rod worth.

Procedure 4.1.5 partial ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 219000 A3.01 Importance Rating 3.3 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): System Operating Procedure 2.2.69 (Rev 91)

(Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank # 1728 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 33 1728 02 05/16/2011 06/15/2005 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 2 1 2 Multiple Choice Topic Area Description Systems COR0022302001030A, COR0022302001050B, RHR SPC response to high drywell pressure signal.

Related Lessons COR0022302 RESIDUAL HEAT REMOVAL Related Objectives COR0022302001030A Describe RHR System design feature(s) and/or interlocks which provide for the following: Automatic initiation/injection COR0022302001050B Briefly describe the following concepts as they apply to the RHR system: Valve operation COR0022302001150D Given plant conditions, determine if the following should occur: RHR valve reposition Related References NONE Related Skills (K/A) 219000 A3.01 Ability to monitor automatic operations of the RHR/LPCI:

TORUS/SUPPRESSION POOL COOLING MODE including: (CFR: 41.7 /

45.7) Valve operation (3.3)

QUESTION: 33 1728 The plant is operating at 75% power with RHR Loop A operating in Suppression Pool Cooling (SPC). A reactor coolant leak develops in the Drywell resulting in the following conditions:

Drywell pressure 3.4 psig and rising slowly

Reactor pressure 700 psig and steady

Reactor water level +36 " (wide range) and steady What is the status of RHR-MO-39A, OUTBOARD SPC VALVE and RHR-MO-66A, HX BYPASS VALVE five (5) minutes later? (Assume NO operator actions taken with RHR Loop A controls).

RHR-MO-39A position RHR-MO-66A position

a. CLOSED OPEN

b. OPEN OPEN

c. CLOSED CLOSED

d. OPEN CLOSED ANSWER: 33 1728

a. CLOSED OPEN Explanation:

NOTE: See attached Procedure 2.2.69 partial.

System Operating Procedure 2.2.69 (Rev 91) Attachment 3 Sections 2.2.14, 2.2.16 and 2.2.19 list the valve interlocks.

In SPC mode, the outboard valve MO-39 is in a throttled position. The Heat exchanger bypass valve MO-66 is closed. On a high drywell pressure signal (>1.84 psig), MO-39 receives a closed signal and MO-66 receives an open signal for 2 minutes.

Distractors:

b. The outboard valve is open during SPC mode. The heat exchanger bypass valve does receive an open signal for 2 minutes on high drywell pressure.

c. The outboard valve receives a closed signal. The heat exchanger bypass valve is closed but receives a open signal for 2 minutes on high drywell pressure.

d. The outboard valve is open during SPC mode. The heat exchanger bypass valve is closed but receives a open signal for 2 minutes on high drywell pressure.

Procedure 2.2.69 partial ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 226001 K2.02 Importance Rating 2.9 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): Electrical Drawing B&R 3002 (Rev N47) Sheet 1 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 34 New 00 08/05/1999 06/15/2005 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 1 1 1 Multiple Choice Topic Area Description Systems COR0022302001020A Residual Heat Removal System Related Lessons COR0022302 RESIDUAL HEAT REMOVAL Related Objectives COR0022302001020A State the electrical power supplies to the following: RHR pump motors Related References NONE Related Skills (K/A) 226001 K2.02 Knowledge of electrical power supplies to the following: (CFR: 41.7) Pumps (2.9*/2.9*)

QUESTION: 34 RHR Pump D is operating and controlling drywell pressure between 2 psig and 10 psig. All systems operate per design. What is the power supply to RHR Pump D at the present time?

a. 4160V Bus 1F via 4160V Bus 1A.

b. 4160V Bus 1G via 4160V Bus 1B.

c. 4160V Bus 1F via the Emergency Station Transformer.

d. 4160V Bus 1G via the Emergency Station Transformer.

ANSWER: 34

b. 4160V Bus 1G via 4160V Bus 1B.

Explanation:

NOTE: See Attached Power supply one line drawing B&R 3002 (Rev N47) Sheet 1.

RHR Pump D is in Containment Cooling Loop B. The pump is powered from critical 4160V Bus 1G. The Startup Service Transformer (SSST) supplies critical buses (fast transfer) before the Emergency Transformer ties to the bus or the DGs come up to rated speed and voltage. The pump is powered form the SSST. Drywell pressure above 2 psig is an RPS scram so the Normal Station Transformer is not available.

Distractors:

a. This is the normal power supply lineup for RHR Pumps A and B with the Startup Transformer.

c. This is the power supply for RHR Pump D if the Normal or Startup Transformer are not available.

d This is the power available for RHR Pumps A and B if the Normal or Startup Transformer are not available.

Emergency Transformer DG 2 Bus 1B Supply Supply Supply

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 239001 K6.09 Importance Rating 3.9 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): GE Drawings 791E266 (Rev N14 for sheet 10 (Attach if not previously provided) and Rev N03 for sheet 14)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank # 93 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 35 93 0 10/08/1997 06/15/2005 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 1 1 1 Multiple Choice Topic Area Description Systems COR0021402 - Main Steam Related Lessons COR0021402 OPS MAIN STEAM Related Objectives COR0021402001040B Describe the Main Steam system design features and/or interlocks that provide for the following: Automatic isolation and opening of drain valves COR0021402001070J Predict the consequences a malfunction of the following would have on the Main Steam system: Loss of electrical power Related References NONE Related Skills (K/A) 239001 K6.09 Knowledge of the effect that a loss or malfunction of the following will have on the MAIN AND REHEAT STEAM SYSTEM: (CFR: 41.7 / 45.7)

PCIS/NSSSS (3.9)

QUESTION: 35 93 A plant startup is in progress with the Reactor Mode Switch in RUN. MCC-L de-energizes.

What effect does this have on the Main Steam system?

a. MO-74 (MSL Drain Inboard Isolation) closes.

b. MO-77 (MSL Drain Outboard Isolation) closes.

c. Inboard MSIVs and MO-74 (MSL Drain Inboard Isolation) closes.

d. Outboard MSIVs and MO-77 (MSL Drain Outboard Isolation) closes.

ANSWER: 35 93

a. MO-74 (MSL Drain Inboard Isolation) closes.

Explanation:

NOTE: See attached GE Drawing partials.

GE Drawings 791E266 (Rev N14 for sheet 10 and Rev N03 for sheet 14) show solenoids for MSIVs and logic for MO-74).

Two solenoid valves (one AC and one DC) control the air supply to the pilot valve actuator. If either solenoid valve is energized, the MSIV will remain open. MCC-L provides power to the RPS 1A MG set which provides power to the inboard MSIV AC solenoid valve so the AC solenoid de-energizes. The DC solenoid remains energized so there is no MSIV valve movement. The logic for MO-74 relay 16A-K56 is powered from RPS 1A so it de-energizes and provides a close signal to MO-74.

Distractors:

b. Valve closes if Div 2 de-energizes.

c. MSIVs close if both AC and DC solenoids de-energize.

d. MSIVs close if both AC and DC solenoids de-energize. MO-77 closes if Div 2 de-energizes.

Pilot moves down and air Opens MSIV X

AC solenoid de-energizes but DC solenoid remains energized and blocks venting top of pilot

MSIV dual solenoids Closed circuit of MO-74. With 16A-K56 de-energized its contact closes and valve closes.

NOTE: Contact to 16A-K56 shown with relay de-energized.

Close coil

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 241000 K3.10 Importance Rating 2.9 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): GOP 2.1.5 (Rev 67)(Rev 24)

(Attach if not previously provided) B&R Drawing 3039 (Rev N23) Sheet 2A (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank # 23729 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 36 23729 00 09/12/2007 06/15/2005 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 1 1 4 Multiple Choice Topic Area Description Systems Operation Loss of 125 VDC to the M. Turbine trip circuit Simulator Guides Related Lessons COR0011402 OPS MAIN TURBINE COR0020702 OPS DC ELECTRICAL DISTRIBUTION Related Objectives COR0011402001030E Describe the Main Turbine and Auxiliaries design features and/or interlocks that provide for the following: Turbine Protection COR0020702001060L Describe the interrelationship between the DC Electrical Distribution System and the following: Main Turbine Generators and Auxiliary Systems Related References NONE Related Skills (K/A) 241000 K3.10 Knowledge of the effect that a loss or malfunction of the REACTOR/TURBINE PRESSURE REGULATING SYSTEM will have on following: (CFR: 41.7 / 45.4) Front standard trip system (2.9)

QUESTION: 36 23729 What effect does de-energizing 125 VDC power panel BB-2 have on the main turbine trip logic?

The main turbine . . .

a. CAN be manually tripped from the Control Room AND can be tripped locally. The main turbine automatic electric trips WILL function.

b. CAN be manually tripped from the Control Room AND can be tripped locally. The main turbine automatic electric trips will NOT function.

c. CANNOT be tripped from the Control Room NOR locally. The main turbine automatic electric trips WILL function.

d. CANNOT be manually tripped from the Control Room, but CAN be tripped locally. The main turbine automatic electric trips will NOT function.

ANSWER: 36 23729

d. CANNOT be manually tripped from the Control Room, but CAN be tripped locally. The main turbine automatic electric trips will NOT function.

Explanation:

NOTE: See attached B&R Electrical Drawing (Rev N23) 3049 Sheet 2A.

General Operating Procedure 2.1.5 (Rev 67) Step 1.2, provides guidance for tripping the main turbine locally if turbine if tripping from the control room fails.

The turbine trip logic is powered from 125VDC BB-2. The relays are required to energize to cause a turbine trip by opening solenoid valves depressurizing the emergency trip header allowing spring pressure to close the stop valves and stop steam admission to the main turbine.

Tripping the turbine from the main control room will not work. Electrical trips (electrical faults, high RPV water level, etc) will not function due to logic power loss. The only two means of tripping the main turbine are local pushbuttons in the trip tricon cabinet and opening two manual valves at the front standard.

a. The diversity of trip solenoids AST-2 and AST-3 still have power but their trip inputs (automatic electric trips) are lost. The control room pushbuttons are the normal manual turbine trip method.

b. The diversity of trip solenoids AST-2 and AST-3 still have power but their trip inputs (automatic electric trips) are lost. The control room pushbuttons are the normal manual turbine trip method.

c. The local trip pushbuttons are the normal alternate trip method and will work. The diversity of trip solenoids AST-2 and AST-3 still have power but their trip inputs (automatic electric trips) are lost.

B&R Drawing 3049 Powered from BB-2 20AST-1 Turbine Trip relays 20 ET 20 AST-2 and AST-3 powered from AC source but their trip inputs are de-energized as shown above.

20AST-2 20AST-3

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 245000 A2.05 Importance Rating 3.6 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): Alarm Procedure 2.3_B-2 (Rev 30)

(Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR Part 55 Content: 55.41 4 55.43 Comments:

Question is more appropriate for a 55.41.4 vice 55.41.5.

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 37 New 00 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 4 1 1 4 Multiple Choice Topic Area Description Systems COR0022302001030O Residual Heat Removal System Related Lessons COR0011402 OPS Main Generator Related Objectives COR0011402001160E Given the below conditions/responses, state the reason for that response: Load Reject Related References NONE Related Skills (K/A) 245000 A2.05 Ability to (a) predict the impacts of the following on the MAIN TURBINE GENERATOR AND AUXILIARY SYSTEMS ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5 / 45.6) Generator trip (3.6)

QUESTION: 37 The plant was operating at rated conditions when the main generator tripped on an NSST fault.

Ten minutes later the following alarm is received:

B-2/C-1, TG EXHAUST HOOD A TEMP HIGH A check of TGI-R-RECCST, CONTROLLED START TEMP, CH 9 - 1ST LP TURBINE EXHAUST, reveals the temperature is 195°F and rising slowly.

What operator action is required per alarm B-2/C-1?

a. If turbine speed is between 1380 and 1710 rpm, on HOOD SPRAY screen, TG EXHAUST HOOD A control, verify hood spray operation by verifying DEMAND is > 0%.

b. Before temperature approaches 225°F, slowly throttle open MC-MO-BMV6A, HOOD SPRAY A BYPASS VLV regardless of turbine speed.

c. If turbine is on turning gear, on HOOD SPRAY screen, TG EXHAUST HOOD A control, verify hood spray operation by verifying DEMAND is > 0%.

d. Lower condenser backpressure by opening AR-MO-150, VACUUM BREAKER.

ANSWER: 37

c. If turbine is on turning gear, on HOOD SPRAY screen, TG EXHAUST HOOD A control, verify hood spray operation by verifying DEMAND is > 0%.

Explanation:

NOTE: See attached Procedure 2.3_B-2 (alarm B-2/C-1) partial Alarm Procedure 2.3_B-2 (Rev 30), specifically alarm B-2/C-1, Step 1.2.1, provides guidance for high exhaust temperatures.

Hood spray can only be initiated on an unloaded turbine if either, the turbine is on the turning gear, or turbine speed is > 95% or 1710 rpm.

Distractors:

a. The 1380 rpm is related to turbine speed for warming up the unit. The 1710 rpm is the lower limit for initiating hood sprays with the turbine unloaded.

b. The 225°F is the High-High exhaust temperature alarm but the corrective guidance is the same. Opening BMV-6A bypass valve is one means of initiating hood spray but the spray limitations are no different.

d. Procedure guidance is to raise condenser vacuum not lower vacuum.

Procedure 2.3_B-2 (alarm B-2/C-1) partial ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 290001 K6.08 Importance Rating 2.7 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): USAR Section V-3.3.3 (Attach if not previously provided) B&R P&ID 2010 (Rev N 95) Sheet 2 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank # 21750 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 38 21750 00 08/09/2005 06/15/2005 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 2 1 4 Multiple Choice Topic Area Description Systems COR0020302, How does a loss of Reliable Inst Air affect RB vent isolation?

Related Lessons COR0020302 OPS CONTAINMENT Related Objectives COR0020302001070A Describe the interrelationship between Secondary Containment and the following: Reactor building ventilation Related References NONE Related Skills (K/A) 290001 K6.08 Knowledge of the effect that a loss or malfunction of the following will have on the SECONDARY CONTAINMENT : (CFR: 41.7 / 45.7) Plant air systems (2.7)

QUESTION: 38 21750 How does a loss of Plant Air affect the operation of Secondary containment Isolation valves HV-259AV and HV-261AV, Reactor Building Vent Exhaust Inboard Isolation valves?

A loss of plant air

a. requires both valves to be closed manually if an auto close signal is received.

b. prevents the closing of HV-259AV with its control switch.

c. prevents the closing of HV-261AV from an automatic isolation signal.

d. requires accumulator air as a motive force for closing both valves.

ANSWER: 38 21750

d. requires accumulator air as a motive force for closing both valves.

Explanation:

NOTE: See attached partial P&ID Drawing of air supply and accumulators to valves.

ESAR Secondary Containment Isolation Systems Accumulators are provided for the air operated Secondary Containment Isolation Valves to ensure the isolation valves can be closed after a loss of the instrument air supply (see USAR Section V-3.3.3).

Distractors:

a. The valves solenoid fails the valve as is on a loss of air. However the accumulator provides enough motive force to close the valve. The valves can be closed with their control switches or they will close automatically.

b. The valves solenoid fails the valve as is on a loss of air. However the accumulator provides enough motive force to close the valve if the control switch is used.

c. The valves solenoid fails the valve as is on a loss of air. However the accumulator provides enough motive force to close the valve in an automatic closure signal is received.

HV-259 HV-261 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 295001 AA1.07 Importance Rating 3.1 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): Abnormal Procedure 2.4RR (Rev 38), Attachment 1 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank # 1063 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR Part 55 Content: 55.41 2 55.43 Comments:

Question more appropriately addresses 55.41.2 vice 55.41.7.

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 39 1063 00 06/22/1999 06/15/2005 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 2 1 4 Multiple Choice Topic Area Description Systems COR0022202, REACTOR RECIRCULATION Related Lessons COR0021502 OPS NUCLEAR BOILER INSTRUMENTATION COR0022202 REACTOR RECIRCULATION SKL0124222 OPS REACTOR RECIRCULATION SYSTEM Related Objectives COR0021502001060K Given a specific NBI malfunction, determine effect on any of the following: Core flow/Jet Pump monitoring COR0022202001060B Given a specific Reactor Recirculation system or the Recirculation Flow Control system malfunction, determine the effect on any of the following: Core Flow (normal and reduced forced flow conditions)

SKL012422200A030E Given plant conditions, predict changes in the following Reactor Recirculation System components/parameters: Core flow Related References 2.4RR Reactor Recirculation Abnormal Related Skills (K/A) 295001 AA1.07 Ability to operate and/or monitor the following as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION : (CFR: 41.7

/ 45.6) Nuclear boiler instrumentation system (3.1)

QUESTION: 39 1063 Given the following conditions:

Recirculation Pump B has tripped.

RR-MO-53B, Recirculation Pump B discharge valve was closed and is now open.

LOOP B JET PUMP FLOW (FI-92B) indicates 2 Mlbm/hr.

LOOP A JET PUMP FLOW (FI-92A) indicates 35 Mlbm/hr.

Annunciator 9-4-3/E-7, RECIRC LOOP B OUT OF SERVICE is NOT alarming.

What is the expected value for indicated Total Core Flow as indicated on Panel 9-5 Recorder DPR/FR-95 AND what is Actual Core Flow?

Indicated total Core Flow Actual core flow

a. 33 Mlbm/hr 33 Mlbm/hr

b. 33 Mlbm/hr 37 Mlbm/hr

c. 37 Mlbm/hr 33 Mlbm/hr

d. 37 Mlbm/hr 37 Mlbm/hr ANSWER: 39 1063

c. 37 Mlbm/hr 33 Mlbm/hr Explanation:

NOTE: See attached Procedure 2.4RR, Attachment 1 partial.

Abnormal Procedure 2.4RR (Rev 38), Attachment 1 (for tripped RR pump) contains a note that describes the correct indications for a properly operating reverse summer.

With one Recirc Pump out of service, a reverse flow will exist through the idle Jet Pumps but the Jet Pump Flow instrumentation will indicate a positive flow. Annunciator 9-4-3/E-7 not in alarm indicates Core Flow circuitry is not functioning properly for single loop (i.e. The Loop Jet Pump flows are being added vice subtracted). Total Core Flow will indicate 37 on DPR/FR-95. Since reverse flow exists in the idle loop, Actual core flow will be the difference between Loop A & B Jet Pump flows. With the Core Flow summing circuit malfunctioning, indicated Total Core Flow will be the sum of Loop A & B Jet Pump flows (37) and Actual Core Flow will be the difference between Loop A & B Jet Pump Flows (33).

Distractors:

a. These are the values obtained if the summer was working.

b. This are the value obtained for indicated core flow if the summer was working. This is the value obtained if the out of service loop flow is added vice subtracted.

d. These are the values if the loop flows are added.

Procedure 2.4RR, Attachment 1 partial ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 295003 G2.2.44 Importance Rating 4.2 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): B&R Electrical 3021 (Rev N08)

(Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR Part 55 Content: 55.41 5 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 40 New 00 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 2 1 4 Multiple Choice Topic Area Description Systems COR0010102 CP restart if C/S not repositioned after loss of AC power.

Related Lessons COR0010102 OPS AC Distribution Related Objectives COR0010102001060B Describe the interrelationship between the AC Electrical Distribution System and the following: Off-site Power Sources COR0010102001130F Predict the consequences of the following events on the AC Electrical Distribution System: De-energizing a plant bus COR0010102001130G Predict the consequences of the following events on the AC Electrical Distribution System: Energizing a dead bus Related References Related Skills (K/A) 295003 Partial or Complete Loss of AC G2.2.44 Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions. (CFR: 41.5 / 43.5 / 45.12) (4.2)

QUESTION: 40 The plant was operating at full power with the Startup Station Transformer (SSST) de-energized for switchyard work. The plant experiences a reactor scram and the electrical systems operate per design. The crew enters Procedure 5.3EMPWR, EMERGENCY POWER DURING MODES 1, 2, OR 3. The crew misses performing step addressing the Main Condensate Pumps and they are in the configuration shown below.

The SSST is repowered. With no manipulation of the above switches, what is the status of the Main Condensate Pumps after the non-critical 4160V buses are re-energized?

a. There is no change to the pumps status as they remained operating throughout the transient.

b. The pump breakers all trip once their respective buses are repowered.

c. The pumps all start with the potential of water hammer damage in the system.

d. The pumps all start with the pumps running deadheaded.

ANSWER: 40

c. The pumps all start with the potential of water hammer damage in the system.

Explanation:

NOTE: See attached B&R Electrical partial drawing 3021 (Rev N08) Sheet 5.

The Main Condensate Pump breakers do not have undervoltage trips so the breakers remain closed on a loss of power to the bus but the pumps are not running because of the loss of power. Once power to the SSST is restored, the 4160V buses are re-energized. With the 4160V buses repowered, the pumps immediately start. The discharge valves on these pumps are manual valves that remain open.

Distractors:

a. The breaker indication (red light) indicates the breaker is closed.

b. All 3 pumps starting put a surge on the 4160V system so it is reasonable a breaker trip could occur.

d. Some pumps have discharge valves that close on pump trips.

There is no trip present with the 4160V bus de-energized. The breaker will trip on Short Circuit (50-1), Locked Rotor (50-2), AC overcurrent with time delay (51) or ground 50-G. None of these conditions are present when the bus is de-energized or re-energized. The control switches have not been touched and remain closed after start (red flagged). The red light contains a drop down resistor so the trickle current goes through the TRIP block but does not have enough current to trip the breaker. Therefore the breaker stays closed and the red closed indicating light remains lit as this logic is powered from DC power.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 295004 AA2.01 Importance Rating 3.2 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): B&R Electrical Drawing 3058 (Rev N56)

(Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 41 New 00 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 2 1 4 Multiple Choice Topic Area Description Systems COR0020702 reason for loss of DC Related Lessons COR0020702 OPS DC DISTRIBUTION Related Objectives COR0020702001060J Describe the interrelationship between the DC Electrical Distribution System and the following: RCIC COR0020702001080C Given a specific DC Electrical Distribution system malfunction, determine the effect on any of the following: Systems with DC components (i.e., valves, motors, solenoids, etc.)

COR0020702001080N Given a specific DC Electrical Distribution system malfunction, determine the effect on any of the following: RCIC Related References NONE Related Skills (K/A) 295004 AA2.01 Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF D.C. POWER: (CFR: 41.10 / 43.5 / 45.13)

Cause of partial or complete loss of D.C. power (3.2)

QUESTION: 41 The plant is operating in Mode 1 in a normal system alignment when the control room operator reports that several RCIC MOVs have lost their indicating lights. What could be the cause of this condition?

a. 125VDC SWGR 1A Bus has a blown fuse which supplies the RCIC Starter Rack.

b. 125VDC SWGR 1B Bus has a blown fuse which supplies the RCIC Starter Rack.

c. 125V Charger 1A sustained a DC output over voltage.

d. 125V Charger 1B sustained a DC output over voltage.

ANSWER: 41

a. 125VDC SWGR 1A Bus has a blown fuse which supplies the RCIC Starter Rack.

Explanation:

NOTE: See B&R Electrical Drawing 3058 (Rev N56) partial attached.

The RCIC starter rack is normally powered from 125VDC Switchgear 1A through a fused disconnect. If the fuse blows, the starter rack is without power. The starter rack can have its power supply transferred to an emergency source (125VDC SGR 1B Bus). The power supplies to 125VDC SWGR 1A Bus are from the 125V Charger 1A and the 125VDC Battery 1A. The charger is the normal supply with the battery as a backup. Should the charger trip, the battery will carry the bus.

Distractors:

b. The 1B bus is the emergency supply to the RCIC starter rack and a blown fuse would not de-energize the RCIC starter rack unless the starter rack were on its emergency source.

c. A DC output over voltage trips the charger AC input breaker causing a loss of the charger as a power supply. The 125VDC Battery 1A would then power the bus.

d. A DC output over voltage trips the charger AC input breaker causing a loss of the charger as a power supply. The 125VDC Battery 1B would then power the bus. The 1B bus is the emergency supply to the RCIC starter rack.

Charger 1A Battery 1A Normal Supply Emerg Supply

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 295005 AA2.07 Importance Rating 3.5 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): USAR Section XIV Figure XIV-5-3 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 42 New 01 09/25/2003 06/15/2005 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 1 1 4 Multiple Choice Topic Area Description Systems COR0011402, RPV level response to a turbine trip Related Lessons COR0011402 OPS MAIN TURBINE Related Objectives COR0011402001040H Describe the interrelationships that exist between the Main Turbine and Auxiliaries and the following: Reactor water level control Related References NONE Related Skills (K/A) 295005 AA2.07 Ability to determine and/or interpret the following as they apply to MAIN TURBINE GENERATOR TRIP : (CFR: 41.10 / 43.5 / 45.13) Reactor water level (3.5)

QUESTION: 42 The plant is operating at 100% power when the main turbine trips.

What is the immediate reactor water level response to this condition?

a. RPV level rapidly lowers greater than 20 inches.

b. RPV level rapidly rises greater than 10 inches.

c. RPV level slowly lowers to 15 inches.

d. RPV level slowly rises to 54.5 inches.

ANSWER: 42

a. RPV level rapidly lowers greater than 20 inches.

Explanation:

USAR Section XIV Figure XIV-5-3 depicts RPV level response on a turbine trip with bypass valves.

Due to the rapid collapsing of voids in the core region there is less resistance to flow. Therefore more water will leave the downcomer region and move inside the shroud so the immediate affect is a rapid drop in level of over 35 on the Reactor water level instruments.

Distractors:

b. RPV level recovers > 10 inches after its initial shrink.

c. RVLC Setpoint Setdown is 15 inches.

d. The high RPV level turbine trip is 54.5 inches.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 295006 AA1.05 Importance Rating 4.2 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): Instrument Operating Procedures 4.1.1 (Rev 21) and (Attach if not previously provided) 4.1.2 (Rev 21)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 43 New 00 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 3 1 4 Multiple Choice Topic Area Description Systems SRM/IRM operation post-scram Related Lessons COR0021202, OPS IRMS COR0023002, OPS SRMS Related Objectives COR0021202001050E Describe the IRM system design features and/or interlocks that provide the following: Changing detector position COR0023002001060D Describe the SRM system design features and/or interlocks that provide for the following: Changing detector position Related References Instrument Operating Procedure 4.1.1 SRMs Instrument Operating Procedure 4.1.2 IRMs Related Skills (K/A) 259006 AA1.05 Ability to operate and/or monitor the following as they apply to SCRAM :

(CFR: 41.7 / 45.6) Neutron monitoring system (4.2)

QUESTION: 43 A reactor scram has just occurred. Procedure 2.1.5, Reactor Scram directs operating the instruments represented by the picture below. What sequence of operator actions are required per Procedures 4.1.1, SRMs and 4.1.2, IRMS?

a. Press POWER ON switch.

Press SRM A through D SELECT switches.

Press IRM A through H SELECT switches.

Press and hold DRIVE IN switch until IN lights all illuminated, then release.

b. Press SRM A through D SELECT switches.

Press IRM A through H SELECT switches.

Press and hold DRIVE IN switch until IN lights all illuminated, then release.

c. Press POWER ON switch.

Press SRM A through D SELECT switches.

Press IRM A through H SELECT switches.

Momentarily press DRIVE IN switch.

d. Press SRM A through D SELECT switches.

Press IRM A through H SELECT switches.

Press POWER ON switch.

Momentarily press DRIVE IN switch.

ANSWER: 43

c. Press POWER ON switch.

Press SRM A through D SELECT switches.

Press IRM A through H SELECT switches.

Momentarily press DRIVE IN switch.

Explanation:

Procedures 4.1.1 (Rev21) SRM System and 4.1.2 (Rev 21) IRM System provide direction for driving in the respective detectors.

The IRMs/SRMs can be selected without pressing the POWER ON switch. However, the POWER ON switch must be depressed before the DRIVE IN switch will insert the detectors.

Both procedures have the following guidance:

1.1 If required, momentarily press and release SRM/IRM DETECTOR POS display (POWER ON) button and check DETECTOR POSITION light turns on.

NOTE - System is designed to allow all detectors to be moved simultaneously in same direction.

1.2 Select detector(s) by momentarily pressing applicable SELECT switch.

1.3 Check applicable SELECT switch(s) light turns on.

1.4 Momentarily press DRIVE IN switch.

Distractors:

a. The DRIVE OUT switch must be continuously depressed when driving the respective detector.

b. The individual detectors can be selected but the POWER ON switch must be used prior to moving the detectors.

d. This will drive in the detectors but it is not in accordance with the procedures.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 295016 AK3.02 Importance Rating 3.7 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): Emergency Procedure 5.1ASD (Rev 15)

(Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR Part 55 Content: 55.41 5 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 44 New 00 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 1 1 4 Multiple Choice Topic Area Description Systems COR0023402 Reason for turbine trip Related Lessons COR0023402 OPS Alternate Shutdown Related Objectives INT0320136O0O0100 Given plant condition(s), determine from memory if a Main Turbine trip is required due to the event(s).

Related References NONE Related Skills (K/A) 295016 AK3.02 Knowledge of the reasons for the following responses as they apply to CONTROL ROOM ABANDONMENT : (CFR: 41.5 / 45.6) Turbine trip (3.7)

QUESTION: 44 What is the reason for manually tripping the main turbine during a control room abandonment event?

a. To cause the main turbine bypass valves to open and control RPV pressure throughout the event.

b. To take positive action to trip the turbine without relying on automatic trips.

c. To initiate a reactor scram and place the reactor in a low energy state.

d. To preclude RPV inventory overfill resulting in RFPT trips.

ANSWER: 44

b. To take positive action to trip the turbine without relying on automatic trips.

Explanation:

NOTE: See attached Procedure 5.1ASD partial Emergency Procedure 5.1ASD (Rev 15) Step 4.2 and sub-steps requires scramming the reactor and tripping the main turbine. These are positive actions as manually tripping the main turbine precludes relying on anti-motoring relaying to cause the trip.

Distractors:

a. One action is to leave the reactor mode switch in RUN which causes MSIV closure so pressure control with the bypass valves is not a priority. SRVs are used to control pressure.

c. The reactor is manually scrammed first so tripping the turbine is not the reason for the scram.

d. RVLC system precludes overfill events and leaving the reactor mode switch in RUN causes MSIV closure and motive force for the RFPTs.

Procedure 5.1ASD partial

4. SUBSEQUENT OPERATOR ACTIONS 4.1 SM or CRS shall perform following:

4.1.1 Ensure event announced over Gaitronics.

4.1.2 lf toxic gas is reason for evacuation, direct all personnel to stay clear of Cable Spreading Room.

4.1.3 Direct Operations personnel to assemble in corridor outside Control Room blast doors at Alternate Shutdown Locker.

4.1.4 Obtain lN-2 Key (Key 1 12 in CR SM Key Locker) for Alternate Shutdown Locker and bring CR SM Key Locker to ASD Locker.

4.1.5 Direct SAS Operator to evacuate SAS and bring key box to Alternate Shutdown Locker.

4.2 Before leaving Control Room, if time permits, perform following:

4.2.1 SCRAM.

4.2.2 Check all control rods in.

4.2.3 Ensure REACTOR MODE switch is in RUN.

4.2.4 Trip following:

4.2.4.1 Main turbine generator.

4.2.4.2 All but one feed pump.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 295018 AA1.03 Importance Rating 3.3 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): Emergency Procedure 5.2REC (Rev 13)

(Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # 10641 (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 45 10641 01 07/07/2004 06/15/2005 NRC Style RO: Y Modified Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 3 1 5 Multiple Choice Topic Area Description Abnormal/Emergency INT0320126, Initiate SW backup to REC Procedures Related Lessons INT0320126 CNS Abnormal Procedures (RO) Cooling Water Related Objectives INT0320126Q0Q0100 Given plant condition(s), and the applicable Abnormal/Emergency Procedure, determine the correct subsequent actions required to mitigate the event(s).

Related References 5.2REC Loss Of REC Related Skills (K/A) 295018 AA1.03 Ability to operate and/or monitor the following as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER: (CFR: 41.7 /

45.6) Affected systems so as to isolate damaged portions (3.3)

QUESTION: 45 10641 Following a LOCA, HPCI is maintaining RPV level and RHR loop B is in suppression pool cooling. The following annunciators now alarm:

M-1/A-1, REC SYSTEM LOW PRESSURE

M-1/A-3, REC SURGE TANK LOW LEVEL The crew performed the following per Procedure 5.2REC, Loss of REC:

Stopped all the operating REC pumps.

Closed REC-MO-709, DRYWELL RETURN ISOLATION.

Inspected the critical loops and found them intact.

Restarted REC Pump 1A.

The following plant and REC system conditions are now present:

REC critical loop supply pressures are 60 psig and oscillating.

Surge tank level is low out of sight and the level control valve is open.

What REC action is required per 5.2REC?

a. Stop REC pump 1A.

b. Split the critical loops.

c. Start two additional REC pumps.

d. Supply SW backup to REC critical loops.

ANSWER: 45 10641

d. Supply SW backup to REC critical loops.

Explanation:

NOTE: See attached Procedure 5.2REC, Attachment 2 The REC surge tank remains low and has not recovered following the automatic isolation of the REC loads. The REC pumps will lose adequate NPSH. The only loads remaining on the system are the critical loops. The critical loops are currently required by the operation of HPCI and RHR. These conditions require that service water backup be initiated to cool the critical loops.

Distractors:

a. This is a good action to take but the procedure is directing service water backup alignment.

b. The loops are split only if one critical loop is intact.

c. There is no advantage to starting additional REC pumps and there is no procedure guidance to do so in this condition.

Procedure 5.2REC, Attachment 2 ATTACHMENT2 REC PIPE BREAK REC PIPE BREAK 1.1 Stop all REC pumps.

1.2 Close REC-MO-709, DRYWELL RETURN ISOLATION.

1.3 Ensure following valves close:

1.3.1 REC-MO-700.

1.3.2 REC-MO-702.

1 .3.3 REC-MO-712.

1.3.4 REC-MO-713.

1.3.5 REC-MO-1329.

1.4 lf Reactor Building accessible, inspect critical loop, including piping between pumps and critical loop supply and return valves.

1.5 lf both critical loops are intact:

1.5.1 Restart one REC pump.

1.5.2 Open one of following:

1.5.2.1 REC-MO-711, NORTH CRITICAL LOOP SUPPLY.

1.5.2.2 REC-MO-714, SOUTH CRITICAL LOOP SUPPLY.

1.5.3 lf REC surge tank not providing adequate NPSH to REC pumps, initiate service water backup per Attachment 6 (Page 12).

MODIFIED QUESTION 10641 Following a small loss of coolant accident HPCI is in operation maintaining normal RPV level.

RHR loop B is in suppression pool cooling. The following annunciators now alarm:

M-1/A-1, REC SYSTEM LOW PRESSURE

M-1/A-3, REC SURGE TANK LOW LEVEL The crew stopped all the operating REC pumps, closed REC-MO-709, DRYWELL RETURN ISOLATION, inspected the critical loops and then restarted REC pump 1A. The following plant and REC system conditions are present:

REC critical loop supply pressures are 60 psig and steady.

Surge tank level is low out of sight and the level control valve is open.

What REC action is required now?

a. Stop REC pump 1A.

b. Split REC critical loops.

c. Start two additional REC pumps.

d. Supply SW backup to REC critical loops.

ANSWER: 10641

b. is correct. The REC surge tank remains low and has not recovered following the automatic isolation of the REC loads. The only loads remaining on the system are the critical loops. The critical loops are currently required by the operation of HPCI and RHR. These conditions require that the REC loops be split.

a. is incorrect. Because even though pressure is low, no pressure fluctuations are present (no indication of cavitation) and the pump has a flowpath.

c. is incorrect. There is no advantage to starting additional REC pumps and there is no procedure guidance to do so in this condition.

d. is incorrect. Because REC pump 1A is supplying the critical loops successfully.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 295019 AK2.06 Importance Rating 2.8 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): Emergency Procedure 5.2AIR (Rev 18)

(Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 46 New 00 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 2 1 4 Multiple Choice Topic Area Description Systems INT0320136 Loss of instrument air effects on offgas Related Lessons INT0320136 OPS-SNS Abnormal Procedures (RO) Miscellaneous Related Objectives INT0320136P0P0100 Given plant condition(s), and the applicable Abnormal/Emergency Procedure, determine the correct subsequent actions required to mitigate the event(s).

Related References 5.1AIR Related Skills (K/A) 295019 AK2.06 Knowledge of the interrelations between PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR and the following: (CFR: 41.7 / 45.8) Offgas System (2.8)

QUESTION: 46 The plant is operating at power when a partial loss of instrument air occurs. The crew enters Emergency Procedure 5.2AIR, Loss of Instrument Air. The procedure directs the crew to enter Procedure 2.1.5, SCRAM, and close the MSIVs and drains. The procedure also directs the crew to ensure the preferred Standby Gas Treatment system is operating.

Why is the crew directed to ensure the preferred Standby Gas Treatment system is operating?

a. To provide dilution flow to the Elevated Release Point due to the Off-Gas dilution fan running deadheaded.

b. To ensure proper Reactor Building Quad room cooling due to room coolers temperature control valves failing closed.

c. To support Reactor Building Kaman operation due to the loss of building ventilation.

d. To aid in HPCI operation.

ANSWER: 46

a. To ensure proper dilution flow out of the Elevated Release Point.

Explanation:

Procedure 5.2AIR (Rev 18) directs ensuring the preferred SGT train operating and the Off-Gas dilution fans secured. The Off-Gas dilution fans are dead headed due to their AOV discharge valve failing closed. SGT is operated to support dilution flow.

The two steps in 5.2AIR are:

1.23 Ensure preferred SGT Subsystem running with SGT-AO-546A(B) open (VBD-K).

1.24 Stop off-gas dilution fan.

Distractors:

b. Quad room cooling is accomplished through their own FCUs. SGT does cause air flow throughout the Reactor Building but Procedure 5.2AIR is concerned with dilution flow.

c. Procedure 5.2AIR directs shutting down the Kaman units due to loss of ventilation flow.

d. The procedure directs SGT operation for dilution flow. With the MSIVs closed, HPCI will be in operation but Procedure 5.2AIR is concerned with dilution flow.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 295021 AA1.01 Importance Rating 3.4 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): Abnormal Procedure 2.4SDC (Rev 12)

(Attach if not previously provided) B&R 2042 (Rev N34) Sheet 1 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank # 24838 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 47 24838 00 07/18/2009 06/15/2005 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 3 1 3 Multiple Choice Topic Area Description Integrated Plant RWCU alternate decay heat removal L/O SDC Related Lessons COR0012002 OPS Reactor Water Cleanup COR0021902 REACTOR EQUIPMENT COOLING Related Objectives COR0012002001090B Describe the RWCU design features and/or interlocks that provide for the following: Maximizing plant efficiency (use of regenerative heat exchanger)

COR0012002001100C Briefly describe the following concepts as they apply to RWCU:

Heat exchanger operation, including relationship between the non-regenerative and regenerative heat exchangers.

Related References ACP 2.4SDC Related Skills (K/A) 295021 AA1.01 Ability to operate and/or monitor the following as they apply to LOSS OF SHUTDOWN COOLING : (CFR: 41.7 / 45.6) Reactor water cleanup system (3.4)

QUESTION: 47 24838 The Plant is in Mode 4 with RCS temperature at 175°F

RHR Pump B is in service in SDC.

RHR Loop A is out of service for maintenance.

Reactor Recirculation Pumps A and B are out of service for maintenance.

RWCU is in service with 1 pump and 2 filters.

RHR Pump B trips due to a motor failure.

What effect does this have on the RWCU system?

a. RWCU can now provide more core forced circulation inside the shroud.

b. The additional cooling in the downcomer precludes flashing in the RWCU suction.

c. The reactor heat removal rate in the regenerative heat exchanger becomes greater.

d. The reactor heat removal rate in the non-regenerative heat exchanger becomes greater.

ANSWER: 47 24838

d. The reactor heat removal rate in the non-regenerative heat exchanger becomes greater.

Explanation:

Note: See drawing B&R 2042 (Rev N34) Sheet 1 attached for Non-regen Hx cooling.

Abnormal Procedure 2.4SDC (Rev 12) Attachment 2, Step 1.5 directs placing RWCU in service for alternate heat removal. The Regen Hxs are bypassed and isolated. The Non-Regen HXs utilize REC cooling to remove decay heat. The Regen Hxs are not a heat sink and it is a common misconception they can be used to cool down the RCS. RWCU suction piping flashing is a common BWR issue. The rise in RCS temp or RWCU flow for a given temperature can result in suction piping flashing.

Distractors:

a. RWCU does not force circulation through the core. The flowpath is from the bottom head to the downcomer.

b. The loss of SDC adds a heat load to the RWCU system and downcomer temperatures rise.

c. The Regen-Hxs are bypassed and isolated.

B&R P&ID Drawing 2042 From Reactor To Reactor REC Cooling This path Isolated.

To/From Filter Demineralizers

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 295023 AK2.07 Importance Rating 3.6 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): USAR XIV Section 6.4 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 48 New 00 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 1 1 4 Multiple Choice Topic Area Description Systems SGT response to a refueling accident.

Related Lessons COR002 OPS Standby Gas Treatment System SKL OPS Transient and Accident Analysis Related Objectives Related References NONE Related Skills (K/A) 295023 AK2.07 Knowledge of the interrelations between REFUELING ACCIDENTS and the following: (CFR: 41.7 / 45.8) Standby gas treatment/FRVS (3.6)

QUESTION: 48 A refueling accident has occurred. What is the Standby Gas Treatment system relationship with this event?

a. The trains start when reactor building exhaust radiation levels reach predetermined levels and any subsequent radiation release is through the elevated release point.

b. The trains start when reactor building exhaust radiation levels reach predetermined levels and any subsequent radiation release is through the reactor building exhaust.

c. The trains remain idle and any radiation release is through the elevated release point.

d. The trains remain idle and any radiation release is through the reactor building exhaust.

ANSWER: 48

a. The trains start when reactor building exhaust radiation levels reach predetermined levels and any subsequent radiation release is through the elevated release point..

Explanation:

NOTE: See attached USAR Chapter XIV, Section 6.4 USAR XIV Section 6.4 describes the refueling accident.

When a refueling accident occurs, the release of radiation is estimated to immediately disperse into secondary containment. The reactor building ventilation isolates and SGT trains start on a PCIS Group 6 isolation signal. The reactor building atmosphere is exhausted through the SGT train filtration system and then through the elevated release point.

Distractors:

b. The reactor building ventilation systems isolate on a high radiation signal.

c. The trains start and release through the ERP.

d. The trains start and the reactor building ventilation systems isolate.

6.4 Fuel Handling Accident Accidents that result in the release of radioactive materials directly to the Secondary Containment can occur when the drywell is open. A survey of the various conditions that could exist when the drywell is open reveals that the greatest potential for the release of radioactive material occurs when the drywell head and reactor vessel head have been removed. In this case, radioactive material released as a result of fuel failure is available for transport directly to the Secondary Containment.

TABLE XIV-6-7 SGT SYSTEM FLOWS AND IODINE REMOVAL EFFICIENCIES Time = 0 seconds â 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months (Both SGT trains running)

Active Heater Failed Heater SGT Flow (cfm) 1492 1492 Elemental Iodine Efficiency 94% 89%

Particulate Iodine Efficiency 98% 98%

Organic Iodine Efficiency 94% 29%

Time = 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months â 30 days (One SGT train running, one train secured)

Single Train Single Train (cross-tie flow)

SGT Flow (cfm) 1492 288 Elemental Iodine Efficiency 94% 89%

Particulate Iodine Efficiency 98% 98%

Organic Iodine Efficiency 94% 29%

{AN1159}

TABLE XIV-6-8

/Q VALUES FOR THE EXCLUSION AREA BOUNDARY AND LOW POPULATION ZONE

/Q Values for the Exclusion Area Boundary Time Period /Q Value (sec/m3) Comments 0 to 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months 5.2E-4 Turbine Building Ground Level Release 0 to 1.3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months 1.6E-5 SGT System Elevated Release Point 1.3 to 1.8 1.2E-4 SGT System Elevated Release Point hours with fumigation 1.8 to 10 1.6E-5 SGT System Elevated Release Point hours

/Q Values for the Low Population Zone Time Period /Q Value (sec/m3) Comments 0 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 2.9E-4 Turbine Building Ground Level release 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 7.3E-5 Turbine Building Ground Level release 1 to 4 days 2.5E-5 Turbine Building Ground Level release 4 to 30 days 5.2E-6 Turbine Building Ground Level release 0 to 1.3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months 4.0E-5 SGT System Elevated Release Point 1.3 to 1.8 1.4E-4 SGT System Elevated Release Point hours with fumigation 1.8 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 4.0E-5 SGT System Elevated Release Point 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 1.6E-5 SGT System Elevated Release Point 1 to 4 days 5.8E-6 SGT System Elevated Release Point 4 to 30 days 1.7E-6 SGT System Elevated Release Point

{AN1165}

TABLE XIV-6-9

/Q VALUES FOR THE CONTROL ROOM INTAKE

/Q Value Occupancy Time Period (sec/m3) Factor Comments 0 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 8.64E-4 1 Turbine Building vent ground level release 2 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 4.66E-4 1 Turbine Building vent ground level release 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 2.32E-4 1 Turbine Building vent ground level release 1 to 4 days 1.53E-4 0.6 Turbine Building vent ground level release 4 to 30 days 1.25E-4 0.4 Turbine Building vent ground level release 0 to 5 minutes 4.15E-3 1 Reactor Building ground level release 0 to 1.3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months 1.00E-10 1 SGT System Elevated Release Point 1.3 to 1.8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 3.03E-4 1 SGT System ground level release with fumigation 1.8 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 1.00E-10 1 SGT System Elevated Release Point 2 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 8.58E-10 1 SGT System Elevated Release Point 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 1.41E-8 1 SGT System Elevated Release Point 1 to 4 days 5.62E-9 0.6 SGT System Elevated Release Point 4 to 30 days 5.69E-9 0.4 SGT System Elevated Release Point

{AN1166}

TABLE XIV-6-10 LOSS-OF-COOLANT ACCIDENT EXCLUSION AREA BOUNDARY, LOW POPULATION ZONE, AND CONTROL ROOM RADIOLOGICAL DOSE CONSEQUENCES TEDE (rem)

EAB LPZ Control Room Primary Containment 0.458 1.559 0.374

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 295024 EA2.06 Importance Rating 4.1 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): USAR XIV Section 6.3.7.2 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 49 New 00 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 1 1 4 Multiple Choice Topic Area Description Accidents High drywell pressure effects on suppression pool temperature Related Lessons INT006-01-15, Accident Analysis Related Objectives Related References EOP-1A, RPV CONTROL EOP-3A, PRIMARY CONTAINMENT CONTROL Related Skills (K/A) 295024 EA2.06 Ability to determine and/or interpret the following as they apply to HIGH DRYWELL PRESSURE: (CFR: 41.10 / 43.5 / 45.13) Suppression pool temperature (4.1)

QUESTION: 49 The plant is operating at rated conditions. A LOCA develops in the drywell causing drywell pressure to rise. What effect does this have on containment parameters over the next five minutes? (Assume no operator actions are taken)

a. Suppression pool temperature rises for one minute and then steadies out.

b. Suppression pool temperature continually rises.

c. Torus pressure lowers for one minute and then begins to rise.

d. Torus pressure equalizes with drywell pressure.

ANSWER: 49

b. Suppression pool temperature continually rises.

Explanation:

NOTE: See attached partial of USAR Chapter XIV, Section 6 USAR XIV 6.3.7.2 describes containment response to a LOCA.

The leak into the drywell results in the steam being condensed in the suppression pool.

Suppression pool water temperature will continually rise as long as energy is being added.

Distractors:

a. Suppression pool temperature continues to rise and is less than drywell pressure due to steam condensing in the suppression pool.

c. Torus pressure rises as inventory temperature rises and swells.

d. Torus pressure will always be lower than drywell pressure due to steam condensing in the suppression pool.

USAR Chapter XIV partial, 6.3.7.2 Long Term Containment Response Analysis In event of a design basis LOCA, the Core Spray system removes decay heat and stored heat from the core, thereby controlling core heatup and limiting metal-water reaction to less than 0.1 percent. The Core Spray water transports the core heat out of the reactor vessel through the broken recirculation line in the form of hot water. This hot water flows into the suppression chamber via the drywell-to-suppression chamber vent pipes. Steam flow is negligible. The energy transported to the suppression chamber water is then removed from the Primary Containment by the RHR heat exchangers.{

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 295025 EK1.03 Importance Rating 3.6 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): GE Drawing 944E689 (Rev N 13) Sheet 1 (Attach if not previously provided) GE Drawing 791E253 (Rev N 28) Sheet 2 (including version/revision number) EOP 1A (Rev 16)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 50 New 00 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 4 2 1 4 Multiple Choice Topic Area Description Emergency Procedures High reactor pressure/SRV tailpipe temperature/pressure relationships Related Lessons COR002-16-02 OPS Nuclear Pressure Relief Related Objectives COR0021602001040C Given a Nuclear Pressure Relief system component manipulation, predict and explain the changes in the following parameters:

Reactor Pressure COR0021602001050J Describe the Nuclear Pressure Relief system design features and/or interlocks that provide for the following: Safety/Relief operating signals Related References NONE Related Skills (K/A) 295025 EK1.03 Knowledge of the operational implications of the following concepts as they apply to HIGH REACTOR PRESSURE : (CFR: 41.8 to 41.10) Safety/relief valve tailpipe temperature/pressure relationships (3.6)

QUESTION: 50 The plant is operating at rated conditions when a DEH failure causes reactor pressure to raise to 1080 psig. The following occur:

The reactor scrams.

The control room operators enter EOP 1A, RPV CONTROL.

SRVs are cycling.

What does the EOPs require for pressure control and how does the control room operator verify proper operation?

a. Lower RPV pressure to 1000 psig by manually opening any SRV with proper operation indicated by the SRVs Red and Amber indicating lights illuminating.

b. Lower RPV pressure to 1000 psig by manually opening any SRV with proper operation indicated by the SRVs Red indicating light illuminating.

c. Verify proper Low-Low Set operation as indicated by SRVs D and F Red, Green, and Amber indicating lights illuminating.

d. Verify proper Low-Low Set operation as indicated by SRVs D and F Green and Red indicating lights illuminating.

ANSWER: 50

c. Verify proper Low-Low Set operation as indicated by SRVs D and F Red, Green, and Amber indicating lights illuminating.

Explanation:

NOTE: See attached partial GE Drawing 944E689 (Rev N13) Sheet 1 for LLS arming, partial GE Drawing 791E253 (Rev N28) Sheet 2 for SRV indicating lights and EOP 1A direction on pressure control.

A reactor scram signal is generated to RPS on RPV pressure at 1050 psig. When reactor pressure reaches 1080 psig, SRVs D and F (Low-Low Set SRVs) open and with an SRV open, LLS arms and SRVs D and F open and close on new pressure setpoints. When an SRV opens, a downstream pressure switch picks up at 30 psig which illuminates the Amber indicating light on Panel 9-3-1 for the SRV. The Red indicating light illuminates when the LLS logic arms and sends an open signal to the SRV. The Green indicating light is on any time the SRV control switch is in its normal CLOSE position. Reactor Pressure above 1050 psig is EOP 1A entry.

The pressure leg on EOP 1A directs the operators verify proper low-low set operation which is described above.

Distractors:

a. Normal reactor pressure is 1000 psig, but the EOPs require RPV pressure lowered below 940 psig after LLS is verified to be operating correctly. The indicating lights are correct for manual valve operation.

b. Normal reactor pressure is 1000 psig, but the EOPs require RPV pressure lowered below 940 psig after LLS is verified to be operating correctly. Manually opening the SRV illuminates the amber light via a down stream pressure switch.

d. Proper verification of the valve being open is an Amber light via a down stream pressure switch.

High RX Pressure Scram Signal SRV Open-30# pressure switch Picked up (SRV A, B, C or D)

K20A energizing inputs to opening the SRV

Control Switch remains in AUTO and contact is closed.

Green light remains lit.

30# pressure switch energizes this relay and K20A (previous page) closes this contact.

inputs to energize the Amber light illuminates SRV solenoid and illuminate Red light.

30# pressure switch closes on valve opening

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 295026 G 2.1.20 Importance Rating 4.6 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): EOP 6A (Rev 14) and EOP Graph 08 (Rev 14)

(Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: Boron Injection Initiation Limit Graph Learning Objective: See Attached (As available)

Question Source: Bank # 5334 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 51 5334 02 02/01/2005 05/23/2010 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 2 1 4 Multiple Choice Topic Area Description Emergency Operating INT0080606, High Suppression Pool temperature and Boron Procedures injection.

Related Lessons INT0080606 FLOWCHART 6A - RPV PRESSURE/POWER (FAILURE-TO-SCRAM)

INT0080618 OPS EOP AND SAG GRAPHS AND CAUTIONS Related Objectives INT00806060010800 Explain the basis for injecting boron before the Boron Injection Initiation Temperature is exceeded and when large periodic neutron flux oscillations in excess of 25% occur.

INT00806180010300 Given plant conditions and the EOP and SAG Graphs Flowchart, determine if operation is within the allowed region of a graph.

INT00806180010200 For each graph used in the flowcharts, identify the action(s) required if the parameters associated indicate operation in the restricted or prohibited area.

Related References INT0080618 EOP and SAG Graphs and Cautions INT0080606 Flowchart 6A - RPV Pressure/Power Failure-to-Scram 5.8 Emergency Operating Procedures (EOPs)

Related Skills (K/A) 295026 Suppression Pool High Water Temperature G2.1.20 Ability to interpret and execute procedure steps. (CFR: 41.10 / 43.5 / 45.12) (4.6)

QUESTION: 51 5334 An Anticipated Transient Without Scram (ATWS) has occurred. Average Reactor Thermal Power is 15% AND Suppression Pool temperature is 96°F and rising 1°F every 5 minutes.

What is the MAXIMUM temperature the Suppression Pool temperature may reach before Boron injection is required?

a. 110°F.

b. 123°F.

c. 139°F

d. 156°F ANSWER: 51 5334

c. 139° F MATERIAL REQUIRED FOR EXAMINATION: Boron Injection Initiation Limit Graph NOTE: See attached BIIT Graph (EOP Graph 08) and partial EOP 6A steps.

Explanation:

EOP 6A, REACTOR POWER (Failure-To-Scram) flow chart step FS/Q-11 states to inject boron before suppression pool temperature reaches BIIT (Graph 08). The BIIT for 15 % power is 140°F.

Distractors:

a. 110° F is the Tech Spec limit but is not the BIIT for 15% power.

b. This is the temperature limit for reactor power of 20% (vertical graph line next to 15%

power).

d. This is the temperature limit for reactor power of 10% (vertical graph line next to 15%

power).

BIIT Graph 8 190 BORON INJECTION INITIATION TEMPERATURE (GRAP08) 184 180 A

170 Average Torus Water Temperature (°F) 160 B

150 140 130 120 110 100 C

0 5 10 15 20 25 30 2.4 Average Thermal Reactor Power (%)

23.5 D

Partial EOP 6A Steps requiring boron injection

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 295028 EK3.02 Importance Rating 3.5 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): AMP-TB00 (Rev 6) PSTG Contingency #4 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank # 23331 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR Part 55 Content: 55.41 5 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 52 23331 00 06/27/2006 05/23/2010 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 2 1 4 Multiple Choice Topic Area Description Abnormal/Emergency INT0080618, LOCA occurred what is action is performed next Procedures and why?

Related Lessons INT0080618 OPS EOP AND SAG GRAPHS AND CAUTIONS Related Objectives INT00806180010300 Given plant conditions and the EOP and SAG Graphs Flowchart, determine if operation is within the allowed region of a graph.

Related References EOP-2B EOP Flow Chart 2B Related Skills (K/A) 295028 EK3.02 Knowledge of the reasons for the following responses as they apply to HIGH DRYWELL TEMPERATURE: (CFR: 41.5 / 45.6) RPV flooding (3.5)

QUESTION: 52 23331 A Loss of Coolant Accident has occurred with the following conditions:

Reactor pressure 200 psig (lowering slowly)

Torus pressure 5 psig (rising slowly)

Drywell temperature 400° F (all points) (steady)

RPV water level can no longer be determined.

What action is performed next and why?

a. Spray the Drywell to lower primary containment pressure.

b. Spray the torus to lower primary containment temperature.

c. Flood the RPV because fuel submergence cannot be assured.

d. Flood the RPV because minimum steam cooling pressure is not met.

ANSWER: 52 23331

c. Flood the RPV because fuel submergence cannot be assured.

Explanation:

NOTE: See attached EOP 1A and 7A overrides Overrides in EOP 1A (non-ATWS) or 7A (ATWS) direct RPV Flooding if RPV level cannot be determined. AMP-TB00 (Rev 6) PSTG Contingency #4 RPV Flooding Overview states:

The RPV flooding strategies in PSTG Contingency #4 are used to cool the core when RPV water level cannot be determined. The specified actions first depressurize the RPV, then control injection to establish and maintain one of the following conditions:

The RPV flooded to the elevation of the main steam lines. The core will then be cooled by full submergence. This condition may ultimately be achieved under either shutdown or failure-to-scram conditions.

Distractors:

a. Torus may be sprayed at this pressure but drywell spray is not initiated until Torus pressure reaches 10 psig.

b. Drywell spray is required before drywell temperature reaches 280° F (containment design temperature) so sprays should already be in service. Spraying the torus air space will not provide cooling to the drywell.

d. The RPV is emergency depressurized if drywell temperature cannot be restored and maintained below 280° F so ED should have already begun. Minimum steam cooling pressure (MSCP) is only applicable during ATWS conditions. Assuming 6 SRVs are open for ED, the MSCP is 135 psig so MSCP is met at 200 psig.

EOP 1A Level leg override EOP 7A Level Leg

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 295030 EK2.04 Importance Rating 3.7 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): EOP Graph 4 (Rev 14) Vortex Limits for RHR (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: EOP/SAG NPSH &

VORTEX GRAPHS Learning Objective: See Attached (As available)

Question Source: Bank # 23333 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 53 23333 00 08/03/2007 05/23/2010 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 3 1 4 Multiple Choice Topic Area Description Abnormal/Emergency INT0080618, What action is required for 1C RHR pump and Procedures why?

Related Lessons INT0080618 OPS EOP AND SAG GRAPHS AND CAUTIONS Related Objectives INT00806180010300 Given plant conditions and the EOP and SAG Graphs Flowchart, determine if operation is within the allowed region of a graph.

Related References EOP/SAG Graphs Related Skills (K/A) 295030 EK2.04 Knowledge of the interrelations between LOW SUPPRESSION POOL WATER LEVEL and the following: (CFR: 41.7 / 45.8) RHR/LPCI (3.7)

QUESTION: 53 23333 A LOCA has occurred with the following conditions:

RHR Pump C is the only low pressure ECCS pump available.

Maximum injection is required

Current RHR flow rate 7000 gpm

Torus pressure 9.57 psig (stable)

Torus average water temp 185°F (rising slowly)

Primary containment water level 5 feet (stable)

The CRS directs that NPSH and Vortex requirements be complied with.

What action is required for RHR Pump C and why?

a. Lower flow to 6000 gpm because of NPSH limits.

b. Lower flow to 6000 gpm because of Vortex limits.

c. Raise flow to 7700 gpm as allowed by Vortex limits.

d. Raise flow to 8600 gpm as allowed by NPSH limits.

ANSWER: 53 23333

b. Lower flow to 6000 gpm because of Vortex limits.

REFERENCE TO BE SUPPLIED TO THE STUDENTS: EOP/SAG NPSH and VORTEX GRAPHS Explanation:

NOTE: See attached EOP Graph 4 Vortex Limits and Graph 5 NPSH limits.

The flow must be reduced to comply with the vortex limit. For a Torus level of 5 feet, the limit is 6000 gpm. With torus pressure high and water temperature at 185°F one pump is limited to its design capacity, 7700 gpm.

Distractors:

a. NPSH is not limiting in this case.

c. Raising flow is not allowed because Vortex limits have to be complied with. Also 7700 gpm is the design flow for one pump.

d. Raising flow is not allowed because Vortex limits have to be complied with. Also 7700 gpm is the design flow for one pump.

Vortex is limiting to 6,000 gpm.

Torus Overpressure is calculated to be 10 psig NPSH NOT limiting 10 psig overpressure line 185°F SP temperature

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 295031 EK3.02 Importance Rating 4.4 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): AMP-TBD00 (Rev 6) Contingency #1 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR Part 55 Content: 55.41 5 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 54 New 00 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 1 1 4 Multiple Choice Topic Area Description EOPs Adequate core cooling by core submergence Related Lessons INT008-06-09 OPS EOP Flowchart 1A - RPV Control, RPV Level Related Objectives INT00806090010100 Describe the three mechanisms specified in the EOPs to assure adequate core cooling including the RPV water level band required and which is the preferred method.

Related References NONE Related Skills (K/A) 295031 EK3.02 Knowledge of the reasons for the following responses as they apply to REACTOR LOW WATER LEVEL : (CFR: 41.5 / 45.6) Core coverage (4.4)

QUESTION: 54 Given the following conditions:

The plant has experienced a LOCA.

CRD is the only system injecting into the RPV.

All control rods have fully inserted.

RPV pressure is stable at 900 psig utilizing SRVs.

RPV water level has stabilized at -156 inches (Corrected Fuel Zone).

What is the status of core cooling?

a. Adequate core cooling does NOT exist because the core is uncovered.

b. The core is covered so adequate core cooling exists.

c. The core is adequately cooled because Minimum Steam Cooling Pressure conditions exist.

d. Steam cooling is providing adequate core cooling at this RPV level.

ANSWER: 54

b. The core is covered so adequate core cooling exists.

Explanation:

NOTE: See attached PSTG and Procedure 4.6.1 partial.

AMP-TBD00 (Rev 6) Contingency #1 (CNS PSTGs) define top of active fuel as -158 inches (Fuel Zone instruments). If the core is covered, adequate core cooling exists. With level stabilized at -156 inches RPV water level is above the top of active fuel.

Distractors:

a. RPV level below -158 inches is considered adequate core cooling with injection. The RPV level is above the top of active fuel so the core is adequately cooled by submergence.

c. MSCP is only appropriate during ATWS conditions. The SRVs are controlling pressure but there is no steam cooling as the core is covered.

d. There is no steam cooling as the core is covered.

PSTG PSTG/SATG Step (First Override)

If while executing the following steps:

It has not been determined that the reactor will remain shutdown under all conditions without boron, enter Contingency #5.

RPV water level cannot be determined, enter Contingency #4.

RPV water level can be restored and maintained above -150 in. (top of fuel as indicated on Wide Range), enter the RPV Control Guideline at Step RC/L.

Procedure 4.6.1 partial

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 295037 EK1.02 Importance Rating 4.1 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): AMP-TB00 (Rev 6) Contingency #5 (CNS PSTGs)

(Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank # 14478 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR Part 55 Content: 55.41 5 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 55 14478 02 12/12/2008 06/15/2005 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 2 1 3 Multiple Choice Topic Area Description Emergency Operating INT0080610, Lowering RPV level effects on reactor power.

Procedures Related Lessons INT0080610 OPS EOP FLOWCHART 7A - RPV LEVEL (FAILURE-TO-SCRAM)

Related Objectives INT00806100010900 Given an EOP flowchart 7A, RPV LEVEL (FAILURE TO SCRAM) step, state the reason for the actions contained in the step.

INT00806100010800 Given plant conditions and EOP flowchart 7A, RPV LEVEL (FAILURE TO SCRAM), determine required actions.

Related References None Related Skills (K/A) 295037 EK1.02 Knowledge of the operational implications of the following concepts as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN :(CFR: 41.8 to 41.10) Reactor water level effects on reactor power. (4.1)

QUESTION: 55 14478 The plant is operating at 100% power when the following occur:

Main Generator fault with a Main Turbine trip.

Very little rod motion on the scram.

The crew initiates ARI.

Both RR pumps are tripped.

The Main Turbine Bypass Valves control pressure.

The feedwater system remains in operation.

Current plant conditions are:

Reactor power 15% (stable)

Reactor Pressure 935 psig (stable)

Reactor water level +13"(NR) (stable)

The CRS orders RPV injection stopped and prevented.

Why is stop and prevent injection ordered?

a. To lower reactor water level which mitigates the consequences of thermal-hydraulic instabilities.

b. To prevent uncontrolled injection of large amounts of cold unborated water and fuel damage due to overpower conditions.

c. To prevent injection from sources inside the shroud which cause large power excursions and fuel damage.

d. To raise the subcooling of the cold unborated feedwater injection which lowers reactor power and heat addition to containment.

ANSWER: 55 14478

a. To lower reactor water level which mitigates the consequences of thermal-hydraulic instabilities.

Explanation:

NOTE: See attached AMP-TB00 (Rev 6) Contingency #5 explanation on level power effects.

Lowering reactor water level brings the boiling region lower in the core and causes reactor power to lower. Level is lowered below -60 inches to lower the subcooling of the injection from the feedwater spargers which in turn prevents large irregular neutron flux oscillations induced by neutronic/thermal-hydraulic instabilities.

Distractors:

b. Stopping and preventing injection prior to emergency depressurization ensures no uncontrolled injection of relatively cold water. There is no requirement to emergency depressurize.

c. Core Spray is not utilized for injection during an ATWS unless only as a last resort. Only outside shroud injection systems are utilized. Core Spray is prevented from injection to lower level in this case.

d. Subcooling is lowered not raised by lowering RPV water level. This raises the temperature of the unborated feedwater injection and lessens neutronic/thermal-hydraulic instabilities.

PSTG/SATG Step C5-3 If reactor power is above 3% (APRM downscale trip) or cannot be determined and RPV water level is above -60 in. (24 in. below the feedwater sparger nozzles), lower RPV water level to below -60 in. (24 in. below the feedwater sparger nozzles) by terminating and preventing all injection into the RPV except from boron injection systems, RCIC, and CRD, defeating interlocks if necessary.

Discussion To prevent or mitigate the consequences of any large irregular neutron flux oscillations induced by neutronic/thermal-hydraulic instabilities, RPV water level is lowered sufficiently below the elevation of the feedwater sparger nozzles. This places the feedwater spargers in the steam space providing effective heating of the relatively cold feedwater and eliminating the potential for high core inlet subcooling. For conditions that are susceptible to oscillations, the initiation and growth of oscillations is principally dependent upon the subcooling at the core inlet; the greater the subcooling, the more likely oscillations will commence and increase in magnitude.

If reactor power is at or below the APRM downscale trip setpoint, it is highly unlikely that core bulk boiling boundary would be below that which provides suitable stability margin for operation at high powers and low flows. (A minimum boiling boundary of 4 ft above the bottom of fuel has been shown to be effective as a stability control because a relatively long two-phase column is required to develop a coupled neutronic/ thermal-hydraulic instability.) Furthermore, flow/density variations would be limited with reactor power this low since the core has a relatively low average void content. Therefore, there is significant stability margin with power at or below the APRM downscale trip setpoint.

Twenty-four inches below the lowest nozzle in the feedwater sparger has been selected as the upper bound of the RPV water level control band. This water level is sufficiently low that steam heating of the injected water will be at least 65% to 75% effective (i.e., the temperature of the injected water will be increased to 65% to 75% of its equilibrium value in the steam environment. This water level is sufficiently high that the capability to bypass the low RPV water level MSIV isolation should be able to control RPV water level with feedwater pumps to preclude the isolation.

Lowering RPV water level is accomplished by terminating and preventing all injection into the RPV, except from boron injection systems, RCIC, and CRD. Boron injection systems, RCIC, and CRD are relatively low flow systems.

With RPV injection terminated, RPV water level and reactor power decrease at the maximum possible rate allowed by boil off. Failure to completely stop RPV injection flow (with the exception of CRD, RCIC, and SLC) would delay the reduction in core inlet subcooling, thus increasing the potential for flux oscillations.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 295038 EK3.04 Importance Rating 3.6 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): AMP-TB00 (Rev 6) Contingency #2 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank # 19785 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR Part 55 Content: 55.41 5 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 56 19785 00 03/19/2003 06/15/2005 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 1 1 3 Multiple Choice Topic Area Description Emergency Operating INT00806170010700, OPS FLOWCHART 5A - SECONDARY Procedures CONTAINMENT AND RADIOACTIVITY RELEASE CONTROL Related Lessons INT0080617 OPS FLOWCHART 5A - SECONDARY CONTAINMENT AND RADIOACTIVITY RELEASE CONTROL Related Objectives INT00806170010700 Given plant conditions and EOP flowchart 5A, SECONDARY CONTAINMENT CONTROL and RADIOACTIVITY RELEASE CONTROL, state the reasons for the actions contained in the steps.

Related References NONE Related Skills (K/A) 295038 EK3.04 Knowledge of the reasons for the following responses as they apply to HIGH OFF-SITE RELEASE RATE: (CFR: 41.5 / 45.6) Emergency depressurization (3.6)

QUESTION: 56 19785 During the execution of EOP 5A, RADIOACTIVITY RELEASE CONTROL, the CRS directs Emergency Depressurization (ED) to be performed.

What is the basis for Emergency Depressurizing while performing EOP 5A, Radioactive Release Control actions?

a. ED ensures the availability of equipment in the turbine building, which is necessary to mitigate the event, is not challenged.

b. ED ensures the energy level of the radiation and the atmospheric dispersion factors fall within the bounds of the accident analysis.

c. ED ensures the isotopic mixture of radioactive materials deposited off-site will be within the bounds of the accident analysis.

d. ED ensures the lowest possible driving head and flow of primary systems that are discharging outside of containment.

ANSWER: 56 19785

d. ED ensures the lowest possible driving head and flow of primary systems that are discharging outside of containment.

Explanation:

NOTE: See attached partial of PSTG Contingency #2 explanation.

Per AMP-TB00 (Rev 6) Contingency #2, ED is performed to minimize the release of radiation to sources external to primary and secondary containment. ED lowers the RPV pressure which is the driving force of the radiation release. Placing the reactor in the lowest energy state possible lowers the release rate.

Distractors:

a. The availability of turbine building equipment is not an EOP consideration, but sounds like the reactor building ED reason.

b. This is a reason to use the DOSE program for the projections vice the ODAM calculations, but is not the reason for the ED.

c. This is not the basis for the ED.

10. PSTG Contingency #2 Emergency RPV Depressurization Overview The actions specified in Contingency #2 rapidly depressurize the RPV and maintain the RPV in a depressurized state. The steps of this contingency may be required to:

Establish or maintain adequate core cooling.

Terminate or minimize the discharge of reactor coolant from unisolable primary system breaks.

Reduce the energy within the RPV before reaching plant conditions for which the pressure suppression system may not be able to safely accommodate an SRV opening or condense steam discharged through the downcomers.

Minimize radioactivity release from the RPV to the primary containment and secondary containment, or to areas external to the primary containment and secondary containment.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 600000 AK1.01 Importance Rating 2.5 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): B&R P&ID 2016 (Rev N 06) Sheet 4 (Attach if not previously provided) MSDS for Halon (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 57 New 00 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 1 1 4 Multiple Choice Topic Area Description Emergency Plant Evolutions Fire Classification Related Lessons GEN005-10-03 Fires and Extinguishing Agents Related Objectives GEN005-10-03 Objective 5 Related References NONE Related Skills (K/A) 600000 AK1.01 Knowledge of the operation applications of the following concepts as they apply to Plant Fire On Site: Fire Classifications by type (2.5)

QUESTION: 57 An active electrical fire has been automatically extinguished in the Service Water Pump room.

What class of fire existed and what are the potential hazards associated with entering the Service Water Pump room?

a. Class D fire and the hazard is electrocution due to water deluge discharge.

b. Class C fire and the hazard is electrocution due to water deluge discharge.

c. Class C fire and the hazard is physical symptoms such as dizziness, headache, and confusion due to Halon discharge.

d. Class D fire and the hazard is physical symptoms such as dizziness, headache, and confusion due to Halon discharge.

ANSWER: 57

c. Class C fire and the hazard is physical symptoms such as dizziness, headache, and confusion due to Halon discharge.

Explanation:

NOTE: See partial B&R P&ID 2016 (Rev N 06) Sheet 4 attached.

An electrical fire is classified as a type C fire. The Service Water Pump room is protected by a Halon system. The Service Water Pump room H&V shuts down and the ventilation dampers all close on Halon discharge. These systems must be manually reset. The MSDS for Halon states:

Human health effects of overexposure by inhalation may include nonspecific discomfort, such as nausea, headache, or weakness; or temporary central nervous system depression with effects such as dizziness, headache, confusion, un-coordination, and loss of consciousness. Higher exposures by inhalation may cause temporary alteration of the heart's electrical activity with irregular pulse, palpitations, or inadequate circulation. skin or eye contact with the liquid may cause frostbite.

Distractors:

a. Class D fires are metal fires. This is an electrical fire. SW Pump room has no water deluge fire protection system.

b. The electrical fire is a class C, but there is no water deluge fire protection system in the SW Pump room.

d. Class D fires are metal fires. This is an electrical fire.

Service Water Pump Room Halon system ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 700000 G2.4.2 Importance Rating 4.5 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): EOP 1A (Rev 16)

(Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 58 New 00 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 2 1 4 Multiple Choice Topic Area Description EOPs Electrical and Grid Disturbances and EOP Entry Conditions Related Lessons INT008-06-09 OPS EOP Flowchart 1A RPV Control, RPV Level Related Objectives INT0080605001010A List the entry conditions of Flowchart 1A: Describe the importance of each in an emergency situation Related References NONE Related Skills (K/A) 700000 Generator Voltage and Electric Grid Disturbances G2.4.2 Knowledge of system set points, interlocks and automatic actions associated with EOP entry conditions.(CFR: 41.7 / 45.7 / 45.8) (4.5)

QUESTION: 58 The plant is operating at 96% power and grid disturbances cause the turbine generator to trip.

What EOP(s) require entry within the first 5 minutes after the transient?

a. EOPs 1A and 3A on high drywell pressure.

b. EOP 1A on low RPV level and high RPV pressure.

c. EOP 5A on secondary containment temperature.

d. EOP 3A on high suppression pool level.

ANSWER: 58

b. EOP 1A on low RPV level and high RPV pressure.

Explanation:

The turbine generator trip is on over-excitation. A generator trip at this reactor power level causes a turbine trip. Closure of the main turbine stop valves causes a pressure spike in the system. RPV high pressure trip of 1050 psig is EOP 1A entry condition. The scram causes RPV level to shrink due to void collapse and an RPV low level below +3 inches is reached which is another EOP 1A entry condition. The main turbine bypass valves trip open and then throttle to control RPV pressure.

Distractors:

a. There is no leak into containment nor is there a loss of cooling so drywell pressure will not rise to EOP entry levels of 1.84 psig.

c. There will be a PCIS Group 6 isolation on low RPV level and reactor building ventilation will isolate. SGT will start and maintain containment pressure negative.

d. The main turbine bypass valves control RPV pressure so no energy will go into the suppression pool.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # 295002 AK2.02 Importance Rating 3.1 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): Abnormal Procedure 2.4VAC (Rev 23)

(Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank # 16653 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 59 16653 04 02/17/2011 06/15/2005 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 2 1 4 Multiple Choice Topic Area Description None INT0320132, CNS Abnormal Procedures (RO) Off Gas/Vacuum Related Lessons INT0320132 CNS Abnormal Procedures (RO) Off Gas/Vacuum Related Objectives INT0320132K0K0100 Given plant condition(s), determine from memory if a Main Turbine trip is required due to the event(s).

INT0320132I0I0100 Given plant condition(s), determine from memory all immediate operator actions required to mitigate the event(s).

INT0320132G0G0100 Given plant condition(s), determine from memory the appropriate Abnormal/Emergency Procedure(s) to be utilized to mitigate the event(s).

INT0320132J0J0100 Given plant condition(s), determine from memory if a manual reactor scram or an emergency shutdown from power is required due to the event(s).

Related References 2.4VAC Loss of Condenser Vacuum Related Skills (K/A) 295002 AK2.02 Knowledge of the interrelations between LOSS OF MAIN CONDENSER VACUUM and the following: (CFR: 41.7 / 45.8) Main turbine (3.1)

QUESTION: 59 16653 A plant shutdown is in progress. With the Reactor operating at 19% power (main turbine still on the line), the BOP operator notices that steam pressure to the Steam Jet Air Ejectors is lowering rapidly and subsequently determines that opening the SJAE steam supply bypass valves does not correct the problem.

The crew enters Abnormal Procedure 2.4VAC, LOSS OF CONDENSER VACUUM and vacuum is now 22Hg.

What immediate action(s) is/are required?

a. Rapidly reduce Reactor power per Procedure 2.1.10.

b. Place Mechanical Vacuum Pumps in service per Procedure 2.2.55.

c. Scram the Reactor, then trip the Main Turbine per Procedure 2.1.5.

d. Trip the Main Turbine and enter Procedure 2.2.77.

ANSWER: 59 16653

d. Trip the Main Turbine and enter Procedure 2.2.77.

Explanation:

NOTE: See attached Procedure 2.4VAC partial.

The conditions of the question will cause entry into 2.4VAC. Rx power is less than 30% so annunciator 9-5-2/C-4 is present. Step 3.1 requires tripping the MT and entering 2.2.77.

Distractors:

a. Rapid power reduction is not an action given in 2.4VAC. Many abnormal procedures have immediate operator actions to rapidly reduce reactor power. The procedure allows lowering power per Procedure 2.1.10 to maintain vacuum and power is already at 19%.

With loss of SJAE flow, lowering power will not restore condenser vacuum to an acceptable level.

b. Rx power is above 5% so Mech Vac pump operation is prohibited.

c. A RX scram is not required with annunciator 9-5-2/C-4 present.

Procedure 2.4VAC partial ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # 295007 G 2.1.7 Importance Rating 4.4 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): Technical Specification (Table 3.3.6.3-1) Function 2 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank # 5425 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR Part 55 Content: 55.41 5 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 60 5425 01 04/22/2007 06/15/2005 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 1 1 4 Multiple Choice Topic Area Description Systems Nuclear Pressure Relief Related Lessons COR0021602 OPS NUCLEAR PRESSURE RELIEF Related Objectives COR0021602001050J Describe the Nuclear Pressure Relief system design features and/or interlocks that provide for the following: Safety/Relief operating signals Related References None Related Skills (K/A) 295007 High Reactor Pressure G 2.1.7 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

(CFR: 41.5 / 43.5 / 45.12 / 45.13) (4.4)

QUESTION: 60 5425 The plant has been operating at 100% power for the past 90 days following a refueling outage, when the following events occur:

An MSIV isolation occurs for unknown reasons.

The Reactor is shutdown.

RPV Pressure rises to 1090 psig, then lowers to 875 psig.

20 minutes later, pressure is cycling between 990 AND 825 psig.

Which of the following statements describes the status of Low Low Set (LLS)?

a. Neither LLS valve is controlling pressure within its designed range.

b. LLS is correctly controlling pressure with SRV F. SRV D is NOT responding as designed.

c. LLS is correctly controlling pressure with SRV D. SRV F is NOT responding as designed.

d. LLS is correctly controlling pressure with BOTH LLS valves.

ANSWER: 60 5425

a. Neither LLS valve is controlling pressure within its designed range.

Explanation:

NOTE: See attached TS Table listing LLS allowable values.

Technical Specification (Table 3.3.6.3-1) Function 2 allowable values are:

SRV D Open between 995 and 1035 and close between 855 and 895.

SRV F Open between 1015 and 1045 and close between 855 and 895.

Both SRVs are opening too soon but are closing within their allowable value.

Distractors:

b. Neither SRV is opening at the correct pressure range.

c. Neither SRV is opening at the correct pressure range.

d. Both closing pressures are within specifications but neither valve is opening within specification.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # 295008 AK1.01 Importance Rating 3.0 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): Per Technical Specification SR 3.3.2.2.2 (Attach if not previously provided) USAR VII 10.3 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR Part 55 Content: 55.41 3*

55.43 Comments:

10CFR55.41.3 is more appropriate for 295008 AK1.01 than the listed 41.8 to 41.10.

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 61 New 00 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 1 1 4 Multiple Choice Topic Area Description Systems RPV level effects on carryover Related Lessons COR0011502 OPS Nuclear Boiler Related Objectives COR0011502001010M State the purpose of the following items related to Nuclear Boiler:

Shroud Head and Steam Separator Assembly Related References NONE Related Skills (K/A) 295008 AK1.01 Knowledge of the operational implications of the following concepts as they apply to HIGH REACTOR WATER LEVEL : (CFR: 41.8 to 41.10) Moisture carryover (3.0)

QUESTION: 61 The plant is operating at power when actual RPV level rises to +48 inches on the narrow range instruments for unknown reasons. RPV level remains steady at this level.

What operational concerns exist for this condition?

a. Instrument inaccuracies result in RPV level being above a procedurally driven scram setpoint.

b. Instrument inaccuracies result in high RPV level turbine trips being non-conservative.

c. RPV level being high enough that moisture carryover causes steam quality to lower.

d. RPV level being high enough that moisture carryover causes steam quality to rise.

ANSWER: 61

c. RPV level being high enough that moisture carryover causes steam quality to lower.

Explanation:

NOTE: See attached USAR excerpts.

Per Technical Specification SR 3.3.2.2.2, the allowable value is 54.0 inches for main and feedwater high level trip. USAR VII 10.3 describes carryover. RPV level above 48 inches starts blocking RPV steam separator return paths to the downcomer so water droplet carryover becomes a problem. More moisture carryover results in steam quality being lower.

Distractors:

a. Instrument inaccuracies are taken into account when setting administrative limits on operator actions. Procedure 2.4RXLVL, (Rev 25), RPV Water Level Control Trouble contains a scram action if RPV level cannot be maintained below +50 inches on the narrow range.

b. Instrument inaccuracies are taken into account when selecting turbine trip setpoints.

The main turbine high RPV level automatic trip setpoint is 52.5 inches.

d. Steam quality lowers and does not rise. Steam quality will rise if RPV level is lowered back to its desired level of +35 inches narrow range.

USAR VII Sections 10.3 Description The feedwater control system, during planned operation, automatically regulates feedwater flow into the reactor vessel. The system is capable of being manually operated.

The optimum reactor vessel water level range is determined by the requirements of the steam separators, which limit the water carryover with the steam going to the turbines and limit the steam carryunder with the water returning to the core. The water level in the reactor vessel is maintained within +/-2 inches of the optimum level during normal power operation. This control capability is achieved during plant load changes by balancing the mass flow rate of feedwater to the reactor vessel with the steam flow from the reactor vessel. The feedwater flow regulation is achieved by adjusting the reactor feed pump turbine control valves or, when in service, adjusting the feedwater startup bypass control valves to deliver the required feedwater flow to the reactor vessel.

During a reactor vessel overfill event, the feedwater control system causes the Main Turbine and the feedwater pump turbines to trip on high RPV water level (Level 8).

10.3.6 Reactor Overfill Protection In Generic Letter 89 19, the NRC stated that BWR plants should provide automatic reactor vessel overfill protection due to feedwater control system failures at power. The CNS design provides for a non essential overfill protection system initiated on a high reactor water level signal (Level 8). Three separate reactor water level channels are arranged in a 2 out of 3 logic which trip the feedwater pump turbines and the Main Turbine. While some of the components in the overfill protection channels share a common power supply with the feedwater control system, no single power supply failure will cause the loss of both the feedwater control system and the reactor overfill protection function. The 2 out of 3 initiating logic provides inherent protection against inadvertent or spurious actuations. The actuation setpoint is low enough to prevent gross moisture carryover to the Main Turbine, while providing sufficient margin above normal reactor water level to prevent spurious turbine trips. Accordingly, adequate protection is provided against a reactor vessel overfill event, as described in Generic Letter 89 19. See also USAR Section XIV 5.8.1 and Appendix G, Section G 5.3, Event 32).

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # 295010 AA2.02 Importance Rating 3.8 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): Technical Specifications Table 3.3.1.1-1 Function 6 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 62 New 00 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 1 1 3 Multiple Choice Topic Area Description Abnormal Procedures High drywell pressure scram Related Lessons COR0022102 OPS REACTOR PROTECTION SYSTEM Related Objectives COR0022102001100K Describe the interrelationship between the RPS and the following:

Primary Containment Related References NONE Related Skills (K/A) 295010 AA2.02 Ability to determine and/or interpret the following as they apply to HIGH DRYWELL PRESSURE : (CFR: 41.10 / 43.5 / 45.13) Drywell pressure (3.8)

QUESTION: 62 A leak has developed in the drywell. What MINIMUM drywell pressure will cause an automatic reactor scram?

a. 0.6 psig

b. 1.5 psig

c. 1.84 psig

d. 2.0 psig ANSWER: 62

c. 1.84 psig Explanation:

NOTE: See attached TS Table 3.3.1.1-1 for Function 6 Technical Specifications Table 3.3.1.1-1 Function 6 allowable value of RPS instrumentation for drywell pressure-high is 1.84 psig.

Distractors:

a. This represents the high drywell pressure alarm.

b. This represents the drywell pressure limit of Abnormal Procedure 2.4PC requiring a manual scram insertion.

d. This represents the drywell pressure drywell spray limitation.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # 295017 AK3.04 Importance Rating 3.6 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s):

5.2FUEL (Rev 17), FUEL FAILURE (Attach if not previously provided) 2.4OG, (Rev 19), OFF-GAS ABNORMAL (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR Part 55 Content: 55.41 5 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 63 New 00 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 2 1 4 Multiple Choice Topic Area Description Abnormal Procedures Reason for lowering power on high off-site release rate Related Lessons INT032-01-30 CNS Abnormal Procedures-High Radiation Related Objectives INT0320130F0F0100 Given plant condition(s), and the applicable Abnormal/Emergency Procedure, determine the correct subsequent actions required to mitigate the event(s).

Related References 5.2FUEL, FUEL FAILURE 2.4OG, OFF-GAS ABNORMAL Related Skills (K/A) 295017 AK3.04 Knowledge of the reasons for the following responses as they apply to HIGH OFF-SITE RELEASE RATE : (CFR: 41.5 / 45.6) Power reduction (3.6)

QUESTION: 63 Power has been raised to 90% after a control rod swap evolution. The control room operator notices SJAE Radiation monitor indication slowly rising. Annunciator 9-4-1/C-4, OFFGAS TIMER INITIATED, alarms 15 minutes later. The CRS enters Procedures 5.2FUEL, FUEL FAILURE and 2.4OG,OFF-GAS ABNORMAL.

What is the appropriate operator action and why?

a. Scram the reactor and place OFFGAS TIMER switch to CLOSE to contain the radiation release.

b. Lower reactor power to lower the radiation release.

c. Scram the reactor to lower the fission product gap release and lower the radiation release.

d. Lower reactor power and close the MSIVs to contain the radiation release.

ANSWER: 63

b. Lower reactor power to lower the radiation release.

Explanation:

NOTE: See attached Procedure 5.2FUEL partial The control rod swap and raising power has resulted in a fuel clad leaking. In order to lower the radiation release rate, reactor power must be lowered as a first step. Lowering reactor power reduces the fission product gases being produced in the fuel pin that is leaking. Lowering the fission product release lowers the release of radiation.

Distractors:

a. Procedure guidance is to isolate off-gas by use of the OFFGAS TIMER control switch which immediately isolates off-gas. Procedure 2.4OG only requires a reactor scram after off-gas isolates or is manually isolated. The scram is not performed first.

c. There is no procedure guidance to scram the reactor first. Reactor power is lowered first to minimize the radiation released.

d. Procedure 5.2FUEL only requires scramming the reactor if a valid main steam line Hi-Hi radiation alarm is received. There is no main steam line HI-Hi radiation alarm present.

The main steam line Hi-Hi radiation alarm comes in around 900 mR/hr. The Off-Gas Hi radiation alarm comes in around 75 mR/hr.

Procedure 5.2FUEL partial ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # 295029 EA1.01 Importance Rating 3.4 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): AMP-TB00 (Rev 6) CNS PSTGs (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank # 24586 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 64 24586 00 02/19/2009 05/23/2010 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 3 1 4 Multiple Choice Topic Area Description Emergency Operating INT0080613, Actions required with primary containment water Procedures level at 16 ft.

Related Lessons INT0080613 OPS EOP FLOWCHART 3A - PRIMARY CONTAINMENT CONTROL Related Objectives INT00806130011100 Given plant conditions and EOP Flowchart 3A, PRIMARY CONTAINMENT CONTROL, determine required actions.

Related References 5.8 Emergency Operating Procedures (EOPs)

Related Skills (K/A) 295029 EA1.01 Ability to operate and/or monitor the following as they apply to HIGH SUPPRESSION POOL WATER LEVEL: (CFR: 41.7 / 45.6) HPCI: Plant-Specific. (3.4)

QUESTION: 64 24586 Following a LOCA and SCRAM the following conditions are noted:

HPCI high suppression pool water level suction transfer logic is defeated per 5.8.20.

HPCI is injecting at 4000 gpm with suction from the ECST.

Reactor pressure is 650 psig.

Reactor water level is -155 inches (corrected FZ) and steady.

Drywell temperature is 185°F and rising slowly.

Drywell pressure is 4.0 psig and rising slowly.

Primary Containment water level is 15 ft and rising an inch per minute.

Suppression pool temperature is 120°F and steady.

What action is required?

a. Initiate drywell sprays.

b. Emergency depressurize.

c. Align HPCI suction to the Suppression Pool.

d. Vent primary containment with drywell vent line.

ANSWER: 64 24586

c. Align HPCI suction to the Suppression Pool.

Explanation:

NOTE: See attached AMP-TB00 (Rev 6) CNS PSTGs on terminating injection from sources external to PC on a high SP water level.

Since suppression pool level is now at 15 feet and rising, EOP-3A requires that injection from sources outside the primary containment be stopped. Since adequate in containment injections sources are available to ensure adequate core cooling, this action is required by SP/L-5.

Distractors:

a. is incorrect because drywell sprays are not permitted with torus pressure at the current 2.6 psig.

b. is incorrect because emergency depressurization is not required yet because step SP/L has not yet been accomplished. And there are sufficient indications available to provide the operator with indications that this step will be effective.

d. is incorrect because venting the drywell is not allowed with the LOCA signal present and pressure well below PCPL-A.

PSTG/SATG Step (SP/L-3.1, paragraphs 2 & 3)

If suppression pool water level cannot be maintained below 16 ft (SRV Tail Pipe Level Limit), enter the RPV Control Guideline at Step RC-1 and execute it concurrently with this procedure.

If adequate core cooling is assured, terminate injection into the RPV from sources external to the primary containment except from systems required to shut down the reactor.

Discussion Entry to the RPV Control guideline requires the initiation of a reactor scram if one has not yet been initiated and permits a controlled depressurization of the RPV to proceed in accordance with Step RC/P-3 (depending on the status of other plant parameters). This action reduces core heat and the steam generation rate in the RPV to decay heat levels (assuming the scram is successful), thereby assisting in maintaining plant conditions below the SRV Tail Pipe Level Limit.

A reactor scram is effected indirectly, through entry of the RPV Control guideline, rather than through an explicit direction in the Primary Containment Control guideline to ensure that RPV water level, RPV pressure, and reactor power are properly coordinated following the scram and to avoid potential conflicts with alternate rod insertion strategies in Step RC/Q if the RPV Control guideline is already in use. (Note that Step RC-1 of the RPV Control guideline requires initiation of a reactor scram only if a scram has not yet been initiated.)

An explicit direction to enter the RPV Control guideline must be provided since conditions requiring entry of the Primary Containment Control guideline do not necessarily require entry of the RPV Control guideline. A scram may have therefore not yet been initiated even if suppression pool water level is high.

Consistent with the definition of the phrase cannot be maintained, entry to the RPV Control Guideline can be made as soon as suppression pool water level exceeds the maximum Technical Specification LCO if it is determined that restoration of suppression pool water level to the normal range cannot be achieved. This RPV Control Guideline entry thus provides the guidance necessary for the operator to comply with technical specification action time requirements for achieving depressurized conditions in a controlled cooldown manner when suppression pool water level is above the maximum technical specification LCO.

A break in the RPV may be contributing to the high suppression pool water level condition being addressed in Step SP/L-3.1; water being injected into the RPV may be spilling out a break and accumulating in the suppression pool. Accordingly, injection from sources outside the primary containment is terminated to prevent any further increase in suppression pool water level that may occur through this mechanism. Assuring adequate core cooling takes precedence over terminating injection into the RPV from external sources since additional action can still be taken to prevent SRV system damage and containment failure. Operation of systems used to inject boron or insert control rods need not be terminated if the systems are being used to shut down the reactor.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # 295035 EA2.01 Importance Rating 3.8 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): Alarm Procedure 2.3_R-2 (Rev 14)

(Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR Part 55 Content: 55.41 9 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 65 New 00 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 1 1 4 Multiple Choice Topic Area Description Abnormal Procedures Secondary containment pressure Related Lessons COR0010802 OPS HEATING, VENTILATION AND AIR CONDITIONING Related Objectives COR0010802001090D Given a specific HVAC malfunction, determine the effect on any of the following: Reactor Building pressure COR0010802001110C Describe the HVAC design features and interlocks that provide for the following: Automatic starting and stopping of fans Related References NONE Related Skills (K/A) 295035 EA2.01 Ability to determine and/or interpret the following as they apply to SECONDARY CONTAINMENT HIGH DIFFERENTIAL PRESSURE:(CFR:

41.8 to 41.10) Secondary containment pressure: Plant-Specific (3.8)

QUESTION: 65 The Reactor Building HVAC is in a normal lineup. Outside temperature is 70°F and winds are steady at 5 miles per hour. The running Reactor Building exhaust fan vortex damper closes.

Per Panel R Annunciator R-2 alarm procedure, when does the Reactor Building exhaust and supply fans trip for the present conditions?

a. Immediately with Reactor Building pressure at -0.45 wg or below.

b. After a 45 second time delay with Reactor Building pressure at -0.45 wg or below.

c. Immediately with Reactor Building pressure at -0.15 wg or above.

d. After a 45 second time delay with Reactor Building pressure at -0.15 wg or above.

ANSWER: 65

d. After a 45 second time delay with Reactor Building pressure at -0.15 wg or above.

Explanation:

NOTE: See attached alarm R-2/A-4 Per Alarm Procedure 2.3_R-2 (Rev 14), alarm R-2/A-4, the high secondary containment d/p of

-0.15 wg pressure trips the supply and exhaust fans after a 45 second time delay. With calm wind conditions, all HVAC fans are in their normal control switch configurations (AUTO) and no fan control switches are in RUN. The exhaust fan vortex damper maintains building dP. With a constant supply flow and a lowering exhaust flow, the reactor building pressure rises to the point where the high pressure alarm comes in at -0.15 wg pressure.

Distractors:

a. There is a time delay for tripping fans. A low building dP also trips booster fans.

b. This condition also trips booster fans. With only the supply fan running, the building pressure will rise and the dP will also rise and not go down.

c. There is a time delay for tripping fans on building low dP.

Alarm R-2/A-4 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 1 K/A # 2.1.4 Importance Rating 3.3 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): Conduct of Operations Procedure 2.0.7 (Rev 7)

(Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank # 16476 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 66 16476 00 09/05/2003 06/15/2005 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 1 1 4 Multiple Choice Topic Area Description None Active License Maintenance Related Lessons INT0320101 CNS Administrative Procedures Volume 0, Administrative Procedures (Formal Classroom/Pre-OJT Training)

Related Objectives INT032010100S010C Procedure 2.0.7, Licensed Operator Active/Reactivation/Medical Status Maintenance Program; Discuss the following as described in 2.0.7 Licensed Operator Active/Reactivation/Medical Status Maintenance Program: Maintaining Active Status Related References Procedure 2.0.7, LICENSED OPERATOR ACTIVE/REACTIVATIO N/M ED I CAL STATUS MAI NTENANCE PROGRAM Related Skills (K/A) 2.1.4 Knowledge of individual licensed operator responsibilities related to shift staffing, such as medical requirements, no-solo operation, maintenance of active license status, 10CFR55, etc. (CFR: 41.10 / 43.2) (3.3)

QUESTION: 66 16476 An RO left shift work on 6/3. The RO worked all scheduled workdays this year as BOP until leaving shift. Since leaving shift, the RO performed the following shifts as the BOP:

7/19 worked 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

8/18 worked 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

8/30 worked 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

9/10 worked 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months What is this operator's license status on 9/11?

a. License is active and only stays active with another 12-hour shift before 10/1.

b. License is active and only stays active with another 12-hour shift before 11/1.

c. License became inactive on 6/30.

d. License became inactive on 7/31.

ANSWER: 66 16476

a. License is active and only stays active with another 12-hour shift before 10/1.

Explanation:

NOTE: See attached Procedure 2.0.7 partial.

Maintenance of an active license requires that five 12-hour shifts be stood per calendar quarter. As long as one more 12-hour shift is stood before the end of September (before 10/1/01) the license remains active.

Distractors:

b. Calendar quarter ends September 30, not October 31.

c. License is still active.

d. License is still active.

Procedure 2.0.7 partial ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 1 K/A # 2.1.17 Importance Rating 3.9 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): Procedure 2.3.1 (Rev 60) Section 4 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 67 New 00 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 1 1 4 Multiple Choice Topic Area Description Systems Related Lessons INT032-0103 OPS CNS Admin Procedures Conduct of Ops General Alarm Procedures Related Objectives INT032010300E010B Discuss the following as described in Alarm Procedure 2.3.1, General Alarm Procedure: Alarm acknowledgement Related References NONE Related Skills (K/A) 2.1.17 Ability to make accurate, clear, and concise verbal reports. (CFR: 41.10 / 45.12 /

45.13) (3.9)

QUESTION: 67 The plant is operating at power when the following alarm is received:

A-4/B-6, SERVICE WATER PUMP A TRIP What information is to be relayed to Control Room supervision concerning this alarm? (Assume the alarm is a valid alarm)

a. Read or paraphrase the annunciator descriptor and referenced EOP procedures.

b. Report alarm card automatic actions and referenced Technical Specifications.

c. Read or paraphrase the annunciator descriptor and referenced Technical Specifications.

d. Report the alarm as an unexpected alarm and referenced EOP procedures.

ANSWER: 67

c. Read or paraphrase the annunciator descriptor and referenced Technical Specifications.

Explanation:

NOTE: See attached Procedure 2.3.1 partial.

Procedure 2.3.1 (Rev 60) Section 4 provides alarm acknowledgement requirements.

The alarm is an unexpected alarm. Unexpected alarms are announced by reading or paraphrasing the annunciator descriptor (Step 4.13). If alarm card Reference Section contains reference to TS/TRM/ODAM, then this is communicated to the Shift Manager (Step 4.16)

There are no automatic actions that take place according to the A-4/B-6 alarm card information and no EOP entry conditions are expected to occur for this alarm.

Distractors:

a. There are no EOP procedures referenced on a SW pump trip alarm. Some alarms have EOP procedures referenced.

b. There are no automatic actions associated with this alarm. Automatic actions arent required to be read to control room supervision.

d. The alarm is an unexpected alarm and should be reported. There are no EOP procedures referenced on a SW pump trip alarm. Some alarms have EOP procedures referenced.

Procedure 2.3.1 partial ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 1 K/A # 2.1.42 Importance Rating 2.5 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): Procedure 10.25.1 (rev 26) Section 5, NOTE 1 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 68 New 00 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 4 1 1 4 Multiple Choice Topic Area Description Procedures Bypassing withdrawn control rod refueling interlocks Related Lessons SKL010-01-02 Initial License Self-Study Related Objectives SKL0100102001460D Given conditions, determine actions required per the precautions and limitations of the following procedures: 5) 10.25.1, REFUELING -

CORE REFUELING SUPPORT OPERATIONS Related References Nuclear Performance Procedure 10.25.1, Refueling-Core Refueling Support Operations Related Skills (K/A) 2.1.42 Knowledge of new and spent fuel movement procedures. (CFR: 41.10 / 43.7 /

45.13) (2.5)

QUESTION: 68 Refueling operations are taking place per Procedure 10.25.1, REFUELING-CORE REFUELING SUPPORT OPERATIONS. A control rod is withdrawn and a bypass jumper for refueling interlocks is installed for the RPIS probe buffer card of the withdrawn control rod.

What indication is available to inform the At the Controls operator the control rods refueling interlocks are bypassed?

a. The Panel 9-5 Full Core Display FULL IN indicating light illuminates and the FULL OUT indicating light extinguishes.

b. The Panel 9-5 Full Core Display FULL IN indicating light extinguishes and the FULL OUT indicating light illuminates.

c. Both Panel 9-5 Full Core Display FULL IN and FULL OUT indicating lights extinguish.

d. Both Panel 9-5 Full Core Display FULL IN and FULL OUT indicating lights illuminate.

ANSWER: 68

d. Both Panel 9-5 Full Core Display FULL IN and FULL OUT indicating lights illuminate.

Explanation:

NOTE: See procedure excerpt of Section 5 attached.

Procedure 10.25.1 (rev 26) Section 5, NOTE 1 describes bypassing the refueling interlocks and the indicating light response.

Distractors:

a. The FULL IN light does illuminate but the FULL OUT light illuminates as well.

b. The FULL IN light does illuminate and the FULL OUT indicating light remains illuminated.

c. Both lights illuminate.

Procedure 10.25.1 Section 5

5. CONTROL ROD WITHDRAW/BYPASS (WD/BYP) OPERATIONS NOTE 1 - WD/BYP operations are not allowed during in-core fuel movement. Verify the core is defueled or that fuel moves have been terminated.

NOTE 2-For WD/BYP operations, LCO 3.10.6 applies. Tracking of SR 3.10.6.1 and SR 3.10.6.2 will be done per Procedure 6.LOG.602. SR 3.10.6.3 is not tracked as WD/BYP operations are not performed during in-core fuel movement.

5.1 The following sequence of steps shall be executed for each control rod which will require bypassing of refueling interlocks recording data on Attachment 1:

5.1.1 A Licensed Operator and Reactor Engineering shall verify all fuel has been removed from control cell.

5.1.2 Verify all control rods with cells containing fuel assemblies are fully inserted.

5.1.3 Reactor Operator will withdraw control rod from 00 to 48.

NOTE 1 - Monitor for bulbs burning out in Panel 9-5 Full Core Display, as bypassing of refueling interlocks shall result in FULL lN and FULL OUT lights illuminating. Bulbs should be replaced as needed.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 2 K/A # 2.2.1 Importance Rating 4.5 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): General Operating Procedure 2.1.1 (Rev 164),

(Attach if not previously provided) Step 2.22 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank # 16466 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR Part 55 Content: 55.41 5 &10 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 69 16466 02 03/16/2007 06/15/2005 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 3 1 4 Multiple Choice Topic Area Description Procedures INT032010400A0100, CNS ADMINISTRATIVE PROCEDURES GENERAL OPERATING PROCEDURES Related Lessons INT0320104 CNS Administrative Procedures General Operating Procedures (Startup and Shutdown) Procedures (Formal Classroom/Pre-OJT Training)

Related Objectives INT032010400A0100 Discuss Precautions and Limitations outlined in General Operating Procedure 2.1.1, Startup Procedure.

Related References 2.1.1 Startup Procedure Related Skills (K/A) 2.2.1 Ability to perform pre-startup procedures for the facility, including operating those controls associated with plant equipment that could affect reactivity. (CFR: 41.5 /

41.10 / 43.5 / 43.6 / 45.1) (4.5)

QUESTION: 69 16466 During a reactor startup and heatup after a refueling outage, reactor period was infinity after withdrawing a control rod. The following conditions are present with NO control rod movement for the last two (2) minutes:

The reactor is on range 5 of the IRMs (rising)

Reactor period is +120 seconds (getting shorter)

Reactor coolant temperature is 180°F (rising)

What action is required?

a. Take conservative action and insert a manual scram.

b. Range IRMs as necessary to keep them on scale until the Point of Adding Heat (POAH) is reached.

c. Fully insert control rods per the Emergency Power Reduction section of NPP 10.13 until the reactor is subcritical.

d. Wait and see if reactor period shortens to under 50 seconds, then insert the last withdrawn control rod until period is longer than 50 seconds.

ANSWER: 69 16466

a. Take conservative action and insert a manual scram.

Explanation:

See Attached partial of Procedure 2.1.1 Step 2.22 (Precautions and Limitations Section) of Procedure 2.1.1 (Rev 164) states that conservative action is required whenever an unexpected situation arises with respect to reactivity, criticality, power level, or any other anomalous behavior of reactor core. This conservative action should include rod insertion to reduce power or a reactor scram without hesitation whenever such unanticipated or anomalous behavior is encountered. In this case indicated power is below the POAH yet temperature is rising and even with this negative reactivity feedback, power is rising and period is getting shorter. All of which are indications of a significant anomaly.

Distractors:

b. This action is not conservative, and would just allow the anomalous reactor behavior to continue.

c. This action may be taken when the flow instability region of the Power to Flow Map is entered, but is not appropriate for the conditions given. In addition, at this point in the startup, operation would be below the 80% rod line and the emergency power reduction control rods are not available

d. While this action is correct for reactor period shorter than 50 seconds, it is NOT conservative (or correct) to wait for this value of reactor period to exist before taking corrective actions to stop the power anomaly from continuing.

Procedure 2.1.1 Precautions and Limitations (Partial) 2.21 During plant heatup with a negative moderator temperature coefficient, it is possible that periods of low control rod worth withdrawal or short (< t hour) delays in rod movement may result in temporary subcriticality. ln such cases, control rod withdrawal may continue and criticality re-achieved provided Shift Management and the Reactor Engineer concur. Proper neutron instrumentation response shall be maintained. Notch withdrawal will be used until criticality is again achieved.

2.21.1 A subcritícal reactor is typically indicated by:

2.21.1.1 Neutron count rate continuously lowering without rod insertions and not due to moderator temperature changes.

2.21.1.2 Neutron count rate is below Point of Adding Heat (POAH). The POAH is generally observed on IRM Range 7 or - 0.1% to 1%/o of rated power.

2.21.1.3 Multiple lRMs ranged down at least two (2) ranges.

2.22 Conservative action is required whenever an unexpected situation arises with respect to reactivity, criticality, power level, or any other anomalous behavior of reactor core. This conservative action should include rod insertion to reduce power or a reactor scram without hesitation whenever such unanticipated or anomalous behavior is encountered.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 2 K/A # 2.2.36 Importance Rating 3.1 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): Technical Specification 1.1, Definitions (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 70 New 00 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 2 1 4 Multiple Choice Topic Area Description Tech Specs TS Required system inoperable due to loss of cooling Related Lessons INT007-05-01, OPS Introduction to Technical Specifications Related Objectives INT0070501001030C From memory, define the following terms: OPERABLE Related References Technical Specifications 1.1 Related Skills (K/A) 2.2.36 Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations. (CFR: 41.10 / 43.2 /

45.13) (3.1)

QUESTION: 70 The SE Quad fan coil unit is having its belts replaced by maintenance personnel.

Does the Core Spray subsystem have to be declared inoperable? Why or why not?

a. No; support equipment does not need to be maintained in an operable status to maintain the TS Required Equipment Operable.

b. No; if maintenance can return the support system to a functional state, then the TS required equipment can remain Operable.

c. Yes; any work on a support system that is associated with a Technical Specification piece of equipment affects the operability of that system.

d. Yes; LCO 3.0.6 requires the supported system to be declared inoperable.

ANSWER: 70

c. Yes; any work on a support system that is associated with a Technical Specification piece of equipment affects the operability of that system.

Explanation:

TS definition of OPERABILITY: A system, subsystem, division, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function( s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, division, component, or device to perform its specified safety function( s) are also capable of performing their related support function( s ).

Distractors:

a. LCO 3.0.6 can make an exception on the support/supported function. However the room cooling is required for TS required equipment to perform its safety function so the LCO 3.0.6 exception is not allowed. The supported system must be declared inoperable.

b. The supported system is inoperable due to loss of room cooling. There is no allowance for maintenance to return the support system to a functional state to allow not declaring the supported system inoperable.

d. LCO 3.0.6 makes an allowance to not enter the conditions and required actions of the supported system, but does not require the supported system to be declared inoperable.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 3 K/A # 2.3.11 Importance Rating 3.8 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): EOP 5A (Rev 15) Step SC-1 (Attach if not previously provided) AMP-TB00 (Rev 6) (CNS PSTGs)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR Part 55 Content: 55.41 12 55.43 Comments:

Question more appropriately satisfies 55.41.12 vice 55.41.11.

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 71 New 00 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 2 1 4 Multiple Choice Topic Area Description Emergency Operating Radioactivity release control Procedures Related Lessons INT0080617 OPS FLOWCHART 5A - SECONDARY CONTAINMENT AND RADIOACTIVITY RELEASE CONTROL Related Objectives INT00806170010600 Given plant conditions and EOP flowchart 5A, SECONDARY CONTAINMENT CONTROL and RADIOACTIVITY RELEASE CONTROL, determine required actions.

INT00806170010700 Given plant conditions and EOP flowchart 5A, SECONDARY CONTAINMENT CONTROL and RADIOACTIVITY RELEASE CONTROL, state the reasons for the actions contained in the steps.

Related References AMP-TB00 (Rev 6) CNS PSTGs Related Skills (K/A) 2.3.11 Ability to control radiation releases. (CFR: 41.11 / 43.4 / 45.10) (3.8/4.3)

QUESTION: 71 The plant is operating at power when a leak in the Reactor Building causes a building high radiation alarm. The Rx Bldg HVAC isolates on a valid high radiation signal and 10 minutes later the Reactor Building Vent radiation monitors are all reading 5 mR/hr and steady.

What does EOP 5A, Secondary Containment Control require for building ventilation control?

a. Verify the HVAC isolation valves all closed and then start Standby Gas Treatment system.

b. Install EOP PTMs to override the high radiation signal and then restart Rx Bldg HVAC.

c. Restart the Rx Bldg HVAC and leave Standby Gas Treatment system operating.

d. Verify the radiation levels and then restart Rx Bldg HVAC.

ANSWER: 71

d. Verify the radiation levels and then restart Rx Bldg HVAC.

Explanation:

NOTE: See attached AMP-TB00 (Rev 6) (CNS PSTGs) reasoning for restarting RB HVAC.

EOP 5A (Rev 15) SC-1 second override directs verifying radiation levels are below 10 mR/hr and then restarting Rx Bldg HVAC per Procedure 5.8.20. The HVAC is restarted to ensure secondary containment pressure is maintained negative. With vent radiation levels at 5 mR/hr, there is residual radiation in the building atmosphere. Controlling the building negative ensures an elevated release of radiation. Procedure 5.8.20 has the operator reset the isolation signals which causes the Rx Bldg HVAC to automatically start and SGT to stop without having to manually operate the systems.

Distractors:

a. Verifying the HVAC isolation valves all close is required by Procedure 2.1.22, Group Isolations and not EOP 5A. SGT automatically started so it doesnt need to be manually started.

b. The only time high radiation signals can be overridden is for containment venting per Procedure 5.8.22.

c. The radiation levels must be verified to be below 10 mR/hr before restarting Rx Bldg HVAC. SGT can be left running but EOP 5A does not direct it.

PSTG/SATG Step (Third SC override)

If while executing the following steps:

Rx Bldg HVAC isolates, and,

Rx Bldg exhaust plenum radiation level is below 10 mR/hr (secondary containment HVAC isolation setpoint),

restart Rx Bldg HVAC, defeating high drywell pressure and low RPV water level isolation interlocks if necessary.

Discussion Secondary Containment HVAC is the system normally used to maintain secondary containment temperature and differential pressure within operational limits. If isolated, it is appropriate to restart this system and use it to restore and maintain control of secondary containment temperature and pressure once it has been confirmed that restart will not result in an excessive release of radioactivity to the environment.

Defeating high drywell pressure and low RPV water level isolation interlocks is appropriate, if needed, since application of these isolations to secondary containment HVAC is for the sole purpose of limiting radioactivity release to the environment. Once assurance is provided that excessive release of radioactivity will not occur, these two isolation interlocks become dispensable.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 3 K/A # 2.3.12 Importance Rating 3.2 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): Technical Specifications Section 5.7 (Attach if not previously provided) Procedure 9.EN-RP-108 (Rev 6)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR Part 55 Content: 55.41 12 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 72 New 00 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 1 1 3 Multiple Choice Topic Area Description Administrative INT0320115, (DDE) of 1250 mrem/hour. How should the entrance to this room be posted?

Related Lessons INT0320115 OPS CNS Administrative Procedures Radiation Protection and Chemistry Procedures (Formal Classroom/Pre-OJT Training)

Related Objectives INT0320115H0H010D Discuss the following as described in Radiological Protection Procedure 9.EN-RP-108, Area Posting and Access Control:

RCA/Satellite Area controls and postings Related References TS 5.7 High radiation area 9.EN-RP-108 Area Posting and Access Control Related Skills (K/A) 2.3.12 Knowledge of radiological safety principles pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc. (CFR: 41.12 / 45.9 / 45.10) (3.2)

QUESTION: 72 One of the rooms in the Reactor Building contains an area in which a person could receive a deep dose equivalent (DDE) of 1250 mR/hr.

Why is the entrance required to be posted as a Locked High Radiation Area?

a. To preclude 10CFR100, REACTOR SITE CRITERIA, radiation limits from being exceeded.

b. To preclude 10CFR20, STANDARDS FOR PROTECTION AGAINST RADIATION, radiation limits from being exceeded.

c. To ensure CNS Planned Special Exposure (PSE) limits are not exceeded during an accident.

d. To ensure dose to the control room personnel does not exceed federal limits during a fuel handling accident.

ANSWER: 72

b. To preclude 10CFR20, STANDARDS FOR PROTECTION AGAINST RADIATION, radiation limits from being exceeded.

Explanation:

Per Technical Specifications, section 5.7, an area in which an individual could pick up a DDE in excess of 1000 mR In 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months shall be classified as a locked high radiation area.

Reference:

Procedure 9.EN-RP-108 (Rev 6) for posting requirements.

Distractors:

a. 10CFR100 sets limits for radiation releases from the facility.

c. Planned Special Exposure limits are above the 10CFR20 limits so 10CFR 20 is more limiting.

d. The limit for control room personnel exposure for the fuel handling accident is not connected to the reason for locked high radiation areas.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 4 K/A # 2.4.18 Importance Rating 3.3 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): AMP-TB00 (Rev 6) ( CNS PSTG) Step SP/L-2.1 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank # 19328 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 73 19328 03 03/02/2010 05/23/2010 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 1 1 4 Multiple Choice Topic Area Description Emergency Operating Why HPCI must be PTL & not RCIC on low SP level Procedures Related Lessons INT0080613 OPS EOP FLOWCHART 3A - PRIMARY CONTAINMENT CONTROL Related Objectives INT00806130010900 Explain why HPCI but not RCIC must be secured at a primary containment water level of 11 feet.

Related References 5.8 Emergency Operating Procedures (EOPs)

Related Skills (K/A) 2.4.18 Knowledge of the specific bases for EOPs. (CFR: 41.10 / 43.1 / 45.13) (3.3)

QUESTION: 73 19328 For what reasons must HPCI, but not RCIC, injection be stopped and prevented if Primary Containment water level cannot be maintained above 11 feet?

Below Primary Containment water level of 11 feet

a. the HPCI exhaust sparger will begin to uncover, while the height of the RCIC suction strainer precludes its operation.

b. the HPCI and RCIC exhaust spargers will begin to uncover, but RCIC exhaust pressure does not challenge Primary Containment integrity.

c. the HPCI exhaust sparger will begin to uncover, but the RCIC exhaust sparger is located below the PC level at which Emergency Depressurization is required.

d. the HPCI and RCIC exhaust spargers will begin to uncover, but RCIC exhaust steam is condensed in the Barometric Condenser prior to discharge to the torus.

ANSWER: 73 19328

b. the HPCI and RCIC exhaust spargers will begin to uncover, but RCIC exhaust pressure does not challenge Primary Containment integrity.

Explanation:

NOTE: See attached PSTG step.

If suppression pool water level cannot be maintained above the top of the HPCI exhaust opening (10.99 feet), HPCI must be secured, irrespective of whether the core is adequately cooled, in accordance with PSTG Step SP/L-2.1. Operation with the exhaust exposed would directly pressurize the suppression chamber. No such restriction exists for RCIC operation since the RCIC exhaust flow rate is much smaller and elevated suppression chamber pressure will trip the RCIC turbine before primary containment integrity is challenged.

Distractors:

a. The height of the RCIC suction strainer has a 6 foot vortex limit.

c. RCIC sparger uncovers at the same height as the HPCI sparger.

d. Barometric condenser does not condense exhaust steam. Turbine shaft seal leakage is routed to the Barometric condenser.

PSTG/SATG Step SP/L-2.1 Maintain suppression pool water level above 10.99 ft (elevation of the top of the HPCI exhaust).

If suppression pool water level cannot be maintained above 10.99 ft (elevation of the top of the HPCI exhaust), secure HPCI irrespective of whether adequate core cooling is assured.

Discussion Operation of the HPCI System with its exhaust discharge device not submerged will directly pressurize the suppression chamber. HPCI operation is therefore secured as required to preclude the occurrence of this condition. The consequences of not doing so may extend to failure of the primary containment from overpressurization, and thus HPCI must be secured irrespective of adequate core cooling concerns.

No comparable instruction regarding RCIC operation is provided because:

The exhaust flow rate of RCIC is no greater than the steam generated by decay heat after reactor shutdown. The basis for determining Primary Containment Pressure Limit A (PCPL-A) assumes the operability of a containment vent capable of removing decay heat ten minutes after reactor shutdown. (Refer to Section 16 of this appendix for a detailed discussion of PCPL-A.) Thus, any steam discharged by RCIC into the suppression chamber airspace can be removed through the primary containment vent and will not cause suppression chamber pressure to exceed PCPL-A even if the RCIC exhaust is not submerged.

Elevated suppression chamber pressure will cause the RCIC turbine to trip much sooner than the HPCI turbine. (Refer to the discussion of Caution #4 in Section 4 of this appendix.)

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 4 K/A # 2.4.23 Importance Rating 3.4 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): Procedure 5.8 (Rev 28), Step 3.5 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 74 New 00 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 1 1 4 Multiple Choice Topic Area Description Emergency Procedures Prioritizing procedure implementation during emergencies Related Lessons INT0080601 OPS EOP Overview Related Objectives INT00806010010400 Identify the required characteristics of Emergency Operating Procedures (EOP's) prepared from the BWR EPG/SAGs.

Related References Procedure 5.8, Emergency Operating Procedures Related Skills (K/A) 2.4.23 Knowledge of the bases for prioritizing emergency procedure implementation during emergency operations. (CFR: 41.10 / 43.5 / 45.13) (3.4)

QUESTION: 74 During EOP actions, the order is given to Stop and Prevent all injection into the RPV. What priority is given when performing the order and why?

Priority is given to the

a. systems that are injecting or about to inject because the objective is to lower RPV water level or prevent a power excursion.

b. systems that are injecting or about to inject because RPV level control would be more difficult in the given level band.

c. highest capacity systems because a rapid cooldown could exceed procedural cooldown rate limitations.

d. low pressure systems because they can rapidly be secured with individual control switches.

ANSWER: 74

a. systems that are injecting or about to inject because the objective is to lower RPV water level or prevent a power excursion.

Explanation:

NOTE: See attached Procedure 5.8 partial.

The direction to Stop and Prevent is given during EOP ATWS conditions to either lower RPV water level which lowers reactor power or prior to Emergency Depressurizing so an uncontrolled injection of cold, unborated water does not cause a power excursion which could damage the vessel. When the order is given Procedure 5.8 (Rev 28), Step 3.5 states that priority is given to systems that are injecting or about to inject. The step also allows steps to be performed in any order in the Stop and Prevent section (Steps 3.5.1 and 3.5.2). Prevent injection is directed in different conditions but Stop and Prevent all injection is unique to AWTS only conditions.

Distractors:

b. This is the reason to Prevent injection for RPV level control.

c. There is no priority given based on capacity during an ATWS.

d. There is no priority given to low pressure systems during an ATWS. In a non-ATWS condition, the systems are Prevented form injection if RPV pressure is above their shutoff head and a high drywell pressure initiation signal is keeping them running dead headed.

Procedure 5.8 partial ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 4 K/A # 2.4.27 Importance Rating 3.4 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): Emergency Procedure 5.1INCIDENT (Rev 22)

(Attach if not previously provided) Attachment 1 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 75 New 00 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 2 1 4 Multiple Choice Topic Area Description Emergency Procedures Knowledge of fire in the plant procedures Related Lessons INT032-01-34 OPS CNS Abnormal Procedures (RO) Fire Related Objectives INT0320134H0H0100 Given plant condition(s), and the applicable Abnormal/Emergency Procedure, determine the correct subsequent actions required to mitigate the event(s).

Related References NONE Related Skills (K/A) 2.4.27 Knowledge of fire in the plant procedures. (CFR: 41.10 / 43.5 / 45.13) (3.4)

QUESTION: 75 A fire has been confirmed in the Turbine Building. The CRS has entered Procedure 5.1INCIDENT, SITE EMERGENCY INCIDENT. As the control room operator you are directed to announce the fire to station personnel. What are your actions for these conditions?

Sound the Fire Alarm Pulse Tone for ten (10) seconds and direct the

a. Fire Brigade and Utility Fire Brigade to a specific fire locker designated by the control room operator and the Turbine Building NLO to the scene of the fire.

b. Fire Brigade to a specific fire locker designated by the control room operator and the Utility Fire Brigade and Fire Brigade Leader to the location of the fire.

c. Fire Brigade and Utility Fire Brigade to an exterior entrance closest to the fire and the Incident Commander to the scene of the fire.

d. Fire Brigade to a specific fire locker designated by the control room operator; the Utility Fire Brigade to an exterior entrance closest to the fire and the Turbine Building NLO to the scene of the fire.

ANSWER: 75

d. Fire Brigade to a specific fire locker designated by the control room operator; the Utility Fire Brigade to an exterior entrance closest to the fire and the Turbine Building NLO to the scene of the fire.

Explanation:

NOTE: See attached 5.1INCIDENT, Attachment 1 partial guidance:

Per Emergency Procedure 5.1INCIDENT (Rev 22) Attachment 1 (Control Room Operator), Step 1.2.1 the Fire Alarm Pulse Tone is sounded for 10 seconds and personnel are directed as follows: Fire Brigade to a specific fire locker chosen by control room operator, Utility Fire Brigade to the location of an exterior entrance closest to the fire, chosen by the control room operator, and Building Operator chosen by the control room operator to respond to the fire.

Distractors:

a. Utility fire brigade reports to an entrance closest to the fire as determined by the control room operator.

b. Utility fire brigade reports to an entrance closest to the fire as determined by the control room operator. Fire brigade leader responds to the scene of the fire but is not directed by the control room operator.

c. Fire brigade reports to the fire locker determined by the control room operator. Incident commander responds to the scene of the fire but is not directed by the control room operator.

5.1INCIDENT, ATTACHMENT 1 CONTROL ROOM OPERATOR

1. lf incident is a fire or a red tile annunciator on Control Room Fire Panel:

1.1 Sound Fire Alarm Pulse Tone for ten (10) seconds 1.2 Make following Gaitronics announcement:

1.2.1 ATTENTION, A FIRE HAS BEEN DETECTED lN THE (location of fire).

FIRE BRIGADE RESPOND TO THE (location of fire locker) FIRE EQUIPMENT LOCKER. UTILITY FIRE BRIGADE RESPOND TO (location of exterior entrance closest to fire). (Building Operator)

RESPOND TO (location of fire) AND REPORT TO THE CONTROL ROOM.

ES-401 Site-Specific SRO Written Examination Form ES-401-8 Cover Sheet U.S. Nuclear Regulatory Commission Site-Specific SRO Written Examination Applicant Information Name:

Date: 11-05-2012 Facility/Unit: Cooper Nuclear Station Region: I II III IV Reactor Type: W CE BW GE Start Time: Finish Time:

Instructions Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. To pass the examination you must achieve a final grade of at least 80.00 percent overall, with 70.00 percent or better on the SRO-only items if given in conjunction with the RO exam; SRO-only exams given alone require a final grade of 80.00 percent to pass. You have 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to complete the combined examination, and 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months if you are only taking the SRO portion.

Applicant Certification All work done on this examination is my own. I have neither given nor received aid.

Applicants Signature Results RO/SRO-Only/Total Examination Values 75 / 25 / 100 Points Applicants Scores / / Points Applicants Grade / / Percent

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 209001 A2.09 Importance Rating 3.3 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): AMP-TB00 (Rev 6) Caution 3 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: EOP Graph 4 vortex limits Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date S1 New 00 Licensed RO: N 76 Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 2 1 4 Multiple Choice Topic Area Description EOPs Low SP level consequences on CS operation Related Lessons INT0080613 EOP 3A PRIMARY CONTAINMENT CONTROL INT0080609 EOP 1A, RPV CONTROL, RPV LEVEL Related Objectives INT0080613001040C State the basis for primary containment control actions as they apply to the following: Graphs reference on Flowchart 3A INT00806130011200 Given plant conditions and EOP flowchart 3A, PRIMARY CONTAINMENT CONTROL, state the reasons for the actions contained in the steps.

INT00806090011100 Given plant conditions and EOP flowchart 1A, RPV CONTROL, determine required actions.

Related References NONE Related Skills (K/A) 209001 A2.09 Ability to (a) predict the impacts of the following on the LOW PRESSURE CORE SPRAY SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5 / 45.6) Low suppression pool level (3.3)

QUESTION: S 1 76 Due to a leak in the suppression pool the reactor is scrammed and the following conditions are present:

All control rods are fully inserted.

Suppression pool is 4.5 feet and lowering 0.1 ft every 5 minutes.

CS Pump A is injecting and controlling RPV level steady at +10 inches.

Five minutes later, CS Pump A indicated flow and discharge pressure start oscillating full scale of their instruments.

What is the consequence of continuing with the given conditions?

What procedure guidance is used to mitigate the consequence?

a. CS Pump A is becoming air bound.

EOP 3A directs the use of HPCI to maintain RPV level.

b. CS Pump A vortex limits are being exceeded.

EOP 3A directs the use of Main Condensate to maintain RPV level.

c. CS Pump A is becoming air bound.

EOP 1A directs the use of HPCI to maintain RPV level.

d. CS Pump A vortex limits are being exceeded.

EOP 1A directs the use of Main Condensate to maintain RPV level.

ANSWER: S 1 76

d. CS Pump A vortex limits are being exceeded.

EOP 1A directs the use of Main Condensate to maintain RPV level.

Explanation:

NOTE: See attached AMP-TB00 (Rev 6) Caution 3 basis.

The CS vortex limits are exceeded when SP level drops below 4.5 ft. Caution 3 warns the operator that exceeding vortex limits may result in equipment damage. All the control rods inserted on the scram so it can be assumed no energy was added to the suppression pool. As suppression pool level lowered, the RPV was emergency depressurized and it can be assumed the suppression pool temperature rose to around 110°F. At these temperatures, NPSH limits would not be a concern for CS pump operation. The vortex limits are the consequence. To resolve equipment damage, main condensate (a Table 3 injection system listed in EOP 1A) can be utilized for injection as it takes suction external from primary containment. HPCI is not a viable injection source as it would have been prevented from operating at a suppression pool level of 11 feet due to exhaust sparger uncovering. With suppression pool at 4.4 feet as stated in the stem, the CS suction piping has not yet uncovered so air binding is not a consequence.

Distractors:

a. CS does not become air bound at this SP level but it does when the suction piping becomes uncovered at 4 ft. EOP 3A directs use of HPCI for suppression pool makeup but not RPV injection.

b. CS vortex limits are exceeded at 4.5 ft. EOP 3A directs the use of main condensate for as a suction source for RHR for drywell sprays.

c. CS does not become air bound at this SP level but it does when the suction piping becomes uncovered at 4 ft. EOP 1A does direct the use of HPCI but it is prevented from use by EOP 3A when its exhaust sparger becomes uncovered.

PSTG/SATG Step Operation of HPCI, RCIC, Core Spray, or RHR with suction from the suppression pool and pump flow above the NPSH or vortex limit may result in equipment damage.

Discussion The NPSH (Net Positive Suction Head) limits are defined to be the highest suppression pool temperature which provides adequate net positive suction head for pumps taking suction on the pool. The NPSH Limits are functions of pump flow and suppression chamber overpressure (airspace pressure plus the hydrostatic head of water over the pump suction). It is utilized to preclude pump damage from cavitation.

The vortex limits are defined to be the lowest suppression pool water level above which air entrainment is not expected to occur in pumps taking suction on the pool. These levels are functions of pump flow. Exceeding the limits can lead to air entrainment at the pump suction strainers.

Refer to Section 16 of this appendix for detailed discussions of the NPSH and vortex limits.

The NPSH and vortex limits are addressed through a caution for the following reasons:

It is difficult to define in advance exactly when the limits should be observed and when pumps should be operated outside the limits.

Pumps to which the limits apply are used in more than one parameter control path. RHR pumps, for example, may be used in PSTG Steps RC/L, SP/T, DW/T, and PC/P. HPCI and RCIC may be used in both PSTG Steps RC/L and RC/P. Authorizing operation of the pumps outside NPSH and vortex limits in one path may conflict with instructions in another path where flow would normally be controlled below the limits.

Pump characteristics, and the shape of the NPSH and vortex limit curves, vary from system to system. If a limit is relatively flat, throttling pump flow will be of little benefit; the operator can only choose whether or not to operate the pump.

Note that the x-axis flow through the pump suction will be greater than indicated system flow if the minimum flow valve is open.

Where references to this caution occur, the identified systems should be operated within the NPSH and vortex limits if possible. If the situation warrants, however, the limits may be exceeded. A judgment as to whether a pump should be operated beyond its limits in a particular event should consider such factors as:

The availability of other systems

The current trend of plant parameters

The anticipated time such operation will be required

The degree to which the limit will be exceeded

The sensitivity of the pump to operation beyond the limit

The consequences of not operating the pump beyond the limit Immediate and catastrophic failure is not expected if a pump is operated beyond the NPSH or vortex limit. The undesirable consequences of uncovering the reactor core could thus outweigh the risk of equipment damage.

Caution #3 is referenced in PSTG Steps RC/L-2, RC/P-2, SP/T, C1-3, C2-1.3, C2-2, C3-2, C4.2.3, C4.3.2, C5-5, and C5-5.2, the overrides at the beginning of the Primary Containment Control PSTG, and in SATG Steps RC/F-1 through RC/F-6, RC/P-2, RC/P-3 and SP/T.

PSTG/SATG Step RC/L-2 Restore and maintain RPV water level between 3 in. (low level scram setpoint (3 in.) or shutdown cooling RPV water level interlock (3 in.),

whichever is higher) and 54 in. (high level trip setpoint) with one or more of the following systems:

MC/RF

CRD

RCIC with suction from ECST if available, defeating low RPV pressure and high area temperature isolation interlocks if necessary.

HPCI with suction from ECST if available, defeating high area temperature isolation interlocks and high suppression pool water level suction transfer logic if necessary.

CS

LPCI with injection through the HXs as soon as possible.

Discussion Step RC/L-2 defines the preferred range in which RPV water level should be established and maintained, and specifies the preferred systems to use in doing so. Maintaining RPV water level below the high end of the identified control band preserves the availability of steam-driven equipment (HPCI, RCIC and feedwater). Maintaining RPV water level above the low end of the identified control band permits the scram to be reset (barring the existence of other scram signals), and allows the use of the normal shutdown cooling system to establish and maintain cold shutdown conditions.

The wide RPV water level control band permitted by this step is sufficient to assure adequate core cooling yet avoid unwarranted demands on an operators attention. If unnecessarily constrained within narrower limits, an operator may be less effective in performing concurrent duties.

Most of the listed systems start or operate automatically, and in many cases an operator need only verify system lineups when this step is reached. If, for example, RPV water level is low due to shrink following a turbine trip, the feedwater system is capable of restoring water level automatically. If a break occurs with the RPV at high pressure, the HPCI and RCIC are designed to automatically inject. Note that Step RC/L-2 may also be reached when a high RPV water level condition exists, recovery from the RPV flooding evolution (PSTG Contingency #4) for example. In this situation, restoring RPV water level requires that water level be lowered through appropriate control of the listed systems.

Step RC/L-2 is purposely written to provide an operator with the flexibility to control available systems as most appropriate for existing plant conditions. As a transient progresses, the most effective modes of system operation may change. (For example: all ECCS pumps may inject following a large pipe break, but a single pump may be sufficient to maintain RPV water level once it is restored within the specified range.) Continued manual control and adjustment of system lineups and injection flows may thus be required in order to remain within the preferred RPV water level control band.

Included in Step RC/L-2 is a list of motor-driven and steam-driven injection systems appropriate for controlling RPV water level. The listed systems include both those used for RPV water level control during normal plant operations at power and those categorized as emergency makeup.

Since symptom-oriented guidelines must address a full spectrum of initial plant conditions and postulated transients, this step does not unconditionally prioritize use of one system over another.

The preferred suction source for HPCI and RCIC is always the emergency condensate storage tank (ECST). While the ECST is a smaller volume than the suppression pool, it provides higher quality water, is at a higher elevation, and is not affected by containment heatup or steam discharges from the RPV. NPSH, vortex, and component cooling limitations are thus less likely to be challenged.

The ECST should be refilled as water is depleted to maintain the suction source. If the ECST is unavailable, however, HPCI and RCIC suctions may be aligned to the suppression pool.

The HPCI and RCIC high area temperature isolations and HPCI high suppression pool water level transfer logic and the RCIC low RPV pressure isolation may be defeated, if necessary, to permit continued use of the systems:

Defeating the HPCI high suppression pool water level transfer logic allows continued use of the preferred suction. (RCIC is not equipped with high suppression pool water level suction transfer logic.)

Defeating the HPCI and RCIC high area temperature isolations permits continued use of the systems following loss of area coolers or ventilation, such as might occur during a station blackout. The high temperature isolations provide protection from primary containment leakage. If high area temperatures are due to loss of cooling or ventilation, the isolations are unnecessary. If a leak does exist, the Secondary Containment Control PSTG specifies the appropriate actions and provides a backup for the isolations: if the systems are not needed for core cooling, they are isolated; if they are needed, they are kept in service.

The RCIC low RPV pressure isolation may be defeated since RCIC can provide some flow as long as pressure is above the value at which the turbine will stall. (Similar authorization is not provided for HPCI since its relatively high steam demand prevents sustained operation at low pressure.)

If suppression pool water level cannot be maintained above the top of the HPCI exhaust opening, HPCI must be secured, irrespective of whether the core is adequately cooled, in accordance with PSTG Step SP/L-2.1. Operation with the exhaust exposed would directly pressurize the suppression chamber. No such restriction exists for RCIC operation since the RCIC exhaust flowrate is much smaller and elevated suppression chamber pressure will trip the RCIC turbine before primary containment integrity is challenged.

Directing flow through the RHR heat exchangers as soon as possible promotes rapid removal of decay heat from the primary containment, thus minimizing suppression pool heatup and prolonging the availability of the suppression pool as a heat sink. Although the heat exchangers

introduce an additional pressure drop across the RHR system, the resulting decrease in flow rate is small. Thermal stresses which result from injecting through the heat exchangers have been considered in the design of the RHR System and are within acceptable limits. The phrase with suction through HXs does not require that all injection must pass through the heat exchangers. If the heat exchanger bypass can be throttled, heat exchanger flow may be controlled to within heat exchanger flow limits.

As used in this step, the phrase as soon as possible means the earliest practicable time within the constraints imposed by system conditions, valve control logic, and concurrently required operator actions. In most cases, injecting through the heat exchangers may be accomplished by simply closing the RHR heat exchanger bypass valve and starting appropriate service water pumps following expiration of a time delay interlock. Detailed instructions, if necessary, are relegated to plant-specific procedures.

Three cautions apply to operation of injection systems in this stepCautions #2, #3, and #4.

Caution #2 applies only to operation of RCIC. The caution reminds the operator that elevated suppression chamber pressure may trip the RCIC turbine on high turbine exhaust pressure.

Caution #3 applies to operation of RCIC, HPCI, CS, and RHR. NPSH and vortex limits for these systems should be observed, if possible, but may be exceeded if necessary to establish and maintain adequate core cooling.

Caution #4 applies to operation of both RCIC and HPCI. Since lube and control oil for these systems are cooled by the water being pumped, operation with high suction temperature may result in bearing damage or loss of control capability.

Cautions applicable to operation of the listed injection systems are referenced next to the associated action statement in the first paragraph of Step RC/L-2.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 215004 A2.03 Importance Rating 3.3 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): Instrument Operating Procedure 4.1.1 (Rev 21)

(Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date S2 New 00 Licensed RO: N 77 Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 3 1 4 Multiple Choice Topic Area Description Systems Mitigate consequences of a stuck SRM detector Related Lessons SKL0124230 Source Range Monitoring System Related Objectives SKL012423000B0500 Given an alarm associated with the Source Range Monitoring system, verify the alarm setpoints and operate the system controls identified in the appropriate alarm response procedures.

Related References Procedure 4.1.1, SRM System Related Skills (K/A) 215004 A2.03 Ability to (a) predict the impacts of the following on the SOURCE RANGE MONITOR (SRM) SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5 / 45.6) Stuck Detector (3.3)

QUESTION: S 2 77 The RO reports that SRM B detector will not withdraw and alarm 9-5-1/F-7, SRM UPSCALE/INOP, has been received for the SRM. Per alarm card guidance SRM B has been bypassed on Panel 9-5. Troubleshooting activities have determined the detector drive cannot be repaired in a timely manner.

What actions are required to mitigate this condition, and what procedure provides the guidance?

a. The CRS directs the power rise to continue and then the SRM B drawer supply fuses are removed after the Reactor MODE switch is in RUN in accordance with Procedure 2.1.1, STARTUP.

b. The CRS directs the power rise to continue and then the SRM B drawer supply fuses are removed after the Reactor MODE switch is in RUN in accordance with Procedure 4.1.1, SRM System.

c. The GMPO must authorize the startup to continue and the SRM B drawer supply fuses are removed after the IRMs are on Range 3 or above in accordance with Procedure 2.1.1, STARTUP.

d. The GMPO must authorize the startup to continue and the SRM B drawer supply fuses are removed after the IRMs are on Range 3 or above in accordance with Procedure 4.1.1, SRM System.

ANSWER: S 2 77

b. The CRS directs the power rise to continue and then the SRM B drawer supply fuses are removed after the Reactor MODE switch is in RUN in accordance with Procedure 4.1.1, SRM System.

Explanation:

NOTE: See attached Section 6 of Procedure 4.1.1 for stuck SRM detector.

Per Instrument Operating Procedure 4.1.1 (Rev 21) Steps 6.2 through 6.6.2.2, the CRS or SM must authorize any power rise and the affected SRM drawer supply fuses are removed after the Reactor Mode switch is in the RUN position.

a. Procedure 2.1.1 directs withdrawing the SRMs and directs Procedure 4.1.1 be used if an SRM detector cannot be withdrawn.

c The CRS or SM authorizes the power rise to continue. The GMPO does authorize some operator activities. The SRMs can be fully withdrawn only after all IRMS are on Range 3 or above. Procedure 2.1.1 directs withdrawing the SRMs and directs Procedure 4.1.1 be used if an SRM detector cannot be withdrawn.

d. The CRS or SM authorizes the power rise to continue. The GMPO does authorize some operator activities. The SRMs can be fully withdrawn only after all IRMS are on Range 3 or above.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 218000 2.4.21 Importance Rating 4.6 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): Tech Specs 3.3.5.1 and Table 3.3.5.1-1 Function 4b (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: TS LCO 3.5.1 and 3.3.5.1 partial Learning Objective: See Attached (As available)

Question Source: Bank # 25336 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR Part 55 Content: 55.41 55.43 2 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date S3 16522 01 04/19/2000 06/15/2005 Licensed RO: N 78 Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 1 1 4 Multiple Choice Topic Area Description Technical Specifications, ADS pushbutton broken, how long to fix ODAM, TRM Related Lessons INT0070504 CNS Tech. Spec. 3.3, Instrumentation COR0021602 OPS NUCLEAR PRESSURE RELIEF Related Objectives COR0021602001060A Briefly describe the following concepts as they apply to NPR: ADS logic operation INT00705040010300 Given a set of plant conditions that constitutes non-compliance with a Section 3.3 LCO, determine the ACTIONS that are required.

Related References 3.5.1 ECCS Operating Related Skills (K/A) 218000 Automatic Depressurization System 2.4.21 Knowledge of the parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc. (CFR:

41.7 / 43.5 / 45.12) (4.0/4.6)

QUESTION: S 3 78 16522 The ADS LOGIC A TMR pushbutton failed, causing its contacts to remain open (electrically equivalent to holding the pushbutton in the depressed position). All other functions of the ADS system are NOT affected.

What action is required by Technical Specifications?

a. Reactor pressure must be below 150 psig within 37 hours4.282407e-4 days 0.0103 hours 6.117725e-5 weeks 1.40785e-5 months .

b. Restore the channel to OPERABLE status within 8 days.

c. Reactor pressure must be below 150 psig within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

d. Restore the channel to OPERABLE status within 14 days.

ANSWER: S 3 78 16522

b. Restore the channel to OPERABLE status within 8 days.

Explanation:

NOTE: See attached figures of ADS logic for A logic timer and typical ADS logic.

Timer Reset Pushbutton A inoperable still allows ADS to initiate from B trip system. The A trip system is inoperable and must be restored to OPERABLE within 8 days since HPCI and RCIC are OPERABLE. It is NOT necessary to perform a plant shutdown at this time because TS 3.3.5.1, Required Action G.1, is based on losing the trip capability in BOTH trip systems.

With Timer A Pushbutton contacts open, relays 2E-K6A and 2E-K7A will not energize on an automatic system initiation. Relays 2E-K6B and 2E-K7B of logic B will energize on an automatic system initiation and these relays have contacts in all ADS valves. With logic B operable ADS can still perform its safety function.

GE Drawing 791E253 (Rev N28), Sheet 2 Pushbutton depressed for Logic A de-energizes relays 2E-K6A and 2E-K7A. See next page for typical SRV logic. B logic arrangement is the same.

GE Drawing 791E253 (Rev N28), Sheet 2 This logic is typical for all ADS valves.

2E-K6A & 7A contacts wont close.

2E-K6B & 7B contacts will close and open valves.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 262002 2.4.9 Importance Rating 4.2 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): Procedure 2.4RXPWR (Rev 4)

(Attach if not previously provided) Procedure 5.3NBPP (Rev12)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date S4 New 00 Licensed RO: N 79 Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 3 1 4 Multiple Choice Topic Area Description Systems Loss of NBPP at low power and reactor scram due to control rod drop.

Related Lessons INT0320123 CNS Abnormal Procedures (RO) Reactivity INT0060115 Accident Analysis Related Objectives INT0320123H0H0100 Given plant condition(s), determine from memory if a manual reactor scram or an emergency shutdown from power is required due to the event(s).

INT00601150010200 Given an accident and a set of conditions, select those conditions that would tend to make the consequences of the given accident more severe.

Related References 2.4RXPWR, REACTOR POWER ANOMALIES Related Skills (K/A) 262002 Uninterruptable Power Supply (A.C. /D.C.)

2.4.9 Knowledge of low power/shutdown implications in accident (e.g., loss of coolant accident or loss of residual heat removal) mitigation strategies. (CFR: 41.10 / 43.5

/ 45.13) (4.2)

QUESTION: S 4 79 Control rod withdrawal is being performed and reactor power is 4%. The Panel 9-5 Full Core Display indicating lights all go out. One minute later a control rod drop event occurs which causes LPRM upscale alarms to be received for LPRMs 28-21, 20-21 and 28-13.

What procedures must the CRS enter to mitigate the transient and what operator actions are required?

Enter Procedures

a. 2.4RPIS and 2.4RXPWR and select the control rod and insert it using the emergency-in switch.

b. 5.3NBPP and 2.4CRD and select the control rod and insert it using the emergency-in switch.

c. 2.4RPIS and 2.4RXPWR and scram the reactor.

d. 5.3NBPP and 2.4CRD and scram the reactor.

ANSWER: S 4 79

c. 2.4RPIS and 2.4RXPWR and scram the reactor.

Explanation:

NOTE: See attached partial of Attachment 1 for Procedure 2.4RXPWR and Entry Conditions for 2.4RPIS.

The loss of RPIS causes a rod select block and the Full Core Display indicating lights. Control rods cannot be selected and inserted using the emergency-in switch. With the loss of full core display indicating lights and entry into Abnormal Procedure 2.4RPIS must be made. Upscale LPRM alarms indicate the LPRMs indicate above 100 watts/cm2 and Abnormal Procedure 2.4RXPWR, REACTOR POWER ANOMALIES, contains a note to assume a control rod drop accident has occurred. The procedure directs the reactor scrammed if two or more LPRMs indicate this high. Emergency Procedure 5.3NBPP does not direct a reactor scram but is a possibility to be entered due to the loss of the NBPP panel and the procedure lists RPIS and Rod Select Matrix as one loads being lost. There isnt enough information in the question stem to verify that NBPP power is lost. Procedure 2.4RXPWR, directs unexplained power rises be mitigated by reducing power. Reactor power cannot be lowered by control rod insertion due to loss of control rod select power. With reactor power at 4%, Reactor Recirculation pumps are at minimum so they cannot be used to lower power.

Distractors:

a. 2.4RPIS is required to be entered due to the loss of indication. 2.4RXPWR is required to be entered due to an unexpected reactor power rise. 2.4RXPWR directs lowering reactor power but the only method available is to scram.

b. 5.3AC120 is entered on loss of 120VAC instrument power so it is a reasonable procedure to enter. This procedure does not require entry because the power loss to the Panel 9-5 instruments is not one of the power supplies covered by 5.3AC120. Because of the control rod drop, Procedure 2.4CRD is a reasonable procedure to enter. Both procedures contain guidance for lowering reactor power under certain conditions.

d. 5.3AC120 is entered on loss of 120VAC instrument power so it is a reasonable procedure to enter. This procedure does not require entry because the power loss to the Panel 9-5 instruments is not one of the power supplies covered by 5.3AC120. Because of the control rod drop, Procedure 2.4CRD is a reasonable procedure to enter. This procedure does not require entry on control rod drop symptoms. Both procedures have the requirement to manually scram the reactor under certain conditions.

PROCEDURE 2.4RXPWR ATTACHMENT 1

1. UNEXPLAINED RISE IN REACTOR POWER 1.1 If LPRM upscale annunciator alarms, select control rod(s) so alarming LPRM(s) can be observed on four rod display:

1.1.1 If two or more LPRMs read 100 watts/cm2, SCRAM and concurrently enter Procedure 2.1.5.

NOTE - Assume a rod drop accident has occurred.

1.1.1.1 Contact Reactor Engineering for assistance.

1.1.1.2 Check off-gas monitors for entry conditions to Procedure 2.4OG.

PROCEDURE 2.4RPIS partial

2. ENTRY CONDITIONS 2.1 Loss of some or all control rod position indication.

3. AUTOMATIC ACTIONS 3.1 Rod select block.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 300000 A2.01 Importance Rating 2.8 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): Emergency Procedure 5.2AIR (Rev 18), Steps 4.2 and (Attach if not previously provided) 4.2.1 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank # 24771 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date S5 24771 00 07/14/2009 05/23/2010 Licensed RO: N 80 Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 2 1 2 Multiple Choice Topic Area Description Abnormal/Emergency Lowering IA mitigating actions due to filter plugging Procedures Related Lessons COR0011702 OPS Plant Air Related Objectives COR0011702001070A Given a specific Plant Air system malfunction, determine the effect on any of the following: Plant Operation Related References PR 5.2AIR Related Skills (K/A) 300000 A2.01 Ability to (a) predict the impacts of the following on the INSTRUMENT AIR SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation: (CFR:

41.5 / 45.6) Air dryer and filter malfunctions (2.8)

QUESTION: S 5 80 24771 The plant is operating at power with the following conditions:

The Annunciator AIR DRYER FILTER HIGH D/P (A-4/F-4) is received.

Instrument air pressure is 97 psig and slowly lowering Select the action necessary to mitigate the lowering Instrument Air pressure and the procedure that must be entered to perform this action:

a. OPEN IA-MO-80 NON CRITICAL INSTRUMENT AIR ISOLATION in accordance with Alarm Procedure for AIR DRYER PANEL FILTER HIGH D/P (A-4/F-4).

b. OPEN SA-MO-81 SA TO IA CROSSTIE in accordance with Alarm Procedure for AIR DRYER PANEL FILTER HIGH D/P (A-4/F-4).

c. OPEN IA-MO-80 NON CRITICAL INSTRUMENT AIR ISOLATION in accordance with 5.2AIR LOSS OF INSTRUMENT AIR.

d. OPEN SA-MO-81 SA TO IA CROSSTIE in accordance with 5.2AIR LOSS OF INSTRUMENT AIR.

ANSWER: S 5 80 24771

d. OPEN SA-MO-81 SA TO IA CROSSTIE in accordance with 5.2AIR LOSS OF INSTRUMENT AIR.

Explanation:

See attached partial of Procedure 5.2AIR Emergency Procedure 5.2AIR (Rev 18), Steps 4.2 and 4.2.1 states the following: If air drying/filtering components at fault: Open SA-MO-81, SA TO IA CROSSTIE (PANEL A).

Distractors:

a. IA-80 is closed to isolate the non-critical instrument air header and preserve instrument air in the reliable air header. The valve is normally open. Alarm A-4/F-4 procedure guidance does not direct operating IA-80.

b. IA-81 is opened to bypass faulty air dryers but Procedure 5.2AIR provides the guidance.

The alarm A-4/F-4 procedure directs placing standby dryer and filter in service. There isnt enough time with instrument air pressure lowering.

c. IA-80 is closed to isolate the non-critical instrument air header and preserve instrument air in the reliable air header. The valve is normally open. Procedure 5.2AIR provides direction to close the valve on loss of instrument air.

Procedure 5.2AIR excerpt.

4. SUBSEQUENT OPERATOR ACTIONS 4.1 If more than one rod is drifting, SCRAM and concurrently enter Procedure 2.1.5.

4.2 If air drying/filtering components at fault:

4.2.1 Open SA-MO-81, SA TO IA CROSSTIE (PANEL A).

4.2.2 Place standby dryer and filters in service per Procedure 2.2.59.

4.2.3 If necessary, manually bypass any obstructed component(s).

4.2.4 When dryer and filter flow returned to service, close SA-MO-81.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 216000 A2.12 Importance Rating 2.9 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): Procedure 2.4RXLVL (Rev 25)

(Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date S6 New 00 Licensed RO: N 81 Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 2 1 4 Multiple Choice Topic Area Description Systems Excess flow check valve closure effects on NBI Related Lessons COR0021502 OPS Nuclear Boiler Instrumentation Related Objectives COR0021502001040D Briefly describe the following concepts as they apply to NBI: Vessel DP measurement COR0021502001060I Given a specific NBI malfunction, determine effect on any of the following: Vessel level monitoring Related References Abnormal Procedure 2.4RXLVL Related Skills (K/A) 216000 A2.12 Ability to (a) predict the impacts of the 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: (CFR: 41.5 / 45.6) Instrument isolation valve closures (2.9)

QUESTION: S 6 81 The plant is at operating conditions when the excess flow check valve seats that serves the low pressure side of the level transmitter that serves RFC-LI-94B on Panel 9-5.

One minute later RPV level lowers 5 inches for unknown reasons. How is RFC-LI-94B affected, and what procedure must be entered to correct the condition?

RFC-LI-94B indicates 5 inches

a. higher than RFC-LI-94A and RFC-LI-94C. The alarm procedure for 9-5-2/G-4 RVLC SYSTEM TROUBLE directs bypassing RFC-LI-94A and RFC-LI-94C.

b. lower than RFC-LI-94A and RFC-LI-94C. The alarm procedure for 9-5-2/G-4 RVLC SYSTEM TROUBLE directs bypassing RFC-LI-94B.

c. higher than RFC-LI-94A and RFC-LI-94C. Abnormal Procedure 2.4RXLVL, RPV WATER LEVEL CONTROL TROUBLE directs bypassing RFC-LI-94B.

d. lower than RFC-LI-94A and RFC-LI-94C. Abnormal Procedure 2.4RXLVL, RPV WATER LEVEL CONTROL TROUBLE directs bypassing RFC-LI-94A and RFC-LI-94C.

ANSWER: S 6 81

c. higher than RFC-LI-94A and RFC-LI-94C. Abnormal Procedure 2.4RXLVL, RPV WATER LEVEL CONTROL TROUBLE directs bypassing RFC-LI-94B.

Explanation:

NOTE: See attached B&R P&ID Drawing 2026 and excerpt from Procedure 4.6.1.

The excess flow check valve closing isolates the variable leg of NBI-LT-52B which inputs to RFC-LI-94B. The instrument low pressure legs pressure will not change as RPV level is lowered so the dP sensed will not change. The high pressure side (reference leg) does not change so indicated level on the instrument does not change. RFC-LI-94A & C are not effected by the valve closure as they have a different variable leg tap and reference let. As RPV water level lowers 5 inches, RFC-LI-94A & C lower to indicate the new level. The result is the B instrument indicating 5 inches higher than the A & C instruments. The abnormal procedure 2.4RXLVL (Rev 25) must be entered because the RPV level change is unexpected. The procedure directs validating the A and C instruments trending opposite the B instrument, which is occurring. The procedure directs comparing RPV level with the steam nozzle instrument which is off a different reference leg altogether and follows the level change. The procedure directs bypassing the higher level instrument if it isnt trending with the steam nozzle, which it isnt. Alarm 9-5-2/G-4, RVLC SYSTEM TROUBLE only comes in if the level disparity is 8 inches or greater. With a level disparity of 5 inches this alarm would not be present. The alarm procedure directs entering 2.4RXLVL.

Distractors:

a. RFC-LI-94B does indicate 5 inches higher but the alarm procedure only directs entering the instrument procedure for RVLC alarms.

b. RFC-LI-94B does not indicate 5 inches lower and the alarm procedure only directs entering the instrument procedure for RVLC alarms.

d. RFC-LI-94B does indicate 5 inches higher but 2.4RXLVL directs bypassing the higher indicators.

B&R P&ID Drawing 2026 (Rev N65), Sheet 1 Reference leg Excess flow check valve closes on variable leg. Low Pressure leg NBI-LT-52A inputs to RFC-LI-94B Below is from Procedure 4.6.1, Attachment 2

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 259001 G2.4.3 Importance Rating 3.9 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): Tech Spec LCO 3.3.3.1 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: TS LCO 3.3.3.1 and Table 3.3.3.1-1 Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date S7 New 00 Licensed RO: N 82 Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 2 1 4 Multiple Choice Topic Area Description Tech Specs Feedwater and PAM Related Lessons INT0070504 CNS Tech. Spec. 3.3, Instrumentation Related Objectives INT00705040010300 Given a set of plant conditions that constitutes non-compliance with a Section 3.3 LCO, determine the ACTIONS that are required.

Related References NONE Related Skills (K/A) 295001 Reactor Feedwater G 2.4.3 Ability to identify post-accident instrumentation (CFR: 41.6 / 45.4) (3.9)

QUESTION: S 7 82 The plant is at rated conditions and NBI-LI-92, STEAM NOZZLE LEVEL, is indicating upscale and all other control room level instrumentation are indicating normal.

What are the required Technical Specification actions?

Declare NBI-LI-92 inoperable and

a. be in MODE 3 in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b. restore the channel to an operable status in 7 days.

c. immediately initiate action to submit a report within the following 14 days.

d. restore the channel to an OPERABLE status in 30 days.

ANSWER: S 7 82

b. restore the channel to an operable status in 7 days.

Explanation:

NOTE: See attached LCO 3.3.3.1 and Table 3.3.3.1-1 The steam nozzle instrument is used as a comparison to the narrow range and wide range instruments for decision making on the control of the RFPs. Therefore, the steam nozzle instrument is indirectly related to the feedwater system. Tech Spec LCO 3.3.3.1, Table 3.3.3.1-1 Function 2c is for the steam nozzle instrument. The required number of channels is 1 and there is only one steam nozzle instrument. Condition C is applicable and the required action is to restore it to an OPERABLE status within 7 days.

Distractors:

a. This is the required action if the Condition A Completion Time is not met.

d. 30 days is for all the Function 2a and 2b level instruments.

c. This is the required action if the Condition C Completion Time is not met.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 290002 G2.4.46 Importance Rating 4.2 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): 2.4RXPWR (Rev 4)

(Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date S8 New 00 Licensed RO: N 83 Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 3 1 4 Multiple Choice Topic Area Description Systems Jet pump failure requires rapid shutdown.

Related Lessons INT0320123, CNS Abnormal Procedures (RO) Reactivity Related Objectives INT0320123F0F0100 Given plant condition(s), determine from memory the appropriate Abnormal/Emergency Procedure(s) to be utilized to mitigate the event(s).

Related References NONE Related Skills (K/A) 290002 Reactor Vessel Internals G2.4.46 Ability to verify that the alarms are consistent with the plant conditions.

(CFR: 41.10 / 43.5 / 45.3 / 45.12) (4.2)

QUESTION: S 8 83 The plant is operating at rated conditions and SP 6.CRD.301, WITHDRAWN CONTROL ROD OPERABILITY IST TEST is ongoing. A control rod is selected but has not yet been moved.

The following alarms are received:

9-5-1/A-4, ROD WITHDRAWAL BLOCK

9-5-1/E-4, RBM DOWNSCALE The RO reports the following:

Reactor power lowered by 9%

Core flow rose by 4 mlbm/hr

Core plate dP lowered by 7 psid The BOP operator determines a displaced jet pump/mixer exists on Jet pumps 5 and 6.

What action is required by procedure?

What procedure is providing the guidance?

a. Shut down per Procedure 2.1.4.1, RAPID SHUTDOWN.

2.4RXPWR, REACTOR POWER ANOMALIES.

b. Perform Rapid Power Reduction.

Procedure 2.1.5, REACTOR SCRAM.

c. Shut down per Procedure 2.1.4.1, RAPID SHUTDOWN.

Procedure 2.1.5, REACTOR SCRAM.

d. Perform Rapid Power Reduction.

2.4RXPWR, REACTOR POWER ANOMALIES.

ANSWER: S 8 83

a. Shut down per Procedure 2.1.4.1, RAPID SHUTDOWN.

2.4RXPWR, REACTOR POWER ANOMALIES.

Explanation:

NOTE: See attached 2.4RXPWR, Attachment 2.

The hold down beam for jet pumps 5 and 6 has broken and the jet pump mixers have displaced.

The flow through the jet pumps has reversed and some core flow is now bypassing the core from the other Recirculation loop. The core flow rises because more flow is going through the broken jet pumps. The flow instrumentation sees the rise in flow and displays this on the core flow recorder. Core plate dP drops because less flow is going through the core which also causes a drop in reactor power. Procedure 2.4RXPWR (Rev 4) Attachment 2 provides direction to determine the status of the jet pumps. The known parameters are compared to Attachment 4 information and the conclusion is the jet pump mixer is displaced. Attachment 2 Step 1.3.1.4 directs shutdown per Procedure 2.1.4.1.

Distractors:

b. Procedure 2.1.5 has rapid power reduction guidance but there is no direction to enter this procedure.

c. Procedure 2.1.5 does not have guidance to perform a rapid shutdown.

d. Procedure 2.4RXPWR provides guidance to lower reactor power but not a rapid power reduction.

Procedure 2.4RXPWR, Attachment 2

5. UNEXPLAINED DROP IN REACTOR POWER 5.1 If RTP drops > 2% with no corresponding reduction in total core flow or total core flow rises, perform following:1,2,3 5.1.1 Obtain following data and determine if parameter changes occurred in conjunction with power drop:

5.1.1.1 Total core flow on NBI-FRDPR-95 has risen.

5.1.1.2 Core plate P on NBI-FRDPR-95 lowered by 6 psid or more.

5.1.1.3 Rod line lowered by 2% or more with no control rod movement.

5.1.1.4 RR pump suction temperature on indicator or recorder (RR-TI-151A, RR-TI-151B, or RR-TR-165) has risen by 5°F or more.

5.1.1.5 RPV level initially rose and then returned to normal.

5.1.2 Compare data obtained in Step 1.1.1 to information presented in Attachment 3, and determine if conditions indicate shroud crack and separation, and crack location.

5.1.3 If shroud cracking and separation is indicated or if results are unknown, perform following:

5.1.3.1 Reduce power using recirculation flow per Procedure 2.1.10 until one of following is met:

a. Core pressure drop on NBI-FRDPR-95 is ~ 6.8 psid or

b. Rod line recovers to value before power drop.

5.1.3.2 Insert control rods per Procedure 10.13 to < 70% rod line.

5.2 If core shroud cracking and separation is confirmed or if reduction in reactor power cannot be explained, perform normal shutdown per Procedure 2.1.4.

5.3 Perform Jet Pump Operability determination per Procedure 6.LOG.601 to validate jet pump condition.

5.3.1 Compare data obtained in Step 5.3 to information presented in Attachment 4.

5.3.1.1 If loose jet pump mixer, determine if shutdown per Procedure 2.1.4 is necessary.

5.3.1.2 If plugged jet pump/riser, perform following:

a. If 1 jet pump completely plugged, shut down per Procedure 2.1.4.

b. If partial plugging occurs, determine if shutdown per Procedure 2.1.4 is necessary.

5.3.1.3 If instrument line failure, use alternate flow indication of other jet pump on associated riser.

NOTE - LCO 3.4.2 requires to be in Mode 3 in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for an inoperable jet pump.

5.3.1.4 If confirmed displaced jet pump/mixer, shut down per Procedure 2.1.4 or 2.1.4.1.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 295001 AA2.02 Importance Rating 3.2 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): Abnormal Procedure 2.4RR (Rev 38)

(Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date S9 New 00 Licensed RO: N 84 Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 2 1 1 Multiple Choice Topic Area Description Emergency Procedures Partial loss of forced recirculation and determining neutron monitoring response Related Lessons INT0320124 CNS Abnormal Procedure (RO) Reactor Recirculation COR0022202 REACTOR RECIRCULATION Related Objectives COR0022202001050I Briefly describe the following concepts as they apply to the Reactor Recirculation system or to the Recirculation Flow Control system:

Power to Flow Map (including normal operation/startup/shutdown and Stability Exclusion Region)

COR0022202001070J Predict the consequences a malfunction of the following would have on the Reactor Recirculation system or the Recirculation Flow Control system: Recirculation Flow Control System (under normal and reduced forced flow conditions)

Related References Procedure 2.4RR Related Skills (K/A) 295001 AA2.02 Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION : (CFR:

41.10 / 43.5 / 45.13) Neutron Monitoring (3.2)

QUESTION: S 9 84 The plant is operating at 104% rod line and 69% power when the RR Pump A trips. The RO reports reactor power has lowered to 55%.

What procedure is required to be entered, and what condition is indicative of thermal neutron instabilities?

a. Procedure 2.4RXPWR, REACTOR POWER ANOMALIES, and LPRM upscale or downscale indications alarming and clearing with an annunciation period of 10 seconds.

b. Procedure 2.4RR, REACTOR RECIRCULATION ABNORMAL, and LPRM upscale or downscale indications alarming and clearing with an annunciation period of 10 seconds.

c. Procedure 2.4RXPWR, REACTOR POWER ANOMALIES, and SRM period positive to negative SRM period swings with a fluctuation time of 2 seconds.

d. Procedure 2.4RR, REACTOR RECIRCULATION ABNORMAL, and SRM period positive to negative SRM period swings with a fluctuation time of 2 seconds.

ANSWER: S 9 84

d. Procedure 2.4RR, REACTOR RECIRCULATION ABNORMAL, and SRM period positive to negative SRM period swings with a fluctuation time of 2 seconds.

Explanation:

NOTE: See attached power-to-flow map.

Emergency Procedure 2.4RR (Rev 38), Attachment 3 describes abnormal neutron flux oscillations as SRM period positive to negative SRM period swings with a characteristic fluctuation time of < 3 seconds. The plant is just outside the buffer region when the RR pump trips and plant operation is deep into the stability exclusion region of the power to flow map.

Distractors:

a. Due to reactor power changing 2.4RXPWR is a reasonable procedure choice. The reason for the reactor power change is a RR pump trip so 2.4RR is the procedure to enter. The LPRM upscale and downscale annunciation period must be < 3 seconds to be considered abnormal flux oscillations.

b. Procedure 2.4RR is the correct procedure to enter. The LPRM upscale and downscale annunciation period must be < 3 seconds to be considered abnormal flux oscillations.

c. Due to reactor power changing 2.4RXPWR is a reasonable procedure choice. The reason for the reactor power change is a RR pump trip so 2.4RR is the procedure to enter. The period swings of 2 seconds is considered abnormal flux oscillations.

Beginning operation here.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 295005 G2.1.32 Importance Rating 4.0 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): Abnormal Procedure 2.4GEN-H2 (Rev 27),

(Attach if not previously provided) Attachment 6 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date S 10 New 00 Licensed RO: N 85 Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 2 1 4 Multiple Choice Topic Area Description Abnormal Procedures Generator warm gas temperature requiring power reduction Related Lessons INT0320127 CNS Abnormal Procedures (RO) Turbine/Generator Related Objectives INT0320127Q0Q0100 Given plant condition(s) and the applicable Abnormal/Emergency procedure, determine the correct subsequent actions required to mitigate the event(s).

Related References 2.4GEN-H2 Related Skills (K/A) 295005 Main Turbine Generator G 2.1.32 Ability to explain and apply system limits and precautions. l (CFR: 41.10 /

43.2 / 45.12) (4.0)

QUESTION: S10 85 The plant is at 65% power steady state. The following occur:

B-2/B-3, GEN HYDROGEN TO COOLER HIGH TEMP.

TGI-TR-100, TRANSFORMER TEMP, Channels 21-24,Warm Gas H2 Cooler In, are indicating 185°F and rising 1°F every 3 minutes.

What action is required and what procedure directs the action?

a. Lower reactor power to maintain the warm gas 194°F.

Procedure 2.4GEN-H2 provides the direction.

b. Lower reactor power to maintain the warm gas 194°F.

Procedure 2.4TURB provides the direction.

c. Throttle closed TEC-MO-149, HYDROGEN COOLER TCV-428 BYPASS valve.

Procedure 2.4GEN-H2 provides the direction.

d. Throttle closed TEC-MO-149, HYDROGEN COOLER TCV-428 BYPASS valve.

Procedure 2.4TURB provides the direction.

ANSWER: S10 85

a. Lower reactor power to maintain the warm gas 194°F.

Procedure 2.4GEN-H2 provides the direction.

Explanation:

NOTE: See attached Procedure 2.4GEN-H2, Attachment 6 partial.

Abnormal Procedure 2.4GEN-H2 (Rev 27), Attachment 6 directs lowering reactor power to maintain warm gas temperatures 194°F.

Distractors:

b. This is the correct action to take. Procedure 2.4TURB does not provide guidance on generator hydrogen cooling. Procedure 2.4TURB does direct lowering reactor power for other turbine generator problems.

c. The proper action would be to throttle open TEC-MO-149 which would provide more cooling flow to the heat exchangers. This is the correct procedure.

d. The proper action would be to throttle open TEC-MO-149 which would provide more cooling flow to the heat exchangers. Procedure 2.4TURB does not provide direction on generator hydrogen cooling but it does provide direction on other turbine generator support system temperatures.

Procedure 2.4GEN-H2, Attachment 6 partial ATTACHMENT 6 ABNORMAL STATOR OR GAS TEMPERATURES

6. Lower reactor power per Procedure 2.1.10 to maintain following temperatures:

NOTE 1 - Equivalent PMIS points may be substituted for TGI-R-200 or TGI-R-100 channels if TGI-R-200 or TGI-R-100 is unavailable. TGI-R-200 Channels 1 through 24 equivalent PMIS points can be viewed by displaying PMIS Group Displays CRLOG1 and CRLOG2.

NOTE 2 - If stator temperature rising, validate temperature using equivalent PMIS point by displaying PMIS Group Displays CRLOG1 or CRLOG2.

6.1 Warm gas, Channels 21 through 24 (PMIS G013 through G016), on TGI-TR-100, 194°F.

6.2 Cold gas, Channels 25 through 28 (PMIS G017 through G020), on TGI-TR-100

< 115°F.

6.3 Generator stator temperature, Channels 1 through 24, on TGI-TR-200 194°F.

6.3.1 If stator temperature 210°F:

6.3.1.1 If Annunciator 9-5-2/C-4 is clear, SCRAM and enter Procedure 2.1.5.

6.3.1.2 Trip Main Turbine.

6.3.1.3 If reactor was not scrammed, enter Procedure 2.2.77.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 295021 G 2.2.38 Importance Rating 4.5 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): Tech Spec LCO 3.4.7 Condition A (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: LCO 3.4.7 and 3.4.8 Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR Part 55 Content: 55.41 55.43 2 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date S 11 New 00 Licensed RO: N 86 Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 2 1 4 Multiple Choice Topic Area Description Technical Specifications COR0022302001030O Residual Heat Removal System Related Lessons INT0070505 CNS Tech. Spec. 3.4, Reactor Coolant System (RCS)

Related Objectives INT00705050010100 Given a set of plant conditions, recognize non-compliance with a Chapter 3.4 LCO.

INT00705050010200 Discuss the applicable Safety Analysis in the Bases associated with each Chapter 3.4 Specification.

INT00705050010300 Given a set of plant conditions that constitutes non-compliance with a Chapter 3.4 LCO, determine the ACTIONS that are required.

Related References TS LCO 3.4.7 Related Skills (K/A) 295021 Loss of Shutdown Cooling G2.2.38 Knowledge of conditions and limitations in the facility license. (CFR: 41.7 / 41.10 /

43.1 / 45.13) (4.5)

QUESTION: S 11 86 The plant is shutdown with RHR Loop A operating in shutdown cooling. The following conditions exist:

RHR Pump A is operating.

Both RR pumps are secured.

Reactor coolant temperature is 220°F and lowering 1°F/minute.

RHR Pump C supply breaker is de-energized for maintenance.

On October 6 at 0815, an electrical perturbation causes RHR Pump A and Service Water Pump A to trip.

What is the MAXIMUM time allowed before reactor coolant must be 212°F or less?

a. October 6 at 2015.

b. October 6 at 2215.

c. October 7 at 0815.

d. October 7 at 1015.

ANSWER: S 11 86

c. October 7 at 0815.

Explanation:

NOTE: See Attached partial of LCO 3.4.7 Conditions, Required Actions and Completion Times.

The plant is in Mode 3 and LCO 3.4.7 is applicable. Per bases, both pumps in one loop or one pump in each of the two loops must be OPERABLE. RHR Pumps A and C are in Loop A and one pump must be operable. When RHR Pump A tripped, LCO 3.4.7, Condition A applies.

Required Action A.3 requires the plant to be in Mode 4 with a Completion Time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months time is reached on October 7 at 0815.

Distractors:

a. This is the time for a 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months mode change. Mode changes with 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months limits are common in Tech Specs.

b. This is the time for a 14 hour1.62037e-4 days 0.00389 hours 2.314815e-5 weeks 5.327e-6 months mode change (12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months plus 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months that might be misapplied from Note 1 or 2 of the LCO).

d. This is the time for a 26 hour3.009259e-4 days 0.00722 hours 4.298942e-5 weeks 9.893e-6 months mode change (24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months plus 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months that might be misapplied from Note 1 or 2 of the LCO).

LCO 3.4.7 and Bases BASES ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 295024 EA2.03 Importance Rating 3.8 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): AMP-TB00 (Rev 6)

(Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: EOP GRAPH 7 and 10 Learning Objective: See Attached (As available)

Question Source: Bank # 21304 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date S 12 21304 02 02/10/2010 06/15/2005 NRC Style RO: N 87 Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 3 1 3 Multiple Choice Topic Area Description Emergency Operating INT0080618, Usafe region of the HCTL Graph Procedures Related Lessons INT0080618 OPS EOP AND SAG GRAPHS AND CAUTIONS Related Objectives INT00806180010300 Given plant conditions and the EOP and SAG Graphs Flowchart, determine if operation is within the allowed region of a graph.

Related References 5.8 Emergency Operating Procedures (EOPs)

Related Skills (K/A) 295024 EA2.03 Ability to determine and/or interpret the following as they apply to HIGH DRYWELL PRESSURE: (CFR: 41.10 / 43.5 / 45.13) Suppression pool level (3.8)

QUESTION: S 12 87 21304 An accident occurred. All control rods fully inserted and the following conditions exist:

Both CRD pumps are injecting.

Reactor pressure is 800 psig (stable).

Drywell pressure is 24 psig (rising slowly).

Torus pressure is 23 psig (rising slowly).

Drywell temperature is 230°F (stable).

Suppression Pool temperature is 210°F (rising slowly).

Primary Containment level is 12.5' (stable).

Reactor water level is -168 inches (FZ) indicated (steady).

The CRS must ...

a. direct the control room operator to initiate a < 100°F/hr cooldown.

b. exit EOP-1A pressure leg and Emergency Depressurize per EOP-2A.

c. direct the control room operator to rapidly depressurize the RPV using the main turbine bypass valves.

d. exit EOP-1A and enter SAG-1.

ANSWER: S 12 87 21304

b. exit EOP-1A pressure leg and Emergency Depressurize per EOP-2A.

Explanation:

NOTE: See attached AMP-TB00 (Rev 6) PSTG Step SP/T-3 and HCTL Graph With reactor pressure at 800 psig, PC level at 12.5 feet and SP temperature at 210°F, the unsafe area of the HCTL curve is entered. Emergency Depressurization is required.

Distractors:

a. RPV depressurization is required and the cooldown rate is going to be exceeded. The EOP 1A pressure leg is exited when ED is required. A cooldown rate should have already been ordered by the CRS per EOP 1A but can no longer be followed.

c. Cannot anticipate ED because ED is required. RPV level is below the MSIV closure interlock so the MSIVs are closed.

d. RPV level is steady above the point where exiting to SAGs is required. Actual level is higher (about 30 inches higher) than the indicated level.

PSTG/SATG Step SP/T-3 When average suppression pool temperature and RPV pressure cannot be maintained below the Heat Capacity Temperature Limit, EMERGENCY RPV DEPRESSURIZATION IS REQUIRED.

Discussion The CNS Heat Capacity Temperature Limit (HCTL) is the highest suppression pool temperature from which emergency RPV depressurization will not raise:

Suppression chamber temperature above maximum temperature capability of the suppression chamber, or

Suppression chamber pressure above Primary Containment Pressure Limit A, while the rate of energy transfer from the RPV to the containment is greater than the capacity of the containment vent. Depressurizing the RPV when suppression pool temperature and RPV pressure cannot be maintained below the HCTL thus avoids failure of the containment and equipment necessary for the safe shutdown of the plant. Refer to Section 16 of this appendix for a detailed discussion of the HCTL.

Control of RPV pressure relative to the HCTL is directed in the RPV Control guideline, entry to which is specified in Step SP/T-2. Therefore, the structure of the PSTG provides for controlling both suppression pool temperature and RPV pressure before reaching a requirement to initiate emergency RPV depressurization based on plant conditions relative to the HCTL.

Limiting reactor pressure curve 800 psig Operating Point

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 295026 G2.2.39 Importance Rating 4.5 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): Technical Specifications LCO 3.6.2.1, Condition D (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR Part 55 Content: 55.41 55.43 2 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date S 13 New 00 Licensed RO: N 88 Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 2 1 4 Multiple Choice Topic Area Description Tech Specs Less than one hour TS action statement Related Lessons INT0070507 CNS Tech Spec 3.6, Containment Systems Related Objectives INT00705070010500 From memory in MODES 1,2, and 3, state the actions required in less than or equal to one hour if suppression pool average temperature >

110 degrees F but is either less than or equal to 120 degrees F (LCO 3.6.2.1)

Related References TS 3.6.2.1 Related Skills (K/A) 295026 High Suppression Pool Temperature G 2.2.39 Knowledge of less than or equal to one hour Technical Specification action statements for systems. (CFR: 41.7 / 41.10 / 43.2 / 45.13) (4.5)

QUESTION: S 13 88 The plant is starting up and the operators are performing SP 6.ADS.201, ADS MANUAL VALVE ACTUATION (IST). When tested, SRV 71A sticks open resulting in the following:

Suppression Pool level is 12 feet 11 inches and rising slowly.

Suppression Pool temperature is 111°F rising slowly.

What is the MAXIMUM time allowed to complete the Technical Specification Required Actions?

a. Place reactor mode switch in shutdown immediately and verify suppression pool temperature 120°F once per 30 minutes.

b. Suspend all testing that adds heat to the suppression pool immediately and verify suppression pool temperature 120°F once per hour.

c. Reduce thermal power to < 1% RTP within one hour and verify suppression pool temperature 120°F once per hour.

d. Place reactor mode switch in shutdown within one hour and verify suppression pool temperature 120°F once per 5 minutes.

ANSWER: S 13 88

a. Place reactor mode switch in shutdown immediately and verify suppression pool temperature 120°F once per 30 minutes.

Explanation:

NOTE: See attached TS LCO 3.6.2.1.

Per Technical Specifications LCO 3.6.2.1, Condition D, when suppression pool temperature is above 120°F but 120°F, the mode switch must be placed in shutdown immediately AND verify the temperature is < 120°F once per 30 minutes. With the SRV stuck open, testing is technically suspended.

Distractors:

b. Suspending testing is the correct action, however, the valve is stuck open so testing is technically not being performed. The temperature is verified once per hour when testing is adding heat to the suppression pool but temperatures are less than 110°F.

c. Thermal power is lowered < 1% if the completion time of Condition A is not met.

Suppression pool temperature is verified < 110°F in Required Action A.

d. Placing the mode switch to shutdown in one hour is reasonable as it gives the organization time to plan for the shutdown. SR 3.6.2.1.1 requires verifying suppression pool temperature every 5 minutes when performing testing that adds heat to the suppression pool.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 295030 EA2.04 Importance Rating 3.7 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s):

AMP-TB00 (Rev 6) CNS PSTGs (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date S 14 New 00 Licensed RO: N 89 Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 2 1 4 Multiple Choice Topic Area Description EOPs Low SP level and containment dP Related Lessons INT0080613 OPS EOP Flowchart 3A-Primary Containment Control Related Objectives INT00806130011100 Given plant conditions and EOP Flowchart 3A, PRIMARY CONTAINMENT CONTROL, determine required actions.

INT00806130011200 Given plant conditions and EOP flowchart 3A, PRIMARY CONTAINMENT CONTROL, state the reasons for the actions contained in the steps.

Related References NONE Related Skills (K/A) 295030 EA2.04 Ability to determine and/or interpret the following as they apply to LOW SUPPRESSION POOL WATER LEVEL : (CFR: 41.10 / 43.5 / 45.13)

Drywell/suppression chamber differential pressure: Mark-I & II (3.7)

QUESTION: S 14 89 The following conditions exist:

Reactor is at 50% power.

Suppression Pool level is 9.8 feet and lowering 0.1ft/min.

The CRS has entered EOP 3A and is directing actions. What is required and why?

a. Enter EOP 2A, scram the reactor and emergency depressurize before the drywell and suppression pool pressure equalize.

b. Enter EOP 1A, scram the reactor and transition to EOP 2A to emergency depressurize before the drywell and suppression pool pressure equalize.

c. Enter EOP 1A, scram the reactor and anticipate emergency depressurization and depressurize within cooldown rate before the SRV Tail Pipe Level Limit is exceeded.

d. Enter EOP 2A, scram the reactor and emergency depressurize before the SRV Tail Pipe Level Limit is exceeded.

ANSWER: S 14 89

b. Enter EOP 1A, scram the reactor and transition to EOP 2A to emergency depressurize before the drywell and suppression pool pressure equalize.

Explanation:

NOTE: See attached PSTG When SP level goes below 9.6 ft. the downcomers are uncovered and drywell and suppression pool air space pressure equalize. The CRS must enter EOP 1A and scram the reactor and then transition to EOP 2A to emergency depressurize before the pressure suppression function is lost.

Distractors:

a. EOP 1A is entered and directs entering 2A to ED.

c. Although anticipating ED is allowed there is no time due to the SP leak rate. The SRV Tail Pipe Level Limit is for high SP levels not low.

d. EOP 1A is entered and directs entering 2A to ED. The SRV Tail Pipe Level Limit is for high SP levels not low.

PSTG/SATG Step SP/L-2.2 Maintain suppression pool water level above 9.58 ft (elevation of the downcomer openings).

(1) Before suppression pool water level drops to 9.58 ft (elevation of the downcomer openings), enter the RPV Control Guideline at Step RC-1 and execute it concurrently with this procedure.

(2) If suppression pool water level cannot be maintained above 9.58 ft (elevation of the downcomer openings),

EMERGENCY RPV DEPRESSURIZATION IS REQUIRED.

Discussion When suppression pool water level decreases to the elevation of the downcomer openings, any further drop in water level could result in direct exposure of the drywell atmosphere to the suppression chamber airspace thus compromising the pressure suppression function of the primary containment. Suppression pool water level should therefore be maintained above this elevation.

Energy in the RPV should be discharged outside the primary containment if possible and thereby reduce or limit the energy added to the suppression pool if emergency RPV depressurization becomes necessary. Entry to the RPV Control guideline is therefore specified before reaching the elevation of the downcomer openings so that the override before Step RC/P-1 can be used to anticipate emergency RPV depressurization and rapidly depressurize the RPV, irrespective of the resulting cooldown rate.

Entering the RPV Control guideline at Step RC-1 assures that, if possible, the reactor is scrammed before RPV depressurization is initiated.

While the increase in core void fraction following emergency RPV depressurization would temporarily shut down the reactor, a potential for subsequent core damage exists and sudden insurges of cold water could result in power spikes as RPV pressure decreases below the shutoff head of low pressure injection systems. Emergency depressurization with the reactor at power should therefore be avoided.

The override before Step RC/P-1 permits rapid depressurization through the main turbine bypass valves in anticipation of emergency RPV depressurization.

If emergency RPV depressurization is required, EPG Contingency #2 is entered through a transfer from Step RC/P of the RPV Control guideline.

A reactor scram is effected indirectly, through entry of the RPV Control guideline, rather than through an explicit direction in the Primary Containment Control guideline to ensure that RPV water level, RPV pressure, and reactor power are properly coordinated following the scram and to avoid potential conflicts with alternate rod insertion strategies in Step RC/Q if the RPV Control guideline is already in use. (Note that Step RC-1 of the RPV Control guideline requires initiation of a reactor scram only if a scram has not yet been initiated.)

An explicit direction to enter the RPV Control guideline must be provided since conditions requiring entry of the Primary Containment Control guideline do not necessarily require entry of the RPV

Control guideline. A scram may have therefore not yet been initiated even if suppression pool water level is low.

The numbered substeps clarify that the scram and emergency RPV depressurization action levels are sequential, independent judgments. Consistent with the definition of before in Section 2 of this appendix, a scram may be performed at any time with no particular margin to the limiting suppression pool water level intended. After the scram is performed, a second judgment is required to determine the need for emergency RPV depressurization. Consistent with the definition of can/cannot be maintained, emergency RPV depressurization may be performed immediately following the scram if it is apparent that suppression pool water level will ultimately drop below the limiting elevation or delayed until the limit is actually reached. The appropriate timing of the two actions is event-dependent and requires an evaluation of system performance and availability in relation to parameter values and trends.

As long as suppression pool water level remains at or above the elevation of the downcomer vent openings, the need to emergency depressurize the RPV due to suppression pool heatup is dictated by the Heat Capacity Temperature Limit.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 295038 EA2.01 Importance Rating 4.3 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): Procedure 5.7.1 (Rev 45)

(Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: 5.7.1 Attachment 4 Category A Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date S 15 New 00 Licensed RO: N 90 Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 2 1 4 Multiple Choice Topic Area Description Emergency Plan Off-site release Related Lessons GEN0030402 OPS EAL Training Part 1, Category A Related Objectives GEN0030401C0C050A Concerning event classification: Given a copy of EPIP 5.7.1 and an EAL identification code, determine the EAL and its associated emergency classification.

GEN0030402001050A Concerning event classification: Given a copy of EPIP 5.7.1 and an EAL identification code, determine the EAL and its associated emergency classification.

Related References Emergency Procedure 5.7.1 Emergency Classification Related Skills (K/A) 295038 EA2.01 Ability to determine and/or interpret the following as they apply to HIGH OFF-SITE RELEASE RATE : (CFR: 41.10 / 43.5 / 45.13) Off-site (4.3)

QUESTION: S 15 90 A radiation release event has occurred. A field survey at 1.2 miles from the reactor building and on NPPD property has revealed the following:

Thyroid CDE is 600 mR for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months of inhalation.

What is the highest EAL?

Is the location of the survey inside the site boundary or beyond the site boundary?

HIGHEST EAL INSIDE SITE BOUNDARY OR BEYOND

a. ALERT INSIDE Site Boundary

b. ALERT BEYOND Site Boundary

c. SAE INSIDE Site Boundary

d. SAE BEYOND Site Boundary ANSWER: S 15 90

d. SAE BEYOND Site Boundary Explanation:

NOTE: See attached EAL wall chart section for Rad Release and Procedure 5.7.1 for EAL AS1.3.

Procedure 5.7.1 (Rev 45) defines site boundary as 1 mile radius around the plant. NPPD property line is outside the 1 mile radius. EAL AS 1.3 for field survey being above 0.5 Rem is met and the location is beyond the site boundary.

Distractors:

a. EAL AA1.3 is met but the higher EAL AS1.3 applies. At 1.2 miles the dose is beyond the site boundary.

b. EAL AA1.3 is met but the higher EAL AS1.3 applies. At 1.2 miles the dose is beyond the site boundary.

c. The correct EAL but the does is beyond the site boundary.

Category: A - Abnormal Rad Release/Rad Effluent Subcategory: 1 - Off-Site Rad Conditions Initiating Condition: Off-site dose resulting from an actual or imminent release of gaseous radioactivity greater than 0.1 Rem TEDE or 0.5 Rem thyroid CDE for the actual or projected duration of the release EAL:

AS1.3 Site Area Emergency Field survey indicates closed window dose rates > 0.1 Rem/hr that is expected to continue for 60 min. at or beyond the site boundary (NOTE 1)

OR Field survey sample analysis indicates thyroid CDE > 0.5 Rem for 1 hr of inhalation at or beyond the site boundary NOTE 1 - The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition will likely exceed the applicable time.

If dose assessment results are available, declaration should be based on dose assessment instead of radiation monitor values (see EAL AS1.2). Do not delay declaration awaiting dose assessment results.

Mode Applicability:

All NEI 99-01 Basis:

This EAL addresses radioactivity releases that result in doses at or beyond the site boundary that exceed 10% of the EPA Protective Action Guides (PAGs). Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public.

CNS Basis:

The 0.5 Rem integrated CDE thyroid dose was established in consideration of the 1:5 ratio of the EPA Protective Action Guidelines for TEDE and thyroid exposure. In establishing the field survey emergency action levels, a duration of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months is assumed. Therefore, the dose rate EALs are based on a Site Boundary dose rate of 0.1 Rem/hr TEDE or 0.5 Rem for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months of inhalation CDE thyroid, whichever is more limiting.

For the purposes of this EAL, the Site Boundary for CNS is a 1 mile radius around the plant.

CNS Basis Reference(s):

1. CNS Drawing DWG.2.2 (P3-A-45).

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # 295009 AA2.03 Importance Rating 2.9 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): Procedure 2.2.66 (Rev 97)

(Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date S 16 New 00 Licensed RO: N 91 Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 2 1 4 Multiple Choice Topic Area Description Systems Adjusting blowdown flow due to lowering RPV level.

Related Lessons COR0022302 RESIDUAL HEAT REMOVAL Related Objectives COR0012002001040L Describe the interrelationship between the RWCU system and the following: Reactor Water Level COR0012002001050D Briefly describe RWCU operation under the following conditions:

Low Water Level COR0012002001100B Briefly describe the following concepts as they apply to RWCU:

Blowdown control valve controller operation including automatic closure.

Related References NONE Related Skills (K/A) 295009 AA2.03 Ability to determine and/or interpret the following as they apply to LOW REACTOR WATER LEVEL: (CFR: 41.10 / 43.5 / 45.13) Reactor water cleanup blowdown rate. (2.9)

QUESTION: S 16 91 The plant is starting up and control rods are being withdrawn to bring the reactor critical. The CRD system filter begins to plug, RPV level lowers until annunciator 9-5-2/G-1, REACTOR LOW WATER LEVEL sounds.

What operator action must be taken next to mitigate the condition, and what procedure is providing the guidance?

a. Shift CRD filters per Procedure 2.4CRD.

b. Start idle CRD pump per Procedure 2.4CRD.

c. Lower RWCU blowdown flow per Procedure 2.2.66.

d. Raise RWCU blowdown flow per Procedure 2.2.66.

ANSWER: S 16 91

c. Lower RWCU blowdown flow per Procedure 2.2.66.

Explanation:

NOTE: See attached Procedure 2.1.1 partial.

CRD is injecting water and RWCU is rejecting water in the blowdown mode. Procedure 2.1.1, STARTUP, directs maintaining RPV water level by rejecting through RWCU per Procedure 2.2.66 (Rev 97).

Distractors:

a. Guidance to shift CRD filters is per Procedure 2.2.8.

b. Guidance to start another CRD pump is per Procedure 2.2.8.

d. Blowdown flow must be lowered due to less RPV injection flow.

Procedure 2.1.1 partial.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # 295014 G2.4.35 Importance Rating 4.0 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): 2.4EX-STM (Rev 17)

(Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date S 17 New 00 None RO: N 92 SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 2 1 5 Multiple Choice Topic Area Description None INT0320135 , Abnormal Procedures - Condensate &

Feedwater Related Lessons INT0320135 CNS Abnormal Procedures (RO) - Condensate/Feedwater Related Objectives INT0320135L0L0100 Given plant condition(s), and the applicable Abnormal/Emergency Procedure, determine the correct subsequent actions required to mitigate the event(s).

Related References NONE Related Skills (K/A) 295014 Inadvertent Reactivity Addition G 2.4.35 Knowledge of local auxiliary operator tasks during an emergency and the l resultant operational effects. (CFR: 41.10 / 43.5 / 45.13) (4.0)

QUESTION: S 17 92 The plant is operating at rated conditions. The NLO is responding to an A-3 feedwater heater high level condition when the following alarm/CRT messages are received:

A-2/C-4, HEATER LOW LEVEL Alarm CRT indicates Heaters A-3 and A-2 are in a low level condition.

The BOP operator updates the crew that feedwater temperature has lowered by 13°F. The RO has lowered reactor power to rated.

What must the NLO do to correct the condition and what procedure provides the direction?

The NLO must throttle

a. closed the affected heater LCVs to raise heater level to the normal band as directed by Abnormal Procedure 2.4MC-RF, CONDENSATE AND FEEDWATER ABNORMAL.

b. closed the affected heater LCVs to raise heater level to the normal band as directed by Abnormal Procedure 2.4EX-STM, EXTRACTION STEAM ABNORMAL.

c. open the affected heater LCVs to raise heater level to the normal band as directed by Abnormal Procedure 2.4MC-RF, CONDENSATE AND FEEDWATER ABNORMAL.

d. open the affected heater LCVs to raise heater level to the normal band as directed by Abnormal Procedure 2.4EX-STM, EXTRACTION STEAM ABNORMAL.

ANSWER: S 17 92

b. closed the affected heater LCVs to raise heater level to the normal band as directed by Abnormal Procedure 2.4EX-STM, EXTRACTION STEAM ABNORMAL.

Explanation:

NOTE: See attached portion of abnormal procedure 2.4EX-STM and Attachment 2 for the normal feedwater heating region and loss of feedwater heating region.

Abnormal procedure 2.4EX-STM (Rev 17) provides direction for loss of feedwater heating. The low Heater level alarm directs entry into 2.4EX-STM. Per the abnormal procedure, the NLO has two hours to restore feedwater temperature by raising feedwater heater tank levels. At reactor power above 25%, the loss of feedwater heating region of 2.4EX-STM, Attachment 2 is reached when feedwater temperature lowers by 10°F. The loss of feedwater heating region of was reached as reported by the RO. The LCVs must be closed to raise the heater level within band and raise feedwater temperature and exit the loss of feedwater heating region of Attachment 2. The restoration of the feedwater heater level will lower feedwater temperature within a few minutes.

Distractors:

a. The correct action to take but the incorrect procedure. Condensate does flow through the tube side of the feedwater heaters.

c. The valve must be closed to lessen the flow to the next feedwater heater. The incorrect procedure is listed. Condensate does flow through the tube side of the feedwater heaters.

d. The valve must be closed to lessen the flow to the next feedwater heater. The correct procedure is listed.

EXCERT OF ABNORMAL PROCEDURE 2.4EX-STM 2.4EX-STM ATTACHMENT 2

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # 500000 G2.4.8 Importance Rating 4.5 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): Emergency Procedure 5.3AC120 (Rev 24)

(Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date S 18 New 00 Licensed RO: N 93 Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 2 1 4 Multiple Choice Topic Area Description Abnormal/EOPs Abnormal/EOP Procedure concurrent usage Related Lessons INT0320131 CNS Abnormal Procedures-Electrical Related Objectives INT0320131W0W0100 Given plant condition(s), and the applicable Abnormal/Emergency Procedure, determine the correct subsequent actions required to mitigate the event(s).

Related References NONE Related Skills (K/A) 500000 High CTMT Hydrogen Concentration G 2.4.8 Knowledge of how abnormal operating procedures are used in conjunction with EOPs. (CFR: 41.10 / 43.5 / 45.13) (4.5)

QUESTION: S 18 93 A LOCA has occurred with some fuel being uncovered. EOPs 1A, 2A and 3A are being utilized to respond to the event. There is an indication of hydrogen in the containment with Div 2 indicating 0.05% on SPDS. A subsequent short causes CCP-1B to become de-energized.

How is the hydrogen concentration monitored and what procedure is providing the direction?

Hydrogen concentration is available on

a. the recorder on VBD-P2 after Emergency Procedure 5.3AC480, 480 VAC BUS FAILURE directs transferring instrument power with MANUAL TRANSFER switch on VBD-H.

b. SPDS after Emergency Procedure 5.3AC480, 480 VAC BUS FAILURE directs transferring instrument power with MANUAL TRANSFER switch on VBD-H.

c. SPDS after Emergency Procedure 5.3AC120, LOSS OF 120 VAC directs transferring DIV 1 hydrogen/oxygen monitor from standby to operation.

d. the recorder on VBD-P2 after Emergency Procedure 5.3AC120, LOSS OF 120 VAC directs transferring DIV 1 hydrogen/oxygen monitor from standby to operation.

ANSWER: S 18 93

c. SPDS after Emergency Procedure 5.3AC120, LOSS OF 120 VAC directs transferring DIV 1 hydrogen/oxygen monitor from standby to operation.

Explanation:

NOTE: See attached Procedure 5.3AC120, Attachment 6 partial DIV 2 hydrogen/oxygen monitoring system is the normal system that is in service. Its remote panel is powered from CCP-1B. The DIV 1 hydrogen/oxygen monitor is normally in standby and is powered from CCP-1A. With a loss of CCP-1B, Emergency Procedure 5.3AC120 (Rev 24) is entered which directs transferring the DIV 1 monitor from standby to operation per its operating procedure (Attachment 6, Step 1.7). Once the DIV 1 monitor is in service, hydrogen concentration is indicated on SPDS, PMIS, and the recorder on VBD-P1.

Distractors:

a. The recorder on VBD-P2 is without power and EP 5.3AC480 does discuss loss of power to CCP-1B but provides no direction to respond to the loss of hydrogen/oxygen monitoring. The transfer switch on VBD-H is for instrument power transfer for loss of CCP-1A.

b. The remote panel for DIV 2 hydrogen/oxygen monitoring is without power so there is no hydrogen concentration indication on SPDS. EP 5.3AC480 does discuss loss of power to CCP-1B but provides no direction to respond to the loss of hydrogen/oxygen monitoring. The transfer switch on VBD-H is for instrument power transfer for loss of CCP-1A.

d. The recorder on VBD-P2 is without power but the recorder on VBD-P1 is available for hydrogen monitoring. Procedure 5.3AC120 is the correct procedure.

Procedure 5.3AC120, Attachment 6 partial ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 1 K/A # 2.1.5 Importance Rating 3.9 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): Procedure 0.12 (Rev 29)

(Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # 3934 (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date S 19 3934 00 10/26/1999 05/23/2010 Licensed RO: N 94 Modified Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 1 1 4 Multiple Choice Topic Area Description Administrative INT0320101F0F0100 CNS Administrative Procedures (RO)

Related Lessons INT0320101 CNS Administrative Procedures Volume 0, Administrative Procedures (Formal Classroom/Pre-OJT Training)

Related Objectives INT032010100F0100 State the working hours limitations and approval requirements associated with Administrative Procedure 0.12, Working Hours Limitations and Personnel Fatigue Management.

Related References 0.12 Working Hour Limitations Related Skills (K/A) 2.1.5 Ability to use procedures related to shift staffing, such as minimum crew complement, overtime limitations, etc. (CFR: 41.10 / 43.5 / 45.12) (3.9)

QUESTION: S 19 94 3934 Given the following conditions:

The plant entered a 14 day LCO at 0900 on Sunday 6/22.

An I&C Tech assigned for repairs worked 10 hour1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months days starting on 6/22.

Work is almost complete on 6/30 and the I&C Tech states one more 10 hour1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months day will complete the task.

(Assume the I&C Tech has worked 40 hour4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months work weeks the previous 5 weeks).

Can the I&C Tech perform the task on the next day?

What procedure controls the decision for completing the work?

a. Yes, if a working hour waiver and fatigue assessment forms are approved.

Procedure 0.12, WORKING HOURS LIMITATIONS AND PERSONNEL FATIGUE MANAGEMENT.

b. Yes, if a working hour waiver and fatigue assessment forms are approved.

Procedure 0-FFD-01, NPG FITNESS FOR DUTY PROGRAM AND BEHAVIOR OBSERVATION PROGRAM.

c. No, a waiver is NOT allowed and another I&C Tech must complete the task.

Procedure 0.12, WORKING HOURS LIMITATIONS AND PERSONNEL FATIGUE MANAGEMENT.

d. No, a waiver is NOT allowed and another I&C Tech must complete the task.

Procedure 0-FFD-01, NPG FITNESS FOR DUTY PROGRAM AND BEHAVIOR OBSERVATION PROGRAM.

ANSWER: S 19 94 3934

a. Yes, if a working hour waiver and fatigue assessment forms are approved.

Procedure 0.12, WORKING HOURS LIMITATIONS AND PERSONNEL FATIGUE MANAGEMENT.

Explanation:

NOTE: See attached Procedure 0.12 partial.

Per Procedure 0.12 (Rev 29), a 34 consecutive hours break must be taken after the eighth day.

A waiver and fatigue assessment must be performed. No other limitations are in effect.

Distractors:

b. The waiver and assessment must be completed but the procedure directing this is 0.12, WORKING HOURS LIMITATIONS AND PERSONNEL FATIGUE MANAGEMENT.

c. The waiver and assessment must be completed, but the procedure listed is correct.

d. A waiver is possible so another I&C Tech is needed and the procedure is not correct.

ORIGINAL QUESTION: 3934 Given the following conditions:

- The plant entered a 14 day LCO at 0900 on 6/22.

- An I&C Tech assigned for repairs has worked 10 hour1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months days starting on 6/22.

- Work is almost complete on 7/4 and the I&C Tech submits an overtime request to complete work by 0700 on 7/5 Exceptions to the working hour limitations shall be approved by...

a. a General Manager, or designee.

b. a Department Manager, or designee.

c. a Department Supervisor, or designee.

d. the Chief Nuclear Officer, or designee.

ANSWER: 3934

a. a General Manager, or designee.

REFERENCES:

Administrative Procedure 0.12

Procedure 0.12 partial 4.1 REST BREAKS (between work periods)

NOTE 1 - One period of shift turnover, either before or after shift, may be included in the rest break, but not both.

NOTE 2 - Incidental duties are (unscheduled) work activities occasionally performed off-site (including phone calls and work required by Supervisor to complete off-site) that are required by the Licensee but do not exceed a nominal, cumulative 30 minutes in a single break period. These occasional 30 minute incidental duties shall be excluded from calculation of working hours. If duties performed off-site exceed the nominal, cumulative 30 minutes or are scheduled activities, this time is counted in the individual's working hours and must be followed by a 10 hour1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months break.

4.1.1 A rest break of at least 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months shall be required between work periods unless approved per Section 0, Waivers, of this procedure.

4.1.2 A rest break of at least 34 hours3.935185e-4 days 0.00944 hours 5.621693e-5 weeks 1.2937e-5 months shall be required in any 9 day period.

6. WAIVERS 6.1 To the extent practicable, CNS shall rely on the granting of waivers only to address circumstances that could not have been reasonably controlled.

6.2 Waivers to work hour controls may be granted to an individual (not groups) under the following conditions:

6.2.1 The Operations Shift Manager determines that the waiver is necessary to mitigate or prevent a condition adverse to safety, the Security Manager or Security Shift Supervisor determines that the waiver is necessary to maintain site security, or a Site Senior-Level Manager with requisite signature authority makes either determination.

6.3 The waiver and fatigue assessment shall be documented by using the TimeWay Software Program.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 2 K/A # 2.2.37 Importance Rating 4.6 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): Tech Spec LCO 3.5.1 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: LCO 3.5.1 and 3.5.2 Learning Objective: See Attached (As available)

Question Source: Bank # 16422 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR Part 55 Content: 55.41 55.43 2 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date S 20 16422 02 08/19/2003 06/15/2005 NRC Style RO: N 95 Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 4 2 1 4 Multiple Choice Topic Area Description Technical Specifications, INT0070506, OPS Tech. Spec. 3.5, Emergency Core Cooling ODAM, TRM (ECCS) and React Related Lessons INT0070501 OPS Introduction to Technical Specifications INT0070506 OPS Tech. Spec. 3.5, Emergency Core Cooling (ECCS) and Reactor Core Isolation Cooling (RCIC) System Related Objectives INT00705010010200 Given plant conditions and a Specification, apply the rules of Section 3.0 to determine appropriate actions.

INT00705010010600 Explain the rules for Completion Times and apply these rules to determine the time allowed to complete Required Actions.

INT00705060010300 Given a set of plant conditions that constitutes non-compliance with a Section 3.5 LCO, determine the ACTIONS that are required.

INT00705060010100 Given a set of plant conditions, recognize non-compliance with a Section 3.5 LCO.

INT00705060010200 Discuss the applicable Safety Analysis in the Bases associated with each Section 3.5 Specification.

Related References 1.3 Completion times 3.0.3 LCO Applicability 3.5.1 ECCS Operating 3.5.1.C HPCI Subsystem Related Skills (K/A) 2.2.37 Ability to determine operability and/or availability of safety related equipment.

(CFR: 41.7 / 43.5 / 45.12) (4.6)

QUESTION: S 20 95 16422 The plant is at 100% power with the following past status:

January 1 at 12:00, RHR Pump A is declared inoperable

January 3 at 12:00, RHR Pump C is declared inoperable

January 6 at 06:00, RHR Pump A is restored to OPERABLE status

January 6 at 08:00, HPCI system is declared inoperable (Assume that equipment currently inoperable remains inoperable).

Including any completion time extension permitted by Technical Specifications, what is the LATEST time and date allowed to place the plant in MODE 3?

a. January 6, at 21:00

b. January 8, at 24:00

c. January 9, at 20:00

d. January 9, at 24:00 ANSWER: S 20 95 16422

b. January 8, at 24:00 Explanation:

NOTE: See attached TS LCO 3.5.1 Conditions and Required Actions.

CONDITION "A" of LCO 3.5.1 is not an inoperable pump, it is an inoperable subsystem. The same subsystem is inoperable when the second pump becomes inoperable and there is no allowed extension. The limit to complete RA "A.1" without the extension is 1/8 at 1200. Then entry into Condition B which allows 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to be in MODE 3. The time for MODE 3 is 1/8 at 2400.

When HPCI is inoperable, entry into Condition C and D is required. After 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months (Condition D) which is 1/9 at 0800, entry into Condition G is required. Condition G requires MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . This is 1/9 at 2000. Condition A would require entry into Condition B at 12:00 on 1/8, which would require placing the plant in MODE 3 by 24:00 and is more restrictive.

Distractors:

a. Assumes entry into Condition H and LCO 3.0.3 when HPCI is declared inop. LCO 3.0.3 requires MODE 3 within 13 hours1.50463e-4 days 0.00361 hours 2.149471e-5 weeks 4.9465e-6 months . This is 1/6 at 2100.

c. Assumes a 24-hour completion time extension is permitted for Condition A because the first inoperable component is fixed first. However, the CONDITION is not an inoperable pump, it is an inoperable subsystem. The same subsystem is inoperable when the second pump becomes inoperable and there is no allowed extension.

d. Assumes entry into Condition B following the 7-day CT for the first inoperable pump plus a 24-hour extension for the second pump. This time with the extension is 1/9 at 1200.

The entry into Condition C allows 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to be in MODE 3. This time is 1/9 at 2400 which is greater than that for Condition D and Condition G (HPCI).

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 2 K/A # 2.2.18 Importance Rating 3.9 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): Procedure 0.50.5 (Rev 25) Attachment 2 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: Procedure 0.50.5 Attachment 2 Learning Objective: See Attached (As available)

Question Source: Bank # 26179 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date S 21 26179 00 04/03/2012 08/22/2012 Licensed RO: N 96 Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 3 1 5 Multiple Choice Topic Area Description Administrative What is the level of risk and what action must be in place before work is to begin?

Related Lessons INT0231001 OPS Shutdown Risk Management Related Objectives INT02310010000500 Explain the term "Shutdown Safety Contingency Plan" as it relates to configuration and Defense in Depth.

Related References 0.50.5 OUTAGE SHUTDOWN SAFETY Related Skills (K/A) 2.2.18 Knowledge of the process for managing maintenance activities during shutdown operations, such as risk assessments, work prioritization, etc.

(CFR: 41.10 / 43.5 / 45.13) (2.6/3.9)

QUESTION: S 21 96 26179 The plant is in MODE 5, activities are planned on the Inventory Control Systems affecting that Key Safety Function. A contingency plan is required for those activities before they can begin.

The following concerns were raised:

Current approach has identifiable short comings that often result in this undesirable event occurring.

The loss of several key safety functions with recovery unable to be accomplished and potential entry into EAL level is highly likely but there are no alternatives.

What is the level of risk and what action must be in place before work is to begin?

a. MEDIUM; Key stakeholders are made aware that an issue could arise during the outage.

b. MEDIUM; Tag-outs in place, supplemental equipment installed, and compensatory measures in place.

c. HIGH; Key stakeholders are made aware that an issue could arise during the outage.

d. HIGH; Tag-outs in place, supplemental equipment installed, and compensatory measures in place.

ANSWER: S 21 96 26179

d. HIGH; Tag-outs in place, supplemental equipment installed, and compensatory measures in place.

Explanation:

NOTE: See Procedure 0.50.5, Attachment 2 partial.

From Procedure 0.50.5 OUTAGE SHUTDOWN SAFETY (Rev 25) - Attachment 2 Shutdown Safety Contingency Plan s Guide, apply the two concerns identified to the matrix and you come up with a level of risk of HIGH (3 and a 5). From the descriptions of each plan requirements, Tag-outs in place, supplemental equipment installed, and compensatory measures in place are required.

Distractors:

a. MEDIUM; is incorrect.

b. MEDIUM; is incorrect.

c. HIGH; Key stakeholders are made aware that an issue could arise during the outage is incorrect. This is a LOW risk requirement.

GRADED APPROACH TO CONTINGENCY PLANNING Determine the probability of requiring a contingency for the specific activity based on the following matrix:

LEVEL INTERPRETATION DEFINITION Event is not known to have occurred under similar 1 - Low Highly Unlikely circumstances using current approach.

Event is known to have occasionally happened under similar circumstances. Current approach 2 - Minor Unlikely does not have identified short comings that increase the likelihood.

Current approach has identifiable short comings 3 - Moderate Likely that often result in this undesirable event occurring.

4- Current approach has usually led to this event Highly Likely Significant occurring under similar circumstances.

Past experience indicates this event almost always 5 - High Near Certainty occurs under similar circumstances using current approach.

Determine the consequence of the event requiring a contingency plan using the following matrixes:

CONSEQUENCE TABLE LEVEL DEFINITION SHORT DEFINITION No impact or challenges on key safety functions 1 - Low No impact.

is expected.

There may be minor challenges to a key safety 2 - Minor function, but processes and procedures are in Minor challenges.

place to deal with the minor challenge.

There may be moderate performance shortfall; a Moderate challenges a 3 - Moderate key safety function may be affected with key safety function is degraded Defense-In-Depth consequences. affected.

There will be a major loss of a key safety Significant challenge 4 - Significant function, with significant consequences to the to a key safety stations shutdown safety Defense-In-Depth. function.

The loss of several key safety function with Unacceptable; no 5 - High recovery unable to be accomplished and alternatives exist.

potential entry into EAL level.

CONTINGENCY PLAN REQUIREMENTS LOW

Formally document contingency plans and include in Outage Contingency Plan Book.

Key stakeholders are made aware that an issue could arise during the outage.

MEDIUM

Fully develop contingency plans that are formally documented and included in Outage Contingency Plan Book.

Time sensitive actions are ready to implement as needed.

Activities requiring contingency plan activities are annotated in schedule.

HIGH

People are trained, on standby, practiced at the contingency, and ready to be implemented within the time restraints (such as time to boil).

Contingency plan developed that meet all of the requirements of this document.

Tag-outs in place, supplemental equipment installed, and compensatory measures in place.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 3 K/A # 2.3.5 Importance Rating 2.9 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): Procedure 5.7.12 (Rev 15)

(Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR Part 55 Content: 55.41 55.43 4 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date S 22 New 00 Licensed RO: N 97 Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 4 2 1 4 Multiple Choice Topic Area Description Administrative Radiological instrument to use for emergency exposure Related Lessons GEN0030105 RADIOLOGICAL MONITORING TEAMS Related Objectives GEN00301050200500 Identify the emergency dose limits Related References 5.7.12 EMERGENCY RADIATION EXPOSURE CONTROL Related Skills (K/A) 2.3.5 Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc. (CFR:

41.11 / 41.12 / 43.4 / 45.9) (2.9)

QUESTION: S 22 97 Due to an accident the OSC has put together a team to protect valuable NPPD property. The task will take 1.2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to complete and the expected dose for the task is 9 rem for each worker.

What preferred survey instrument(s) must the individuals have to do the work and what is the procedure that directs this requirement?

a. A low range portable survey instrument, 0 to 10 rem/hr.

EPIP Procedure 5.7.12, EMERGENCY RADIATION EXPOSURE CONTROL.

b. A low range portable survey instrument, 0 to 10 rem/hr.

EPIP Procedure 5.7.15, OSC TEAM DISPATCH.

c. A high range portable survey instrument, 0 to 1000 rem/hr is preferred and a low range portable survey instrument, 0 to 50 rem/hr should be available.

EPIP Procedure 5.7.12, EMERGENCY RADIATION EXPOSURE CONTROL.

d. A high range portable survey instrument, 0 to 1000 rem/hr is preferred and a low range portable survey instrument, 0 to 50 rem/hr should be available.

EPIP Procedure 5.7.15, OSC TEAM DISPATCH.

ANSWER: S 22 97

c. A high range portable survey instrument, 0 to 1000 rem/hr is preferred and a low range portable survey instrument, 0 to 50 rem/hr should be available.

EPIP Procedure 5.7.12, EMERGENCY RADIATION EXPOSURE CONTROL.

Explanation:

NOTE: See attached Procedure 5.7.12, Step 4.2. partial.

Per EPIP Procedure 5.7.12 (Rev 15), Step 4.2.1.1.a, the instrument of choice is the high range portable survey instrument 0 to 1000 rem/hr. A low range survey instrument of 0 to 50 rem/hr should also be available.

Distractors:

a. The 0 to 10 rem/hr instrument is not an option per Procedure 5.7.12. The procedure listed is the correct procedure.

b. The 0 to 10 rem/hr instrument is not an option per Procedure 5.7.12. The procedure listed is not the correct procedure although this procedure directs dispatching teams in an emergency.

d. This is the correct instrument per Procedure 5.7.12. The procedure listed is not the correct procedure although this procedure directs dispatching teams in an emergency.

Procedure 5.7.12 Section 4.2 partial 4.2 PERSONNEL EXPOSURE CONTROL 4.2.1 Individuals shall not enter any area where dose rates are unknown or unmeasurable with instruments immediately available.

4.2.1.1 If possible, the following survey instruments should be used:

a) High range portable survey instrument, 0 to 1000 rem/hr (0 to 10 Sv/hr); this should be the instrument of choice.

b) Low range portable survey instrument, 0 to 50 rem/hr (0 to 0.5 Sv/hr).

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 3 K/A # 2.3.6 Importance Rating 3.8 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): Procedure 8.8.11 (Rev 29) Attachment 1 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank # 20526 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR Part 55 Content: 55.41 55.43 4 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date S 23 20526 00 04/02/2004 05/23/2010 NRC Style RO: N 98 Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 2 1 3 Multiple Choice Topic Area Description Administrative Liquid Release Authorization/Approval and Actions for a Lost CW Pump During Release Related Lessons INT0320115 OPS CNS Administrative Procedures Radiation Protection and Chemistry Procedures (Formal Classroom/Pre-OJT Training)

Related Objectives INT0320115B0B0100 State who, by title, authorizes releases of radioactive liquid effluents from CNS.

INT0320115B0B0300 State the number of Circulating Water Pumps required to be in service during liquid radioactive discharges.

Related References 8.8.11 Liquid Radioactive Waste Discharge Authorization Related Skills (K/A) 2.3.6 Ability to approve release permits. (CFR: 41.13 / 43.4 / 45.10) (2.0/3.8)

QUESTION: S 23 98 20526 The plant is operating at low power with 2 Circulating Water pumps running. De-icing is in progress. The Radwaste Operator indicates that the Floor Drain Sample Tank requires discharging.

1) Whose approvals/authorizations is/are required in order to accomplish this discharge?

2) What action is required and why if one of the two operating circulating water pumps trip during the discharge?

a. 1) Chemistry department authorizes the release and the duty Shift Manager approves the release.

2) Continue the discharge at a lower rate as sufficient dilution flow exists.

b. 1) Duty Shift Manager authorizes and approves the release.

2) Continue the discharge at a lower rate as sufficient dilution flow exists.

c. 1) Chemistry department authorizes the release and the duty Shift Manager approves the release.

2) Terminate the discharge as insufficient dilution flow exists.

d. 1) Duty Shift Manager authorizes and approves the release.

2) Terminate the discharge as insufficient dilution flow exists.

ANSWER: S 23 98 20526

c. 1) Chemistry department authorizes the release and the duty Shift Manager approves the release.

2) Terminate the discharge as insufficient dilution flow exists.

NOTE: See attached Procedure 8.8.11, Attachment 1 Sections 3 and 4 Explanation:

Chemistry Procedure 8.8.11 (Rev 29), Attachment 1 requires that chemistry authorizes the release and the duty Shift Manager approves the release. The loss of one CW pump would reduce flow to less than the minimum required 159,000 gpm and the discharge should be terminated. Part of the CW discharge is routed back to the intake structure for de-icing and does not flow to the discharge canal so the credited dilution flow is lowered. When de-icing is in operation, two CW pumps can be credited for 193,200 gpm and one CW pump can be credited with a dilution flow of 118, 800 gpm.

Distractors:

a. The discharge should be terminated not lowered.

b. The discharge should be terminated, not lowered and the chemistry department authorizes the release.

d. Chemistry authorizes the release.

Procedure 8.8.11 Attachment 1 Sections 3 and 4 Section 3. AUTHORIZATION TO RELEASE RADIOACTIVE LIQUID WASTE To: Shift Manager From: Chemistry Release Authorization Number:

Total µCi/ml:

Total Concentration is < 1.0E-02 µCi/ml YES/NO Signature:

31 Day Dose, Percent Of Annual Limit For Each Value Is 2.0E+00 YES/NO Signature:

You Are Authorized To Release Subject Tank With Either Of Following Restrictions:

Maximum Liquid Waste Discharge Rate (gpm)

1) 2) 3)

Minimum Dilution Flow To Canal (gpm)

1) 159,000 2) 159,000 3) 159,000 Discharge Monitor Alarm Setpoint (µCi/ml)

1) 2) 3)

NOTE - Terminate Discharge If Above Specifications Cannot Be Maintained.

Contents Of This Tank Are Within Chemical Parameters For Discharge.

Chemistry: Time: Date:

Section 4. SHIFT MANAGER APPROVAL TO RELEASE 4.1 Circle Appropriate Discharge Canal Flow Rate:

NUMBER OF AVERAGE CW DISCHARGE FLOWRATE OPERATING CW (gpm)

PUMPS DE-ICING NO DE-ICING 4 378,600 631,000 3 308,400 514,000 2 193,200 322,000 1 118,800 198,000

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 4 K/A # 2.4.6 Importance Rating 3.8 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): EOP-1A Steps RC/L-14 and RC/L-15.

(Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: See Attached (As available)

Question Source: Bank # 4220 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date S 24 21807 01 07/08/2008 05/23/2010 NRC Style RO: N 99 Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 4 2 1 5 Multiple Choice Topic Area Description Emergency Operating INT0080611, What do you do when RPV Level Instruments are Procedures restored?

Related Lessons INT0080612 OPS EOP FLOWCHART 7B - RPV FLOODING (FAILURE-TO-SCRAM)

Related Objectives INT00806120010400 Given plant conditions and EOP Flowchart 7B, RPV FLOODING (FAILURE TO SCRAM) and REACTOR POWER (FAILURE TO SCRAM), determine required actions.

INT00806120010500 Given plant conditions and EOP flowchart 7B, RPV FLOODING (FAILURE TO SCRAM) and REACTOR POWER (FAILURE TO SCRAM), state the reasons for the actions contained in the steps.

Related References EOP 7B RPV Flooding Failure To Scram Related Skills (K/A) 2.4.6 Knowledge of EOP mitigation strategies. (CFR: 41.10 / 43.5 / 45.13) (3.7/4.7)

QUESTION: S 24 99 21807 During the execution of EOPs during an ATWS, water level indication is lost and flooding is entered. Maintenance restores water level indication and the TSC notifies the control room that water level indication is restored.

What action is required?

a. Exit EOP-7B. Enter EOP-1A and restore water level +3 inches to +54 inches per level leg and maintain the reactor depressurized per the pressure leg.

b. Exit EOP-7B. Enter EOP-7A and control level -183 inches FZ to +54 inches with outside shroud injection and Enter EOP-6A and maintain the reactor depressurized.

c. Enter EOP-7A and control level -183 inches FZ to +54 inches with outside shroud injection and continue actions in EOP-7B to maintain the reactor depressurized.

d. Enter EOP-1A and maintain the reactor depressurized in the pressure leg. Enter EOP-7A and control level -183 inches FZ to +54 inches with outside shroud injection.

ANSWER: S 24 99 21807

b. Exit EOP-7B. Enter EOP-7A and control level -183 inches FZ to +54 inches with outside shroud injection and Enter EOP-6A and maintain the reactor depressurized.

Explanation:

NOTE: See attached EOP Flowchart 7B override, Step F-S/F-1.

The 4th override contained in EOP-7B FS/F-1 states that if water level can be determined and the it has not been determined that the reactor remain shutdown under all conditions without boron then exit and go to EOP-6A and 7A for pressure and level control actions.

Distractors:

a. EOP-1A actions are not taken when the reactor is Not shutdown under all conditions without boron.

c. EOP-7B is exited and EOP-7B actions are no longer taken.

d. EOP-1A actions are not taken when the reactor is Not shutdown under all conditions without boron.

EOP-7B Overrides ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 3 K/A # 2.4.32 Importance Rating 4.0 Proposed Question: See Attached Proposed Answer: See Attached Explanation (Optional): See Attached Technical Reference(s): Procedure 5.7.1 (Rev 45) and Attachment 4 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: Category S of EAL WALL CHART Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date S 25 New 00 Licensed RO: N 100 Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 2 1 4 Multiple Choice Topic Area Description Emergency Plan Loss of all control room annunciators and PMIS Related Lessons GEN0030401 Emergency Plan for Licensed Operators Related Objectives GEN0030401C0C050E Concerning event classification: Given a copy of EPIP 5.7.1 and hypothetical abnormal plant symptoms, indications, or events, determine any and all EALs which have been exceeded and specifiy the appropriate emergency classification.

GEN0030403001050A Concerning event classification: Given a copy of EPIP 5.7.1 and an EAL identification code, determine the EAL and its associated emergency classification.

GEN0030403001050E Concerning event classification: Given a copy of EPIP 5.7.1 and hypothetical abnormal plant symptoms, indications, or events, determine any and all EALs which have been exceeded and specify the appropriate emergency classification.

Related References Procedure 5.7.1 Related Skills (K/A) 2.4.32 Knowledge of operator response to loss of all annunciators (CFR: 41.10 /

43.5 / 45.13) (4.0)

QUESTION: S 25 100 The plant is operating at 85% power and making plans to raise power to 100% per the load schedule. At 1250, PMIS fails and the IT department reports it will take one hour to being it back. At 1300, the following annunciator is received:

A-1/F-1, ANNUNCIATOR SYSTEM FAILURE The BOP reports that both lamps on the annunciator are illuminated. Maintenance is contacted and at 1305 report the annunciator problem will be repaired in 30 minutes.

What the HIGHEST EAL that is required and what is the MAXIMUM TIME allowed to make the declaration?

The Shift Manager must declare

a. a NOUE no later than 1320.

b. a NOUE no later than 1330.

c. an ALERT no later than 1320.

d. an ALERT no later than 1330.

ANSWER: S 25 100

c. an ALERT no later than 1320.

Explanation:

NOTE: See attached 5.7.1 (Rev 45) for EAL SA4.1 The loss of PMIS by itself does not require emergency planning to be involved. The annunciator A-1/F-1 having both lights lit indicates there is a total loss of all annunciators in the control room. The loss of PMIS means compensatory indications are unavailable. These two events require an ALERT (SA4.1) to be declared. Note 3 of the EAL table states the Emergency Director not wait until the applicable time has elapsed (15 minutes in this case) but should declare the event as soon as it is determined the condition will likely exceed the applicable time. When maintenance reports (at 1305) the system will be repaired in 30 minutes the Shift Manager has a maximum of 15 minutes to declare which is 1320 (1305 + 15 minutes).

Distractors:

a. A NOUE is applicable at 1305 but the ALERT is the highest.

b. A NOUE is applicable and it applies at 1305. If the Shift Manager waited the 15 minutes to declare then the time would be 1320 but NOTE 3 applies.

d. If the Shift Manager waited the 15 minutes to declare then the time would be 1320 but NOTE 3 applies.

PROCEDURE 5.7.1 (PARTIAL)

Category: S - System Malfunction Subcategory: 4 - Instrumentation Initiating Condition: Unplanned loss of safety system annunciation or indication in the Control Room with EITHER (1) a significant transient in progress, or (2) compensatory indicators unavailable EAL:

SA4.1 Alert Unplanned loss of > approximately 75% of annunciators or indicators associated with safety systems on Control Room Panels 9-3, 9-4, 9-5, and C (NOTE 3) for 15 min. (NOTE 3)

AND EITHER:

Any significant transient is in progress, Table S-1 OR Compensatory indications are unavailable NOTE 3 - The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition will likely exceed the applicable time.

Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Shutdown NEI 99-01 Basis:

This EAL is intended to recognize the difficulty associated with monitoring changing plant conditions without the use of a major portion of the annunciation or indication equipment during a SIGNIFICANT TRANSIENT.

"PLANNED" loss of annunciators or indicators includes scheduled maintenance and testing activities.

Quantification is arbitrary, however, it is estimated that if approximately 75% of the safety system annunciators or indicators are lost, there is an increased risk that a degraded plant condition could go undetected. It is not intended that plant personnel perform a detailed count of the instrumentation lost but use the value as a judgment threshold for determining the severity of the plant conditions. It is also not intended that the Shift Manager be tasked with making a judgment decision as to whether additional personnel are required to provide increased monitoring of system operation.

It is further recognized that most plant designs provide redundant safety system indication powered from separate uninterruptible power supplies. While failure of a large portion of annunciators is more likely than a failure of a large portion of indications, the concern is included in this EAL due to difficulty associated with assessment of plant conditions. The loss of specific or several safety system indicators should remain a function of that specific system or component operability status. This will be addressed by the specific Technical Specification.

The initiation of a Technical Specification imposed plant shutdown related to the instrument loss will be reported via 10CFR50.72. If the shutdown is not in compliance with the Technical Specification action, the UE is based on EAL SU3.1, Inability to Reach Required Shutdown Within Technical Specification Limits.

"Compensatory indications" in this context includes computer based information such as PMIS/SPDS. If both a major portion of the annunciation system and all computer monitoring are unavailable, the Alert is required.

Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.

This Alert will be escalated to a Site Area Emergency if the Operating Crew cannot monitor the transient in progress due to a concurrent loss of compensatory indications with a SIGNIFICANT TRANSIENT in progress during the loss of annunciation or indication.

CNS Basis:

PMIS and SPDS serve as redundant compensatory indicators which may be utilized in lieu of normal Control Room indicators. Safety system annunciation and indication considered in this EAL is found on Control Room Panels 9-3, 9-4, 9-5, and C. The other annunciators and indicators are important to plant operation but are not important to safety (Reference 1-14).

Significant transients are listed in Table S-1 and include response to automatic or manually initiated functions such as scrams, runbacks involving > 25% thermal power change, electrical load rejections of > 25% full electrical load, ECCS injections, or thermal power oscillations of 10% or greater.

ML12297A439

Text

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference: