Text

11 Final Written Exam
	ML17355A650
Person / Time
Site:	Wolf Creek
Issue date:	11/15/2017
From:	Vincent Gaddy; Operations Branch IV
To:	; Wolf Creek
References
	50-482/17-11 50-482OL-17
	Download: ML17355A650 (241)
	v • d • e

Examination Outline Cross-Reference Level RO 003 Reactor Coolant Pump Tier # 2 Group # 1 Knowledge of bus power supplies to the K/A # K2.02 following: K2.02 CCW pumps Rating 2.5 Question 1 The plant is operating at 100% power. Annunciator 00-018A, NB01 BUS LOCKOUT, alarms.

NB01 bus voltage reads 0 Volts. This will result in a loss of power to which of the following components?

A. Service Water Pump A (1WS01PA)

B. Service Water Pump B (1WS01PB)

C. Component Cooling Water Pump A (PEG01A)

D. Component Cooling Water Pump B CCW (PEG01B)

Answer: C Explanation:

A is wrong because Service Water Pump A is powered from SL41. Plausible because like the CCW pumps, Service Water Pump A is powered from a 4.16kV source.

B is wrong because Service Water Pump B is powered from SL31. Plausible because like the CCW pumps, Service Water Pump B is powered from a 4.16kV source.

C is correct because NB01 is the power supply to Component Cooling Water Pump A.

D is wrong because Component Cooling Water Pump B is powered from NB02. Plausible because CCW pump B is powered from a 4.16kV source.

Technical

References:

Lesson Plan LO1400800, Component Cooling Water System Rev 10, Page 42/49 References to be provided to applicants during exam: None.

Learning Objective: Lesson Plan LO1400800, Objective 11: Identify the electrical power supplies for the CCW pumps.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental 2 Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)7

55.43 Examination Outline Cross-Reference Level RO 003 Reactor Coolant Pump Tier # 2 Group # 1 Ability to predict and/or monitor changes in K/A # A1.09 parameters (to prevent exceeding design Rating 2.8 limits) associated with operating the RCPS controls including: A1.09 Seal flow and D/P Question 2 According to OFN BB-005, RCP MALFUNCTIONS, when Reactor Coolant Pump #1 Seal Leak Off reaches a MAXIMUM of 1) , you should determine the required action per Attachment E. When the seal differential pressure reaches a MINIMUM of 2) , immediate shutdown of the pump is required.

A. 1) 5.7 gpm

2) 209 psid B. 1) 5.7 gpm

2) 200 psid C. 1) 6.0 gpm

2) 209 psid D. 1) 6.0 gpm

2) 200 psid Answer: D Explanation:

A is wrong because part 1 is wrong but plausible because the #1 seal leak off alarm occurs at 5.7 gpm. Part 2 is wrong but plausible because the D/P alarm occurs at 209 psid.

B is wrong because part 1 is incorrect (see A above). Part 2 is correct.

C is wrong because part 1 is correct, but part 2 is incorrect (see A above).

D is correct.

Technical

References:

Document where the correct answer is found SY1300300, Reactor Coolant Pumps, Rev 24, page 26, OFN BB-005, Rev 27, page 8 References to be provided to applicants during exam: None.

Learning Objective: SY1300300, Reactor Coolant Pumps, Objective R4 Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X

Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental 2 Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b).5

Examination Outline Cross-Reference Level RO 004 Chemical and Volume Control Tier # 2 Group # 1 Knowledge of the operational implications of K/A # K5.09 the following concepts as they apply to the Rating 3.7 CVCS: thermal shock: high component stress due to rapid temperature change: K5.09 Question 3 Which of the following correctly describes the basis for a limitation that applies when the crew performs a plant shutdown from MODE 3 to MODE 5 in accordance with GEN 00-006, HOT STANDBY TO COLD SHUTDOWN?

When 1) is in operation, the crew must plot the T between the pressurizer and CVCS using STS BB-011, REACTOR COOLANT SYSTEM AND PRESSURIZER HEATUP/COOLDOWN SURVEILLANCE and maintain the T between the pressurizer and CVCS less than 583°F to prevent 2) .

A. 1) Normal pressurizer spray

2) Subjecting the pressurizer spray nozzle to undue thermal stress B. 1) Normal pressurizer spray

2) Pressurizer surge line stratification C. 1) Auxiliary pressurizer spray

2) Subjecting the pressurizer spray nozzle to undue thermal stress D. 1) Auxiliary pressurizer spray

2) Pressurizer surge line stratification Answer: C Explanation:

A (1) is wrong because GEN 00-006, Attachment G requires plotting the T using STS BB-011 when aux spray is in operation, not normal spray. Plausible because information must be recorded in STS BB-011 if T exceeds 280°F when normal spray is in use per GEN 00-006A, Attachment G. (2) Is the correct reason why the T must be maintained less than 583°F (refer to the bases for TSR 3.4.3.3).

B (1) is wrong because GEN 00-006, Attachment G requires plotting the T using STS BB-011 when aux spray is in operation, not normal spray. Plausible because information must be recorded in STS BB-011 if T exceeds 280°F when normal spray is in use per GEN 00-006A, Attachment G. (2) Is plausible but incorrect because it is the reason why the T during normal and aux spray operation must be maintained below 250°F, which is the more restrictive administrative limit identified in GEN 00-006A, Attachment G.

C is correct. (1) is correct because GEN 00-006 Attachment G requires plotting the T using STS BB-011 only when aux spray is in operation. (2) is correct because the 583°F limit is to guard against undue thermal stress on the spray nozzle (refer to the bases for TSR 3.4.3.3).

D (1) is correct because GEN 00-006 Attachment G requires plotting the T using STS BB-011 only when aux spray is in operation. (2) Is plausible but incorrect because it is the reason why the T during normal and aux spray operation must be maintained below 250°F, which is the more restrictive administrative limit identified in GEN 00-006A, Attachment G.

Technical

References:

Technical Requirements Manual Bases, TSR 3.4.3.3, Rev 31, Page B 3.4.3-4 GEN 00-006, Hot Standby to Cold Shutdown, Rev 98, Pages 78-79/116 References to be provided to applicants during exam: None.

Learning Objective: Lesson Plan LO1300400, Objective 18, Describe the operation of the CVCS during plant shutdown, reduced inventory, and refueling operations.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis 4 10CFR Part 55 Content: 55.41(b)5

Examination Outline Cross-Reference Level RO 005 Residual Heat Removal Tier # 2 Group # 1 Knowledge of RHRS design feature(s) and/or K/A # K4.07 interlock(s) which provide or the following: Rating 3.2 K.07 System protection logics, including high-pressure interlock, reset controls, and valve interlocks Question 4 The interlock for the Reactor Coolant System Hot Leg Suction Valves to the RHR Pumps (EJ HV 8701A/B, and BB PV-8702A/B) should 1) unless RCS Pressure is less than a MINIMUM of 2) .

A. 1) Automatically close

2) 360 psig B. 1) Prevent opening

2) 360 psig C. 1) Automatically close

2) 433 psig D. 1) Prevent opening

2) 433 psig Answer: B Explanation:

A is wrong because part 1 is wrong but plausible, as it is reasonable to assume than a pressure interlock would automatically close a valve. Part 2 is correct.

B is correct because the valves will not open unless pressure is less than 360 psig.

C is wrong because part 1 is wrong (see A above). Part 2 is wrong but plausible because 433 psig is the pressure at which a control room annunciator comes in if any RHR-RCH isolation valve is open.

D is wrong because part 1 is correct, but part 2 is wrong (see C above).

Technical

References:

SY1300500, Rev. 21, page 8 References to be provided to applicants during exam: None.

Learning Objective: SY1300500, Rev 21, Objective R2 Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental 3 Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b).7

Examination Outline Cross-Reference Level RO 005 Residual Heat Removal Tier # 2 Group # 1 Ability to manually operate and/or monitor in K/A # A4.02 the control room: Heat exchanger bypass flow Rating 3.4 control Question 5 Given:

The unit is in MODE 5 with RHR Train 'A' in service.

Temperature is being maintained at 160°F using both the HX outlet and bypass controllers in MANUAL.

Nitrogen backup supply has NOT been installed.

A loss of Instrument Air to the RHR system occurs.

EJ FCV-618, RHR Heat Exchanger A Bypass Valve, will ___1)___, and EJ HCV-606, RHR Heat Exchanger A Outlet Flow Control Valve, will ___2)___.

A. 1) Close

2) Open B. 1) Close

2) Close C. 1) Open

2) Close D. 1) Open

2) Open Answer: A Explanation: The RHR system is designed to maintain decay heat removal capabilities if a loss of instrument air were to occur. The valves that regulate bypass flow around the heat exchanger are air operated, fail closed valves. The flow control valves for the system are air-operated, fail open valves.

A is correct. See explanation above.

B (1) is correct. (2) is wrong because EJ HCV-606 fails open on a loss of instrument air.

Plausible if a candidate confuses the fail position of the two different kinds of valves.

C (1) is wrong because when there is a loss if IA, the bypass valves fail closed. (2) is wrong because EJ HCV-606 fails open on a loss of instrument air. Plausible if a candidate confuses the fail position of the two different kinds of valves.

D (1) is wrong because when there is a loss if IA, the bypass valves fail closed. Plausible if a candidate confuses the fail position of the two different kinds of valves. (2) is correct.

Technical

References:

Lesson Plan LP1300500, Rev 14, Pages and 11/36 and 30/36.

References to be provided to applicants during exam: None.

Learning Objective: LP 1300500, Objective 7, Describe system and integrated plant response to transient and equipment failures.

Question Source: Bank ID# 88678 X (Made minor changes to the wording of Modified Bank #

the question and answer choices.

Changed stem conditions to say loss of IA to RHR system, not loss if IA to EJ FCV-618.)

New Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis 2 10CFR Part 55 Content: 55.41(b)(7)

Examination Outline Cross-Reference Level RO 006 Emergency Core Cooling Tier # 2 Group # 1 2.1.20 Ability to interpret and execute K/A # 2.1.20 procedure steps. Rating 4.6 Question 6 The crew is at step 19 of EMG E-0, REACTOR TRIP OR SAFETY INJECTION, Check If ECCS Flow Should Be Reduced.

RCS Pressure is 1450 psig and rising

RCS Temperature is 500°F and stable

Pressurizer level is 20% and rising

RCPs are running

A, B, and D S/G NR levels are 20% and stable

C S/G NR level is 25% and stable

Total AFW flow is 250,000 lbm/hr

Containment pressure is currently at 0.5 psig, after reaching a maximum of 7.3 psig

Containment radiation is currently at 7X103 R/hr, after reaching a maximum of 6X104 R/hr

The TSC is NOT activated Can the crew proceed to EMG ES-03, SI TERMINATION?

A. Yes. All criteria are satisfied.

B. No. Pressurizer level must be raised ONLY.

C. No. Either total AFW flow must be raised or S/G levels must be raised ONLY.

D. No. Either total AFW flow must be raised or S/G levels must be raised, AND pressurizer level must be raised.

Answer: A Explanation:

A is correct because the criteria to reduce ECCS flow is RCS subcooling greater than 30°F, AFW flow greater than 270,000 lbm/hr OR level in at least one steam generator is greater than 6%, RCS pressure stable or increasing, and pressurizer level greater than 6%.

B is wrong because pressurizer level is adequate. It is plausible if one believes the adverse containment value of 29% applies, since there is adverse containment.

C is wrong because AFW flow and S/G levels are adequate. It is plausible if one believes that both AFW flow and S/G levels must be adequate, or if one believes that the adverse containment value of 29% applies, since there is adverse containment.

D is wrong because the parameters are all adequate. It is plausible if one believes adverse containment applies. Even though containment parameters are below 5 psig and 105 R/hr, adverse containment still applies since the staff has not determined if the instrumentation has received an integrated radiation dose of 106 R (since the TSC is not manned.)

Technical

References:

EMG E-0, Reactor Trip or Safety Injection, step 19, Revision 39 References to be provided to applicants during exam: None.

Learning Objective: LO1732313 Revision 012, Objective 4 Question Source: Bank # Diablo Canyon 2014-08 Question 4 (note changes; attach parent) Modified Bank #

New Question History: Last NRC Exam 2014 Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis 4 10CFR Part 55 Content: 55.41(b).10

Examination Outline Cross-Reference Level RO 007 Pressurizer Relief/Quench Tank Tier # 2 Group # 1 K/A # A3.01 Ability to monitor automatic operation of the Rating 2.7 PRTS, including: Components which discharge to the PRT Question 7 Given:

The unit is in MODE 1 at 100% power

Annunciators 00-034F PRT LEV HILO and 00-0034E PRT PRESS HI are in alarm

PRT level and pressure are rising slowly Which of the following could cause these indications?

A. Leakoff from RCP seal number 2 B. The CVCS letdown relief valve lifted C. Leakoff from the reactor vessel flange D. The A RHR pump discharge relief valve lifted Answer: B Explanation: The CVCS letdown relief inside containment discharges to the PRT.

A is wrong because RCP #2 seal leak-off goes to the RCDT. Plausible because RCP #1 seal returns relief valve goes to the PRT.

B is correct. See explanation.

C is wrong because rx vessel flange leakoff goes to the RCDT. Plausible because like CVCS letdown, it is a contaminated source that is collected inside containment, and a candidate could think it goes to the PRT.

D is wrong because the A and B RHR pump discharge relief valves go to the RHUT.

Plausible because A and B RHR pump suction relief valves go to the PRT.

Technical

References:

Lesson Plan LO1300200, Rev 9, P26/46 References to be provided to applicants during exam: None.

Learning Objective: Lesson Plan LO1300200, Determine where auxiliary systems/components interface with the RCS and the function of the interfacing system.

Question Source: Bank #105 (ID: 100493): X Replaced distractor PORV packing leakoff to RHR discharge relief lifted to avoid cueing applicants to the correct answer (because the correct answer is does NOT have leakoff in it and because PORV packing leakoff is not plausible (it doesnt exist).

Modified Bank #

New Question History: Last NRC Exam No (Seabrook 2013)

Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis 2 10CFR Part 55 Content: 55.41(b)(7)

Examination Outline Cross-Reference Level RO 008 Component Cooling Water Tier # 2 Group # 1 Ability to (a) predict the impacts of the K/A # A2.02 following malfunctions or operations on the Rating 3.2 CCWS, and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: A2.02 High/low surge tank level Question 8 According to the Foldout Page for OFN EG-004, CCW MALFUNCTIONS, if Surge Tank level on the train supplying the safety loop lowers to less than a MAXIMUM of 1) , then after placing the affected CCW pumps in pull-to-lock, you should 2) .

A. 1) 15%

2) Trip the reactor B. 1) 19%

2) Trip the reactor C. 1) 15%

2) Place supported equipment in pull-to-lock or normal-after-stop D. 1) 19%

2) Place supported equipment in pull-to-lock or normal-after-stop Answer: C Explanation:

A is wrong because part 1 is correct. Part 2 is incorrect but plausible because tripping the reactor is directed if the surge tank level supplying the service loop falls below 15%.

B is wrong because part 1 is incorrect but plausible because 19% is the level at which the CCW Surge Tank Low Level alarm comes in. Part 2 is incorrect (see A above).

C is correct because if the surge tank level supplying the safety loop falls below 15%, the foldout page directs placing equipment in pull-to-lock or normal-after-stop.

D is wrong because part 1 is incorrect (see B above). Part 2 is correct.

Technical

References:

OFN EG-004, CCW System Malfunctions, Rev. 18A, page 3. ALR 00-051D, CCW SRG TK A LEV HILO, Rev. 6A, page 1 References to be provided to applicants during exam: None.

Learning Objective: LO4710520, Component Cooling Water Malfunctions, Objective R2

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental 3 Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b).5

Examination Outline Cross-Reference Level RO 010 Pressurizer Pressure Control Tier # 2 Group # 1 Knowledge of the operational implications of K/A # K5.01 the following concepts as Rating 3.5 the apply to the PZR PCS: Determination of condition of fluid in PZR, using steam tables Question 9 Following a reactor trip in which SI was actuated, the following conditions exist in the PZR:

Level is 25% and rising.

Liquid temperature is 604°F.

Vapor temperature is 618°F.

Pressure is 1575 psig.

Based on these conditions, the PZR liquid is 1) and the PZR vapor is 2) .

A. 1) superheated

2) saturated B. 1) saturated

2) superheated C. 1) saturated

2) saturated D. 1) superheated

2) superheated Answer: B Explanation:Pressure must be converted to psia by adding 14.7 psi (1575 psig + 14.7 psi is about 1590 psia). Saturation temperature for 1590 psia is 604.09°F. The liquid space is 604°F, and i is therefore saturated. The vapor space is greater than 604°F, and it is therefore superheated.

A (1) and (2) are wrong because the vapor is superheated, not saturated. Plausible if the candidate fails to convert psig to psia because saturation temperature for 1575 psia is

~602F° and the liquid conditions are superheated. The second part is also wrong since PZR vapor is superheated for the given pressure.

B is correct. See explanation.

C (1) is correct. (2) is wrong because vapor space is superheated. Plausible if the candidate does not use the steam tables correctly.

D is wrong because the liquid is saturated. Plausible if the candidate does not use the steam tables correctly. (2) is correct.

Technical

References:

Steam tables.

References to be provided to applicants during exam: Steam tables Learning Objective:

Question Source: Bank #

(No changes) Modified Bank #135 (ID: X 108008)

New Question History: Last NRC Exam (2013 No DCPP)

Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis 2 10CFR Part 55 Content: 55.41(b)(5)

Examination Outline Cross-Reference Level RO 010 Pressurizer Pressure Control Tier # 2 Group # 1 Knowledge of the effect of a loss or K/A # K6.03 malfunction of the following will have on the Rating 3.2 PZR PCS: K6.03 PZR sprays and heaters Question 10 The Unit is at 100% power under the following conditions

Pressurizer Pressure is at 2100 psig and lowering

Both loop spray valves are open

Variable heaters are ON at minimum output

Back-Up heaters are OFF What failure would cause the plant to be in the conditions as stated above?

A. Pressurizer Pressure Master Controller Output High B. Pressurizer Spray Valve Controller Output High C. Pressurizer Pressure Master Controller Output Low D. Pressurizer Pressure Spray Valve Controller Output Low Answer: A Explanation:

A is correct B is wrong because this failure would result in opening of the affected pressurizer spray valve, not both and heaters would be unaffected.

C is wrong because this failure would result in both spray valves closing with both variable and back up heaters on.

D is wrong because it would result in closing of only the affected spray valve. The unaffected spray valve would still operate normally. Heaters would be unaffected.

Technical

References:

BD-OFN SB-008, Instrument Malfunctions, Rev. 14, Pages 72-73 References to be provided to applicants during exam: None.

Learning Objective: LO1124501, Controllers and Positioners, Obj. 7, 9 Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X

Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis 2 10CFR Part 55 Content: 55.41(b)7

Examination Outline Cross-Reference Level (S)RO 000007 Reactor Trip-Stabilization Tier # 1 Group # 1 Knowledge of the reasons for the following as K/A # EK3.01 the apply to a reactor trip: Actions contained in Rating 4.0 EOP for reactor trip: EK3.01 Question 11 Given:

The plant was operating at 75% power when a small break LOCA occurred

The reactor tripped

Safety Injection actuated

Annunciator 018A NB01 BUS LOCKOUT is lit

CCP B started and then tripped on overcurrent

SI Pump B did not start when Safety Injection actuated

RHR Pump B started when Safety Injection actuated

RCS pressure is 1395 PSIG and lowering

The crew is performing EMG E-0, REACTOR TRIP OR SAFETY INJECTION Which of the following describes 1) the procedurally required action the crew should take in accordance with EMG E-0 and 2) the reason for the action?

A. 1) Trip the RCPs

2) To prevent uncovering the core if RCPs were tripped later B. 1) Trip the RCPs

2) The RHR pump provides core heat removal C. 1) Maintain RCPs running

2) To prevent opening the PORVs later in the recovery D. 1) Maintain RCPs running

2) The RCPs provide core heat removal Answer: D Explanation:

A (1) is wrong because tripping RCPs is the wrong action. The stem says that neither the CCP or SI pump is running. A CCP or SI pump must be running to meet RCP trip criteria on the foldout page. EMG E-0 directs tripping RCPs only when ALL RCP trip criteria are met.

Part (2) is plausible because this is the reason given in BD-EMG E-0 for tripping RCPs when all of the RCP trip criteria on the foldout page are met.

B (1) is wrong because tripping RCPs is the wrong action. The stem says that neither the CCP or SI pump is running. A CCP or SI pump must be running to meet RCP trip criteria on the foldout page. EMG E-0 directs tripping RCPs only when ALL RCP trip criteria are met.

Part (2) is plausible because the RHR pump starts on a Safety Injection for decay heat

removal. However, the RHR pump would only be recirculating until RCS pressure dropped to 600#.

C (1) is correct. The information in the stem indicates no ECCS pumps are running.

Therefore, the RCPs are NOT to be tripped because RCP trip criteria on the foldout page are not met. Part (2) is not correct because this is not the reason for leaving the RCPs running per BD-EMG E-0. It is plausible because in EMG ES-03, precluding use of a PORV is part of the reason for starting the RCPs following recovery from a LOCA per EMG ES-03, Step 36.

D (1) is correct because the information in the stem indicates no ECCS pumps are running.

Therefore, the RCPs are NOT to be tripped because RCP trip criteria on the foldout page are not met. Part (2) is correct because when no SI or CCP pumps are running, then all trip criteria are not met, and BD-EMG E-0 explains that RCPs are left running because RCPs can provide RCS heat removal in these conditions.

Technical

References:

BD-EMG E-0, REACTOR TRIP OR SAFETY INJECTION, Rev 27, Page 101/108 EMG E-0 REACTOR TRIP OR SAFETY INJECTION Rev 39, Foldout Page References to be provided to applicants during exam: None.

Learning Objective: Lesson Plan LO1732313 R4: Explain the bases and any knowledge requirements for selected procedure steps.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis 3 10CFR Part 55 Content: 55.41(b)10 55.43

Examination Outline Cross-Reference Level RO 008 Pressurizer Vapor Space Accident Tier # 1 Group # 1 2.4.4 Ability to recognize abnormal indications K/A # 2.4.4 for system operating parameters that are Rating 4.5 entry-level conditions for emergency and abnormal operating procedures.

Question 12 Reactor Power is 100% when the following indications are observed.

PZR Pressure is 2100 psig and slowly lowering

PZR level is 57% and slowly rising

PRT Pressure is 1 psig and steady

CTMT Rad Levels 3 Rad/hour and slowly rising

CTMT Humidity is 20% and rising

CTMT Pressure is 1.1 psig and slowly rising

RCS Tavg is 585.3°F and steady What event is in progress?

A. RCS Cold Leg leak B. Pressurizer vapor space leak C. Fully OPEN Pressurizer Safety Valve D. Faulted Steam Generator inside containment Answer: B Explanation:

A is wrong because while several of the parameters are responding like there is a SBLOCA in progress (RCS pressure, all of the Containment parameters,) making this a plausible choice but the fact that PZR level is rising and not lowering make this an incorrect choice.

B is correct because all of these indication are consistent with a PZR Vapor space accident.

C is wrong because while several of the parameters are consistent with a fully open PZR safety, making it plausible, the fact that PRT pressure is not rising makes this an incorrect choice.

D is wrong because while a steam line break inside containment causes some of the indication (all of the containment parameters and RCS/PZR pressure) but the fact that PZR level is rising and Tavg is steady means that this is not the accident in progress.

Technical

References:

LO 1610500, Rev. 9, page 8

References to be provided to applicants during exam: None.

Learning Objective: LO 1610500, Objective R3 Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis 3 10CFR Part 55 Content: 55.41(b).10

Examination Outline Cross-Reference Level RO 000011 Large Break LOCA Tier # 1 Group # 1 Knowledge of the reasons for the following responses as K/A # EK3.13 the apply to the Large Break LOCA: Hot-leg Rating 3.8 injection/recirculation Question 13 Given:

The reactor tripped and SI actuated at 1200 as a result of a large break LOCA.

The crew is performing EMG E-1, LOSS OF REACTOR OR SECONDARY COOLANT.

All RCS hot leg temperatures are 420°F.

The time is now 2200.

In accordance with EMG E-1, LOSS OF REACTOR OR SECONDARY COOLANT, the crew should now __ 1) __ in order to avoid __ 2) __.

A. 1) Transfer to hot leg recirculation

2) Core flow channel blockages caused by boron precipitation B. 1) Isolate the SI accumulators

2) Injecting nitrogen into the RCS, which may damage the ECCS pumps C. 1) Isolate the SI accumulators

2) Injecting nitrogen into the RCS, which may impede RCS depressurization D. 1) Transfer to hot leg recirculation

2) Core flow channel blockages caused by debris from thermal insulation in containment Answer: A Explanation: EMG E-1, Step 33 directs going to EMG ES-13, Transfer to Hot Leg Recirculation, when 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months since SI actuation has elapsed. Lesson Plan LO1732323, Rev 012, says, One of the most important Operator fundamentals during a LOCA event is setting high standards for precise control of plant evolutions for assuring proper alignment of the RHR and SI pumps to Hot Leg Recirculation 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months after the initiation of safety injection in order to flush concentrated boric acid solution out RCS break to preventing the solubility limit of boron from being reached and avoiding potential flow channel blockages, which may be caused by boron precipitation.

A is correct. See explanation.

B (1) is wrong because per EMG E-1, Step 26, SI accumulators will be isolated when at least two hot legs are less than 410°F, and the stem says all RCS hot legs are 420°F. Plausible because this is an action directed by EMG E-1 for a large break LOCA, and the candidate may choose it because he or she may not realize the significance of 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months having passed since SI actuated (and therefore, transfer to hot leg recirc is required by EMG E-1).

(2) is wrong because this is not the basis for Step 26 or for transferring to hot leg recirc; however, it is plausible because gas could cause gas-binding and damage to pumps.

C (1) is wrong because per EMG E-1, Step 26, SI accumulators will be isolated when at least two hot legs are less than 410°F, and the stem says all RCS hot legs are 420°F. Plausible because this is an action directed by EMG E-1 for a large break LOCA, and the candidate may choose it because he or she may not realize the significance of 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months having passed since SI actuated (and therefore, transfer to hot leg recirc is required by EMG E-1). (2) is wrong because it is not the reason for transferring to hot leg recirc; however, it is plausible because this is the reason for isolating the SI accumulators.

D (1) is correct. (2) is wrong because this is not the reason for transferring to hot leg recirculation; however, it is plausible because the UFSAR, Page 6.2-45, describes that during a large break LOCA, thermal used inside containment will be a significant source of debris captured in the containment sump.

Technical

References:

EMG E-1, Rev 27, Step 33, Page 47/53 Lesson Plan LO1732323, Rev 12, page 5/12 References to be provided to applicants during exam: None.

Learning Objective: Lesson Plan LO1732323, Rev 12, Objective R4, EXPLAIN the bases and knowledge requirements for selected procedure steps.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis 3 10CFR Part 55 Content: 55.41(b)(10)

Examination Outline Cross-Reference Level RO 015/017 Reactor Coolant Pump (RCP) Tier # 1 Malfunctions Group # 1 K/A # AK1.02 Knowledge of the operational implications of Rating 3.7 the following concepts as they apply to Reactor Coolant Pump Malfunctions (Loss of RC Flow): AK1.02 Consequences of an RCPS failure Question 14 The unit is at 45% power, and RCP B trips on overcurrent.

Initially, UNAFFECTED steam generator PRESSURES should 1) . According to ALR 00 071D, RCP B OVERLOAD TRIP, a manual reactor trip 2) required.

A. 1) rise

2) is B. 1) rise

2) is not C. 1) lower

2) is D. 1) lower

2) is not Answer: D Explanation:

A is wrong because both parts 1) and 2) are incorrect. Part 1) is incorrect because The affected steam generator will shrink with steam and feed flow decreasing. The nonaffected steam generators will assume a higher steam load, resulting in a higher steam and feed flow and decreased steam pressure, not increased steam pressure. Part 2) is incorrect because according to ALR 00-071D, a reactor trip is required if one RCP is tripped above 48% power.

B is wrong because part 1) is incorrect and part 2) is correct. See explanations in A above.

C is wrong because part 1) is correct but part 2) is incorrect. See explanations in A above.

D is correct because both parts 1) and 2) are correct. See explanations in A above.

Technical

References:

ALR 00071D, RCP B OVERLOAD TRIP, Rev. 10, Page 2 References to be provided to applicants during exam: None.

Learning Objective: LO17322415, objective R2

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis 3 10CFR Part 55 Content: 55.41(b).10

Examination Outline Cross-Reference Level RO 000025 Loss of RHR System Tier # 1 Group # 1 Knowledge of the interrelations between the K/A # AK2.03 Loss of Residual Heat Removal System and Rating 2.7 the following: Service water or closed cooling water pumps Question 15 Given:

The unit is in MODE 5.

RHR Train 'A' is in service.

A loss of offsite power occurs.

Both EDGs start and load onto their respective buses.

MCB Annunciator 055C, ESW PMP TROUBLE, is LIT.

ESW 'A' Pump is NOT running.

Which of the following identifies actions the crew would be required to perform by procedure?

A. Cross tie ESW and then start RHR A Pump B. Secure EDG A and then start RHR B Pump C. Start CCW B Pump and then start RHR 'B' Pump D. Start ESW A Pump and then start RHR A Pump Answer: B Answer Explanation: The conditions in the stem indicate that ESW A Pump has tripped on overcurrent, and therefore, it is not going to be available within 3 minutes. OFN EF-033, LOSS OF ESW, Foldout Page Item 1, directs unloading and stopping a running EDG if ESW is not available on for the EDG within 3 minutes. OFN EJ-15, LOSS OF RHR COOLING, Step 10, directs starting at least one RHR pump. RHR B pump should be started because it is the only RHR pump with ESW, CCW, and power available.

A. (Cross tie ESW and start RHR A Pump) is wrong because there is no direction in OFN EF-033, OFN EJ-015, or ALR 55C to crosstie ESW. Plausible because ALR 55C directs opening service water to ESW cross connect valves, and OFN EJ-015 Step 10 directs starting the previously running RHR pump. A candidate could think that ESW should be cross-connected to start the previously running ESW pump B. CORRECT (Secure EDG A and Start RHR B). See explanation.

C. (Start CCW B pump manually and start RHR B pump) is wrong because the stem says that all other plant equipment (i.e., other than ESW Tr A pump) responds as expected, and the CCW B Pump will start automatically by the Train B Shutdown Sequencer. Therefore, this is not an action the crew will need to perform for the given conditions. The distractor is

plausible because OFN EJ-015 directs starting CCW pumps manually when one is not already running in each train. It is possible that a candidate may not recall that the Shutdown Sequencer starts CCW pumps automatically following a LOOP without a SI or CSAS.

D. (Start ESW A pump manually and then start RHR A Pump) is INCORRECT because ESW A pump is not running, and because ALR 055C is lit, this means that the pump tripped on overcurrent and cannot be restarted. Therefore, RHR A pump cannot be started because CCW Train A has no heat sink.

Technical

References:

OFN EJ-15, LOSS OF RHR COOLING, Step 10, Rev 29, Page 27/181 OFN EF-033, LOSS OF ESW, Foldout Page Item 1, Rev 18. Page 3/122 ALR 00-55C, ESW PUMP TROUBLE, Rev 11, Page 1/5 References to be provided to applicants during exam: None.

Learning Objective: Document learning objective if possible.

Question Source: Bank # # 271 (ID:19483) X (changed distractors to include two Modified Bank actions and to reduce psychometric flaws and revised stem to state an alarm is lit vs the pump failed to start automatically and manually to improve operational validity)

New Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis 3 10CFR Part 55 Content: 55.41(b)(7)

Examination Outline Cross-Reference Level RO 026 Loss of Component Cooling Water Tier # 1 Group # 1 Knowledge of system purpose and/or function: K/A # 2.1.27 2.1.27 Rating 3.9 Question 16 The Unit is at 100% power

CCW Pump A and B are currently running supplying both Safety Loops

CCW Pump D is tagged out for maintenance and is unavailable

CCW Service Loop is aligned to Train B Subsequently a seismic event occurs which trips CCW Pump B and EG HV-53, CCW TRN A SPLY VLV, is unable to be electrically or mechanically opened. These events result in a loss of the __1)__ which __2)__ procedurally require a manual reactor trip.

A. 1) Safety Loop

2) would B. 1) Service Loop

2) would not C. 1) Safety Loop

2) would not D. 1) Service Loop

2) would Answer: D Explanation:

A is wrong because the combination of losing all flow in Train B with service loop being unable to be realigned to Train A due to failure of HV-53 results in a loss of the Service Loop. Second part is correct. OFN EG-004, CCW System Malfunctions, directs a manual trip of the reactor if the Service Loop is lost and unable to be realigned to the other train.

B is wrong because second part is incorrect.

C is wrong because both parts are incorrect.

D is correct because the loss of Train B coupled with the inability to realign flow to the Service Loop due to failure of HV-53 results in the loss of the Service Loop which will require a manual reactor trip per OFN EG-004.

Technical

References:

SY1400800, COMPONENT COOLING WATER SYSTEM, Rev. 029, Page 26 OFN EG-004, CCW SYSTEM MALFUNCTIONS, Rev. 18B, Page 6

References to be provided to applicants during exam: None.

Learning Objective: SY1400800, COMPONENT COOLING WATER SYSTEM, Obj. R3, R4 Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis 2 10CFR Part 55 Content: 55.41(b)3

Examination Outline Cross-Reference Level RO 000027 Pressurizer Pressure Control System Tier # 1 Malfunction Group # 1 K/A # AK2.03 Knowledge of the interrelations between the Rating 2.6 Pressurizer Pressure Control Malfunctions and the following: Controllers and positioners (CFR 41.7 / 45.7)

Question 17 Given:

The unit is at 100% power.

BB PS-455F, PZR Pressure Control Selector Switch, is selected to P457/P456.

BB PT-457 fails HIGH.

Which of the following describes plant response if NO operator actions are taken?

A. Both PORVs will open, and the reactor will trip on high pressure B. Only one PORV will open, and the reactor will trip on high pressure C. Both spray valves and one PORV will open, and the reactor will trip on low pressure D. Only one spray valve and both PORVs will open, and the reactor will trip on low pressure Answer: C Explanation: PT-457 is the controlling channel. When it fails high, then the master PZR pressure output goes to 100%, which results in both spray valves and PORV PCV-455A opening. Therefore, actual RCS pressure drops. The reactor will trip on low pressure without any operator action at 1940#.

A is wrong because only one PORV will open for failure of PT-457, and because actual plant pressure will lower, which will result on a reactor trip on low pressure (not high pressure).

Plausible because the failure will cause two spray valves to open, and a candidate could confuse the PORV and spray valve response to the failed instrument. Also plausible because if PT-457 failed low, then the reactor pressure would rise to the PORV setpoint (the plant would not trip though because the PORV setpoint is lower than the reactor trip setpoint). A candidate could confuse how the plant responds for PT-457 failing low and high and when the reactor trip and PORV setpoints are reached.

B is wrong because actual plant pressure will lower, which will result on a reactor trip on low pressure (not high pressure). Plausible because one PORV will open as a result of PT-457 failing high. Also plausible because if PT-457 failed low, then pressure will rise, and it will rise to the PORV open setpoint A candidate could confuse how the plant responds for PT-457 failing low and high and when the reactor trip and PORV setpoints are reached.

C is correct. See explanation.

D is wrong because both spray valves will open. Plausible because the reactor will trip on low pressure. Also plausible because only both spray valves open, and the candidate may confuse PORV and PZR spray valve response to the instrument failing high.

Technical

References:

LO4701000, Rev 3, Pages 26-27/30 References to be provided to applicants during exam: None.

Learning Objective: LO1301000, Rev. 010, Obj. 11: Describe system and integrated plant response to transient and equipment failures.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis 3 10CFR Part 55 Content: 55.41(b)(7)

Examination Outline Cross-Reference Level RO 038 Steam Generator Tube Rupture Tier # 1 Group # 1 Ability to determine or interpret the following K/A # EA2.01 as they apply to a SGTR: EA2.01 When to Rating 4.1 isolate one or more S/Gs Question 18 According to EMG E-3, STEAM GENERATOR TUBE RUPTURE, which of the following is/are the MINIMUM indication(s) that would procedurally REQUIRE a ruptured S/G (steam generator) to be isolated?

A. S/G level rising in an uncontrolled manner

-AND-S/G has abnormal radiation B. S/G has abnormal radiation

-OR-NR (narrow range) level in affected S/G is greater than 6% [29%]

C. S/G level rising in an uncontrolled manner

-AND-NR level in affected S/G is greater than 6% [29%]

D. S/G level rising in an uncontrolled manner

-OR-NR level in affected S/G is greater than 6% [29%]

Answer: C Explanation:

A is wrong because EMG E-3 requires either one of A AND NR level in affected S/G be greater than 6% [9%].

B is wrong because it is an OR statement. If it was an AND statement it would be correct.

C is correct.

D is wrong because it is an OR statement like B. If it was an AND statement it would be correct.

Technical

References:

LO4710501, Introduction to Steam Generator Tube Rupture, Rev. 004, Pg. 10 EMG E-3, Steam Generator Tube Rupture,, Rev. 34B, Pg. 5 References to be provided to applicants during exam: None.

Learning Objective: LO4710501, Introduction to Steam Generator Tube Rupture, Obj. T1 Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis 3 10CFR Part 55 Content: 55.41(b)10

Examination Outline Cross-Reference Level RO 000040 (W/E12) Steam Line Rupture - Tier # 1 Excessive Heat Transfer Group # 1 K/A # 2.4.21 Knowledge of the parameters and logic used Rating 4.0 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.

Question 19 The crew is monitoring the critical safety functions using EMG F-0, CRITICAL SAFETY FUNCTION STATUS TREES, following a reactor trip and SI actuation that occurred 40 minutes ago.

The crew observes the following indications:

Source range instruments are NOT energized.

IR SUR is - 0.1 DPM and lowering.

S/G A, B, 'C NR levels are 22% and rising.

S/G D NR level is 5% and lowering.

S/G A, B, 'C pressures are 700 psig and stable.

S/G D pressure is 400 psig and lowering.

RCS pressure is 1000 psig and lowering.

RCS cold leg temperatures in Loops A, B, and C are 400°F and lowering.

RCS cold leg temperature in Loop D is 220°F and lowering.

NO containment spray pumps are available.

Containment pressure is 28 psig and rising.

The crew should go to A. EMG FR-S2, RESPONSE TO LOSS OF CORE SHUTDOWN B. EMG FR-Z1, RESPONSE TO HIGH CONTAINMENT PRESSURE C. EMG FR-H5, RESPONSE TO STEAM GENERATOR LOW LEVEL D. EMG FR-P1, RESPONSE TO IMMINENT PRESSURIZED THERMAL SHOCK CONDITIONS Answer: D Explanation: An orange path exists for CSF F-04, Integrity, because (1) a temperature decrease greater than 100F occurred for all RCS cold legs in less than 60 minutes, (2) all RCS pressures and cold legs temperatures are to the right of Limit A in Figure 5 of EMG F-0, and (3) one RCS cold leg temperature is less than 240F. No red path or other orange paths exist for the indications provided in the stem. EMG F-0, Foldout Page #2, says that when an orange path is diagnosed and no red path exists, then the crew should go to the functional

restoration procedure for the affected safety function. Therefore, the crew should go to FR-P1 to address the orange path on integrity.

A is wrong because an orange path exists for integrity, and EMG F-0 directs going to the functional restoration procedure for the highest priority safety function with the orange path.

Plausible because a yellow path exists for the subcriticality safety function, which is a higher priority than integrity. A candidate may not recognize the orange path conditions for integrity exist and think it is necessary to address the yellow path for subcriticality and select this distractor.

B is wrong because an orange path exists for integrity, and EMG F-0 directs going to the functional restoration procedure for the highest priority safety function with the orange path, and integrity is a higher safety function than containment. Plausible because an orange path exists for containment. A candidate may not recognize an orange path exists for integrity but recognize an orange path exists for containment and select this distractor.

C is wrong because an orange path exists for integrity, and EMG F-0 directs going to the functional restoration procedure for the safety function with the orange path. Plausible because a yellow path exists for the heat sink safety function, which is a higher priority than integrity. A candidate may not recognize the orange path conditions for integrity exist and think it is necessary to address the yellow path for heat sink and select this distractor.

D is correct. See explanation.

Technical

References:

EMG F-0, Page 3/22, and Page 16/22, Revision 17.

References to be provided to applicants during exam: None.

Learning Objective: LO1732338, Revision 013, Objective T1, Given a set of parameters, from memory, EXPLAIN the purpose, entry conditions, and major actions of EMG F-0, CRITICAL SAFETY FUNCTION STATUS TREES (CSFST).

Question Source: Bank #

Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis 4 10CFR Part 55 Content: 55.41(b)(7)

Examination Outline Cross-Reference Level RO 054 Loss of Main Feedwater Tier # 1 Group # 1 Knowledge of the operational implications of K/A # AK1.02 the following concepts as they apply to Loss of Rating 3.6 Main Feedwater (MFW): AK1.02 Effects of feedwater introduction on dry S/G Question 20 The plant is in a loss of secondary heat sink

The crew is performing EMG FR-H1, RESPONSE TO LOSS OF SECONDARY HEAT SINK

All Steam Generators are dry

Plant is now in Bleed and Feed

RCS temperature is 552F

AFW has now become available

1) What is/are the MINIMUM number of steam generator(s) that need to be fed to remove decay heat?

2) What is the MOST important concern with feeding a dry steam generator?

A. 1) 1

2) Thermal stress leading to S/G failure or rupture B. 1) 2

2) Thermal stratification leading to S/G failure or rupture C. 1) 1

2) Thermal stratification leading to S/G failure or rupture D. 1) 2

2) Thermal stress leading to S/G failure or rupture Answer: A Explanation:

A is correct. Only one S/G is required to remove all decay heat and thermal stresses leading to S/G rupture or failure is the most important concern with an empty S/G and primary side temperature over 550F.

B is wrong because only one S/G is required to remove all decay heat and feeding two at one time is not allowed due to the possibility of steam generator rupture or failure. Thermal stratification would not be the most important concern at this point since the plant is in a feed and bleed situation it has already experienced a rapid cooldown due to the SI.

C is wrong because thermal stratification is not the primary concern.

D is wrong because feeding two S/Gs at the same time is not allowed or necessary.

Technical

References:

BD-EMG FR-H1, Background Response to Loss of Secondary Heat Sink, Rev. 015, Pg. 80-81 References to be provided to applicants during exam: None.

Learning Objective: LO1505900, Main Feedwater System, Obj. 14 Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental 3 Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)10

Examination Outline Cross-Reference Level RO 056 Loss of Offsite Power Tier # 1 Group # 1 Knowledge of the reasons for the following K/A # AK3.02 responses as they apply to the Loss of Offsite Rating 4.4 Power: Actions contained in EOP for loss of offsite power Question 21 Given:

A loss of offsite power and reactor trip occurred.

Offsite power has been restored.

The crew attempted to start RCP D in accordance with EMG ES-02, REACTOR TRIP RESPONSE, and it could NOT be started.

Per EMG ES-02, the crew should start A. RCPs A, B, and C to provide adequate spray flow.

B. RCP A ONLY to provide adequate spray flow.

C. RCPs A, B and C to remove decay heat from the core.

D. RCP A ONLY to remove decay heat from the core.

Answer: A Explanation: BD-EMG ES-02, Page 39/57, says, At Wolf Creek, there are PZR connections to one RCS hot leg via the surge line and to two RCS cold legs via the spray lines. Single pump operation in the loop that provides the best spray is preferred to obtain normal PZR spray capability. This is loop D. If the D RCP cannot be started, all remaining RCPs should be started to provide adequate spray flow. If only two RCPs can be started, the spray valve associated with the running RCP should be opened, this will provide minimal spray flow.

Running RCPs A, B, or C by themselves will not provide spray flow.

A is correct. See explanation.

B is wrong because running only RCP A, by itself will NOT provide spray flow. Plausible because running RCP D by itself will provide adequate spray flow. A candidate may think that running any RCP other than RCP D by itself will also provide adequate spray flow.

C is wrong because the RCPs are started to provide adequate spray flow not for heat removal. Plausible because EMG ES-02 directs starting all three RCPs if D RCP is NOT running.

D is wrong because all three RCPs are directed to be started by EMG ES-02 to provide adequate spray flow. Plausible because operation of a single RCP would assist in removing decay heat from the core.

Technical

References:

BD-EMG ES-02, Page 39/57, Revision 26 EMG ES-02, Page 47/59, Revision 34 References to be provided to applicants during exam: None.

Learning Objective: LO1732315, Revision 14, R4, Explain the bases and knowledge requirements for selected procedure steps.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis 2 10CFR Part 55 Content: 55.41(b)(5)

Examination Outline Cross-Reference Level RO 000057 Loss of Vital AC Inst. Bus Tier # 1 Group # 1 Ability to operate and / or monitor the following K/A # AA1.06 as they apply to the Loss of Vital AC Rating 3.5 Instrument Bus: Manual control of components for which automatic control is lost Question 22 Given:

The unit is at full power

MCB Annunciator 025A, NN01 INST BUS UV alarmed

Pressurizer Pressure Switch PS-455F is selected to the P455/P456 position

Pressurizer Level Switch LS-459D is selected to the L459/L460 position The crew entered OFN-NN-021, LOSS OF VITAL 120VAC INSTRUMENT BUS, and verified a loss of NN01 has occurred. As the RO you are procedurally required to ___1)____ and

____2)____ .

A. 1) Select the alternate level channel position L461/L460 with LS-459D

2) Place Pressurizer Master Controller BB PK-455A in manual and select channel P457/P456 to regain pressure control B. 1) Select the alternate level channel position L459/L461 with LS-459D

2) Place Pressurizer Master Controller BB PK-455A in manual and select channel P455/P458 with PS-455F to regain pressure control C. 1) Select the alternate level channel position L459/L461 with LS-459D

2) Place Pressurizer Master Controller BB PK-455A in manual and select channel P457/P456 with PS-455F to regain control over pressure control D. 1) Select the alternate level channel position L461/L460 with LS-459D

2) Place Pressurizer Master Controller BB PK-455A in manual and select channel P455/P458 to regain pressure control Answer: A Explanation:

A is correct because in the normal lineup the instruments for both control systems are selected to use the red train instruments, or 459 and 459. So both of these must be swapped. The only piece left to discern between A and B as correct is B contains 455/458 for pressure control and this is not the correct position since the red train failed, making A correct.

B is wrong because see A above.

C is wrong because loss of NN01 (red train) affects both of these systems with the switches in their normal lineup and they are given that way in the stem.

D is wrong because loss of NN01 (red train) affects both of these systems with the switches in their normal lineup and they are given that way in the stem.

Technical

References:

OFN-NN-021, rev 27A, pages 4-5.

References to be provided to applicants during exam: None.

Learning Objective: LO1732431 XX Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis 3 10CFR Part 55 Content: 55.41(b)7

Examination Outline Cross-Reference Level RO 000058 Loss of DC Power Tier # 1 Group # 1 Ability to determine and interpret the following K/A # AA2.03 as they apply to the Loss of DC Power: DC Rating 3.5 loads lost; impact on ability to operate and monitor plant systems Question 23 The Unit is in MODE 1 at 50% power.

Which of the following identifies an action the crew should perform per procedure as a result of a loss of Non Vital 125V DC Switchboard PK03?

A. Place Excess Letdown in service B. Place AMSAC Master Bypass Switch to On C. Manually trip the condensate pumps when hotwell level reaches the lo-lo level setpoint D. Manually trip the circulating water pumps when the condenser pit level reaches the high level setpoint Answer: B Explanation: OFN PK-029, Attachment C, OPERATOR ACTIONS FOR FAILURE OF BUS PK03, Step C1, states, Place AMSAC Master Bypass Switch To On.

A is wrong because loss of PK03 does not result in letdown isolation. Plausible because loss of PK51 or PK01 does cause letdown to isolate per Att F, and OFN PK-029 directs placing excess letdown in service when letdown isolates. A candidate could think this action must occur for loss of PK03.

B is correct. See explanation.

C is wrong because a loss of PK03 does not result in loss of the automatic trip function for the condensate pumps on lo-lo hotwell level. Plausible because OFN PK-029, Att E, directs this action for loss of PK41, and a candidate could think this action occurs for loss of PK03.

D is wrong because loss of PK03 does not result in loss of the automatic trip function for the circ water pumps. Plausible because loss of PK51 or PK01 results in a loss of the automatic trip function for the circulating water pumps when there is a high condensate pit level per Att F of OFN PK-29, and a candidate could think this action occurs for loss of PK03.

Technical

References:

OFN PK-029, Attachment C, OPERATOR ACTIONS FOR FAILURE OF BUS PK03, Step C1, Page Rev 25 References to be provided to applicants during exam: None.

Learning Objective: LO1732439, Rev 11, Objective R2, RECOGNIZE the available situations which are addressed by procedure OFN PK-029.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis 3 10CFR Part 55 Content: 55.43(b)(5)

Examination Outline Cross-Reference Level RO 000062 Loss of Nuclear Svc Water Tier # 1 Group # 1 Ability to operate and / or monitor the following K/A # AA1.05 as they apply to the Loss of Nuclear Service Rating 3.1 Water (SWS): The CCWS surge tank, including level control and level alarms, and radiation alarm Question 24 Given:

The unit is at full power when the crew completed responding to a fault on transformer XNB01

The BOP notes Train A CCW surge tank level is at 64% and rising

EG HIS-1 DI WTR TO CCW SURGE TANK A is closed

The Service Loop is aligned to Train A CCW The event is a ___1)____ and the RO should ___2)____.

A. 1) seal Water HX tube leak

2) isolate the Seal Water HX B. 1) letdown HX tube leak

2) isolate Letdown C. 1) thermal Barrier HX tube leak

2) isolate the Thermal Barrier HX D. 1) CCW HX tube leak

2) swap the service loop Answer: D Explanation:

A is wrong because the pressure gradient is not high enough to cause level to increase at that rate.

B is wrong because there were no radiation alarms given in the stem nor an increase in radiation values.

C is wrong because wrong because there were no radiation alarms given in the stem nor an increase in radiation values.

D is correct because with no radiation alarms and increasing surge tank level and auto makeup off, the leak has to be ESW across the tubes for the in-service loops CCW HX. The applicant must recognize that ESW is aligned due to the transformer failure.

Technical

References:

LO4710520, rev 13, page 5.

References to be provided to applicants during exam: None Learning Objective: LO4710520 R3 Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis 2 10CFR Part 55 Content: 55.41(b)7

Examination Outline Cross-Reference Level RO 065 Loss of Instrument Air Tier # 1 Group # 1 Ability to determine and interpret the following K/A # AA2.06 as they apply to the Loss of Instrument Air: Rating 3.6 When to trip reactor if instrument air pressure is decreasing Question 25 Given:

The Unit is at 100% power.

The crew has just entered OFN KA-19, LOSS OF INSTRUMENT AIR.

When air pressure lowers to a MINIMUM of __________, it is expected that equipment required for plant operation will be impacted and a reactor trip is required.

A. 70 psig B. 75 psig C. 90 psig D. 95 psig Answer: A Explanation: OFN KA-019, Step 2, LOSS OF INSTRUMENT AIR, directs the crew to trip the reactor if air pressure lowers to 70 PSIG and if equipment required for plant operation cannot be controlled.

A is correct.

B is wrong because the crew should trip the reactor at 70 PSIG, not 75 PSIG. Plausible because 75 PSIG is a very low value for instrument air, and a candidate may select it if he or she does not remember 70 PSIG exactly.

C is wrong because the crew should trip the reactor at 70 PSIG, not 90 PSIG. Plausible because at 90 PSIG, the N2 backup accumulators for the feedwater control valves provide motive force. A candidate may associate 90 PSIG with the value for tripping the reactor.

D is wrong because the crew should trip the reactor at 70 PSIG, not 95 PSIG. Plausible because 95 PSIG is a very low value for instrument air, and a candidate may select it if he or she does not remember 70 PSIG exactly.

Technical

References:

OFN KA-019, Rev 15, Page 5/79.

References to be provided to applicants during exam: None.

Learning Objective: Document learning objective if possible.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental 2 Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(5)

W/E04 LOCA Outside Containment Tier # 1 Group # 1 Knowledge of the interrelations between the K/A # EK2.2 (LOCA Outside Containment) and the Rating 3.8 following: Facilitys heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility Question 26 Given:

The unit was operating at full power when a reactor trip and Safety Injection occurred.

The crew is responding using EMG C-12, LOCA OUTSIDE CONTAINMENT.

MCB Annunciator 047D, RWST LEV LOLO1 AUTO XFR actuates.

What happens to the Containment Recirculation Sumps as a result of this condition?

A. Sumps remain dry since check valves prevent flow from the RWST to the Containment Recirculation Sump through the suction of the Residual Heat Removal pumps.

B. Sumps remain dry since the RWST suction to the RHR pumps close before the Containment Recirculation Sump suction valves open.

C. Water enters the sumps when RCS inventory flows back from the RCS through the RHR System to Hot Legs.

D. Water enters the sumps from RWST due to gravity draining while the Containment Recirculation Sump and RWST RHR valves are repositioning.

Answer: D Explanation:

The shift from the injection mode to the cold leg recirculation mode starts automatically but must be completed manually when RWST level reaches the LoLo-1 level Setpoint (36 %).

The Containment recirculation sump isolation to RHR pump suction valves, EJ HV-8811A/B, automatically open. When these valves are full open, the RWST to RHR pump suction isolation valves, BN HV-8812A/B, will close.

Level in the RWST will drain into the Sump due to elevation differential between the RWST and the sump.

A is wrong but plausible. There are check valves in the RWST suction line to the RHR pump system, but it allows for flow from the RWST to CTMT while both valve are open, there are also other RCS System penetrations that contain check valves (See M-12BB01 and M12EJ01)

B is wrong but plausible. This is the opposite response , CTMT Recirc Sump suction valves open first and are interlocked to allow closure of RWST Suction to RHR pumps so that a suction source is maintained.

C is wrong but plausible. This might be true if RHR Suction were aligned to the RCS system in a cooldown lineup prior to the break outside of containment. The RCS to RHR Suction valves are interlocked to prevent opening until <350 psig when conditions allow for placing RHR in shutdown cooling lineup. Wrong since the unit was in normal operation prior to the event. These valves are checked closed in EMG C-12, Step 4.

D Correct, see above.

Technical

References:

ALR 00-047D, Rev 10B, Page 1 SY1300600, Rev 12, Pages 42, 47, 60, EMG C-12, Rev 15, Page 10.

Drawings M-12BB01, M-12EJ01, E-13EJ06A/B and E-13BN03/A References to be provided to applicants during exam: None.

Learning Objective: LO1300600 R5, Explain ECCS and ECCS subsystem components, including automatic operation during all phases of injection/recirculation, power supplies, and failure modes.

Question Source: Bank #16560 X (note changes; attach parent) Modified Bank #

New Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis 3 10CFR Part 55 Content: 55.41(b)7

Examination Outline Cross-Reference Level RO E11 Loss of Emergency Coolant Recirc Tier # 1 Group # 1 Ability to operate and / or monitor the following K/A # EA1.3 as they apply to the (Loss of Emergency Rating 3.7 Coolant Recirculation): Desired operating results during abnormal and emergency situations Question 27 The plant experienced a loss of coolant accident and has entered EMG C-13, CONTROL ROOM RESPONSE TO SUMP BLOCKAGE.

Which pumps are NOT secured as an immediate response to recirculation sump blockage when suction is lost or signs of cavitation occur?

A. SI B. RHR C. CCP D. CTMT Spray Answer: B Explanation: EMG C-13, CONTROL ROOM RESPONSE TO SUMP BLOCKAGE, Foldout Page #1. CCP, SI, AND CONTAINMENT SPRAY PUMP STOPPING CRITERIA: IF any Charging pump, SI pump or Containment Spray pump loses suction or shows indication of cavitation, THEN stop the affected pump.

A is wrong because SI pumps must be secured per the FOP in EMG C-13.

B is correct. See explanation.

C is wrong because CCPs must be secured per the FOP in EMG C-13.

D is correct because CTMT Spray pumps must be secured per the FOP in EMG C-13.

Technical

References:

EMG C-13, Rev 8, Page 3/104 References to be provided to applicants during exam: None.

Learning Objective: Document learning objective if possible.

Question Source: Bank # 784 (ID: 19790) X Modified Bank #

New Question History: Last NRC Exam No

Question Cognitive Level: Memory/Fundamental 2 Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(7)

Examination Outline Cross-Reference Level RO 000077 Generator Voltage and Electric Grid Tier # 1 Disturbances Group # 1 K/A # AK1.02 Knowledge of the operational implications of Rating 3.3 the following concepts as they apply to Generator Voltage and Electric Grid Disturbances: Over-excitation Question 28 The unit is at 70% power and rising (after an outage) with the following generator conditions:

generator voltage regulator is in AUTO

50 MVAR lag/out on the display Severe weather starts to develop in the area very quickly, causing a grid disturbance and a subsequent alarm to occur:

ALR 131B, GEN FIELD OV

1) Which of the following describes the grid disturbance that has occurred?

2) What is the required operator action in response to the alarm and the MAXIMUM amount of time permitted to perform this action to avoid a generator trip?

A. 1) Grid voltage has risen causing an over-excitation condition

2) Lower excitation within 5 seconds B. 1) Grid voltage has lowered causing an under-excitation condition

2) Raise excitation within 5 seconds C. 1) Grid voltage has risen causing an over-excitation condition

2) Lower excitation within 10 seconds D. 1) Grid voltage has lowered causing an under-excitation condition

2) Raise excitation within 10 seconds Answer: C Explanation:

A is wrong because (see answer C).

B is wrong because (see answer C).

C is correct because the generator was already at lag/out conditions (given in stem) so the grid disturbance caused the generator to increase excitation more (over-excited) and therefore voltage/excitation must be lowered on the generator to prevent a trip in 10 seconds.

D is wrong because (see answer C)

Technical

References:

Lesson Plan LO4704502, Rev 2, pages 13 - 16. ALR 131B, GEN FIELD OV References to be provided to applicants during exam: None.

Learning Objective: LO4704502 Obj 1 Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis 3 10CFR Part 55 Content: 55.41(b)4

Examination Outline Cross-Reference Level RO 001 Continuous Rod Withdrawal Tier # 1 Group # 2 Ability to determine and interpret the following K/A # AA2.05 as they apply to the Continuous Rod Rating 4.4 Withdrawal: Uncontrolled rod withdrawal, from available indications Question 29 Given :

The unit is operating at 88% power with generator load stable.

The Rod Control System is in Automatic.

Reactor power is RISING

Tavg is greater than Tref

Pressurizer level is RISING Which of the following would cause these symptoms to occur?

A. Power Range Channel N-43 fails high B. Failed OPEN S/G safety valve C. Uncontrolled rod withdrawal D. Turbine load rejection Answer: C Explanation: Continuous rod withdrawal would cause reactor power to rise, and Thot would rise, which would cause Tave to rise above Tref for a constant steam demand. RCS pressure would rise due to pressurizer level rising (RCS less dense).

A is wrong because a failure of the PR channel high would not cause any plant parameters to change initially. It would cause rod withdrawal to be blocked at C-2 (103%) on 1/4 channels. As a result of the input of the failed PR channel high to the rate comparator of the Reactor Control Unit, which would anticipate Tave higher than Tref, rods would move in to reduce power to lower Thot and therefore Tavg. Plausible if the candidate does not understand plant response on failure of the PR channel.

B is wrong because this would increase steam demand, and therefore Tc would lower, causing Tave to be less than Tref. Plausible because reactor power would rise.

C is correct. See explanation.

D is wrong because load rejection would cause steam demand to lower, and less heat would be taken out of the RCS, and Tc would increase, which would make Tave > Tref, and power would lower. Plausible because Tave would initially be greater than Tref.

Technical

References:

UFSAR Section 15.4.2, Rev 29, page 15.4-7 LO13 001 00, Rev 11, Pages 35-37/61 References to be provided to applicants during exam: None.

Learning Objective: LO13 001 00, Rev 11, Objective 14: Discuss off-normal and emergency operations for the Rod Control System.

Question Source: Bank #8, ID: 71199 X (note changes; attach parent) Modified Bank #

New Question History: Last NRC Exam No (2011 WC exam)

Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis 3 10CFR Part 55 Content: 55.41(b)(7)

Examination Outline Cross-Reference Level RO 000003 Dropped Control Rod Tier # 1 Group # 2 Ability to operate and / or monitor the following K/A # AA1.07 as they apply to the Dropped Control Rod: In- Rating 3.8 core and ex-core instrumentation Question 30 The unit is at 25 percent power when Rod F-8 drops into the core. Which bistable in the nuclear instrumentation circuitry should actuate to cause a reactor trip for this event?

A. Intermediate range, high negative rate trip at - 2% for four seconds B. Intermediate range, high negative rate trip at - 4% for two seconds C. Power range, high negative rate trip at - 2% for four seconds D. Power range, high negative rate trip at - 4% for two seconds Answer: D Explanation:

A is wrong because both parts are incorrect.

B is wrong because it is not in the IR circuit, second part is correct.

C is wrong because the second part is incorrect (this rate is for rod ejection event)

D is correct because it is the PR circuit and it is a negative rate of - 4 % for two seconds.

Technical

References:

LO 1301501, rev 9, page 53. ALR 0-85B References to be provided to applicants during exam: None.

Learning Objective: LO 1301501 Obj. 11 Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis 2 10CFR Part 55 Content: 55.41(b)7

Examination Outline Cross-Reference Level RO 000028 Pressurizer Level Malfunction Tier # 1 Group # 2 Knowledge of the reasons for the following K/A # AK3.02 responses as they apply to the Pressurizer Rating 2.9 Level Control Malfunctions: AK3.02 Relationships between PZR pressure increase and reactor makeup/letdown imbalance Question 31 The unit is operating at 100% power, and PZR LEV CTRL SEL, BB LS-459D, is in the L 459/L 460 position when the following take place:

Letdown has isolated

Annunciator 042A, CHG LINE FLOW HILO is in alarm, and charging flow is rising

PZR pressure is 2255 psig and rising Which of the following describes the instrument failure that has occurred?

A. The upper selected PZR level channel failed to 0%.

B. The upper selected PZR level channel failed to 100%.

C. The lower selected PZR level channel failed to 0%.

D. The lower selected PZR level channel failed to 100%.

Answer: A Explanation:

A is correct because the upper selected PZR level channel controls charging flow valve control, heater interlocks, letdown orifice isolation auto closure, and Letdown Isolation Valve BG LCV-459D. Charging flow goes to a maximum because the upper selected PZR level channel has failed low at 0%, which closes the charging flow valve. All heaters de-energize when either selected PZR level channel is less than 17%. All three letdown orifice isolation valves close when either selected PZR level channel is less than 17%. BG LCV-459D automatically closes when the upper selected PZR level channel is less than 17%. With letdown flow isolated and charging flow at max, actual level in the PZR rises, causing RCS pressure to rise, eventually reaching the setpoint for the spray valves to open.

B is wrong because when the upper selected PZR level channel fails high to 100%, letdown does not isolate, the charging flow valve would modulate to reduce charging flow, and actual PZR level would lower. Because the PZR level deviation between the controlling pressurizer level channel and programmed pressurizer level is > 5%, the all backup heaters would turn on. B is plausible if the candidate confuses the plant response for a failure of the upper and the lower selected PZR level channels.

C is wrong because the lower selected PZR level channel does not control the charging flow valve, and charging flow would remain steady if the lower selected PZR level control channel

failed low. Letdown would still isolate and thus pressure would still rise, but not as much as for a failure of the upper selected PZR level channel. It is plausible because a candidate may not recall that the upper selected PZR level channel controls the charging flow valve.

D is wrong because the lower selected PZR level channel does not control the charging flow valve, and charging flow would remain steady if the lower selected PZR level control channel failed high. Letdown would not isolate because the channel failed high (above 17%). It is plausible because a candidate may not recall that the upper selected PZR level channel controls the charging flow valve.

Technical

References:

Lesson Plan LO4701000 Rev 003, Pages 25-26/30 OFN SB-008, Revision 44A, App J, Pages 48/122 References to be provided to applicants during exam: None.

Learning Objective: Lesson plan LO4701000, Objective 3: Explain integrated plant response between the Pressurizer Pressure and Pressure Level Control Systems and other interfacing systems.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis 3 10CFR Part 55 Content: 55.41(b)(5) 55.43

Examination Outline Cross-Reference Level RO 000033 Loss of Intermediate Range NI Tier # 1 Group # 2 Knowledge of the operational implications of K/A # AK1.01 the following concepts as they apply to Loss of Rating 2.7 Intermediate Range Nuclear Instrumentation:

Effects of voltage changes on performance Question 32 Given:

The unit is shutting down to MODE 3

Power is 35% power

MCB Annunciator 078D, IR CMP VOLT FAIL, alarms

At N35 NI Rack, LOSS OF COMP VOLT amber light is LIT Which of the following describes the effect this condition will have on Intermediate Range (IR) N35 channel performance as the shutdown continues?

A. At the present power, Intermediate Range channels will agree, and as power goes below 1E- 8 amps, channel N35 will begin to indicate higher than N36.

B. At the present power, N35 will indicate higher than channel N36, and the difference will rise as power is lowered.

C. At the present power, Intermediate Range channels will agree, and as power goes below 1E- 8 amps, channel N35 will begin to indicate lower than N36.

D. At the present power, N35 will indicate lower than channel N36, and the difference will rise as power is lowered.

Answer: A Explanation:

A is correct because at the present power, Intermediate Range channels will agree, and as power goes below 1E-8 amps, channel N35 will begin to indicate higher than N36.

B is wrong because at power, until gammas are significant, the channels will agree. Effect when shutdown is correct.

C is wrong because the error caused by loss of compensating voltage is only manifested when gamma radiation is a significant contributor to IR detector current (ie below about 1E-8 amps. However, a loss of compensating voltage would cause N35 to indicate HIGHER than N36, not less.

D is wrong because no change in IR indication would be expected at 35% power and shutdown effect is reversed.

Technical

References:

WC Exam Bank; BD-OFN SB-008, INSTRUMENTATION MALFUNCTION, Rev. 14, Attachment Q, Page 51 of 73 References to be provided to applicants during exam: None.

Learning Objective: LO108064. (This is also the question number)

Question Source: Bank # X (note changes; attach parent) Modified Bank #

New Question History: Last NRC Exam Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis 2 10CFR Part 55 Content: 55.41(b) 8

Examination Outline Cross-Reference Level (S)RO 000037 Steam Generator Tube Leak Tier # 1 Group # 2 Knowledge of EOP mitigation strategies: 2.4.6 K/A # 2.4.6 Rating 3.7 Question 33 At 85% power during a unit startup following a refueling outage, the crew received an alarm on Main Condenser Air Discharge Radiation Monitor GE RE-92. The crew entered OFN BB-07A, STEAM GENERATOR TUBE LEAKAGE. The following conditions exist in the control room:

PZR level lowered to 51% and then stabilized

Letdown flow is 75 gpm

Charging flow is 100 gpm

VCT level is 7% and lowering, and makeup is not available Which of the following describes the procedurally required actions the crew should perform?

A. Trip the reactor, initiate safety injection, and go to EMG E-0, REACTOR TRIP OR SAFETY INJECTION, Step 1, while continuing in OFN BB-07A.

B. Trip the reactor, initiate safety injection, exit OFN BB-07A, and go to EMG E-0, REACTOR TRIP OR SAFETY INJECTION, Step 1.

C. Align the charging pump suction to the RWST, trip the reactor, and go to EMG E-0, REACTOR TRIP OR SAFETY INJECTION, Step 1, while continuing in OFN BB-07A.

D. Align the charging pump suction to the RWST, trip the reactor, exit OFN BB-07A, and go to EMG E-0, REACTOR TRIP OR SAFETY INJECTION, Step 1.

Answer: C.

Explanation: (Note: If VCT level goes below 5%, the charging pump suctions swap to the RWST.) The stem provides indication that VCT level is not being maintained. Per Step 5 of OFN BB-07A, if VCT level is not being maintained, then the RNO directs aligning the charging pump to the suction of the RWST, tripping the reactor, and then going to E-0 while continuing in OFN BB-07A. BD-OFN BB-07A, Page 15/184 states that once the suction has swapped to the RWST, orderly shutdown will be difficult, and therefore the procedure directs tripping and going to EMG E-0. Further, because PZR level never lowered below 6% and has stabilized, the leak is within the capacity of the charging pump, and therefore, SI is not required.

A is wrong because PZR level never lowered below 6%, PZR level has stabilized, and therefore the procedure does not require initiating SI. The distractor is plausible because tripping the reactor and initiating SI are required if PZR level is below 6%. Going to E-0 and continuing in OFN BB-07A are part of the correct answer.

B is wrong because PZR level never lowered below 6%, PZR level has stabilized, and therefore the procedure does not require initiating SI. The distractor is plausible because tripping the reactor and initiating SI are required if PZR level is below 6%. Exiting OFN BB-07A and going to E-0 is correct only if the reactor has been tripped and SI has been actuated because of low PZR level.

C is correct. See explanation.

D is wrong because the procedure directs tripping the reactor and going to EMG E-0 while continuing in OFN BB-07A. Plausible because aligning the charging pump suction to the RWST is correct.

Technical

References:

BD-OFN BB-07A Steam Generator Tube Leakage background document, Rev 8, Page 16/184.

OFN-07A, Steam Generator Tube Leakage, Rev References to be provided to applicants during exam: None.

Learning Objective: LO1732436, Rev 7, Objective R3, Given a procedural flow path, EXAMINE the available options for procedure actions in OFN BB-07A.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis 3 10CFR Part 55 Content: 55.41(b)10 55.43

Examination Outline Cross-Reference Level RO 068 Control Room Evacuation. Tier # 1 Group # 2 Ability to determine and interpret the following K/A # AA2.09 as they apply to the Control Room Evacuation: Rating 4.1 Saturation margin (CFR: 43.5 / 45.13)

Question 34 Given:

The Control Room has been evacuated.

A plant cooldown from the Auxiliary Shutdown Panel has been initiated per OFN RP-013A, HOT STANDBY TO COLD SHUTDOWN FROM OUTSIDE THE CONTROL ROOM with the following limits:

Subcooling >50F

Steam Generator Level 45-55%

Which of the following describes how these limits are monitored per OFN RP-013A?

Subcooling is determined by 1) .

Actual Steam Generator level is determined by 2) .

A. 1) calculating RCS Subcooling based on RCS WR Temperatures and Pressures

2) reading Steam Generator level meter B. 1) reading RCS Subcooling meter

2) using temperature correction of indicated S/G level C. 1) reading RCS Subcooling meter

2) reading Steam Generator level meter D. 1) calculating RCS Subcooling based on RCS WR Temperatures and Pressures

2) using temperature correction of indicated level Answer: A Explanation:

Per OFN RP-013A, Step 14, direct maintaining RCS subcooling based on RCS wide range temperature and pressure. Per Step 15, S/G Levels (AE LI-517X, 528X, 537X, and 548X) are monitored using provided gages at the shutdown panel.

A Correct, see above..

B is wrong but plausible, there is no subcooling monitor in the Aux Shutdown Panel and there is no requirement to temperature correct indicated levels. Plausible since Core Subcooling Monitor system contains Reference RTDs that are used for Tcold corrections.

C is wrong but plausible, there is no subcooling monitor in the Aux Shutdown Panel, using the given indicated levels is right for part 2).

D is wrong but plausible. The first part is correct, subcooling must be calculated using WR temperatures and pressures. Parts 2) is wrong since there is no required temperature correction for the indicated S/G level. Plausible since Core Subcooling Monitor system contains Reference RTDs that are used for Tcold corrections Technical

References:

OFN RP-013A, Rev 5B, pages 9 and 10.

References to be provided to applicants during exam: None.

Learning Objective: LO1732424 R4 Question Source: Bank #98757 X (note changes; attach parent) Modified Bank #

New Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental 3 Comprehensive/Analysis 10CFR Part 55 Content: 55.43(b)5

Examination Outline Cross-Reference Level RO 067 Plant Fire Onsite Tier # 1 Group # 2 Knowledge of the operational implications of K/A # AK1.02 the following concepts as they apply to Plant Rating 3.1 Fire on Site: Fire-fighting Question 35 During an outage, a fire in the startup transformer will result in automatic A. Deluge, and NB02 will remain energized B. Deluge, and NB01 will remain energized C. Halon actuation, and NB02 will remain energized D. Halon actuation, and NB01 will remain energized Answer: B Explanation: Lesson Plan LO1408600, Page 31/66, says the startup transformers are protected by Automatic (Deluge) Water Spray Systems. Page 54/66, states, Automatic operation of the Water Spray Systems protecting A, B, and C Main Transformers, and the Startup Transformer, will cause an isolation of the power feeds to the appropriate transformer. A loss of A, B, or C Main Transformer will result in a turbine trip. A loss of the Startup Transformer during an outage will result in a loss of AC power to all but the NB01 bus.

A is wrong because the startup transformer powers NB02. When the fire occurs, automatic deluge will occur, and power to the transformer will be isolated. Power to NB02 will be lost.

Plausible if a candidate thinks power to NB01 will be lost.

B is correct. See explanation.

C is wrong because the startup transformer is not protected by Halon. Plausible because Halon does provide protection to other electrical equipment, and a candidate may think the startup transformer is protected by Halon. C is also wrong because power to NB02 will be lost. Plausible if a candidate thinks power to NB01 will be lost.

D is wrong because the startup transformer is not protected by Halon. Plausible because Halon does provide protection to other electrical equipment, and a candidate may think the startup transformer is protected by Halon.

Technical

References:

Lesson Plan LO1408600, Rev 7, Pages 36 and 66/83 References to be provided to applicants during exam: None.

Learning Objective: LO1408600, R4, Describe the functional relationship of the Fire Protection System with interfacing systems.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis 2 10CFR Part 55 Content: 55.41(b)(8)

Examination Outline Cross-Reference Level RO W/E13 Steam Generator Over-pressure Tier # 1 Group # 2 Ability to operate and/or monitor the following K/A # EA1.1 as they apply to the (Steam Generator Rating 3.1 Overpressure): EA1.1 Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Question 36 The unit is at 100% power when a 30% load rejection occurs. During the transient, the steam dumps do NOT open.

What failure in the Steam Dump System would cause this to occur?

A. AB PT-507 failing High B. AC PT-505 failing Low C. AC PT-506 failing High D. Tavg input failing Low Answer: C Explanation:

A is wrong as there will be no response in the normal Tavg mode of operation due to improper operation of AB PT-507. If steam dumps were operating in the Pressure Control mode of operation, Failure of AB PT-507 high will cause steam dump valves to OPEN in an attempt to re-establish Demand Setpoint as indicated on the erroneous pressure sensor, not closed.

B wrong as a failure of the AC PT-505 will only affect the steam dump system if operating in the Tavg mode AND an arming signal exists (C-7 or plant trip). Also if AC PT-505 failed low with C-7 armed, then a false Tavg/Tref mismatch would exist and the steam dumps would OPEN to bring Tavg down.

C is correct. . For the given scenario, the steam dumps did NOT open since the steam dumps did NOT arm (C-7) which would be caused by AC PT-506 failing high. The C-7 Arming signal comes from AC PT-506 ONLY and is NOT selectable.

D is wrong because Auctioneered high Tavg input is used for control of steam dumps. For a Tavg input failing low, there is NO steam dump response. If Tavg input failed high, steam dumps would open if armed (C7 actuated). -

Technical

References:

SY1504100, Steam Dump System, Rev. 18, Pg. 28-29 BD-OFN SB-008, INSTRUMENT MALFUNCTIONS, Rev 14. Attachment D (page 13 of 73) and Attachment B (page 5 of 73)

References to be provided to applicants during exam: None.

Learning Objective: SY1504100, Steam Dump System, Obj. 5 Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis 3 10CFR Part 55 Content: 55.41(b)7

Examination Outline Cross-Reference Level RO W/E16 High Containment Radiation Tier # 1 Group # 2 Knowledge of the interrelations between the K/A # EK2.2 (High Containment Radiation) and the Rating 2.6 following: Facilitys heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility Question 37 Given:

A LOCA occurred in Containment

Containment radiation is 10 R/hr

Containment Pressure peaked at 12 psig and is lowering.

The crew has entered EMG FR-Z3, RESPONSE TO HIGH CONTAINMENT RADIATION LEVEL.

Per EMG FR-Z3, Containment Spray Pumps ____1)____be started because ____2)_____.

A. 1) Should NOT

2) It is not appropriate to use Containment Spray to reduce radioactivity at low containment pressures.

B. 1) Should NOT

2) Containment Spray pumps are NOT procedurally allowed to be started with Containment Pressure <27 psig.

C. 1) Should

2) Establishing Containment Spray distributes sodium hydroxide which is effective in reducing Iodine activity produced from failed fuel.

D. 1) Should

2) Establishing Containment Spray is effective at scrubbing noble gasses from Containment Atmosphere.

Answer: A Explanation:

Per BD-EMG FR-Z3, Step 2. It should be noted that the use of containment spray to reduce radioactivity has been considered. However, since containment spray is designed for containment heat removal at high containment pressures, it has been determined that it would NOT be appropriate to use containment spray in this procedure to reduce radioactivity at low containment pressure.

A is correct. See above.

B is wrong because there is no prohibition to starting CTMT Spray pumps at <27 psig. EMG E-0, Step F12 checks Containment Pressure REMAINED <27 psig and directs actuation of Containment Spray even if <27 psig at the time of Step F12 performance. Plausible since CTMT spray pumps are NOT started per EMG FR-Z3 C is wrong because pumps are NOT started. Plausible because NaOH is used for Iodine Removal at high containment pressures. .

D is wrong because pumps are NOT started. Plausible because the design of the Containment Spray system (Spray header location and use of 197 nozzles per header) maximize the iodine removal at high containment pressures, but not Noble gasses..

Technical

References:

BD-EMG FR-Z3, Page 11/77, Rev 5 EMG E-0, Step F12, Page 95 of 99, Rev 39A Lesson Plan LO1302600, Page 5/40, Rev 8 References to be provided to applicants during exam: None.

Learning Objective: Lesson Plan LO1302600, Objective 1: Discuss the purpose of the Containment Spray and Containment Cooling Systems.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis 4 10CFR Part 55 Content: 55.41(b)7

Examination Outline Cross-Reference Level RO 012 Reactor Protection Tier # 2 Group # 1 Knowledge of the effect of a loss or K/A # K3.01 malfunction of the RPS will have on the Rating 3.9 following : CRDS Question 38 The unit is operating at 100% power with all rods fully withdrawn when Vital 120 VAC Instrument Bus, NN04 is lost.

Due to the loss of NN04 __ _1)___ train of SSPS power supply is impacted and this

___2)___ cause the reactor to trip?

A. 1) B

2) should B. 1) A

2) should not C. 1) B

2) should not D. 1) A

2) should Answer: C Explanation:

A is wrong because although the B train of SSPS is impacted the reactor does not trip due to being fed power from NN02.

B is wrong because the B train of SSPS is impacted, not A. The second part is correct.

C is correct because a failure of NN04 impacts the B train of SSPS. But 48VDC power supply fed from NN03 maintain the undervoltage coils for RTB B and Bypass Breaker A energized. So the reactor does not trip.

D is wrong because the B train is impacted and also the reactor does not trip.

Technical

References:

SY1301200, Reactor Protection System, Rev. 011, Pg. 45 References to be provided to applicants during exam: None.

Learning Objective: SY1301200, Reactor Protection System, Obj. R8 Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental 3 Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)7

Examination Outline Cross-Reference Level RO 013 Engineered Safety Features Actuation Tier # 2 Group # 1 Ability to predict and/or monitor changes in K/A # A1.06 parameters (to prevent exceeding design Rating 3.6 limits) associated with operating the ESFAS controls, including: RWST level Question 39 Given:

A LOCA occurred

The reactor tripped

Safety injection actuated

Annunciator 00-047B, RWST EMPTY, is in alarm.

Which of the following describes the procedurally required action the crew should perform NEXT?

A. Refill the RWST B. Stop any pump taking suction from the RWST C. Align the containment spray system for recirculation D. Swap the CCP and SI pump suctions to the RHR pump discharge Answer: B Explanation: Annunciator 00-047B, RWST EMPTY, actuates when RWST level is 6% or less. EMG ES-12 Foldout Page directs stopping any pump taking suction from the RWST when level is less than 6%. ALR 00-047B also directs the same action when the annunciator alarms. Per BD-EMG ES-12, the reason is to protect damage to pumps taking suction from the RWST.

A is wrong because EMG ES-12 Foldout Page and ALR 00-047B require stopping any pump taking suction from the RWST when level is less than 6%. Plausible because Annunciator 00-047E, RWST LEV HILO, actuates on low level (96.9%), and ALR 00-047E directs the crew to fill the RWST. A candidate could choose this answer if he or she does not remember the setpoint associated with Annunciator 00-047B and the required actions.

B is correct. See explanation.

C is wrong because EMG ES-12 Foldout Page and ALR 00-047B require stopping any pump taking suction from the RWST when level is less than 6%. Plausible because EMG ES-12 directs taking this action when Annunciator 00-047C, RWST LEV LOLO 2, is lit (i.e., RWST level is <12%). A candidate could choose this answer if he or she does not remember the setpoint associated with Annunciator 00-047B and the required actions.

D is wrong because EMG ES-12 Foldout Page and ALR 00-047B require stopping any pump taking suction from the RWST when level is less than 6%. Plausible because EMG E-0 directs going to EMG ES-12 when Annunciator 00-047D, RWST LEV LOLO 1 AUTO XFR is lit (i.e., RWST level is <36%). A candidate could choose this answer if he or she does not remember the setpoint associated with Annunciator 00-047B and the required actions.

Technical

References:

EMG ES-12 Foldout Page, Rev 22A, Page 4/31 ALR 00-047B, RWST EMPTY, Rev 8, Page 2/2 BD-EMG ES-12, Rev 17, Page 52/76 References to be provided to applicants during exam: None.

Learning Objective: Lesson Plan LO4700600, Rev. 003, Objective 2, Determine entry conditions and required actions with alarm response procedures associated with the Emergency Core Cooling System.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis 2 10CFR Part 55 Content: 55.41(b)5

Examination Outline Cross-Reference Level RO 013 Engineered Safety Features Actuation Tier # 2 Group # 1 Ability to monitor automatic operation of the K/A # A3.01 ESFAS including : Input channels and logic Rating 3.7 Question 40 Which of the following is the correct input signal AND logic for a Safety Injection Signal?

A. High containment pressure of 2.5 psig / (2/3)

B. Low steam line pressure of 515 psig / (2/4 on 1/4 S/Gs)

C. High containment pressure of 3.5 psig / (2/4)

D. Low steam line pressure of 615 psig / (2/3 on 1/4 S/Gs)

Answer: D Explanation:

A is wrong because high containment pressure signal comes in at 3.5 psig. Two of three is correct.

B is wrong because low steam line pressure signal comes in at 615 psig. Two of four is incorrect as well as the logic is two of three.

C is wrong because the logic is two of three.

D is correct.

Technical

References:

SY1301301, Engineered Safety Features Actuation System (ESFAS), Rev. 015, Pgs. 15-16 References to be provided to applicants during exam: None.

Learning Objective: SY1301301, Engineered Safety Features Actuation System (ESFAS), Obj. 3 Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental 3 Comprehensive/Analysis

10CFR Part 55 Content: 55.41(b)7 Examination Outline Cross-Reference Level RO 022 Containment Cooling Tier # 2 Group # 1 Knowledge of CCS design feature(s) and/or K/A # K4.04 interlock(s)which provide for the following: Rating 2.8 Cooling of control rod drive motors Question 41 Given:

CRDM Cooling Fans C and D were in operation

A reactor trip and SI actuated due to a large break LOCA What is the status of 1) CTMT fan coolers and 2) CRDM cooling fans at one minute after the SI actuation?

A. 1) Running in FAST speed

2) 'C' Tripped, 'D' Running B. 1) Running in FAST speed

2) 'D' Tripped, 'C' Running C. 1) Running in SLOW speed

2) 'D' Tripped, 'C' Running D. 1) Running in SLOW speed

2) 'C' Tripped, 'D' Running Answer: C Explanation:

For (1), the four containment fan coolers start in slow speed if not already running in slow upon receipt of an SIS. If running in fast speed, they are tripped by the SIS signal and are started in slow speed by the LOCA Sequencer at +35 seconds.

For (2), the B & D CRDM Cooling Fans are load shed upon an SIS. The C CRDM is unaffected because it is on non-safety related power, and therefore it will remain running following the trip and SIS. (The A CRDM Cooling Fan was removed during the Rx vessel head modification.)

A (1) is wrong because following an SIS, the containment fan coolers start in slow speed. If they were running in fast speed, they are tripped and then started in slow speed by the LOCA Sequencer. (2) is wrong because the B and D CDRM fans are stripped, and C is running because it was running before the SI, and it is powered from a non-safety source.

Plausible if the candidate forgets that CTMT fan coolers start in slow speed and not fast speed during emergencies, or if the candidate forgets which CRDM fan is powered from the non-safety bus (or that one fan IS powered from the non-safety bus).

B (1) is wrong because following an SIS, the containment fan coolers start in slow speed. If they were running in fast speed, they are tripped and then started in slow speed by the LOCA Sequencer. Plausible if the candidate forgets that CTMT fan coolers start in slow speed and not fast speed during emergencies. (2) is correct.

C is correct. See explanation.

D (1) is correct. (2) is wrong because the B and D CDRM fans are stripped, and C is running because it was running before the SI, and it is powered from a non-safety source. Plausible if the candidate forgets that CTMT fan coolers start in slow speed and not fast speed during emergencies, or if the candidate forgets which CRDM fan is powered from the non-safety bus (or that one fan IS powered from the non-safety bus).

Technical

References:

Lesson Plan LO1302600, Rev 8, Pages 30, 34-35, and 45 of 46.

References to be provided to applicants during exam: None.

Learning Objective: Document learning objective if possible.

Question Source: Bank #187 (ID: 100500) x (note changes; attach parent) Modified Bank #

New Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis 2 10CFR Part 55 Content: 55.41(b)(7)

Examination Outline Cross-Reference Level RO 026 Containment Spray Tier # 2 Group # 1 Knowledge of the physical connections and/or K/A # K1.01 cause effect relationships between the CSS Rating 4.2 and the following systems: ECCS Question 42 The crew is mitigating a LOCA event and is transferring ECCS suction to the Containment Sumps per EMG ES-12, TRANSFER TO COLD LEG RECIRCULATION.

If Containment Spray actuates after the ECCS recirculation is aligned, what procedurally required actions should the crew take for Containment Spray?

A. Realign Containment Spray Pump suction to Containment Sump when RWST Level is <12% while maintaining Containment Spray pump running.

B. Realign Containment Spray Pump suction to Containment Sump when RWST Level is <6% while maintaining Containment Spray pump running.

C. Stop the Containment Spray Pumps, Realign suction to Containment Sump when RWST Level is <12% and re-start Containment Spray Pump.

D. Stop the Containment Spray Pumps, Realign suction to Containment Sump when RWST Level is <6% and re-start Containment Spray Pump.

Answer: A Explanation:

Per BD-EMG ES-12 Step 12, Once cold leg recirculation is established, any water remaining in the RWST is reserved for Containment Spray. If Containment Spray is not running at this time, it should remain aligned to the RWST so that this water can be utilized should spray be required. If Containment Spray is in service, the Operator will normally wait until RWST level reaches 12% before aligning spray for recirculation.

Since the ECCS swap over occurs at 36% and the transfer allowance is above 12% the Operator should allow the RWST to deplete prior to switching suction to the CTMT sumps.

Rules of usage require the Operator to loop step 12 until the RWST is 12%.

A. Correct, See above.

B. wrong but plausible. 6% corresponds to the foldout page direction of EMG ES-12 when the crew is required to stop any pump taking suction from RWST to prevent damage. Wrong since the procedurally required action is to swap suction source when RWST Level <12% per EMG ES-12, Step 12.

C. Wrong but plausible since the procedurally required action is to swap suction source when RWST level is <12% per EMG ES-12, Step 12. Wrong since stopping the pump is NOT procedurally required. Plausible because the SI pumps are stopped in EMG ES-13, Step 5 when aligning the SI for Hot Leg Recirculation. (SI pump mini-flow is isolated in EMG ES-12 since it recircs back to the RWST).

D. Wrong but plausible. The procedurally required action is to swap suction source when RWST level is <12% per EMG ES-12, Step 12, not <6%. This value corresponds to foldout page direction to stop any pump taking suction from the RWST to prevent pump damage.

Also wrong since stopping the pump is NOT procedurally required to swap suction source.

Plausible because the SI pumps are stopped in EMG ES-13, Step 5 when aligning the SI for Hot Leg Recirculation. (SI pump mini-flow is isolated in EMG ES-12 since it recircs back to the RWST).

Technical

References:

BD-EMG ES-12, Rev 15, Pages 29, 30 EMG ES-12, Rev 22A, Pages 4 and 19 EMG ES-13, Rev 13, Page 10 SY1302600, Rev 19, Pages 18 and 19.

References to be provided to applicants during exam: None.

Learning Objective: SY1302600, Containment Spray and Cooling Systems, Obj. R6 Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental 2 Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)8

Examination Outline Cross-Reference Level RO 039 Main and Reheat Steam Tier # 2 Group # 1 Ability to monitor automatic operation of the K/A # A3.02 MRSS, including: Isolation of the MRSS Rating 3.1 Question 43 While cooling down to cold shutdown conditions in accordance with GEN 00-006, HOT STANDBY TO COLD SHUTDOWN, the following plant conditions existed:

RCS Temperature 487°F

RCS Pressure 1890 psig

CTMT Pressure 0.2 psig

PZR Level 35%

S/G Pressures: A: 623 psig B: 621 psig C: 625 psig D: 620 psig Forty-five seconds later, the following plant conditions exist:

RCS Temperature 472°F

RCS Pressure 1810 psig

CTMT Pressure 3.1 psig

PZR Level 17%

S/G Pressures A: 530 psig B: 525 psig C: 522 psig D: 523 psig Which of the following identifies the expected plant response and the reason?

A. MSIVs will close due to low steamline pressure B. The safety injection pumps will start due to low PZR pressure C. MSIVs will close due to steamline high pressure - negative rate D. The safety injection pumps will start due high containment pressure Answer: C Explanation: The steamline high pressure - negative rate signal closes the MSIVs when a rate of 100 psig in 50 seconds is exceeded. S/G C pressure has dropped 102 psig in 45 seconds, which should cause the MSIVs to close. Also, prior to commencing the cooldown IAW GEN 00-006, the crew would have manually blocked SI signals when P-11 activated at 1920 psig.

A is wrong because MSIVs will close due to the steamline high pressure - negative rate signal. Plausible because the MSIVs close on low steamline pressure at 615 psig, and all steamline pressures have dropped below this setpoint; however, the low steamline pressure signal is blocked below 1970 psig (P-11).

B is wrong because the SI pumps will NOT start due to low PZR pressure (i.e., 1830 psig) because below 1970 psig (P-11), automatic SI is blocked. Therefore, the SI pumps will not start even though pressure has fallen below the SI actuation setpoint of 1830 psig. Plausible if the candidate does not realize that P-11 prevents the SI actuation from occurring.

C is correct. See explanation.

D is correct because the SI pumps will NOT start because the high containment pressure setpoint is 3.5 psig, and this pressure has not yet been exceeded. Plausible if the candidate does not remember the high containment pressure setpoint for SI actuation.

Technical

References:

LO1503900, Rev 11, Pages 35 and 36/52 LO1301200, Rev. 12, Page 30/49 References to be provided to applicants during exam: None.

Learning Objective: LO1503900, Rev 11, Objective 5, Discuss the operation of the Main Steam Isolation Valves.

Question Source: Bank #

Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis 3 10CFR Part 55 Content: 55.41(b)(5)

Examination Outline Cross-Reference Level RO 059 Main Feedwater (MFW) System Tier # 2 Group # 1 Knowledge of the effect that a loss or K/A # K3.03 malfunction of the MFW will have on the Rating 3.5 following: K3.03 S/GS Question 44 With the unit operating at full power

1) Which of the following failures of AE PT-508, Main Feed Header Pressure Channel, should cause an INITIAL DROP in feedwater flow to all S/G's?

2) What action should the crew take in accordance with OFN SB-008, INSTRUMENT MALFUNCTIONS to mitigate the failure?

A. 1) Fails LOW

2) Take manual control of Main Feedwater pump speed B. 1) Fails HIGH

2) Take manual control of Main Feedwater pump speed C. 1) Fails LOW

2) Take manual control of Main Feedwater Regulating valves D. 1) Fails HIGH

2) Take manual control of Main Feedwater Regulating valves Answer: B Explanation:

A is wrong, but plausible because failure LOW will cause main feed pump speed to RISE to restore program differential pressure for the main feed pump. This is the opposite response and is plausible if the student confuses what inputs to the main feed pump speed and what affect it would have on the system.. The OFN Action is correct, OFN SB-008, ATT B, Step B1.

B is correct because this is an input to the feed pump speed control circuit so when this fails HIGH the MFP will slow down causing an initial SG level drop. To fix this per OFN SB-008 take manual control of the MFP speed controller per OFN SB-008, ATT B, Step B1.

C is wrong, but plausible because the failure LOW will cause a main feed pump to RISE to restore program differential pressure for the main feed pump. This is the opposite response and is plausible if the student confuses what inputs to the main feed pump speed and what affect it would have on the system.. This is also the wrong action since OFN SB-008, ATT B, Step B1 directs taking manual control of MFP speed controller and NOT taking manual control of individual Main Feed Water Reg Valves. Plausible since this is an action that would be taken per OFN SB-008 for different failures within the S/G Water Level Control System.

D is wrong but plausible because this is the failure HIGH will cause the MFP to slow down causing an initial SG level drop. but the OFN will have the BOP take manual control of the pump not each feed reg valve. This is also wrong action since OFN SB-008, ATT B, Step B1 directs taking manual control of MFP speed controller and NOT taking manual control of individual Main Feed Water Reg Valves. Plausible since this is an action that would be taken per OFN SB-008 for different failures within the S/G Water Level Control System.

Technical

References:

OFN SB-008, Instrument Malfunctions, Rev. 44A, page 10 References to be provided to applicants during exam: None.

Learning Objective: SY1505900, Main Feedwater System, Rev 25, Objective R11 Question Source: Bank #

Opposite Plant Response, Different Modified Bank #98358 X Correct Answer New Question History: Last NRC Exam Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis 3 10CFR Part 55 Content: 55.41.(b).7

Examination Outline Cross-Reference Level RO 059 Main Feedwater Tier # 2 Group # 1 Ability to manually operate and monitor in the K/A # A4.03 control room: Feedwater control during power Rating 2.9 increase and decrease Question 45 When performing a unit startup in accordance with GEN 00-003, HOT STANDBY TO MINIMUM LOAD, the crew should transfer from the MFRV bypass valves to the MFRVs prior to exceeding a MAXIMUM of _____ power.

A. 20%

B. 25%

C. 30%

D. 35%

Answer: C Explanation: GEN 00-003 says, Per USAR 10.4.7.2.3 the swap between the MFR Bypass Valves and the MFRV's is to be performed between 20 and 30% power. Power increase should continue while swapping Feed Reg Valves as this improves the capability to maintain Steam Generator levels in band. If the swap is NOT completed by 30% power, then power increased is to be stopped.

A is wrong because GEN 00-003 directs the swap to commence at 20% power and says power may increase while swapping (as long as power does not increase above 30% before the swap is complete). Plausible because on a downpower, reduction in power cannot continue if the swap from MFRVs to MFR bypass valves has not occurred. A candidate could confuse the power limit on a down power and an up power.

B is wrong because GEN 00-003 directs the swap to commence at 20% power and says power may increase while swapping (as long as power does not increase above 30% before the swap is complete). Plausible because GEN 00-003 directs securing feedwater heating at 25%, which is the step in the procedure following swapping of the valves, and a candidate could confuse these actions.

C is correct. See explanation.

D is wrong because GEN 00-003 says to stop the power increase if the swap has not occurred by the time power is at 30%. Plausible because 35% is associated with another startup milestone (i.e., starting two condensate pumps between 30% and 35%), and a candidate could remember this in place of when to swap the valves.

Technical

References:

GEN 00-003, HOT STANDBY TO MINIMUM LOAD, Rev 99, Page 29/103 References to be provided to applicants during exam: None.

Learning Objective: LO4710202, R2: Shift from the Main Feedwater bypass valves to the Main Feedwater Regulating valves IAW GEN 00-003.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental 2 Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(7)

Examination Outline Cross-Reference Level RO 061 Auxiliary / Emergency Feedwater Tier # 2 Group # 1 Knowledge of the effect of a loss or K/A # K6.01 malfunction of the following will have on the Rating 2.5 AFW components: K6.01 Controllers and positioners Question 46 Given:

The Governor on the Terry Turbine failed to the full steam flow position, causing an overspeed trip.

Trip and throttle valve, FC HV-312, has been reset.

Which of the following is a procedurally allowed alternate method of controlling TDAFP speed?

A. Control speed at the Local Control Panel B. Transfer control to the Aux Shutdown Panel C. Manually throttle with the Trip & Throttle Valve, FC HV-312 D. Locally throttle with the local Steam Supply Valves, AB V-085/087 Answer: C Explanation:

The turbine has its own integral oil system with a shaft-driven oil pump and cooler that serve the turbine bearings and Governor Control Oil System. The drains from the two bearings connect to a large common drain on the same side of the turbine as the cooler (west side). A loss of Governor oil would cause the TDAFWP to overspeed as the oil pressure is used to close the Governor.

Turbine speed is controlled by a governor system that controls FC FV-312 (Governor Control Valve). Turbine speed is normally set at 3850 rpm and may be controlled from either the Main Control Board or from the ASP. FC FV-312 is left in MANUAL both on the MCB and ASP. Auto does not function.

Due to the type of operator on AB HV-5 and 6 the valves can only be closed locally or returned to a neutral position. The operator can be damaged if local opening is attempted.

The governor valve springs to open position upon loss of oil or power.

Control of the governor valve is accomplished either at the B ASP using HIK-313B with RP HIS-1 in ISOLATE or the Main Control Board using HIK-313A with RP HIS-1 in REMOTE.

Both controllers are the same. A slide bar (Joystick) is adjusted to control turbine speed. A remote light on the Main Control Board speed controller tells the operator when control is selected to Remote or Isolate. The upper limit for TDAFW speed is set using the joystick on the controller. Turbine speed cannot be raised above 3924 rpm. If the joystick is held in the raise position, indicated speed may momentarily increase above this value but will immediately return.

A is wrong to control speed at the local control panel, but plausible, since TDAFW Pump Control PANEL FC219 was recently installed. Wrong because there are no controls to throttle FC HV-312 B is wrong to transfer control to the Aux shutdown panel. but plausible since the transfer switches still exist on ASP. Incorrect because the governor has failed to full open C. Correct - see above D wrong but plausible as AB HV-085/087 have local opening capability. Incorrect since 085/087 cannot be throttled procedurally.

Meets the K/A because Question establishes AFW system operations and a failed turbine controller.

RO level because AFW system design Technical

References:

SY1406100, Auxiliary Feedwater System, Revision 29, pages 16, 25 SYS AL120, Rev 53, Attachment A, Page 27 References to be provided to applicants during exam: None.

Learning Objective: SY1406100, objective R1 Question Source: Bank 112168# X (note changes; attach parent) Modified Bank #

New Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental 2 Comprehensive/Analysis 10CFR Part 55 Content: 55.41.(b).7

Examination Outline Cross-Reference Level RO 062 AC Electrical Distribution Tier # 2 Group # 1 Ability to manually operate and/or monitor in K/A # A4.04 the control room: Local operation of breakers Rating 2.6 Question 47 Given:

OFN KJ-032, LOCAL EMERGENCY DIESEL STARTUP, is in progress.

EDG A has just been started by removing the break glass cover from the EMERGENCY START pushbutton.

The EDG A output breaker is OPEN.

Bus NB01 voltage is 0 Volts.

Which of the following operator actions should be performed NEXT in accordance with OFN KJ-032?

A. Close the EDG A output breaker locally B. Place the Master Transfer Switch to LOC/MAN C. Close the EDG A output breaker in the control room D. Place NE HIS-25, NB01 Emergency Supply Breaker handswitch, to the TRIP position Answer: B Explanation: OFN KJ-032, Step 4, directs resetting/clearing the EDG output breaker anti-pumping relay locally by placing the Master Transfer Switch to LOC/MAN for at least 4 seconds, and then placing the Master Transfer Switch to AUTO when the EDG has started and the output breaker is still open.

A is wrong because OFN KJ-032, Step 4, directs resetting/clearing the EDG output breaker anti-pumping relay locally to close the breaker. Plausible because the breaker should be closed, and it is possible to close the breaker locally (once the anti-pumping relay has been reset). Also, OFN KJ-032, Step 5 directs closing the breaker locally if the anti-pumping relay has been reset, no lockouts are on the bus, the normal and alternate supply breakers are open, and bus NB01 is deenergized.

B is correct. See explanation.

C is wrong because OFN KJ-032, Step 4, directs resetting/clearing the EDG output breaker anti-pumping relay locally to close the breaker. Plausible because the breaker can be closed from the control room (once the anti-pumping relay has been reset).

D is wrong because OFN KJ-032, Step 4, directs resetting/clearing the EDG output breaker anti-pumping relay locally to close the breaker. Plausible because OFN NB-030 directs

resetting the anti-pumping relay by placing NE HIS-25 to trip momentarily when the EDG is running, and the output breaker has not closed automatically.

D is correct. See explanation.

Technical

References:

OFN KJ-032, Rev 13B, Page 4/17 References to be provided to applicants during exam: None.

Learning Objective: Document learning objective if possible.

Question Source: Bank # Question ID# X 98796 (The stem has been changed to Modified Bank #

remove information about resetting the anti-pumping relay to test whether the candidate realizes that action is required during local operation when the EDG output breaker has not closed and the EDG is running.)

New Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis 3 10CFR Part 55 Content: 55.41(b)(7)

Examination Outline Cross-Reference Level RO 063 DC Electrical Distribution Tier # 2 Group # 1 Knowledge of bus power supplies to the K/A # K2.01 following: K2.01 Major DC Loads Rating 2.9 Question 48 A ground has occurred, rendering Panel NK51 inoperable. Which of the following loads is no longer available:

A. Pressurizer PORV 455A B. Pressurizer PORV 456A C. Train A Reactor Trip Breaker Control D. Train B Reactor Trip Breaker Control Answer: A Explanation:

A is correct since PORV 455A is powered from Panel NK51, from bus NK01 B is wrong but plausible since PORV 456 A is powered from Panel NK44, from bus NK04 C is wrong but plausible since Train A Reactor Trip Breaker is powered from Panel NK41, from bus NK01 D is wrong but plausible since Train B Reactor Trip Breaker is powered from Panel NK54, from bus NK04 Technical

References:

SY1506300, DC and Instrument Power, Class 1E, pages 28 and 30 References to be provided to applicants during exam: None.

Learning Objective: SY 1506300, objective R1 Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental 2 Comprehensive/Analysis 10CFR Part 55 Content: 55.41.(b).7

Examination Outline Cross-Reference Level RO 064 Emergency Diesel Generator Tier # 2 Group # 1 Knowledge of ED/G system design feature(s) K/A # K4.02 and/or interlock(s) which provide for the Rating 3.9 following: Trips for ED/G while operating (normal or emergency)

Question 49 Following a normal start of the emergency diesel generator, which of the following conditions should cause an automatic shutdown of the engine?

A. Overcurrent B. Reverse power C. Under frequency D. Generator differential Answer: D Explanation: A generator differential overcurrent condition will cause the engine to trip. The other conditions listed will result in the EDG output breaker opening, but they are not engine trips.

A is wrong because overcurrent does not trip the diesel engine. Plausible because it is a generator trip and results in the opening of the EDG output breaker. Therefore, a candidate may think this condition results in an engine trip.

B is wrong because the reverse power trip never sends a diesel engine shutdown signal.

Plausible because it is a generator trip and results in the opening of the EDG output breaker.

Therefore, a candidate may think this condition results in an engine trip.

C is wrong because the under frequency trip never sends a diesel engine shutdown signal.

Plausible because it is a generator trip and results in the opening of the EDG output breaker.

Therefore, a candidate may think this condition results in an engine trip.

D is correct. See explanation.

Technical

References:

Lesson Plan LO 1406400, Rev 15, Page 55/65 Lesson Plan LO 1406401, Rev 0, Page 22 and 23/57 References to be provided to applicants during exam: None.

Learning Objective: Lesson Plan LO 1406401, Objective 2, Discuss the instrumentation and controls associated with the Emergency Diesel Generator System.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental 2 Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(7)

Examination Outline Cross-Reference Level RO 064 Emergency Diesel Generator Tier # 2 Group # 1 Ability to predict and/or monitor changes in K/A # A1.03 parameters (to prevent exceeding design Rating 3.2 limits) associated with operating the ED/G system controls including: Operating voltages, currents, and temperatures Question 50 During implementation of OFN NB-030, LOSS OF AC EMERGENCY BUS NB01 (NB02),

cooling water is lost to the Emergency Diesel.

With the diesel running unloaded, the MAXIMUM time required to stop it prior to a ___1)___

trip is ___2)___ minutes A. 1) high jacket water temperature

2) 3 B. 1) high engine lube oil temperature

2) 3 C. 1) high jacket water temperature

2) 30 D. 1) high engine lube oil temperature

2) 30 Answer: C Explanation:

A is wrong because part 2 is incorrect. 3 minutes is plausible because OFN-NB-030 foldout page requires any Emergency Diesel running loaded to be unloaded if cooling flow can not be established is 3 minutes. Part 1 is correct.

B is wrong because both parts are incorrect. Part 1 is plausible because high engine lube oil temperature is a valid engine condition should cooling water be lost but incorrect because there is not an Emergency Diesel trip associated with it. Part 2 is plausible (see A above).

C is correct because both parts are correct. For part 1, ALR-501 identifies high jacket water temperature as one of the causes for a diesel engine shutdown (high engine lube oil is not identified). For part 2, OFN-NB-030 foldout page states if an Emergency Diesel is running unloaded and cooling water can not be established within 30 minutes, then the diesel is stopped.

D is wrong because part 1 is incorrect but plausible (see B above). Part 2 is correct.

Technical

References:

OFN-NB-030, LOSS OF AC EMERGENCY BUS NB01 (NB02), Rev 33A, page 4 ALR-501, STANDBY DIESEL ENGINE SYSTEM CONTROL PANEL KJ-121 , Rev 26, page 10 References to be provided to applicants during exam: None.

Learning Objective: LO1406400 Obj 8, 12 Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental 3 Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)4

Examination Outline Cross-Reference Level RO 073 Process Radiation Monitoring Tier # 2 Group # 1 Knowledge of the purpose and function of K/A # 2.1.28 major system components and controls. Rating 4.1 Question 51 Which of the following identifies ALL of the control functions performed by Steam Generator Blowdown Effluent Monitor BM RE-52 when the Hi Hi setpoint is reached?

A. Close Sample Isolation Valves BM HV-5, 6, 7, and 8 Close Blowdown Isolation Valves BM HV-1, 2, 3, and 4 B. Close Upper Sample Valves BM HV 19, 20, 21 and 22 Close Lower Sample Valves BM HV-35, 36, 37 and 38 Close Blowdown Sample Isolation BM HV 65, 66, 67, and 68 C. Close Sample Isolation Valves BM HV-5, 6, 7, and 8 Close Blowdown Isolation Valves BM HV-1, 2, 3, and 4 Stop the SGBD Discharge Pumps PBM01A and B Close Steam Generator Blowdown Discharge Valve BM LV-56 D. Close Upper Sample Valves BM HV 19, 20, 21 and 22 Close Lower Sample Valves BM HV 35, 36, 37 and 38 Close Blowdown Sample Isolation BM HV 65, 66, 67, and 68 Stop the SGBD Discharge Pumps PBM01A and B Close Steam Generator Blowdown Discharge Valve BM LV-56 Answer: C Explanation: Lesson Plan LO1503800, Rev 8, Page 27, says that Steam Generator Blowdown Process Monitor, BM RE-52, measures the liquid effluent from the steam generator blowdown discharge pumps for radioactivity when discharging to the Coffey County Lake. This monitor will actuate a Blowdown and Sample Process Isolation Signal (BSPIS) and close BM HV-1, 2, 3, 4 CTMT isolation valves and BM HV-5, 6, 7, 8 sampling isolation valves, close BM LV-56, and secure the SGBD discharge pumps on high radiation.

A is wrong because it does not list all of the functions performed when RE-52 senses hi hi radiation. Plausible because Steam Generator Blowdown (SGBD) Sampling Monitor, SJ RE-02, will actuate a Blowdown and Sample Process Isolation Signal (BSPIS) ONLY, which will cause BM HV-1, 2, 3, 4 and BM HV-5, 6, 7, 8 sampling isolation valves to close, and a candidate may confuse the functions of RE-02 and RE-52.

B is wrong because it does not list all of the functions performed when RE-52 senses hi hi radiation, and it also lists functions that are NOT performed by RE-52 on high radiation.

Plausible because a Steam Generator Blowdown Sample Isolation Signal (SGBSIS) will perform these functions (as well as closing BM HV-1-4 and 5-8, which are not included in this distractor to avoid a psychometric flaw) when either AFAS-M, SIS, or undervoltage on NB01

or NB02 occur. A candidate may confuse the actions that occur on a SGBSIS with a BSPIS and think that RE-52 performs functions associated with a SGBSIS, but knows that RE-.

C is correct.

D is wrong because it does not list all of the functions performed when RE-52 senses hi hi radiation, and it also lists functions that are NOT performed by RE-52 on high radiation.

Plausible if a candidate confuses the valves that are closed by SGBSIS and BSPIS, but knows that RE-52 hi radiation will stop the discharge pumps and close the discharge valve.

Technical

References:

Lesson Plan LO1503800, Rev 8, Page 27 and 39/52 References to be provided to applicants during exam: None.

Learning Objective: LO1407300, Rev 11, Objective 4: Discuss the function of the radiation monitors in the Process and Effluent Radiation Monitoring System, including any protective interlocks.

Question Source: Bank #

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New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental 2 Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(7)

Examination Outline Cross-Reference Level RO 076 Service Water Tier # 2 Group # 1 Knowledge of the physical connections and/or K/A # K1.19 cause- effect relationships between the SWS Rating 3.6 and the following systems: SWS emergency heat loads Question 52 With the unit at 100% power, the crew started A ESW pump to perform SYS EF-300, ESW SERVICE WATER MACROFOUL TREATMENT.

Annunciator 055A, ESW PMP A PRESS LO, is LIT

ESW Pump Discharge pressure is 105 psig

CCW HX Outlet temperatures are RISING

Annunciator 060E, CTMT SUMP A/B LEV HI, is LIT

Containment temperature is slowly rising Which of the following is causing the Containment conditions?

A. CCW Thermal Barrier leak B. Containment Cooler supply line break C. Steam leak inside Containment D. RCS leak inside Containment Answer: B Explanation:

A is wrong because there are no CCW surge tank/Process Rad Monitor alarms to indicate CCW leakage. CCW temps would be rising on a TB leak with no effect on Ctmt temp.

B is correct (Containment Cooler supply line break) because ESW pressure is below the alarm setpoint, which indicates system leakage as a possible cause of the low header pressure. That, coupled with containment sump alarm and increasing temperature indicates that the leakage is from ESW supply line to coolers.

C is wrong because Sump alarms would be in for a steam leak condensation, containment temps would be rising on a steam leak, and these do not explain why ESW pressure is low.

D is wrong (RCS leak inside Containment) because there are no rad monitor readings/alarms to indicate a RCS leak. CTMT temperature rising could indicate a primary leak, however without a corresponding change in humidity or radiation an RCS leak is not indicated. Also doesn't account for low ESW Pressure.

Technical

References:

ALR 00-055A, Rev 11A, OFN EF-33, Rev 18, LO1732443, Rev 7.

References to be provided to applicants during exam: None.

Learning Objective: LO1732443 R2/R3 Question Source: Bank # X (note changes; attach parent) Modified Bank #

New Question History: Last NRC Exam WC-2011, Q26 Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis 3 10CFR Part 55 Content: 55.41(b)8

Examination Outline Cross-Reference Level RO 078 Instrument Air Tier # 2 Group # 1 Knowledge of the effect that a loss or K/A # K3.01 malfunction of the IAS will have on the Rating 3.1 following: Containment air system Question 53 While operating at power, an extended loss of instrument air occurs. Which of the following identifies plant response if NO operator action is taken?

A. PZR pressure will rise; charging pump suction will shift to the RWST B. PZR pressure will rise; charging pump suction will remain aligned to the VCT C. PZR pressure will remain stable; charging pump suction will shift to the RWST D. PZR pressure will remain stable; charging pump suction will remain aligned to the VCT Answer: A Explanation: Lesson Plan LO1407800, Rev 11, Page 22, says, All instrument air to Containment is supplied through air-operated valve KA FV-29. This valve fails closed on a loss of instrument air and automatically closes on a Containment Isolation Signal Phase A (CISA).

Lesson Plan LO1301000, Rev. 009, Page 46, says, Upon a complete loss of instrument air, plant equipment and valves begin to fail to their safe position below 70 psig air pressure.

Without operator action, once pressure in the air header bleeds down sufficiently, letdown will isolate due to letdown orifice isolation valves failing closed. Charging will slowly back off due to level control, but not fast enough to prevent depletion of VCT level. Auto makeup to the VCT fails when boric acid and reactor makeup water valves fail close, resulting in a swap over to the RWST supplying the charging header. Charging without letdown has been filling the pressurizer and increasing primary pressure. PZR spray valves will open to control RCS pressure until air pressure is low enough that the spray valves fail closed. Air pressure will continue to bleed off, failing open the normal-closed Regenerative HX to RCS Cold Leg valve and the NCP discharge control valve, thereby maximizing charging to the pressurizer.

A is correct. See explanation.

B is wrong because VCT will lower to the setpoint that will result in a shift of CCP suction to the RWST. Plausible because initially on loss of IA, charging flow will initially lower because of letdown isolation, and a candidate may think either that VCT level will lower but not reach the setpoint of the swap to the RWST suction or not realize that reactor makeup water valves fail closed on loss of IA, and there will not be any automatic makeup to the VCT.

C is wrong because pressure will rise as a result of PZR level rising above the setpoint for automatic operation of the PZR heaters (they will turn on and raise pressure). Plausible

because initially, PZR spray valves will be able to operate, but eventually will fail closed when air pressure falls below 70 psig.

D is wrong because pressure will rise as a result of PZR level rising above the setpoint for automatic operation of the PZR heaters (they will turn on and raise pressure). Plausible because initially, PZR spray valves will be able to operate, but eventually will fail closed when air pressure falls below 70 psig. It is also wrong because VCT will lower to the setpoint that will result in a shift of CCP suction to the RWST. Plausible because initially on loss of IA, charging flow will initially lower because of letdown isolation, and a candidate may think either that VCT level will lower but not reach the setpoint of the swap to the RWST suction or not realize that reactor makeup water valves fail closed on loss of IA, and there will not be any automatic makeup to the VCT.

Technical

References:

Lesson Plan LO1407800, Rev 11, Page 22 Lesson Plan LO1301000, Rev 9, Page 46 References to be provided to applicants during exam: None.

Learning Objective: Lesson Plan LO1301000, Rev. 009, Objective 11: Describe system and integrated plant response to transient and equipment failures.

Question Source: Bank #

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New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis 3 10CFR Part 55 Content: 55.41(b)(7)

Examination Outline Cross-Reference Level RO 103 Containment Tier # 2 Group # 1 Knowledge of the physical connections and/or K/A # K1.02 cause effect relationships between the Rating 3.9 containment system and the following systems: Containment isolation/containment integrity.

Question 54 What is the basis for Component Cooling Water (CCW) System's valve alignment for a Containment Isolation Phase 'B' actuation?

A. Minimizes the consequences of and/or terminates the mass and energy releases associated with high containment pressure.

B. Ensures CCW System is NOT an additional potential radioactive release path from containment.

C. Reduces heat load on the CCW System by eliminating unnecessary cooling requirements.

D. Ensures the CCW System can still meet its design cooling function for loads within containment during a Design Basis Accident LOCA.

Answer: B Explanation:

A is wrong but plausible since this is why main steamlines are isolated when containment pressure exceeds Hi-2 17 psig pressure setpoint.

B is correct per BD-EMG E-0, Step F12, Containment isolation phase B valves are closed to isolate additional potential release paths from containment.

C is wrong but plausible since several CCW valves reposition with SIS signal for this reason.

Wrong because the purpose of repositioning for CIS Phase B is to isolate potential leak paths from containment for the containment integrity challenge that exists when CTMT Pressure rises to 27 psig setpoint.

D is wrong but plausible since several CCW valves reposition with SIS signal for this reason.

Wrong because the purpose of repositioning for CIS Phase B is to isolate potential leak paths from containment for the containment integrity challenge that exists when CTMT Pressure rises to 27 psig setpoint.

Technical

References:

SY1303200, rev 20, page 18.

BD-EMG E-0, Rev 27, Page 95 BD-EMG FR-Z1, Rev 7, Page 18.

References to be provided to applicants during exam: None.

Learning Objective: LO1303200 R7 Question Source: Bank #116325 X (note changes; attach parent) Modified Bank #

New Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental 3 Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)9

Examination Outline Cross-Reference Level RO 103 Containment Tier # 2 Group # 1 Ability to (a) predict the impacts of the K/A # A2.03 following malfunctions or operations on the Rating 3.5 containment system and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Phase A and B isolation Question 55 The crew has just declared Train A of Containment Isolation Signal Phase A (CISA)

INOPERABLE due to a failed surveillance of the automatic actuation logic.

If a Loss of Coolant Accident that required Safety Injection Actuation were to occur, 1) which procedure would be used to ensure containment was properly isolated, and 2) which component will the crew have to manually isolate?

A. 1) EMG E-0, REACTOR TRIP OR SAFETY INJECTION

2) Containment atmosphere radiation monitors B. 1) EMG E-0, REACTOR TRIP OR SAFETY INJECTION

2) Hydrogen analyzers C. 1) EMG E-1, LOSS OF REACTOR OR SECONDARY COOLANT

2) Hydrogen analyzers D. 1) EMG E-1, LOSS OF REACTOR OR SECONDARY COOLANT

2) Containment atmosphere radiation monitors Answer: A Explanation:

1) EMG E-0, Attachment F, Step F3 verifies proper alignment of CIS A valves and directs actions to reposition those valves by manual actuation of CISA signal or manually repositioning the valves using EMG E-0, Attachment B.

2) EMG E-0, Attachment B lists both answer choices. Applicants have to analyze given components and apply system knowledge to determine Containment Atmosphere Radiation Monitors are in service during normal operation while Hydrogen analyzers are not normally inservice and are not placed in service until directed to do so using EMG procedure direction after CISA signal is reset.

A is correct. See explanation.

B 1) is correct, 2) is wrong but plausible since the crew will transition to EMG E-1 for the given LOCA. Step 19 of EMG E-1 places the Hydrogen Analyzer in service after resetting CISA Signal in Step 5.

C 1) is wrong but plausible in that the crew will transition to EMG E-1 for the given LOCA, wrong since the crew will have addressed the failure of automatic isolation of CISA at Step F3 of EMG E-0, ATTACHMENT F.2) is wrong, but plausible since the crew will transition to EMG E-1 for the given LOCA. Step 19 of EMG E-1 places the Hydrogen Analyzer in service after resetting CISA Signal is Step 5.

D 1) is wrong but plausible in that the crew will transition to EMG E-1 for the given LOCA, wrong since the crew will have addressed the failure of automatic isolation of CISA at Step F3 of EMG E-0, ATTACHMENT F.2) is correct.

Technical

References:

EMG E-0, Rev 39, Attachments B, and Page 63/99 and F, Page 81/99.

LO1301301, Rev 11, Page 20/51 References to be provided to applicants during exam: None.

Learning Objective: LO1301301, Rev 11, Objective 7: Describe system and integrated plant response to transient and equipment failures.

Question Source: Bank #

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New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis 3 10CFR Part 55 Content: 55.41(b)(5)

Examination Outline Cross-Reference Level RO 001 Control Rod Drive Tier # 2 Knowledge of the following operational Group # 2 implications as they apply to the CRDS: K/A # K5.07 Effects of an asymmetric rod configuration on Rating 3.3 power distribution Question 56 Unit power is steady at 70% as the Reactor Operator withdraws control rods for temperature control.

Based on these conditions, you would expect Axial Flux Distribution to be ____1)_____

because there is more power being produced in the _____2)_____.

A. 1) positive

2) lower section of the core B. 1) negative

2) lower section of the core C. 1) positive

2) upper section of the core D. 1) negative

2) upper section of the core Answer: B Explanation:

A is wrong because AFD would be more negative with rods inserted and more power would be towards the bottom of the core.

B is correct (see A above) more negative with more power shifted to the lower section of the core for this event C is wrong because (see A above)

D is correct because (see A above)

Technical

References:

SY1301501, rev 15, page 61.

References to be provided to applicants during exam: None.

Learning Objective: LO1301501 R12 Question Source: Bank #

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New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis 3 10CFR Part 55 Content: 55.41(b)5

Examination Outline Cross-Reference Level RO 011 Pressurizer Level Control Tier #

Group #

Knowledge of the effect that a loss or K/A # K3.02 malfunction of the PZR LCS will have on the Rating 3.5 following: RCS Question 57 Given:

The Unit is steady at 30% power.

PZR Heaters are in Auto.

Control Rods are in MANUAL

Loop 1 Tavg instrument BB TI-412 fails to 630°F.

Which of the following identifies the expected plant response?

A. PZR level will RISE, and all PZR backup heaters will turn ON B. PZR level will RISE, and all PZR backup heaters will remain OFF C. PZR level will not change, and all PZR backup heaters will turn ON D. PZR level will not change, and all PZR backup heaters will remain OFF Answer: B Explanation: At 30% power, Tref is 566F (per OFN SB-008, Page 22). Tavg equals Tref because turbine load and power have not changed. PZR program level is 36%, and PZR actual level would be at the program level. When the Loop 1 Tavg instrument fails high, PZR program level changes to the full load program level (57%) because program level is determined by auctioneered high Tavg. The deviation between program level and actual level is (-)21%.

Lesson Plan LO1301000, Rev 9, Page 24/55 says, The master controller compares actual level with programmed level (derived from auctioneered high Tavg above no-load) and transmits any resulting error signal via a PI controller. The level error signal causes charging flow to increase or decrease by repositioning a flow control valve for the centrifugal charging pumps (CCP) BG FCV-121 or the Normal Charging Pump (NCP) BG FCV-462. Therefore, PZR level will rise.

Page 28/55, says, Pressurizer level error uses bistables for plus or minus five percent

(+/-5%) level deviation signals. The plus five percent bistable operates an annunciator alarm and turns on all backup heaters. This ensures in-surge from the RCS is heated to saturation conditions. The minus five percent bistable operates the annunciator alarm only. Therefore, PZR backup heaters will not be energized.

A is wrong because backup heaters will be off. Plausible because for a level deviation where program level is 5% or more less than actual level, heaters will energize, which is the

opposite response, and a candidate may select this distractor if he or she confuses the level deviation.

B is correct. See explanation.

C is wrong because charging flow will increase and PZR level will rise in response to program level being higher than actual level, and backup heaters will not be on. Plausible because if the same instrument failure occurred at 100% power, program level is constant above the full load Tavg setpoint, and there would be no level deviation and no increase in charging flow. Plausible because for a level deviation where program level is 5% or more less than actual level, heaters will energize, which is the opposite response, and a candidate may select this distractor if he or she confuses the level deviation.

D is wrong because charging flow will increase and PZR level will rise in response to program level being higher than actual level, and backup heaters will not be on. Plausible because if the same instrument failure occurred at 100% power, program level is constant above the full load Tavg setpoint, and there would be no level deviation and no increase in charging flow.

Technical

References:

Lesson Plan LO1301000, Rev 9, Page 24 and 28 of 55 References to be provided to applicants during exam: None.

Learning Objective: Document learning objective if possible.

Question Source: Bank #

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New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis 2 10CFR Part 55 Content: 55.41(b)(7)

Examination Outline Cross-Reference Level RO 015 Nuclear Instrumentation Tier # 2 Group # 2 Ability to recognize system parameters that K/A # 2.2.42 are entry-level conditions for Technical Rating 3.9 Specifications Question 58 The reactor is MODE 1 at 75% power. The CRS has just entered Technical Specification (TS) LCO 3.2.3 for AFD (Axial Flux Distribution).

1) What is the MINIMUM number of excore channels outside limits that require TS entry?

2) What is MAXIMUM amount of time allowed by TS to reduce power after TS entry?

A. 1) One of four.

2) 30 minutes B. 1) Two of four.

2) 60 minutes.

C. 1) One of four.

2) 60 minutes.

D. 1) Two of four.

2) 30 minutes.

Answer: D Explanation:

A is wrong because the first part is incorrect.

B is wrong because the second part is incorrect.

C is wrong because two channels are required to be outside limits and thermal power must be reduced to less than 50% within 30 minutes.

D is correct.

Technical

References:

SY1301501, Excore Nuclear Instrumentation System, Rev. 015, Pg. 61 Technical Specifications 3.2.3, AFD ALR 00-079D, Rev 10, Page 1 References to be provided to applicants during exam: None.

Learning Objective: SY1301501, Excore Nuclear Instrumentation System, Obj. R13

Question Source: Bank #

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New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental 3 Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)10

Examination Outline Cross-Reference Level RO 017 In-Core Temperature Monitor Tier # 2 Group # 2 Knowledge of ITM system design feature(s) K/A # K4.01 and/or interlock(s) which provide for the Rating 3.4 following: Input to subcooling monitors Question 59 Several minutes following a steam line break, the following parameters are indicated on the MCBs or NPIS.

Loop 1 WR Thot - 517°F

Loop 2 WR Thot - 517°F

Loop 3 WR Tcold - 514°F

Loop 4 WR Tcold - 514°F

Pzr Press BBPT-0455 - 1932 psig

Pzr Press BBPT-0457 - 1932 psig

WR RCS Press BBPT-0405 - 1990 psig

1st Hottest Ch A CETC - 541°F

2nd Hottest Ch A CETC - 539°F

3rd Hottest Ch A CETC - 538°F

4th Hottest Ch A CETC - 537°F

5th Hottest Ch A CETC - 535°F What should the 'A' train MCB subcooling monitor read?

A. 98°F Subcooled B. 113°F Subcooled C. 115°F Subcooled D. 118°F Subcooled Answer: C Explanation:

A. (98°F Subcooled) is wrong, but plausible. This is the result when using the wrong pressure and the average of 5 CETC values (1990 psig 2005 psia 636°F - 538°F =

98°F) This is plausible since some procedures direct Operators to use the average of the 5 hottest CETCs (for example in Step 13 of EMG E-3) Wrong since the Thermocouple/Core Cooling Monitor (TC/CCM) uses RTD auctioneered high input for MCB Subcooling meter indication. Additionally wrong since the lowest pressure is used in subcooling calculations.

B. (113F Subcooled) is wrong, but plausible. This is the result of using the right pressure, but converting to psia wrong while using the right temperature (1932 psig 1917 psia 630°F - 517°F = 113°F)

C. (115 F Subcooled) is correct. Using the lowest pressure (1932 psig) gives a saturation temperature of 632F. The hottest loop temperature (517F) gives a subcooling of 115°F.

(1932 psig 1947 psia 632°F - 517°F = 115°F)

D. (118°F Subcooled) is wrong, but plausible. This is the result when using the correct pressure, but the lower loop temperatures (1932 psig 1947 psia 632°F - 514°F= 118°F)

Technical

References:

Lesson plan LO1301700, Rev 7, Page 26 References to be provided to applicants during exam: None.

Learning Objective: Lesson Plan LO1301700, Objective 4: Explain the operation of the Incore Thermocouple System major components.

Question Source: Bank #

(note changes; attach parent) Modified Bank #116261 X New Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis 2 10CFR Part 55 Content: 55.41(b)(7)

Examination Outline Cross-Reference Level RO 033 Spent Fuel Pool Cooling System (SFPCS) Tier # 2 Group # 2 Ability to predict and/or monitor changes in K/A # A1.02 parameters (to prevent exceeding design Rating 2.8 limits) associated with Spent Fuel Pool Cooling System operating the controls including: A1.02 Radiation monitoring systems Question 60 Fuel handling operations are in progress. As spent fuel pool level falls below its MINIMUM required level, high radiation alarms alarm at a MINIMUM of 1) mrem/hr on the area radiation monitors at the 2) elevation.

A. 1) 100

2) 2047-6 B. 1) 15

2) 2047-6 C. 1) 100

2) 2032-0 D. 1) 15

2) 2032-0 Answer: B Explanation:

A is wrong but plausible because 2.5 mrem per hour is the setpoint for SD RE-35 and SD RE-36, New Fuel Storage Area Rad Monitors, The 2nd part is correct as SD RE-37 and 38 are the area radiation monitors that will alarm and are located on the 2047-6 elevation.

B is correct as 15 mrem per hour the alarm setpoint for high radiation level, and SD RE-37 or 38, the correct area radiation monitors and they are located on the 2047-6 elevation.

C is wrong but plausible because 2.5 mrem per hour is the setpoint for SD RE-35 and SD RE-36, New Fuel Storage Area Rad Monitors. These monitors are on the 2032 level, not the 2047-6 elevation which are the area rad monitors that will alarm with lowering SFP level..

D is wrong because even though 15 mrem per hour is the alarm setpoint for high radiation level,on SD RE-37 and SD RE-38 , these monitors are physically located on the 2047-6 elevation. Plausible since new fuel storage area radiation alarms are located on the 2032 level.

Technical

References:

SY1403300, Fuel Pool Cooling and Cleanup System, Rev. 20, Section 5.1 (page 24, rad monitors) and Section 11 (page 35, radiation levels), Updated Safety Analysis Report, Section 9.1.2.5

References to be provided to applicants during exam: None.

Learning Objective: LO1407200, objectives T1 and R2.

Question Source: Bank #

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New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental 3 Comprehensive/Analysis 10CFR Part 55 Content: 55.41.(b).5

Examination Outline Cross-Reference Level RO 045 Main Turbine Generator Tier # 2 Group # 2 Ability to monitor automatic operation of the K/A # A3.11 MT/G system, including: Generator trip Rating 2.6 Question 61 The crew should expect a 30-second time delay for automatic opening of the generator output breakers after an automatic turbine trip resulting from which of the following conditions?

A. High bearing vibration B. High level in the MSR C. Low condenser vacuum D. Low bearing oil pressure Answer: B Explanation: LO1504502, Rev 6, Page 21, says, The generator vital trip circuit responds to turbine trips on low condenser vacuum, journal bearing high vibration, abnormal thrust bearing position (wear), and the low bearing oil pressure. Vital trips still need to meet the Reverse Power and Steam Valve Sequential Logic Circuit requirements, but vital trips do bypass the 30-second time delay (for opening the generator output breakers) to rapidly stop the turbine-generator unit. Leaving the Main Generator loaded will not help slow down the Turbine. When synched to the grid, the grid will maintain Turbine speed (even as a motor).

Therefore, the 30-second time delay is bypassed and the Generator Output Breakers open as soon as the Reverse Power and Sequential Logic circuits are met The generator non-vital trip circuit responds to all other turbine trips and delays generator tripping for 30 seconds after onset of motoring.

LO1504800, Rev 8, Page 24, lists high MSR level as a turbine trip that is not a vital trip, and therefore, there will be a 30 second time delay for opening the generator output breakers.

A is wrong because the high vibration turbine trip is a vital trip, and the 30 second time delay is bypassed. Plausible if the candidate does not know that this trip bypasses the time delay.

B is correct.

C is wrong because low condenser turbine trip is a vital trip, and the 30 second time delay is bypassed. Plausible if the candidate does not know that this trip bypasses the time delay.

D is wrong because low bearing oil pressure turbine trip is a vital trip, and the 30 second time delay is bypassed. Plausible if the candidate does not know that this trip bypasses the time delay.

Technical

References:

LO1504800, Rev 8, Page 24 LO1504502, Rev 6, Page 21 References to be provided to applicants during exam: None.

Learning Objective: Document learning objective if possible.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental 2 Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(7)

Examination Outline Cross-Reference Level RO 068 Liquid Radwaste Tier # 2 Group # 2 Knowledge of the effect of a loss or K/A # K6.10 malfunction on the following will have on the Rating 2.5 Liquid Radwaste System : Radiation monitors Question 62 Which of the following will result in an automatic closure of HB RV-18, Liquid Radwaste Discharge Valve?

A. 120 VAC Panel HF-132 loses power B. Liquid Radwaste Radiation Monitor, HB-RE-18, fails high C. One Circulating Water Pump trips D. Service Water dilution flow is secured Answer: B Explanation:

A is wrong because RV-18 is powered from 125 VDC. Plausible as this is the power supply for RV-45, Secondary Liquid Waste Discharge Valve.

B is correct because the liquid radwaste radiation monitor will send a signal to isolate RV-18 upon a high alarm C is wrong because RV-18 closes automatically on the loss of all circ water pumps.

Plausible if applicant believes auto closure occurs on loss of one pump.

D is wrong because when service water is used for dilution flow, there is no automatic isolation function. Plausible because a manual isolation is required in this case.

Technical

References:

LO1406904, page 9 of 10 References to be provided to applicants during exam: None.

Learning Objective: LO1406904 3: Explain how releases to the environment are terminated automatically and the criteria for manual termination.

Question Source: Bank #

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New X Question History: Last NRC Exam No

Question Cognitive Level: Memory/Fundamental 2 Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)11

Examination Outline Cross-Reference Level RO 071 Waste Gas Disposal Tier # 2 Group # 2 Ability to manually operate and/or monitor in K/A # A4.14 the control room: WDGS status alarms Rating 2.8 Question 63 If Radwaste Building Radiation Monitor, GH RE-10A is in Accident Isolate mode, then the monitor title on SP056A, Digital Radiation Monitor Panel will be A. Half Intensity Cyan B. White C. Magenta D. Dark Blue Answer: A Explanation: Lesson Plan LO1408804, Rev 4, Page 24, explains that GH RE-10A monitors particulate and iodine, and GH RE-10B monitors gas activity. Page 24 also says, GH RE-10B provides gaseous activity indication A high activity alarm on this channel will cause GH RE-10A to go into a purge mode, also referred to as accident isolate... When GH RE-10A is in the purge mode, the Chemistry surveillance requirements for continuous monitoring cannot be met. The monitoring requirement is for projecting radiation levels at or beyond the Site Boundary. This condition is also noted on the SP056A by displaying the monitor title in a gray colorA spike on GH RE-10B has caused GH RE-10A to go into a purge mode on occasion. It is important for operators in the Control Room and Radwaste to be aware of indications for this occurrence. At Wolf Creek during a gaseous release in 1995, GH RE-10A went into accident isolate upon receipt of an Alert (Process Radiation High) alarm on GH RE-10B, only minutes after the release was commenced. Almost 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months passed before it was noticed that the monitor was not reset, during an awareness tour.

Also, SYS SP-121, Rev 28, Page 49, lists the various colors displayed in SP056A and the meaning of those colors. Note that half intensity cyan is displayed as gray, which is how the color is described in the ILO lesson plan.

A is correct. See explanation.

B is wrong because the title will be gray, not white. Plausible because white is a color used on SP056A to indicate the monitor is offline, and a candidate may think white indicates the monitor is in accident isolate mode (i.e., no longer sampling particulate and iodine).

C is wrong because the title will be gray, not magenta. Plausible because magenta is a color used on SP056A to indicate communications failure, and a candidate may think magenta indicates the monitor is in accident isolate mode (i.e., no longer sampling particulate and iodine).

D is wrong because the title will be gray, not dark blue. Plausible because dark blue is a color used on SP056A to indicate channel failure, and a candidate may think dark blue indicates the monitor is in accident isolate mode (i.e., no longer sampling particulate and iodine).

Technical

References:

Lesson Plan LO1408804, Rev 4, Page 24 References to be provided to applicants during exam: None.

Learning Objective: Document learning objective if possible.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental 2 Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(7)

Examination Outline Cross-Reference Level RO 075 Circulating Water Tier # 2 Group # 2 Ability to (a) predict the impacts of the K/A # A2.03 following malfunctions or operations on the Rating 2.5 circulating water system; and (b) based on those predictions, use Procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Safety features and relationship between condenser vacuum, turbine trip, and steam dump Question 64 The unit is at 25% power when a circulating water malfunction occurs and condenser vacuum degrades to 8.0 in-HgA .

1) What automatically occurs from this malfunction?

2) Select the next action that is procedurally required to mitigate this event.

A. 1) BOTH a turbine trip and reactor trip;

2) Verify both tripped and enter EMG E-0 REACTOR TRIP OR SAFETY INJECTION B. 1) ONLY a turbine trip;

2) Verify turbine tripped and enter OFN MA-001 LOAD REJECTION OR TURBINE TRIP C. 1) ONLY steam dumps are lost;

2) Verify C-9, COND AVAILABLE, is out and enter OFN-AF 025 UNIT LIMITATIONS D. 1) BOTH a turbine trip and steam dumps are lost;

2) Manually trip the reactor and enter EMG E-0 REACTOR TRIP OR SAFETY INJECTION Answer: D Explanation:

A is wrong because the turbine trip for low vacuum is at 7.5 in-HgA but the reactor does not auto trip on turbine trip until >50% (P9).

B is wrong because although the turbine trips the steam dumps are also lost due to C-9 at 5 in-HgA and since reactor is less than 30% power, the crew is required to manually trip the reactor and enter EMG E-0, per ONF AF-025, ATT F, Step F2..

C is wrong because both dumps are lost and turbine is tripped. Procedure is also wrong.

D is correct because the turbine trip for low vacuum is at 7.5 in-HgA and because the steam dumps are lost when vacuum reaches 5 in-HgA. Per OFN AF-025, ATT F, Step F2, the crew is required to manually trip the reactor since power is <30% at this vacuum per FIGURE 2; therefore EMG E-0 is required to be entered.

Technical

References:

OFN-MA-001, LOAD REJECTION OR TURBINE TRIP, revision 25 OFN-AF-025, UNIT LIMITATIONS, revision 50 EMG-E-0, REACTOR TRIP OR SAFETY INJECTION, revision 39 References to be provided to applicants during exam: None.

Learning Objective: LO1732411 Obj. R1 and R5 Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis 4 10CFR Part 55 Content: 55.41(b)5

Examination Outline Cross-Reference Level RO 056 Condensate Tier # 2 Group # 2 Knowledge of the physical connections and/or K/A # K1.03 cause-effect relationships between the Rating 2.6 Condensate System and the following systems: MFW Question 65 Given:

The crew is responding to a loss of A Main Condensate Pump

The unit is at 90% power.

Main Feedwater Pump Suction Pressures are each at 330 psig.

B and C Main Condensate Pump Motor Currents are each at 430 amps.

Which of the following actions should the crew procedurally perform NEXT?

A. Continue reducing power until <62% (760 MWE).

B. Verify all Low Pressure Heater Strings are unisolated.

C. Continue reducing power until Main Feedwater Pumps Suction Pressure >340 psig.

D. Continue reducing power until Main Condensate Pump Motor Current is <420 amps.

Answer: C Explanation:

Per OFN AF-025, UNIT LIMITATIONS, ATTACHMENT A, Page 3.of 7, for the loss of one Main Condensate Pump, reduce power to 90%, MFP suction pressure >340 psig and Main Condensate Pump current <440 amps.

A is wrong, but plausible in that this is the maximum power level allowed if only one MFP is in service.

B is wrong, but plausible in that this is a required action per OFN AF-025 if Reactor power were >65% with a loss of one heater drain pump.

C is correct, see answer explanation above.

D is wrong but plausible in that there is a specified maximum current loading limit of 440 amps, which is met for the given conditions...

Technical

References:

OFN AF-025, Rev 52, page 17 of 52.

Lesson Plan LO1505600, Rev 9, Page 39/42 References to be provided to applicants during exam: None.

Learning Objective: Lesson Plan LO1505600, Rev 9, Objective 10, Describe system and integrated plant response to transient and equipment failures.

Question Source: Bank #

(note changes; attach parent) Modified Bank # X LO134526 New Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis 3 10CFR Part 55 Content: 55.41(b)(4)

Examination Outline Cross-Reference Level RO Tier # 3 Knowledge of conduct of operations Group # 1 requirements. K/A # 2.1.1 Rating 3.8 Question 66 According to AP 21-001, CONDUCT OF OPERATIONS, Control Room Operators should, at a MINIMUM, perform main control board (MCB) walk-downs __________.

A. twice per shift not including shift turnover B. once per hour not including shift turnover C. twice per shift including shift turnover D. once per hour including shift turnover Answer: A.

Explanation:

A is correct because the Conduct of Ops directs MCB walkdowns should be performed at least twice per shift by each Control Room Operator not including shift turnover.

B is wrong because the minimum frequency is twice per shift. Plausible if applicant believes the Conduct of Ops directs more frequent walkdowns.

C is wrong because turnover is not included.

D is wrong because the minimum frequency is twice per shift. Plausible if applicant believes the Conduct of Ops directs more frequent walkdowns. Also, turnover is not included.

Technical

References:

AP 21-001, Page 31 References to be provided to applicants during exam: None.

Learning Objective: None - General Knowledge of AP 21-001, Conduct of Operations.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental 3 Comprehensive/Analysis

10CFR Part 55 Content: 55.41(b)10 Examination Outline Cross-Reference Level RO Conduct of Operations Tier # 3 Group # n/a Knowledge of procedures and limitations K/A # 2.1.36 involved in core alterations. Rating 3.0 Question 67 According to GEN 00-009, REFUELING, when moving control rods in containment, the crew should verify refueling pool water level is at a MINIMUM of at least 23 feet ___1)___.

While monitoring the Main Control Board Wide Range Loop Level indicators, the channels should indicate within a MINIMUM of___2)___ of each other.

A. 1) Over the top of the reactor vessel flange

2) 10 inches Correct answer changed from B to C during post-exam B. 1) Over the top of the reactor vessel flange comments. See Exam Report.

2) 3 inch C. 1) Above the top of the fuel assemblies in the reactor vessel

2) 10 inches D. 1) Above the top of the fuel assemblies in the reactor vessel

2) 3 inch Answer: C Explanation: GEN 00-009, Rev 39, in multiple locations states refueling pool water level must be at least 23 feet above the top of the fuel assemblies. Precaution and Limitation 4.7.2 specifies NR Loop Level channels should indicate within 3 inch of each other if using NPIS and 10 inches of each other if using Main Control Board indicators.

A 1) is wrong because refueling pool water level must be a minimum of at least 23 feet above the fuel assemblies when moving control rods. 2) is correct See explanation.

B 1) is correct 2) is wrong but plausible in that 3 inches is required range if using NPIS.

C 1) is wrong because refueling pool water level must be a minimum of at least 23 feet above the fuel assemblies when moving control rods. 2) is correct D 1) is wrong because refueling pool water level must be a minimum of at least 23 feet above the fuel assemblies moving control rods. (2) is wrong but plausible in that 3 inches is required range is using NPIS.

Technical

References:

GEN 00-009, REFUELING, Rev 39, References to be provided to applicants during exam: None.

Learning Objective:

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental 3 Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(10)

Examination Outline Cross-Reference Level RO Tier # 3 Ability to identify and interpret diverse Group # 1 indications to validate the response of another K/A # 2.1.45 indication. Rating 4.3 Question 68 During post-accident conditions, while evaluating an elevated response from Source Range Nuclear Instruments, the Reactor Operator should look at _ 1)_ _, which can be used to CONFIRM the existence of ___2)___.

A. 1) RVLIS indication

2) fuel relocation to the downcomer B. 1) RVLIS indication

2) coolant voiding or core uncovery C. 1) subcooling margin

2) coolant voiding or core uncovery D. 1) subcooling margin

2) fuel relocation to the downcomer Answer: B Explanation:

A is wrong since RVLIS would NOT confirm existence of fuel relocation to the downcomer.

Plausible since if RVLIS showed the reactor vessel to be full, then the elevated counts would be due to either a loss of shutdown margin, or fuel damage.

B is Correct if RVLIS indication is used to detect voiding/uncovery in the core. A value <45%

is used in EMG F-0, CRITICAL SAFETY FUNCTION STATUS TREES for Core Cooling Safety Function Evaluation C is wrong because subcooling margin <30F, along with rising SR counts would be indicative of coolant voiding or the core becoming uncovered, along with RVLIS indication lowering..

Subcooling margin >30F would ELIMINATE the existence of coolant voiding or core uncovery, not confirm it.

D is wrong because a reduction in subcooling margin along with a rise in SR counts would not ALONE be used to confirm fuel relocation to the downcomer. Plausible as a fuel relocation is a possible reason for rising SR counts.

Meets generic requirements because the question is not asking specifics of nuclear instrument overlap ranges. The question is asking what to do with indications.

Technical

References:

LO1610711, pages 5-8 References to be provided to applicants during exam: None.

Learning Objective: LO1610711 T1: Explain the response of nuclear instrumentation in a post-accident situation.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis 4 10CFR Part 55 Content: 55.41(b)1

Examination Outline Cross-Reference Level RO Equipment control Tier # 3 Group # n/a Knowledge of tagging and clearance K/A # 2.2.13 procedures. Rating 4.1 Question 69 According to AP 21E-001, CLEARANCE ORDERS, all of the following can be credited for meeting Tags Plus EXCEPT for a A. Breaker with a lock hasp B. Breaker containing D-rings C. Valve in series with another valve D. Valve located in Containment when in MODE 3 Answer: B Explanation: AP 21E-001, CLEARANCE ORDERS, Rev 40B, Page 24/85, lists examples for meeting Tags Plus. It says, 120 volt breakers inside electrical breaker panels may contain D-rings to allow secure DNO tag placement. The D-ring prevents use of Tags plus.

120 volt breakers containing D-rings are exempted from the tags plus requirement.

A is wrong because a breaker with a restraining device such as a lock hasp is a way to meet Tags Plus.

B is correct. See explanation.

C is wrong because two valves in series is a way to meet Tags Plus.

D is wrong because inaccessible location (e.g., in Containment when Containment is closed for Mode 1-4) is a way to meet Tags Plus.

Technical

References:

AP 21E-001, CLEARANCE ORDERS, Rev 40B, Page 24/85 References to be provided to applicants during exam: None.

Learning Objective:

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No

Question Cognitive Level: Memory/Fundamental 2 Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(10)

Examination Outline Cross-Reference Level RO Tier # 3 Knowledge of conditions and limitations in the Group # 1 facility license. K/A # 2.2.38 Rating 3.6 Question 70 The plant is in the following condition:

The reactor is shutdown with fuel in the vessel

Average coolant temperature is 210°F The reactor is in MODE __ 1)_ _. In this condition, Wolf Creek Technical Specifications, Section 5.2 Organization, requires that a MINIMUM of _ _(2)__ _ nuclear station operators be assigned to the unit staff.

A. 1) 5

2) 1 B. 1) 5

2) 2 C. 1) 4

2) 1 D. 1) 4

2) 2 Answer: D Explanation:

A is wrong because the plant is in Mode 4. Plausible if applicant believes that less than boiling temperature is Mode 5. With fuel in the reactor and in mode 4, two NSOs must be assigned. Plausible as only 1 NSO is required in Mode 5 with fuel in the vessel.

B is wrong because the plant is in Mode 4. Plausible if applicant believes that less than boiling temperature is Mode 5.

C is wrong because two ROs are required in Mode 4. Plausible if applicant believes that a reduction to 1 NSO is allowed in Mode 4 instead of Mode 5.

D is correct because the reactor is in Mode 4 and two NSOs are required.

Technical

References:

Technical Specifications Table 1.1-1 and Section 5.2 References to be provided to applicants during exam: None.

Learning Objective: LO1732700 4: List the six operational modes as defined in Technical Specifications.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental 3 Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)10

Examination Outline Cross-Reference Level RO Radiation Control Tier # 3 Group # n/a Knowledge of radiation exposure limits under K/A # 2.3.4 normal or emergency conditions. Rating 3.2 Question 71 Given:

An Operator is planning to perform a task in a location with a general dose rate of 181 mrem/hour.

The Operator needs to transit through an area with a general dose rate of 12 Rem/hour for 1 minute. He will return using the same path after he completes the task.

The Operator has an accumulated annual dose of 557 mrem (all from WCGS).

Which of the following is the MAXIMUM time, the Operator has to complete the task without exceeding any administrative limits?

A. 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months B. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months C. 11 hours1.273148e-4 days 0.00306 hours 1.818783e-5 weeks 4.1855e-6 months D. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Answer: A Explanation: The annual administrative limit is 2000 mrem for WCNOC exposure. The operator has already accumulated 557 mrem this year; therefore, the operator can receive 1443 mrem during this task and not exceed the administrative limit.

For this task, the candidate must account for the total transit time and the task performance time. The total transit time is 2 minutes in an area of 12,000 mrem/60min, which will provide 400 mrem. 1443 mrem - 400 mrem = 1043 mrem that can be received during the task.

Therefore, the operator can stay in the location to perform the task for 1043 mrem/181 mrem/hour = 5.76 hours8.796296e-4 days 0.0211 hours 1.256614e-4 weeks 2.8918e-5 months , which rounded to the nearest hour is 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months to not exceed any exposure limits.

A is correct. See explanation.

B is wrong because the max time is 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months to not exceed any administrative limits.

Plausible because 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is the value that results from rounding the answer up. (However, if the operators stays for 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , then the administrative limit will be exceeded.)

C is wrong because the max time is 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months . Plausible because 3000 mrem/year is the annual administrative dose limit to include exposure earned at other sites, and 11.28 hours3.240741e-4 days 0.00778 hours 4.62963e-5 weeks 1.0654e-5 months is the time calculated if the candidate uses 3000 mrem as the exposure limit.

D is wrong because the max time is 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months . Plausible because 3000 mrem/year is the annual administrative dose limit to include exposure earned at other sites, and 11.28 hours3.240741e-4 days 0.00778 hours 4.62963e-5 weeks 1.0654e-5 months is the time calculated if the candidate uses 3000 mrem as the exposure limit, and the candidate may round the answer up to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Technical

References:

AP 25B-100, RADIATION WORKER GUIDELINES, Rev 50, Page 12 References to be provided to applicants during exam: None.

Learning Objective: LO1733204. Rev 11, Obj. 1, Discuss the requirements of procedure AP 25B-100, Radiation Worker Guidelines as pertaining to the responsibilities of rad.

workers, exposure limits, and contamination controls.

Question Source: Bank #

(Similar to Q71 from the 2015 WC NRC Modified Bank # X (ID 116331)

Exam and Question ID 116331 in the exam bank)

New Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis 2 10CFR Part 55 Content: 55.41(b)(12)

Examination Outline Cross-Reference Level RO Tier # 3 Knowledge of radiation monitoring systems, Group #

such as fixed radiation monitors and alarms, K/A # 2.3.15 portable survey instruments, personnel Rating 2.9 monitoring equipment, etc.

Question 72

1) Which of the following monitors is considered part of the RCS Leakage Detection system in accordance with Technical Specification 3.4.15?

2) Which channel(s) of this monitor (particulate, iodine, gas) are REQUIRED to meet the Technical Specifications for RCS leakage detection?

A. 1) GT RE-59 Containment High Range Radiation Monitor;

2) particulate ONLY B. 1) GT RE-31 Containment Atmosphere Monitor;

2) particulate and gas C. 1) GT RE-59 Containment High Range Radiation Monitor;

2) particulate and gas D. 1) GT RE-31 Containment Atmosphere Monitor;

2) particulate ONLY Answer: D Explanation:

A is wrong because GT-RE-59 is not part of the RCS leakage detection system, GT RE-31 is correct. For the second part, both the particulate and gas channels are part of the RCS leak detection system per the lesson plan however the above the line info only includes the particulate portion of the monitor and the bases confirms that to meet the LCO requirements, only the particulate portion is required.

B is wrong because the second part is wrong (see above for A explanation).

C is wrong because the first and second parts are wrong (see A above).

D is correct because GT RE-31 is the correct monitor. For the second part, both the particulate and gas channels are part of the RCS leak detection system per the lesson plan however the above the line info in the TS only includes the particulate portion of the monitor and the bases confirms that to meet the LCO requirements, only the particulate portion is required.

Technical

References:

TS 3.4.15, page 3.4-38, amendment 212 LO14 073 00, revision 11, page 26.

References to be provided to applicants during exam: None.

Learning Objective: LO14 073 00, Obj 4 and 9 Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental 3 Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)12

Examination Outline Cross-Reference Level RO Emergency procedures/plan Tier # 3 Group # n/a Knowledge of the organization of the operating K/A # 2.4.5 procedures network for normal, abnormal and Rating 3.7 emergency evolutions Question 73 Given:

An event occurred on the unit

The crew entered EMG FR-P1, RESPONSE TO IMMINENT PRESSURIZED THERMAL SHOCK due to a red path on the Integrity Status Tree

The crew is at Step 3, Try to Stop the RCS Cooldown, and is throttling feedwater flow

A RED path on the Heat Sink Status Tree now appears Which of the following describes the proper use of procedures for these conditions?

A. Continue throttling feedwater flow until the RCS cooldown stops, and then transition to EMG FR-H1, RESPONSE TO LOSS OF SECONDARY HEAT SINK, if the red path on the Heat Sink Status Tree still exists.

B. Perform EMG FR-P1 RESPONSE TO IMMINENT PRESSURIZED THERMAL SHOCK and EMG FR-H1, RESPONSE TO LOSS OF SECONDARY HEAT SINK concurrently to address both red path conditions.

C. Stop throttling feedwater flow, and then transition to EMG FR-H1, RESPONSE TO LOSS OF SECONDARY HEAT SINK, if the red path on the Heat Sink Status Tree still exists.

D. Complete EMG FR-P1, RESPONSE TO IMMINENT PRESSURIZED THERMAL SHOCK prior to evaluating the Heat Sink Status Tree.

Answer: D Explanation: Administrative Procedure AP 15C-003, Page 18/60, states, In procedure EMG C-21, UNCONTROLLED DEPRESSURIZATION OF ALL STEAM GENERATORS and EMG FR-P1, RESPONSE TO IMMINENT PRESSURIZED THERMAL SHOCK, it is possible that steam generator levels are below the narrow range and total feed flow is throttled to less than the minimum flow required to maintain a heat sink. These actions are taken to limit RCS cooldown and therefore, do not represent a true loss of heat removal capability. For this reason, the "higher priority" procedure EMG FR-H1, RESPONSE TO LOSS OF SECONDARY HEAT SINK, should not be performed.

A is wrong because the red path on the heat sink tree exists as a result of actions the crew is taking to stop the RCS cooldown per FR-P1. AP 15C-003 contains direction to not perform the higher priority FR-H1 because there isnt a true loss of heat sink. Plausible because the candidate may know that stopping the RCS cooldown by continuing to throttle feed flow is the correct action to do, and may also think that it is necessary to then transition to FR-H1 if the red path still exists because heat sink is a higher priority safety function.

B is wrong because the red path on the heat sink tree exists as a result of actions the crew is taking to stop the RCS cooldown per FR-P1. AP 15C-003 contains direction to not perform the higher priority FR-H1 because there isnt a true loss of heat sink. Also, EMG F-0, Foldout Page, directs the operators to implement the FRP for the higher priority safety function (vs performing them concurrently). Plausible if the candidate knows that the crew will need to perform something differently than it normally would (i.e., NOT perform the higher priority FRP in this instance), but does not know exactly what the correct use of procedures is for this instance, and therefore selects concurrent implementation.

C is wrong because the red path on the heat sink tree exists as a result of actions the crew is taking to stop the RCS cooldown per FR-P1. AP 15C-003 contains direction to not perform the higher priority FR-H1 because there isnt a true loss of heat sink. It is also wrong because there is no direction in FR-P1 to stop throttling the feedwater flow if there is a red path on the heat sink tree. Plausible because the candidate may realize that the crews action per FR-P1 caused the red path, and may want to stop doing the action to restore the higher priority safety function.

D is correct. See explanation.

Technical

References:

AP 15C-003, PROCEDURE USER'S GUIDE FOR ABNORMAL PLANT CONDITIONS, Rev 34, Page 18/60 References to be provided to applicants during exam: None.

Learning Objective: LO1732329, Revision 015, Objective 9, Discuss the priority of procedures that deal with abnormal conditions IAW AP 15C-003.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis 3 10CFR Part 55 Content: 55.41(b)(10)

Examination Outline Cross-Reference Level RO Tier # 3 Knowledge of operator response to loss of all Group #

annunciators. K/A # 2.4.32 Rating 3.6 Question 74 The unit is at full power, when the following occurs:

Main Control Board Annunciator 0-17B, PK01/02 PK03/04 TROUBLE illuminates.

The power failure event results in a loss of 72% of the annunciators.

To determine the affected panel the Operator would ______1)________ and then dispatch the field operator to the affected PK panel to perform the appropriate response procedure

_______2)______.

A. 1) use the NPIS computer

2) ALR 401 for loss of PK01 B. 1) use the cross reference table that lists alarms lost versus PK bus in ALR-00-017B

2) ALR 401 for loss of PK01 C. 1) use the NPIS computer

2) ALR 402 for loss of PK02 D. 1) use the cross reference table that lists alarms lost versus PK bus in ALR-00-017B

2) ALR 402 for loss of PK02 Answer: C Explanation:

A is wrong because although the NPIS computer is used to determine that it is PK02 (ie first part is correct), the second part is incorrect because the table in the OFN-PK-029 lists PK02 as causing the biggest loss of annunciators, specifically loss of PK-52, and this is also OE for WC.

B is wrong because you have to go to the NPIS first to determine that PK02 is the problem, The cross reference table is not contained in the ALR and this table is contained in the OFN-PK-25 not the ALR. With operator knowledge you might guess it is PK-52 but to verify you would need the NPIS check, then dispatch to the ALR 402 for loss of PK-52, this is the correct local alarm procedure for this event, so the second part is also incorrect with PK01 (it causes around 42% loss of annunciators via loss of PK-51).

C is correct because see explanation for A above (both first and second parts are correct).

D is wrong because the first part is wrong, the second part is correct.

Technical

References:

OFN-PK-029, revision 25, (pages 59-60)

ALR-00-017B, rev 5B References to be provided to applicants during exam: None.

Learning Objective: LO1732439 Obj. R6 Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental 3 Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)10

Examination Outline Cross-Reference Level RO

4. Emergency Procedures / Plan Tier # 3 Group # 4 Knowledge of RO tasks performed outside the K/A # 2.4.34 main control room during an emergency and Rating 4.2 the resultant operational effects.

Question 75 Following a fire in the control room that results in evacuation, OFN RP-017, CONTROL ROOM EVACUATION, directs the reactor operator to perform which of the following tasks?

A. Start the TDAFW Pump to provide feed flow to S/G B B. Locally Trip all RCPs C. Perform breaker lineup on NG04C D. Perform breaker lineup for select NK loads Answer: D Explanation: OFN RP-017, Appendix C, lists Reactor Operator Actions. Steps 1-3 are memory action steps. Step C2 provides a list of NK disconnects that must be taken to OFF position.

A is wrong because in OFN RP-017 does not direct the reactor operator to start the TDAFW Pump. Plausible because the OFN RP-017, Appendix A, lists SRO actions, and one of those actions (Steps A8 through A17) is for the SRO to start the TDAFW pump and establish feed flow to S/G B.

B is wrong because in OFN RP-017 does not direct the reactor operator to trip all RCPs.

Plausible because the OFN RP-017, Appendix B, lists Turbine Building Actions, and one of those actions (Step B1) is for the Operator to Locally Trip RCPs.

C is wrong because OFN RP-017 does not direct the reactor Operator to reposition loads on NG04C. Plausible the OFN RP-017, Appendix D lists the Auxiliary Building actions and one of those actions (Step D4) is to perform breaker lineup on NG04C.

D is correct. See explanation.

Technical

References:

OFN RP-017, Rev 49, Pages 15-17, 19, 15 and 36 References to be provided to applicants during exam: None.

Learning Objective: LO1732427, Rev 13, Examine the available options for procedure actions.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental 3 Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(10)

Examination Outline Cross-Reference Level SRO 000009 Small Break LOCA Tier # 1 Group # 1 Ability to determine or interpret the following K/A # EA2.11 as they apply to a small break LOCA: Rating 4.1 Containment temperature, pressure, and humidity Question 76 According to the Technical Specification (TS) Basis document for LCO 3.6.6, Containment Spray and Cooling Systems, in post-accident operation following an actuation signal, the Containment Cooling System fans are designed to automatically shift to _ __1) ___ speed in order to __ 2)_ __.

A. 1) slow

2) prevent motor overload from the higher mass atmosphere B. 1) fast

2) limit post-accident temperature in containment to less than the design value C. 1) slow

2) reduce loading on the diesel generators in the event of a loss of offsite power D. 1) fast

2) limit post-accident pressure in containment to less than the design value Answer: A Explanation:

A is correct because the TS basis states that motor overload is the concern. SRO-only due to required knowledge of TS basis.

B is wrong because the fans shift to slow. Plausible if applicant believes that high speed is required for a LOCA containment temp control.

C is wrong because the concern is motor overload. Plausible if applicant believes that increased fan loading is a diesel loading concern.

D is wrong because the fans shift to slow. Plausible if applicant believes that high speed is required for a LOCA containment pressure control.

Technical

References:

TS 3.6.6 and basis References to be provided to applicants during exam: None.

Learning Objective: TS Lectures T1: Given a set of conditions determine Technical Specification Operability.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental 3 Comprehensive/Analysis 10CFR Part 55 Content:

55.43(b)(2)

Examination Outline Cross-Reference Level SRO 022 Loss of Reactor Coolant Makeup Tier # 1 Group # 1 Ability to determine and interpret the following K/A # AA2.01 as they apply to the Loss of Reactor Coolant Rating 3.8 Makeup: Whether charging line leak exists Question 77 Given:

The unit is in MODE 1 at 100% power

Normal 75 gpm letdown is in service

Seal injection is 8 gpm to each RCP

Pressurizer level is lowering slowly

Charging flow has risen to 120 gpm

VCT level is lowering

Regen Heat Exchanger Letdown outlet temperature has risen

Containment sump level and containment pressure are normal

No action has been taken by the crew Which of the following states the location of the leak in progress and the action, if any, the crew may need to take in accordance with the Emergency Plan?

(Reference Provided)

A. There is a leak in the charging line. No EAL will need to be declared because the leak can be isolated successfully.

B. There is a leak in the RCS. No EAL will need to be declared because the leak can be isolated successfully.

C. There is a leak in the charging line. The Emergency Plan will need to be entered for loss of reactor coolant boundary.

D. There is a leak in the RCS. The Emergency Plan will need to be entered for loss of reactor coolant boundary.

Answer: A Explanation: The rise in regen heat exchanger letdown outlet temperature indicates the leak is in the charging line, which is isolable. Step 8 of OFN BB-007, RCS LEAKAGE HIGH would direct the crew to isolate charging and letdown, and the leak would stop. No EAL will need to be declared because leak isolation will be successful.

A is correct. See explanation.

B is wrong because the rise in regen heat exchanger letdown outlet temperature indicates the leak is in the charging line, not the RCS. The second part of the distractor is true.

Plausible because a leak in the RCS would result in PZR level lowering and charging flow rising, and it is possible that a candidate might not correctly interpret that the rise in regen heat exchanger letdown outlet temperature indicates the leak is in the charging line.

C is wrong because Step 8 of OFN BB-007, RCS LEAKAGE HIGH would direct the crew to isolate charging and letdown, and the leak would stop. Per EAL-3 in APF 06-002-01, if leak isolation is successful, then no EAL has been exceeded.

D is wrong because the rise in regen heat exchanger letdown outlet temperature indicates the leak is in the charging line, not the RCS. Plausible because a leak in the RCS would result in PZR level lowering and charging flow rising, and it is possible that a candidate might not correctly interpret that the rise in regen heat exchanger letdown outlet temperature indicates the leak is in the charging line. Also plausible because the leak is >25GPM and

>10GPM, which could result in a NOUE being declared if the leak cannot be isolated by taking actions in accordance with OFN BB-007.

Technical

References:

EAL-3 in APF 06-002-01, Rev 17A, Page 8/37 Lesson Plan LO4710522, Rev 16, Page 10/15 References to be provided to applicants during exam: APF 06-002-01 Learning Objective: Document learning objective if possible.

Question Source: Bank # 108381 (with X some edits)

(note changes; attach parent) Modified Bank New Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis 2 10CFR Part 55 Content: 55.43(b)(5)

Examination Outline Cross-Reference Level SRO 000029 ATWS Tier # 1 Group # 1 Knowledge of the emergency action level K/A # 2.4.41 thresholds and classifications. Rating 4.6 Question 78 The plant was operating at 100% power when a valid SSPS actuation occurred. The reactor failed to trip automatically. The reactor operators take immediate actions per EMG E-0, SAFETY REACTOR TRIP OR INJECTION, and the reactor does not trip.

In accordance with EMG E-0, Step 1 RNO, the CRS shall next direct entry into _ _1) ___.

The crew takes action to deenergize rod drive motor generators. Upon completion of this action, all rod bottom lights are lit and reactor power is less than 5% with a negative Intermediate Range SUR. Currently no red paths exist.

The Shift Manager should declare a(n) __ 2) ___.

(Reference Provided)

A. 1) OFN BG-009, EMERGENCY BORATION

2) Alert B. 1) OFN BG-009, EMERGENCY BORATION

2) Site Area Emergency C. 1) EMG FR-S1, RESPONSE TO NUCLEAR POWER GENERATION ATWS

2) Alert D. 1) EMG FR-S1, RESPONSE TO NUCLEAR POWER GENERATION/ATWS

2) Site Area Emergency Answer: D.

Explanation:

A is wrong because FR-S1 is entered and SAE is declared. Plausible if applicant believes that as long as the reactor is tripped, it is an Alert. Plausible if applicant believes E-0 directs initiation of Emergency Boration prior to FR-S1 entry. Emergency Boration is directed by FR-S1, step 6.

B is wrong because FR-S1 is entered. Plausible if applicant believes E-0 directs initiation of Emergency Boration prior to FR-S1 entry. Emergency Boration is directed by FR-S1.

C is wrong because an SAE is declared. Plausible if applicant believes that as long as the reactor is tripped, it is an Alert.

D is correct because FR-S1 is required and SAE should be declared if actions in the control room fail to trip the reactor. SRO-only due to procedure direction and EP declaration.

Technical

References:

E-0, FR-S1, APF 06-002-01, pages 26-27 References to be provided to applicants during exam: APF 06-002-01 Learning Objective: LO1733215 T1: Given a set off parameters, from memory, gain knowledge of the Emergency Plan responsibilities assigned to licensed Control Room personnel.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis 3 10CFR Part 55 Content: 55.43(b)5

Examination Outline Cross-Reference Level SRO 000038 Steam Generator Tube Rupture Tier # 1 Group # 1 Ability to interpret control room indications to K/A # 2.2.44 verify the status and operation of a system, Rating 4.4 and understand how operator actions and directives affect plant and system conditions:

2.2.44 Steam Generator Tube Rupture Question 79 Given:

A steam generator tube rupture occurred on SG A

The crew manually tripped the reactor, manually actuated Safety Injection, and entered EMG E-0, REACTOR TRIP OR SAFETY INJECTION

The following indications existed immediately following the reactor trip:

o Instrument Air Pressure on KA PI-40 was 120 PSIG and stable o All four SG levels were off-scale low Subsequently:

The crew transitioned to EMG E-3, STEAM GENERATOR TUBE RUPTURE and completed the RCS cooldown to below the target temperature using steam dumps

The following indications exist:

o Instrument Air Pressure on KA PI-40 is 60 PSIG and lowering o S/G A level is 78% WR and rising o S/G B level is 38% NR and stable o S/G C level is 80% NR and rising o S/G D level is 36% NR and stable Which one of the following describes the NEXT action the CRS should perform?

A. Commence depressurizing the RCS using normal spray B. Commence depressurizing the RCS using one PORV C. Go to EMG E-3, STEAM GENERATOR TUBE RUPTURE, Step 1 D. Go to EMG C-33, SGTR WITHOUT PRESSURIZER PRESSURE CONTROL, Step 1 Answer: C Explanation:

A is wrong because the stem indicates that SG radiation is higher than normal on more than one SG following the RCS cooldown. The Foldout Page in EMG E-3 directs stopping any cooldown or depressurization and going to EMG E-3, Step 1 for multiple tube ruptures. This is plausible because the next major step in the procedure is to depressurize the RCS. Normal spray is a method of depressurizing the RCS.

B is wrong because the stem indicates that SG radiation is higher than normal on more than one SG following the RCS cooldown. The Foldout Page in EMG E-3 directs stopping any cooldown or depressurization and going to EMG E-3, Step 1 for multiple tube ruptures. This is plausible because the next major step in the procedure is to depressurize the RCS. Using a PORV is a method of depressurizing the RCS when normal spray is not available. The stem indicates there is a loss of instrument air, which would mean the normal spray valves are not available.

C is correct because the stem indicates that SG radiation is higher than normal on more than one SG following the RCS cooldown. The Foldout Page in EMG E-3 directs stopping any cooldown or depressurization and going to EMG E-3, Step 1 when there are multiple tube ruptures.

D is wrong because the stem indicates that SG radiation is higher than normal on more than one SG following the RCS cooldown. The Foldout Page in EMG E-3 directs stopping any cooldown or depressurization and going to EMG E-3, Step 1 for multiple tube ruptures. This is plausible because EMG E-3 directs going to EMG C-33 when normal spray, aux spray, and at least one PORV are not available to depressurize the RCS.

Technical

References:

EMG E-3, Steam Generator Tube Rupture, Rev 34, Foldout Page Item 3, Page 5/146 References to be provided to applicants during exam: None.

Learning Objective: LO 1732325 Revision 014, Objective R2, Recognize the major actions which are accomplished by EMG E-3.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis 3 10CFR Part 55 Content: 55.43(b)(5)

Examination Outline Cross-Reference Level SRO 055 Station Blackout Tier # 1 Group # 1 Ability to determine or interpret the following as K/A # EA2.02 they apply to a Station Blackout: RCS core cooling Rating 4.6 through natural circulation cooling to S/G cooling Question 80 The plant is in a station blackout with natural circulation flow established.

The CRS has directed the crew to depressurize the steam generators using the ARVS in accordance with step 32 of EMG C-0, LOSS OF ALL AC POWER. All S/G levels are stable at 30% NR.

During the depressurization, pressurizer level goes off scale low with Reactor Vessel Level Natural Circulation Range Indication at 65%. The CRS should:

A. Stop the depressurization and enter EMG CS-02 LOSS OF AC POWER RECOVERY WITH SI REQUIRED.

B. Continue the depressurization and enter EMG ES-06 NATURAL CIRCULATION COOLDOWN WITH STEAM VOID IN THE VESSEL (WITH RVLIS).

C. Stop the depressurization and enter EMG FR-I2 RESPONSE TO LOW PRESSURIZER LEVEL.

D. Continue the depressurization and stay in EMG C-0 LOSS OF ALL AC POWER.

Answer: D Explanation:

A is wrong because the note before step 32 states that PZR level may be lost and head voiding may occur but do not stop the depressurization.

B is wrong because although you do continue with depressurization there is no transition direction to EMG ES-06 in EMG C-0 at this point in the procedure.

C is wrong because you do not stop the depressurization (see A explanation above).

D is correct because the note before step 32 states that PZR level may be lost and head voiding may occur but do not stop the depressurization and there is no transition direction to EMG ES-06 in EMG C-0 at this point in the procedure.so you stay in EMG C-0.

Technical

References:

EMG C-0, Revision 38, page 60 (Notes box).

References to be provided to applicants during exam: None.

Learning Objective: LO1732329 Obj. R3, LO100912 Obj. R3

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis 3 10CFR Part 55 Content: 55.43(b)5

Examination Outline Cross-Reference Level SRO BW/E04; W/E05 Inadequate Heat Transfer - Tier # 1 Loss of Secondary Heat Sink Group # 1 K/A # 2.2.37 Ability to determine operability and/or Rating 4.6 availability of safety related equipment Question 81 Initially:

A loss of offsite power occurred

EDG A started and then tripped

EDG B is tagged out for maintenance

Total AFW flow is 280,000 lbm/hr

All S/G NR levels are off-scale low

Containment pressure is 20 psig and rising

The crew transitioned to EMG C-0, LOSS OF ALL AC POWER

The CRS sent an operator to attempt to locally start EDG A Now:

NB01 bus voltage has just been restored

Alarm 129C TD AFP OVSPD/SYS FAULT TRIP is lit

Alarm 129A MD AFP A TROUBLE is lit

SPDS indicates a RED PATH on the Heat Sink Status Tree

All S/G WR levels are 20% and lowering Given these conditions, which of the following describes the actions the CRS should perform next?

A. Go to EMG FR-H1, RESPONSE TO LOSS OF SECONDARY HEAT SINK, and establish RCS heat removal by RCS bleed and feed B. Go to EMG FR-H1, RESPONSE TO LOSS OF SECONDARY HEAT SINK, and start any available AFW pump C. Continue in EMG C-0, LOSS OF ALL AC POWER, and establish RCS heat removal by RCS bleed and feed D. Continue in EMG C-0, LOSS OF ALL AC POWER, and start any available AFW pump Answer: D

Explanation:

A is wrong because EMG C-0, Notes above Step 1, say, CSF status trees shall be monitored for information only. Function Restoration (FR) Procedures shall not be implemented during this procedure. Additionally, EMG C-0 Foldout Page Item #6 says, If AFW Flow is lost after Step 11 has been performed AND is NOT immediately recoverable, THEN perform EITHER of the following: Start non-safety related AFW Pump or perform Attachment G, LOW PRESSURE FEEDWATER FLOW. Step 11 is to check that AC power has been restored. The stem provides indication that is has (i.e., NB01 voltage is normal).

The crew will proceed to Step 11, and this Foldout Page item will be applicable. However, it is plausible because a red path exists, and conditions are met for feed and bleed (with adverse containment), and SI equipment is now available. A candidate could forget that functional recovery procedures are not implemented when in EMG C-0.

B is wrong because EMG C-0, Notes above Step 1, say, CSF status trees shall be monitored for information only. Function Restoration (FR) Procedures shall not be implemented during this procedure. Plausible because the candidate could forget that FRPs are not implemented during EMG C-0.

C is wrong because the crew is still in EMG C-0, which directs starting the non-safety AFW pump on the Foldout Page to address loss of heat sink. EMG C-0 does not direct initiating feed and bleed. Plausible because conditions are met for feed and bleed and SI equipment is now available, and the candidate could forget that feed and bleed is a last resort.

D is correct because EMG C-0, Notes above Step 1, say, CSF status trees shall be monitored for information only. Function Restoration (FR) Procedures shall not be implemented during this procedure. Additionally, EMG C-0 Foldout Page Item #6 says, If AFW Flow is lost after Step 11 has been performed AND is NOT immediately recoverable, THEN perform EITHER of the following: Start non-safety related AFW Pump or perform Attachment G, LOW PRESSURE FEEDWATER FLOW. Step 11 is to check that AC power has been restored. The stem provides indication that is has (i.e., NB01 voltage is normal).

The crew will proceed to Step 11, and this Foldout Page item will be applicable.

Technical

References:

EMG C-0, LOSS OF ALL AC POWER, Rev 38, Steps 1-14 and Foldout Page References to be provided to applicants during exam: None.

Learning Objective: Lesson Plan LO1406100, Objective 11 Describe system and integrated plant response to transient and equipment failures.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis 4 10CFR Part 55 Content: 55.43(b)(5)

Examination Outline Cross-Reference Level SRO 000005 Inoperable/Stuck Control Rod Tier # 1 Group # 2 Knowledge of the bases in Technical K/A # 2.2.25 Specifications for limiting conditions for Rating 4.2 operations and safety limits.

Question 82 Per the Technical Specification Basis for LCO 3.1.4, Reactivity Control Systems, a control rod is considered INOPERABLE if __ _1)__ __.

When performing a SDM margin calculation to account for an INOPERABLE control rod per LCO 3.1.4, the basis states that the SDM verification must include ___ 2)___ _.

A. 1) the rod lift coil fails

2) the worth of the inoperable rod, as well as a rod of maximum worth B. 1) the rod has a drop time of 3.5 seconds

2) the worth of the inoperable rod, as well as a rod of maximum worth C. 1) the rod lift coil fails

2) the worth of the inoperable rod, only D. 1) the rod has a drop time of 3.5 seconds

2) the worth of the inoperable rod, only Answer: B Explanation:

A is wrong because if a rod lift coil fails it is still trippable, and therefore operable. Plausible if applicant believes an unmovable, but trippable rod is inoperable.

B is correct because minimum rod drop time is 2.7 seconds and both the inoperable rod and a rod of max worth is required in the SDM calculation. SRO-only due to required knowledge of TS basis.

C is wrong because if a rod lift coil fails it is still trippable, and therefore operable. Also an additional rod of max worth is required. Plausible if applicant believes unmoveable, but trippable rod is inoperable and only the inoperable rod must be included in the calculation.

D is wrong because both the inoperable rod and a rod of max worth is required in the SDM calculation.

Plausible if applicant believes that only the inoperable rod (one stuck rod criterion) must be included.

Technical

References:

TS 3.1.4 and basis, B 3.1.4, pages 1-5 References to be provided to applicants during exam: None.

Learning Objective: TS Lectures T1: Given a set of conditions determine Technical Specification Operability.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental 4 Comprehensive/Analysis 10CFR Part 55 Content: 55.43(b)2

Examination Outline Cross-Reference Level SRO W/E15 Containment Flooding Tier # 1 Group # 2 Ability to recognize abnormal indications for K/A # 2.4.4 system operating parameters that are entry- Rating 4.7 level conditions for emergency and abnormal operating procedures..

Question 83 Given:

A large break LOCA has occurred.

Containment Spray has actuated

A Train Containment Spray Pump tripped on overcurrent.

The crew is performing the final step of EMG ES-12, TRANSFER TO COLD LEG RECIRCULATION, when the STA reports the following conditions:

o Containment Pressure is 26 psig and lowering.

o Containment Radiation is 20 Rad/hr.

o Recirculation sump level is 2004'.

To which procedure will the crew transition and why?

A. EMG FR-Z1, RESPONSE TO HIGH CONTAINMENT PRESSURE to ensure Containment structural integrity is maintained.

B. EMG FR-Z2, RESPONSE TO CONTAINMENT FLOODING to address flooding of vital equipment in containment.

C. EMG FR-Z3, RESPONSE TO HIGH CONTAINMENT RADIATION LEVEL to address erroneous instrument readings due to high Containment radiation.

D. EMG C-13, CONTROL ROOM RESPONSE TO SUMP BLOCKAGE to protect the ECCS pumps for long term decay heat removal.

Answer: B Explanation. Per EMG F-0, FIGURE 6, CSF F-05 CONTAINMENT, conditions are met to transition to ORANGE PATH EMG FR-Z2 to address containment flooding. This procedure has the highest priority per EMG F-0 Foldout page criteria for ORANGE PATH and YELLOW PATH procedure priority setting.

A. is wrong, but plausible. Entry conditions would be met to go to EMG FR-Z1 on a YELLOW PATH if CTMT Pressure were >27 psig with 1 CSS pump running. If given pressure was higher, this procedure could be entered at CRS discretion if no other RED or ORANGE Path conditions were met.

Wrong since entry to the procedure is NOT required.

B. CORRECT, see explanation above.

C. is wrong, but plausible. Entry conditions would be met to transition to EMG FR-Z3 for a YELLOW PATH with given CTMT Radiation >4 R/HR. The procedure could be entered at CRS discretion if no other RED or ORANGE Path conditions were met. Wrong since an ORANGE Path conditions exists for CTMT Flooding.

D. is wrong, but plausible. Sump blockage could be a reason for why CTMT Level is high and there is a procedure transition from EMG ES-12, Step 22 to EMG C-13 IF both indications of pump cavitation are present (None given) AND Containment Recirc Sump Levels >2001 feet. Wrong since there are no given indications of cavitation, and per EMG FR-Z3, for level to be >2004 feet, there has to be an additional source of water besides what was in the RCS and RWST.

Technical

References:

EMG ES-12, Rev 22A, Page 31 EMG C-13, Rev 8, Page 1 EMG F-0, Rev 17, Pages 3 and 20 EMG FR-Z2, Rev 12, Page 1 BD-EMG FR-Z2, Rev 9, Page 10 LO1732351, Rev 014, Page 7 References to be provided to applicants during exam: None.

Learning Objective:

Lesson Plan 1732351, Objective 1. Identify Procedure entry conditions.

Question Source: Bank #

(3) Stem conditions, Distractor D Modified Bank #72038 X New Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental 3 Comprehensive/Analysis 10CFR Part 55 Content: 55.43(b)(2)

Examination Outline Cross-Reference Level SRO 069 or (W/E14) Loss of CTMT Integrity Tier # 1 Group # 2 Ability to determine and interpret the following K/A # AA2.01 as they apply to the Loss of Containment Rating 4.3 Integrity: Loss of containment integrity Question 84 THIS QUESTION DELETED. SEE EXAMINATION REPORT 05000482/2017301 FOR DETAILS Per Technical Specification Bases for LCO 3.6.3 for Containment Isolation Valves, the single failure criterion imposed during the plant safety analysis for a loss of containment integrity is:

A. RCP seal injection valves B. 36-inch shutdown purge valves C. 18-inch purge isolation valves D. Category 1 containment Isolation valves Answer: C Explanation:

A is wrong because...these are left open per a note in the LCOs.

B is wrong because these valves must be closed with a blind flange before entry to mode 4.

They are so large that they are assumed to be unable to be shut against DBA pressure.

C is correct because this is the single failure criterion of this TS and these valves per the TS bases. They are opened intermittently at power for various reasons so the inboard and outboard isolation valves have separate power supplies, etc.

D is wrong because this is a group of valves listed in the table that have the shortest LCO completion time due to risk and other factors but these are not the single failure criterion for TS 3.6.3.

Technical

References:

TS Bases rev 75, page 3.6.3-3.

References to be provided to applicants during exam: None.

Learning Objective: LO1732700 Obj. 7 Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No

Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis 3 10CFR Part 55 Content: 55.43(b)2

Examination Outline Cross-Reference Level SRO W/E01 & E02 Rediagnosis & SI Termination Tier # 1 Group # 2 EA2.1: Ability to determine and interpret the K/A # EA2.1 following as they apply to the (Reactor Trip or Rating 4.0 Safety Injection Rediagnosis): Facility conditions and selection of appropriate procedures during abnormal and emergency operations. (CFR: 43.5 /

45.13)

Question 85 Given:

The unit was at 100% power.

The crew observed RCS pressure was 2200 PSIG and lowering.

The reactor tripped and SI actuated.

The crew entered EMG E-0, REACTOR TRIP OR SAFETY INJECTION.

Subsequently, the crew stopped RCPs A and D.

PZR pressure stabilized at 1733 PSIG.

The crew observed RCS subcooling was 59°F.

Which of the following identifies 1) the fully stuck open PZR spray valve that caused this event and 2) the procedure the crew should transition to from EMG E-0, REACTOR TRIP OR SAFETY INJECTION?

A. 1) PCV-445B

2) EMG ES-03, SI TERMINATION B. 1) PCV-445C

2) EMG ES-03, SI TERMINATION C. 1) PCV-445B

2) EMG E-1, LOSS OF REACTOR OR SECONDARY COOLANT D. 1) PCV-445C

2) EMG E-1, LOSS OF REACTOR OR SECONDARY COOLANT Answer: A Explanation: Following a reactor trip and SI, the crew would enter EMG E-0. Step 12 directs the crew to check that the spray valves are closed. If they are not closed, then the RNO directs the crew to secure at least two RCPs depending on which spray valve is open. If PCV-445B is open, then the procedure directs the crew to stop RCPs A and D. The stem provides information that the crew has taken these steps, and the candidate must realize that these are the actions taken for a failed open PZR spray valve PCV-455B to answer the question correctly. Continuing in EMG E-0 after securing two RCPs, the crew would transition to EMG ES-03, SI TERMINATION, to reduce ECCS flow at Step 19. The values

for pressure and subcooling after stopping RCPs A and D were determined using the desktop simulator.

A is correct. See explanation.

B (1) is wrong because the crew would stop RCPs A and D for a failed open PZR spray valve PCV-445B. Plausible because PCV-455C is associated with Loop 2 and RCPs B and C, and a candidate may think that PCV-455C is associated with Loop 1 and the RCPs A and D in Loop 1. (2) is correct.

C (1) is correct. (2) is wrong because E-1 is not entered for a failed open spray valve once the RCPs have been secured and pressure stabilized. Plausible because if a PZR PORV is stuck, EMG E-0 directs going to E-1. A candidate who selects this answer may realize this is the appropriate action for a failed open PORV and think this may also be required for a failed open spray valve.

D (1) is wrong because the crew would stop RCPs A and D for a failed open PZR spray valve PCV-445B. Plausible because PCV-455C is associated with Loop 2 and RCPs B and C, and a candidate may think that PCV-455C is associated with Loop 1 and the RCPs A and D in Loop 1. (2) is wrong because E-1 is not entered for a failed open spray valve once the RCPs have been secured and pressure stabilized. Plausible because if a PZR PORV is stuck, EMG E-0 directs going to E-1. A candidate who selects this answer may realize this is the appropriate action for a failed open PORV and think this may also be required for a failed open spray valve.

Technical

References:

EMG E-0, Rev 39, Step 12 and 19, Pages 25 and 33 of 99 This question is similar to DCPP 2008 SRO Question 10 (ADAMS Accession No. ML091270008) and Wolf Creek SRO Tier 1, Gr 2 Exam Bank ID 47021.

References to be provided to applicants during exam: None.

Learning Objective: Document learning objective if possible.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis 3 10CFR Part 55 Content: 55.43(b)(5)

Examination Outline Cross-Reference Level SRO 004 Chemical and Volume Control Tier # 2 Group # 1 Ability to perform specific system and K/A # 2.1.23 integrated plant procedures during all modes Rating 4.4 of plant operation Question 86 The crew is attempting to emergency borate due to failure of two control rods to fully insert following a reactor trip. The crew was unable to establish emergency boration flow through BG HIS-8104, Emergency Borate to Charging Pump Suction Valve. After aligning the RWST to the Charging Pump suction, maximum available RWST flow through the charging system has been established at 85 gpm.

The CRS should direct performance of ...

A. OFN BG-009, EMERGENCY BORATION, Attachment A, ESTABLISHING ALTERNATE BORATION FLOWPATH, to establish 120 gpm letdown flow B. OFN BG-009, EMERGENCY BORATION, Attachment B, EMERGENCY BORATION USING SI AS A FLOWPATH, to utilize an SI pump for hot leg injection C. OFN BG-009, EMERGENCY BORATION, Attachment A, ESTABLISHING ALTERNATE BORATION FLOWPATH, to establish manual boration D. OFN BG-009, EMERGENCY BORATION, Attachment B, EMERGENCY BORATION USING SI AS A FLOWPATH, to utilize an SI pump for cold leg injection Answer: C Explanation:

A is wrong because manual boration must be established if RWST system flow is less than 90 gpm. Plausible as 120 gpm of letdown is directed earlier in procedure if using RWST.

B is wrong because transition to Attachment A is correct. Plausible if applicant believes SI is first option.

C is correct because manual boration should be established per Attachment A. SRO-only due to procedure direction of an AOP subsection/attachment.

D is wrong because transition to Attachment A is correct. Plausible if applicant believes SI is first option.

Technical

References:

OFN BG-009, page 6; EMG ES-02, REACTOR TRIP RESPONSE

References to be provided to applicants during exam: None.

Learning Objective: LO4710524 R1: DETERMINE major procedure flowpaths for conditions requiring emergency boration IAW OFN BG-009 Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis 3 10CFR Part 55 Content:

55.43(b)5

Examination Outline Cross-Reference Level SRO 012 Reactor Protection Tier # 2 Group # 1 Ability to (a) predict the impacts of the K/A # A2.01 following malfunctions or operations on the Rating 3.6 RPS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Faulty bistable operation Question 87 Given:

At 0830

The Unit was at 100% power.

The bistable for Channel B PZR Pressure Low failed.

The crew bypassed Channel B PZR Pressure Low to allow I&C to perform a Channel Operational Test on Channel A PZR Pressure Low.

At 0930

I&C completed the Channel Operational Test on Channel A PZR Pressure Low.

I&C reports the trip setpoint was 1920 psig and was unable to be adjusted.

To comply with Technical Specifications, the crew must A. Initiate action within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to be in MODE 3 within a MAXIMUM of 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months B. Initiate action within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to be in MODE 3 within a MAXIMUM of 13 hours1.50463e-4 days 0.00361 hours 2.149471e-5 weeks 4.9465e-6 months C. Place Channel B PZR Pressure Low in TRIP within a MAXIMUM of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months D. Place Channel B PZR Pressure Low in TRIP within a MAXIMUM of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Answer: A Explanation: At 0830, the crew must enter LCO 3.3.1, Condition M, and LCO 3.3.2, Condition D, because Channel B of PZR Pressure Low is inoperable due to the failed bistable. To comply with LCO 3.3.1, the crew must either place Channel B to trip within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or reduce power below P-7 within 78 hours9.027778e-4 days 0.0217 hours 1.289683e-4 weeks 2.9679e-5 months . To comply with LCO 3.3.2, the crew must either place Channel B to trip within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in Mode 3 in 78 hours9.027778e-4 days 0.0217 hours 1.289683e-4 weeks 2.9679e-5 months and in Mode 4 in 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months . LCO 3.3.1, Condition M, and LCO 3.3.2, Condition D required actions say that the inoperable channel may be bypassed to allow for surveillance (SR) testing for up to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

The stem says that the crew has bypassed inoperable Channel B PZR Pressure Low for SR testing (i.e., the channel operational test on Channel A).

At 0930, per TS Bases 3.3.1, Channel A of PZR Pressure Low is also now inoperable because the trip setpoint (i.e., 1920#) exceeds the allowable value (i.e., 1930#) in Table 3.3.1-1 for PZR Pressure Low for the Reactor Trip System (note that the ESFAS setpoint in Table 3.3.2-1 for PZR Pressure Low has NOT been exceeded - it is 1820psig). Therefore, there are two channels of Reactor Trip System instrumentation inoperable. Because there

are no conditions in LCO 3.3.1 for two channels of Reactor Trip System instrumentation inoperable, the crew must enter LCO 3.0.3. LCO 3.0.3 states, action shall be initiated within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to place the unit, as applicable, in Mode 3 within 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months , Mode 4 within 13 hours1.50463e-4 days 0.00361 hours 2.149471e-5 weeks 4.9465e-6 months , and Mode 5 within 37 hours4.282407e-4 days 0.0103 hours 6.117725e-5 weeks 1.40785e-5 months .

A is correct. See explanation.

B is wrong because the unit must be in Mode 3 within 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months , not 13 hours1.50463e-4 days 0.00361 hours 2.149471e-5 weeks 4.9465e-6 months . Plausible because LCO 3.0.3 directs the unit to be in Mode 4 within 13 hours1.50463e-4 days 0.00361 hours 2.149471e-5 weeks 4.9465e-6 months , and a candidate may mistakenly associate 13 hours1.50463e-4 days 0.00361 hours 2.149471e-5 weeks 4.9465e-6 months with Mode 3.

C is wrong because the crew must take the actions for LCO 3.0.3 because two channels of RTS instrumentation are inoperable. Also, LCO 3.3.1 and LCO 3.3.2 say that one inoperable channel may be bypassed for up to 12 channels, which does not mean that the channel must be placed in trip within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . Plausible because if the candidate does not recognize that Channel A is also inoperable and also thinks that the channel must be placed in trip within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when it has been bypassed.

D is wrong because the crew must take the actions for LCO 3.0.3 because two channels of RTS instrumentation are inoperable. Plausible because if only one channel is inoperable, then to comply with LCO 3.3.1, Condition M, the crew must either place Channel B to trip within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or reduce power below P-7 within 78 hours9.027778e-4 days 0.0217 hours 1.289683e-4 weeks 2.9679e-5 months , and to comply with LCO 3.3.2, Condition D, the crew must either place Channel B to trip within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in Mode 3 in 78 hours9.027778e-4 days 0.0217 hours 1.289683e-4 weeks 2.9679e-5 months and in Mode 4 in 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months . A candidate may not recognize that Channel A is also inoperable.

Technical

References:

Technical Specifications, Table 3.3.1-1 (Page 2/6), Amendment #140, Page 3.3-16 Technical Specifications, Amendment #156, Page 3.3-6 (LCO 3.3.1, Condition M)

Technical Specifications, Amendment #183, Page 3.3-31 (LCO 3.3.2, Condition D)

Technical Specifications Bases, Rev 1, Page 3.3.1-30 Technical Specifications Bases, Rev 0, Page 3.3.1-4 Lesson Plan LO1301200, Rev 12, Page 10/49 References to be provided to applicants during exam: None.

Learning Objective: Document learning objective if possible.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis 3 10CFR Part 55 Content: 55.43(b)(2)

Examination Outline Cross-Reference Level SRO 026 Containment Spray Tier # 2 Group # 1 Ability to apply Technical Specifications for a K/A # 2.2.40 system. Rating 4.7 Question 88 Given the following timeline with the unit operating at 100% power:

At 0900 on November 2, the A Containment Spray Pump coupling was replaced.

At 0900 on November 12, the Auxiliary Building Watch identified that the A Containment Spray pump coupling is cracked in three places Which of the following is the REQUIRED action?

(Reference Provided)

A. Restore the A pump to OPERABLE status by 0900 on November 15.

B. Be in MODE 3 no later than 1600 on November 12.

C. Be in MODE 3 no later than 1500 on November 12.

D. Restore the A pump to OPERABLE status by 0900 on November 22 Answer: A Explanation:

A is correct because once it is discovered and with the other train of CS OPERABLE, the 72 hrs applies, so 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months from 0900 on nov 12 is 0900 on nov 15.

B is wrong because this would apply if you went straight to 3.03, but you dont.

C is wrong because this would be correct if you misinterpret the 10 days from discovery portion of the LCO for Action A1.

D is wrong because if you just add 10 days to the discovery of the event then this would be correct but the other train is still OPERABLE, so the 10 days is not in force, only the 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

Technical

References:

TS Amendment 167, page 3.6-16 and TS section 1.3 Amendment 127.

References to be provided to applicants during exam: TS 3.6.6 Learning Objective: LO17327024 Obj. R1

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis 3 10CFR Part 55 Content: 55.43(b)2

Examination Outline Cross-Reference Level SRO 061 Auxiliary/Emergency Feedwater Tier # 2 Group # 1 Ability to (a) predict the impacts of the following K/A # A2.05 malfunctions or operations on the AFW; and (b) Rating 3.4 based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Automatic control malfunction Question 89 Given:

The Unit is operating at 100% power on November 1st, 2017.

At 0600, the TDAFP is declared INOPERABLE due to water in the oil.

At 0700, the TDAFP is tagged out for repairs.

At 0800, AL FT-009, Auxiliary Feedwater to S/G B Flow Transmitter, failed HIGH.

Which of the following describes 1) AFW flow to S/G B if a loss of offsite power were to occur at 0801 and 2) the action and EARLIEST completion time the crew must take to comply with Technical Specifications?

(Reference Provided)

A. 1) AFW flow to S/G B would NOT occur without Operator action.

2) Be in MODE 4 by November 1st at 2000.

B. 1) AFW flow to S/G B would NOT occur without Operator action.

2) Be in MODE 5 by November 2nd at 2100.

C. 1) AFW flow to S/G B would occur automatically without operator action

2) Restore the TDAFP to OPERABLE status by November 8th at 0600.

D. 1) AFW flow to S/G B would occur automatically without operator action

2) Restore the TDAFP to OPERABLE status by November 4th at 0600.

Answer: A Explanation: AL FT-9 provides automatic control for AL HV-009 based on flow rate. If this flow transmitter failed HIGH, then AL HV-009 would CLOSE stopping all flow to B S/G. Since two trains of AFW are INOPERABLE, then LCO 3.7.5, CONDITION C applies and the required action is to be in MODE 4 in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months since the unit is already in MODE 3 for the given loss of off site power at 0801.

A is correct. See explanation.

B 1) is correct See explanation.

B 2) is wrong because the crew has 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to reach MODE 4 at which time you are no longer in the MODE of applicability. Plausible since 37 hours4.282407e-4 days 0.0103 hours 6.117725e-5 weeks 1.40785e-5 months corresponds to the time required to be in MODE 5 for an LCO 3.03 event, which is NOT the case.

C 1) is wrong. Plausible if student doesnt recognize the failure of the controlling flow transmitter closes the applicable flow control valve (AL HV-009) It senses high flow as being delivered from the TDAFW Pump and closes the MD AFWP Flow control Smart valve.

C 2) is wrong because Condition A is not applicable for the given reason why the TDAFW Pump was declared INOPERABLE. Plausible since this would be true for an INOPERABLE steam supply valve to the TDAFW Pump.

D 1) is wrong. Plausible if student doesnt recognize the failure of the controlling flow transmitter closes the applicable flow control valve (AL HV-009) It senses high flow as being delivered from the TDAFW Pump and closes the MD AFWP Flow control Smart valve.

D 2) is wrong because Condition C is applicable and the plant is required to be placed in MODE 4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . Plausible since Condition B is applicable for the given reason why the TDAFW Pump was declared INOPERABLE and would be true if not for the flow transmitter failure would yields the train second train INOPERABLE.

Technical

References:

Lesson Plan LO1406100, Rev 15, Pages 13/64 and 23-25/64 LCO 3.7.5 Condition C, Amendment 184, Page 3.7-15 TS Bases, Page B 3.7.5-6, Rev 44 References to be provided to applicants during exam: LCO 3.7.5, pages 3.7-15 and 3.7-16.

Learning Objective: Document learning objective if possible.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis 3 10CFR Part 55 Content: 55.43(b)(5)

Examination Outline Cross-Reference Level SRO 062 AC Electrical Distribution Tier # 2 Group # 1 Ability to (a) predict the impacts of the K/A # A2.12 following malfunctions or operations on the ac Rating 3.6 distribution system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Restoration of power to a system with a fault on it Question 90 Given:

Unit is at full power

Main Control Board Annunciators 016A, XPB03/04 XFMR LOCKOUT and 016B, PB03/04 BUS UV actuate.

XPB03 transformer has a valid lockout due to transformer overcurrent.

The blue lockout light above the 4.16 KV bus feeder breaker for PB03 is lit and the bus is NOT energized.

What procedure provides the necessary instructions to physically re-energize the affected bus?

A. SYS PB-131, ENERGIZING PB BUSES B. ALR-00-0016A, XPB03/04 XFMR LOCKOUT C. SYS PB-200, ENERGIZING PB BUSES THROUGH TIE BREAKER D. OFN-NB-035, LOSS OF OFF-SITE POWER RESTORATION Answer: C Explanation:

A is wrong because the bus tie breaker should have auto-closed but if it doesnt then the ALR 00-016A directs using SYS-200 to re-energize the bus to prevent overloading it.

B is wrong the ALR directs the use of SYS-200.

C is correct - see A above D is wrong because this is not the correct procedure to use to mitigate this event. Attachment A in this procedure has guidance on restoration of non-vital buses and it includes this but this is not the set of conditions in the stem (no LOSP event in progress)

Note that the (a) part of the KA is implied from the information given in the stem to get to the correct procedure selection.

Technical

References:

ALR-00-016A, page 3 and 4, revision 14.

References to be provided to applicants during exam: None.

Learning Objective: LO4706205 Obj. 3, LO100912 Obj. R3 Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental 3 Comprehensive/Analysis 10CFR Part 55 Content: 55.43(b)5

Examination Outline Cross-Reference Level SRO 002 Reactor Coolant Tier # 2 Group # 2 Knowledge of the specific bases for EOPs. K/A # 2.4.18 Rating 4.0 Question 91 One hour ago:

A reactor trip and SI actuation occurred.

RCS pressure was 1800 psig and lowering.

The crew entered EMG E-0, REACTOR TRIP OR SAFETY INJECTION.

Now:

RCS pressure is 350 psig and stable.

The crew is performing Step 24 in EMG E-1, LOSS OF REACTOR OR SECONDARY COOLANT, Check If RCS Cooldown and Depressurization Is Required.

Per the background document for this step, because RCS pressure is ___1)___, the CRS should ___2)___.

A. 1) Below the shutoff head of the SI pumps

2) Remain in EMG E-1, LOSS OF REACTOR OR SECONDARY COOLANT B. 1) Above the shutoff head of the RHR pumps

2) Remain in EMG E-1, LOSS OF REACTOR OR SECONDARY COOLANT C. 1) Below the shutoff head of the SI pumps

2) Transition to EMG ES-11, POST-LOCA COOLDOWN AND DEPRESSURIZATION D. 1) Above the shutoff head of the RHR pumps

2) Transition to EMG ES-11, POST-LOCA COOLDOWN AND DEPRESSURIZATION Answer: D Explanation: The question requires the applicant to assess the conditions and apply knowledge of the bases for EMG E-1 to select the correct answer. The question stem says that RCS pressure is 350 psig, which is above RHR pump shutoff head (193 psig). At Step 24 in EMG E-1, the crew must make a decision based on RCS pressure. Lesson Plan LO1732320 describes Step 24 as a transition step and says the crew should only remain in this procedure for a Large Break LOCA in which RCS pressure is < RHR pump shutoff head and flow from the RHR pumps has been verified. For any Small Break LOCA in the RCS for which the RCS pressure stabilizes above the shutoff head of the RHR pumps, the crew should transition to EMG ES-11.

A (1) is wrong because the determination of whether to stay in EMG E-1 or transition to EMG ES-11 is based on whether or not the RHR pumps are injecting. RHR pumps will inject when SI has actuated and RCS pressure is below the shutoff head of the RHR pumps. Plausible because the decision of whether to transition to EMG ES-11 from EMG ES-03 is based on whether or not RCS pressures is below shutoff pressure of the SI pumps (i.e., 1578 psig), and the question stem says RCS pressure is below the shutoff head of the SI pumps. A candidate may select this distractor if he or she does not know the basis for EMG E-1.

(2) is wrong because when RCS pressure is above the shutoff head of the RHR pumps, EMG E-1 directs the crew to transition to EMG ES-11. Plausible if the candidate thinks that when RCS pressure is below the shutoff head of the SI pumps, the candidate should continue in E-1.

B (1) is correct (its the correct basis for making the decision at Step 24 in EMG E-1). (2) is wrong because when pressure is stable above RHR shutoff head, EMG E-1 directs a transition to EMG ES-

11. Plausible if the student thinks the crew should remain in EMG E-1 to mitigate the LOCA.

C (1) is wrong because the determination of whether to stay in EMG E-1 or transition to EMG ES-11 is based on whether or not the RHR pumps are injecting. RHR pumps will inject when SI has actuated and RCS pressure is below the shutoff head of the RHR pumps. Plausible because the decision of whether to transition to EMG ES-11 from EMG ES-03 is based on whether or not RCS pressures is below shutoff pressure of the SI pumps (i.e., 1578 psig), and the question stem says RCS pressure is below the shutoff head of the SI pumps. A candidate may select this distractor if he or she does not know the basis for EMG E-1. (2) is correct.

D (1) and (2) are correct. See explanation.

Technical

References:

Lesson Plan LO1732320, Rev 14, Page 18/26 BD-EMG E-1, Rev 17, Page 51/76 References to be provided to applicants during exam: None.

Learning Objective: Lesson Plan LO1732320, Objective 4, Explain the bases and any knowledge requirements for selected procedure steps.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis 3 10CFR Part 55 Content: 55.43(b)(5)

Examination Outline Cross-Reference Level SRO 014 Rod Position Indication Tier # 2 Group # 2 Ability to (a) predict the impacts of the following K/A # A2.06 malfunctions or operations on the RPIS; and (b) Rating 3.0 based on those on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Loss of LVDT Question 92 Reactor Power is at 80%

Main Control Board Annunciator 080B, RPI NON URG ALARM, is received

DRPI LEDs indicate Data B failure 1, 2, and 3 AND GENERAL WARNIN for ALL control rods

1) Per Technical Specification Bases for LCO 3.1.7, what is the status of DRPI?

2) What is the current accuracy of the system?

A. 1) OPERABLE

2) +4 / -10 B. 1) OPERABLE

2) -4 / +10 C. 1) INOPERABLE

2) +4 / -10 D. 1) INOPERABLE

2) -4 / +10 Answer: A Explanation:

A is correct per TSB for 3.1.7 and ALR-00-080B, a loss of data B for DRPI (which is the equivalent of LVDT in the KA catalog and therefore meets this KA) causes a half accuracy but is still operable (see also TSB 3.1.4) and the accuracy per the ALR-00-080B is +4 / -10 for Data B.

B is wrong because the signs are wrong on the accuracy.

C is wrong because it is still operable, second part is correct.

D is incorrect because both first and second parts are incorrect (see A above).

Technical

References:

TSB 3.1.7 (page B 3.1.7-2), rev 0, and ALR-00-080B, rev 7A.

References to be provided to applicants during exam: None.

Learning Objective: LO1732700 Obj. 7 Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis 3 10CFR Part 55 Content: 55.43(b)2

Examination Outline Cross-Reference Level SRO 035 Steam Generator Tier # 2 Group # 2 Ability to (a) predict the impacts of the following K/A # A2.01 malfunctions or operations on the GS; and (b) Rating 4.6 based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Faulted or ruptured S/Gs Question 93 Given:

A reactor trip and safety injection have occurred

RCS pressure is 1600 psig and lowering

PZR level is off-scale low

Tavg is 500ºF and lowering

Containment pressure is 3 psig and rising

S/G A pressure is 620 psig and lowering

SG B, C, and D pressures are 900 psig and stable Which of the following actions and procedures will be performed immediately following transition from EMG E-0, REACTOR TRIP OR SAFETY INJECTION?

A. Align normal charging and stop ECCS pumps in accordance with EMG ES-03, SI TERMINATION B. Initiate a controlled RCS cooldown and depressurization in accordance with EMG ES-11, POST LOCA COOLDOWN AND DEPRESSURIZATION C. Isolate SG A in accordance with EMG E-2, FAULTED STEAM GENERATOR ISOLATION D. Restore pressurizer level in accordance with EMG FR-I2, RESPONSE TO LOW PRESSURIZER LEVEL Answer: C Explanation: Because SG A pressure is lowering in an uncontrolled manner, the crew will transition to EMG E-2 at EMG E-0, Step 16 RNO.

A is wrong RCS pressure is lowering, and therefore, SI termination criteria is not met.

Plausible because EMG E-0, Step 19, directs the crew to transition to EMG ES-03, SI TERMINATION, if SI termination criteria are met, and a candidate may think the criteria are met.

B is wrong because EMG ES-11 is entered from EMG E-1 or EMG ES-03, not EMG E-0.

Plausible because ultimately, the crew will complete EMG E-2, transition to EMG E-1, LOSS OF REACTOR OR SECONDARY COOLANT, and then transition to EMG ES-11 to cooldown and depressurize the RCS.

C is correct. See explanation.

D is wrong because EMG FR-I2 directs the operator to return to the procedure and step in effect if SI has not been terminated, and SI has not yet been terminated (nor is the criteria met for it to be terminated). Plausible because for these conditions, a yellow path on the Integrity Status Tree would exist, and the crew would have initiated EMG F-0 in EMG E-0, Step 15, which would direct monitoring the safety functions while continuing with the procedure in effect.

Technical

References:

EMG E-0, Step 16 RNO (directs transition to EMG E-2), Rev 39, Page 27/99 References to be provided to applicants during exam: None.

Learning Objective:

Question Source: Bank #60058 (SRO Tier 2, X Gr 2)

(note changes; attach parent) Modified Bank #

New Question History: Last NRC Exam No (last on 2006 WC NRC Exam)

Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis 3 10CFR Part 55 Content: 55.43(b)(5)

Examination Outline Cross-Reference Level SRO Tier # 3 Knowledge of the fuel-handling responsibilities of Group # 1 SROs. K/A # 2.1.35 Rating 3.9 Question 94 According to FHP 03001, REFUELING MACHINE OPERATING INSTRUCTIONS:

Prior to movement of fuel assemblies using the Refueling Machine, STS KE001, REFUELING MACHINE OPERABILITY TEST shall be performed within a MAXIMUM of

__1)__ hours.

During refueling operations, to bypass Refueling Machine interlocks with a fuel assembly suspended, permission must be obtained from __2)____.

A. 1) 72

2) the refueling SRO only B. 1) 72

2) the refueling SRO with concurrence from the Shift Manager or Control Room Supervisor C. 1) 100

2) the refueling SRO only D. 1) 100

2) the refueling SRO with concurrence from the Shift Manager or Control Room Supervisor Answer: D.

Explanation:

A is wrong because the surveillance must be conducted within 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months and SM or CRS concurrence is required. Plausible if the applicant thinks 3 days is the requirement. Plausible since refueling SRO permission only is required to bypass with no fuel.

B is wrong because the surveillance must be conducted within 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months . Plausible if the applicant thinks 3 days is the requirement.

C is wrong because SM or CRS concurrence is required. Plausible since refueling SRO permission only is required to bypass with no fuel.

D is correct because the surveillance must be conducted within 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months and SM or CRS concurrence is required. SRO-only due to refueling supervisor requirements being tested.

Technical

References:

FHP 03001, pages 2-4 References to be provided to applicants during exam: None.

Learning Objective: LO1403404 R2: DISCUSS the operation of the Manipulator Crane, including electrical power supply and associated interlocks.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental 3 Comprehensive/Analysis 10CFR Part 55 Content: 55.43(b)7

Examination Outline Cross-Reference Level SRO Conduct of Operations Tier # 3 Group # n/a K/A: Ability to use procedures to determine the K/A # 2.1.43 effects on reactivity of plant changes, such as Rating 4.3 reactor coolant system temperature, secondary plant, fuel depletion, etc.

Question 95 Given:

The Unit is operating at 95% power.

The crew just returned Heater Drain Pump A to service in accordance with SYS AF-121, HEATER DRAIN PUMP OPERATION.

Following the pump start, the crew observed a power reduction of a greater magnitude than expected.

The crew determined that an error was made when the Main Turbine Load Control mode was selected prior to starting Heater Drain Pump A.

The crew should have placed Main Turbine Load Control in ___1)___ mode, and the classification for this event in accordance with AP 19E-002, REACTIVITY MANAGEMENT, is

___2)___.

A. 1) Megawatt

2) Level 4 (unplanned reactivity change caused by personnel error)

B. 1) Open Loop

2) Level 4 (unplanned reactivity change caused by personnel error)

C. 1) Megawatt

2) Level 2 (a human performance action bypassed a reactivity control system resulting in improper reactivity control)

D. 1) Open Loop

2) Level 2 (a human performance action bypassed a reactivity control system resulting in improper reactivity control)

Answer: B Explanation:

A (1) is wrong because SYS AF-21 directs placing Main Turbine Load Control in Open Loop Mode to avoid unplanned changes in power. In MEGAWATT (MW) mode, the turbine control system will monitor main generator output and adjust turbine control valve position as necessary to maintain generator output constant. Any factor that would impact generator output (RCS or feedwater temperature, SG pressure, main steam line leaks/ruptures) will result in the control system opening or closing the turbine control valves to maintain a constant generator output. In OPEN LOOP, the main turbine control system will maintain a steady control valve position without any feedback mechanism.

Plausible if a candidate does not understand what happens in MW mode and Open Loop Mode when starting or stopping a heater drain pump. (2) is correct.

B (1) is correct. SYS AF-121, Precaution/Limitation 4.6, states, When securing a Heater Drain Pump above 60% RTP, Main Turbine Load Control should be in Open Loop mode to minimize undesired reactor power changes. Step 6.1.9 for Heater Drain Pump A startup also directs the crew to ensure Open Loop mode is selected. Wolf Creek also has OE for this event (refer to CR 00085896, Optimum Mode of Turbine Operation To Prevent Adverse Reactivity Management Events, which is listed as CR #17 in Section 3.1.3 of GEN 00-004, POWER OPERATION.) (2) Is correct per Attachment A, Step A.11, Level 4 Table, #4.6, Unplanned Reactivity Change Directly Caused by Equipment Problem or Personnel Error.

C (1) is wrong because SYS AF-21 directs placing Main Turbine Load Control in Open Loop Mode to avoid unplanned changes in power. In MEGAWATT (MW) mode, the turbine control system will monitor main generator output and adjust turbine control valve position as necessary to maintain generator output constant. Any factor that would impact generator output (RCS or feedwater temperature, SG pressure, main steam line leaks/ruptures) will result in the control system opening or closing the turbine control valves to maintain a constant generator output. In OPEN LOOP, the main turbine control system will maintain a steady control valve position without any feedback mechanism.

Plausible if a candidate does not understand what happens in MW mode and Open Loop Mode when starting or stopping a heater drain pump. (2) is wrong because a Level 2 event is defined as an event that places the plant outside of the Design, Analysis, or Licensing Basis or significant events that compromise fuel-related limits, or directly result in fuel failure. This is not a Level 2 event (there is a power reduction). Plausible because a candidate might consider ovation turbine control system a Reactivity Control System since changes to the secondary effect reactivity, and performance of the evolution was improper in that MEGAWATT Mode was selected for turbine control.

D (1) is correct (2) is wrong because a Level 2 event is defined as an event that places the plant outside of the Design, Analysis, or Licensing Basis or significant events that compromise fuel-related limits, or directly result in fuel failure. This is not a Level 2 event (there is a power reduction).

Plausible because a candidate might consider ovation turbine control system a Reactivity Control System since changes to the secondary effect reactivity, and performance of the evolution was improper in that MEGAWATT Mode was selected for turbine control.

Technical

References:

AP 19E-002, REACTIVITY MANAGEMENT PROGRAM, Rev 19, ATTACHMENT A, Page 20/26 References to be provided to applicants during exam: NONE Learning Objective:

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis 3 10CFR Part 55 Content: 55.43(b)(6)

Examination Outline Cross-Reference Level SRO Tier # 3 Knowledge of the process for controlling temporary Group # 2 design changes. K/A # 2.2.11 Rating 3.3 Question 96 According to AP 21I-001, TEMPORARY CONFIGURATION CHANGES:

The ___1)___ is the senior Operations Department representative responsible for approving installation of Temporary Configuration Changes (TCCs).

A 50.59 screening must be performed no later than __ 2)___ days after a TCC is entered into the TCC log.

A. 1) Shift Manager

2) 90 B. 1) Shift Manager

2) 60 C. 1) Control Room Supervisor

2) 90 D. 1) Control Room Supervisor

2) 60 Answer: A Explanation:

A is correct because SM approval is required and 90 days is the limit. SRO-only due to administrative responsibilities of the Shift Manager and oversight of 50.59 requirements.

B is wrong because 90 days is the limit. Plausible if applicant thinks 60 days.

C is wrong because SM approval is required. Plausible if applicant thinks CRS approves changes.

D is wrong because SM approval is required and 90 days is the limit. Plausible if applicant thinks CRS approval only and a 60 day limit.

Technical

References:

AP 21I-001, pages 7-8 References to be provided to applicants during exam: None.

Learning Objective: LO1734019 T1: Explain the SROs knowledge and responsibility in Changes to the Facility.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental 3 Comprehensive/Analysis 10CFR Part 55 Content: 55.43(b)3

Examination Outline Cross-Reference Level SRO

2. Equipment Control Tier # 3 Group # n/a Knowledge of the process for managing K/A # 2.2.18 maintenance activities during shutdown operations, Rating 3.9 such as risk assessments, work prioritization, etc.

Question 97 Which of the following is a responsibility of the Operations SRO according to AP 21D-002, EVALUATION FOR POTENTIAL ENERGY/FLUID TRANSFER PATHS?

A. Evaluate the impact on the schedule of any identified fluid/energy transfers B. Evaluate the impact on plant personnel of any identified fluid/energy transfers C. Perform review as verification that all potential fluid or energy transfer paths have been identified and positive controls are in place to control configuration D. Perform review as verification that all potential fluid or energy transfer paths have been identified and measures have been taken to protect plant personnel Answer: C Explanation: AP 21D-002, Section 5.0, Responsibilities, Subsection 5.4, Operations Senior Reactor Operator (SRO), lists the responsibilities of the Operations SRO. One of these responsibilities is to perform review as verification that all potential fluid or energy transfer paths have been identified and positive controls are in place to control configuration.

A is wrong because AP 21D-002, Section 5.0, Responsibilities, Subsection 5.4, Operations Senior Reactor Operator (SRO), does NOT list this as a responsibility of the Operations SRO. Plausible because AP 21D-002, Section 5.0, Responsibilities, Subsection 5.3, Evaluator, states that a responsibility of the Evaluator is to review the schedule and evaluate the impact to the plant of any identified fluid/energy transfers, and a candidate may think this is a responsibility of the SRO.

B is wrong because AP 21D-002, Section 5.0, Responsibilities, Subsection 5.4, Operations Senior Reactor Operator (SRO), does NOT list this as a responsibility of the Operations SRO. AP 21D-002, Section 5.0, does not list this activity as a responsibility for anyone. Further, AP 21D-002, Section 1.0, Purpose, states, A FPE is not for personnel protection. . Plausible because a candidate may think one purpose of an FPE is to protect plant personnel.

C is correct. See explanation.

D is wrong because AP 21D-002, Section 5.0, Responsibilities, Subsection 5.4, Operations Senior Reactor Operator (SRO), does NOT list this as a responsibility of the Operations SRO. AP 21D-002, Section 5.0, does not list this activity as a responsibility for anyone. Further, AP 21D-002, Section 1.0, Purpose, states, A FPE is not for personnel protection. Plausible because a candidate may think one purpose of an FPE is to protect plant personnel.

Technical

References:

AP 21D-002, Rev 12A, Page 4/9 Lesson Plan LO1734018, Rev 6, Page 26-27/46 (NOTE: this lesson plan is labeled as SRO Only.)

References to be provided to applicants during exam: None.

Learning Objective: Describe the purpose, scope and operator responsibilities of procedure AP 21D-002, EVALUATION FOR POTENTIAL ENERGY/FLUID TRANSFER PATHS (R12).

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental 2 Comprehensive/Analysis 10CFR Part 55 Content: 55.43(b)(5)

Examination Outline Cross-Reference Level SRO Tier 3 Tier # 3 Group #

Knowledge of radiological safety procedures K/A # 2.3.13 pertaining to licensed operator duties, such as Rating 3.8 response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc Question 98 A fire has been reported in the Aux Building. The Shift Manager needs to direct an Operator into the RCA to complete a time critical action per OFN KC-016, FIRE RESPONSE. What procedure contains the guidance for this specific type of RCA entry process?

A. AI 21-016, OPERATOR TIME CRITICAL ACTIONS VALIDATION B. AP 21-001, CONDUCT OF OPERATIONS C. AP 25A-100, CONTAINMENT ENTRY D. EPP 06-013, EXPOSURE CONTROL AND PERSONNEL PROTECTION Answer: B Explanation:

A is wrong see B.

B is correct because the RCA fast entry process during emergency conditions is in section 6.25.1 of the conduct of operations procedure, AP 21-001, page 66.

C is wrong see B.

D is wrong because it provides no guidance for RCA entry. Plausible in that the scope of the procedure applies to WCGS Emergency Response Organization for reentry and recovery activities required to restore from an emergency. This procedure is applicable with Shift Manager Responsibilities for recovery operations from a NOUE.

Technical

References:

AP 21-001, page 66, rev 78.

LO1733258, Rev 2 Page 8 of 12, References to be provided to applicants during exam: None.

Learning Objective: LO1733258, Objective 1 Discuss the purpose / scope and selected knowledge requirements of procedure AP 21-001, Conduct of Operations.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental 3 Comprehensive/Analysis 10CFR Part 55 Content: 55.43(b)5

Examination Outline Cross-Reference Level SRO

4. Emergency Procedures/Plan Tier # 3 Group # n/a Knowledge of EOP implementation hierarchy K/A # 2.4.16 and coordination with other support Rating 4.4 procedures or guidelines such as, operating procedures, abnormal operating procedures, and severe accident management guidelines.

Question 99 While performing EMG ES-03, SI TERMINATION the crew observes alarms are present on Fire Alarm Control Panel KC008. A few minutes later, an NSO calls the control room and reports there is a fire in MDAFW Pump Room B.

The CRS A. Is required to concurrently implement both EMG ES-03 and OFN KC-016, FIRE

RESPONSE

B. Is required to complete the actions in EMG ES-03 and then transition to OFN KC-016, FIRE RESPONSE C. Is allowed by procedure to implement OFN KC-016, FIRE RESPONSE as long as it does not interfere with performance of EMG ES-03 D. Is allowed by procedure to implement the actions in EMG ES-03 as long as they do not interfere with the actions of OFN KC-016, FIRE RESPONSE Answer: C Explanation: AP 15C-003, Section 6.2.3, says, While performing EMGs, plant conditions may indicate the need to correct problems not directly related to the event mitigation strategy. The operator may perform OFNs and ALRs which address these problems as long as the actions do not interfere with performance of the EMGs. Therefore, the SRO may concurrently perform the Fire OFN as long as it does not interfere with performance of the EMGs.

A is wrong because AP 15C-003 says the SRO may perform OFN and EMG actions concurrently; not that the SRO is required to do so. Plausible because if the fire is not extinguished within 15 minutes, the crew will need to declare an EAL. Therefore, extinguishing the fire as soon as possible will be a priority for the crew; however, it is not required by AP 15C-003 to concurrently perform the OFN with the EMG, even where there is a fire.

B is wrong because AP 15C-003 allows the SRO to perform the OFN concurrently as long as it does not interfere with performance of the EMG. There is no requirement to wait to perform the OFN after performing the EMG actions. Plausible because performance of actions in EMGs are a higher priority than performance of actions directed by an OFN.

C is correct. See explanation.

D is wrong because the EMGs are a higher priority than OFNs. Plausible since concurrent implementation is allowed, and a candidate may think that addressing fire response is a higher priority because a fire has the potential to injure plant personnel, to spread and cause additional damage, which could complicate recovery with the EMGs, and to require entry into the Emergency Plan.

Technical

References:

AP 15C-003, Rev 34, Page 19/60 Lesson Plan LO1733203, Rev 013, Page 16/33 References to be provided to applicants during exam: None.

Learning Objective: Lesson Plan LO1733203, Objective R14, Discuss the priority of procedures that deal with abnormal conditions IAW AP 15C-003.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis 2 10CFR Part 55 Content: 55.43(b)(5)

Examination Outline Cross-Reference Level SRO Tier 3 Tier # 3 Group #

Knowledge of events related to system K/A # 2.4.30 operation/status that must be reported to Rating 4.1 internal organizations or external agencies, such as the State, the NRC, or the transmission system operator.

Question 100 The Shift manager is required to notify the Transmission Service Operator (TSO) of a change in status or capability of the automatic voltage regulator as soon as practical, but not to exceed _____1)____ in accordance with procedure ____2)_____.

A. 1) Thirty minutes

2) OFN AF-25, UNIT LIMITATIONS B. 1) One hour

2) OFN AF-25, UNIT LIMITATIONS C. 1) Thirty minutes

2) AP 21C-001, WOLF CREEK SUBSTATION D. 1) One hour

2) AP 21C-001, WOLF CREEK SUBSTATION Answer: C Explanation:

A is wrong because part 2, the procedure, is incorrect. Part 1 is correct for the time.

B is wrong because both 1 and 2 are incorrect.

C is correct because a change in the AVR is required to be communicated to the TSO as soon as practical but within 30 minutes IAW procedure AP-21C-001, section 5.3.26, page 11, rev 20.

D is wrong because first part is incorrect. The procedure is correct.

Technical

References:

AP-21C-001, section 5.3.26, page 11, rev 20.

References to be provided to applicants during exam: None.

Learning Objective: LO100912 Obj. R3 Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental 3 Comprehensive/Analysis 10CFR Part 55 Content: 55.43(b)5

Containment Spray and Cooling Systems 3.6.6 3.6 CONTAINMENT SYSTEMS 3.6.6 Containment Spray and Cooling Systems LCO 3.6.6 Two containment spray trains and two containment cooling trains shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One containment spray A.1 Restore containment 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months train inoperable. spray train to OPERABLE status. AND 10 days from discovery of failure to meet the LCO B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition A not AND met.

B.2 Be in MODE 5. 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months C. One containment cooling C.1 Restore containment 7 days train inoperable. cooling train to OPERABLE status. AND 10 days from discovery of failure to meet the LCO (continued)

Wolf Creek - Unit 1 3.6-16 Amendment No. 123, 167

Containment Spray and Cooling Systems 3.6.6 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. Two containment cooling D.1 Restore one containment 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months trains inoperable. cooling train to OPERABLE status.

E. Required Action and E.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition C or D AND not met.

E.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months F. Two containment spray F.1 Enter LCO 3.0.3. Immediately trains inoperable.

OR Any combination of three or more trains inoperable.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.6.1 ---------------------------------NOTE-------------------------------

Not required to be met for system vent flow paths opened under administrative control.

Verify each containment spray manual, power 31 day operated, and automatic valve in the flow path that is not locked, sealed, or otherwise secured in position is in the correct position.

(continued)

Wolf Creek - Unit 1 3.6-17 Amendment No. 123, 167, 212

AFW System 3.7.5 3.7 PLANT SYSTEMS 3.7.5 Auxiliary Feedwater (AFW) System LCO 3.7.5 Three AFW trains shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS

NOTE-----------------------------------------------------------

LCO 3.0.4b. is not applicable when entering MODE 1.

CONDITION REQUIRED ACTION COMPLETION TIME A. One steam supply to A.1 Restore steam supply to 7 days turbine driven AFW pump OPERABLE status.

inoperable. AND 10 days from discovery of failure to meet the LCO B. One AFW train inoperable B.1 Restore AFW train to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months for reasons other than OPERABLE status.

Condition A. AND 10 days from discovery of failure to meet the LCO (continued)

Wolf Creek - Unit 1 3.7-15 Amendment No. 123, 155, 171, 177, 184

AFW System 3.7.5 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and C.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time for Condition A or B AND not met.

C.2 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months OR Two AFW trains inoperable.

D. Three AFW trains D.1 --------------NOTE-------------

inoperable. LCO 3.0.3 and all other LCO Required Actions requiring MODE changes are suspended until one AFW train is restored to OPERABLE status.

Initiate action to restore Immediately one AFW train to OPERABLE status.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.5.1 -------------------------------NOTE---------------------------------

Not required to be performed for the AFW flow control valves until the system is placed in standby or THERMAL POWER is > 10% RTP.

Verify each AFW manual, power operated, and 31 days automatic valve in each water flow path, and in both steam supply flow paths to the steam turbine driven pump, that is not locked, sealed, or otherwise secured in position, is in the correct position.

(continued)

Wolf Creek - Unit 1 3.7-16 Amendment No. 123, 171, 177, 184

APF 06-002-01, Rev. 17A [Commitment Step 3.2.2] Page 1 of 37 EMERGENCY ACTION LEVELS EAL-1, RADIOACTIVE EFFLUENT RELEASE START 1-RER1 1-RER2 1-RER3 1-RER4 Mode: ALL Mode: ALL Mode: ALL Mode: ALL Rapidly increasing readings

Unplanned release of airborne

Any of the following monitors exceed

Any of the following monitors exceed or liquid radioactivity at greater limit listed for greater than or equal to limit listed for greater than or equal to OR than the HI alarm limit for 15 minutes: 15 minutes:

greater than or equal to 15

GT RE 21B (Channel 213) greater

GT RE 21B (Channel 213) greater ALERT alarm on any of the minutes on any of the following than or equal to 1.9E+06 uCi/sec than or equal to 1.9E+07 uCi/sec following process rad monitors:

GH RE 10B (Channel 103) greater process rad monitors:

GH RE 10B (Channel 103) greater

GT RE 21A (Channel 211) than or equal to 1.9E +06 uCi/sec than or equal to 1.9E +07 uCi/sec

GT RE 21B (Channel 213) OR OR

GT RE 21A (Channel 212)

GH RE 10B (Channel 103)

Gamma radiation measurements

Gamma radiation measurements

GT RE 21B (Channel 213)

BM RE 52 (Channel 526) indicates dose rates at site indicates dose rates at site

GH RE 10A (Channel 101)

Yes

LE RE 59 (Channel 596) Yes boundary in excess of 20 mR/Hr Yes boundary in excess of 200 mR/Hr Yes

GH RE 10A (Channel 102)

HF RE 95 (Channel 956) OR OR

GH RE 10B (Channel 103)

OR

Valid dose assessment indicates a

Valid dose assessment indicates a

BM RE 52 (Channel 526)

As indicated by chemistry dose greater than 100 mR TEDE dose greater than 1000 mR TEDE

LE RE 59 (Channel 596) sample analysis. or 500 mR thyroid (CDE) at site or 5000 mR thyroid (CDE) at the

HF RE 95 (Channel 956) boundary site boundary NOTE: If rad monitor reading can not be validated within 15 NOTE: If rad monitor reading can not NOTE: If rad monitor reading can not minutes, declaration must be be validated within 15 minutes, be validated within 15 m inutes, based on actual monitor reading. declaration must be based on actual declaration must be based on actual monitor reading. monitor reading.

No No No No 1-RER5 Mode: ALL

Valid Control Room area rad monitor reading greater than 15 mR/Hr (SD RE 33)

OR GENERAL

Unplanned or unexplained area rad monitor increases by EMERGENCY Yes a factor of 1000 over normal levels in areas that require access to locally operate equipment to maintain safe operation or perform a safe shutdown. Average values are contained in the Bases for this EAL.

SITE AREA EMERGENCY No 1-RER6 Mode: ALL

Unplanned release of airborne or liquid radioactivity at greater than the ALERT alarm limit for greater than or equal to 60 ALERT minutes on any of the following process rad monitors:

GT RE 21B (Channel 213)

GH RE 10B (Channel 103)

BM RE 52 (Channel 526)

NOTIFICATION OF Yes

LE RE 59 (Channel 596) UNUSUAL EVENT

HF RE 95 (Channel 956)

OR

As indicated by chemistry sample analysis.

NO ACTION THIS No NOTE: If rad monitor reading can not be validated within 60 CATEGORY minutes, declaration must be based on actual monitor reading.

APF 06-002-01, Rev. 17A [Commitment Step 3.2.2] Page 2 of 37 EMERGENCY ACTION LEVELS BASES-1, RADIOACTIVE EFFLUENT RELEASE PROCESS RAD MONITORS GTRE 21B Unit Vent WRGM GHRE 10B Rad Waste WRGM GTRE 21A Unit Vent P-I monitor GHRE 10A Rad Waste Vent P-I monitor BMRE 52 SG Blowdown EFF monitor LERE 59 Turbine Building Drain Effluent HFRE 95 Disch to Waste Water Treatment 1-RER 1. - MODES; ALL This box is used to indicate unexpected liquid and gaseous release rates or releases greater than ODCM allowable values.

1-RER 2. - MODES: ALL Valid means that a radiation monitor reading has been confirmed by the operators. Validation is not limited to chemistry sampling, but may include methods such as radiation surveys or review of area radiation monitors.

Monitor indications are calculated on the basis of the methodology of the site Off-site Dose Calculation Manual (ODCM), or other site procedures that are used to demonstrate compliance with 10 CFR 20 and/or 10 CFR 50 Appendix I requirements. The High alarm setpoint is based on the 10CFR20 maximum radioisotopic release limit. Annual average meteorology is used where allowed.

1-RER 3.- MODES; ALL Valid means that a radiation monitor reading has been confirmed by the operators. Validation is not limited to chemistry sampling, but may include methods such as radiation surveys or review of area radiation monitors.

The 100 mR TEDE in this initiating condition is based on the 10 CFR 20 annual average population exposure. This value also provides a desirable difference (one order of magnitude) between the Alert, Site Area Emergency, and General Emergency classes. It is deemed that exposures less than this limit are not consistent with the Site Area Emergency class description.

The 500 mR thyroid (CDE) was established in consideration of the 1:5 ratio of the EPA Protective Action Guidelines for whole body and thyroid.

Integrated doses are generally not monitored in real-time. In establishing the emergency action levels, a duration of one hour was assumed, and that the EAL is based on a site boundary dose of 100 mR/hour TEDE or 500 mR/hour thyroid (CDE), whichever is more limiting.

Unit Vent and Radwaste Vent numbers were obtained using the WCGS Emergency Dose Calculation Program (EDCP). Meteorological data was obtained from Wolf Creek USAR Table 2.3-25 for wind speed and 2.3-37 for stability class. Ventilation flow rates selected are design flow rates.

1-RER 4 - MODES: ALL Valid means that a radiation monitor reading has been confirmed by the operators. Validation is not limited to chemistry sampling, but may include methods such as radiation surveys or review of area radiation monitors.

The 1000 mR TEDE and the 5000 mR thyroid (CDE) dose are based on the EPA protective action guidance which indicates that public protective actions are indicated if the dose exceeds 1 Rem TEDE or 5 Rem thyroid (CDE). This is consistent with the emergency class description for a General Emergency. This level constitutes the upper level of the desirable difference for the Site Area Emergency. Actual meteorology is specifically identified in the initiating condition since it gives the most accurate dose assessment. Actual meteorology should be used whenever possible.

Integrated doses are generally not monitored in real-time. In establishing the emergency action levels, a duration of one hour was assumed, and that the EAL is based on site boundary doses for either TEDE or thyroid (CDE), whichever is more limiting.

Unit Vent and Radwaste Vent numbers were obtained using the WCGS Emergency Dose Calculation Program (EDCP). Meteorological data was obtained from Wolf Creek USAR Table 2.3-25 for wind speed and 2.3-37 for stability class. Ventilation flow rates selected are design flow rates.

APF 06-002-01, Rev. 17A [Commitment Step 3.2.2] Page 3 of 37 EMERGENCY ACTION LEVELS BASES-1, RADIOACTIVE EFFLUENT RELEASE 1-RER 5. - MODES: ALL Valid means that a radiation monitor reading has been confirmed by the operators to be correct.

This IC addresses increased radiation levels that impede necessary access to WCGS, or other areas containing equipment that must be operated locally, in order to maintain safe operation or perform a safe shutdown. It is this impaired ability to operate the plant that results in the actual or potential substantial degradation of the level of safety of the plant. The cause and/or magnitude of the increase in radiation levels is not a concern of this IC. Consider the source or cause of the increased radiation levels and determine if any other IC may be involved. For example, a dose rate of 15 mR/hr in the Control Room may be a problem in itself. However, the increase may also be indicative of high dose rates in the Containment Building due to a LOCA. In this latter case, an SAE or GE may be indicated by the fission product barrier matrix ICs. A YES answer to box 1-RER5 requires that the site area boundary be monitored and if limits are exceeded an SAE or GE should be declared.

Areas requiring continuous occupancy include the Control Room and the Central Alarm Station.Section III.D.3 of NUREG-0737, "Clarification of TMI Action Plan Requirements",

provides that the 15 mR/hr value can be averaged over 30 days. The value is used here without averaging, as a 30 day duration implies an event potentially more significant than an Alert.

For areas requiring infrequent access the value is based on radiation levels which could result in exposure control measures intended to maintain doses within normal occupational exposure guidelines and limits (i.e., 10 CFR 20), and in doing so, will impede necessary access.

The following is a list of the Area Rad Monitors and their average readings obtained by archive search for the period 7/1/04 - 9/31/04. The list also has a column indicating the x1000 mrem/hr reading.

Instrument Number Location Average (mrem/hr) x1000 (mrem/hr) 0-SD-RE-1 Radwaste Building Corridor, Basement .130 130.

0-SD-RE-2 Radwaste Building Corridor, Basement .112 112.

0-SD-RE-3 Radwaste Building Corridor, Basement .267 267.

0-SD-RE-4 Radwaste Building Corridor, Ground Flr .113 113.

0-SD-RE-5 Radwaste Building Corridor, Ground Flr .136 136.

0-SD-RE-6 Solid Radwaste Area .137 137.

0-SD-RE-7 Truck Space .394 394.

0-SD-RE-8 Sample Laboratory .102 102.

0-SD-RE-9 RW Bldg Valve Room Corridor .285 285.

0-SD-RE-10 RW Bldg Valve Room Corridor .244 244.

0-SD-RE-11 RW Bldg HVAC Filter Unit .135 135.

0-SD-RE-12 Aux Bldg Corridor Basement .115 115.

0-SD-RE-13 Aux Bldg Corridor Basement .323 323.

0-SD-RE-14 Aux Bldg Corridor Basement .311 311.

0-SD-RE-15 Aux Bldg Corridor Basement .128 128.

0-SD-RE-16 Aux Bldg Corridor Basement .182 182.

0-SD-RE-17 Pipe Tunnel & Personnel Access .361 361.

0-SD-RE-18 Aux Bldg Ground Floor Corridor .295 295.

0-SD-RE-19 Aux Bldg Ground Floor Corridor .181 181.

0-SD-RE-20 Aux Bldg Valve Room Corridor Ground Flr. .273 273.

0-SD-RE-21 Aux Bldg Valve Room Corridor Ground Flr. .334 334.

0-SD-RE-22 Aux Bldg Corridor Ground Floor .409 409.

0-SD-RE-23 Aux Bldg Corridor Ground Floor .153 153.

0-SD-RE-24 Sample Room 1.055 1055.

0-SD-RE-25 Filter Unit Aux Bldg .132 132.

0-SD-RE-26 RHR Heat Exchanger Outside .182 182.

(Continued)

APF 06-002-01, Rev. 17A [Commitment Step 3.2.2] Page 4 of 37 EMERGENCY ACTION LEVELS BASES-1, RADIOACTIVE EFFLUENT RELEASE 1-RER 5. - MODES: ALL Instrument Number Location Average (mrem/hr) x1000 (mrem/hr) 0-SD-RE-27 Ctmt Purge Filter Unit .187 187.

0-SD-RE-28 Personnel Hatch .296 296.

0-SD-RE-29 Hot Machine Shop .118 118.

0-SD-RE-30 Hot Instrument Shop .171 171.

0-SD-RE-31 Hot Laboratory .157 157.

0-SD-RE-32 Control Bldg Corridor .181 181.

0-SD-RE-33 Control Room .103 103.

0-SD-RE-34 Cask Handling Area .154 154.

0-SD-RE-35 New Fuel Storage Area .125 125.

0-SD-RE-36 New Fuel Storage Area .127 127.

0-SD-RE-37 Spent Fuel Pool Area .142 142.

0-SD-RE-38 Spent Fuel Pool Area .142 142.

0-SD-RE-39 ** Seal Table Area 9.64 9640.

0-SD-RE-40 Personnel Access Hatch Area 6.61 6610.

0-SD-RE-41 Manipulator Bridge Crane 23.5 10000.

0-SD-RE-42 Containment Building 2.28 2280 0-SD-RE-43 Technical Support Center N/A N/A 0-SD-RE-44 Emergency Off-site Facility N/A N/A 0-SD-RE-47 Pass Sampling Room .004 4.

- This data was modified from the above archive review to reflect changes made to the transmitter/indication system for this point.

1-RER 6. - MODES: ALL Valid means that a radiation monitor reading has been confirmed by the operators. Validation is not limited to chemistry sampling, but may include methods such as: radiation surveys or review of area radiation monitors.

The term "Unplanned", as used in this context, includes any release for which a radioactive discharge permit was not prepared, or a release that exceeds the conditions (e.g., minimum dilution flow, maximum discharge flow, alarm setpoints, etc.) on the applicable permit.

Unplanned releases in excess of 10CFR50 limits that continue for 60 minutes or longer represent an uncontrolled situation and hence, a potential degradation in the level of safety. The final integrated dose (which is very low in the Unusual Event emergency class) is not the primary concern here; it is the degradation in plant control implied by the fact that the release was not isolated within 60 minutes. Therefore, it is not intended that the release be averaged over 60 minutes. For example, a release of 4 times T/S for 30 minutes does not exceed this initiating condition. The Emergency Manager should not wait until 60 minutes has elapsed, but should declare the event as soon as it is determined that the release duration has or will likely exceed 60 minutes.

APF 06-002-01, Rev. 17A [Commitment Step 3.2.2] Page 5 of 37 EMERGENCY ACTION LEVELS EAL-2, STEAM GENERATOR TUBE FAILURE START 2-SGTF1 2-SGTF2 2-SGTF3 2-SGTF4 2-SGTF5 Mode: 1-4 Mode: 1-4 Mode: 1-4 Mode: 1-4 Mode: 1-4 SG tube rupture is Yes Is any faulted SG releasing steam SG tube leakage is Failed fuel is indicated by Is ruptured SG also Yes Yes greater than capacity outside CTMT which cannot be Yes Yes greater than 150 any of the following: a faulted SG?

of one CCP No stopped until SG has blown dry?

GPD through any

Core Cooling Orange discharging to one SG Path 2-SGTF7 Yes normal charging No OR Mode: 1-4 header.

Core Cooling Red Path Any of the following conditions exists:

OR

One or more CISA, CISB, OR CPIS CTMT penetrations

Valid Heat Sink Red No 2-SGTF6 CANNOT be isolated after receipt of an automatic isolation GENERAL Path Mode: 1-4 signal EMERGENCY OR OR Is any faulted SG

CTMT Red Path No

2 CET Greater than releasing steam o OR 1400 inside CTMT

Core Cooling Red Path for greater than 15 minutes OR Yes which cannot be OR 2-SGTF8 Yes

Greater Than 2% Fuel stopped until SG

Unexplained CTMT pressure decrease Mode: 1-4 Clad Damage has blown dry? OR Is GT RE 59 or GT RE 60

CTMT hydrogen greater than or equal to 4%

OR No reading greater than or No

CTMT pressure greater than 27 psig AND less than one train of equal to 2500 R/Hr?

No CTMT depressurization equipment available 2-SGTF9 2-SGTF10 No Mode: 1-4 Mode: 1-4 NOTE: One train of CTMT depressurization is defined as one train of CTMT spray AND one train of CTMT coolers.

Is any faulted SG SG tube rupture is greater releasing steam outside than capacity of one CCP Yes CTMT which cannot be Yes discharging to normal stopped until SG has Yes charging header. 2-SGTF13 2-SGTF11 blown dry?

Mode: 1-4 Mode: 1-4 Any of the following conditions exists:

No No

One or more CISA, CISB, OR CPIS CTMT Is ruptured SG SITE AREA Yes penetrations CANNOT be isolated after receipt of an also a faulted SG? EMERGENCY automatic isolation signal OR 2-SGTF12

CTMT Red Path No Mode: 1-4 OR Is any faulted SG

Core Cooling Red Path for greater than 15 minutes releasing steam inside OR CTMT which cannot be Yes

Unexplained CTMT pressure decrease No ALERT stopped until SG has OR blown dry?

CTMT hydrogen greater than or equal to 4%

OR No

CTMT pressure greater than 27 psig AND less than one train of CTMT depressurization equipment available NOTIFICATION OF UNUSUAL EVENT NOTE: One train of CTMT depressurization is defined as one train of CTMT spray AND one train of CTMT coolers NO ACTION THIS CATEGORY

APF 06-002-01, Rev. 17A [Commitment Step 3.2.2] Page 6 of 37 EMERGENCY ACTION LEVELS BASES-2, STEAM GENERATOR TUBE FAILURE 2-SGTF 1 - MODES: 1 THROUGH 4 This initiating condition indicates the SG tube leakage is at a point where Technical Specifications require a shutdown. As such, it is a precursor of a potentially worse condition.

2-SGTF 2 - MODES: 1 THROUGH 4

l. This EAL is for using Critical Safety Function Status Tree (CSFST) monitoring and functional recovery procedures. RED path indicates an extreme challenge to the safety function.

ORANGE path indicates a severe challenge to the safety function.

Core Cooling - ORANGE indicates subcooling has been lost and that some clad damage may occur. Heat Sink - RED indicates the steam generator heat sink function is under extreme challenge and thus these two items indicate potential loss of the Fuel Clad Barrier. A "Valid" Heat Sink RED requires RCS Pressure greater than or equal to SG pressure.

Core Cooling - RED indicates significant superheating and core uncovery and is considered to indicate loss of the Fuel Clad Barrier.

2. Two core exit thermocouples with valid readings greater than 1400F indicates significant clad heating and thus the Fuel Clad Barrier is considered lost. The core exit thermocouples provide an adequate measure of core temperatures to estimate core temperatures at which potential cladding damage (i.e. 1400F) and core over-temperature (above about 2400F) may be occurring.

3 A core damage assessment (EPP 06-017) corresponding to about 2% to 5% fuel clad damage indicates significant degradation and thus the Fuel Clad Barrier is considered lost.

2-SGTF 3. - MODES: 1 THROUGH 4 RCS Leak Rate EAL is based on the inability to maintain normal liquid inventory within the Reactor Coolant System (RCS) by normal operation of the Chemical and Volume Control System which is considered as one centrifugal charging pump discharging to the charging header at maximum rate.

2-SGTF 4. - MODES: 1 THROUGH 4 A check for SG fault is made to determine if the next fission product boundary is under challenge or lost. The release path looked for is either a faulted, ruptured SG or a faulted SG to a challenged Containment.

2-SGTF 5. - MODES: 1 THROUGH 4 Once a faulted SG has been determined, a release path via a faulted, ruptured SG is checked.

2-SGTF 6. - MODES: 1 THROUGH 4 This box checks for an unisolable secondary side steam release to the Containment atmosphere.

2-SGTF 7. - MODES: 1 THROUGH 4

1. Containment Isolation Valve Status After Receipt of Automatic or Manual Phase A, Phase B, or CPIS Isolation: The failure to complete Phase A, Phase B, or CPIS isolation of each containment penetration by at least one device in each penetration if by design that penetration is assumed to isolate means the containment barrier must be considered breached. Note that ESW does not receive an automatic or manual isolation signal. The intent of this basis is only applicable to those penetrations designed to isolate by an automatic signal. Local isolation of one of the required devices in a timely manner would meet the intent and would be a NO answer for this box.

2. This EAL is for Using Critical Safety Function Status Tree (CSFST) Monitoring and Functional recovery procedures: RED path indicates an extreme challenge to the safety function derived from appropriate instrument readings and/or sampling results, and thus represents a potential loss of containment. Conditions leading to a containment RED path result from RCS barrier and

/or Fuel Clad Barrier Loss. Thus, this EAL is primarily a discriminator between Site Area Emergency and General Emergency representing a potential loss of the third barrier.

In this EAL, the function restoration procedures are those emergency operating procedures that address the recovery of the core cooling critical safety functions. The procedure is considered effective if the core temperature is decreasing or if the vessel level is increasing.

The conditions in this potential loss EAL represent imminent melt sequence which, if not corrected, could lead to vessel failure and an increased potential for containment failure. In conjunction with the core exit thermocouple EALs in the Fuel and RCS barriers columns, this EAL would result in the declaration of a General Emergency - loss of two barriers and the potential loss of a third. If the function restoration procedures are ineffective, there is no "success" path.

Severe accident analyses (e.g., NUREG-1150) have concluded that function restoration procedures can arrest core degradation within the reactor vessel in a significant fraction of the core damage scenarios, and that the likelihood of CTMT failure is very small in these events. Given this, it is appropriate to provide a reasonable period to allow function restoration procedures to arrest the core melt sequence. Whether or not the procedures will be effective should be apparent within 15 minutes. The declaration should be made as soon as it is determined that the procedures have been, or will be ineffective.

(Continued)

APF 06-002-01, Rev. 17A [Commitment Step 3.2.2] Page 7 of 37 EMERGENCY ACTION LEVELS BASES-2, STEAM GENERATOR TUBE FAILURE 2-SGTF 7. - MODES: 1 THROUGH 4

3. An unexplained CTMT pressure decrease is a check of CTMT integrity. After the CTMT is initially pressurized, a decrease in CTMT pressure can indicate loss of CTMT. Likewise, at the onset of an event determined to be inside Containment, failure of the containment to pressurize may be indicative of a significant leak path. Care must be exercised here as action of CTMT Coolers and Sprays as well as RCS pressure decrease can cause a slow decrease in CTMT pressure. To determine if there is an unexplained CTMT pressure decrease, CTMT pressure and temperature should be trended for a short period of time (<15 minutes) to check for a deviation from the expected pressure/temperature relationship to determine the status of CTMT integrity.

Other items to check to verify the presence of leakage are the area and process rad monitors for the auxiliary building to see if an unexplained increase in radiation readings is noted. Also the equipment and emergency escape hatches should be considered and checked for possible leakage. This condition is to provide indication of loss of CTMT.

4. Existence of an explosive mixture means an H2 Concentration greater than 4% and the potential for an explosive mixture and possible damage to Containment exists.

5. Having less than one train of CTMT Depressurization Systems is a potential loss of CTMT in that the heat removal/ depressurization system (i.e. CTMT Spray and CTMT Coolers) are either lost or performing in a degraded manner, as indicated by CTMT pressure greater than 27 PSIG (setpoint at which the equipment was supposed to operate). Having a Cooler set on one train and Spray Pump on the other train constitutes one train and does not yield a YES answer.

2-SGTF 8. - MODES: 1 THROUGH 4 This reading is a value which indicates significant fuel damage well in excess of the EALs associated with both loss of Fuel Clad and loss of RCS Barriers. A major release of radioactivity requiring offsite protective actions from core damage is not possible unless a major failure of fuel cladding allows radioactive material to be released from the core into the reactor coolant.

Regardless of whether Containment is challenged, this amount of activity in Containment, if released, could have such severe consequences that it is prudent to treat this as a potential loss of Containment, such that a General Emergency declaration is warranted. NUREG-1228, "Source Estimations During Incident Response to Severe Nuclear Power Plant Accidents," indicates that such conditions do not exist when the amount of clad damage is less than 20%. Reading from AN 98-029, Basis for Wolf Creek Core Damage Assessment Guidance.

2-SGTF 9. - MODES: 1 THROUGH 4 RCS Leak Rate EAL is based on the inability to maintain normal liquid inventory within the Reactor Coolant System (RCS) by normal operation of the Chemical and Volume Control System which is considered as one centrifugal charging pump discharging to the charging header at maximum rate. For leakage less than the capacity of one centrifugal charging pump but greater than Technical Specification allowed leakage, an NUE is classified for increased awareness of a possible rapid loss of reactor coolant volume.

2-SGTF 10. - MODES: 1 THROUGH 4 See 2-SGTF 4.

2-SGTF 11. - MODES: 1 THROUGH 4 See 2-SGTF 5.

2-SGTF 12. - MODES: 1 THROUGH 4 See 2-SGTF 6.

2-SGTF 13. - MODES: 1 THROUGH 4 See 2-SGTF 7.

APF 06-002-01, Rev. 17A [Commitment Step 3.2.2] Page 8 of 37 EMERGENCY ACTION LEVELS EAL-3, LOSS OF REACTOR COOLANT BOUNDARY START 3-LRCB1 3-LRCB2 3-LRCB3 3-LRCB4 Mode: 1-4 Mode: 1-4 Mode: 1-4 Mode: 1-4 Any of the following conditions exists:

RCS break is indicated by any of the Failed fuel is indicated by any of the Safety Injection Initiated Yes Yes Yes

One or more CISA, CISB, OR CPIS CTMT penetrations following: following:

AND CANNOT be isolated after receipt of an automatic

RCS leak greater than capacity of one

Core Cooling Orange Path RCS leakage is greater than any of isolation signal CCP discharging to normal charging OR the following: OR header

Core Cooling Red Path

Unidentified leakage greater than

CTMT Red Path OR OR 10 GPM OR

Safety Injection Initiated

Valid Heat Sink Red Path OR

Core Cooling Red Path for greater than 15 minutes OR OR

Pressure boundary leakage OR

Valid Heat Sink Red Path

2 CET Greater than 1400o greater than 10 GPM

Unexplained CTMT pressure decrease OR OR OR OR

Integrity Red Path

GT RE 59 or GT RE 60 reading

Identified leakage greater than 25

Indications of LOCA outside CTMT OR greater than or equal to 250 R/hr GPM (except S/G tube leakage) OR

Both CTMT atmospheric rad monitors OR

CTMT hydrogen greater than or equal to 4%

increase off scale high

Greater Than 2% Fuel Clad OR (GT RE 31 & GT RE 32) Damage

CTMT pressure greater than 27 psig AND less than one train of CTMT depressurization equipment available No 3-LRCB9 No 3-LRCB5 No 3-LRCB6 Yes NOTE: One train of CTMT depressurization is defined as Mode: 1-4 Mode: 1-4 Mode: 1-4 one train of CTMT spray AND one train of CTMT coolers Leak isolations in OFN BB--007 or ECCS flow ECCS flow less Yes Yes OFN BB-031 are unsuccessful required than 225 gpm 3-LRCB8 No Yes AND Mode: 1-4 RCS leakage is greater than any of the following: 3-LRCB7 No No GT RE 59 or GT RE 60

Unidentified leakage greater than Mode: 1-4 GENERAL reading greater than or Yes 10 GPM Any of the following conditions exists: equal to 2500 R/Hr EMERGENCY OR

One or more CISA, CISB, OR CPIS CTMT penetrations CANNOT be

Pressure boundary leakage isolated after receipt of an automatic isolation signal greater than 10 GPM OR No OR

CTMT Red Path SITE AREA

Identified leakage greater than 25 OR Yes EMERGENCY GPM (except S/G tube leakage)

Unexplained CTMT pressure decrease OR

Indications of LOCA outside CTMT OR No ALERT No

CTMT pressure greater than 27 psig AND less than one train of CTMT depressurization equipment available NOTE: One train of CTMT depressurization is defined as one train of CTMT spray AND one train of CTMT coolers.

NOTIFICATION OF UNUSUAL EVENT NO ACTION THIS CATEGORY

APF 06-002-01, Rev. 17A [Commitment Step 3.2.2] Page 9 of 37 EMERGENCY ACTION LEVELS BASES-3, LOSS OF REACTOR COOLANT BOUNDARY 3-LRCB 1.- MODES: 1 THROUGH 4 This IC is included as an Unusual Event because it may be a precursor of more serious conditions and, as a result, is considered to be a potential degradation of the level of safety of the plant. The 10 gpm value for the unidentified and pressure boundary leakage was selected as it is observable with normal Control Room indications. The terms unidentified, pressure boundary, and identified are the same as those in the Definitions Section of the Technical Specifications. Lesser values must generally be determined through time-consuming surveillance tests. The value for "identified" leakage recognizes the allowable 10 gpm Technical Specification Limit, excluding S/G tube leakage, and the 25 gpm value is sufficiently above this to allow determination from normal indication. Since personnel would try to isolate the leak, a provision has been added to see if efforts were successful. Successful leak isolation results in "No Action" and is appropriate. Leak isolation should be considered unsuccessful if any of the following occurs: 1) the OFN is completed and the leak is not stopped, 2) the OFN loops while trying to isolate the leak, OR 3) the foldout page directs the Operator to the EMG network. Also, leak isolation should be considered as unsuccessful, if PZR level cannot be maintained > 6%

OR if Rx is tripped and sub cooling is <30º. S/G tube leakage is dealt with in EAL-2.

It is not intended that this tree be used during accident mitigation strategies where the operator induces the LOCA as part of the event mitigation (e.g. RCS bleed and feed in response to loss of all secondary heat sink).

3-LRCB 2.- MODES: 1 THROUGH 4

1. RCS Leak Rate: The "Potential Loss" EAL is based on the inability to maintain normal liquid inventory within the Reactor Coolant System (RCS) by normal operation of the Chemical and Volume Control System which is considered as one centrifugal charging pump discharging to the charging header at maximum rate. A Pressurizer vapor space leak could cause pressure to decrease to the Safety Injection setpoint which should be answered YES because this should be considered a loss of liquid inventory.

2. Critical Safety Function Status: This EAL is for using Critical Safety Function Status Tree (CSFST) monitoring and functional recovery procedures. RED path indicates an extreme challenge to the safety function derived from appropriate instrument readings, and these CSFs indicate a potential loss of RCS barrier. "Valid" Heat Sink RED requires RCS pressure greater than or equal to SG pressure.

3. Containment Radiation Monitoring: Containment Atmospheric monitors (GT RE 31 and 32) are used because they will sense an RCS leak before the CHARMS (GTRE 59 and 60). It is intended to include any of the monitor channels i.e. either Gas, Particulate or Iodine. With an RCS leak with RCS Activity at "Normal" values the CHARMS would not detect this leak in a timely fashion. Off-scale high readings on GT RE 31 & 32 monitors are as follows: Gas > 10E-2 C/cm3, Iodine > 10E-6 C/cm3, & Particulate > 10E-7 C/cm3 3-LRCB 3.- MODES: 1 THROUGH 4

1. Critical Safety Function Status : This EAL is for using Critical Safety Function Status Tree (CSFST) monitoring and functional recovery procedures. RED path indicates an extreme challenge to the safety function. ORANGE path indicates a severe challenge to the safety function.

Core Cooling - ORANGE indicates subcooling has been lost and that some clad damage may occur. Heat Sink - RED indicates the SG heat sink function is under extreme challenge and thus these two items indicate potential loss of the Fuel Clad Barrier. "Valid" Heat Sink RED requires RCS Pressure greater than or equal to SG pressure.

Core Cooling - RED indicates significant superheating and core uncovery and is considered to indicate loss of the Fuel Clad Barrier.

2. Two core exit thermocouples with valid readings greater than 1400F indicates significant clad heating and thus the Fuel Clad Barrier is considered lost. The core exit thermocouples provide an adequate measure of core temperatures to estimate core temperatures at which potential cladding damage (i.e. 1400F) and core over-temperature (above about 2400F) may be occurring.

3. Reading from EPP 06-017, Figure 3, Curve with RCS Pressure >1600 psig with Containment Spray (AN 98-029, Basis For Wolf Creek Core Damage Assessment Guidance, Appendix 6 CRM3 Case D (CRM3D): 2% @ 0 hours0 days 0 hours 0 weeks 0 months after reactor trip OR 5% @ 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after reactor trip).

4. A core damage assessment (EPP 06-017) corresponding to about 2% to 5% fuel clad damage indicates significant degradation and thus the Fuel Clad Barrier is considered lost.

3-LRCB 4 - MODES: 1 THROUGH 4

1. Containment Isolation Valve Status After Receipt of Automatic or Manual Phase A, Phase B, or CPIS Isolation: The failure to complete Phase A, Phase B, or CPIS isolation of each containment penetration by at least one device in each penetration if by design that penetration is assumed to isolate means the containment barrier must be considered breached. Note that ESW does not receive an automatic or manual isolation signal. The intent of this basis is only applicable to those penetrations designed to isolate by an automatic signal. Local isolation of one of the required devices in a timely manner would meet the intent and would be a NO answer for this box.

(Continued)

APF 06-002-01, Rev. 17A [Commitment Step 3.2.2] Page 10 of 37 EMERGENCY ACTION LEVELS BASES-3, LOSS OF REACTOR COOLANT BOUNDARY 3-LRCB 4 - MODES: 1 THROUGH 4

2. Using Critical Safety Function Status Tree (CSFST) Monitoring and Functional recovery procedures: RED path indicates an extreme challenge to the safety function derived from appropriate instrument readings and/or sampling results, and thus represents a potential loss of containment. Conditions leading to a containment RED path result from RCS barrier and/or Fuel Clad Barrier Loss. Thus, this EAL is primarily a discriminator between Site Area and General Emergency representing a potential loss of the third barrier.

In this EAL, the function restoration procedures are those emergency operating procedures that address the recovery of the core cooling critical safety function. The procedure is considered effective if the core temperature is decreasing or if the vessel level is increasing.

The conditions in this potential loss EAL represent imminent melt sequence which, if not corrected, could lead to vessel failure and an increased potential for containment failure. In conjunction with the core exit thermocouple EALs in the Fuel and RCS barriers, this EAL would result in the declaration of a General Emergency - loss of two barriers and the potential loss of a third. If the function restoration procedures are ineffective, there is no "success" path.

Severe accident analyses (e.g., NUREG-1150) have concluded that function restoration procedures can arrest core degradation within the reactor vessel in a significant fraction of the core damage scenarios, and that the likelihood of CTMT failure is very small in these events. Given this, it is appropriate to provide a reasonable period to allow function restoration procedures to arrest the core melt sequence. Whether or not the procedures will be effective should be apparent within 15 minutes. The declaration should be made as soon as it is determined that the procedures have been, or will be ineffective.

3. An unexplained CTMT pressure decrease is a check of CTMT integrity. After the CTMT is initially pressurized, a decrease in CTMT pressure can indicate loss of CTMT. Likewise, at the onset of an event determined to be inside Containment, failure of the containment to pressurize may be indicative of a significant leak path. Care must be exercised here as action of CTMT Coolers and Sprays as well as RCS pressure decrease can cause a slow decrease in CTMT pressure. To determine if there is an unexplained CTMT pressure decrease, CTMT pressure and temperature should be trended for a short period of time (<15 minutes) to check for a deviation from the expected pressure/temperature relationship to determine the status of CTMT integrity.

Other items to check to verify the presence of leakage are the area and process rad monitors for the auxiliary building to see if an unexplained increase in radiation readings is noted. Also the equipment and emergency escape hatches should be considered and checked for possible leakage. This condition is to provide indication of loss of CTMT.

4. "Indications of LOCA outside CTMT" is intended to allow the use of Area Radiation and Process Radiation Monitors, radiological surveys, sump level increase outside CTMT, etc. to determine loss of CTMT. It is not necessary to be in EMG C-12, although if that is where the plant status puts the operator then CTMT is lost.

5. Existence of an explosive mixture means an H2 Concentration greater than 4% and the potential for an explosive mixture and possible damage to Containment exists.

6. Having less than one train of CTMT Depressurization Systems is a potential loss of CTMT in that the heat removal/ depressurization system (i.e. CTMT Spray and CTMT Coolers) are either lost or performing in a degraded manner, as indicated by CTMT pressure greater than 27 PSIG (setpoint at which the equipment was supposed to operate). Having a Cooler set on one train and Spray Pump on the other train constitutes one train and does not yield a YES answer.

3-LRCB 5 - MODES: 1 THROUGH 4 This indicates a need for ECCS flow to maintain RCS inventory.

3-LRCB 6 - MODES: 1 THROUGH 4 This IC is used to determine if any ECCS system is capable of delivering a sufficient volume of water to the core for acceptable core cooling. 225 gpm was chosen because it is conservatively larger then Tech Spec delta P requirement of 210 gpm at 2400 PSID which is what a centrifugal charging pump could deliver to smallest size break. Seal injection flow should not be used as part of ECCS flow.

3-LRCB 7 - MODES: 1 THROUGH 4 See 3-LRCB 4. Containment hydrogen is not applicable due to no Fuel Failure on this path.

3-LRCB 8 - MODES: 1 THROUGH 4 This reading is a value which indicates significant fuel damage well in excess of the EALs associated with both loss of Fuel Clad and loss of RCS Barriers. A major release of radioactivity requiring offsite protective actions from core damage is not possible unless a major failure of fuel cladding allows radioactive material to be released from the core into the reactor coolant.

Regardless of whether Containment is challenged, this amount of activity in Containment, if released, could have such severe consequences that it is prudent to treat this as a potential loss of Containment, such that a General Emergency declaration is warranted. NUREG-1228, "Source Estimations During Incident Response to Severe Nuclear Power Plant Accidents," indicates that such conditions do not exist when the amount of clad damage is less than 20%. Reading from AN 98-029, Basis for Wolf Creek Core Damage Assessment.

APF 06-002-01, Rev. 17A [Commitment Step 3.2.2] Page 11 of 37 EMERGENCY ACTION LEVELS 3-LRCB 9.- MODES: 1 THROUGH 4 This IC is included as an Unusual Event because it may be a precursor of more serious conditions and, as a result, is considered to be a potential degradation of the level of safety of the plant. The 10 gpm value for the unidentified and pressure boundary leakage was selected as it is observable with normal Control Room indications. The terms unidentified, pressure boundary, and identified are the same as those in the Definitions Section of the Technical Specifications. Lesser values must generally be determined through time-consuming surveillance tests. The value for "identified" leakage recognizes the allowable 10 gpm Technical Specification Limit, excluding S/G tube leakage, and the 25 gpm value is sufficiently above this to allow determination from normal indication. Since personnel would try to isolate the leak, a provision has been added to see if efforts were successful. Successful leak isolation results in "No Action" and is appropriate. Leak isolation should be considered unsuccessful if any of the following occurs: 1) the OFN is completed and the leak is not stopped, 2) the OFN loops while trying to isolate the leak, OR 3) the foldout page directs the Operator to the EMG network. Also, leak isolation should be considered as unsuccessful, if PZR level cannot be maintained > 6%

OR if Rx is tripped and sub cooling is <30º. S/G tube leakage is dealt with in EAL-2.

It is not intended that this tree be used during accident mitigation strategies where the operator induces the LOCA as part of the event mitigation (e.g. RCS bleed and feed in response to loss of all secondary heat sink).

APF 06-002-01, Rev. 17A [Commitment Step 3.2.2] Page 12 of 37 EMERGENCY ACTION LEVELS EAL-4, MAIN STEAM LINE BREAK 4-MSLB4 START Mode: 1-4 4-MSLB2 4-MSLB3 o SGTR exists in a faulted SG Yes 4-MSLB1 Mode: 1-4 Mode: 1-4 AND Mode: 1-4 Is any faulted SG 4-MSLB10 Yes Failed fuel is indicated by o SGTR is greater than capacity releasing steam outside Mode: 1-4 Is any SG any of the following: Any of the following conditions exists: of one CCP discharging to Yes Yes CTMT which cannot be normal charging faulted?

Core Cooling Orange

One or more CISA, CISB, OR CPIS CTMT penetrations CANNOT be stopped until SG has Path isolated after receipt of an automatic isolation signal blown dry? No Yes OR OR

Core Cooling Red Path

CTMT Red Path No 4-MSLB9 No OR OR

Valid Heat Sink Red Mode: 1-4

Core Cooling Red Path for greater than 15 minutes Path OR Is SG faulted inside OR Yes

Unexplained CTMT pressure decrease CTMT? GENERAL

2 CET Greater than OR 1400 o

CTMT hydrogen greater than or equal to 4% EMERGENCY OR No OR

Greater Than 2% Fuel

CTMT pressure greater than 27 psig AND less than one train of Clad Damage CTMT depressurization equipment available NOTE: One train of CTMT depressurization is defined as one train of No CTMT spray AND one train of CTMT coolers.

4-MSLB11 No 4-MSLB12 Mode: 1-4 Mode: 1-4 o SGTR exists in a faulted SG AND GT RE 59 or GT RE 60 o SGTR is greater than capacity of Yes reading greater than or Yes 4-MSLB5 one CCP discharging to normal equal to 2500 R/Hr Mode: 1-4 charging Is any faulted SG releasing steam outside CTMT which No No Yes cannot be stopped until SG 4-MSLB6 has blown dry? 4-MSLB8 Mode: 1-4 Mode: 1-4 o SGTR exists in a faulted SG Any of the following conditions exists: AND

One or more CISA, CISB, OR CPIS SITE AREA No Yes o SGTR is greater than capacity of Yes 4-MSLB7 CTMT penetrations CANNOT be isolated EMERGENCY one CCP discharging to normal Mode: 1-4 after receipt of an automatic isolation charging Is SG faulted inside CTMT? Yes signal OR No 4-MSLB13

CTMT Red Path Mode: 1-4 OR ALERT No

Unexplained CTMT pressure decrease Is a main steam line or feed water line No broken in the NOTIFICATION OF Turbine Building or Yes Area 5? UNUSUAL EVENT No NO ACTION THIS CATEGORY

APF 06-002-01, Rev. 17A [Commitment Step 3.2.2] Page 13 of 37 EMERGENCY ACTION LEVELS BASES-4, MAIN STEAM LINE BREAK 4-MSLB 1. - MODES: 1 THROUGH 4 IF NO S/G pressure is decreasing in an uncontrolled manner or is completely depressurized, THEN the S/Gs should be considered as NOT FAULTED. A Main Steam Line break inside or outside containment could cause a potential degradation of the level of safety of the plant.

4-MSLB 2. - MODES: 1 THROUGH 4

1. Critical Safety Function Status: This EAL is for using Critical Safety Function Status Tree (CSFST) monitoring and functional recovery procedures. RED path indicates an extreme challenge to the safety function. ORANGE path indicates a severe challenge to the safety function.

Core Cooling - ORANGE indicates subcooling has been lost and that some clad damage may occur.

Heat Sink - RED indicates the steam generator heat sink function is under extreme challenge and coupled with Core Cooling - ORANGE indicates potential loss of the Fuel Clad Barrier.

"Valid" Heat Sink RED requires that RCS pressure must be greater than or equal to intact SG pressure.

Core Cooling - RED indicates significant superheating and core uncovery and is considered to indicate loss of the Fuel Clad Barrier.

2. Two core exit thermocouples with valid readings greater than 1400F indicates significant clad heating and thus the Fuel Clad Barrier is considered lost. The core exit thermocouples provide an adequate measure of core temperatures to estimate core temperatures at which potential cladding damage (i.e. 1400F) and core over-temperature (above about 2400F) may be occurring.

3. A core damage assessment (EPP 06-017) corresponding to about 2% to 5% fuel clad damage indicates significant degradation and thus the Fuel Clad Barrier is considered lost.

4-MSLB 3. - MODES: 1 THROUGH 4 A check for SG fault outside containment is made to determine if the next fission product boundary is under challenge or lost. The release path looked for is either a faulted, ruptured SG or a faulted SG to a challenged Containment.

4-MSLB 4. - MODES: 1 THROUGH 4 Once a faulted SG has been determined, a release path via a faulted, ruptured SG is checked. Leakage greater than charging capacity of one CCP to a SG constitutes failure of the RCS fission product boundary.

4-MSLB 5. - MODES: 1 THROUGH 4 See 4-MSLB 3.

4-MSLB 6. - MODES: 1 THROUGH 4 See 4-MSLB 4.

4-MSLB 7. - MODES: 1 THROUGH 4 This box checks for an unisolable secondary side steam release to Containment.

4-MSLB 8. - MODES: 1 THROUGH 4

1. Containment Isolation Valve Status After Receipt of Automatic or Manual Phase A, Phase B, or CPIS Isolation: The failure to complete Phase A, Phase B, or CPIS isolation of each containment penetration by at least one device in each penetration if by design that penetration is assumed to isolate means the containment barrier must be considered breached. Note that ESW does not receive an automatic or manual isolation signal. The intent of this basis is only applicable to those penetrations designed to isolate by an automatic signal. Local isolation of one of the required devices in a timely manner would meet the intent and would be a NO answer for this box.

2. Using Critical Safety Function Status Tree (CSFST) Monitoring and Functional recovery procedures: RED path indicates an extreme challenge to the safety function derived from appropriate instrument readings and/or sampling results, and thus represents a potential loss of containment. Conditions leading to a containment RED path can result from RCS barrier and/or Fuel Clad Barrier Loss. In this box the main threat to Containment is from the energy of the steam, thus this box is primarily a discriminator between proceeding on or declaration of NUE.

3. An unexplained CTMT pressure decrease is a check of CTMT integrity. After the CTMT is initially pressurized, a decrease in CTMT pressure can indicate loss of CTMT. Likewise, at the onset of an event determined to be inside Containment, failure of the containment to pressurize may be indicative of a significant leak path. Care must be exercised here as action of CTMT Coolers and Sprays as well as RCS pressure decrease can cause a slow decrease in CTMT pressure. To determine if there is an unexplained CTMT pressure decrease, CTMT pressure and temperature should be trended for a short period of time (<15 minutes) to check for a deviation from the expected pressure/temperature relationship to determine the status of CTMT integrity.

Other items to check to verify the presence of leakage are the area and process rad monitors for the auxiliary building to see if an unexplained increase in radiation readings is noted. Also the equipment and emergency escape hatches should be considered and checked for possible leakage. This condition is to provide indication of loss of CTMT.

4-MSLB 9. - MODES: 1 THROUGH 4 See 4-MSLB-7.

APF 06-002-01, Rev. 17A [Commitment Step 3.2.2] Page 14 of 37 EMERGENCY ACTION LEVELS BASES-4, MAIN STEAM LINE BREAK 4-MSLB 10. - MODES: 1 THROUGH 4

1. Containment Isolation Valve Status After Receipt of Automatic or Manual Phase A, Phase B, or CPIS Isolation: The failure to complete Phase A, Phase B, or CPIS isolation of each containment penetration by at least one device in each penetration if by design that penetration is assumed to isolate means the containment barrier must be considered breached. Note that ESW does not receive an automatic or manual isolation signal. The intent of this basis is only applicable to those penetrations designed to isolate by an automatic signal. Local isolation of one of the required devices in a timely manner would meet the intent and would be a NO answer for this box.

2. Using Critical Safety Function Status Tree (CSFST) Monitoring and Functional recovery procedures: RED path indicates an extreme challenge to the safety function derived from appropriate instrument readings and/or sampling results, and thus represents a potential loss of containment. Conditions leading to a containment RED path result from RCS barrier and/or Fuel Clad Barrier Loss, thus this EAL is primarily a discriminator between Site Area Emergency and General Emergency representing a potential loss of the third barrier.

In this EAL, the function restoration procedures are those emergency operating procedures that address the recovery of the core cooling critical safety function. The procedure is considered effective if the core temperature is decreasing or if the vessel level is increasing.

The conditions in this potential loss EAL represent imminent melt sequence which, if not corrected, could lead to vessel failure and an increased potential for containment failure. In conjunction with the core exit thermocouple EALs in the Fuel and RCS barriers, this EAL would result in the declaration of a General Emergency - loss of two barriers and the potential loss of a third. If the function restoration procedures are ineffective, there is no "success" path.

Severe accident analyses (e.g., NUREG-1150) have concluded that function restoration procedures can arrest core degradation within the reactor vessel in a significant fraction of the core damage scenarios, and that the likelihood of containment failure is very small in these events. Given this, it is appropriate to provide a reasonable period to allow function restoration procedures to arrest the core melt sequence. Whether or not the procedures will be effective should be apparent within 15 minutes. The declaration should be made as soon as it is determined that the procedures have been, or will be ineffective.

3. An unexplained CTMT pressure decrease is a check of CTMT integrity. After the CTMT is initially pressurized, a decrease in CTMT pressure can indicate loss of CTMT. Likewise, at the onset of an event determined to be inside Containment, failure of the containment to pressurize may be indicative of a significant leak path. Care must be exercised here as action of CTMT Coolers and Sprays as well as RCS pressure decrease can cause a slow decrease in CTMT pressure. To determine if there is an unexplained CTMT pressure decrease, CTMT pressure and temperature should be trended for a short period of time (<15 minutes) to check for a deviation from the expected pressure/temperature relationship to determine the status of CTMT integrity.

Other items to check to verify the presence of leakage are the area and process rad monitors for the auxiliary building to see if an unexplained increase in radiation readings is noted. Also the equipment and emergency escape hatches should be considered and checked for possible leakage. This condition is to provide indication of loss of CTMT.

4. Existence of an explosive mixture means an H2 Concentration greater than 4% and the potential for an explosive mixture and possible damage to Containment exists.

5. Having less than one train of Containment Depressurization Systems is a potential loss of Containment in that the heat removal/depressurization system (i.e. Containment Spray and Containment Coolers) are either lost or performing in a degraded manner, as indicated by Containment pressure greater than 27 PSIG (setpoint at which the equipment was supposed to operate). Having a Cooler set on one train and Spray Pump on the other train constitutes one train and does not yield a YES answer.

4-MSLB 11. - MODES: 1 THROUGH 4 See 4-MSLB 4.

4-MSLB 12. - MODES: 1 THROUGH 4 This reading is a value which indicates significant fuel damage well in excess of the EALs associated with both loss of Fuel Clad and loss of RCS Barriers. A major release of radioactivity requiring offsite protective actions from core damage is not possible unless a major failure of fuel cladding allows radioactive material to be released from the core into the reactor coolant.

Regardless of whether Containment is challenged, this amount of activity in Containment, if released, could have such severe consequences that it is prudent to treat this as a potential loss of Containment, such that a General Emergency declaration is warranted. NUREG-1228, "Source Estimations During Incident Response to Severe Nuclear Power Plant Accidents," indicates that such conditions do not exist when the amount of clad damage is less than 20%. Reading was taken from EPP 06-017, CORE DAMAGE ASSESSMENT METHODOLOGY.

4-MSLB 13. - MODES: 1 THROUGH 4 Rapid depressurization of the secondary due to a MSL break which is isolable from the SG's. (i.e. downstream of the MSIV's) A main steam line or feed water break in the Turbine Building or Area 5 could cause a potential degradation of the level of safety of the plant.

APF 06-002-01, Rev. 17A [Commitment Step 3.2.2] Page 15 of 37 EMERGENCY ACTION LEVELS EAL-5, FUEL ELEMENT FAILURE START 5-FEF1 5-FEF2 5-FEF3 Mode: ALL Mode: ALL Mode: ALL Any of the following conditions exists:

Failed fuel is indicated by any of the RCS break is indicated by any of

One or more CISA, CISB, OR CPIS CTMT penetrations CANNOT be following: the following: isolated after receipt of an automatic isolation signal

Core Cooling Orange Path

RCS leak greater than capacity of OR OR one CCP discharging to normal

CTMT Red Path

Core Cooling Red Path charging header OR OR OR

Core Cooling Red Path for greater than 15 minutes

Valid Heat Sink Red Path

Valid Heat Sink Red Path OR OR OR

Unexplained CTMT pressure decrease

2 CET Greater than 1400o

Integrity Red Path GENERAL Yes Yes OR Yes OR

Indication of LOCA outside CTMT EMERGENCY

Greater Than 2% Fuel Clad OR Damage

CTMT hydrogen greater than or equal to 4%

OR OR

SJL016 Reading Greater Than or

CTMT pressure greater than 27 psig AND less than one train of Equal to 1000 uCi/cc CTMT depressurization equipment available OR

GT RE 59 or GT RE 60 reading NOTE: One train of CTMT depressurization is defined as one train of greater than 250 R/Hr CTMT spray AND one train of CTMT coolers.

5-FEF5 No No No 5-FEF6 5-FEF4 Mode: ALL Mode: ALL Mode: ALL o SJL016 HI HI alarm Is GT RE 59 or GT RE 60 reading Any of the following conditions Yes AND greater than or equal to 2500 R/Hr?

exists:

o Analysis shows an increase Yes

One or more CISA, CISB, OR greater than 63 uCi/gm gross CPIS CTMT penetrations activity CANNOT be isolated after receipt No of an automatic isolation signal OR SITE AREA Yes No

Core Cooling Red Path for greater EMERGENCY 5-FEF7 than 15 minutes Mode: ALL OR

Unexplained containment RCS activity greater than pressure decrease Tech Spec 3.4.16 as indicated by:

DEI exceeds limit for greater No than 48 Hr during one continuous interval ALERT OR

DEX exceeds limit for greater than 48 Hr during one continuous interval NOTIFICATION OF Yes UNUSUAL EVENT No NO ACTION THIS CATEGORY

APF 06-002-01, Rev. 17A [Commitment Step 3.2.2] Page 16 of 37 EMERGENCY ACTION LEVELS BASES-5, FUEL ELEMENT FAILURE 5-FEF 1. - MODES: ALL

1. Critical Safety Function Status: This EAL is for using Critical Safety Function Status Tree (CSFST) monitoring and functional recovery procedures. RED path indicates an extreme challenge to the safety function. ORANGE path indicates a severe challenge to the safety function.

Core Cooling - ORANGE indicates subcooling has been lost and that some clad damage may occur.

Core Cooling - RED indicates significant superheating and core uncovery and is considered to indicate loss of the Fuel Clad Barrier.

Heat Sink - RED indicates the steam generator heat sink function is under extreme challenge and thus these two items indicate potential loss of the Fuel Clad Barrier. "Valid" Heat Sink RED means to meet the requirements for entry into FR-H1.

2. Two core exit thermocouples with valid readings greater than 1400F indicates significant clad heating and thus the Fuel Clad Barrier is considered lost. The core exit thermocouples provide an adequate measure of core temperatures to estimate core temperatures at which potential cladding damage (i.e. 1400F) and core over-temperature (above about 2400F) may be occurring.

3. A core damage assessment (EPP 06-017) corresponding to about 2% to 5% fuel clad damage indicates significant degradation and thus the Fuel Clad Barrier is considered lost.

4. A reactivity excursion or mechanical damage event in and of itself may cause fuel damage that is first detected by the Letdown Radiation Monitor (SJL016). Use of a confirmed Letdown Radiation Monitor reading leads to an earlier Alert classification based upon the reactor core status, and therefore, earlier issuance of Protective Action Recommendations.

5. Reading from EPP 06-017, Figure 3, Curve with RCS Pressure >1600 psig with Containment Spray (AN 98-029, Basis For Wolf Creek Core Damage Assessment Guidance, Appendix 6 CRM3 Case D (CRM3D): 2% @ 0 hours0 days 0 hours 0 weeks 0 months after reactor trip OR 5% @ 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after reactor trip).

5-FEF 2. - MODES: ALL

1. RCS Leak Rate: The "Potential Loss" EAL is based on the inability to maintain normal liquid inventory within the Reactor Coolant System (RCS) by normal operation of the Chemical and Volume Control System which is considered as one centrifugal charging pump discharging to the charging header at maximum rate.

2. Critical Safety Function Status: This EAL is for using Critical Safety Function Status Tree (CSFST) monitoring and functional recovery procedures. RED path indicates an extreme challenge to the safety function derived from appropriate instrument readings, and these CSFs indicate a potential loss of RCS barrier.

Heat Sink - RED indicates the steam generator heat sink function is under extreme challenge and thus the Fuel Clad and RCS are threatened. "Valid" Heat Sink RED means to meet the requirements for entry into FR-H1.

Integrity - RED indicates the RCS boundary is extremely challenged and thus with Failed Fuel two barriers would be gone.

5-FEF 3. - MODES: ALL

1. Containment Isolation Valve Status After Receipt of Automatic or Manual Phase A, Phase B, or CPIS Isolation: The failure to complete Phase A, Phase B, or CPIS isolation of each containment penetration by at least one device in each penetration if by design that penetration is assumed to isolate means the containment barrier must be considered breached. Note that ESW does not receive an automatic or manual isolation signal. The intent of this basis is only applicable to those penetrations designed to isolate by an automatic signal. Local isolation of one of the required devices in a timely manner would meet the intent and would be a NO answer for this box.

2. Using Critical Safety Function Status Tree (CSFST) Monitoring and Functional recovery procedures: RED path indicates an extreme challenge to the safety function derived from appropriate instrument readings and/or sampling results, and thus represents a potential loss of containment. Conditions leading to a containment RED path result from RCS barrier and/or Fuel Clad Barrier Loss. Thus, this EAL is primarily a discriminator between Site Area Emergency and General Emergency representing a potential loss of the third barrier.

In this EAL, the function restoration procedures are those emergency operating procedures that address the recovery of the core cooling critical safety function. The procedure is considered effective if the core temperature is decreasing or if the vessel level is increasing.

The conditions in this potential loss EAL represent imminent melt sequence which, if not corrected, could lead to vessel failure and an increased potential for containment failure. In conjunction with the core exit thermocouple EALs in the Fuel and RCS barriers, this EAL would result in the declaration of a General Emergency - loss of two barriers and the potential loss of a third. If the function restoration procedures are ineffective, there is no "success" path.

Severe accident analyses (e.g., NUREG-1150) have concluded that function restoration procedures can arrest core degradation within the reactor vessel in a significant fraction of the core damage scenarios, and that the likelihood of containment failure is very small in these events. Given this, it is appropriate to provide a reasonable period to allow function restoration procedures to arrest the core melt sequence. Whether or not the procedures will be effective should be apparent within 15 minutes. The declaration should be made as soon as it is determined that the procedures have been, or will be ineffective.

(Continued)

APF 06-002-01, Rev. 17A [Commitment Step 3.2.2] Page 17 of 37 EMERGENCY ACTION LEVELS BASES-5, FUEL ELEMENT FAILURE 5-FEF 3. - MODES: ALL

3. An unexplained CTMT pressure decrease is a check of CTMT integrity. After the CTMT is initially pressurized, a decrease in CTMT pressure can indicate loss of CTMT. Likewise, at the onset of an event determined to be inside Containment, failure of the containment to pressurize may be indicative of a significant leak path. Care must be exercised here as action of CTMT Coolers and Sprays as well as RCS pressure decrease can cause a slow decrease in CTMT pressure. To determine if there is an unexplained CTMT pressure decrease, CTMT pressure and temperature should be trended for a short period of time (<15 minutes) to check for a deviation from the expected pressure/temperature relationship to determine the status of CTMT integrity.

Other items to check to verify the presence of leakage are the area and process rad monitors for the auxiliary building to see if an unexplained increase in radiation readings is noted. Also the equipment and emergency escape hatches should be considered and checked for possible leakage. This condition is to provide indication of loss of CTMT.

4. "Indications of LOCA outside Containment" is intended to allow the use of Area Radiation and Process Radiation Monitors, radiological surveys, sump level increase outside containment, etc.

to determine loss of containment. It is not necessary to be in EMG C-12, although if that is where the plant status puts the operator then containment is lost.

5. Existence of an explosive mixture means an H2 Concentration greater than 4% and the potential for an explosive mixture and possible damage to Containment exists.

6. Having less than one train of Containment Depressurization Systems is a potential loss of Containment in that the heat removal/depressurization system (i.e. Containment Spray and Containment Coolers) are either lost or performing in a degraded manner, as indicated by Containment pressure greater than 27 PSIG (setpoint at which the equipment was supposed to operate). Having a Cooler set on one train and Spray Pump on the other train constitutes one train and does not yield a YES answer.

5-FEF 4. - MODES: ALL

1. Containment Isolation Valve Status After Receipt of Automatic or Manual Phase A, Phase B, or CPIS Isolation: The failure to complete Phase A, Phase B, or CPIS isolation of each containment penetration by at least one device in each penetration if by design that penetration is assumed to isolate means the containment barrier must be considered breached. Note that ESW does not receive an automatic or manual isolation signal. So the intent of this basis is only applicable to those penetrations designed to isolate by an automatic signal. Local isolation of one of the required devices in a timely manner would meet the intent and would be a NO answer for this box.

2. Using Critical Safety Function Status Tree (CSFST) Monitoring and Functional recovery procedures: RED path indicates an extreme challenge to the safety function derived from appropriate instrument readings and/or sampling results, and thus represents a potential loss of containment. Conditions leading to a containment RED path result from RCS barrier and/or Fuel Clad Barrier Loss, thus this EAL is primarily a discriminator between Site Area Emergency and General Emergency representing a potential loss of the third barrier.

In this EAL, the function restoration procedures are those emergency operating procedures that address the recovery of the core cooling critical safety function. The procedure is considered effective if the core temperature is decreasing or if the vessel level is increasing.

The conditions in this potential loss EAL represent imminent melt sequence which, if not corrected, could lead to vessel failure and an increased potential for containment failure. In conjunction with the core exit thermocouple EALs in the Fuel and RCS barriers, this EAL would result in the declaration of a General Emergency - loss of two barriers and the potential loss of a third. If the function restoration procedures are ineffective, there is no "success" path.

Severe accident analyses (e.g., NUREG-1150) have concluded that function restoration procedures can arrest core degradation within the reactor vessel in a significant fraction of the core damage scenarios, and that the likelihood of containment failure is very small in these events. Given this, it is appropriate to provide a reasonable period to allow function restoration procedures to arrest the core melt sequence. Whether or not the procedures will be effective should be apparent within 15 minutes. The declaration should be made as soon as it is determined that the procedures have been, or will be ineffective. The reactor vessel level chosen should be consistent with the emergency response procedures applicable to the facility.

3. An unexplained CTMT pressure decrease is a check of CTMT integrity. After the CTMT is initially pressurized, a decrease in CTMT pressure can indicate loss of CTMT. Likewise, at the onset of an event determined to be inside Containment, failure of the containment to pressurize may be indicative of a significant leak path. Care must be exercised here as action of CTMT Coolers and Sprays as well as RCS pressure decrease can cause a slow decrease in CTMT pressure. To determine if there is an unexplained CTMT pressure decrease, CTMT pressure and temperature should be trended for a short period of time (<15 minutes) to check for a deviation from the expected pressure/temperature relationship to determine the status of CTMT integrity.

Other items to check to verify the presence of leakage are the area and process rad monitors for the auxiliary building to see if an unexplained increase in radiation readings is noted. Also the equipment and emergency escape hatches should be considered and checked for possible leakage. This condition is to provide indication of loss of CTMT.

5-FEF 5. - MODES: ALL This reading is a value which indicates significant fuel damage well in excess of the EALs associated with both loss of Fuel Clad and loss of RCS Barriers. A major release of radioactivity requiring offsite protective actions from core damage is not possible unless a major failure of fuel cladding allows radioactive material to be released from the core into the reactor coolant.

Regardless of whether Containment is challenged, this amount of activity in Containment, if released, could have such severe consequences that it is prudent to treat this as a potential loss of Containment, such that a General Emergency declaration is warranted. NUREG-1228, "Source Estimations During Incident Response to Severe Nuclear Power Plant Accidents," indicates that such conditions do not exist when the amount of clad damage is less than 20%. Reading was taken from EPP 06-017, CORE DAMAGE ASSESSMENT METHODOLOGY.

APF 06-002-01, Rev. 17A [Commitment Step 3.2.2] Page 18 of 37 EMERGENCY ACTION LEVELS BASES-5, FUEL ELEMENT FAILURE 5-FEF 6. - MODES: ALL This IC is included as an Unusual Event because it is considered to be a potential degradation in the level of safety of the plant and a potential precursor of more serious problems. This IC addresses coolant samples exceeding coolant technical specifications for iodine spike. Escalation of this IC to the Alert level is via the Fission Product Barrier Degradation Monitoring ICs.

An increase of greater than 63 uCi/gm on the RCS Sample with the HI HI alarm on the letdown monitor (SJL016) is to ensure the alarm is valid.

5-FEF 7. - MODES: ALL This IC addresses RCS Technical Specification Activity Limits being exceeded as an indication of a precursor to possibly worse activity levels. Note that on some Reactor Trips towards end-of-life with activity levels elevated but less than Technical Specification limits spikes can occur above these limits. The intent is for the levels to remain above these limits for extended time intervals in order to classify an NUE. DEI = Dose Equivalent Iodine; DEX = Dose Equivalent Xenon.

APF 06-002-01, Rev. 17A [Commitment Step 3.2.2] Page 19 of 37 EMERGENCY ACTION LEVELS START EAL-6, LOSS OF ELECTRICAL POWER/ASSESSMENT CAPABILITY 6-LEP/AC1 Mode: ALL

Loss of all offsite power to both 6-LEP/AC2 NB TRANSFORMERS which lasts Mode: ALL 6-LEP/AC3 6-LEP/AC4 or is predicted to last greater Mode: 1-4 Mode: 1-4 than 15 minutes Loss or projected loss of all Any of the following conditions exists: GENERAL Yes OR Is plant in

Loss or projected loss of all AC EMERGENCY Yes AC power to NB01 and NB02 Yes Yes

Loss of both DIESEL GENERATORS Modes 1 - 4? power for greater than 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for greater than 15 minutes which lasts or is predicted to last OR greater than 15 minutes

Core Cooling Orange Path No OR SITE AREA No

Core Cooling Red Path No 6-LEP/AC6 No EMERGENCY Mode: 1-4 o Plant in Modes 1-4 6-LEP/AC5 AND Mode: 1-4 o All NK busses less than or equal to Yes Only one NB bus energized Yes ALERT 105 vdc as indicated on RL016 for greater than 15 minutes No 6-LEP/AC7 No NOTIFICATION OF Mode: 5-6 UNUSUAL EVENT o Plant in Modes 5-6 AND o Unplanned loss of required NK bus SITE AREA voltages (NK01/3 or NK02/4) to less than or equal to 105 vdc as indicated EMERGENCY on RL016 for greater than 15 minutes Yes 6-LEP/AC8 No Mode: 1-4 6-LEP/AC9

Unplanned loss of greater than or equal to 75% of MCB Mode: 1-4 annunciators or indication for greater than 15 minutes 6-LEP/AC10 OR A major transient is in progress Mode: 1-4

Unplanned loss of PK01 or PK02 for greater than 15 causing a major RCS Yes Yes NPIS is inoperable Yes temperature and/or pressure minutes change OR

Shift Manager has determined that additional personnel are required to monitor plant systems No No 6-LEP/AC11 NOTE: If a major transient is in progress, the 15 minute time Mode: 1-4 limits are not applicable. NPIS is inoperable Yes ALERT 6-LEP/AC12 No No Mode: ALL

Unplanned complete loss of all onsite communication NOTIFICATION OF Yes capability UNUSUAL EVENT OR

Unplanned complete loss of all offsite communication capability NO ACTION THIS No NOTE: Refer to bases for definitions CATEGORY

APF 06-002-01, Rev. 17A [Commitment Step 3.2.2] Page 20 of 37 EMERGENCY ACTION LEVELS BASES-6, LOSS OF ELECTRICAL POWER/ASSESSMENT CAPABILITY 6-LEP/AC 1 - MODES: ALL Prolonged partial loss of AC power reduces required redundancy and potentially reduces the level of safety by rendering the plant more vulnerable to a complete Loss of AC Power (Station Blackout). The first part of this block asks has a loss of all off-site power to both NB transformers occurred? If power is available to any NB transformer from any off site source and that power can be supplied to either NB bus within 15 minutes, then the answer is NO. The second part of the block asks has a loss of both DGs occurred? If DG is available and can energize either bus the answer is NO. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.

6-LEP/AC 2 - MODES: ALL Loss of all AC power compromises all plant safety systems requiring electric power, including RHR, ECCS, Containment Heat Removal, and the Ultimate Heat Sink. It also makes RCS leakage more likely due to failure of RCP seals. Prolonged loss of all AC power will cause core uncovering and loss of containment integrity, thus this event can escalate to a General Emergency. The 15 minute time duration was selected to exclude transient or momentary power loss.

6-LEP/AC 3 - MODES: 1 THROUGH 4 When in cold shutdown, refueling, or defueled mode the event can be classified as an Alert, because of the significantly reduced decay heat, lower temperature and pressure, increasing the time to restore one of the emergency busses, relative to that specified for the Site Area Emergency EAL. The RCP seals are not challenged as much due to lower RCS pressure and temperature. Escalating to Site Area Emergency, if appropriate, is by Radioactive Effluent Release, or Administrative ICs. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.

6-LEP/AC 4 - MODES: 1 THROUGH 4 The 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to restore AC power was based on the site blackout coping analysis performed in conformance with 10 CFR 50.63 and Reg. Guide 1.155, "Station Blackout". Although this IC may be viewed as redundant to the Fission Barrier Degradation IC, its inclusion is necessary to better assure timely recognition and emergency response.

This IC is specified to assure that in the unlikely event of a prolonged station blackout, timely recognition of the seriousness of the event occurs and that declaration of a General Emergency occurs as early as possible, based on a reasonable assessment of the event progression.

The likelihood of restoring at least one emergency bus should be based on a realistic appraisal of the situation since a delay in an upgrade decision based on only a chance of mitigating the event could result in a loss of valuable time in preparing and implementing public protective actions.

In addition, under these conditions, fission product barrier monitoring capability may be degraded. Although it may be difficult to predict when power can be restored, it is necessary to give a reasonable idea of how quickly to declare a General Emergency based on two major considerations:

1. Are there any present indications that core cooling is already degraded to the point that Loss or Potential Loss of Fission Product Barriers is IMMINENT? (CSFST shows Red or Orange path on Core Cooling)

2. If there are no present indications of such core cooling degradation, how likely is it that power can be restored in time to assure that a loss of two barriers with a potential loss of the third barrier can be prevented?

Thus, indication of continuing core cooling degradation must be based on Fission Product Barrier monitoring with particular emphasis on Emergency Manager judgment as it relates to IMMINENT Loss or Potential Loss of fission product barriers and degraded ability to monitor fission product barriers.

6-LEP/AC 5 - MODES: 1 THROUGH 4 This IC and the associated EAL are intended to provide an escalation from LEP/AC 1. The condition indicated by this IC is the degradation of the offsite and onsite power systems such that any additional single failure would result in a station blackout. This condition could occur due to a loss of offsite power with a concurrent failure of one emergency generator to supply power to its emergency busses. The subsequent loss of this single power source would escalate the event to a Site Area Emergency in accordance with 6-LEP/AC 2.

APF 06-002-01, Rev. 17A [Commitment Step 3.2.2] Page 21 of 37 EMERGENCY ACTION LEVELS BASES-6, LOSS OF ELECTRICAL POWER/ASSESSMENT CAPABILITY 6-LEP/AC 6 - MODES: 1 THROUGH 4 Loss of all Class IE DC power compromises ability to monitor and control plant safety functions. Prolonged loss of all DC power will cause core uncovering and possible loss of containment integrity when there is significant decay heat and sensible heat in the reactor system. Escalation to a General Emergency would occur due to Abnormal Rad Levels/ Radiological Effluent, Fission Product Barrier Degradation, or Administrative ICs. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.

105 VDC is based on the minimum bus voltage necessary for the operation of safety related equipment. This voltage value incorporates a margin of at least 15 minutes of operation before the onset of inability to operate those loads. This voltage is usually near the minimum voltage selected when battery sizing is performed. Typically the minimum value for the entire battery set is approximately 105 VDC. For a 60 cell string of batteries the cell voltage is 1.75 Volts per cell. Reading in the Control Room should be verified valid before declaration. Valid means that a reading has been confirmed by the operators to be correct.

6-LEP/AC 7 -MODES: 5 and 6 The purpose of this IC and its associated EALs is to recognize a loss of DC power compromising the ability to monitor and control the removal of decay heat during Cold Shutdown or Refueling operations. This EAL is intended to be anticipatory in as much as the operating crew may not have necessary indication and control of equipment needed to respond to the loss.

Unplanned is included in this IC and EAL to preclude the declaration of an emergency as a result of planned maintenance activities. Routinely plants will perform maintenance on a Train related basis during shutdown periods. It is intended that the loss of the operating (operable) train only is to be considered.

105 VDC is based on the minimum bus voltage necessary for the operation of safety related equipment. This voltage value incorporates a margin of at least 15 minutes of operation before the onset of inability to operate those loads. This voltage is usually near the minimum voltage selected when battery sizing is performed. Typically the minimum value for the entire battery set is approximately 105 VDC. For a 60 cell string of batteries the cell voltage is 1.75 Volts per cell. Reading in the Control Room should be verified valid before declaration. Valid means that a reading has been confirmed by the operators to be correct.

6-LEP/AC 8 - MODES: 1 THROUGH 4 This IC and its associated EAL are intended to recognize the difficulty associated with monitoring changing plant conditions without the use of a major portion of the annunciation or indication equipment.

Quantification is arbitrary, however, it is estimated that if approximately 75% of the Main Control Board 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 annunciation lost but use the value as a judgment threshold for determining the severity of the plant conditions. This judgment is supported by the specific opinion of the Shift Manager that additional operating personnel will be required to provide increased monitoring of system operation to safely operate the unit.

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 10 CFR 50.72. If the shutdown is not in compliance with the Technical Specification action, the Unusual Event is based on ADM 1 "Inability to Reach Required Shutdown Within Technical Specification Limits."

Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.

"Unplanned" loss of annunciators or indicator excludes scheduled maintenance and testing activities.

6-LEP/AC 9 - MODES: 1 THROUGH 4 "Major Transient" includes response to automatic or manually initiated functions such as scrams, runbacks involving greater than 25% thermal power change, ECCS injections, or thermal power oscillations of 10% or greater. Monitoring of major transients is increasingly more difficult with large numbers of annunicators/indicators out of service.

6-LEP/AC 10 -MODES: 1 THROUGH 4 This IC and its associated EAL are intended to recognize the inability of the control room staff to monitor the plant response to a transient. A Site Area Emergency is considered to exist if the Control Room staff cannot monitor safety functions needed for protection of the public.

Indications needed to monitor safety functions necessary for protection of the public must include Control Room indications, computer generated indications and dedicated annunciation capability. The specific indications should be those used to determine such functions as the ability to shut down the reactor, maintain the core cooled and in a coolable geometry, to remove heat from the core, to maintain the reactor coolant system intact, and to maintain Containment intact.

APF 06-002-01, Rev. 17A [Commitment Step 3.2.2] Page 22 of 37 EMERGENCY ACTION LEVELS BASES-6, LOSS OF ELECTRICAL POWER/ASSESSMENT CAPABILITY 6-LEP/AC 11 -MODES:

See 6-LEP/AC10. Since a major transient is not in progress, the inoperability of NPIS merits an ALERT classification.

6-LEP/AC 12 - MODES: ALL Communications - The purpose of this IC and its associated EALs is to recognize a complete loss of communications capability that either defeats the plant operations staff ability to perform routine tasks necessary for plant operations or the ability to communicate problems with offsite authorities. The loss of offsite communications ability is expected to be significantly more comprehensive than the condition addressed by 10 CFR 50.72.

The onsite communications loss includes all of the following:

l. Complete failure of the plant telephone system.

2. Complete failure of the Gaitronics system.

3. Complete failure of the plant radio system.

The offsite communications loss includes all of the following:

1. Complete failure of the ENS line.

2. Complete failure of offsite telephone service (inability to receive or call a location offsite).

3. Complete failure of onsite fax machines.

This EAL is intended to be used only when extraordinary means are being utilized to make communications possible ( relaying of information from radio transmissions, individuals being sent to offsite locations, the use of any cell phone, etc.)

APF 06-002-01, Rev. 17A [Commitment Step 3.2.2] Page 23 of 37 EMERGENCY ACTION LEVELS EAL-7, FUEL HANDLING ACCIDENT START 7-FHA1 7-FHA2 Mode: ALL Mode: ALL Radioactivity is released to the Fuel Building as indicated by HI HI alarm on any of the following rad Valid fuel handling accident has monitors:

occurred.

GG RE 27 (Channel 273)

NOTE: A fuel handling accident is valid Yes

GG RE 28 (Channel 283) Yes ALERT

SD RE 38 (SFP area) when any symptom or entry condition for

SD RE 37 (SFP bridge crane area)

OFN KE-018 has been met.

SD RE 35 (New Fuel Storage area)

SD RE 34 (Cask Handle crane area)

No No 7-FHA3 Mode: ALL Radioactivity is released to the Reactor Building as indicated by HI HI alarm on any of the following rad monitors:

GT RE 22 (Channel 223)

GT RE 31 (Channel 313)

GT RE 32 (Channel 323)

Yes

GT RE 33 (Channel 333)

GT RE 59 (CTMT Hi area )

GT RE 60 (CTMT Hi area)

SD RE 41 (Manipulator Crane Monitor)

SD RE 42 (Containment area)

SD RE 40 (Access Hatch area)

No NOTIFICATION OF UNUSUAL EVENT NO ACTION THIS CATEGORY

APF 06-002-01, Rev. 17A [Commitment Step 3.2.2] Page 24 of 37 EMERGENCY ACTION LEVELS BASES-7, FUEL HANDLING ACCIDENT 7-FHA 1. - MODES: ALL This procedure provides the necessary instructions to minimize the release of airborne activity following a fuel handling accident which indicates a potential degradation of the level of safety of the plant. In this case the specific symptoms/entry conditions of OFN KE-018 are used to answer YES or NO.

Valid means that the Fuel Handling accident has been confirmed by the Operators.

7-FHA 2 & 3. - MODES: ALL NUREG-0818, "Emergency Action Levels for Light Water Reactors," forms the basis for these EALs. There is time available to take corrective actions, and there is little potential for substantial fuel damage. In addition, NUREG/CR-4982, "Severe Accident in Spent Fuel Pools in Support of Generic Safety Issue 82," July 1987, indicates that even if corrective actions are not taken, no prompt fatalities are predicted, and that risk of injury is low.

Thus, an Alert Classification for this event is appropriate. Escalation, if appropriate, would occur via Abnormal Rad Level/Radiological Effluent or Emergency Manager judgment ICs.

APF 06-002-01, Rev. 17A [Commitment Step 3.2.2] Page 25 of 37 EMERGENCY ACTION LEVELS EAL-8, SAFETY SYSTEM FAILURE OR MALFUNCTION START 8-SSFM1 Mode: 1-2 8-SSFM2 8-SSFM3 Mode: 1-2 Mode: 1-3 SSPS failed to trip the reactor Reactor trip breakers cannot be opened Yes Yes Any of the following exists:

automatically with Control Room trip switches

Core Cooling Red Path GENERAL

SB HS-1 on RL003 Yes OR EMERGENCY No OR 8-SSFM4

Valid Heat Sink Red Path

SB HS-42 on RL006 No Mode: 1-3 Unable to feed any SG at an adequate feed rate with any of the 8-SSFM5 8-SSFM6 No following systems: Mode: 1-3 Mode: 1-3

Main Feed Water Yes o All SG NR levels are less than 6%

OR

Condensate AND o "Feed and Bleed" initiated OR o Total AFW flow is less than 270 klbm/ per EMG FR-H1

Aux Feed Water hr Yes AND Yes AND o Total ECCS flow is less than No o Secondary heat sink is required per 250 gpm No EMG FR H-1 8-SSFM7 Mode: 1-4 o Failure to bring the reactor subcritical with the control rods No fully inserted Yes AND o Complete loss of all boron injection flow paths 8-SSFM8 No Mode: 1-4 o All SG levels <10% wide range AND o Total loss of ability to use steam dumps SITE AREA AND Yes o Total loss of ability to use SG ARV's EMERGENCY AND o Complete loss of both trains of RHR or CCW or ESW 8-SSFM9 No ALERT Mode: 1-4 8-SSFM12

UHS level <1070 Mode: 5 & 6 8-SSFM11 OR Yes Yes Mode: 5 & 6 Time for onset of core uncovery

UHS >90° F and main lake is lost has been exceeded or is o OFN EJ-015 or OFN BB-031 has been Yes expected to be exceeded 8-SSFM10 No entered No Mode: 1-4 AND Complete loss of function of both trains Yes o RCS temperature exceeds or is NO ACTION THIS of RHR or CCW or ESW (except due expected to exceed 200 °F No to loss of NB01 and NB02 for less than CATEGORY No 15 minutes)

APF 06-002-01, Rev. 17A [Commitment Step 3.2.2] Page 26 of 37 EMERGENCY ACTION LEVELS BASES-8, SAFETY SYSTEM FAILURE OR MALFUNCTION 8-SSFM 1. - MODE: 1 THROUGH 2 This condition indicates failure of the automatic protection system to trip the reactor. This condition is more than a potential degradation of a safety system in that a front line automatic protection system did not function in response to a plant transient and thus the plant safety has been compromised, and design limits of the fuel may have been exceeded. An Alert is indicated because conditions exist that lead to potential loss of fuel clad or RCS. Reactor protection system setpoint being exceeded (rather than limiting safety system setpoint being exceeded) is specified here because failure of the automatic protection system is the issue.

8-SSFM 2.- MODE: 1 THROUGH 2 Automatic and manual trip are not considered successful if action away from the reactor control or turbine control console was required to trip the reactor. A manual trip is any set of actions by the reactor operator(s) at the reactor control or turbine control console which causes control rods to be rapidly inserted into the core and brings the reactor subcritical (e.g., reactor trip switch). Failure of manual trip would escalate the event to a Site Area Emergency. Under these conditions, the reactor is producing more heat than the maximum decay heat load for which the safety systems are designed. A Site Area Emergency is indicated because conditions exist that lead to imminent loss or potential loss of both fuel clad and RCS. Although this IC may be viewed as redundant to the Fission Product Barrier Degradation IC, its inclusion is necessary to better assure timely recognition and emergency response. Escalation of this event to a General Emergency would be via Fission Product Barrier Degradation or Administrative ICs.

8-SSFM 3. - MODES: 1 THROUGH 3 The extreme challenge to the ability to cool the core is intended to mean that the core exit temperatures are at or approaching 1200F or that the reactor vessel water level is below the top of active fuel. For CSFSTs, this EAL equates to a Core Cooling RED condition. Another consideration is the inability to initially remove heat during the early stages of this sequence. If emergency feedwater flow is insufficient to remove the amount of heat required by design from at least one steam generator, an extreme challenge should be considered to exist. This EAL equates to a Heat Sink RED condition on the CSFSTs. Note that the SG heat sink is required if RCS pressure is greater than SG pressure. In the event either of these challenges exist at a time that the reactor has not been brought below the power associated with the safety system design (typically 3 to 5% power) a core melt sequence exists. In this situation, core degradation can occur rapidly. For this reason, the General Emergency declaration is intended to be anticipatory of the fission product barrier matrix declaration to permit maximum offsite intervention time.

8-SSFM 4. - MODES: 1 THROUGH 3 A complete loss of SG heat sink is indicated. Adequate flow means enough flow is provided by any of the three systems to prevent entry onto FR-H1.

8-SSFM 5. - MODES: 1 THROUGH 3 In combination with a loss of all feedwater capability (addressed in box 8-SSFM 4) a complete loss of SG heat sink is indicated. A check to see if the heat sink is required was added because for larger LOCA break sizes, the RCS will depressurize below the intact steam generator pressures. The steam generators no longer function as a heat sink and the core decay heat is removed by the RCS break flow. For this range of LOCA break sizes, the SG heat sink is not required and actions to restore SG heat sink are not necessary. It is not necessary to be in EMG FR-H1 to answer the portion of the "AND" statement YES. For these cases, the operator transfers to EMG E-1, LOSS OF REACTOR OR SECONDARY COOLANT, to address a loss of reactor coolant. CSFST indicators are used as determining factors.

8-SSFM 6. - MODES: 1 THROUGH 3 This indicates a need for ECCS flow based on actual automatic or manual signal. 250 gpm is chosen because it is greater than the flow from one CCP at the Pressurizer PORV Lift setpoint. It is anticipated that conditions leading to this box will require operator initiation of RCS bleed and feed in accordance with EMG FR-H1. Functional Restoration Procedures should raise ECCS flow to the required value for adequate heat removal.

It is not intended that once the bleed and feed is initiated that the LRCB tree be used. While a LOCA has been induced it is part of planned mitigation strategy and is not unplanned or beyond the capability of the operator to control.

8-SSFM 7. - MODES: 1 THROUGH 4 This EAL addresses the loss of reactivity control required for hot shutdown with the reactor at pressure and temperature. Under these conditions, there is an actual major failure of a system intended for protection of the public, thus declaration of a Site Area Emergency is warranted. Escalation to a General Emergency would be via Abnormal Rad Levels/Radiological Effluent, Emergency Manager judgment, or Fission Product Barrier Degradation ICs.

APF 06-002-01, Rev. 17A [Commitment Step 3.2.2] Page 27 of 37 EMERGENCY ACTION LEVELS 8-SSFM 8. - MODES: 1 THROUGH 4 This EAL addresses complete loss of functions required for hot shutdown with the reactor at pressure and temperature. The loss of all of these functions would prevent core cooling which would lead to core boiling and possible core damage. Under these conditions, there is an actual major failure of systems intended for protection of the public, thus declaration of a Site Area Emergency is warranted.

8-SSFM 9. - MODES: 1 THROUGH 4 This EAL addresses the loss of design function of the UHS. With low level there is insufficient cooling or makeup water available to ensure core cooling. With temperature greater than 90F there is insufficient heat sink to ensure core cooling, thus declaration of a Site Area Emergency is warranted.

8-SSFM 10. - MODES: 1 THROUGH 4 This EAL addresses complete loss of functions required to maintain cold shutdown for accidents starting in Modes 1 through 4. The complete loss of function means that the system can not pump water or remove heat. This does not apply if the system is inoperable but can still pump water and remove heat. The inability of RHR, CCW, or ESW to provide cooling for any reason which prevents the pumping of water or removal of heat in both trains of any of the three systems should cause an Alert classification. The exception for the loss of NB01 and NB02 is due to their loss being covered by the loss of electrical power EAL chart which allows for a fifteen minute delay. Steam Generators are still available to remove heat and UHS is still available as a source of water using alternate methods of supplying the water.

8-SSFM11. - MODES: 5 & 6 This EAL addresses complete loss of functions required for core cooling for accidents starting in Modes 5 and 6. Escalation to Site Area Emergency or General Emergency would be via Abnormal Rad Levels/Radiological Effluent, Emergency Manager judgment, or Fission Product Barrier Degradation ICs.

This IC and its associated EAL are based on concerns raised by Generic Letter 88-17, "Loss of Decay Heat Removal." A number of phenomena such as pressurization, vortexing, steam generator U-tube draining, RCS level differences when operating at a mid-loop condition, decay heat removal system design, and level instrumentation problems can lead to conditions where decay heat removal is lost and core uncovery can occur. NRC analyses show that sequences exist which can cause core uncovery in 15 to 20 minutes and severe core damage within an hour after decay heat removal is lost. Under these conditions, RCS integrity is lost and fuel clad integrity is lost or potentially lost, which is consistent with a Site Area Emergency.

"Uncontrolled" means that system temperature increase is not the result of planned actions by the plant staff.

The EAL guidance related to uncontrolled temperature rise is necessary to preserve the anticipatory philosophy of NUREG-0654 for events starting from temperatures much lower than the cold shutdown temperature limit.

8-SSFM12. - MODES: 5&6 Under the conditions specified by this IC, severe core damage can occur and reactor coolant system pressure boundary integrity may not be assured. This IC covers sequences such as prolonged boiling following loss of decay heat removal, thus declaration of a Site Area Emergency is warranted under the conditions specified by the IC.

Escalation to a General Emergency is via radiological effluent IC. Refer to OFN EJ-015 for guidance to determine times for core uncover.

APF 06-002-01, Rev. 17A [Commitment Step 3.2.2] Page 28 of 37 EMERGENCY ACTION LEVELS

APF 06-002-01, Rev. 17A [Commitment Step 3.2.2] Page 29 of 37 EMERGENCY ACTION LEVELS BASES-9, LOSS OF PLANT CONTROL/SECURITY COMPROMISE 9-LPC/SC 1. - MODES: ALL This EAL is based on the WCGS Site Security Plan. Security events which do not represent at least a potential degradation in the level of safety of the plant, are reported under 10 CFR 73.71 or in some cases under 10 CFR 50.72. The plant Protected Area Boundary is typically that part within the security isolation zone and is defined in the WCGS security plan. Bomb devices discovered within the plant Vital Area could result in EAL escalation.

9-LPC/SC 2. - MODES: ALL This class of security events represents an escalated threat to plant safety above that contained in the Unusual Event. For the purposes of this IC, a civil disturbance which penetrates the protected area boundary can be considered a hostile force. Intrusion into a protected area by a hostile force will escalate this event to a Site Area Emergency. [RCMS 05-115]

9-LPC/SC 3. - MODES: ALL This class of security events represents an escalated threat to plant safety above that contained in the Unusual Event. For the purposes of this IC, a civil disturbance which penetrates the protected area boundary can be considered a hostile force. Intrusion into a protected area by a hostile force will escalate this event to a Site Area Emergency. [RCMS 05-115]

9-LPC/SC 4. - MODES: ALL This IC encompasses conditions under which a HOSTILE FORCE has taken physical control of vital areas required to reach and maintain safe shutdown. The loss of control of vital equipment or controls of vital equipment needed to maintain the functions of reactivity control, RCS inventory, secondary heat removal, and spent fuel pool cooling system (if imminent fuel damage is likely) would cause this IC condition to be met requiring a General Emergency to be declared. [RCMS 05-115]

9-LPC/SC 5. - MODES: ALL With the Control Room evacuated, additional support, monitoring and direction through the Technical Support Center and/or the Emergency Operations Facility is necessary. Inability to establish plant control from outside the Control Room will escalate this event to a Site Area Emergency. While OFN RP-013, RP-014, RP-017 are governing procedures when the Control Room is evacuated, they need not be officially entered in order to answer YES.

9-LPC/SC 6. - MODES: ALL Expeditious transfer of safety systems has not occurred but fission product barrier damage may not yet be indicated. WCGS time for transfer based on analysis or assessments as to how quickly control must be reestablished without core uncovering and/or core damage. This time should not exceed 15 minutes. In cold shutdown and refueling modes, operator concern is directed toward maintaining core cooling such as is discussed in Generic Letter 88-17, "Loss of Decay Heat Removal." In power operation, hot standby, and hot shutdown modes, operator concern is primarily directed toward maintaining critical safety functions and thereby assuring fission product barrier integrity. Escalation of this event, if appropriate, would be by Fission Product Barrier Degradation, Abnormal Rad Levels/Radiological Effluent, or Administrative ICs.

APF 06-002-01, Rev. 17A [Commitment Step 3.2.2] Page 30 of 37 EMERGENCY ACTION LEVELS EAL-10, FIRE START 10-FR1 10-FR2 10-FR3 10-FR4 Mode: ALL Mode: ALL Mode: ALL Mode: ALL Fire in any of the following buildings: Is the fire located in or near Fire affecting the operability of Fire inside the protected area Yes

Reactor Building Yes plant equipment? Yes Plant Safety Systems required Yes ALERT

Control Building for the current operating mode

Fuel Building NOTE: This does NOT include

Auxiliary Building such items as wastebasket

Diesel Generator Building fires and other small fires of no No safety consequence. No

Diesel FOST Vault

Turbine Building

Communications Corridor

Hot Machine Shop No

ESW Pumphouse

ESW Electrical Manholes

ESW Valve House

ESW Access Vaults No 10-FR5 Mode: ALL The fire has lasted :

Greater than 15 minutes from time of notification NOTIFICATION OF OR Yes

Greater than 15 minutes UNUSUAL EVENT from verification of fire alarm (OFN KC-016)

No NO ACTION THIS CATEGORY

APF 06-002-01, Rev. 17A [Commitment Step 3.2.2] Page 31 of 37 EMERGENCY ACTION LEVELS BASES-10, FIRE 10-FR 1. - MODES: ALL The purpose of this IC is to address the magnitude and extent of fires that may be potentially significant precursors to damage to safety-related systems which would render them non-functional. Hence, only the protected area is considered. Smoke from a fire is considered a part of this chart.

10-FR 2. - MODES: ALL This IC is necessary to address fires in areas of the plant which could spread to vital areas or other significant areas. Transformers located next to the Turbine Building such as the NB, PB, MA, and MR are to be considered as part of the Turbine Building and classified as such. It is not intended to include warehouses, etc. not immediately adjacent to or connected to plant vital areas. The Radwaste Building has fire barriers at each end of the Radwaste tunnel and thus is not listed in this box.

10-FR 3. - MODES: ALL This IC excludes fires of non-safety-related consequence.

10-FR 4. - MODES: ALL This IC is used to determine extent of damage to safety-related equipment to determine if the ALERT level of clarification is necessary. Fire in safety related equipment or its support systems which cause the safety related equipment to be inoperable for a required mode will cause an Alert to be declared. As long as the affected NB bus is energized, fire affecting offsite power should not cause an escalation to an Alert on this tree. Whether the safety related equipment is operable or not at the time of the fire has no affect in making a determination to declare an Alert, only whether it is required for that mode.

10-FR 5. - MODES: ALL This IC is necessary because fires inside the protected area located near equipment that last greater than 15 minutes can result in taxing the site Fire Brigade and also in the callout of local fire department. This represents a degradation in plant operational status.

APF 06-002-01, Rev. 17A [Commitment Step 3.2.2] Page 32 of 37 EMERGENCY ACTION LEVELS EAL-11, NATURAL PHENOMENA START 11-NP1 Mode: ALL Continuous winds of greater than or equal to 95 MPH Yes ALERT 11-NP3 No Mode: ALL Visible or other in-plant indication of damage to any of the following:

Reactor Building

Control Building 11-NP2

Fuel Building Mode: ALL

Auxiliary Building

Diesel Generator Building Report of tornado striking within the

Diesel FOST Access Vaults Yes Yes protected area

Turbine Building (structural framing integrity)

Communication Corridor (structural framing integrity)

No

ESW Pumphouse

ESW Electrical Manholes

ESW Valve House

ESW Access Vaults No 11-NP4 11-NP5 Mode: ALL Mode: ALL o Earthquake felt in plant Operating Basis Earthquake limits exceeded AND Yes as indicated by annunciator 00-98D, OBE, in Yes o Annunciator 00-98E, SEISMIC alarm RECORDER ON, in alarm 11-NP6 No No Mode: ALL ANY natural phenomena which caused the loss of a safety related train and NOTIFICATION OF Yes has the potential to be a common UNUSUAL EVENT mode failure No NO ACTION THIS CATEGORY

APF 06-002-01, Rev. 17A [Commitment Step 3.2.2] Page 33 of 37 EMERGENCY ACTION LEVELS BASES-11, NATURAL PHENOMENA 11-NP 1. - MODES: ALL NP 1 is based on WCGS USAR Section 3.3.1.1. Wind loads of this magnitude can cause damage to safety functions. Continuous wind is defined as the fastest observed one minute value.

11-NP 2. - MODES: ALL NP 2 is based on the assumption that a tornado striking (touching down) within the protected boundary may have potentially damaged plant structures containing functions or systems required for safe shutdown of the plant. If such damage is confirmed visually or by other in-plant indications, the event may be escalated to Alert.

11-NP 3. - MODES: ALL This EAL specifies structure containing systems and functions required for a safe shutdown of the plant. This list was obtained from WCGS USAR Table 3.3-1. A structure will have framing integrity when its main support features (I-Beams, Floors, Concrete Pedestals) are substantially intact.

11-NP 4. - MODES: ALL NP 4 was developed on WCGS basis. Damage may be caused to some portions of the site, but should not affect ability of safety functions to operate. Method of detection can be based on instrumentation, validated by a reliable source, or operator assessment. As defined in the EPRI-sponsored "Guidelines for Nuclear Plant Response to an Earthquake", dated October 1989, a "felt earthquake" is: An earthquake of sufficient intensity such that: (a) the vibratory ground motion is felt at the nuclear plant site and recognized as an earthquake based on a consensus of Control Room operators on duty at the time, and (b) for plants with operable seismic instrumentation, the seismic switches of the plant are activated. If the seismic instrumentation is inoperable, contact the Oklahoma Geological Office in Tulsa, OK for verification of an earthquake. If there is verification of an earthquake, this box should be answered YES.

11-NP 5. - MODES: ALL NP 5 based on WCGS USAR design basis. Seismic events of this magnitude can cause damage to safety functions. If the seismic instrumentation is inoperable, contact the Oklahoma Geological Office in Tulsa, OK for verification of an earthquake. If there is verification of an earthquake, this box should be answered YES.

11-NP MODES: ALL Natural Phenomena can be any of the following but not limited to excessive wind, temperature extremes, excessive precipitation, icing, fish kill, vegetation problems, or other events of nature.

APF 06-002-01, Rev. 17A [Commitment Step 3.2.2] Page 34 of 37 EMERGENCY ACTION LEVELS EAL-12, OTHER HAZARDS START 12-OH1 12-OH2 Mode: ALL Mode: ALL Uncontrolled entry of toxic, IDLH or flammable gas from onsite or offsite sources into the plant that can affect Toxic or flammable gases within a plant structure that normal operation. Yes will be life threatening to plant personnel Yes ALERT NOTE: Smoke from a fire is considered in the Fire EAL.

No No 12-OH3 12-OH6 Mode: ALL Mode: ALL Vehicle crash into plant structures or systems within protected area Yes NOTE: Refer to bases

Toxic or flammable gases within a plant structure that will affect safe operation of the plant OR No

Visible or other in plant indication of damage within 12-OH4 any of the following:

Mode: ALL

Reactor Building

Control Building

Fuel Building Explosion in the protected area

Auxiliary Building resulting in visible damage to Yes Yes

Diesel Generator Building permanent structures or equipment

Diesel FOST Access Vault

Turbine Building (structural framing integrity)

Communications Corridor (structural framing integrity)

No

ESW Pumphouse

ESW Electrical Manholes NOTIFICATION OF No 12-OH5 UNUSUAL EVENT

ESW Valve House Mode: ALL

ESW Access Vaults Turbine failure resulting in casing penetration or damage to turbine or Yes generator seals No NO ACTION THIS CATEGORY

APF 06-002-01, Rev. 17A [Commitment Step 3.2.2] Page 35 of 37 EMERGENCY ACTION LEVELS BASES-12, OTHER HAZARDS 12-OH 1. - MODES: ALL This IC is based on releases in concentrations within the site boundary coming from onsite or offsite sources that will affect the health of plant personnel or affecting the safe operation of the plant (i.e., tanker truck accident releasing toxic gases, IDLH atmospheres preventing access to vital areas / equipment, etc.). Smoke from an onsite fire is covered in the fire EALs.

12-OH 2. - MODES: ALL This IC is based on gases that have entered a plant structure affecting the safe operation of the plant. This IC applies to buildings and areas contiguous to plant Vital Areas or other significant buildings or areas (i.e., Service Water Pump house). The intent of this IC is not to include buildings (i.e., warehouses) or other areas that are not contiguous or immediately adjacent to plant Vital Areas. It is appropriate that increased monitoring be done to ascertain whether consequential damage has occurred. Escalation to a higher emergency class, if appropriate, will be based on System Malfunction, Fission Product Barrier Degradation, Abnormal Rad Levels/Radioactive Effluent, or Administrative ICs. Life threatening is defined as a IDLH "Immediately Dangerous to Life and Health" or explosive atmosphere.

12-OH 3. - MODES: ALL This EAL is intended to address such items as plane or helicopter crash, or train derailment that may potentially damage plant structures containing functions and systems required for safe shutdown of the plant. If the crash is confirmed to affect a plant vital area, the event may be escalated to Alert. Minor vehicle accidents into plant structure (i.e. pickup backs into the CST) which causes minor damage to the structure are not applicable. Unauthorized vehicle intrusion and any resulting damage within the protected area is covered in Loss of Plant Control/Security Compromise flow charts.

12-OH 4. - MODES: ALL For this EAL only those explosions of sufficient force to damage permanent structures or equipment within the protected area should be considered. As used here, an explosion is a rapid, violent, unconfined combustion, or a catastrophic failure of pressurized equipment, that potentially imparts significant energy to near-by structures and materials. No attempt is made in this EAL to assess the actual magnitude of the damage. The occurrence of the explosion with reports of evidence of damage (e.g., deformation scorching) is sufficient for declaration. All security aspects of the explosion also need to considered, if applicable.

12-OH 5. - MODES: ALL This EAL is intended to address main turbine rotating component failures of sufficient magnitude to cause observable damage to the turbine casing or to the seals of the turbine generator. Of major concern is the potential for leakage of combustible fluids (lubricating oils) and gases (hydrogen cooling) to the plant environs. Actual fires and flammable gas build up are appropriately classified under OH 1 or the Fire IC for Emergency Classification. This EAL is consistent with the definition of an Unusual Event while maintaining the anticipatory nature desired and recognizing the risk to non-safety related equipment. Escalation of the emergency classification is based on potential damage done by missiles generated by the failure, or in conjunction with a steam generator tube rupture. These latter events would be classified by the radiological ICs or Fission Product Barrier ICs.

12-OH 6. - MODES: ALL This EAL specifies structures containing systems and functions required for safe shutdown of the plant. This list was obtained from WCGS USAR Table 3.3-1.

APF 06-002-01, Rev. 17A [Commitment Step 3.2.2] Page 36 of 37 EMERGENCY ACTION LEVELS EAL-13, ADMINISTRATIVE START 13-ADM1 13-ADM2 13-ADM3 13-ADM4 Mode: ALL Mode: ALL Mode: ALL Mode: ALL Unusual events are in process Events are in process or have Events are in process or have Events are in process or have or have occurred which occurred which involve an occurred which involve actual occurred which involve actual or indicate a potential degradation actual or potential substantial or likely major failures of plant imminent substantial core of the level of safety of the degradation of the level of functions needed for protection degradation or melting with potential plant.

safety of the plant. of the public. for loss of containment integrity.

OR GENERAL Yes OR Yes OR Yes OR Yes Releases of radioactive EMERGENCY Releases are expected to Releases are expected to Releases can be reasonably material requiring offsite exceed a small fraction of the result in exposure levels which expected to exceed EPA Protective response or monitoring could EPA Protective Action exceed EPA Protective Action Action Guideline exposure levels be expected if further Guideline exposure levels near Guideline exposure levels near offsite for more than the immediate degradation of safety systems the site boundary. the site boundary. site area.

occurs.

No No No No 13-ADM5 Mode: 1-4 SITE AREA EMERGENCY o Plant shutdown required by Technical Specifications AND o Required operating mode Yes NOT achieved within Technical Specification ALERT action statement time limits No NOTIFICATION OF UNUSUAL EVENT NO ACTION THIS CATEGORY

APF 06-002-01, Rev. 17A [Commitment Step 3.2.2] Page 37 of 37 EMERGENCY ACTION LEVELS BASES-13, ADMINISTRATIVE 13-ADM 1. - MODES: ALL Potential degradation of the level of safety of the plant is indicated primarily by exceeding plant Technical Specification Limiting Condition of Operation (LCO) allowable action statement time for achieving required mode change. Precursors of more serious events should also be included because precursors do represent a potential degradation in the level of safety of the plant.

Minor releases of radioactive materials are included. In this emergency class, however, releases do not require monitoring or offsite response (e.g., dose consequences of less than 10 millirem).

13-ADM 2. - MODES: ALL Potential degradation of the level of safety of the plant is indicated primarily by exceeding plant Technical Specification Limiting Condition of Operation (LCO) allowable action statement time for achieving required mode change. Precursors of more serious events should also be included because precursors do represent a potential degradation in the level of safety of the plant.

A determination that increased monitoring of plant functions is warranted due to safety system degradation requires an Alert. Minor releases of radioactive materials are included. In this emergency class, however, releases do not require monitoring or offsite response (e.g., dose consequences of less than 10 millirem).

13-ADM 3. - MODES: ALL The discriminator (threshold) between Site Area and General Emergency is whether or not the EPA PAG plume exposure levels are expected to be exceeded outside the site boundary. This threshold, in addition to dynamic dose assessment considerations discussed in the EAL guidelines, clearly addresses NRC and offsite emergency response agency concerns as to timely declaration of a General Emergency.

13-ADM 4. - MODES: ALL The bottom line for the General Emergency is whether evacuation or sheltering of the general public is indicated based on EPA PAGs, and therefore should be interpreted to include radionuclide release regardless of cause. In addition, it should address concerns as to uncertainties in systems or structures, such as CTMT, response; and also events such as waste gas tank releases and severe spent fuel pool events postulated to occur at high population density sites. To better assure timely notification, EALs in this category must primarily be expressed in terms of plant function status, with secondary reliance on dose projection. In terms of fission product barriers, loss of two barriers with potential loss of the third barrier constitutes a General Emergency.

13-ADM 5. - MODES: 1 THROUGH 4 Limiting Conditions of Operation (LCOs) require the plant to be brought to a required shutdown mode when the Technical Specification required configuration cannot be restored. Depending on the circumstances, this may or may not be an emergency or precursor to a more severe condition. In any case, the initiation of plant shutdown required by the site Technical Specifications requires a four hour report under 10 CFR 50.72 (b) Non-emergency events. The plant is within its safety envelope when being shut down within the allowable action statement time in the Technical Specifications. An immediate Notification of an Unusual Event is required when the plant is not brought to the required operating mode within the allowable action statement time in the Technical Specifications. Declaration of an Unusual Event is based on the time at which the LCO-specified action statement time period elapses under the site Technical Specifications and is not related to how long a condition may have existed. Other required Technical Specification shutdowns that involve precursors to more serious events are addressed by other System Malfunction, Hazards, or Fission Product Barrier Degradation ICs.

ML17355A650

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