examination answer key · 2019-12-16 · main turbine bypass steam (40% of main steam design flow...

EXAMINATION ANSWER KEY 2017 NRC

BYR OPS ILT EXAM Page: 1 of 256 21 September 2017

1 ID: 2017 NRC Q1 Points: 1.00 Which of the following is the basis for verifying the turbine is tripped at Step 2 in BEP-0, REACTOR TRIP OR SAFETY INJECTION?

A. Prevents Condenser damage.

B. Prevents low Pressurizer pressure.

C. Prevents an uncontrolled RCS cooldown.

D. Ensures turbine overspeed will NOT occur.

Answer: C

Answer Explanation A. Incorrect: This is plausible because the steam dumps, another source of steam to the condenser, are isolated on C-9 which protects the condenser. B. Incorrect: Prevents low Pressurizer pressure is not listed in the background document as a basis for step 2 of BEP-0. Plausible because Pressurizer pressure will drop from both the Reactor trip and the cooldown. C. CORRECT: Prevents an uncontrolled RCS cooldown is correct because E-0 Step 2 WOG ERG BKG document states: The turbine is tripped to prevent an uncontrolled cooldown of the RCS due to steam flow that the turbine would require. D. Incorrect: Ensures a turbine overspeed will NOT occur is not listed in the background document as a basis for step 2 of BEP-0. Plausible because Turbine TV and GV will close as a result of an overspeed. Meets K/A, examinee must understand the basis for closing the main turbine governor valve and the main turbine stop valve after a reactor trip. Technical References: BEP-0 REACTOR TRIP OR SAFETY INJECTION, E-0 REACTOR TRIP OR SAFETY INJECTION WOG ERG BKG document, FR S.1 RESPONSE TO NUCLEAR POWER GENERATION/ATWS WOG ERG BKG document



Question 1 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 4.00

System ID: 1741572 User-Defined ID: 2017 NRC Q1 Cross Reference Number: T.EP01-03

Topic: Basis for Step 2 of BEP-0 RO Importance: 3.7 SRO Importance: 4.0 K/A: 007EK1.03 Comments: Bank question from LORT Exam bank (1596633)

License Level: RO Cognitive Level: Low, from memory EPE: 007 Reactor Trip K1.03 Knowledge of the operational implications of the following concepts as they apply to the reactor trip: Reasons for closing the main turbine governor valve and the main turbine stop valve after a reactor trip Objective: T.EP01-03, Given the procedure, be able to DESCRIBE the Intent/Basis of each step and how it is performed



2 ID: 2017 NRC Q2 Points: 1.00 A Loss of Coolant Accident (LOCA) occurred. • The Reactor TRIPPED. • An automatic Safety Injection Signal occurred. • High Head SI Flow is 500 gpm • Reactor Coolant System pressure is 150 psia. • RWST level is 58% and LOWERING • Pressurizer level is 0% The crew has transitioned to 1BEP-1, LOSS OF REACTOR OR SECONDARY COOLANT and is currently at step 11, Initiate Evaluation of Plant Status. 1. Reactor Coolant Pumps (RCP) will be ___(1)___ .

2. Residual Heat Removal Pumps (RH) will be ___(2)___ .

A. 1. running

2. running

B. 1. running 2. stopped

C. 1. stopped

2. running

D. 1. stopped 2. stopped

Answer: C

Answer Explanation



The question places the examinee in a situation where a Large Break LOCA has occurred. During the execution of BEP-1, Step 1 directs the Operator to review RCP Trip Criteria. In this case, a trip criterion is met based on RCS pressure less than 1425 psig with High Head Injection flow greater than 100 gpm. The RH pumps will continue to run because the transition criteria to 1BEP ES-1.1 SI Termination (step 6 of 1BEP-1) has not been met based on RCS inventory, specifically PZR level. Another "gate" for stopping RH pumps is at step 8, Check if RH pumps should be stopped. That gate is not met because RCS pressure is less than 325 psig. A. Incorrect: The first part is incorrect but plausible because in BFR-S.1 (another Emergency

Operating Procedure) there is a caution to NOT Trip RCPs even if trip criteria is met. The second part is correct as stated above.

B. Incorrect: The first part is incorrect but plausible because in BFR-S.1 (another Emergency

Operating Procedure) there is a caution to NOT Trip RCPs even if trip criteria is met. The second part is incorrect but plausible as there are procedural steps contained in 1BEP-1 for tripping the RH pumps.

C. Correct: See above. D. Incorrect: The first part is correct as stated above. The second part is incorrect but plausible

as there are procedural steps contained in 1BEP-1 for tripping the RH pumps. The question meets the K/A by placing the examinee in a situation where a large break LOCA has occurred. The examinee has to assess the stated conditions to make this determination. In addition, the examinee during the assessment must understand that the (1) RCP which were running, need to be tripped and (2) the RH pumps which were in standby prior to the event would automatically start and run and again by assessment would need to continue to run during the present condition. Reference: 1BEP-1, Operator Action Summary (RCP trip criteria), steps 6 and 8 1BFR-S.1, first caution of procedure, page 2




System ID: 1741710 User-Defined ID: 2017 NRC Q2 Cross Reference Number: 8D.EP-007-B

Topic:

A Loss of Coolant Accident (LOCA) occurred. The Reactor TRIPPED. An automatic Safety Injection

RO Importance: 2.6 SRO Importance: 2.7 K/A: EPE011K2.02 Comments: New question for 2017 NRC exam

RO level High Cog Level due to performing analysis EPE011K2.02, EPE011, Large Break LOCA, EK2 Knowledge of the interrelations between the following Large Break LOCA: EK2.02, Pumps Objective: 8D.EP-007-B, Demonstrate the proper usage of EP-1



3 ID: 2017 NRC Q3A Points: 1.00 Following a tube rupture on the 1B S/G, all RCP's were manually tripped at step 1 of 1BEP-3, STEAM GENERATOR TUBE RUPTURE. AT step 37, the Unit Supervisor directs start of one RCP per 1BOA ESP-1, RCP RE-START. 1A RCP 1B RCP 1C RCP 1D RCP No. 1 seal ΔP 300 PSID 250 PSID 300 PSID 250 PSID

lower radial bearing temperature 220°F 240°F 220°F 240°F

All RCP BRNG CC WTR FLOW LOW alarms are lit. All RCP SEAL WTR INJ Filter DP HIGH alarms are clear and Seal Injection flows are all 9 GPM. Which RCP is preferred to be started FIRST?

A. 1A RCP

B. 1B RCP

C. 1C RCP

D. 1D RCP

Answer: C

Answer Explanation



The Preferred RCP start sequence is D, C, B, A per 1BOA ESP-1 RCP RE-START. Parameters for restart: No.1 seal ΔP GREATER THAN 275 PSID, Seal injection flow BETWEEN 8 GPM and 13 GPM and RCP SEAL WTR INJ FILTER DP HIGH alarm (1-7-A2) is NOT LIT. Both the required parameter values AND the preferred start sequence must be satisfied for pump start; therefore, extreme high or low parameter values bounding other parameter values (subsets) are an insufficient reason alone to throw out any distractor. Also, step 4 RNO requires a different pump be started if lower radial bearing temperature is above 225°F. A. Incorrect because it is not the first RCP to be started per the preferred start sequence. Plausible as all parameters are met for start. B. Incorrect because it is not the first RCP to be started per the preferred start sequence and the No. 1 seal ΔP is NOT met for start. Plausible because the trip setpoint for the RCP is 200 PSID, and pressure is well above that value, this may lead the candidate to conclude pump start is acceptable. Also, lower radial bearing temperature is not one of the initial parameters checked for start of the RCP. C. CORRECT because it is the first available RCP to be started per the preferred start sequence and all parameters are met for start. D. Incorrect because the No. 1 seal ΔP is NOT met for start. Plausible because this is the first RCP to be started per the preferred start sequence. Also plausible because the trip setpoint for the RCP is 200 PSID, and pressure is well above that value, this may lead the candidate to conclude pump start is acceptable. Also, lower radial bearing temperature is not one of the initial parameters checked for start of the RCP. Meets K/A, examinee must evaluate plant conditions for the Reactor Coolant Pumps and select which pump is acceptable and preferred to start per 1BOA ESP-1, RCP RE-START. Technical References: 1BEP-3 STEAM GENERATOR TUBE RUPTURE, 1BOA ESP-1 RCP RE-START.




System ID: 1845772 User-Defined ID: 2017 NRC Q3A Cross Reference Number: T.OA09B-03

Topic: Following a tube rupture on the 1B S/G, all RCP's were manually tripped at step 1 of 1BEP-3, STEAM

RO Importance: 4.3 SRO Importance: 4.4 K/A: 015AG2.1.23 Comments: New Question for 2017 NRC Exam

License Level: RO Cognitive Level: High, Multiple analyses are required to answer question. APE: 015 Reactor Coolant Pump Malfunctions G2.1.23 Ability to perform specific system and integrated plant procedures during all modes of plant operation. Objective: T.OA09B-03, ANALYZE a given set of plant conditions and DETERMINE the required actions per 1/2BOA ESP-1, Reactor Coolant Pump Startup During Abnormal Conditions



4 ID: 2017 NRC Q4 Points: 1.00 1BOA PRI-15, CVCS ABNORMALITIES, step 9, Restore Charging Flow, requires the operator to throttle 1CV121 and 1CV182 to establish 70 GPM total charging flow. When 1CV182, CHG HDR BACK PRESSURE CONTROL VLV, is throttled...

A. open,

seal injection flow rises and charging flow rises.

B. open, seal injection flow lowers and charging flow lowers.

C. closed,

seal injection flow rises and charging flow lowers.

D. closed, seal injection flow lowers and charging flow rises.

Answer: C

Answer Explanation The seal injection line tap is upstream of 1CV182 and the charging line is downstream of 1CV182. Therefore, 1CV182 functions as a backpressure regulator for seal injection and a normal throttle valve for changing flow A. Incorrect: when 1CV182 is opened, seal injection will lower because 1CV182 is a

backpressure regulator for seal injection flow. The distractor is plausible because if 1CV182 functioned as a normal throttle valve for both lines, opening 1CV182 would raise flow in both cases.

B. Incorrect: When 1CV182 is opened, charging flow will rise. The distractor is plausible

because if 1CV182 functioned as a backpressure regulator for both lines, 1CV182 being opened would lower flow in both cases.

C. CORRECT: When 1CV182 is closed seal injection will rise and charging flow will lower D. Incorrect: When 1CV182 is closed seal injection will rise. Plausible because if the charging

line were upstream of CV182, charging flow would rise as the valve is throttled closed, and if the seal injection line were downstream of 1CV182, injection flow would lower as the valve is throttled closed.

The question meets the K/A by testing the examinee's knowledge of adjustment of RCP seal backpressure regulator valve to obtain normal flow. References: 1BOA PRI-15, CVCS ABNORMALITIES P&ID M-64 sheet 3A (1CV121) and 3B (1CV182)




System ID: 1741687 User-Defined ID: 2017 NRC Q4 Cross Reference Number: S.CV1-05-H

Topic: 1BOA PRI-15, CVCS ABNORMALITIES, step 9, Restore Charging Flow, requires the operator to throttle 1

RO Importance: 2.7 SRO Importance: 3.1 K/A: 022AK3.01 Comments: New Question for 2017 NRC Exam

License Level: RO Cognitive Level: Low, from memory APE: 022 Loss of Reactor Coolant Makeup K3.01 Adjustment of RCP seal backpressure regulator valve to obtain normal flow. Objective: S.CV1-05-H, PREDICT how CVCS/plant parameters will respond to manipulation of the following CVCS local/remote controls: CV-182



5 ID: 2017 NRC Q5A Points: 1.00 At 1000 Unit 1 is in MODE 4 • 1A Residual Heat Removal (RHR) train aligned for Shutdown Cooling. • 1B Residual Heat Removal (RHR) train aligned for Cold Leg Injection. At 1005 Pressurizer level starts to lower and the crew enters 1BOA S/D-2, SHUTDOWN LOCA. The Radwaste Equipment Operator reports the on-line Hold Up Tank (HUT) is showing an unexplained level RISE. 1. What is the MOST probable source for the RCS leak?

2. The leak ...

A. 1. 1CV8119, U-1 LTDWN HX OUTLET HDR RLF VLV.

2. will NOT be stopped if letdown is isolated.

B. 1. 1CV8119, U-1 LTDWN HX OUTLET HDR RLF VLV 2. WILL be stopped if letdown is isolated.

C. 1. 1RH8708A, 1A RH PP SUCT HDR RLF VLV.

2. will NOT be stopped if the 1A RH pump suction isolation valves are closed.

D. 1. 1RH8708A, 1A RH PP SUCT HDR RLF VLV. 2. WILL be stopped if the 1A RH pump suction isolation valves are closed.

Answer: D

Answer Explanation



The letdown line relief valves discharge to the PRT (1CV8117).The RH pump suction relief valve relieves to the Hold Up Tanks (HUTs). The RH pumps are normally crosstied, but with one train in shutdown cooling and the other aligned for cold leg injection, the trains are split which results in isolation of the leak when the steps of BOA S/D-2. SHUTDOWN LOCA are taken. Per BOA S/D-2. SHUTDOWN LOCA, page 2 NOTE: "Monitor for possible RCS leakage from the RH suction relief to HUT." The Abnormal Procedure directs the crew to: Step 1 to stop the running RH pump; step 2, to isolate RCS letdown; step 3, raise charging flow; step 6, actuate CNMT Phase A isolation and IN step 10 to isolate the RCS loop suction valve(s) to both trains of Shutdown cooling. A. Incorrect: The first part is plausible because letdown is placed in service when on RH Cooling.

Incorrect because the relief discharges to the PRT and the normal letdown orifice are not used when on RH letdown.

B. Incorrect: The first part is plausible because letdown is placed in service when on RH Cooling.

Incorrect because the relief discharges to the PRT and the normal orifice are not used when on RH letdown.

C. Incorrect: The first part is correct making this distractor plausible. The second part is incorrect

but is plausible if the candidate concludes the relief valve is located upstream of the suction isolation valve as is the case for the SI pumps.

D. Correct: This is correct as the suction valve relieves to the HUT. The second part is correct,

as closing the pump suction isolation valve will isolate the relief valves from the RCS. The question meets the K/A by placing the examinee in a situation where a Loss of RHRS has occurred and then determine which system, when isolated, will cause the leak to stop. Note: The major actions of the procedure dictate that attempts to isolate the leak are taken. References: BOP RH-6 Operation of the RH system in Shutdown Cooling Horse Notes: RH-1, RHR Cooldown, CV-1, CVCS and ECCS-1




System ID: 2029205 User-Defined ID: 2017 NRC Q5A Cross Reference Number: T.OA35-03

Topic: At 1000 Unit 1 is in MODE 4 on Shutdown Cooling via the 1A Residual Heat Removal (RHR) train.

RO Importance: 3.3 SRO Importance: 3.6 K/A: APE025AA2.04 Comments: New Question for 2017 NRC Exam

RO level High Cog level K/AAPE025, Loss of Residual Heat Removal System (RHRS), AA2: Ability to determine and interpret the following as they apply to the Loss of Residual Heat Removal System, 04: Location and isolability of leaks Objective: T.OA35-03, Analyze a given set of plant conditions and Determine the required actions per 1/2BOA S/D-2, Shutdown LOCA



6 ID: 2017 NRC Q6 Points: 1.00 At 1000 The U-0 CC pump is being aligned to replace the 2B CC pump. • 2A CC pump is running. • 2B CC pump is racked out with control power fuses NOT yet installed, and the control switch is in the

PULL TO LOCK position. • U-0 CC pump is racked in to Bus 242 with control power fuses installed and in standby. At 1002 The Red First Out annunciator 2-11-C1, PZR PRESS LOW SI/RX TRIP, is in alarm. At 1004 Catastrophic failure causes CC surge tank level to LOWER to 10%. The U-0 CC pump ONLY is running.

Which of the following describes the reasons for the final CC pump status?

A. The U-0 CC pump STARTED on the SI.

The 2A CC pump TRIPPED on system pressure.

B. The U-0 CC pump STARTED on the SI. The 2A CC pump TRIPPED on low surge tank level.

C. The U-0 CC pump STARTED on system pressure.

The 2A CC pump TRIPPED on system pressure.

D. The U-0 CC pump STARTED on system pressure. The 2A CC pump TRIPPED on low surge tank level.

Answer: B

Answer Explanation



When an SI occurs the standby CC pump aligned to the unit will start, in this case the Unit 0 pump. This is accomplished by having the U-0 CC pump control power fuses installed. The question states that the realignment is in progress and that the control power fuses are not yet installed on the 2B CC pump. These fuses instate the Low Surge Tank Level pump trip feature for the standby pump, in this case U-0. Since the fuses have not yet been installed, the U-0 pump will NOT trip on Low Surge Tank Level. A. Incorrect: The first part is plausible because it is correct. The second part is incorrect but

plausible as the CC system has pressure switches that provide control functions to the CC pumps. These pressure switches are for a standby pump auto start on low system pressure, but other ESF pumps do trip on low suction pressure.

B. Correct: Both pumps will receive a start signal on a Safety Injection Signal. The 2A pump will

trip on Low Surge Tank Level, but the U-0 pump will not trip because the 2B pump control power fuses are not installed.

C. Incorrect: The first part is incorrect but plausible as the CC system has pressure switches that

provide control functions to the CC pumps. These pressure switches are for a standby pump auto start on low system pressure. The second part is incorrect but plausible as other ESF pumps do trip on low suction pressure.

D. Incorrect: The first part is incorrect but plausible because when the 2A pump tripped on low

surge tank level, system pressure will lower which will cause a start signal to the U-0 pump. The second part is plausible because it is correct.

This question meets the K/A as it places the examinee in a situation in which there is a Loss of Component Cooling water (malfunction) requiring knowledge of the reasons as to how the CC system responds during an ESFAS actuation. References: BOP CC-10 6E-0-4030 CC04 ESFAS L-P Annunciators 2-2-A5, CC SURGE TANK LEVEL HIGH LOW 2-11-C1, PZR PRESS LOW SI/RX TRIP




System ID: 1741851 User-Defined ID: 2017 NRC Q6 Cross Reference Number: 8D.OA-016-B

Topic:

At 1000 The U-0 CC pump is being aligned to replace the 2B CC pump. 2A CC pump is running. 2B C

RO Importance: 3.6 SRO Importance: 3.9 K/A: APE026AK3.02 Comments: New Question for 2017 NRC Exam

RO Level High Cog Level due to analysis K/A APE 026: Loss of Component Cooling Water AK3: Knowledge of the reasons for the following responses as they apply to Loss of Component Cooling Water, .02: The automatic actions (alignments) within CCWS resulting from the actuation of the ESFAS. Objective: 8D.OA-016-B, Direct Actions per OA Pri-6



7 ID: 2017 NRC Q7 Points: 1.00 Unit 1 is at 100% power. • Rods are in manual. • 1PK-455A, Master PZR Pressure Controller output, is failed “AS IS.” • 1PK-455A will NOT transfer to Manual control. A 100 MWe load rejection occurs over 10 minutes.

Which of the following will occur FIRST?

A. OPDT trip setpoints will rise.

B. PZR Backup Heaters will energize.

C. 1RY455B, PZR Spray Valve, will throttle open.

D. 1CV121, CV Pump Flow Control Valve, will throttle closed.

Answer: D

Answer Explanation A. Incorrect. OPDT doesn’t use pressure input to calculate the trip setpoint. Plausible misconception that pressure rising will cause the OPDT setpoint to rise because OTDT does use pressure as an input. B. Incorrect. PZR Backup Heaters will energize at 5% deviation in PZR level, but this will occur after 1LK-459 has lowered demand on 1CV121. Plausible because PZR Backup Heaters will energize at 5% deviation in PZR level. C. Incorrect. 1PK-455A, the master PZR Pressure controller, is failed as is so the spray valve would not change position. Plausible because this is what would occur if the master PZR pressure controller was in auto and worked as designed. D. Correct. When a 100 MWe load reject occurs, Tave will go up and PZR level will rise. 1LK-459, the master PZR Level controller, will sense that PZR level is rising and cause 1CV121 to throttle close to lower PZR level. Meets K/A, examinee must understand that as the load rejection is occurring RCS temperature will rise, as a result, PZR level will also rise. The impact of this is PZR level control will throttle closed on 1CV121. References: BOP RY-100, PRESSURIZER OPERATION.




System ID: 1741698 User-Defined ID: 2017 NRC Q7 Cross Reference Number: T.OA11-23

Topic: Unit 1 is at 100% power. Rods are in manual. 1PK-455A, Master PZR Pressure Controller output

RO Importance: 2.8 SRO Importance: 3.1 K/A: 027AK1.02 Comments: Bank Question from BWD 2014 NRC Exam Q 45 (ID:

RE1027-N14-45) <QQ 1138143(1412)><<License Level: RO Cognitive Level: High, Multiple analyses are required to answer question. APE: 027 Pressurizer Pressure Control System (PZR PCS) Malfunction K1.02 Knowledge of the operational implications of the following concepts as they apply to Pressurizer Pressure Control Malfunctions: Expansion of liquids as temperature increases Objective: T.OA11-23, DESCRIBE the actions necessary to stabilize the plant following a Process Instrumentation malfunction.



8 ID: 2017 NRC Q8D Points: 1.00 An ATWS has occurred. • The crew has entered 1BFR-S.1, RESPONSE TO NUCLEAR POWER GENERATION/ATWS. • At step 5, VERIFY CNMT VENTILATION ISOLATION, the Group 6 CNMT Vent Isol monitor light for

1VQ005A, Miniflow Purge Inside Cnmt Exh Vlv, is dark. What is the required action per step 5, VERIFY CNMT VENTILATION ISOLATION, if any?

A. No action is required.

B. Manually initiate Safety Injection.

C. Locally close 1VQ005A.

D. Shutdown running VQ fans and then manually close 1VQ005A.

Answer: D

Answer Explanation A. Incorrect. See below. Plausible because the examinee may assume that 1VQ005A does not have to be closed with 1VQ005B, 1VQ005C and 1VQ003 all closed as this would isolate the flowpath. B. Incorrect. See below. Plausible because manually initiating Safety Injection will initiate a second Containment Ventilation Isolation signal for 1VQ005A. C. Incorrect. See below. Plausible because many of the emergency procedures require a valve to be closed locally if it is not in the required position. D. Correct. Step 5 of BFR S.1, VERIFY CNMT VENTILATION ISOLATION, states: Group 6 CNMT Vent Isol monitor lights - LIT. If not, the RNO states: Perform the following: 1) Stop any running VQ fans. 2) Manually close VQ isol valve(s) as necessary. The question meets the K/A by requiring the examinee to determine the position of a valve from the status light indication for containment ventilation isolation during the performance of BFR S.1. Reference: 1 BFR-S.1, RESPONSE TO NUCLEAR POWER GENERATION/ATWS.




System ID: 2029213 User-Defined ID: 2017 NRC Q8D Cross Reference Number: T.FR01-07

Topic: An ATWS has occurred. The crew has entered 1BFR-S.1, RESPONSE TO NUCLEAR POWER GENERATION/ATWS.

RO Importance: 3.4 SRO Importance: 3.4 K/A: EPE029EA2.05 Comments: New Question

RO Level Low Cog Level K/A: EPE029: Anticipated Transient Without Scram, EA2: Ability to determine or interpret the following as they apply to an ATWS, 05: System component valve position indications T.FR01-07 ANALYZE a given set of conditions and DETERMINE the appropriate operator actions to respond to an ATWS event



9 ID: 2017 NRC Q9 Points: 1.00 <QQ 1138040(1410)><<<QQ 1138040(1410)><<Unit 1 has experienced a SGTR on the 1A SG. • Failures of ALL SI and CV pumps occurred during the event. • The initial RCS cooldown has been completed. • It has been determined that the RCS is experiencing reflux boiling. 1. The RCS has...

2. Which of the following mitigating strategies apply? • 1BEP-3: STEAM GENERATOR TUBE RUPTURE • 1BCA-3.1: SGTR WITH LOSS OF REACTOR COOLANT-SUBCOOLED RECOVERY DESIRED • 1BCA-3.2: SGTR WITH LOSS OF REACTOR COOLANT-SATURATED RECOVERY DESIRED

A. 1. LOST natural circulation. 2. 1BEP-3 ONLY applies

B. 1. NOT lost natural circulation.

2. 1BEP-3 ONLY applies.

C. 1. LOST natural circulation. 2. 1BEP-3 AND 1BCA-3.1 and/or 1BCA-3.2 apply.

D. 1. NOT lost natural circulation.

2. 1BEP-3 AND 1BCA-3.1 and/or 1BCA-3.2 apply.

Answer: C

Answer Explanation



Explanation: Reflux boiling is generally associated with SBLOCA theory; however, it could conceivably occur during a SGTR wherein the RCS loses sufficient inventory via the SGTR or in combination with a LOCA. Reflux boiling occurs when the RCS is saturated and the SG U-tubes are predominantly steam on BOTH hot and cold leg sides of the SG. This is the point at which natural circulation is lost due to the small difference in water density between hot and cold leg sides of the RCS. Per Byron E series procedures (e.g. ES-0.1, REACTOR TRIP RECOVERY, attachment B: "Conditions that support or indicate natural circulation", the first of 5 "assessment steps" verifies RCS subcooling is acceptable. With reflux boiling occurring, the RCS is saturated and by definition there is no subcooling which is required for natural circulation. Following the initial cooldown of the RCS in 1BEP-3, the RCS subcooling is checked. With no subcooling present, the mitigating strategies of 1BCA-3.1 SGTR WITH LOSS OF REACTOR COOLANT -SUBCOOLED RECOVERY DESIRED and potentially 1BCA-3.2 SGTR WITH LOSS OF REACTOR COOLANT-SATURATED RECOVERY DESIRED. A. Incorrect. Natural Circulation has been lost, and the strategies of 1BEP-3 won’t address this. Distractor is plausible as 1BEP-3 is the procedure that is entered for any SGTR. B. Incorrect. Natural Circulation has been lost. Plausible because if natural circulation had NOT been lost, as indicated in part one, then 1BEP-3 would apply. C. Correct. See explanation. D. Incorrect. Natural Circulation has been lost. Plausible because 1BEP-3 does have a kick out at step 3 RNO to 1BCA-3.1. This is prior to the check for natural circulation. Meets K/A, examinee must assess the operational mitigating strategies that apply to a SGTR with reflux boiling occurring. Technical References: Lesson Plan MCDSBLOCA-LP-I1-MI-XL-14, Small Break LOCA Theory, Lesson Plan I1-CA-XL-04 CA-3.1, 3.2, 3.3 Contingency Action Procedures. 1BCA-3.1,SGTR WITH LOSS OF REACTOR COOLANT-SUBCOOLED RECOVERY DESIRED. 1BCA-3.2,SGTR WITH LOSS OF REACTOR COOLANT-SATURATED RECOVERY DESIRED. 1BEP-3, STEAM GENERATOR TUBE RUPTURE, Background Information, ECA-3.1, SGTR WITH LOSS OF REACTOR COOLANT - SUBCOOLED RECOVERY DESIRED



Question 9 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 0 Difficulty: 0.00 System ID: 1741709 User-Defined ID: 2017 NRC Q9 Cross Reference Number: T.CA4-04-A

Topic: Unit 1 has experienced a SGTR on the 1A SG. Failures of ALL SI and CV pumps occurred during the

RO Importance: 3.1 SRO Importance: 3.3 K/A: 038EK1.04 Comments: <QQ 1138040(1412)><<Bank Question from BWD 2013 NRC

Exam Q 47 (ID: RE10038-N02) License Level: RO Cognitive Level: Low, from memory EPE 038 Steam Generator Tube Rupture K1 Knowledge of the operational implications of the following concepts as they apply to the SGTR: K1.04 Reflux boiling RO 3.1 SRO 3.3 Objective: T.CA4-04-A, Given a set of plant conditions, DIAGNOSE and ANALYZE a SGTR with Loss of Reactor Coolant - Subcooled Recovery Desired



10 ID: 2017 NRC Q10B Points: 1.00 Unit 1 was at 100% power.

At 1000 • A rupture of the Main Feedwater piping in the Turbine Building has occurred. • The Reactor was manually TRIPPED due to lowering Steam Generator levels. • 1A Aux. Feed pump did NOT automatically start. • 1B train total Aux. Feed Flow is 650 gpm. • All Steam Generator NR levels are 7% and dropping.

At 1002 • Aux. Feed Suction Pressure is 17.1 psia to both Aux Feed pumps.

At 1004, the crew is executing 1BEP-0, REACTOR TRIP OR SAFETY INJECTION, at step 6, Verify AF System. 1. What action, if any, is required to satisfy step 6 criteria, Verify AF System?

2. The suction source to the AF system is the ...

A. 1. None, all criteria are met.

2. CST

B. 1. None, all criteria are met. 2. SX system

C. 1. Start the "A" AF pump.

2. CST

D. 1. Start the "A" AF pump. 2. SX system

Answer: D

Answer Explanation



The question refers to an auto start of the AF system due to a loss of MFW. The S/G levels are <10%, which requires >500 gpm AF flow to satisfy the heat sink safety function. Currently AF flow is listed as 650 gpm. The stem lists that AF suction pressure is 17.1 psia. Per BAR 1-3-E7 the SX supply valves to AF open at 18.1 psia. This is indicative that the SX supply valves have opened, supplying the safety related water supply to the suction of the AF pumps. Step 6a. requires the operator to verify both AF pumps operating and if not to manually start non-running pumps. A. Incorrect: This first part is incorrect as step 6a requires the operator to verify both AF pumps operating and if not to manually start non-running pumps. The first part is plausible because the total AF flow requirement is met (i.e. >500gpm). The second part is incorrect but plausible as annunciator 1-3-A7 (AF pump suction pressure LOW) will be lit on low AF pump suction pressure of 20.1 psia which does NOT cause the AF suction source to shift. B. Incorrect: This first part is incorrect as step 6a. requires the operator to verify both AF pumps operating and if not to manually start non-running pumps. Plausible because the total AF flow requirement is met (i.e. >500gpm). The second part is correct making this part of the distractor plausible. C. Incorrect: The first part is correct making this part of the distractor plausible. The second part is incorrect but plausible as annunciator 1-3-A7 (AF pump suction pressure LOW) will be lit on low AF pump suction pressure of 20.1 psia which does NOT cause the AF suction source to shift. D. Correct: as described above. This question meets the K/A as it tests the examinee on a Loss of Feedwater event in conjunction with what controls are available to maintain S/G level. It further queries the examinee on indications of the auto swap over of the alternate AFW sources, in this case the Essential Service water system. References: 1BEP-0, Reactor Trip or Safety Injection BAR 1-3-A7 AF PP SUCT PRESS LOW BAR 1-3-E7 AF PP SX SUCT VLVS ARMED




System ID: 1845794 User-Defined ID: 2017 NRC Q10B Cross Reference Number: S.AF1-14

Topic:

Unit 1 was at 100% power. At 1000 A rupture of the Main Feedwater piping in the Turbine Building

RO Importance: 4.5 SRO Importance: 4.4 K/A: APE054AA1.01 Comments: New Question

RO Level High Cog Level due to analysis K/A: APE054: Loss of Main Feedwater, AA1: Ability to operate and/or monitor the following as they apply to the Loss of Main Feedwater, 01: AFW controls, including the use of alternate AFW sources Objective: S.AF1-14 DISCUSS Auxiliary Feedwater System operations. Include when the system is used and when it can be secured.



11 ID: 2017 NRC Q11C Points: 1.00 Unit 1 has entered 1BCA-0.0, LOSS OF ALL AC POWER.

Step 5, Try To Restore Power To Any/Both Unit 1 4KV ESF Busses, is in progress.

1. To restore power to the 4KV ESF bus(es), the FIRST activity is to ...

2. The Critical Safety Function Status Trees should be monitored ...

A. 1. manually initiate SI to auto-start Emergency Diesel Generators.

2. and Functional Restoration Guidelines should be implemented, as required.

B. 1. manually initiate SI to auto-start Emergency Diesel Generators. 2. for information only. Functional Restoration Guidelines should NOT be implemented.

C. 1. crosstie to any available Unit 2 4KV ESF bus to load the emergency bus(es) from any

other appropriate ESF bus(es). 2. and Functional Restoration Guidelines should be implemented, as required.

D. 1. crosstie to any available Unit 2 4KV ESF bus to load the emergency bus(es) from any

other appropriate ESF bus(es). 2. for information only. Functional Restoration Guidelines should NOT be implemented.

Answer: B

Answer Explanation



Per BCA 0.0, step 5, if NO ESF bus is energized, the operator will initiate SI from 1PM05J and 06J. If this step is not successful the operator continues in the procedure and in the succeeding steps attempt to perform the crosstie to the other unit. Initiating SI is performed first. The third note prior to step 1 states that the status trees are monitored for information only and that the BFRs should NOT be implemented. The WOG background document states the reason for initiating the SI is to allow the D/G to start by bypassing the normal starting interlocks. It goes on to say one method of doing that is to SI. Also step 5 of the background document states that if the emergency bus cannot be loaded on the D/G, then the operator is to attempt to load the emergency bus on any other appropriate plant specific power supply. A. Incorrect. The first part is correct which makes this part of the distractor plausible. The second

part is incorrect but plausible because when E-0 is exited, monitoring of the status trees is performed and in almost all cases the FRGs will be implemented when required CA-0.0 and ES-1.3 (transfer to cold leg recirc, first 6 steps) are 2 examples of when the status trees are monitored for information only. In all other cases, if the status trees indicate a red or orange path, the FRG are implemented.

B. Correct as describe above. C. Incorrect: The first part is incorrect but is plausible as this step may be performed while

executing CA-0.0 to restore power but is only performed AFTER the SI is initiated. The second part is incorrect but plausible because when E-0 is exited, monitoring of the status trees is performed. In almost all cases the FRGs will be implemented when required. CA-0.0 and ES-1.3 (transfer to cold leg recirc, first 6 steps) are 2 examples of when the status trees are monitored for information only. In all other cases, if the status trees indicate a red or orange path, the FRG are implemented, making this distractor plausible.

D. Incorrect: The first part is incorrect but is plausible as this step may be performed while

executing CA-0.0 to restore power, but only performed AFTER the SI is initiated. The second part is correct making this part of the distractor plausible.

Reference: BCA-0.0 WOG background document for ECA-0.0 Question meets the K/A by examining the candidate on actions contained in BCA-0.0 along with the reason for those actions on a Loss of Offsite and Onsite Power. Question is RO level as it exams the candidate on the major actions contained in the BCA-0.0.




System ID: 2025534 User-Defined ID: 2017 NRC Q11C Cross Reference Number: T.CA1-06

Topic: Unit 1 has entered 1BCA-0.0, LOSS OF ALL AC POWER. Step 5, Try To Restore Power To Any/Both Unit

RO Importance: 4.3 SRO Importance: 4.6 K/A: 055EK3.02 Comments: New

License Level: RO Cognitive Level: Low, from memory EPE 055 Loss of Offsite and Onsite Power (Station Blackout) K3 Knowledge of the reasons for the following responses as the apply to the Station Blackout: EK3.02 Actions contained in EOP for loss of offsite and onsite power Objective: T.CA1-06, EXPLAIN the purpose of each step, note and caution of the CA-0 series procedures



12 ID: 2017 NRC Q12 Points: 1.00 Unit 1 is in MODE 1 at 100% power. • Instrument Inverter 112 is Out of Service for maintenance. • Instrument Bus 112 is energized by the associated Constant Voltage Transformer. Which Technical Specification LCO(s) are entered, if any?

A. NO Tech Spec LCOs apply.

B. 3.8.7, Inverters- Operating ONLY

C. 3.8.9, Distribution Systems- Operating ONLY

D. 3.8.7, Inverters- Operating AND 3.8.9, Distribution Systems- Operating

Answer: B

Answer Explanation With the 112 inverter O.O.S. for maintenance, T.S. 3.8.7 is entered for the loss of inverter. 3.8.9 is not entered even though the inverter is O.O.S. because 112 bus voltage is assumed to be within specification when energized by the Constant Voltage Transformer. A. Incorrect: A plausible misconception that no Tech Spec LCOs apply because Instrument Bus

112 is energized in the stem and 112 bus voltage is assumed to be within specification. B. Correct: per the explanation above: with one inverter inoperable due to maintenance, Tech

Spec 3.8.7 ONLY applies C. Incorrect: Plausible misconception that T.S. 3.8.9 would apply because T.S. 3.8.7 directs

entry into to T.S. 3.8.9 with any Instrument Bus deenergized. D. Incorrect: Plausible distractor because with one inverter inoperable due to maintenance,

Tech Spec 3.8.7 applies. Also plausible because 3.8.7 directs entry into to T.S. 3.8.9 with any Instrument Bus deenergized.

This question meets the K/A as it places the examinee in a situation in which maintenance is being performed on an instrument inverter and asking which LCO the unit will enter. References: T.S. 3.8.7 and 3.8.9 and bases




System ID: 1742412 User-Defined ID: 2017 NRC Q12 Cross Reference Number: S.AP1-15

Topic: Unit 1 is in MODE 1 at 100% power. Instrument Inverter 112 is Out of Service for maintenance.

RO Importance: 3.1 SRO Importance: 4.2 K/A: APE057G2.2.36 Comments: New Question

RO Level Low Cog Level K/A: APE057; Loss of Vital AC Electrical Instrument Bus, G2.2.; Generic Equipment Control; 36 Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations. Objective: S.AP1-15: Given a set of plant conditions, DETERMINE applicable A.C. Electrical Distribution System Tech Spec/TRM operability requirements. (S.AP1-15)



13 ID: 2017 NRC Q13A Points: 1.00 Unit 1 is at 100% power. • Annunciator 1-21-E8: 125V DC BATT CHGR 111 TROUBLE is in alarm • Annunciator 1-21-E7: 125V DC BATT CHGR 111 FD BRKR TRIP is in alarm. • Annunciator 1-21-E10: 125V DC PNL 111/113 VOLT LOW is in alarm. • Indicator 1EI-DC001: Unit 1 DC BUS 111 VOLTAGE is 122 volts. • Indicator 2EI-DC001: Unit 2 DC BUS 211 VOLTAGE is 130 volts. EO reports that DC Bus 113 voltage is 122 volts. The Unit 1 reactor ___ (1) ___ be TRIPPED.

DC Bus 111 ___ (2) ___ be crosstied with DC Bus 211

A. 1. WILL

2. CAN

B. 1. WILL 2. can NOT

C. 1. will NOT

2. CAN

D. 1. will NOT 2. can NOT

Answer: C

Answer Explanation



BAR 1-21-E10: 125V DC PNL 111/113 VOLT LOW is an entry condition into 1 BOA ELEC-1. With voltage present on DC Bus 111, the reactor is not tripped in accordance with Attachment B. Per BOP DC-7, 125V DC BUS CROSSTIE/RESTORATION, the crosstie can be accomplished if differential voltage between the DC busses does not exceed 20 volts. The charger trouble alarm annunciated due to a loss of AC input voltage was caused by the battery charger feed breaker tripping. A. Incorrect: The first part is incorrect but is plausible as attachment A of the procedure directs that the reactor be tripped on a loss of both DC bus 111 and 113. The second part is correct as differential bus voltage is within the 20 volt limit. B. Incorrect: The first part is incorrect but is plausible as attachment A of the procedure directs that the reactor be tripped on a loss of both DC bus 111 and 113. The second part is incorrect but plausible as the busses cannot be crosstied if greater than a 20 volt differential is present or one bus has greater than a 100 volt ground. C. Correct: Per 1 BOA ELEC-1, with voltage present on DC Bus 111, the reactor is not tripped in accordance with Attachment B. Per BOP DC-7, 125V DC BUS CROSSTIE/RESTORATION, the crosstie can be accomplished if differential voltage between the DC busses does not exceed 20 volts. D. Incorrect: The first part is correct because with voltage present on DC Bus 111, the reactor is not tripped in accordance with 1 BOA ELEC-1. The second part is incorrect but plausible as the busses cannot be crosstied if greater than a 20 volt differential is present or one bus has greater than a 100 volt ground. The question meets the K/A by placing the examinee in a situation of loss of DC power; specifically, in this case, a loss of AC input to the charger. In this condition the DC busses between units must crosstie to maintain DC voltage available to the "A" Train of Unit 1. The examinee must assess the annunciators and DC voltage indications to confirm that a crosstie with the other unit is possible. References: 1 BOA ELEC-1, LOSS OF DC BUS UNIT 1 BOP DC-7, 125V DC BUS CROSSTIE/RESTORATION BAR 1-21-E8: 125V DC BATT CHGR 111 TROUBLE has LIT. BAR 1-21-E7: 125V DC BATT CHGR 111 FD BRKR TRIP has LIT. BAR 1-21-E10: 125V DC PNL 111/113 VOLT LOW




System ID: 1799135 User-Defined ID: 2017 NRC Q13A Cross Reference Number: S.DC1-05-D

Topic:

Unit 1 is at 100% power. Annunciator 1-21-E8: 125V DC BATT CHGR 111 TROUBLE is in alarm

RO Importance: 3.4 SRO Importance: 3.5 K/A: APE058AA1.01 Comments: This is the original version of question 13 w/o pictures

New question RO Level High Cog Level due to analysis K/A: APE058; Loss of DC Power, AA1, Ability to operate and/or monitor the following as they apply to the Loss of DC Power, .01; Cross-tie of the affected dc bus with the alternate supply. Objective: S.DC1-05, Discuss 125VDC system operation during the following modes of operation, including normal voltages; D: Battery Charger failure



14 ID: 2017 NRC Q14 Points: 1.00 On Unit 1, Safety Injection occurred. • Both DGs are running. • 1B SX pump failed to start. • 1PI-SX007 and 1PI-SX008, SX pump discharge pressures, have dropped to 92 PSIG. The 1B DG ...

A. must be manually shutdown due to low SX system flow

B. will automatically trip due to high Jacket Water temperature

C. is unaffected due to the SX system being crosstied between Trains

D. is unaffected due to the SX system being crosstied between Units

Answer: C

Answer Explanation On an SI, both DGs will start and SX flow will be initiated to both DG’s. With only one SX pump running, SX header pressure will drop about 10 PSIG. At 92 PSIG, the SX system is still above the low pressure alarm. The SX system is crosstied between trains on both units and only 1 pump is needed to supply the unit loads. A. Incorrect. See above. Plausible because low SX flow is an alarm on the local DG panel. Also, the 1B SX pump is not running and B train SX supples cooling to the 1B DG. B. Incorrect. See above. Plausible because SX flow is used to cool the Jacket Water on the DG and the DG will trip on high Jacket Water temperature. C. Correct. See above. D. Incorrect. See above. Plausible because the SX system is crosstied between the units on the return header. Meets K/A, Requires examinee to evaluate the SX alignment from the given indications and determine the effect, if any, on the components cooled by SX . Technical References: BAR 1-2-A2, SX PUMP DSCH HDR PRESS LOW. BAR 1PL08J-1-C2, ESS SERVICE WTR FLOW LOW




System ID: 1742181 User-Defined ID: 2017 NRC Q14 Cross Reference Number: S.SX1-15

Topic:

On Unit 1, Safety Injection occurred. Both DGs are running. 1B SX pump failed to start. 1PI-SX0

RO Importance: 2.9 SRO Importance: 3.0 K/A: 062AA1.07 Comments: New Question

License Level: RO Cognitive Level: High, multiple analysis required. APE: 062 Loss of Nuclear Service Water AA1. Ability to operate and / or monitor the following as they apply to the Loss of Nuclear Service Water (SWS): AA1.07 Flow rates to the components and systems that are serviced by the SWS; interactions among the components Objective: S.SX1-15, PREDICT the effect on plant systems of an SX component or system failure



15 ID: 2017 NRC Q15 Points: 1.00 Unit 1 is at 100% • EO reports that the Instrument Air piping to 1RH606, HX 1A FLOW CONT VLV, has ruptured. • 1RH606 remains open. Which of the following Tech Spec LCOs are entered, if any?

A. NO Tech Spec LCO applies

B. LCO 3.5.2, ECCS - Operating ONLY

C. LCO 3.6.3, Containment Isolation Valves ONLY

D. LCO 3.5.2, ECCS - Operating AND LCO 3.6.3, Containment Isolation Valves

Answer: A

Answer Explanation A. Correct. 1RH606 fails open on loss of IA. The 1A RH train is operable because the system can still perform its designed safety function. B. Incorrect. 1RH606 fails open on loss of IA. The 1A RH train is operable because the system can still perform its designed safety function. Plausible because LCO 3.5.2, Two ECCS trains shall be OPERABLE requires both RH trains be operable, so if the 1A RH train was determined to be inoperable, LCO 3.5.2 would apply. C. Incorrect. LCO 3.6.3, Each containment isolation valve shall be OPERABLE does not apply because 1RH606 is not a containment isolation valve. Plausible because the RH system does penetrate containment. D. Incorrect but plausible. See B and C above. Meets K/A, Requires examinee to determine operability and/or availability of safety related equipment. This is RO level, From the ES-401 Attachment 2 Figure 1: Screening for SRO-only linked to 10 CFR 55.43(b)(2) (Tech Specs): • Can NOT be answered by knowing less than 1 hour Tech Specs. • Can NOT be answered by knowing information listed "above-the-line". YES it can be

answered by above the line information. • Can NOT be answered by knowing the TS Safety Limits. • Does involve one or more of the following for TS, TRM or ODCM: Technical References: LCO 3.5.2 and 3.6.3.




System ID: 1742210 User-Defined ID: 2017 NRC Q15 Cross Reference Number: S.RH1-13

Topic: Unit 1 is at 100% EO reports that the Instrument Air piping to 1RH606, HX 1A FLOW CONT VLV, has

RO Importance: 3.6 SRO Importance: 4.6 K/A: 065G2.2.37 Comments: New Question

License Level: RO Cognitive Level: Low, from memory. APE: 065 Loss of Instrument Air 2.2.37 Ability to determine operability and/or availability of safety related equipment. Objective: S.RH1-13, ANALYZE a given set of plant conditions and DETERMINE RH System Tech Spec/TRM operability requirements.



16 ID: 2017 NRC Q16B Points: 1.00 Unit 1 is operating at approximately 35% power when a grid transient occurs. • The reactor tripped on low frequency of 56.5Hz 1. The Main Generator FIRST tripped on ...

2. The breakers that OPENED on the Main Generator Trip were ...

A. 1. reverse power 2. Bus Tie breaker 3-4 and 4-5 ONLY.

B. 1. reverse power

2. Bus Tie breaker 3-4, 4-5 and the PMG output breaker.

C. 1. anti-motoring 2. Bus Tie breaker 3-4 and 4-5 ONLY.

D. 1. anti-motoring

2. Bus Tie breaker 3-4, 4-5 and the PMG output breaker.

Answer: B

Answer Explanation



The transient described in the question stem caused the reactor to trip on 2 of 4 RCP Bus low frequency of 57 Hz. On a reactor trip the turbine receives a trip signal. When the turbine trips the generator will undergo a reverse power situation and current will flow into the generator. 2 relays sense this condition; they are, (1) - reverse power and (2) - anti-motoring. The reverse power condition results in actuation of the reverse power relay after a 30 second time delay which then energizes the 86G1A and 86G1B relays which cause the main generator output breakers to trip along with the PMG output breaker. If the reverse power relay malfunctioned the anti-motoring relay would energize the 86G1A and B relays after a 60 second time delay. The reverse power will actuate first. A. Incorrect. The first part is correct which makes this part of the distractor plausible. The second part is incorrect but plausible as the generator output breakers trip, but they are not the only main generator breakers that open. The Permanent Magnet Generator output breaker opens also. B. Correct as described above. C. Incorrect. The first part is incorrect but plausible as this trip relay will trip the main generator, but it has a 60 second time delay versus the 30 second time delay with the reverse power relay. The second part is incorrect but plausible as the generator output breakers trip, but they are not the only main generator breakers that open. The Permanent Magnet Generator output breaker opens also. D. Incorrect. The first part is incorrect but plausible as this trip relay will trip the main generator, but it has a 60 second time delay versus the 30 second time delay with the reverse power relay. The second part is correct making this part of the distractor plausible. Meets the K/A, the examinee must understand what trips the Main Generator first during the grid transient. Then the examinee must know which breakers will open from that trip. Technical References: BAR 0-35-B2, GCB 3-4 TRIP. BAR 0-35-B1, OCB 4-5 TRIP BAR 1-18-E1, REACTOR TRIP TURBINE TRIP BAR 1-11-B5, RCP BUS UNDER FREQ. REACTOR TRIP 6E-1-4030MP02, 03, 04 and 05.




System ID: 2027098 User-Defined ID: 2017 NRC Q16B Cross Reference Number: S.MP1-06

Topic: Unit 1 is operating at approximately 35% power when a grid transient occurs. The reactor tripped

RO Importance: 3.1 SRO Importance: 3.3 K/A: APE077AK2.02 Comments: New Question

License Level: RO Cognitive Level: High, requires multiple analysis. APE: 077 Generator Voltage and Electric Grid Disturbances AK2. Knowledge of the interrelations between Generator Voltage and Electric Grid Disturbances and the following: AK2.02 Breakers, relays S.MP1-06, STATE all the main generator trips including their effect on overall plant operation



17 ID: 2017 NRC Q17 Points: 1.00 An SI occurred on Unit 1. • The crew entered 1BEP-1 "LOSS OF REACTOR OR SECONDARY COOLANT". The crew has transitioned to 1BEP ES-1.3, TRANSFER TO COLD LEG RECIRCULATION. • 1A RH pump is tripped (Overcurrent). • 1B RH pump is running. • 1SI8811A, Containment Sump Isolation valve, is CLOSED and energized. • 1SI8811B, Containment Sump Isolation valve, is CLOSED, and CANNOT be energized. Which procedure will be used to mitigate the event?

A. 1BEP ES-1.1, SI TERMINATION.

B. 1BEP-1, LOSS OF REACTOR OR SECONDARY COOLANT.

C. 1BCA-1.1, LOSS OF EMERGENCY COOLANT RECIRCULATION.

D. 1BEP ES-1.2, POST LOCA COOLDOWN & DEPRESSURIZATION.

Answer: C

Answer Explanation



A. Incorrect. 1BEP ES-1.3 will direct transition to 1BCA-1.1, LOSS OF EMERGENCY COOLANT RECIRCULATION and the actions of 1BCA-1.1 will mitigate this event. Plausible because 1BEP ES-1.3 does direct transition to 1BEP-1 which would then direct transition to 1BEP ES-1.1. B. Incorrect. 1BEP ES-1.3 will direct transition to 1BCA-1.1, LOSS OF EMERGENCY COOLANT RECIRCULATION, and the actions of 1BCA-1.1 will mitigate this event. Plausible because 1BEP ES-1.3 does direct transition to 1BEP-1. C. Correct. 1BEP ES-1.3 will direct transition to 1BCA-1.1, LOSS OF EMERGENCY COOLANT RECIRCULATION and the actions of 1BCA-1.1 will mitigate this event. D. Incorrect. 1BEP ES-1.3 will direct transition to 1BCA-1.1, LOSS OF EMERGENCY COOLANT RECIRCULATION and the actions of 1BCA-1.1 will mitigate this event. Plausible because 1BEP ES-1.3 does direct transition to 1BEP-1 which would then direct transition to 1BEP ES-1.2. Meets K/A, examinee must be able to evaluated the given indications and select 1BCA-1.1, Loss of Emergency Coolant Recirculation, as the procedure that will mitigate the event. Question is RO Level: From the ES-401 Attachment 2 Figure 2: Screening for SRO-only linked to 10CFR55.43(b)(5) (Assessment and selection of procedures): 1) assessing plant conditions and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. Using the flowchart, this question can: • NOT be answered solely by knowing “systems knowledge”, i.e., how the system works, flowpath, logic, component location. • NOT be answered solely by knowing immediate operator actions. • NOT be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs. • NOT be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure. Knowing the basic mitigative strategy of 1BEP ES-1.3 and with the stem setup of no RH pumps available an RO would know that a different procedure would be needed. With the knowledge of the purpose of 1BCA-1.1 the RO would be able to select that as the procedure to use. Technical References: Lesson Plan: I1-CA-XL-02, 1BEP ES-1.3, TRANSFER TO COLD LEG RECIRCULATION.




System ID: 1741931 User-Defined ID: 2017 NRC Q17 Cross Reference Number: T.CA2-01

Topic:

An SI occurred on Unit 1. The crew entered 1BEP-1 "LOSS OF REACTOR OR SECONDARY COOLANT". The

RO Importance: 3.4 SRO Importance: 4.2 K/A: E11EA2.1 Comments: <QQ 1138040(1412)><<NewnnnNModified Bank Question<QQ

1138040(1412)><<NewnnnN BYLC3DCA02B010 (535692) License Level: RO Cognitive Level: High, requires multiple analysis. E11 Loss of Emergency Coolant Recirculation EA2. Ability to determine and interpret the following as they apply to the (Loss of Emergency Coolant Recirculation) EA2.1 Facility conditions and selection of appropriate procedures during abnormal and emergency operations. Objective: T.CA2-01, DISCUSS the purpose of the contingency action procedure, CA-1.1 and CA-1.2.



18 ID: 2017 NRC Q18 Points: 1.00 Unit 1 reactor TRIPPED A loss of secondary heat sink condition exists. • RCS pressure is 2250 psig and stable. • S/G pressures are all 1100 psig and stable. • S/G WR levels are 45% and slowly lowering. • Both CV pumps have tripped and will NOT start from the MCR. • Both AF pumps have tripped and will NOT start from the MCR. Which of the following actions must be performed NEXT?

A. Trip ALL RCP's.

B. Go to 1BEP-1, LOSS OF REACTOR OR SECONDARY COOLANT.

C. Establish AF to at least one steam generator.

D. Establish main FW to at least one steam generator.

Answer: A

Answer Explanation



A is Correct. Per 1BFR-H.1, RESPONSE TO LOSS OF SECONDARY HEAT SINK, step 2, CHECK CENT CHG PUMP STATUS: CENT CHG pumps - AT LEAST ONE. IF NEITHER pump will start, THEN perform the following: 1) Stop all RCPs. 2) GO TO Step 14 which has the operator actuate SI. This is also the required action for the OAS page. This is RO level because of the OAS page requirements. B is incorrect. Per explanation above this is incorrect. Plausible because transition to1BEP-1 is directed in this procedure. C is incorrect. Per explanation above this is incorrect. Plausible because establishing AF to at least one steam generator is an action taken later in this procedure. D is incorrect. Per explanation above this is incorrect. Plausible because establishing main FW to at least one steam generator is an action taken later in this procedure. Meets K/A, examinee must have knowledge of the interrelations between the (Loss of Secondary Heat Sink) and the facility*s 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 is NOT SRO Level: From the ES-401 Attachment 2 Figure 2: Screening for SRO-only linked to 10CFR55.43(b)(5) (Assessment and selection of procedures): 1) assessing plant conditions and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. Using the flowchart, this question can: • NOT be answered solely by knowing “systems knowledge”, i.e., how the system works, flowpath, logic, component location. • NOT be answered solely by knowing immediate operator actions. • NOT be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs. • NOT be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure. This can be answered by knowledge of the required actions for the OAS page of 1BFR H-1. The OAS page actions are required knowledge for an RO. Technical References: 1BFR-H.1, RESPONSE TO LOSS OF SECONDARY HEAT SINK




System ID: 1741941 User-Defined ID: 2017 NRC Q18 Cross Reference Number: T.FR03-02

Topic:

Unit 1 reactor TRIP A loss of secondary heat sink condition exists. RCS pressure is 2250 psig

RO Importance: 3.9 SRO Importance: 4.2 K/A: E05EK2.2 Comments: Bank Q<QQ 1138040(1412)><<NewnnnNuestion

BYLC3DFR08B003 (422972) License Level: RO Cognitive Level: Low E05 Loss of Secondary Heat Sink EK2. Knowledge of the interrelations between the (Loss of Secondary Heat Sink) and the following: EK2.2 Facility*s 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. Objective: T.FR03-02, Without the use of the H-Series Procedure, DESCRIBE the steps required to restore the critical safety function to within specifications.



19 ID: 2017 NRC Q19 Points: 1.00 At 1000 The plant is operating at 100% power. • Control rod H-10 (Control Bank A) dropped to 84 steps. • Delta flux (ΔI) is determined to be out of limits. • PDMS is INOPERABLE 1. Compared to the Delta flux (ΔI) prior to the rod misalignment, Delta flux (ΔI) is currently more ...

2. At 1100, to address Delta flux (ΔI) out of limits, BOA ROD-3, DROPPED OR MISALIGNED ROD, directs the operator to ...

A. 1. positive.

2. pull rod H-10 out to the bank position.

B. 1. positive. 2. reduce thermal power to <50%.

C. 1. negative.

2. pull rod H-10 out to the bank position.

D. 1. negative. 2. reduce thermal power to <50%.

Answer: D

Answer Explanation



Delta flux (ΔI) is defined as the difference in normalized flux signals between the top and bottom halves of a two section excore neutron detector. This is commonly viewed as the power in the upper portion (of the core) minus the power in the lower portion. When the control rod dropped to 80 steps, negative reactivity will be added to the top portion of the core. This will cause proportionally more power to be produced in the lower regions of the core which will make delta flux more negative. BOA ROD-3, DROPPED OR MISALIGNED ROD, directs the operator to either: (1) IF rod motion does NOT cause further misalignment, then rods may be adjusted in MANUAL to restore ΔI to within limits or (2) Reduce thermal power to <50%. Moving rod H-10 out to restore ΔI is not an option per BOA ROD-3, DROPPED OR MISALIGNED ROD. A. Incorrect: See above. Plausible because if the rod had dropped all the way into the core, ΔI

would actually be more positive due to Tave dropping. Also, BOA ROD-3, DROPPED OR MISALIGNED ROD, will recover rod M-4 after the initial steps of stabilizing the plant.

B. Incorrect: See above. Plausible because if the rod had dropped all the way into the core, ΔI

would actually be more positive due to Tave dropping. Also, BOA ROD-3, DROPPED OR MISALIGNED ROD does direct reducing power to <50% in this condition.

C. Incorrect: See above. Plausible because ΔI would actually be more negative and BOA

ROD-3, DROPPED OR MISALIGNED ROD will recover rod M-4 after the initial steps of stabilizing the plant.

D. Correct: See above. The question meets the K/A, the examinee must know the operational implications of the misaligned rod. The question is high cognitive level due to requiring the examinee to evaluate the conditions and determining how Delta flux will move. Then the examinee will need to use the conditions given and his knowledge of the procedure to determine what actions will be taken. Reference: BCB-1 Figure 19 which depicts the "target" AFD with regards to reactor power. BOA ROD-3, DROPPED OR MISALIGNED ROD.




System ID: 1742452 User-Defined ID: 2017 NRC Q19 Cross Reference Number: T-OA34-05

Topic: The plant is operating at 100% power. Control rod H-10 (Control Bank A) dropped to 84 steps.


RO Level High Cog Level APE 003; Dropped Rod AK1; Knowledge of the operational implications of the following concepts as they apply to a Dropped Control Rod 1.21 Delta flux Objective: T-OA34-05, DESCIRBE the actions necessary to stabilize the plant during a dropped or misaligned rod event.



20 ID: 2017 NRC Q20 Points: 1.00 During performance of the MOVEABLE CONTROL ASSEMBLIES QUARTERLY SURVEILLANCE, Rod P4 became misaligned from the other rods in Shutdown Bank A. • Maintenance completed repairs in 5 hours. • Rod P4 indicates 210 steps on DRPI. • All other rods in Shutdown Bank A indicate 228 steps on both the Step Counters and DRPI. 1. Prior to recovery of Rod P4 per BOA ROD-3, Dropped or Misaligned Rod, Rx power must be less than...

AND

2. This power limit ...

A. 1. 75%

2. ensures Rx power level remains <100% during recovery of the rod.

B. 1. 75% 2. prevents localized power problems when realigning the rod.

C. 1. 50%.

2. ensures Rx power level remains <100% during recovery of the rod.

D. 1. 50%. 2. prevents localized power problems when realigning the rod.

Answer: D

Answer Explanation



A. Incorrect. See below. Plausible because the procedure allows reactor power level of up to 100% during a recovery of a rod within 4 hours that is misaligned less than or equal to 12 steps, the rod is misaligned by 6 steps. The initial power reduction in the procedure is power <70%. 75% is plausible that is the limit during recovery of the dropped rod when the rod has been misaligned by greater than 12 steps for less than 4 hours. B. Incorrect. See below. 75% is plausible that is the limit during recovery of the dropped rod when the rod has been misaligned by greater than 12 steps for less than 4 hours. Plausible because Xenon changes over a long period can cause localized power problems when realigning the rod. The initial power reduction in the procedure is power <70%. C. Incorrect. See below. Plausible because the procedure requires reactor power level to be less than 50% prior to recovery of a rod within 4 hours. Also, the procedure allows reactor power level of up to 100% during a recovery of a rod within 4 hours that is misaligned less than or equal to 12 steps, the rod is misaligned by 6 steps. D. Correct. BOA Rod-3 requires that power level be reduced to less than 50% for recovery of a misaligned rod when repairs and restoration cannot be completed within 4 hours. Xenon changes over a long period can cause localized power problems when realigning the rod; local power density is limited when power is restricted to < 50%. Meets K/A, requires examinee to have knowledge of the 50% power limit on rod misalignment for greater than 4 hours and the reasons for that limit. Technical References: BOA ROD-3, Dropped or Misaligned Rod. 11-0A-XL-34/ BOA ROD-3.





Topic: During performance of the MOVEABLE CONTROL ASSEMBLIES QUARTERLY SURVEILLANCE, Rod P4 became

RO Importance: 3.4 SRO Importance: 4.2 K/A: 005AK3.05 Comments: New Question

License Level: RO Cognitive Level: Low. APE: 005 Inoperable/Stuck Control Rod AK3. Knowledge of the reasons for the following responses as they apply to the Inoperable / Stuck Control Rod: AK3.05 Power limits on rod misalignment Objective: T.OA34-03, ANALYZE a given set of plant conditions and DETERMINE the required actions per 1/2BOA ROD-3, Dropped or Misaligned Rod



21 ID: 2017 NRC Q21B Points: 1.00 Which of the following is an entry condition for 1BOA PRI-2, EMERGENCY BORATION?

A. Unexplained rise in RCS temperature in Mode 1.

B. Unexplained drop in boron concentration in Mode 1.

C. Uncontrolled cooldown when the reactor is shutdown.

D. Failure of ONE RCCA to fully insert following a reactor trip or shutdown.

Answer: C

Answer Explanation



Per 1BOA PRI-2, the following conditions require emergency boration: • Failure of more than one RCCA to fully insert following a reactor trip or shutdown. • Uncontrolled cooldown when the reactor is shutdown. • Inadequate shutdown margin in Mode 2 (with Keff < 1.0), 3, 4, 5. • During Mode 6 when boron concentration is less than COLR Limit in the refueling canal,

refueling cavity or any filled portions of the RCS. A. Incorrect. See above. Plausible because an unexplained rise in RCS temperature would require the insertion of negative reactivity. The examinee may view this as a reason for emergency boration. B. Incorrect. See above. Plausible because an unexplained drop in boron concentration would require the insertion of negative reactivity. The examinee may view this as a reason for emergency boration. C. Correct. See above. D. Incorrect. See above. Plausible because failure of more than one RCCA to fully insert following a reactor trip or shutdown is a reason for emergency boration. Question meets the K/A. The examinee must have the ability to interpret and execute procedure for Emergency Boration. RO question: From the ES-401 Attachment 2 Figure 2: Screening for SRO-only linked to 10CFR55.43(b)(5) (Assessment and selection of procedures): 1) assessing plant conditions and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. Using the flowchart, this question can: • NOT be answered solely by knowing “systems knowledge”, i.e., how the system works, flowpath, logic, component location. No • NOT be answered solely by knowing immediate operator actions. No • NOT be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs. Yes, can be answered by knowing the entry conditions for 1BOA PRI-2. • NOT be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure. No Reference: BOA PRI-2




System ID: 1992470 User-Defined ID: 2017 NRC Q21B Cross Reference Number: T.OA13-03

Topic: Which of the following is an entry condition for 1BOA PRI-2, EMERGENCY BORATION?

RO Importance: 4.3 SRO Importance: 4.6 K/A: APE024G2.1.37 Comments: New question

RO Level Low Cog Level K/A: APE 024: Emergency Boration, G2: Generic, .1: Conduct of Operations, .37: Knowledge of procedures, guidelines, or limitations associated with reactivity management. Objective: T.OA13-03, ANALYZE a given set of plant conditions and DETERMINE the required actions per 1/2BOA PRI-2, Emergency Boration



22 ID: 2017 NRC Q22 Points: 1.00 A reactor startup is in progress at 100 cps in the Source Range. The control power fuse for NI Channel N-32 BLOWS. Which of the following actions must be taken NEXT?

A. Verify ALL rod bottom lights LIT.

B. Manually reinsert ALL Control and Shutdown banks.

C. RAISE reactor power greater than P-6, then BLOCK Source Range Hi Flux trip.

D. Place the LEVEL TRIP switch for the affected channel in BYPASS.

Answer: A

Answer Explanation A is correct. Below P-6 on a loss of power to Source Range N-32, the channel will fail high (trip) making the 1 out of 2 coincidence for a reactor trip. B is incorrect. Below P-6 and a loss of power to Source Range N-32, the channel will fail high (trip) making the 1 out of 2 coincidence for a reactor trip. The distractor is plausible as the requirement to manually reinsert ALL Control and Shutdown banks is a procedure step for issues that don’t cause a reactor trip. C is incorrect. Below P-6 and a loss of power to Source Range N-32, the channel will fail high (trip) making the 1 out of 2 coincidence for a reactor trip. The distractor is plausible because placing the LEVEL TRIP switch for the affected channel on 1PM07J in BYPASS is procedurally driven if greater than P-6 power. D is incorrect. Below P-6 and a loss of power to Source Range N-32, the channel will fail high (trip) making the 1 out of 2 coincidence for a reactor trip. The distractor is plausible because when raising reactor power greater than P-6, the actions taken for Source Range Hi Flux trip is similar to the action for a loss of an intermediate range detector which is to take the affected channel level trip to bypass. Meets K/A, examinee must have knowledge of the power supplies to the source range instrumentation and the interrelationship between that and a loss of Source Range Instrumentation. This is RO level because the examinee must have knowledge of the power supplies for the SR instruments along with the effects on the plant for the loss of those powers supplies. Technical References: BAR 1-11-A2 for setpoints and 1BOA INST-1 for distracters




System ID: 1741882 User-Defined ID: 2017 NRC Q22 Cross Reference Number: S.NI1-08-B

Topic: A reactor startup is in progress at 100 cps in the Source Range. The control power fuse for NI

RO Importance: 2.7 SRO Importance: 3.1 K/A: 032AK2.01 Comments: <QQ 1138040(1412)><<Bank Question from 2012 NRC Exam

RO Question #21 License Level: RO Cognitive Level: High 032 Loss of Source Range Nuclear Instrumentation AK2 Knowledge of the interrelations between the Loss of Source Range Nuclear Instrumentation and the following: AK2.01 Power supplies, including proper switch positions Objective: S.NI1-08-B, STATE the effect on the Source Range NIs if a loss of the following occurs: Control Power



23 ID: 2017 NRC Q23 Points: 1.00 A Reactor Startup is in progress and power is at the Point Of Adding Heat. Intermediate Range Nuclear Instrument N-35 FAILS LOW. LCO 3.3.1, Reactor Trip Instrumentation ____ (1) ____

BECAUSE

reactor power is _______ (2) _______ the P-6 interlock.

A. 1. will be entered

2. BELOW

B. 1. will NOT be entered 2. BELOW

C. 1. will be entered

2. ABOVE

D. 1. will NOT be entered 2. ABOVE

Answer: C

Answer Explanation



When the IR NI failed low, in the present condition, BOA Inst. 1 Attachment B directs the Operator to EITHER lower power to < P-6 or raise power to >P-10 within 2 hours. Per Technical Specification 3.3.1, Reactor Trip System Indication, Condition F, with 1 channel of IR NI inoperable the action required is to lower thermal power to < P-6 or raise thermal power to >P-10. The MODE applicability contained in Table 3.3.1-1 mimics the previous statement. A. Incorrect: The first part is correct as entry into the LCO is required, which makes this part of

the distractor plausible. The second part is incorrect but plausible based on a possible misconception of the P-6 setpoint and its relationship to the POAH. Also, knowing the relationship between the P-6 interlock and the Intermediate Nuclear Instrument is essential for determining operability.

B. Incorrect: The first part is incorrect but plausible because the LCO is not applicable < P-6. The

second part is incorrect but plausible based one possible misconception of the P-6 setpoint and its relationship to the POAH. Also, knowing the relationship between the P-6 interlock and the Intermediate Nuclear Instrument is essential for determining operability.

C. Correct: When the IR NI failed low, entry into the LCO is required. In the present condition,

BOA Inst. 1 Attachment B directs the Operator to EITHER lower power to < P-6 or raise power to >P-10 within 2 hours.

D. Incorrect: The first part is incorrect but is plausible because the LCO does not apply >P10.

The second part is incorrect but plausible based a possible misconception of the P-6 setpoint and its relationship to the POAH. Also, for both parts, knowing the relationship between P6 and P10 interlocks and the Intermediate Nuclear Instrument is essential for determining operability.

The question meets the K/A as it requires the examinee to evaluate the indications and determine if the plant is above or below the P-6 and if the LCO applies. The question is High Cog Level due to analysis of conditions at the present power condition. Reference: T.S. 3.3.1 Condition F and Table 3.3.1-1 1BOA Instrument 1, attachment B




System ID: 1742490 User-Defined ID: 2017 NRC Q23 Cross Reference Number: T.OA10.04

Topic: A Reactor Startup is in progress and power is at the Point Of Adding Heat.


RO Level High Cog Level due to analysis K/A: APE033; Loss of Intermediate Range Nuclear Instrumentation, AA2; Ability to determine and interpret the following as they apply to the Loss of Intermediate Range Nuclear Instrumentation, 08; Intermediate range channel operability. ANALIZE a given set of plant conditions and DETERMINE: THE INDICATIONS OF A FAILED NUCLEAR INSTRUMENT. (T.OA10.04)



24 ID: 2017 NRC Q24 Points: 1.00 Reactor power is 25% with the Main Generator Synchronized to the grid. Condenser vacuum starts to LOWER due to condenser air in-leakage. 1. The steam dumps will no longer be available when condenser pressure is ...

2. Disabling steam dump operation in this condition prevents ...

A. 1. > 7 inches HG absolute.

2. loss of Megawatts.

B. 1. > 7 inches HG absolute. 2. condenser damage.

C. 1. > 10 inches HG absolute.

2. loss of Megawatts.

D. 1. > 10 inches HG absolute. 2. condenser damage.

Answer: B

Answer Explanation



From the UFSAR section 10.4.1 Main condenser: "There is also sufficient surface to condense the main turbine bypass steam (40% of main steam design flow at design pressure) following a 50% load rejection from maximum load." Per page 8 of the BOA SEC-3 lesson plan, For a given Rx power, if the turbine doesn’t remove some thermal energy, then the heat load on the condenser will rise. Example: 30% Rx Power = 1075.8 Mw 30% Turbine Load = 375 Mw Condenser Heat Load = 700.8 Mw Therefore the heat load on the condenser is increased by 375 MW on a turbine trip. Per UFSAR section 10.4.1 Main condenser, the heat load on the condenser is limited to prevent condenser damage. Byron Alarm Response procedure 1-BP-5.6 lists the conditions for the CDSR NOT AVAL. C-9 to light, which blocks steam dump arming if condenser conditions are not proper to support dumping steam. The condenser is NOT available under the following conditions: a) 1/1 condenser pressure <23” HgV and b) No main circulating water pump breakers shut. A. Incorrect: The first part is correct as steam dump operation will be ceased by the C-9 interlock

at >7 inches of mercury absolute. The second part is incorrect but plausible because if steam dumps are disabled more steam is available for the turbine which would then prevent the loss of megawatts from the main generator. Loss of megawatts is monitored shiftly by performance of 1BOSR 0.1-1,2,3, MODE 1, 2 & 3 SHIFTLY AND DAILY OPERATING SURVEILLANCE, and during the required hourly board walk downs.

B. Correct: as described above. C. Incorrect: The first part is incorrect but plausible as 10 inch mercury absolute is the turbine trip

on Low Condenser Vacuum. This trip also protects the condenser and the turbine. The second part is incorrect but plausible because if steam dumps are disabled more steam is available for the turbine which would then prevent the loss of megawatts from the main generator. Loss of megawatts is monitored shiftly by performance of 1BOSR 0.1-1,2,3, MODE 1, 2 & 3 SHIFTLY AND DAILY OPERATING SURVEILLANCE, and during the required hourly board walk downs.

D. Incorrect: The first part is incorrect but plausible as 10 inch mercury absolute is the turbine trip

on Low Condenser Vacuum. The second part is correct which makes this distractor plausible. Meets K/A, examinee must evaluate the conditions and determine when steam dumps are not available and recall the reason for this interlock. References: USFAR section 10.4 on condenser and steam dumps BOP MS-100, Main Steam Dump Operation, third prerequisite on P-12 BAR 1-BP-5.6, Condenser not available light on Bypass Permissive Panel Steam Dump system Lesson Plan S-24 1BOA SEC-3, Loss of condenser vacuum 1BOA SEC-3 Lesson plan page 8, excessive heat load on condenser.




System ID: 1742547 User-Defined ID: 2017 NRC Q24 Cross Reference Number: III.D.OA-19-A

Topic: Reactor power is 25% with the Main Generator Synchronized to the grid. Condenser vacuum starts to


RO Level High Cog Level due to converting inches HG vac. to inches HG absolute K/A: 051; Loss of Condenser Vacuum, AK3; Knowledge of the reasons for the following responses as they apply to the Loss of Condenser Vacuum, .01 Loss of steam dump capability upon loss of condenser vacuum. Objective: Given a set of plant conditions or parameters indicating a partial loss of vacuum, DISCUSS the integrated plant response to the event/casualty with no operator action. III.D.OA-19-A



25 ID: 2017 NRC Q25C Points: 1.00 Annunciator 0-37-A4, Unit 1 Area Fire has just alarmed. 1. This alarm is actuated by ...

2. The Plant Wide Fire Alarm is ...

A. 1. the suppression system ONLY.

2. automatically actuated by the Unit 1 Area Fire alarm (Annunciator 0-37-A4).

B. 1. the suppression system ONLY. 2. manually actuated by the NSO.

C. 1. either the suppression OR detection system.

2. automatically actuated by the Unit 1 Area Fire alarm (Annunciator 0-37-A4).

D. 1. either the suppression OR detection system. 2. manually actuated by the NSO.

Answer: D

Answer Explanation Per BAR 0-37-A4, UNIT 1 AREA FIRE, the probable cause for for this alarm to be actuated is EITHER (1) - A Fire Detection System has been actuated OR (2) - A Fire Suppression System has been actuated. Also, per the BAR, if the suppression system is in alarm, the NSO notifies the US and actuates the plant wide fire alarm. If the detection system is in alarm, the alarm is verified and after verification the NSO notifies the US and actuates the plant wide fire alarm. A. Incorrect: The first part is incorrect but plausible as the suppression system does input to the

annunciator window actuation. The second part is incorrect but plausible as either a detection or suppression zone actuating will energize the Unit 1 Area Fire Window, and it is plausible to think that it would in turn actuate the Plant Wide Fire Alarm.

B. Incorrect: The first part is incorrect but plausible as the suppression system does input to the

annunciator window actuation. The second part is correct which makes this part of the distractor plausible.

C. Incorrect: The first part is correct which makes this part of the distractor plausible. The second

part is incorrect but plausible as either a detection or suppression zone actuating will energize the Unit 1 Area Fire Window and it is plausible to think that it would in turn actuate the Plant Wide Fire Alarm.

D. Correct as explained above. Reference: BAR 0-37-A4 The question meets the K/A by placing the examinee in a condition in which the Unit Fire Alarm has actuated and tests them on the operation of the Plant Wide Fire Alarm.




System ID: 1845790 User-Defined ID: 2017 NRC Q25C Cross Reference Number: S.FP1-01

Topic: Annunciator 0-37-A4, Unit 1 Area Fire has just alarmed. This alarm is actuated by ...


RO Level Low Cog Level APE 067; Plant fire on site AA1; Ability to operate and/or monitor the following as they apply to the Plant Fire on Site 1.06; Fire Alarm Objective: STATE the purposes of the Fire Protection System (S.FP1-01)



26 ID: 2017 NRC Q26 Points: 1.00 A Safety Injection occurs from 100% power due to a LOCA. The crew has entered 1BEP-0, REACTOR TRIP OR SAFETY INJECTION. Attachment B, Step 5; VERIFY CNMT ISOLATION PHASE A, is in progress with the following indications: • 1CV-8152, LTDOWN LINE CNMT ISOL VLV, indicates OPEN. • 1CV-8160, LTDOWN LINE CNMT ISOL VLV, indicates OPEN. • 1RF026, CNMT FLOOR DRN SUMP INSIDE ISOL VLV, indicates OPEN. • 1RF027, CNMT FLOOR DRN SUMP OUTSIDE ISOL VLV, indicates CLOSED. 1. To satisfy the MINIMUM Containment Integrity requirement, CLOSE ...

2. To satisfy the Procedural requirement of Attachment B, CLOSE ...

A. 1. ONE CV Letdown Isolation Valve.

2. ONE CV Letdown Isolation Valve.

B. 1. ALL OPEN Isolation Valves. 2. ONE CV Letdown Isolation Valve.

C. 1. ONE CV Letdown Isolation Valve.

2. ALL OPEN Isolation Valves.

D. 1. ALL OPEN Isolation Valves. 2. ALL OPEN Isolation Valves.

Answer: C

Answer Explanation



Per T.S. 3.6.3, Containment Isolation Valves, if one containment isolation valve is inoperable the other valve in the two isolation valve flow path must be closed within 4 hours and the valve must be verified closed once per every 31 days. If 2 valves are inoperable in a 2 valve flow path, then 1 valve must be closed within 1 hour. In this case, 1 of the 2 CV valves must be closed to satisfy containment integrity. ALL open containment isolation valves are closed per the procedure. A. Incorrect: The first part is correct. The second part is incorrect but plausible because

containment integrity is satisfied. B. Incorrect: The first part is incorrect but plausible as closing ALL valves is performed per the

procedure and that would satisfy containment integrity, but the question asks the minimum number of valves, which makes this incorrect. The second part is incorrect but plausible as that would satisfy containment integrity.

C. Correct: as described above. D. Incorrect: The first part is incorrect but plausible as closing ALL valves is performed per the

procedure and that would satisfy containment integrity, but the question asks the minimum number of valves, which makes this incorrect. The second part is correct because all open valves are directed to be closed per the procedure.

Technical References: T.S. 3.6.3 and bases BEP-0, attachment B Meets K/A, examinee must be able to monitor for the Loss of Containment Integrity. Question is RO Level as it requires knowledge of T.S. actions of 1 hour or less. The question is also RO by examining on the major actions contained in the procedure.




System ID: 1742689 User-Defined ID: 2017 NRC Q26 Cross Reference Number: S.PC1-10-C

Topic: A Safety Injection occurs from 100% power due to a LOCA. The crew has entered 1BEP-0, REACTOR


License Level: RO Cognitive Level: High, requires multiple analysis. APE: 069 Loss of Containment Integrity AA1. Ability to operate and / or monitor the following as they apply to the Loss of Containment Integrity: AA1.01 Isolation valves, dampers, and electro-pneumatic devices. Objective: Given a set of plant conditions. PREDICT the impact on the plant and/or Containment of various failures including: COMPONENT MALFUNCTIONS (S.PC1-10-C)



27 ID: 2017 NRC Q27 Points: 1.00 A LOCA has occurred on Unit 1. The crew is preparing to initiate a Cooldown to 200°F per 1BEP ES-1.2, POST LOCA COOLDOWN AND DEPRESSURIZATION: • RCS Pressure is 500 psig and STABLE • CETC's are indicating 400°F and STABLE • PZR level is 35% and RISING The crew must ________(1)________ and cooldown at __________(2)__________.

A. 1. allow the RH pumps to run

2. the maximum controllable rate

B. 1. stop the RH pumps 2. the maximum controllable rate

C. 1. allow the RH pumps to run

2. less than 100°F in any 1 hour period

D. 1. stop the RH pumps 2. less than 100°F in any 1 hour period

Answer: D

Answer Explanation



A. Incorrect. The conditions are met that required the RH pumps to be shutdown per the operator action summary page. Plausible because the step to stop the RH pumps is before the cooldown step, and also plausible because the note prior to the cooldown does state that the cooldown should be maximized. Also, plausible because other procedures such as 1BEP-3 direct to cooldown at maximum rate. B. Incorrect. The note before step 8, states that the cooldown should be maximized within allowable limits to reduce time required to reach RH. The allowable limit as stated in the step is less than 100°F in any 1 hour period. Plausible because the note prior to the cooldown does state that the cooldown should be maximized. Also, plausible because other procedures such as 1BEP-3 direct to cooldown at maximum rate. C. Incorrect. The conditions are met that required the RH pumps to be shutdown per the operator action summary page. Plausible because the step to stop the RH pumps is before the cooldown step, and also plausible because the note prior to the cooldown does state that the cooldown should be maximized within allowable limits. D. Correct. Step 6, Check if RH pumps should be stopped, is a continuous action step and the conditions to stop the RH pump are met. Also, the note before step 8, states that the cooldown should be maximized within allowable limits to reduce time required to reach RH; however, the allowable limit as stated in the step is less than 100°F in any 1 hour period. Meets K/A, examinee must have the ability to determine and interpret the following as they apply to the (LOCA Cooldown and Depressurization), adherence to appropriate procedures and operation within the limitations in the facility*s license and amendments. NOTE: Part 2 of question Q27 asks about RCS cooldown rates. Part 1 of Q11 also asks cooldown rates. Q11 is in reference to an extended Loss of All AC power while question 11 concerns Post LOCA Cooldown and Depressurization. The correct answer for both questions, concerning the cooldown rate portion of the question is the same: less than 100 degree per hour. Even though the correct answer for the cooldown rate is the same, they are addressing different circumstances. Q11 asks the reason(s) for the specified cooldown rate while Q27 does not. The correct answer is listed in Q11. The reason for the cooldown rate in Q27 is to preclude violation of the Integrity Status Tree which would direct the operators out of ES 1.2 and into the BSTs. Based on this data, the exam author feels there is a large enough difference between the questions based on (1) the event introduced in the stem and (2) the reason that that cooldown rate applies, to NOT consider Q11 and Q27 to "overlap". The other distractor "maximum controllable rate", stands on its' own for the plausibility as stated in each applicable distractor. Technical References: 1BEP ES-1.2, POST LOCA COOLDOWN AND DEPRESSURIZATION.




System ID: 1742690 User-Defined ID: 2017 NRC Q27 Cross Reference Number: T.EP02-01-E

Topic:

A LOCA has occurred on Unit 1. The crew is preparing to initiate a Cooldown to 200°F per 1BEP ES

RO Importance: 3.5 SRO Importance: 4.1 K/A: E03EA2.2 Comments: New Question

License Level: RO Cognitive Level: High, requires multiple analysis. E03 LOCA Cooldown and Depressurization EA2. Ability to determine and interpret the following as they apply to the (LOCA Cooldown and Depressurization) EA2.2 Adherence to appropriate procedures and operation within the limitations in the facility*s license and amendments. Objective: T.EP02-01-E, Without the use of EP-1, EP ES-1.1, 1.2, 1.3, 1.4: EXPLAIN the Operator Action Summary page of the procedure



28 ID: 2017 NRC Q28 Points: 1.00 Unit 1 is at 28% power when the 1A RCP trips. When plant conditions have STABLIZED: 1. The 1A Steam Generator Feedwater flow will be …

2. The 1B, C and D Steam Generator Feedwater flows will be …

A. 1. HIGHER

2. LOWER

B. 1. LOWER 2. HIGHER

C. 1. HIGHER

2. approximately the SAME

D. 1. LOWER 2. approximately the SAME

Answer: B

Answer Explanation With the loss of an RCP, there will be less heat transfer without forced circulation. The effected Steam Generator will be steaming less and require less feedwater flow to maintain level. In the other Steam Generators; steam flow will go up as the heat transfer rises in those Steam Generators, and feedwater flow will then rise to maintain Steam Generator level. A. Incorrect. See above. Plausible because the INITIAL response is for feedwater flow to rise in the effected Steam Generator due to the temperature drop from reverse RCS flow in that loop and the initial feedwater flow rise in the effected Steam Generator would lead to a temporary feedwater flow reduction to the other Steam Generators. B. Correct. See above. C. Incorrect. See above. Plausible because the INITIAL response is for feedwater flow to rise in the effected Steam Generator due to the temperature drop from reverse RCS flow in that loop. Also, power level has not changed; therefore, it's plausible to conclude that total feedwater flow to all the Steam Generators would remain approximately the same. D. Incorrect. See above. The first part is correct making the distractor plausible. Also, power level has not changed; therefore, it's plausible to conclude that total feedwater flow to all the Steam Generators would remain approximately the same. Meets K/A, examinee must have knowledge of the effect that a loss of an RCP will have on Feedwater. Technical References: Lesson Plan, I1-RC-XL-02.




System ID: 1742751 User-Defined ID: 2017 NRC Q28 Cross Reference Number: S.RC2-07-D

Topic:

Unit 1 is at 28% power when the 1A RCP trips. When plant conditions have STABLIZED: The 1A

RO Importance: 2.8 SRO Importance: 3.1 K/A: 003K3.03 Comments: New Question

License Level: RO Cognitive Level: Low, from memory. 003 Reactor Coolant Pump System (RCPS) K3 Knowledge of the effect that a loss or malfunction of the RCPS will have on the following: K3.03 Feedwater and emergency feedwater. Objective: S.RC2-07-D, ANALYZE and PREDICT the effect that a loss of (a) Reactor Coolant Pumps will have on the following: Feedwater/Auxiliary Feedwater Systems



29 ID: 2017 NRC Q29 Points: 1.00 The Unit is operating at 100% power. • ALL RCP Seal Injection flows are 8 gpm. An INADVERTENT Safety Injection occurs. How will RCP Seal Injection flow respond and why?

A. LOWER because Pressurizer level is RISING.

B. LOWER because CV121, CENT CHG PMPS FLOW CONT VLV, CLOSES on an SI

signal.

C. RISE due to the start of the second CV pump ONLY.

D. RISE due to the start of the second CV pump AND normal charging isolating.

Answer: D

Answer Explanation



On a Safety Injection Signal, normal charging is isolated and the second CV pump will auto start and inject via the CHG PMPS TO COLD LEGS INJ ISOL VALS SI8801A/B opening. This will increase charging flow. 1CV121 may go closed based on increasing PZR level; however, when in auto (as the question posits), there is a minimum "stop" which limits flow to 52 gpm even though the controller output is at minimum. This design ensures the valve will remain open to supply RCP seal injection flow even when normal charging is isolated. As run on the simulator on 7/13/2017 from a critical condition, RCP seal injection flows of approximately 8 gpm were noted. When the Inadvertent SI occurred, seal injection flows were approximately 14 gpm to each RCP. A. Incorrect: Seal injection flows lowering due to pressurizer level rising is plausible as this would

occur under normal operations. As PZR level rises 1CV-121 would throttle closed. Throttling closed CV121 would reduce total charging flow output and lower seal injection flow. This is incorrect, however, as normal charging is isolated on a SI.

B. Incorrect: On a Safety Injection Signal, instrument air is isolated to containment. 1CV121, is

an air operated valve that fails on a loss of air. The valve fails open but, failing closed is plausible as examinees frequently confuse the failing position of the many air operated valves of the nuclear plant. If the valve did fail closed, seal injection to the RCPs would go to "0".

C. Incorrect: Plausible because the start of the second CV pump WOULD cause seal injection

flow to RISE; however, it is incorrect because injection flow rises for 2 reasons; they are: (1) - start of the second charging pump and (2) - the securing of the normal charging line.

D. Correct as stated in the above explanation. Question meets the K/A placing the candidate in a position of monitoring RCP seal injection flow due automatic actions that take place on a Safety Injection Signal. Question is Low Cog due to memory References: EP-0 step 4 for CV pump operation and SI 8801A/B opening E-0 attachment B step 2 Verify ECCS Valve alignment, Group 2 cold leg monitor lights lit that contain normal charging isolation valves: 1CV 8105 and 1CV8106 closed. CVCS Lesson plan (chapter 15) page 31 on CV121 control board controls.




System ID: 1742692 User-Defined ID: 2017 NRC Q29 Cross Reference Number: S.CV1-17-B

Topic:

The Unit is operating at 100% power. ALL RCP Seal Injection flows are 8 gpm. An INADVERTENT SI

RO Importance: 3.3 SRO Importance: 3.2 K/A: 003A3.01 Comments: New question

RO Level Low Cog Level (memory) K/A: 003: Reactor Coolant Pump, A3: Ability to monitor automatic operation of the RCPS, including:, .01: Seal Injection Flow Objective: S.CV1-17-B: PREDICT how each of the following supported systems will be impacted by CVCS failures (after the listed system has been taught): b. RCP



30 ID: 2017 NRC Q30 Points: 1.00 At 1000 A Reactor Trip and Safety Injection are initiated: • RCS pressure is 2300 PSIG and stable • Pressurizer level is 25% and stable • 1CV8152, LTDWN LINE CNMT ISOL VLV, is closed • 1CV8160, LTDWN LINE CNMT ISOL VLV, failed to close At 1001 Letdown orifice isolation valves indicate: • 1CV8149A OPEN • 1CV8149B OPEN • 1CV8149C CLOSED The Letdown orifice isolation valves __________(1)__________

AND

1CV8149A and B have ________(2)________.

A. 1. receive a close signal DIRECTLY from SI 2. failed open

B. 1. receive a close signal DIRECTLY from SI

2. responded as designed

C. 1. are interlocked with 1CV8160 2. failed open

D. 1. are interlocked with 1CV8160

2. responded as designed

Answer: D

Answer Explanation



A. Incorrect. See below. Plausible because SI will generate a Phase A signal. Operators sometimes confuse which valves reposition on a SI signal with which valves reposition on a Phase A signal and confuse the interlocks of 1CV8160 with 1CV459 and 1CV460. If the valves had received an auto close signal then they would appear to be failed open. B. Incorrect. See below. Plausible because the Pressurizer low level alarm would be lit with Pressurizer level at 19%. 1CV8149A, B, C auto close on Pressurizer low level. If 1CV8149A, B, C had received an auto close signal, then they would appear to be failed open with the indications given in the stem of the question. C. Incorrect. See below. Plausible because operators sometimes confuse the open interlocks with the close interlocks for 1CV8149A, B, C. D. Correct. 1CV8149A, B, C (letdown orifice isolation valves) are interlocked with 1CV8160 (letdown containment isolation valve) such that when 1CV8160 closes, 1CV8149A, B, C will auto close. 1CV8160 will receive a Phase A signal to close when SI is actuated. Per the conditions in the stem, 1CV8160 is open, so 1CV8149A, B and C did not get a closed signal. The normal lineup with 120gpm letdown is for two valves to be open. Also, 1CV8149A, B and C will go close on Pressurizer Low Level of 17%. With level at 25%, no close signal was generated. Meets K/A, examinee must have the ability to monitor automatic operation of the CVCS, including letdown orifice isolation valve position. Technical References: BOP CV-1a, STARTUP OF THE CV SYSTEM.


System ID: 1742754 User-Defined ID: 2017 NRC Q30 Cross Reference Number: S.CV1-06-C

Topic: A Reactor Trip and Safety Injection are initiated: RCS pressure is 2300 PSIG and stable Pressuriz

RO Importance: 2.8 SRO Importance: 2.7 K/A: 004A3.18 Comments: New Question

License Level: RO Cognitive Level: High, requires multiple analysis. 004 Chemical and Volume Control System (CVCS) A3 Ability to monitor automatic operation of the CVCS, including: A3.18 Interpretation of letdown orifice isolation valve position indicators Objective: S.CV1-06-C, STATE and EXPLAIN the interlocks associated with the following: CV-8149 A, B, C.



31 ID: 2017 NRC Q31 Points: 1.00 Unit 1 is at 100% power. 1BOSR 0.5-2.SI.2-2.2, 1SI8802B, 1SI8809B, 1SI8811B AND 1SI8923B STROKE TIME AND POSITION INDICATION TEST results: Stroke

Direction Observe Time (Sec)

Administrative Stroke Limit (Sec)

Applicable LCOAR (Admin Limit)

Maximum Stroke Limit (Sec)

Applicable LCOAR (Max Limit)

1SI8809B Open 8.3 5.0 – 8.4 0BOL IST1 10.1 1BOL 5.2 1BOL 5.3

1SI8809B Closed 9.3 5.5 – 9.1 0BOL IST1 11.0 1BOL 5.2 1BOL 5.3

1SI8811B Open 98 79.1 – 100.0 0BOL IST1 100.0 1BOL 5.2 1BOL 5.3

1SI8811B Closed 105 77.4 – 100.0 0BOL IST1 100.0 1BOL 5.2 1BOL 5.3

What Tech Spec LCO(s) apply, if any?

A. NO Tech Spec LCOs apply.

B. Enter Tech Spec 3.5.2 ECCS-OPERATING ONLY for 1SI8811B.

C. Enter Tech Spec 3.5.2, ECCS-OPERATING for 1SI8809B AND 1SI8811B.

D. Enter Tech Spec 3.5.2 ECCS-OPERATING AND Tech Spec 3.5.3 ECCS-SHUTDOWN for 1SI8811B.

Answer: B

Answer Explanation



A. Incorrect. See below. Plausible because 1SI8811B open stroke time is less than the maximum stroke limit. It is a common misconception that these valves only need to meet open stroke times to comply with Tech Specs. B. Correct. 1SI8811B closed stroke time is greater than the maximum limit so entry into Tech Spec 3.5.2 is required. C. Incorrect. See above. Plausible because both 1SI8809B and 1SI8811B closed stroke times are outside the administrative limit. It is a common misconception that not meeting the administrative limit requires entry into the Tech Spec. Also, 1SI8811B closed stroke time is greater than the maximum limit so entry into Tech Spec 3.5.2 is required. D. Incorrect. See above. Plausible because both Tech Specs are listed as applicable by the procedure. Meets K/A, examinee must be able to determine test acceptability by comparison of recorded valve response times with Tech-Spec requirements. Technical References: 1BOSR 0.5-2.SI.2-2.2, 1SI8802B, 1SI8809B, 1SI8811B AND 1SI8923B STROKE TIME AND POSITION INDICATION TEST. 1BOSR 0.5-2.SI.2-2.2.BY01, Acceptance Criteria Data Sheet. Tech Spec 3.5.2 and 3.5.3. 0BOL IST1, IST - ASME STROKE TIMES




System ID: 1742773 User-Defined ID: 2017 NRC Q31 Cross Reference Number: S.RH1-12

Topic:

Unit 1 is at 100% power. 1BOSR 0.5-2.SI.2-2.2, 1SI8802B, 1SI8809B, 1SI8811B AND 1SI8923B STROKE


License Level: RO Cognitive Level: High, requires multiple analysis. 005 Residual Heat Removal System (RHRS) A1 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the RHRS controls including: A1.07 Determination of test acceptability by comparison of recorded valve response times with Tech-Spec requirements Objective: S.RH1-12, Given a set of plant conditions, DETERMINE the applicable RH System Tech Spec/TRM operability requirements.



32 ID: 2017 NRC Q32 Points: 1.00 While placing the 1A RH train in shutdown cooling per BOP RH-6, OPERATION OF THE RH SYSTEM IN SHUTDOWN COOLING, throttle _______ (1) _______,

_______ (2) _______ to adjust the heatup of the RH system?

A. 1. 1CC9412A, CC to RH HX 1A

2. remotely at 1PM06J

B. 1. 1CC9412A, CC to RH HX 1A 2. locally at 364 elevation Aux. Building.

C. 1. 1RH606, RH HX A Outlet Flow Cont Valve

2. remotely from 1PM06J.

D. 1. 1RH606, RH HX A Outlet Flow Cont Valve 2. locally at 364 elevation Aux. Building.

Answer: C

Answer Explanation A. Incorrect. BOP RH-6 directs the operator to throttle 1RH606 to control the heatup of the RH system. Plausible because throttling 1CC9412A would affect the amount of RH cooling; also, BOP RH-6 does direct positioning this valve. Plausible also as this valve can be manipulated from 1PM06J. B. Incorrect. BOP RH-6 directs the operator to throttle 1RH606 to control the heatup of the RH system. Plausible because throttling 1CC9412A would affect the amount of RH cooling; also, BOP RH-6 does direct positioning this valve. Also plausible as this valve can be operated locally in the Aux. Bldg. C. Correct. BOP RH-6 directs the operator to throttle 1RH606 to control the heatup of the RH system. The valve manipulation is from the control room. D. Incorrect. Plausible as BOP RH-6 directs the operator to throttle 1RH606 to control the heatup of the RH system. Also plausible because RH606 can be manipulated locally at the valve as contained in 1BOA PRI-5, Attachment C. Meets K/A, Requires examinee to have the ability to locate and operate components, including local controls for the RH system. Technical References: BOP RH-6, OPERATION OF THE RH SYSTEM IN SHUTDOWN COOLING




System ID: 1742188 User-Defined ID: 2017 NRC Q32 Cross Reference Number: S.RH1-09-A

Topic: While placing the 1A RH train in shutdown cooling per BOP RH-6, OPERATION OF THE RH SYSTEM IN SHUTD

RO Importance: 4.4 SRO Importance: 4.0 K/A: 005G2.1.30 Comments: New Question

License Level: RO Cognitive Level: High, multiple analysis required. 005 Residual Heat Removal System (RHRS) G2.1.30 Ability to locate and operate components, including local controls. Objective: S.RH1-09-A, DISCUSS operation of the RH system during: Plant Cooldown (Include method of Cooldown Rate Control)



33 ID: 2017 NRC Q33A Points: 1.00 The accident in which a Pressurized Thermal Shock condition has the GREATEST probability of occurring IF ECCS pumps are NOT STOPPED when conditions are met, is a

A. steam Line break in containment.

B. small break loss of coolant.

C. large break loss of coolant.

D. pressurizer vapor space loss of coolant.

Answer: A

Answer Explanation The correct answer is steam line break in containment as this event causes a rapid RCS cooldown and if ECCS pumps are allowed to continue to run, a rapid repressurization of the RCS occurs which could result in a PTS event. This event is proceduralized in 1BFR-P.1, RESPONSE TO IMMINENT PRESSURIZED THERMAL SHOCK CONDITION, at step 2, Check RCS Cold Leg Temperatures. The concern is addressed to maintain heat sink yet minimize the cooldown following a Faulted S/G. The subsequent steps check if ECCS can be terminated. There is a note just prior to step 2 that is specifically directed to a faulted S/G, which is the principal accident that this procedure is addressing. A. Correct as explained above. B. Incorrect: This is plausible as a small break LOCA will inject cool water into the RCS, resulting

in the prerequisite of an RCS cooldown and because it is a small break, RCS pressure will remain relatively high.

C. Incorrect: A Large Break LOCA is plausible as this will cool down the RCS significantly due to

the large quantity of cool water injection from the ECCS pumps. This is incorrect, however, because due to the Large Break, the RCS will not repressurize significantly.

D. Incorrect: The Pressurizer Vapor Space Loss of Coolant is plausible as ECCS will inject,

cooling down the RCS. Plausibility is further enhanced by the fact that RCS pressure will remain relatively high.

References: 1BFR-P.1, RESPONSE TO IMMINENT PRESSURIZED THERMAL SHOCK CONDITION The question meets the K/A by placing the candidate in a situation when ECCS pumps are operating and testing the examinee about which accident is of the highest concern in regards to PTS




System ID: 1744212 User-Defined ID: 2017 NRC Q33A Cross Reference Number: T.FR04-04-A

Topic: The accident in which a Pressurized Thermal Shock condition has the GREATEST probability of


RO Level High Cog Level K/A: 006: Emergency Core Cooling System (ECCS), K5: Knowledge of the operational implications of the following concepts as they apply to ECCS, .10: Theory of thermal stress. Objective: T.FR04-04-A, Given a set of plant conditions, DIAGNOSE and ANALYZE a Response to Imminent Pressurized Thermal Shock Condition.



34 ID: 2017 NRC Q34 Points: 1.00 Unit 1 is at 100% power • A Pressurizer PORV fails OPEN. • The associated PZR PORV Block valve will NOT CLOSE. • Operators perform a manual reactor trip and safety injection. At 1000 • The PRT rupture disc RUPTURED. At 1005, which of the following indications can be used to confirm the PRT rupture disk has RUPTURED?

A. PRT level is RISING.

B. RVLIS level is LOWERING.

C. PRT level is RAPIDLY LOWERING.

D. Containment sump leak detection flow rates are RISING.

Answer: D

Answer Explanation A. Incorrect. PRT level rising is incorrect because it should be filled and does not mean the rupture disc has opened. Plausible because the PZR is still discharging into the PRT through the stuck open PORV B. Incorrect. The change in RVLIS level would not be due to the PRT rupture disk being ruptured. Plausible because a Vapor Space LOCA would result in lowering RVLIS level. C. Incorrect. PRT level lowering rapidly is incorrect as the rupture disc is on top of the PRT. There may be a minor level drop but it would not lower rapidly. Plausible because the temperature of the PRT at this point would be high enough for some of the water to flash off, but the level drop would be very minor. D. Correct. The PRT rupture would result in a steam input into containment, this would have a direct effect on containment sump flowrate. The end result would be a rise in flowrate. Meets K/A, the PRT rupture would result in a steam input into containment; this would have a direct effect on containment sump flowrate. Examinee must have knowledge of the effects that the PRT rupture would have on containment. Technical References: Lesson PlanI1-RY-XL-01, CH 14




System ID: 1741885 User-Defined ID: 2017 NRC Q34 Cross Reference Number: S.RY1-15

Topic:

Unit 1 is at 100% power A Pressurizer PORV fails OPEN. The associated PZR PORV Block valve will

RO Importance: 3.3 SRO Importance: 3.6 K/A: 007K3.01 Comments: <QQ 1138040(1412)><<Bank Question from 2013 Cert Exam

RO Question #6 License Level: RO Cognitive Level: High, multiple analysis required. 007 Pressurizer Relief Tank/Quench Tank System (PRTS) K3 Knowledge of the effect that a loss or malfunction of the PRTS will have on the following: K3.01 Containment Objective: S.RY1-15, STATE the internal design pressure of the PRT. DISCUSS how the PRT is protected from exceeding this pressure



35 ID: 2017 NRC Q35 Points: 1.00 Unit 1 is in MODE 5 preparing to draw a bubble in the PZR: • PRT level = 71% • PRT pressure = 4.5 psig • PRT temperature = 85°F Following venting the PZR solid: • PRT level = 77% and STABLE • PRT pressure = 7.2 psig and SLOWLY RISING • PRT temperature = 92°F and STABLE What caused this PRT response?

A. PZR PORV is failed OPEN.

B. Nitrogen Regulator is failed OPEN.

C. RCP Seal Leakoff Relief valve is failed OPEN.

D. Gaseous Waste isolation valve is failed OPEN.

Answer: B

Answer Explanation A. Incorrect. If the PORV had not closed, the level and pressure would be rising. This is plausible because the valve was opened in the question stem. B. Correct. At 6 psig, RY469 auto closes to GW from the PRT. The BAR says probable cause is (1) Valve leakoff or relief valve flow, (2) PORV or Safety Valve lifted, (3) Filling PRT, and (4) N2 Regulator failure. With no additional level or temperature rise the N2 regulator would be the appropriate selection for cause. C. Incorrect. If the RCP Seal Leakoff Relief valve was lifting, pressure and level would be rising. This is plausible because the valve is an input into the PRT. D. Incorrect. The Gaseous Waste isolation valve is an output from the PRT. This is plausible because the valve closing is an automatic action for high pressure in the PRT. Meets K/A, examinee must have knowledge of components that discharge into the PRT. Technical References: Lesson Plan I1-RY-XL-01, CH 14





Topic:

Unit 1 is in MODE 5 preparing to draw a bubble in the PZR: PRT level = 71% PRT pressure = 4.5 psig

RO Importance: 2.7 SRO Importance: 2.9 K/A: 007A3.01 Comments: <QQ 1138040(1412)><<Bank Question from 2016 Cert

Question #30 (1144466) License Level: RO Cognitive Level: High 007 Pressurizer Relief Tank/Quench Tank System (PRTS) A3 Ability to monitor automatic operation of the PRTS, including: A3.01 Components which discharge to the PRT Objective: S.RY1-14, LIST the sources which discharge into the PRT



36 ID: 2017 NRC Q36 Points: 1.00 Unit 1 is at 100% power. 1CC685 inadvertently closes. 1. _____________cooling has been lost.

2. The RCP’s __________ be shut down.

A. 1. All RCP

2. Will

B. 1. ONLY RCP thermal barrier 2. Will

C. 1. All RCP

2. will NOT

D. 1. Only RCP thermal barrier 2. will NOT

Answer: D

Answer Explanation



A. Incorrect. Thermal Barrier isolation requires entry into 1BOA PRI-6, COMPONENT COOLING MALFUNCTION; however, the procedure does not require an RCP trip due to seal injection is still available. Also, the CC system has a single line that feeds all cooling on the RCP but, a separate line coming from the oil coolers and the thermal barrier. 1CC685 is on the return side of the thermal barriers. Plausible because a reactor trip is required for a loss of CC cooling to the RCP. Also plausible because the CC system has a single line that feeds all cooling on the RCP and the examinee may assume that 1CC685 is located on the line that feeds all cooling to the RCP. B. Incorrect. Thermal Barrier isolation requires entry into 1BOA PRI-6, COMPONENT COOLING MALFUNCTION; however, the procedure does not require an RCP trip due to seal injection is still available. Also, the CC system has a single line that feeds all cooling on the RCP but, a separate line coming from the oil coolers and the thermal barrier. 1CC685 is on the return side of the thermal barriers. Plausible because a reactor trip is required for a loss of all CC cooling to the RCP and the examinee may assume that 1CC685 is located on the line that feeds all cooling to the RCP. Part 1 is correct which makes this part plausible. C. Incorrect. Thermal Barrier isolation requires entry into 1BOA PRI-6, COMPONENT COOLING MALFUNCTION; however, the procedure does not require an RCP trip due to seal injection is still available. Also, the CC system has a single line that feeds all cooling on the RCP but, a separate line coming from the oil coolers and the thermal barrier. 1CC685 is on the return side of the thermal barriers. Plausible because the RCPs will not need to be shut down and because the CC system has a single line that feeds all cooling on the RCP. D. Correct. Thermal Barrier isolation requires entry into 1BOA PRI-6, COMPONENT COOLING MALFUNCTION; however, the procedure does not require an RCP trip due to seal injection is still available. Also, the CC system has a single line that feeds all cooling on the RCP but a separate line coming from the oil coolers and the thermal barrier. 1CC685 is on the return side of the thermal barriers. Technical References: 1BOA PRI-6, COMPONENT COOLING MALFUNCTION. Meets K/A, requires examinee knowledge of the cause-effect relationships between the CCWS and the Loads cooled by CCWS.




System ID: 1742153 User-Defined ID: 2017 NRC Q36 Cross Reference Number: S.CC1-18

Topic: Unit 1 is at 100% power. 1CC685 inadvertently closes.

RO Importance: 3.3 SRO Importance: 3.4 K/A: 008K1.02 Comments: Modified ILT Exam Bank (412277)

License Level: RO Cognitive Level: Low, from memory. 008 Component Cooling Water System (CCWS) K1 Knowledge of the physical connections and/or cause-effect relationships between the CCWS and the following systems: K1.02 Loads cooled by CCWS Objective: S.CC1-18, Given an operating mode and/or various plant conditions, PREDICT how supported systems and the CC System will be impacted by various CC System failures and mis-operations.



37 ID: 2017 NRC Q37 Points: 1.00 Unit 1 is at 100% power. • An RCS pressure transient occurs and 1RY8010A, Pressurizer Safety Valve, is stuck open. • RCS pressure is dropping. 1. The LOWEST pressure at which the Pressurizer Spray valves will go fully closed is...

AND

2. The LOWEST pressure at which the Pressurizer Backup heaters will energize is...

A. 1. 2310 PSIG.

2. 2225 PSIG.

B. 1. 2260 PSIG. 2. 2225 PSIG.

C. 1. 2310 PSIG.

2. 2220 PSIG.

D. 1. 2260 PSIG. 2. 2220 PSIG.

Answer: D

Answer Explanation A. Incorrect. See below. Plausible because 2310 PSIG is the pressure at which the Pressurizer Spray valves will begin to close and 2225 PSIG is the pressure at which the Pressurizer Backup heaters would de-energize. B. Incorrect. See below. Plausible because 2260 PSIG is the pressure at which the Pressurizer Spray valves would go fully closed and 2225 PSIG is the pressure at which the Pressurizer Backup heaters de-energize. C. Incorrect. See below. Plausible because 2310 PSIG is the pressure at which the Pressurizer Spray valves begin to close and 2220 PSIG is the pressure at which the Pressurizer Backup heaters energize. D. Correct.1RY8010A will open at 2460 PSIG. As RCS pressure drops, Pressurizer Spray valves will begin to close at 2310 PSIG and be full closed at 2260 PSIG. While Pressurizer Backup heaters will energize at 2220 PSIG. The Pressurizer Backup heaters would de-energize at 2225 PSIG with RCS pressure rising. Meets K/A, examinee must have knowledge of the effect of a loss or malfunction of the Pressurizer, specifically a malfunction of the Pressurizer Safety Valve, will have on the Pressurizer Pressure Control System. Technical References: BOP RY-100, PRESSURIZER OPERATION



Question 37 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 0 Difficulty: 0.00 System ID: 1742775 User-Defined ID: 2017 NRC Q37 Cross Reference Number: S.RY1-20

Topic: Unit 1 is at 100% power. An RCS pressure transient occurs and 1RY8010A, Pressurizer Safety Valve,


License Level: RO Cognitive Level: Low, from memory. 010 Pressurizer Pressure Control System (PZR PCS) K6 Knowledge of the effect of a loss or malfunction of the following will have on the PZR PCS: K6.02 PZR Objective: S.RY1-20, Without the aid of the pressurizer level and pressure control logic diagrams, be able to DISCUSS all inputs, outputs, bistable setpoints, control functions, and interlocks associated with the logic diagrams.



38 ID: 2017 NRC Q38 Points: 1.00 Unit 1 is at 100% power with the following indication:

While bypassing the functions of 1PT-937 per 1BOA INST-2, OPERATION WITH A FAILED INSTRUMENT CHANNEL, the following indications were observed before the operators were STOPPED.

Containment Spray ___(1)___ actuated and the crew must ___(2)___.

A. 1. is 2. resolve the condition using 1BOA INST-2, OPERATION WITH A FAILED

INSTRUMENT CHANNEL

B. 1. is 2. enter 1BEP-0, REACTOR TRIP OR SAFETY INJECTION

C. 1. is NOT

2. resolve the condition using 1BOA INST-2, OPERATION WITH A FAILED INSTRUMENT CHANNEL

D. 1. is NOT

2. enter 1BEP-0, REACTOR TRIP OR SAFETY INJECTION



Answer: C

Answer Explanation A. Incorrect. See below. Plausible because taking TS-1 to test will trip bistables for that channel. A note prior to tripping bistables states: performing the following actions in conjunction with previously tripped bistables may result in inadvertent actuations. B. Incorrect. See below. Plausible because taking TS-1 to test will trip bistables for that channel. A note prior to tripping bistables states: performing the following actions in conjunction with previously tripped bistables may result in inadvertent actuations. Part 2 of this distractor is plausible as the first bistable that was inadvertently tripped places the unit in a "half trip" condition of the Reactor Protection System. C. Correct. Taking TS-1 to test only trips one channel of Safety injection and Main Steam isolation and TS-1 does not affect the Containment Spray bistable; only one channel of Containment Spray is actuated. D. Incorrect. See above. The first part is correct making the distractor plausible. The second part is plausible because taking TS-1 to test will trip bistables for that channel and the examinee may assume that this includes the Containment Spray bistable. Containment Spray is a 2 of 4 logic. Meets K/A, examinee must have the ability to (a) predict the impacts of incorrect channel bypassing on the RPS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations. The inadvertent tripping of bistables has a direct affect on the Reactor on the Reactor Protection System as described in distractor B. Technical References: 1BEP-0, REACTOR TRIP OR SAFETY INJECTION. 1BOA INST-2, OPERATION WITH A FAILED INSTRUMENT CHANNEL



Question 38 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 0 Difficulty: 0.00 System ID: 1744175 User-Defined ID: 2017 NRC Q38 Cross Reference Number: T.OA11-22 Topic: Unit 1 is at 100% power with the following indication: RO Importance: 3.4 SRO Importance: 3.7 K/A: 012A2.03 Comments: New question

License Level: RO Cognitive Level: High, multiple analysis. 012 Reactor Protection System A2 Ability to (a) predict the impacts of the following malfunctions or operations on the RPS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: A2.03 Incorrect channel bypassing. Objective: T.OA11-22, Given a step, note or caution from 1/2BOA INST-2, Operation with Failed Instrument Channel, EXPLAIN the basis of that note, caution or step.



39 ID: 2017 NRC Q39 Points: 1.00 Which power supply, IF LOST, would prevent 1B CC and 1B SX pumps from sequencing on during a loss of offsite power (LOOP)?

A. Bus 131X.

B. DC bus 114.

C. Instrument bus 112.

D. Instrument bus 114.

Answer: D

Answer Explanation 1BOA ELEC-2, LOSS OF INSTRUMENT BUS, Table D lists the functions that are lost on a Loss of Instrument Bus. The first function lists "Train B ESF loads will NOT actuate or reset". A. Incorrect. Loss of 131X would prevent train A ESF equipment from actuating due to loss of

power to some valves. The distractor is plausible because 131X powers train A ESF valves and because 131X is commonly mistaken as a power source for Train B ESF equipment.

B. Incorrect. DC bus 114 does not supply control power for train B equipment. The distractor is

plausible because DC 112 provides control power for train B ESF equipment and DC bus 114 is commonly mistaken as a power source for Train B. Also plausible because, the functions of DC bus 114 are commonly confused with the functions of Instrument bus 114.

C. Incorrect. Instrument bus 114 powers the slave relays in SSPS that function to start B train

ESF equipment. The distractor is plausible as this component does supply B train ESF instrumentation and relays, but NOT the SSPS slave relays.

D. Correct. A loss of instrument bus 114 de-energizes the slave relays in SSPS that would

normally energize during an ESF actuation to start the B train ESF equipment. Meets K/A, requires examinee to identify the bus power supply (to the appropriate portion of the ESFAS/safeguards equipment control circuitry) that would prevent the B train of ESF equipment from sequencing. Technical References: 1BOA ELEC-2, LOSS OF INSTRUMENT BUS





Topic: Which power supply, IF LOST, would prevent 1B CC and 1B SX pumps from sequencing on during a loss

RO Importance: 3.6 SRO Importance: 3.8 K/A: 013K2.01 Comments: <QQ 1138040(1412)><<Bank Question from BWD 2013 NRC

Exam Q 21 (ID: RS10013-N03) License Level: RO Cognitive Level: Low, from memory 013 Engineered Safety Features Actuation System (ESFAS) K2 Knowledge of bus power supplies to the following: K2.01 ESFAS/safeguards equipment control Objective: T.OA02-02, ANALYZE a given set of plant conditions and DETERMINE if entry into 1/2BOA ELEC-2, Loss of Instrument Bus, is required.



40 ID: 2017 NRC Q40 Points: 1.00 At 1000 The plant is operating at 100% power. • The IM Dept. is testing PZR Pressure channel 457 and currently has the channel in the TRIPPED

condition. At 1005 PZR Pressure channel 455 has failed LOW. 1. The reactor ___ (1) ___ TRIPPED.

2. Which procedural flowpath applies, if any?

A. 1. has NOT

2. No procedure entry is required.

B. 1. has NOT 2. Enter BOA INST-2, OPERATION WITH A FAILED INSTRUMENT CHANNEL.

C. 1. HAS

2. Enter BEP-0, REACTOR TRIP OR SAFETY INJECTION, ONLY.

D. 1. HAS 2. Enter BEP-0, REACTOR TRIP OR SAFETY INJECTION, and then transition to BEP

ES-1.1, SI TERMINATION.

Answer: D

Answer Explanation



With 1 PZR Press channel in the tripped condition and then a second channel failing LOW, would make up the 2 of 4 coincidence for an SI Actuation on Low PZR Press. The reactor would trip on SI. This would require entry into E-0 and then the examinee would enter ES-1.1 (SI Termination) directly from E-0, at step 13, to terminate the SI condition. A. Incorrect. The first part is incorrect but plausible because the pressure channel failed low

which did NOT cause a PZR PRESS HIGH Reactor Trip (another ESFAS function from PZR Pressure). This is also plausible as the reactor will not trip on Low Pressurizer Pressure when power is less than P-7. The second part is incorrect but plausible since the reactor did not trip. Also it is plausible that no procedure entry is required because Tech Spec actions apply, and the candidate may consider that as sufficient.

B. The first part is incorrect but plausible because the pressure channel failed low which did NOT

cause a PZR PRESS HIGH Reactor Trip. (another ESFAS function from PZR Pressure.) The second part is incorrect but plausible because if the reactor did not trip, entry into BOA INST-2 would be the correct response.

C. Incorrect. The first part is correct making this part of the distractor plausible. The second part

is incorrect but plausible as the candidate will enter E-0. However, this will not mitigate the SI actuation.

D. Correct. The reactor will trip as described in the explanation above. The correct procedure

flowpath will be from E-0 through step 12 then at step 13 a transition will be made to ES-1.1 (SI Termination).

Meets KA, requires examinee to predict the impacts of inadvertent ESFAS actuation and based on those predictions, use procedures to correct, control, or mitigate the consequences of the inadvertent ESFAS actuation. This is RO level based on required knowledge of the procedure purpose: From the ES-401 Attachment 2 Figure 2: Screening for SRO-only linked to 10CFR55.43(b)(5) (Assessment and selection of procedures): 1) assessing plant conditions and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. Using the flowchart, this question can: • NOT be answered solely by knowing “systems knowledge”, i.e., how the system works, flowpath, logic, component location. No • NOT be answered solely by knowing immediate operator actions. No • NOT be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs. No • NOT be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure. YES, the question can be answered with knowledge of the procedure purpose and system knowledge. Technical References: BEP-0, REACTOR TRIP OR SAFETY INJECTION BEP ES-1.1, SI TERMINATION




System ID: 1742081 User-Defined ID: 2017 NRC Q40 Cross Reference Number: 3D.EP-04-C

Topic: At 1000 The plant is operating at 100% power. The IM Dept. is testing PZR Pressure channel 457


License Level: RO Cognitive Level: High due to analysis 013 Engineered Safety Features Actuation System (ESFAS) A2 Ability to (a) predict the impacts of the following malfunctions or operations on the ESFAS; and (b) based Ability on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations; A2.06 Inadvertent ESFAS actuation Objective: 3D.EP-04-C, Given a set of plant conditions or parameters indicating an Inadvertent Safeguards Actuation and a set of plant procedures, IDENTIFY the correct procedure(s) to be utilized and DISCUSS required operator actions.



41 ID: 2017 NRC Q41A Points: 1.00 During a LOCA which of the following containment parameters is/are the bases RCFCs shift to LOW SPEED operation from HIGH SPEED?

A. High temperature ONLY

B. High pressure ONLY

C. High humidity ONLY

D. High temperature, high pressure AND high humidity

Answer: C

Answer Explanation Per T.S. B 3.6.6, the Containment Spray System and Containment Cooling System limit the temperature and pressure that could be experienced following a DBA. The limiting DBAs considered are the LOCA and the Steam Line Break (SLB). Per T.S. B 3.6.6, page 4, during post-accident operation “the fans shift to low speed to prevent adverse fan conditions (e.g. motor overload, increased blade stresses) from the higher mass atmosphere.” Higher mass implies higher humidity since containment is in post-accident conditions. A. Incorrect. See above. Plausible because the Containment Cooling System is designed to limit

the temperature rise that would be experienced following a LOCA. High temperature can cause higher motor coil resistance which could degrade motor performance.

B. Incorrect. See above. Plausible because the Containment Cooling System is designed to limit

the pressure rise that would be experienced following a LOCA. High pressure is plausible as in a confined volume, as pressure rises, density of a gas would rise. The rising density means that more pounds mass are being circulated per cubic volume of air flow, making this distractor plausible.

C. Correct. See above. D. Incorrect. See above. Plausible because the Containment Cooling System is designed to limit

the temperature and pressure rise that could be experienced following a LOCA and the fans are shifted to Low Speed to prevent adverse fan conditions from the humidity rise. Plausibility of high temperature and pressure is explained in distractors A and B above.

Meets the K/A. The examinee must have knowledge that the RCFCs shift to low speed during a LOCA because they are subject to damage by high temperature, humidity and pressure in containment. Reference: B 3.6.5 B 3.6.6




System ID: 1742778 User-Defined ID: 2017 NRC Q41A Cross Reference Number: S.VP1-13

Topic: During a LOCA which of the following containment parameters is/are the reason(s) RCFCs shift to LOW

RO Importance: 2.9 SRO Importance: 3.2 K/A: 022K3.01 Comments: New question

RO Level Low Cog Level K/A: 022: Containment Cooling System, K3: Knowledge of the effect that a loss or malfunction of the CCS will have on the following, 01: Containment equipment subject to damage by high or low temperature, humidity and pressure Objective: S.VP1-13, Given a set of plant conditions, PREDICT the impact, on the plant and/or Containment ventilation system, of various failures, including instrument malfunctions, electrical supply failures and component malfunctions



42 ID: 2017 NRC Q42A Points: 1.00 The following valves are CLOSED • 1SI8812A, RH Pump RWST Suction Isolation Valve • 1CS001A, CS Pump RWST Suction Valve Concerning Main Control Board Valve Interlocks: What additional valve, if any, must be CLOSED to allow 1SI8811A, Containment Recirc Sump Outlet Isolation Valve, to be manually opened?

A. NO additional valve must be closed.

B. 1CV8804A, RH HX to U-1 CV Pumps Suction Header Isolation Valve.

C. 1RH8701A, RH Pump Suction from 1A HL DWST Isolation Valve.

D. 1RH8702A, RH Pump Suction from 1C HL DWST Isolation Valve.

Answer: C

Answer Explanation A. Incorrect. See below. Plausible because 1SI8812A and 1CS001A are the two RWST flowpath isolations for ‘A’ train, and 1SI8811A is the sump flowpath isolation. It is possible for the examinee to assume that with both RWST flowpaths isolated, the sump flowpath could be unisolated. B. Incorrect. See below. Plausible because 1CV8804A and 1SI8811A closed is the interlock for opening 1SI8812A. It is possible for the examinee to conclude that the same valve combination would be used for opening any of the three valves. C. Correct. Per 1BGP 100-1A2, MAIN CONTROL BOARD VALVE INTERLOCKS, to manually open 1SI8811A, the following interlocks need to be met: 1SI8812A closed and 1RH8701A or 1RH8701B closed and 1CS001A closed. Also shown on drawing 6E-1-4030SI14. D. Incorrect. See above. Plausible because this valve is sometimes believed to be an ‘A’ train valve because of the ‘A’ in its EPN designator. Meets K/A, Requires examinee to have the knowledge of the physical connections and/or cause effect relationships between CS and ECCS. The valve interlocks are there due to the common suction sources for the two systems during both the injection and recirculation phases. Technical References: Drawing 6E-1-4030SI14, 1BGP 100-1A2, MAIN CONTROL BOARD VALVE INTERLOCKS




System ID: 1744277 User-Defined ID: 2017 NRC Q42A Cross Reference Number: S.RH1-06-C

Topic: The following valves are CLOSED 1SI8812A, RH Pump RWST Suction Isolation Valve 1CS001A, CS Pump

RO Importance: 4.2 SRO Importance: 4.2 K/A: 026K1.01 Comments: <QQ 1138040(1412)><<New Question

License Level: RO Cognitive Level: Low, from memory 026 Containment Spray System (CSS) K1 Knowledge of the physical connections and/or cause effect relationships between the CSS and the following systems: K1.01 ECCS S.RH1-06-C, DESCRIBE all interlocks with their setpoints and EXPLAIN the purpose of each for the following: 1MOV-SI8811A & B



43 ID: 2017 NRC Q43A Points: 1.00 At 1000 Unit 1 is at 100% power. • The 1B Containment Spray pump is Out of Service. A Loss of Coolant Accident occurred. The crew has transitioned to 1BEP-1, LOSS OF REACTOR OR SECONDARY COOLANT. At 1015, conditions worsened and currently: • NO Containment Spray pumps are running. • Containment Pressure is 22 psig. 1. Which CLOSED valve is preventing the 1A CS pump from starting?

2. What procedure will be used to INITIATE Containment Spray?

A. 1. 1CS019A, CS EDUCTOR SPRAY ADDITIVE VALVE.

2. 1BEP-0, Attachment C, MANUAL CS ACTUATION.

B. 1. 1CS019A, CS EDUCTOR SPRAY ADDITIVE VALVE. 2. 1BFR Z.1, RESPONSE TO HIGH CONTAINMENT PRESSURE.

C. 1. 1CS010A, CS EDUCTOR INLET FLOW CONTROL VALVE.

2. 1BEP-0, Attachment C, MANUAL CS ACTUATION.

D. 1. 1CS010A, CS EDUCTOR INLET FLOW CONTROL VALVE. 2. 1BFR Z.1, RESPONSE TO HIGH CONTAINMENT PRESSURE.

Answer: B

Answer Explanation



Per drawing 6E-1-4030 CS01, CS pump 1A, If the pump control switch is in the "after trip" position the only valve contact in line with the breaker spring release (close) coil is 1CS019A. Since the crew has left BEP 0, that procedure no longer applies to the situation, and BFR Z.1 would be entered to start the CS pump. A. Incorrect: The first part is correct, making this distractor plausible. The second part is incorrect

but plausible as Attachment C provides actions to manually start the CS pump. B. Correct: As explained above. C. Incorrect: The first part is incorrect but plausible as this valve is verified in the proper position

in either of the two procedures mentioned but it does not provide an interlock feature for a pump start. The procedure has the operator check the eductor inlet flow control valve open so that sodium hydroxide is available. The second part is incorrect but plausible as Attachment C provides actions to manually start the CS pump.

D. Incorrect: The first part is incorrect but plausible as this valve is verified in the proper position

in either of the two procedures mentioned but it does not provide an interlock feature for a pump start. The procedure has the operator check the eductor inlet flow control valve open so that sodium hydroxide is available. The second part is correct making this distractor plausible.

Reference: 6E-1-4030 CS01 1BEP 0 1BFR-Z.1, RESPONSE TO HIGH CONTAINMENT PRESSURE This question meets the K/A by testing the candidate on interlocks associated with the auto start of the Containment Spray pumps by placing the candidate in a situation in which the requirements are met to enter an emergency procedure, in this case the Functional Restoration procedure for High Containment Pressure (BFR-Z.1), resulting from an interlock preventing the CS pump start.




System ID: 2029207 User-Defined ID: 2017 NRC Q43A Cross Reference Number: T.FR07-07-E

Topic: Unit 1 is at 100% power. The 1B Containment Spray pump is Out of Service.

RO Importance: 4.5 SRO Importance: 4.6 K/A: 026G4.2 Comments: New Question

RO Level High Cog Level, multiple analysis K/A: 026: Containment Spray System, G4: Emergency Procedures/Plan, .2: Knowledge of system setpoints, interlocks and automatic actions associated with EOP Entry Conditions Objective: Given a set of plant conditions, diagnose and analyze the containment status tree



44 ID: 2017 NRC Q44A Points: 1.00 <QQ 1138123(1410)><<Unit 1 started up 2 weeks ago following a refueling outage. • The reactor is at 97% power. • Rods are in manual. • A steam leak develops on the 1A S/G. 1. Reactor power will…

and

2. RCS Tave will…

A. 1. lower.

2. lower.

B. 1. lower. 2. rise.

C. 1. rise.

2. lower.

D. 1. rise. 2. rise.

Answer: C

Answer Explanation



Given the initial conditions in the question stem, the additional steam flow from the steam leak would result in a lower RCS temperature, and this would insert positive reactivity, and as a result reactor power would rise. A positive MTC is plausible as the Tech. Specs, specifically LCO 3.1.3 on MTC, allows operation with a positive MTC up to 100% power. This would occur very early in core life. After 2 weeks of almost full power operation it is expected that MTC will be negative due to the rapid boron dilution caused by the fission product poison buildup. A. Incorrect. The first part is incorrect but plausible because early in core life the Moderator

Temperature Coefficient is the least negative, and at lower reactor power levels could be positive. With a positive MTC as RCS temperature lowers reactor power lowers. The second part is correct, as increased steam demand will remove more energy from the RCS causing temperature to lower. The second part being correct makes this part of the distractor plausible.

B. Incorrect. The first part is incorrect but plausible because early in core life the Moderator

Temperature Coefficient is the least negative and at lower reactor power levels could be positive. With a positive MTC as RCS temperature lowers reactor power lowers. A steam leak would cause more energy removal from the RCS which would then lower reactor power due to the decreased temperature. The second part is incorrect but plausible because as power is ramped down steam generator pressure rises. The rising S/G pressure causes S/G steam temperature to rise.

C. Correct. See above. D. Incorrect. The first part is correct which makes this part of the distractor plausible. The second

part is incorrect but plausible because the question stem states that the reactor is in a condition shortly after a refueling. In this condition, with a high boron concentration, the Moderator Temperature Coefficient is the least negative and at lower reactor power levels could be positive. With a positive MTC, as RCS temperature rises reactor power rises.

Meets K/A, examinee must understand the operational implications of more steam removal on reactivity. The effects are opposite for positive and negative MTC. As discussed in SOER 07-01, the safety significance of this is high enough to require training on it yearly. Technical References: Lesson Plan I1-MS-XL-01 Chapter 23 T.S. 3.1.3, MTC



Question 44 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 0 Difficulty: 0.00 System ID: 1997312 User-Defined ID: 2017 NRC Q44A Cross Reference Number: S.MS1-11

Topic: Unit 1 started up 2 weeks ago following a refueling outage. The reactor is at 97% power.

RO Importance: 3.6 SRO Importance: 3.6 K/A: 039K5.08 Comments: <QQ 1138040(1412)><<Bank Question from BWD 2014 NRC

Exam Q 25 (ID: RS1039-N14-25) License Level: RO Cognitive Level: Higher, Multiple analyses are required to answer question. 039 Main and Reheat Steam System (MRSS) K5.08 Knowledge of the operational implications of the following concepts as the apply to the MRSS: Effect of steam removal on reactivity Objective: S.MS1-11. DESCRIBE the relationship between reactor power, steam pressure and impulse pressure.



45 ID: 2017 NRC Q45A Points: 1.00 The Unit has TRIPPED from 100% power. • Reactor Trip Breaker B did NOT open. Tave will be controlled at ___°F.

A. 550

B. 557

C. 560

D. 564

Answer: C

Answer Explanation The steam dumps will control on the Load Rejection controller at 560°F because RTB B did not open. This breaker (RTB B) opening would cause the RCS to be controlled at 557°F. A. Incorrect: See above. This is plausible as this is the P-12 setpoint which would cause all steam dumps to close at 550°F. The dumps would then open when temperature went above 550°F. Thus, the dumps would control at 550°F. B. Incorrect: See above. This is plausible as this is the temperature that the Steam Dumps normally control at on a reactor trip. C. Correct as described above. D. Incorrect: See above. This is the temperature at which a Feedwater Isolation occurs following a reactor trip, making this distractor plausible. This question meets the K/A by testing the candidate on what RCS Tave will be controlled at due to a malfunction of one of the reactor trip breakers (interlock). The distractors include setpoints of RCS Tave limits of when they would close on low temperature and what temperature the RCS would be controlled at if the dumps did not open as required. Reference: Steam Dump Lesson Plan (Chapter 27)




System ID: 1799197 User-Defined ID: 2017 NRC Q45A Cross Reference Number: S.DU1-11

Topic:

The Unit has TRIPPED. Reactor Trip Breaker B did NOT open. Tave will be controlled at ___°F.


License Level: RO Cognitive Level: High, multiple analysis required. 039 Main and Reheat Steam System (MRSS) K4 Knowledge of MRSS design feature(s) and/or interlock(s) which provide for the following: K4.02 Utilization of Tave. program control when steam dumping through atmospheric relief/dump valves, including Tave. limits. Objective: S.DU1-11, Given an operating mode and/or various plant conditions, PREDICT how the steam dump system and/or supported systems will be impacted by various steam dump instrumentation, control circuit, or electrical power failures, without the use of references.



46 ID: 2017 NRC Q46 Points: 1.00 The controlling level channel for 1A Steam Generator fails HIGH. Per BHC 1-SG, STEAM GENERATOR LEVEL, the NSO should FIRST ...

A. allow the Feed Regulating valve to respond in AUTO.

B. swap the 1A Steam Generator level control to the other channel.

C. take manual control of the Feed Regulating valve and RAISE the demand to OPEN the

valve.

D. take manual control of the Feed Regulating valve and LOWER the demand to CLOSE the valve.

Answer: C

Answer Explanation A. Incorrect. The SGWLC System has inputs from both level and flow mismatch. But, it is level dominate and the system would drive the Feed Regulating valve closed. Plausible because there would be a mismatch between steam and feedwater flows that would open the Feed Regulating valve if there were no level mismatch. B. Incorrect. BHC 1-SG, STEAM GENERATOR LEVEL, requires the NSO to take manual control of the Feed Regulating valve and open it. Plausible because swapping controlling level channels would reopen the Feed Regulating valve. C. Correct. The SGWLC System has inputs from both level and flow mismatch. But, it is level dominate, and the system would drive the Feed Regulating valve closed. BHC 1-SG, STEAM GENERATOR LEVEL, requires the NSO to take manual control of the Feed Regulating valve and open it. D. Incorrect. The SGWLC System has inputs from both level and flow mismatch. But, it is level dominate and the system would drive the Feed Regulating valve closed. Plausible because taking manual control of the Feed Regulating and lowering the demand would close the valve. Meets K/A, examinee must have the ability to manually operate and monitor the Feed regulating valve controllers in the control room. Technical References: BHC 1-SG, STEAM GENERATOR LEVEL. OP-BY-101-0004, STRATEGIES FOR SUCCESSFUL TRANSIENT MITIGATION




System ID: 1742920 User-Defined ID: 2017 NRC Q46 Cross Reference Number: S.FW2-16

Topic: The controlling level channel for 1A Steam Generator fails HIGH. Per BHC 1-SG, STEAM GENERATOR


License Level: RO Cognitive Level: Low, from memory. 059 Main Feedwater (MFW) System A4 Ability to manually operate and monitor in the control room: A4.08 Feed regulating valve controller Objective: S.FW2-16, Given a set of plant conditions, ANALYZE these conditions and DETERMINE how they are affected by any SGWLC System instrumentation failure.



47 ID: 2017 NRC Q47 Points: 1.00 Unit 1 was manually TRIPPED from 100% power due to a Steam Generator 1A Tube Rupture. The crew is in 1BEP-3, STEAM GENERATOR TUBE RUPTURE, and Step 4, Check Rupture Steam Generator Level is in progress. • 1A Steam Generator Level is 17% and RISING. • Containment pressure is 0.3 psig and STABLE.

Auxiliary Feedwater (AF) flow ___ (1) ___ be ISOLATED to the Ruptured S/G.

If isolation is required per 1BEP-3, AF flow would be isolated by _________ (2) _________.

A. 1. will NOT

2. CLOSING S/G 1A ISOL VLVs (1AF013A, E) ONLY

B. 1. will NOT 2. CLOSING S/G 1A ISOL VLVs (1AF013A, E) AND SETTING the control pot to ZERO

for the S/G 1A FLOW CONT VLVs (1AF005A, E)

C. 1. will 2. CLOSING S/G 1A ISOL VLVs (1AF013A, E) ONLY

D. 1. will

2. CLOSING S/G 1A ISOL VLVs (1AF013A, E) AND SETTING the control pot to ZERO for the S/G 1A FLOW CONT VLVs (1AF005A, E)

Answer: D

Answer Explanation



1BEP-3 step 4.a directs the Operator to isolate AF flow to the ruptured S/G if NR level is greater than 10% (31% Adverse Containment). Isolation of AF flow to the S/G is accomplished by closing the AF Flow Control valves (AF005s) AND the S/G AF Isolation valves (AF013s). This is stated in step 4.b. The AF005 valves are in a series arrangement with the AF013s. A. Incorrect: The first part is incorrect because BEP-3 step 4 requires S/G levels to >31% prior to

isolating AF when the containment is in an adverse containment condition. The adverse numbers are sometimes mixed up with the normal numbers by students making the first part of the distractor plausible. The second part is incorrect but plausible as the AF Isolation valves are closed to isolate feedwater flow to the S/G: however, these are not the only valves closed per 1BEP-3 to isolate flow.

B. Incorrect: The first part is incorrect because BEP-3 step 4 requires S/G levels to >31% prior to

isolating AF when the containment is in an adverse containment condition. The adverse numbers are sometimes mixed up with the normal numbers by students making the first part of the distractor plausible. The second part is correct making this part of the distractor plausible.

C. Incorrect: The first part is correct making this part of the distractor plausible. The second part

is incorrect but plausible as the AF Isolation valves are closed to isolate feedwater flow to the S/G: however, these are not the only valves closed per 1BEP-3 to isolate flow.

D. Correct as described above. The question meets the K/A by placing the candidate in a situation which they must (or will) operate the AF system controls to minimize S/G level rise to prevent a potential radiological release to the environment. Reference: 1BEP-3, STEAM GENERATOR TUBE RUPTURE




System ID: 1742912 User-Defined ID: 2017 NRC Q47 Cross Reference Number: T.EP04-08

Topic: Unit 1 was manually TRIPPED from 100% power due to a Steam Generator 1A Tube Rupture.


RO Level High Cog Level due to analysis K/A: 061: Auxiliary/Emergency Feedwater (AFW) System, A1: Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the AFW controls including, .01: S/G level Objective: ANALYZE a set of plant conditions and DETERMINE the required operator actions in response to a SGTR.



48 ID: 2017 NRC Q48 Points: 1.00 Unit 1 is at 15% power. A fault on 6.9KV Bus 159 causes a bus lockout. RCP 1D breaker will trip on _____(1)_____, and the reactor will ____(2)____ automatically.

A. 1. undervoltage

2. trip

B. 1. undervoltage 2. NOT trip

C. 1. under frequency

2. trip

D. 1. under frequency 2. NOT trip

Answer: B

Answer Explanation A. Incorrect. An UV on one RCP bus will trip its respective RCP breaker, but it also it takes an UV

on 2/4 RCP buses or 2/4 loop loss of flow for a reactor trip. Plausible because loss of one RCP at higher power levels will result in a reactor trip.

B. Correct. An UV on one RCP bus will trip its respective RCP breaker, and it takes an UV on 2/4

RCP buses or 2/4 loop loss of flow for a reactor trip. C. Incorrect. The UF relays require power to actuate. Also, it takes an UV on 2/4 RCP buses or

2/4 loop loss of flow for a reactor trip. Plausible because RCP bus low frequency will result in a RCP breaker trip as well as reactor trip at higher power levels.

D. Incorrect. The UF relays require power to actuate. Also, it takes an UV on 2/4 RCP buses or

2/4 loop loss of flow for a reactor trip. Plausible because RCP bus low frequency will result in a RCP breaker trip.

Meets K/A, requires examinee to know the 6.9KV bus load breaker automatic trips. Technical References: 1BEP-0, REACTOR TRIP OR SAFETY INJECTION, SYMPTOMS OR ENTRY CONDITIONS.




System ID: 1742094 User-Defined ID: 2017 NRC Q48 Cross Reference Number: S.AP1-10-A

Topic:

Unit 1 is at 15% power. A fault on 6.9KV Bus 159 causes a bus lockout. RCP 1D breaker will

RO Importance: 2.5 SRO Importance: 2.7 K/A: 062K4.02 Comments: 2017 NRC Q48

ILT Exam Bank (419336) License Level: RO Cognitive Level: High, multiple analysis required. 062 A.C. Electrical Distribution K4 Knowledge of ac distribution system design feature(s) and/or interlock(s) which provide for the following: K4.02 Circuit breaker automatic trips Objective: S.AP1-10-A, EVALUATE the response of the 6.9 KV and 4 KV buses for the following condition: Bus undervoltage



49 ID: 2017 NRC Q49B Points: 1.00 Unit 1 is at 100% power. • Bus 143 is deenergized due to a bus fault. What is the effect on unit operations due to the loss of Bus 143 and what action is required?

A. The reactor trips due to Reactor Coolant Flow Low and 1BEP-0, REACTOR TRIP OR

SAFETY INJECTION, will be entered.

B. The Main Turbine will be run back due to loss of Heater Drain Pump per 1BOA SEC-1, SECONDARY PUMP TRIP.

C. The Main Turbine power will be lowered due to reduced Condenser Cooling Water flow

per 1BOA SEC-3, LOSS OF CONDENSER VACUUM.

D. Both Units will enter LCO based on loss of Control Room Chiller per 0BOL 7.11, CONTROL ROOM VENTILATION (VC) TEMPERATURE CONTROL SYSTEM LCOAR.

Answer: C

Answer Explanation Bus 143 (a medium voltage bus) feeds A train Non-ESF 4160 volt loads. These loads include a containment chiller, 2 CW pumps and a Station Air compressor along with Non-ESF 480 volt Unit Sub-Stations. A. Incorrect: This is plausible because the Reactor Coolant Pumps are powered from Medium

voltage busses. A loss of any Non-ESF 6.9 KV bus will result in a Low Flow Reactor Trip. Also plausible because 1BEP-0, REACTOR TRIP OR SAFETY INJECTION, would be entered for the reactor trip.

B. Incorrect: This is plausible because the Heater Drain Pumps are powered from Medium

voltage busses. Also plausible because a loss of bus 156 or 157 (Non-ESF) will result in a loss of at least 1 HD pump and the procedure used for this situation is 1BOA SEC-1, SECONDARY PUMP TRIP.

C. Correct: Bus 143 powers 2 CW pumps. If 2/3 of circ water flow is lost at full power a runback

will be the correct action. 1BOA SEC-3, LOSS OF CONDENSER VACUUM, will direct the runback.

D. Incorrect: The control room chillers being affected is plausible as it is 4160 volt load. Being at

the correct voltage makes this distractor plausible, however this is a safety related component and bus 143 is not. Also plausible because 0BOL 7.11, CONTROL ROOM VENTILATION (VC) TEMPERATURE CONTROL SYSTEM LCOAR is applicable for the control room chillers being inoperable.

Meets K/A, Examinee must predict the loss of bus 143 and know what procedure to use to mitigate the consequences of the loss of bus 143. Technical References: 1BEP-0, REACTOR TRIP OR SAFETY INJECTION, 1BOA SEC-1, SECONDARY PUMP TRIP, 1BOA SEC-3, LOSS OF CONDENSER VACUUM and 0BOL 7.11, CONTROL ROOM VENTILATION (VC) TEMPERATURE CONTROL SYSTEM




System ID: 1952014 User-Defined ID: 2017 NRC Q49B Cross Reference Number: S.AP1-12.B

Topic:

Unit 1 is at 100% power. Bus 143 is deenergized due to a bus fault. What is the effect on unit


RO level High Cog Level- due to performing more than 1 mental manipulation K/A: 062: AC Electrical Distribution System, A2: Ability to (a) predict the impacts of the following malfunctions or operations on the ac distribution system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations, 04: Effect on plant of de-energizing a bus Objective: List the loads supplied by the following busses, 141, 142, 143, 144, 241, 242, 243, 244.



50 ID: 2017 NRC Q50C Points: 1.00 A loss of all AC has occurred on Unit 1. • 111 Battery Charger DC output breaker has OPENED. • DC 111/113 bus voltage is indicating 128 VDC and is slowly LOWERING. Which of the following DC BUS LOAD breakers, IF OPENED, will extend Battery 111 capacity the most?

A. Control power Main- Bus 156

B. Main Turbine- DC Emergency Oil Pump

C. Instrument Inverter 113 DC Input

D. Control power Main- Bus 143

Answer: C

Answer Explanation A. Incorrect. See below. This distractor is plausible because Control power Main- Bus 156 is a DC load; however, it is "B" Train equipment. The question stem is addressing "A" train loads. B. Incorrect. See below. This distractor is plausible because this is a large DC load when the pump is running; however, it is a DC Bus 123 load. C. Correct. Battery capacity will be extended by decreasing the current load on the battery. Reducing the current load on either DC Bus 111 or 113 will reduce the 111 Battery discharge rate. Instrument Inverter 113 is a load fed from DC Bus 111. The inverter is normally fed by AC and DC is the backup; with a loss of all AC, Instrument Inverter 113 is a larger load on DC Bus 111 than Control power to Bus 143. D. Incorrect. See above. Plausible because Control power Main- Bus 143 is a load on DC Bus 111 and Instrument Inverter 113 is normally feed by AC. Reference: DC loads are contained in 1BCA-0.0 Table A Question meets the K/A by having the examinee predict how discharge rate on the battery will change when the DC input breaker is opened for an Instrument Inverter under specific plant conditions and compare that to a smaller load.




System ID: 2029198 User-Defined ID: 2017 NRC Q50C Cross Reference Number: S.DC1-04-B

Topic: 111 Battery Charger DC output breaker has OPENED. DC 111 bus voltage is indicating 128 VDC and is


RO Level High Cog Level due to analysis 063 DC Electrical Distribution System A1 Ability to predict and/or monitor changes in parameters associated with operating the DC electrical system controls including: A1.01 Battery capacity as it is affected by discharge rate Objective: S.DC1-04-B, DISCUSS the function of the following major components for the 125 VDC System: Battery



51 ID: 2017 NRC Q51 Points: 1.00 One air receiver is Out of Service for the 1A Emergency Diesel Generator (DG). • The other receiver ruptures. Both 1A air receivers are at 0 psig. If the 1A DG is in standby, the DG ___ (1) ___ be started.

If the 1A DG were running, an ___ (2) ___ condition would trip the DG.

A. 1. can

2. engine oil pressure low

B. 1. can 2. overspeed

C. 1. can NOT

2. engine oil pressure low

D. 1. can NOT 2. overspeed

Answer: D

Answer Explanation A. Incorrect. See below. Plausible because there is a volume bottle in the starting air system which contains an additional volume of air. This volume is used to bypass the engine oil pressure low trip on a start. A possible misconception is that this extra volume of air would supply both the Air Start solenoids and the engine oil pressure low trip if the DG were running. B. Incorrect. See below. Plausible because there is a volume bottle in the starting air system which contains an additional volume of air. This volume is used to bypass the engine oil pressure low trip on a start. A common misconception is that this extra volume of air would supply both the Air Start solenoids and the engine oil pressure low trip if the DG were running. C. Incorrect. See below. Plausible because there is a volume bottle in the starting air system which contains an additional volume of air. This volume is used to bypass the engine oil pressure low trip on a start. A possible misconception is that this extra volume of air would supply the engine oil pressure low trip if the DG were running. D. Correct. The D/G requires air to the Air Start solenoid valves and accompanying cylinders to provide the motive force to "roll" the engine (start). The air receivers also supply air to the pneumatic protection system to force the BIMBA cylinder to the "fuel off" position to stop the engine. Without air the machine will not start if it is stopped. If it were running it would not trip with the exception of the overspeed butterfly valve closing to starve the machine of combustion air. The volume bottle in the starting air system is used to bypass two trips during a DG start. The volume bottle depressurizes after the start. Meets K/A, examinee must know the effects of loss of air receivers on the DG. Technical References: Lesson Plan I1-DG-XL-01.




System ID: 1741912 User-Defined ID: 2017 NRC Q51 Cross Reference Number: S.DG1-02-F

Topic: One air reciever is Out of Service for the 1A Emergency Diesel Generator (DG). The other receiver

RO Importance: 2.7 SRO Importance: 2.9 K/A: 064K6.07 Comments: <QQ 1138040(1412)><<Modified Bank Question 2016 CERT

EXAM Q59 (1144551) License Level: RO Cognitive Level: Low, from memory. 064 Emergency Diesel Generator (ED/G) System K6 Knowledge of the effect of a loss or malfunction of the following will have on the ED/G system: K6.07 Air receivers Objective: S.DG1-02-F, DISCUSS the function and operation of the following Diesel Generator Auxiliary Support System: Pneumatic Protection



52 ID: 2017 NRC Q52 Points: 1.00 A liquid release is in progress from Release Tank 0WX01T. 0PR01J, LIQUID RADWASTE EFFLUENT RAD MONITOR, has ALARMED on the RMS with a DARK BLUE color. 1. What monitor condition does the color represent?

2. The Liquid Release should be ...

A. 1. Equipment Failure

2. terminated.

B. 1. Operate Failure 2. terminated.

C. 1. Equipment Failure

2. continued.

D. 1. Operate Failure 2. continued.

Answer: B

Answer Explanation Per BOP AR/PR-11, the Operate Failure is designated by the Dark Blue color, and the Equipment Failure is designated by the Cyan color (light blue). Per BOP AR/PR-11, alarm responses for Operate Failures, ensure that the interlock function has occurred on monitors so equipped. For equipment failures, no automatic actions occur. A. Incorrect: The first part is incorrect but plausible as this is an RMS color and function. The second part is correct which makes this part of the distractor plausible. B. Correct as described above. C. Incorrect: The first part is incorrect but plausible as this is an RMS color and function. The second part is incorrect but plausible as not all RMS alarms for a monitor result in an interlock function. D. Incorrect: The first part is correct which makes this distractor plausible. The second part is incorrect but plausible as not all RMS alarms for a monitor result in an interlock function. References: BOP AR/PR-11A6 BOP AR/PR-11A20 The question meets the K/A as it places the candidate in a situation where an liquid effluent release is occurring and queries about the rad monitor equipment in the control room where the release can be monitored from.




System ID: 1744178 User-Defined ID: 2017 NRC Q52 Cross Reference Number: S.AR1-09

Topic: A liquid release is in progress from Release Tank 0WX01T. 0PR01J, LIQUID RADWASTE EFFLUENT RAD

RO Importance: 3.9 SRO Importance: 3.9 K/A: 073.A4.01 Comments: Bank Question (from 2013 Byron NRC Exam question 50)

RO Level Low Cog Level K/A: 073: Process Radiation Monitoring (PRM) System, A4: Ability to manually operate and or monitor in the control room, .01: Effluent release Objective: IDENTIFY the condition of a particular AR/PR Monitor channel given its color code (S.AR1-09)



53 ID: 2017 NRC Q53 Points: 1.00 Which of the following describes an actuation that occurs DIRECTLY as the result of an SI signal ONLY?

A. SX supply valves to BOTH Diesel Generators OPEN. (_SX169A/B)

B. CNMT Chiller Condenser Bypass valves OPEN. (_SX147A/B)

C. AF Pump SX Suction valves OPEN. (_AF006A/B and _AF017A/B)

D. RCFC SX Outlet Isolation valves OPEN. (_SX027A/B)

Answer: B

Answer Explanation A. Incorrect. DG valves open after the DG starts (280 rpm). Plausible because the DG receives an SI signal to start and then the SX valves would open. B. Correct. SI signal directly opens RCFC Bypass valves to ensure SX flowpath through the RCFCs. C. Incorrect. The AF suction valves open with an SI signal and low AF suction pressure. Plausible because the AF pumps receive an SI signal to start, and the SX suction valves can supply a flowpath for the pumps. D. Incorrect. The RCFC SX Outlet Isolation valves open after a start of the RCFC. Plausible because the RCFC would start with an SI signal, and after the start, the outlet valves would open. Meets K/A, examinee must know which SX valves will reposition with an SI signal. Technical References: BAR 1-3-E7, AF PUMP SX SUCT VLVS ARMED. BAR 1PL07J-1-C2, ESS SERVICE WTR FLOW LOW. BOP VP-5, REACTOR CONTAINMENT FAN COOLER STARTUP. Integrated Systems Training/IST-12.




System ID: 1743023 User-Defined ID: 2017 NRC Q53 Cross Reference Number: S.SX1-07-A

Topic: Which of the following describes an actuation that occurs DIRECTLY as the result of an SI signal ON

RO Importance: 3.6 SRO Importance: 3.8 K/A: 076K1.16 Comments: Modified LORT Bank question (1512249)

License Level: RO Cognitive Level: Low, from memory. 076 Service Water System (SWS) K1 Knowledge of the physical connections and/or cause- effect relationships between the SWS and the following systems: K1.16 ESF Objective: S.SX1-07-A, STATE the effect of the following on the Essential Service Water flowpath (valve position changes): Safety Injection.



54 ID: 2017 NRC Q54A Points: 1.00 The MCR received alarms for 0/1/2 SA Receiver pressures low and 0/1/2 IA Receiver pressures low. • SA header pressure LOWERED to 107 psig and is STABLE. • IA header pressure is 88 psig and slowly DROPPING. Per 0BOA SEC-4, LOSS OF INSTRUMENT AIR- UNIT 0, the crew should?

A. Locally align the standby IA Dryer.

B. Verify the standby IA Dryer has automatically aligned.

C. Verify the IA dryer has automatically isolated for any IA Dryer with a failed purge valve.

D. Verify the standby Prefilter and/or Afterfilter has automatically realigned for high ΔP on

the filters.

Answer: A

Answer Explanation A. Correct. 0BOA SEC-4 directs the operator to align the standby IA Dryer when IA header pressure is less than 90 psig. This is done by locally opening valves. B. Incorrect. See above. Plausible because 0BOA SEC-4 does directly align the standby IA Dryer. C. Incorrect. See above. 0BOA SEC-4 directs the operator to locally isolate the affected IA Dryer. Plausible because 0BOA SEC-4 does direct isolation of the affected IA Dryer. D. Incorrect. See above. 0BOA SEC-4 directs the operator to locally align the standby filters when the online filters have a high ΔP. Plausible because 0BOA SEC-4 does direct aligning the standby filters. Meets K/A, examinee must evaluate the conditions and determine the manual actions needed, transfer to the standby IA Dryer. RO Knowledge: From the ES-401 Attachment 2 Figure 2: Screening for SRO-only linked to 10CFR55.43(b)(5) (Assessment and selection of procedures): 1) assessing plant conditions and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. Using the flowchart, this question can: • NOT be answered solely by knowing “systems knowledge”, i.e., how the system works, flowpath, logic, component location. Yes • NOT be answered solely by knowing immediate operator actions. No • NOT be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs. No • NOT be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure. Yes, can be answered by knowing the overall mitigative strategy of 0BOA SEC-4. Technical References: 0BOA SEC-4, LOSS OF INSTRUMENT AIR. BAR 0-38-C2, SA RCVR 0 PRESS LOW, BAR 0-37-C2 SA RCVR 1 PRESS LOW





Topic: The MCR received alarms for 0/1/2 SA Receiver pressures low and 0/1/2 IA Receiver pressures low.


License Level: RO Cognitive Level: Low, from memory. 078 Instrument Air System (IAS) K4 Knowledge of IAS design feature(s) and/or interlock(s) which provide for the following: K4.01 Manual/automatic transfers of control Objective: T.OA39-03, ANALYZE a given set of plant conditions and DETERMINE the required actions per 0/1/2BOA SEC-4, Loss of Instrument Air.



55 ID: 2017 NRC Q55 Points: 1.00 Which Bus powers the 1A CV pump?

A. 131X

B. 132X

C. 141

D. 142

Answer: C

Answer Explanation Bus 141 powers the 1A CV pump. A. Incorrect: This distractor is plausible as it is an ESF power supply to Train A equipment. Also,

plausibility is enhanced because the PDP is powered by 480 VAC. B. Incorrect: This distractor is plausible as it is an ESF power supply. Also, plausibility is

enhanced because the PDP is powered by 480 VAC. C. Correct: Bus 141 is an ESF power supply and is the correct train for the 1A CV pump. D. Incorrect: This distractor is plausible as it is ESF related and the correct voltage (4160). The question meets the K/A as it tests the candidate on the bus power supply to the 1A Charging pump. Reference: 6E-1-4030-CV01




System ID: 1742758 User-Defined ID: 2017 NRC Q55 Cross Reference Number: S.AP1-12-B

Topic: Which Bus powers the 1A CV pump?


RO Level Low Cog Level K/A: 004: Chemical and Volume Control System, K2: Knowledge of Bus power supplies to the following, .03: Charging pumps Objective: From the AC Electrical Power system Lesson Plan (S.AP1-12-B): LIST the loads supplied by the following buses: b. 141, 142 143, 144, 241, 242, 243, and 244 (S.AP1-12-B)



56 ID: 2017 NRC Q56 Points: 1.00 With an RCS average temperature RISE of 3°F, OPDT setpoint will___(1)___ and OTDT setpoint will___(2)___.

A. 1. rise

2. rise

B. 1. rise 2. lower

C. 1. lower

2. rise

D. 1. lower 2. lower

Answer: D

Answer Explanation A. Incorrect. See below. Plausible because both OPDT and OTDT setpoints change with changing RCS average temperature. B. Incorrect. See below. Plausible because both OPDT and OTDT setpoints change with changing RCS average temperature. C. Incorrect. See below. Plausible because both OPDT and OTDT setpoints change with changing RCS average temperature. D. Correct. The Over-temperature Delta T trip Function is provided to ensure that the design limit DNBR is met. The Over-temperature Delta T trip Function uses each loop's Delta T as a measure of reactor power and is compared with a setpoint that is automatically varied with reactor coolant average temperature - the Trip Setpoint is varied to correct for changes in coolant density and specific heat capacity with changes in coolant temperature. The Overpower Delta T trip Function ensures that protection is provided to ensure the integrity of the fuel. It uses the Delta T of each loop as a measure of reactor power with a setpoint that is automatically varied with reactor coolant average temperature - the Trip Setpoint is varied to correct for changes in coolant density and specific heat capacity with changes in coolant temperature. Both OTDT and OPDT will lower the trip setpoint as RCS Tave rises. Meets K/A, examinee must predict how the OPDT and OTDT setpoints will change with an RCS temperature change of 3°F. Technical References: Tech Spec - 3.3.1 Reactor Trip System (RTS) Instrumentation. B 3.3.1, Reactor Trip System (RTS) Instrumentation. CORE OPERATING LIMITS REPORT (COLR)




System ID: 1743062 User-Defined ID: 2017 NRC Q56 Cross Reference Number: T S.RP2-03-B

Topic: With an RCS average temperature RISE of 3°F, OPDT setpoint will___(1)___ and OTDT setpoint will___


License Level: RO Cognitive Level: Low, from memory. 002 Reactor Coolant System (RCS) A1 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the RCS controls including: A1.07 Reactor differential temperature Objective: T S.RP2-03-B, DISCUSS the relationship between the Safety Limits and the following: OP Delta T Reactor Trip.



57 ID: 2017 NRC Q57 Points: 1.00 Unit 1 is at 95% • Control Bank D is indicating 221 steps on the step counters and 222 steps on DRPI. The following annunciators are in alarm: 1-10-E6, ROD AT BOTTOM 1-10-D5, BANK D ROD STOP C-11 Rod D-4 has dropped into the core. DRPI will indicate ____(1)____steps for rod D-4.

Tech Spec ________ (2) ________ must be entered.

A. 1. 222

2. 3.1.4, Rod Group Alignment Limits

B. 1. 222 2. 3.1.7, Rod Position Indication

C. 1. 0

2. 3.1.4., Rod Group Alignment Limits

D. 1. 0 2. 3.1.7, Rod Position Indication

Answer: C

Answer Explanation



When the control rod drops, Control bank D rods will step out to recover Tave, which has dropped. DRPI will indicate 0 steps for the dropped rod. C-11 (Rod Stop) would have stopped rods at 223 steps. T.S. 3.1.4 would be entered for 1 rod not within alignment. A. Incorrect. DRPI indication for rod D-4 will be 0 steps while Control Bank D step counters would follow demanded position for the bank. C-11 (Rod Stop) would have stopped rods at 223 steps. Plausible because the other rods in Control Bank D would have stepped out to 223 steps and their DRPI indication would be 222. The second part is correct, making this part of the distractor plausible. B. Incorrect. DRPI indication for rod D-4 will be 0 steps while Control Bank D step counters would follow demanded position for the bank. C-11 (Rod Stop) would have stopped rods at 223 steps. Plausible because the other rods in Control Bank D would have stepped out to 223 steps, and their DRPI indication would be 222. The second part is incorrect but plausible because the candidate could evaluate the indications and believe there is a DRPI problem with DRPI indicating 0 steps and the Step Counters reading 221 steps for rod D-4. C. Correct. DRPI indication for rod D-4 will be 0 steps while Control Bank D step counters would follow demanded position for the bank. C-11 (Rod Stop) would have stopped rods at 223 steps. The second part is correct as the individual rod is not within alignment limits. D. Incorrect. See above. DRPI indication for rod D-4 will be 0 steps making this part of the distractor plausible. The second part is incorrect but plausible because the candidate could evaluate the indications and believe there is a DRPI problem with DRPI indicating 0 steps and the Step Counters reading 221 steps for rod D-4. Meets K/A, examinee must know that with a dropped rod DRPI will read 0 steps while the Step Counters will still read 221 steps. Technical References: Lesson Plan I1-PI-XL-01. Lesson Plan I1-RD-XL-01. 1BOA ROD-3, DROPPED OR MISALIGNED ROD




System ID: 1743142 User-Defined ID: 2017 NRC Q57 Cross Reference Number: S.PI1-06-B

Topic: Unit 1 is at 95% Control Bank D is indicating 221 steps on the step counters and 222 steps on DRPI


License Level: RO Cognitive Level: High, multiple analysis required. 014 Rod Position Indication System (RPIS) K5 Knowledge of the operational implications of the following concepts as they apply to the RPIS: K5.01 Reasons for differences between RPIS and step counter Objective: S.PI1-06-B, Given the following condition, PREDICT how the Rod Position Indication System will respond: Dropped Rod.



58 ID: 2017 NRC Q58B Points: 1.00 A reactor startup is in progress per 1BGP 100-2A1, REACTOR STARTUP. BOTH N35 and N36 indications have risen to 2.0E-5%. SR BLOCK PERMISSIVE P6 remains LIT. The NSO ___(1)___ BLOCK Source Range N31/N32 HI FLUX TRIP.

Based on current power indication, the Unit is ___(2)___.

A. 1. CAN

2. below the P-6 SETPOINT

B. 1. CAN 2. above the P-6 SETPOINT

C. 1. can NOT

2. below the P-6 SETPOINT

D. 1. can NOT 2. above the P-6 SETPOINT

Answer: D

Answer Explanation A. Incorrect. IR channels are at 2.0E-5%, P6 would normally be met. But the Bypass Permissive panel window is still lit, so P6 cannot be reset. Plausible because IR channels are greater than the P6 setpoint. B. Incorrect. With IR channels at 2.0E-5%, P6 would normally be met. But the Bypass Permissive panel window is still lit, so P6 cannot be reset. Plausible because IR channels are greater than the P6 setpoint. C. Incorrect. With IR channels at 2.0E-5%, P6 would normally be met. But the Bypass Permissive panel window is still lit, so P6 cannot be reset. Plausible because the Bypass Permissive panel window is still lit, so P6 cannot be reset. D. Correct. With IR channels at 2.0E-5%, P6 would normally be met. But the Bypass Permissive panel window is still lit, so P6 cannot be reset. The Unit is above the P6 setpoint. Meets K/A, examinee must recognize with the current indications the plant is above P-6, but there is a failure of the bistable as indicated by SR BLOCK PERMISSIVE P6 light being lit. Technical References: 1BGP 100-2A1, REACTOR STARTUP. BAR 1-BP-3.2, SR BLOCK PERMISSIVE P6.




System ID: 2029422 User-Defined ID: 2017 NRC Q58B Cross Reference Number: S.NI2-10

Topic: A Reactor startup is in progress per 1BGP 100-2A1, REACTOR STARTUP. N35 and N36 indication have


License Level: RO Cognitive Level: High, multiple analysis 015 Nuclear Instrumentation System K6 Knowledge of the effect of a loss or malfunction on the following will have on the NIS: K6.04 Bistables and logic circuits Objective: S.NI2-10, Given a set of plant conditions ANALYZE these conditions and DETERMINE how they are affected by any I.R. NI System instrumentation, control circuit, or electrical power failure without the use of references.



59 ID: 2017 NRC Q59A Points: 1.00

Refer to the above Control Room display. 1. The temperature that is displayed for each Train of CETCs is the average of ...

2. The core location for each thermocouple can be found ...

A. 1. ALL valid reading thermocouples in that Train.

2. by using the menus contained in the SELECT feature pushbutton.

B. 1. 10 HIGHEST valid reading thermocouples in that Train. 2. by using the menus contained in the SELECT feature pushbutton.

C. 1. ALL valid reading thermocouples in that Train.

2. in _BOSR 3.h.1-1 PDMS INSTRUMENTATION CHANNEL CHECKS.

D. 1. 10 HIGHEST valid reading thermocouples in that Train. 2. in _BOSR 3.h.1-1 PDMS INSTRUMENTATION CHANNEL CHECKS.

Answer: D

Answer Explanation



From page 9 of the Inadequate Core Cooling System lesson plan: MCB Displays 1) One display for each train (A and B). 2) Temp range 35o - 2300oF 3) Normal readout is average of 10 highest CET's of each train. 4) CETC/SELECT button: push once, readout shows each CET in that train, one at a time. Per BOP RC-12, PLACING RVLIS/HJTC/CETC IN SERVICE, the select button is used to perform functions on the CETC panel but the procedure directs the operator to BOP RC-12T5, Table 5- CETC Core Locations to find the specific core location for each thermocouple in that train. The locations can also be found in _BOSR 3.h.1-1 PDMS INSTRUMENTATION CHANNEL CHECKS, which is performed weekly A. Incorrect: The first part is plausible because the normal display for the CETCs shows an averaged number, but it does not indicate what is averaged. The second part is plausible because the SELECT button is used to interrogate the CETC monitor. Additionally, the procedure has the operator use the "select" pushbutton during the performance of system operations making this selection plausible. B. Incorrect: The first part is correct making this part of the distractor plausible. The second part is plausible because the SELECT button is used to interrogate the CETC monitor. Additionally, the procedure has the operator use the "select" pushbutton during the performance of system operations making this selection plausible. C. Incorrect: The first part is plausible because the normal display for the CETCs shows an averaged number, but it does not indicate what is averaged. The second part is correct making this part of the distractor plausible. D. Correct as described above. References: _BOSR 3.h.1-1 PDMS INSTRUMENTATION CHANNEL CHECKS Inadequate Core Cooling System lesson plan, Chapter 34b The question meets the K/A as it tests the candidate on design features of the ITM system (average of the 10 highest TCs), and questions the candidate on how to determine core locations of the thermocouples.




System ID: 2029230 User-Defined ID: 2017 NRC Q59A Cross Reference Number: S.IT1-05

Topic: Refer to the above Control Room display. The temperature that is displayed for each Train of


RO Level High Cog Level, multiple analysis K/A: 017, In-Core Temperature Monitoring System (ITM), K4: Knowledge of the ITM system design feature(s) and/or interlock(s) which provide for the following: Sensing and determination of location of core hot spots. Objective: Explain the principles of operation for the Core Exit Thermocouple System



60 ID: 2017 NRC Q60 Points: 1.00 Per Technical Specifications, the minimum required water level above the irradiated fuel assemblies in the Spent Fuel Pool is…

A. 24 feet

B. 23 feet

C. 17 feet

D. 6 feet

Answer: B

Answer Explanation A. Incorrect. See below. Plausible because the Spent Fuel Pool low level alarm is at 24 feet, and for many Tech Spec items, the alarm is also the Tech Spec entry condition. B. Correct. Per Technical Specification 3.7.14 Spent Fuel Pool Water Level minimum is 23 feet. C. Incorrect. See above. Plausible because the suction lines for Spent Fuel Pool Cooling are 7’ below normal water level. With the low level alarm at 24 feet, the examinee may conclude the Tech Spec limit is 17 feet. D. Incorrect. See above. Plausible because the discharge line stops 6’ above the top of the active fuel region to prevent dewatering the pool. This is a design feature, so the examinee may conclude the Tech Spec limit is 6 feet. Meets K/A, examinee must know Technical Specification 3.7.14 Spent Fuel Pool Water Level. RO level: From the ES-401 Attachment 2 Figure 1: Screening for SRO-only linked to 10 CFR 55.43(b)(2) (Tech Specs): • Can NOT be answered by knowing less than 1 hour Tech Specs. No • Can NOT be answered by knowing information listed "above-the-line". Yes • Can NOT be answered by knowing the TS Safety Limits. No Technical References: Technical Specification 3.7.14 Spent Fuel Pool Water Level. BAR 1-1-C1, SPENT FUEL PIT LEVEL HIGH LOW. BOP FC-11, SPENT FUEL POOL & REFUELING CAVITY LEVEL ADJUSTMENT. Lesson Plan I1-FC-XL-01 Chapter 51.




System ID: 1743168 User-Defined ID: 2017 NRC Q60 Cross Reference Number: S.FC1-08

Topic: Per Technical Specifications, the minimum required water level above the irradiated fuel assemblies

RO Importance: 3.6 SRO Importance: 4.5 K/A: 033G2.2.38 Comments: New question

License Level: RO Cognitive Level: Low, from memory. 033 Spent Fuel Pool Cooling System (SFPCS) G2.2.38 Knowledge of conditions and limitations in the facility license. Objective: S.FC1-08, Given a set of plant conditions, DETERMINE from memory, applicable Spent Fuel Pool Cooling and Cleanup System Tech Spec/TRM operability requirements.



61 ID: 2017 NRC Q61A Points: 1.00 Unit 1 has experienced a failure of 1LI-519, 1A SG level failed HIGH. • The crew has taken manual control of 1FW510, 1A SG Feed Reg valve and matched feedwater flow

with steam flow. • 1FW510, 1A SG Feed Reg valve is placed in AUTO. The following indications are observed after the transfer to AUTO.

1. Did the transfer of 1FW510 from manual to auto control occur as expected?

2. The NSO must ...

A. 1. Yes

2. leave 1FW510 controller in auto.

B. 1. Yes 2. take 1FW510 back to manual.

C. 1. No

2. leave 1FW510 controller in auto.

D. 1. No 2. take 1FW510 back to manual.

Answer: D

Answer Explanation



A. Incorrect. See below. Plausible because auto control on 1FW510 would normally maintain SG level and because the controller would be left in auto if the transfer were bumpless as in part 1. B. Incorrect. See below. Part 2 is plausible because FW flow indication is low and taking 1FW510 back to manual would allow the NSO to restore FW flow if it were actually low. C. Incorrect. See below. Plausible because auto control on 1FW510 would normally maintain SG level and because FW flow indication is low, and additional feed flow would be required to raise FW flow if it were actually low. D. Correct. When 1LI-519 failed high, the NSO would take 1FW510 to manual and stabilize the plant per BHC 1-SG, STEAM GENERATOR LEVEL and 1BOA INST-2, OPERATION WITH A FAILED INSTRUMENT CHANNEL. When SG level is stable, FW flow and Steam flow matched, 1FW510 would be taken to auto. A bumpless transfer of control from manual to auto would be indicated by FW flow and Steam flow continuing to be matched. With FW flow failed low, 1FW510 would need to be taken back to manual. Meets K/A, examinee must recognize that the indicated FW flow shows the transfer from manual to auto for SG level control has not been bumpless and must know what actions to take. Technical References: BHC 1-SG, STEAM GENERATOR LEVEL. 1BOA INST-2, OPERATION WITH A FAILED INSTRUMENT CHANNEL.


System ID: 1846466 User-Defined ID: 2017 NRC Q61A Cross Reference Number: S.FW2-16

Topic: Unit 1 has experienced a failure of 1LI-519, 1A SG level failed HIGH. The crew has taken manual


License Level: RO Cognitive Level: High, multiple analysis required. 035 Steam Generator System (S/GS) A4 Ability to manually operate and/or monitor in the control room: A4.01 Shift of S/G controls between manual and automatic control, by bumpless transfer. Objective: S.FW2-16, Given a set of plant conditions, ANALYZE these conditions and DETERMINE how they are affected by any SGWLC System instrumentation failure.



62 ID: 2017 NRC Q62A Points: 1.00 The Unit is in MODE 1, making preparations to roll the main turbine. • Reactor Power is 14%. • Steam Dumps are in STM PRESS mode and in AUTO. 1PT-507, S/G HDR PRESS, fails HIGH. The Steam Dump Valves will ___(1)___

and

S/G pressure will IMMEDIATELY ___(2)___

A. 1. OPEN

2. LOWER due to more heat removal from the RCS.

B. 1. OPEN 2. RISE due to the reactor adding more heat to the RCS.

C. 1. CLOSE

2. RISE due to less heat removal from the RCS.

D. 1. CLOSE 2. LOWER due to the reactor adding less heat to the RCS.

Answer: A

Answer Explanation



The question places the candidate in a situation in which the steam dumps would be in the Steam Pressure Mode. When the failure occurs, the sensed steam pressure is above setpoint and the steam dump valves will go full open. This will increase steam demand and lower S/G pressure. From page 7 of the Steam Dump Lesson Plan (Chapter 27) a. Steam Pressure Controller 1) This controller is active when the steam dump MODE SELECT switch is in the STM PRESS position. 2) Used for no load conditions (S/U & S/D) 4) Controlled from MCB _PM02J c. PT507 - Steam header pressure input From page 22 of the L-P: Concerning PT-507, 1) If failed high or low at power: a) Steam dumps will not be affected because the MS HDR PRESS controller is not active. 2) If failed high while in the STM PRESS mode: a) If RCS temperature is initially above P-12: (1) All steam dumps will open fully. The steam dumps will close when temperature drops to the P-12 setpoint (<550°F). 5) Compares setpoint pressure to steam header pressure (PT-507) A. Correct as described above. B. Incorrect. The first part is correct, which makes this part of the distractor plausible. The second part is plausible because RCS temperature will initially lower due to the increased steam demand, and reactor power rises because the RCS temperature change has added positive reactivity, transferring more heat to the RCS; the candidate may, therefore, conclude that SG pressure would rise as a result of greater heat transfer. C. Incorrect: The first part is incorrect but plausible because S/G pressure will initially lower due to the dumps opening, then based on the pressure being sensed (S/G versus Steam Header pressure), the candidate could reason the Steam Dumps should close. This would result in SG pressure rising with less heat removal from the RCS, making the second part of the distractor plausible. D. Incorrect: The first part is incorrect but plausible because S/G pressure will initially lower due to the dumps opening, then based on the pressure being sensed (S/G versus Steam Header pressure), the candidate could reason the Steam Dumps should close. The second part is incorrect but plausible because if the valves closed, reactor power would decrease and less heat would be transferred to the RCS; the candidate may, therefore, conclude that SG pressure would lower as a result. References: Steam Dump Lesson Plan, Chapter 27 BGP 100-3 Power Ascension, concerning Steam Dumps in steam pressure mode, (page 29). Question meets the K/A as it tests the candidate on the SDS that has an input sensor failure (steam pressure transmitter) and what the effect is on S/G pressure.




System ID: 1798991 User-Defined ID: 2017 NRC Q62A Cross Reference Number: S.DU1-04-D-3

Topic:

The Unit is in MODE 1, making preparations to roll the main turbine. Reactor Power is 14%. Steam


RO Level High Cog Level due to analysis K/A: 041: Steam Dump System (SDS) and Turbine Bypass Control, K3: Knowledge of the effect that a loss or malfunction of the SDS will have on the following, .01: S/G. Objective: EXPLAIN the operation of the Steam Dump System for each of the following: d. Failure of: 3) PT-507 (S.DU1-04-D-3)



63 ID: 2017 NRC Q63 Points: 1.00 Unit 1 is at 60% power. The following indications are observed:

What actions are required?

A. Place rods in MANUAL.

B. Allow rods to drive in AUTO.

C. Enter 1BEP-0, REACTOR TRIP OR SAFETY INJECTION.

D. Enter 1BOA TG-8, TURBINE TRIP BELOW P8.

Answer: C

Answer Explanation



A. Incorrect. See below. Plausible because. BAR 1-BP-4.4, TURBINE TRIP C8 directs placing rods in manual when reactor power is < 15%. B. Incorrect. See below. Plausible because BAR 1-BP-4.4, TURBINE TRIP C8, directs allowing rods to drive in Auto when reactor power is between 15% and P8. C. Correct. All 4 throttle valves fully closed or 2/3 EH header pressures <1000 psig on either train are the inputs for C8, Turbine Trip. Receiving C8 (Turbine Trip) while above P8 (indications show reactor power at 60%) will cause a reactor trip. Entering 1BEP-0, REACTOR TRIP OR SAFETY INJECTION would be required. D. Incorrect. See above. Plausible because the indications given in the stem of the question are for a turbine trip and that is one of the entry conditions for 1BOA TG-8, TURBINE TRIP BELOW P8. Meets K/A, examinee must know that a turbine trip above P8 will cause a reactor trip. Technical References: BAR 1-BP-4.4, TURBINE TRIP C8. BAR 1-BP-3.7, LOW POWER TRIP BLOCKED P8. 1BEP-0, REACTOR TRIP OR SAFETY INJECTION. 1BOA TG-8, TURBINE TRIP BELOW P8.




System ID: 1743241 User-Defined ID: 2017 NRC Q63 Cross Reference Number: S.RP2-04

Topic:

Unit 1 is at 60% power. The following indications are observed: What actions are required


License Level: RO Cognitive Level: High, multiple analysis required. 045 Main Turbine Generator (MT/G) System K1 Knowledge of the physical connections and/or cause effect relationships between the MT/G system and the following systems: K1.18 RPS. Objective: S.RP2-04, Given a set of plant conditions, ANALYZE those conditions and DETERMINE if conditions exist that demand a Reactor Trip, that would allow blocking (Permissives) and/or that would actuate any control systems/devices associated with RPS.



64 ID: 2017 NRC Q64 Points: 1.00 At 1000 A radioactive liquid release is in progress. • Circulating Water Blowdown flow at 11,200 gpm. At 1020 Circulating Water Blowdown drops to 9,900 gpm with the release continuing. What action is required, if any?

A. Continue to MONITOR the release. No additional actions required.

B. REDUCE Release Tank flowrate to maintain the same proportion of release flow to

blowdown flow that is contained in BCP 400-TWX01, LIQUID RADWASTE RELEASE FORM FOR RELEASE TANK 0WX01T.

C. INCREASE CW Blowdown flowrate to restore it to the original value using BOP CW-12

CIRCULATING WATER BLOWDOWN SYSTEM STARTUP,OPERATION AND SHUTDOWN.

D. TERMINATE the release in accordance with BCP 400-TWX01, LIQUID RADWASTE

RELEASE FORM FOR RELEASE TANK 0WX01T.

Answer: D

Answer Explanation



Per the Release Permit, BCP 400-TWX01, page 25; fourth NOTE: Do not adjust CW blowdown flow during the release. If CW blowdown flow must be adjusted terminate the release. Per BOP WX-120, Release Tank Operations Overview, Section D. Precautions: The release is based on dilution by blowdown. The blowdown rate cannot be decreased or the release rate increased during a release or the release must be stopped immediately. A. Incorrect: This is plausible as the operators have control of this parameter, and it has a typical

range of 0-12,000 gpm. B. Incorrect: This is plausible as the Release Permit states in step 9.9.6 for the operator to

ADJUST 0WX897, Rel Tk Low Flow Dsch to CW B/D Flow Control Vlv, with controller 0FK-WX630, to obtain a release rate not to exceed the maximum release rate specified in step 5.8. or the computerized printout.

C. Incorrect: This is plausible as the operators perform adjustments of components to return the

value to the desired range. D. Correct as described above. References: BCP 400-TWX01, Liquid Release Package BOP WX-120, Release Tank Operations Overview BOP CW-12 CIRCULATING WATER BLOWDOWN SYSTEM STARTUP,OPERATION AND SHUTDOWN. BAR 0PL01J-7-B6, CIRC WTR BLOWDOWN FLOW LOW Question meets the K/A because it asks about a liquid radwaste release with a failure of the release to terminate on Low Circulating Water Blowndown flowrate and then has the candidate pick the appropriate procedure action.




System ID: 1743240 User-Defined ID: 2017 NRC Q64 Cross Reference Number: S.WX1-11

Topic: At 1000 A radioactive liquid release is in progress. Circulating Water Blowdown flow at 11,200 gpm


RO Level High Cog Level due to performing more than 1 mental manipulation K/A:068 Liquid Radwaste System, A2: Ability to (a) predict the impacts of the following malfunctions or operations on the Liquid Radwaste System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations, .04: Failure of automatic isolation Objective: State the conditions required to open the Key Lock Valves and what signals will close the valves. S.WX1-11



65 ID: 2017 NRC Q65 Points: 1.00 During maintenance activities, the fire protection system pressure lowers to 138 psig. Which Fire Protection pump(s) respond to this pressure transient?

A. The 0A Jockey pump ONLY starts.

B. Both Jockey pumps ONLY start.

C. Both Jockey pumps and the Motor-driven fire pump ONLY will start.

D. Both Jockey pumps, the Motor-driven and Diesel-driven fire pumps start.

Answer: B

Answer Explanation A. Incorrect. See below. Plausible because all of the pumps start on lowering pressure. B. Correct. The Fire Protection pumps start as pressure is dropping: 0A Jockey pump starts at 145 PSIG, 0B Jockey pump starts at 140 PSIG, Motor Driven pump starts at 135 PSIG and Diesel Driven pump starts at 125 PSIG. C. Incorrect. See above. Plausible because all of the pumps start on lowering pressure. D. Incorrect. See above. Plausible because all of the pumps start on lowering pressure. Meets K/A, examinee must have the ability to monitor automatic operation of the Fire Protection System including starting mechanisms of fire water pumps. Technical References: BAR 0-38-B7, FIRE PUMP 0A RUNNING. BAR 0-38-B8, FIRE PUMP 0B RUNNING




System ID: 1742993 User-Defined ID: 2017 NRC Q65 Cross Reference Number: S.FP1-04-A

Topic: During maintenance activities, the fire protection system pressure lowers to 138 psig.


License Level: RO Cognitive Level: Low, from memory. 086 Fire Protection System (FPS) A3 Ability to monitor automatic operation of the Fire Protection System including: A3.01 Starting mechanisms of fire water pumps Objective: S.FP1-04-A, DISCUSS how the water fire protection subsystem header is kept pressurized. Include: Jockey Pumps.



66 ID: 2017 NRC Q66 Points: 1.00 Per LS-AA-119, FATIGUE MANAGEMENT AND WORK HOUR LIMITS, which of the following is an NSO's responsibility?

A. Initiate a 10 CFR 26 Work Hour Limits Waiver.

B. Determine if a 10 CFR 26 Work Hour Limits Waiver is necessary to prevent a condition

adverse to safety.

C. Verifying all work hours are correctly documented if performing covered work.

D. Perform fatigue assessment and waiver documentation checks.

Answer: C

Answer Explanation A. Incorrect. LS-AA-119, lists determining if a 10 CFR 26 Work Hour Limit Waiver is necessary as an Operations Shift Manager responsibility. Plausible because an employee may feel that once it has been determined that a 10 CFR 26 Work Hour Limit Waiver is necessary, they would then initiate it. B. Incorrect. LS-AA-119, lists determining if a 10 CFR 26 Work Hour Limit Waiver is necessary as an Operations Shift Manager responsibility. Plausible because an assessment will need to be made for exceeding 10 CFR 26 Work Hour Limits. C. Correct. LS-AA-119, lists this as an individual’s responsibility. D. Incorrect. LS-AA-119, lists verifying work hours are correctly documented as a cognizant supervisor responsibility. Plausible because Attachment 1, 10 CFR 26 Work Hour Limits Waiver will need to be verified and transmitted to Security for record retention. Meets K/A, examinee must be able to use procedures related to shift staffing and overtime limitations. Technical References: LS-AA-119, FATIGUE MANAGEMENT AND WORK HOUR LIMITS.




System ID: 1741744 User-Defined ID: 2017 NRC Q66 Cross Reference Number: T.AM03-01

Topic: Per LS-AA-119, FATIGUE MANAGEMENT AND WORK HOUR LIMITS, which of the following is an individual's

RO Importance: 2.9 SRO Importance: 3.9 K/A: G2.1.5 Comments: <QQ 1138040(1412)><<New Question

License Level: RO Cognitive Level: Low, From memory. 2.0 GENERIC KNOWLEDGES AND ABILITIES 2.1.5 Ability to use procedures related to shift staffing, such as minimum crew complement, overtime limitations, etc. Objective: 4C.AM-05: Conduct shift turnover and relief



67 ID: 2017 NRC Q67A Points: 1.00 The RUNNING SX pump TRIPS. The NSO will ...

A. take Prompt Action and THEN announce the SX pump trip. NO repeat back is required.

B. announce the SX pump trip and THEN take Prompt Action. NO repeat back is required.

C. take Prompt Action and THEN announce the SX pump trip. The Unit Supervisor WILL

repeat back the communication.

D. announce the SX pump trip and THEN take Prompt Action. The Unit Supervisor WILL repeat back the communication.

Answer: D

Answer Explanation A. Incorrect. See below. Plausible because communication standards are sometimes lowered during transient situations. B. Incorrect. See below. Plausible because communication standards are sometimes lowered during transient situations. C. Incorrect. See below. Plausible because OP-BY-101-0004, STRATEGIES FOR SUCCESSFUL TRANSIENT MITIGATION does allow for the Unit Supervisor to initiate the 3-way communication during the transient response. D. Correct. OP-BY-101-0004, STRATEGIES FOR SUCCESSFUL TRANSIENT MITIGATION, step 4.2, Prompt Action Communication states when a transient requiring PROMPT OPERATOR ACTION occurs the response actions are governed as follows: The Reactor Operator (RO) will announce the transient requiring PROMPT OPERATOR ACTION, followed by a proper repeat back by the Unit Supervisor validating the action to be taken is correct (preferred method). Meets K/A, examinee must have the ability to make proper communications during transient situations as described in OP-BY-101-0004, STRATEGIES FOR SUCCESSFUL TRANSIENT MITIGATION. Technical References: OP-BY-101-0004, STRATEGIES FOR SUCCESSFUL TRANSIENT MITIGATION. OP-AA-104-101, COMMUNICATIONS.




System ID: 1799001 User-Defined ID: 2017 NRC Q67A Cross Reference Number: T.AM69-01-B

Topic:

Unit 1 is at 100% power. 1A SX pump is running. The 1A SX pump TRIPS. The NSO will ...

RO Importance: 3.9 SRO Importance: 4.0 K/A: G2.1.17 Comments: License Level: RO

Cognitive Level: Low 2.1 Conduct of Operations 2.1.17 Ability to make accurate, clear, and concise verbal reports. Objective: T.AM69-01-B, DESCRIBE the following Operations communications policy requirements and given a specific situation APPLY the appropriate requirements: Proper Protocol for Performing and Participating in Briefs during Transient Conditions



68 ID: 2017 NRC Q68A Points: 1.00 The Pressurizer is designed to accommodate a ... 1. ramp of ___ (1) ____ per minute with control rods in automatic AND

2. a step load RISE of ___ (2) ___ without a reactor trip.

A. 1. 5%

2. 10% WITHOUT steam dumps available.

B. 1. 5% 2. 50% WITHOUT steam dumps available.

C. 1. 10%

2. 10% WITH steam dumps available

D. 1. 10% 2. 50% WITH steam dumps available.

Answer: A

Answer Explanation The purpose (design) of the Pressurizer system is to accommodate: 1) a ramp of 5%/min with control rods in auto without a reactor trip, 2) a step load change of 10% with control rods in auto without a reactor trip. 3) a step load reduction of 50% with control rods in auto AND steam dumps available without a reactor trip, 4) a complete loss of load, control rods and steam dumps unavailable WILL cause a reactor trip but water will NOT reach the Pressurizer safeties. A. Correct as described above. B. Incorrect: The first part is correct making this part of the distractor plausible. The second part

is incorrect but plausible because if the steam dumps were available, 50% would be a correct answer.

C. Incorrect: The first part is incorrect but plausible as that is the step load change design feature

without steam dump availability. The second part is correct making this part of the distractor plausible.

D. Incorrect: The first part is incorrect but plausible as that is the step load change design feature

without steam dump availability. The second part is incorrect but plausible because if the steam dumps were available, 50% would be a correct answer.

The question meets the K/A by requiring the examinee to know the purpose (design) for the Pressurizer. Reference: Pressurizer L-P page 5




System ID: 1799275 User-Defined ID: 2017 NRC Q68A Cross Reference Number: S.RY1-01

Topic: The Pressurizer is designed to accommodate a ... ramp of ___ (1) ____ per minute with control rods

RO Importance: 3.9 SRO Importance: 4.0 K/A: G2.1.27 Comments: New question

License Level: RO Cognitive Level: Low (memory) 2.0 GENERIC KNOWLEDGES AND ABILITIES 2.1.27 Knowledge of system purpose and/or function Objective: STATE the purpose of the Pressurizer system.



69 ID: 2017 NRC Q69 Points: 1.00 There is a large water leak coming from the body of valve 0FP248, Aux Building FP Ring Header Isolation Valve. What combinations of valves will isolate the leak?

A. 0FP209A, 0FP252, 0FP839

B. 0FP209B, 0FP249, 0FP841

C. 0FP249, 0FP252, 0FP839, 0FP841

D. 0FP209A, 0FP209B, 0FP249, 0FP252

Answer: D

Answer Explanation A. Incorrect. See below. Plausible because valves 0FP209A, 0FP252, 0FP839 can be used as partial isolation of 0FP248, but do not provide complete isolation. B. Incorrect. See below. Plausible because valves 0FP209B, 0FP249, 0FP841 can be used as partial isolation of 0FP248, but do not provide complete isolation. C. Incorrect. See below. Plausible because valves 0FP249, 0FP252, 0FP839, 0FP841 can be used as partial isolation of 0FP248, but do not provide complete isolation. D. Correct. Drawing M-52 Sheet 1 show valves 0FP209A, 0FP209B, 0FP249, 0FP252 as isolation points for 0FP248. Meets K/A, examinee must have the ability to determine the expected plant configuration using design and configuration control documentation, such as drawings, line-ups, tag-outs, etc. Technical References: Drawing M-52 Sheet 1.




System ID: 1744119 User-Defined ID: 2017 NRC Q69 Cross Reference Number: A.BP2-03-A

Topic: There is a large water leak coming from the body of valve 0FP248, Aux Building FP Ring Header


License Level: RO Cognitive Level: High, multiple analysis required. 2.2 Equipment Control 2.2.15 Ability to determine the expected plant configuration using design and configuration control documentation, such as drawings, line-ups, tag-outs, etc. Objective: A.BP2-03-A, Given a specific P&ID or C&ID, PERFORM the following: Trace Flow Paths Required reference material: M-52 Sheet 1



70 ID: 2017 NRC Q70B Points: 1.00 1BOSR 8.1.2-1, UNIT ONE 1A DIESEL GENERATOR OPERABILITY SURVEILLANCE, is being performed as a post-maintenance test. • The NSO is preparing to synchronize 1A Diesel Generator to the GRID. Per the 1BOSR 8.1.2-1, who must be notified prior to synchronization?

A. Nuclear Duty Officer (NDO)

B. Electric Operations (TSO)

C. Station Surveillance Coordinator (SSC)

D. Operations Analysis Department (OAD)

Answer: B

Answer Explanation Per 1 BOSR 8.1.2-1 step F.7.a.1 (page 23 of 43) Electric Operations (a.k.a Transmission System Operations) is notified prior to synchoniztion. A. Incorrect: See above. The Nuclear Duty Officer is plausible because the Shift frequently

interfaces with this entity regarding activities performed on shift. B. Correct as described above. C. Incorrect: See above. The Station Surveillance Coordinator is plausible because that position

reviews the completed surveillance activities for the station. D. Incorrect: See above. OAD is plausible because they are responsible for various electrical

maintenance and testing operations/procedures at the station during all MODES of operation. The question meets the K/A by testing the examinee on the performance of post maintenance testing requirements of the Diesel Generator and the required interface with the Transmission System Operator.




System ID: 2029199 User-Defined ID: 2017 NRC Q70B Cross Reference Number:

Topic: 1 BOSR 8.1.2-1, UNIT ONE 1A DIESEL GENERATOR OPERABILITY SURVEILLANCE, is being performed as a post

RO Importance: 2.6 SRO Importance: 3.8 K/A: G2.2.17 Comments: New Question

RO Level Low Cog Level K/A: G2: Generic, 2: Equipment Control, .17: Knowledge of the process of managing maintenance activities during power operations, such as risk assessments, work prioritization, and coordination with the transmission system operator. Objective: 7E.AM-163-A Describe the proper procedure for coordinating load changes with Electric Operations. 9/14/17: New question written based on not meeting the K/A. OTPS and Fac. Rep assisted in writing and approving new question with input from the second round validators: WFH 9/19/17 Changes to the justifications incorporated from OTPS review. TM



71 ID: 2017 NRC Q71 Points: 1.00 MMD has to make a weld repair on the Spent Resin Storage Tank. • Total work time is estimated to be 4 hours • Dose rates at the work site are 850 mrem/hour. What is/are the FEWEST number of worker(s) needed to perform the repair AND remain within the Exelon Administrative Dose Limits?

A. 1

B. 2

C. 3

D. 7

Answer: B

Answer Explanation Per RP-AA-203 Exposure Control and Authorization TABLE 1 the admin limit is 2000 mr/yr. 850 mr/hr times 4 hours is 3400 mr. This would require 2 people to not exceed the admin limits. A. Incorrect. This is plausible as the legal limit is 5000 mr/yr B. Correct. See above. C. Incorrect. This is plausible based on miscalculation of total dose. D. Incorrect. This is plausible based on 3400 total dose with the admin limit of 500 mr/year for a

person that declares they are pregnant. Meets K/A, examinee must have knowledge of radiation exposure limits under normal or emergency conditions. Technical References: RP-AA-203, Exposure Control and Authorization




System ID: 1744120 User-Defined ID: 2017 NRC Q71 Cross Reference Number: T.AM46-01

Topic: MMD has to make a weld repair on the Spent Resin Storage Tank. Total work time is estimated to be


License Level: RO Cognitive Level: High due to calculation. 2.3 Radiation Control 2.3.4 Knowledge of radiation exposure limits under normal or emergency conditions. Objective: T.AM46-01, STATE the 10CFR20 Exposure Limits



72 ID: 2017 NRC Q72 Points: 1.00 At 1000 Both Units are at full power, normal alignment. • The 0A VC train is in standby. • The 0B VC train is in normal operation. At 1005 An event that has the potential for an accidental radioactive release in the Unit 2 Auxiliary Building is reported to the MCR. • No equipment actuations have occurred. • All MCR rad monitor icons on the RMS GRID 2, PROCESS AIR MONITORS, are GREEN. To monitor control room intake air on the RMS, the NSO will trend the...

A. 0PR31J or 0PR32J, OUT AIR IN 0A.

B. 0PR33J or 0PR34J, OUT AIR IN 0B.

C. 0PR35J or 0PR36J, TURB AIR IN 0A.

D. 0PR37J or 0PR38J, TURB AIR IN 0B.

Answer: B

Answer Explanation



A. Incorrect: 0PR31J and 32J sample the outside air intake from Unit 1 (0A train). Although the spill was in the Aux. Bldg, because 0A VC train was not running, this plenum would not experience intake air flow. Plausible because the MCR rad monitor indications are green (normal) even when their sampled plenums are not online, and the sample pumps will continuously sample plenums that have stagnant air flow and not the MCR intake air. Also plausible because all monitors are labelled AIR IN on the RMS.

B. Correct: With the 0B VC system in normal alignment (outside air intake) the only rad monitors

that would have MCR intake air flow through their respective intake plenum are 0PR33J and 34J.

C. Incorrect: 0PR35J and 36J sample the turbine bldg intake from Unit 1 (0A train). Although the

spill was in the Aux. Bldg, this plenum would not experience intake air flow unless the 0A VC system was manually or automatically swapped to emergency mode. Plausible because the MCR rad monitor indications are green (normal) even when their sampled plenums are not online, and the sample pumps will continuously sample plenums that have stagnant air flow and not the MCR intake air. Also, plausible because all monitors are labelled AIR IN on the RMS.

D. Incorrect: 0PR37J and 38J sample the turbine bldg intake from Unit 2 (0B train). Although the

spill was in the Aux. Bldg this plenum would not experience intake air flow unless the 0B VC system was manually started in emergency mode. Plausible because the MCR rad monitor indications are green (normal) even when their sampled plenums are not online, and the sample pumps will continuously sample plenums that have stagnant air flow and not the MCR intake air. Also plausible because all monitors are labelled AIR IN on the RMS.

Meets K/A, the question requires the examinee to know the normal alignment for VC and which rad monitors are used to monitor that alignment. Technical References: BOP VC-1, STARTUP OF CONTROL ROOM HVAC SYSTEM. 0BOL 3.7, LCOAR CONTROL ROOM VENTILATION (VC) FILTRATION SYSTEM ACTUATION INSTRUMENTATION TECH SPEC LCO # 3.3.7




System ID: 1741848 User-Defined ID: 2017 NRC Q72 Cross Reference Number: S.AR1-06

Topic:

At 1000 Both Units are at full power, normal alignment. The 0A VC train is in standby. The 0B

RO Importance: 2.9 SRO Importance: 2.9 K/A: G2.3.5 Comments: <QQ 1138040(1412)><<Bank Question from BWD 2013 NRC

Exam Q 71 (ID: RG30235-N01) License Level: RO Cognitive Level: Low, from memory 2.0 GENERIC KNOWLEDGES AND ABILITIES 2.3.5 Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc. Objective: S.AR1-06, IDENTIFY the systems monitored by Process Radiation Monitors



73 ID: 2017 NRC Q73 Points: 1.00 Unit 1 is in Mode 5 • Both PZR PORVs are OPEN and ENERGIZED per 1BGP 100-5, PLANT SHUTDOWN AND

COOLDOWN • All Loop Stop Isolation valves are OPEN A LOSS of U-1 RH Cooling occurs: • Attempts to restore U-1 RH cooling are unsuccessful • The crew has implemented 1BOA PRI-10, LOSS OF RH COOLING Per 1BOA PRI-10, the PREFERRED method of cooling is ...

A. RCS Bleed and Feed.

B. SI Accumulator Injection.

C. SI Pump Hot Leg Injection.

D. Steaming Intact/Non-isolated SGs.

Answer: D

Answer Explanation



Per BOA PRI-10, LOSS OF RH COOLING, Table A, DETERMINATION OF EFFECTIVE RCS COOLING METHODS: The RCS is ADVERSE if ALL of the following exist: a) The RCS has a Cold leg opening AND b) The Reactor Vessel Head is NOT removed AND c) All RCS Loops are blocked (loop stops closed or hot leg nozzle dams are installed) The RCS is intact if the PZR PORV(s), PZR vent or reactor are open but can be isolated. A. Incorrect. Per BOA PRI-10, Table A note 4, Steaming Intact/Non-isolated SGs is the preferred heat removal method. Plausible because both PZR PORVs are open, as stated in the stem of the question. Attachment C, RCS BLEED AND FEED, uses both the PZR PORVs and charging pumps for the bleed and feed paths. With both PZR PORVs already open the examinee may choose this as an acceptable option. B. Incorrect. Per BOA PRI-10, Table A note 4, Steaming Intact/Non-isolated SGs is the preferred heat removal method. Plausible because using attachment F, ACCUMULATOR INJECTION, is the quickest method of supplying cool water to the RCS. Also, this is an acceptable method per Table A. C. Incorrect. Per BOA PRI-10, Table A note 4, Steaming Intact/Non-isolated SGs is the preferred heat removal method. Plausible because a common misconception is that with the PZR PORVs open, the RCS is not intact. For this reason note 5 of Table A states: The RCS is intact if the PZR PORV(s), PZR vent or reactor are open but can be isolated. This would be an acceptable method per Table A if the RCS was not intact and with both RH pumps tripped the operator would want to start the SI pumps. D. Correct. Per BOA PRI-10, Table A note 4, Steaming Intact/Non-isolated SGs is the preferred heat removal method. Meets K/A, the question requires the examinee to know that the mitigating strategy for 1BOA PRI-10 is to use Steaming Intact/Non-isolated SGs when the RCS is not adverse in Mode 5. This is RO level based on required knowledge of mitigating strategy(ies): From the ES-401 Attachment 2 Figure 2: Screening for SRO-only linked to 10CFR55.43(b)(5) (Assessment and selection of procedures): 1) assessing plant conditions and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. Using the flowchart, this question can: • NOT be answered solely by knowing “systems knowledge”, i.e., how the system works, flowpath, logic, component location. No • NOT be answered solely by knowing immediate operator actions. No • NOT be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs. No • NOT be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure. YES Technical References: BOA PRI-10, LOSS OF RH COOLING





Topic: Unit 1 is in Mode 5 Both PZR PORVs are open per 1BGP 100-5, PLANT SHUTDOWN AND COOLDOWN.

RO Importance: 3.8 SRO Importance: 4.2 K/A: G2.4.9 Comments: Modified Bank Q<QQ 1138040(1412)><<NewnnnNuestion

2013 CERT EXAM Q73 (1146450) License Level: RO Cognitive Level: High, multiple analysis required. 2.4 Emergency Procedures / Plan 2.4.9 Knowledge of low power/shutdown implications in accident (e.g., loss of coolant accident or loss of residual heat removal) mitigation strategies. Objective: T.OA20-07, DISCUSS alternate methods of RCS decay heat removal per procedure 1/2BOA PRI-10, Loss of RH Cooling Requires reference: 1BOA PRI-10 Table A



74 ID: 2017 NRC Q74 Points: 1.00 At 1000 A Reactor trip and SI has occurred as a result of a LOCA. At 1005 The Shift Manager has declared an Alert (FA1). What is the LATEST time that a NARS communication, if any, must be made?

A. 1015

B. 1020

C. 1030

D. NO NARS communication required.

Answer: B

Answer Explanation A. Incorrect. See below. Plausible because this would be 15 minutes from the time that the LOCA occurred and the site is required to make the NARS call in 15 minutes of the declaration of an emergency. B. Correct. EP-MW-114-100, MIDWEST REGION OFF-SITE NOTIFICATIONS step 4.1.1 states that the NARS call has to be made within 15 minutes of the declaration of an emergency. C. Incorrect. See above. Plausible because the Emergency Director has 15 minutes to declare an emergency and then 15 minutes to make the NARS call. D. Incorrect. This is plausible as there are many events at the station that do not require a NARS call to be made. Meets K/A, examinee must have knowledge of events related to system operation/status that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator. Technical References: EP-MW-114-100, MIDWEST REGION OFF-SITE NOTIFICATIONS




System ID: 1744130 User-Defined ID: 2017 NRC Q74 Cross Reference Number: T.ZP1-35

Topic:

At 1000 A Reactor trip and SI has occurred as a result of a LOCA. At 1005 The Shift Manager


License Level: RO Cognitive Level: High, multiple analysis required. 2.4 Emergency Procedures / Plan 2.4.30 Knowledge of events related to system operation/status that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator. Objective: T.ZP1-35, EXPLAIN the use of the nuclear accident reporting system (NARs) including: Initial Notification and Update Requirements (Time Limits)



75 ID: 2017 NRC Q75A Points: 1.00 Unit 2 is at 100% power. Annunciator 2-12-B7, PRT PRESS HIGH, is in alarm. • PRT Level: 74% • PRT Temp: 80°F • PRT Press: 10 psig 2RY469, PRT to GW Isolation Valve, indicates CLOSED. To REDUCE PRT pressure to 3 psig, what actions are required?

A. Vent ONLY.

B. Vent, THEN Drain.

C. Drain, THEN Vent.

D. Vent and Drain SIMULTANEOUSLY.

Answer: C

Answer Explanation



Per the Alarm Response Procedure the PRT vent valve auto closes at a pressure of 6 psig. Once this valve closes, the only way to reduce pressure in the PRT is to lower level. This is accomplished by utilizing BOP RY-4, DRAINING THE PRESSURIZER RELIEF TANK. Draining the PRT will reduce pressure which would then allow the tank to be vented once pressure is less than 6 psig. A. Incorrect: The vent valve will not open at 10 psig. It is plausible as normally vent valves do not

have pressure interlocks to auto close. B. Incorrect: The evolutions stated are correct but they are in the wrong order as described

above making this distractor plausible. C. Correct: Level will first be lowered to reduce pressure to <6 psig, then the vent vale can be

opened to reduce pressure to the normal setpoint of approximately 3 psig. D. Incorrect: This distractor is plausible as it contains the actions necessary to reduce pressure;

however, they cannot be performed simultaneously because the pressure is too high to allow opening the vent valve.

Question meets the K/A because the examinee must have the ability to verify system alarm setpoints and operate controls to obtain the desired response in accordance with the Alarm Response Procedure and knowledge of system interlocks. The auto closure of RY469 occurs at the alarm setpoint (6 psig). The stem of the question states that the alarm is "in" and the valve has closed, as pressure is 10 psig indicating the alarm is valid. References: BAR 1-12-B7, PRT PRESSURE HIGH BAR 1-12-A7, PRT LEVEL HIGH LOW BOP RY-4, DRAINING THE PRESSURIZER RELIEF TANK.




System ID: 1799033 User-Defined ID: 2017 NRC Q75A Cross Reference Number: S.RY1-17

Topic:

Unit 2 is at 100% power. Annunciator 2-12-B7, PRT PRESS HIGH, is in alarm. PRT Level: 74%


License Level: RO Cognitive Level: High, multiple analysis. 2.4 Emergency Procedures / Plan 2.4.50 Ability to verify system alarm setpoints and operate controls identified in the alarm response manual. Objective: S.RY1-17, EXPLAIN how to drain the PRT; include all the interlocks associated with automatically starting the RCDT Pumps



76 ID: 2017 NRC Q76B Points: 1.00 At 1000 A Reactor Trip First Out annunciator is in alarm with the reactor at power. • The reactor did NOT trip. • The crew entered BFR-S.1, RESPONSE TO NUCLEAR POWER GENERATION/ATWS. • Step 7, Check if the Reactor is Subcritical, is in progress. At 1010 • Power Range Nuclear Instruments are indicating 5%. • Intermediate Range Startup Rate is -0.1 DPM. 1. The reactor trip criteria have...

2. Which procedure must be executed?

A. 1. NOT been met due to Power Range indication ONLY.

2. Continue in BFR-S.1 at step 8.

B. 1. NOT been met due to Power Range indication ONLY. 2. Exit BFR-S.1; return to BEP-0, REACTOR TRIP OR SAFETY INJECTION.

C. 1. Been met due to Intermediate Range Startup Rate indication.

2. Continue in BFR-S.1 at step 8.

D. 1. Been met due to Intermediate Range Startup Rate indication. 2. Exit BFR-S.1; return to BEP-0, REACTOR TRIP OR SAFETY INJECTION.

Answer: A

Answer Explanation



Per BFR-S.1, step 7, the conditions that must be met to exit the procedure are: (1) PR channels; <5% and IR channels -SUR. The conditions are NOT met because PR indications are NOT less than 5%. Reactor Power is 5%. IR SUR does meet the criterion because it is negative. Based on the indications given, the correct action is to remain in BFR-S.1 A. Correct. See above. B. Incorrect. See above. Plausible because part one is correct. Part two is also plausible

because the IR SUR being negative and the procedure requiring boration to be continued even after exit of the procedure; the examinee may conclude all required actions have been taken. Also, part two is plausible because the examinee may conclude the items in step seven are open bullets which would then require only one to meet the step.

C. Incorrect. See above. Plausible because the first part, IR SUR being negative, meets part of the step requirement. Part two is also plausible because it is correct.

D. Incorrect. See above. Plausible because the first part, IR SUR being negative, meets part of the step requirement. Part two is also plausible because the IR SUR being negative and the procedure requiring boration to be continued even after exit of the procedure; the examinee may conclude all required actions have been taken. Also, part two is plausible because the examinee may conclude the items in step seven are open bullets which would then require only one to meet the step.

Meets the K/A, the question requires the examinee to assess Power Range and Intermediate Range Startup Rate indications and determine that the reactor is not shutdown. From the ES-401 Attachment 2 Figure 2: Screening for SRO-only linked to 10CFR55.43(b)(5) (Assessment and selection of procedures): 1) assessing plant conditions and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. Using the flowchart, this question can: • NOT be answered solely by knowing “systems knowledge”,i.e., how the system works, flowpath, logic, component location. NO, procedural step question • NOT be answered solely by knowing immediate operator actions. NO, step is not immediate action • NOT be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs. NO • NOT be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure. NO, question asks detailed information on 1 procedure step • CAN be answered with knowledge of ONE or MORE of the following: - Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. YES, if the reactor does not meet the criteria of "subcritical" the procedure is continued at the next step. If it meets the criteria the procedure is exited and BEP-0 is entered. - Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps. - Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures. - Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures.

References: 1BFR-S.1 step 7 BEP-0, step 1 BST-1 Subcriticality




System ID: 1742669 User-Defined ID: 2017 NRC Q76B Cross Reference Number: T.FR01-07

Topic:

At 1000 A Reactor Trip First Out annunciator is in alarm with the reactor at power. The reactor

RO Importance: SRO Importance: 4.3 K/A: EPE007EA2.01 Comments: New Question

SRO Level High Cog level (analysis) K/A: EPE007: (Reactor Trip), EA2 (Ability to determine or interpret the following as they apply to a Reactor Trip), .01 (Decreasing power level from available indication) Objective: Analyze a given set of conditions and determine the appropriate operator actions to respond to an ATWS event



77 ID: 2017 NRC Q77 Points: 1.00 Consider the following conditions for BOTH Unit 1 AND Unit 2.

At 1200 A Loss of Coolant Accident (LOCA) occurs. • The reactor has TRIPPED. • Safety Injection has INITIATED. • _BEP-0, REACTOR TRIP OR SAFETY INJECTION, is in progress.

At 1220: • Containment pressure is 6 psig after peaking at 11 psig. • Containment High Range radiation monitors (_RT-020 and _RT-021) are at the ALERT set point. • All Steam Generator narrow range levels are indicating 33%. At the current time, what actions would the US direct to feed the Steam Generators for each Unit? 1. On Unit 1, feed the Steam Generators to maintain ...

2. On Unit 2, feed the Steam Generators to maintain ...

A. 1. narrow range levels greater than 31%.

2. narrow range levels greater than 31%.

B. 1. narrow range levels greater than 31%. 2. total feed flow greater than 500 gpm until narrow range levels are greater than 34% .

C. 1. total feed flow greater than 500 gpm until narrow range levels are greater than 34% .

2. maintain narrow range levels greater than 31%.

D. 1. total feed flow greater than 500 gpm until narrow range levels are greater than 34% . 2. total feed flow greater than 500 gpm until narrow range levels are greater than 34% .

Answer: B

Answer Explanation



The containment is in an adverse condition (greater than 5 psig containment pressure) and as such, S/G levels will be maintained greater than 31% on Unit 1. Unit 2's value for minimum level under adverse conditions is greater than 34%. The analysis identifies Unit 1 S/Gs meet the minimum requirement and Unit 2 does not. The RNO column of 2BEP-1 directs the operator to maintain total feed flow greater than 500 gpm until level is above the minimum required. A. Incorrect: Unit 2's value for minimum level under adverse conditions is greater than 34%. The

distractor is plausible because the first part is correct. The second part is incorrect but plausible because S/G levels will be maintained greater than 31% on Unit 1.

B. Correct as stated above C. Incorrect: The first part is incorrect but plausible if the examinee confuses the Unit 2 number

with Unit 1. The second part is incorrect but plausible if the examinee confuses the Unit 1 numbers with Unit 2.

D. The first part is incorrect but plausible if the examinee confuses the Unit 1 number with Unit 2. Furthermore, the distractor is plausible because the second part is correct.

Meets the K/A, the question requires the examinee to know, for a LOCA, the different values for S/G minimum levels to maintain the secondary heat sink for both units. Question is SRO level due to assessment of plant conditions and then selection of a procedure or a portion of the procedure to mitigate or recover or with which to proceed. From the ES-401 Attachment 2 Figure 2: Screening for SRO-only linked to 10CFR55.43(b)(5) (Assessment and selection of procedures): 1) assessing plant conditions and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. Using the flowchart, this question can: • NOT be answered solely by knowing “systems knowledge”,i.e., how the system works, flowpath, logic, component location. • NOT be answered solely by knowing immediate operator actions. • NOT be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs. • NOT be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure.

• CAN be answered with knowledge of ONE or MORE of the following: - Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. YES, the examinee is required to evaluate the conditions for both Units and then for Unit 2 use the RNO of BEP-0 step 6 to maintain adequate feedwater flow until adequate SG level is reached.

References: 1BEP-0 and 2BEP-0 Note before step 1 defines adverse containment RNO column for step 6d provides acceptable S/G levels for adverse containment conditions




System ID: 1741695 User-Defined ID: 2017 NRC Q77 Cross Reference Number: T.EP02-02-D

Topic: Consider the following conditions for BOTH Unit 1 AND Unit 2. At 1200 A Loss of Coolant Accident (

RO Importance: SRO Importance: 3.6 K/A: EPE009G2.2 Comments: New Question for 2017 NRC exam

SRO level High Cog level due to analysis K/A: EPE 009 (Small Break LOCA) G 2.2 (Equipment Control) .4 (ability to explain the variations in control board/control room layouts, systems, instrumentation, and procedural actions between units at a facility. Objective: T.EP02-02-D, Define the following term, Adverse Containment



78 ID: 2017 NRC Q78A Points: 1.00 Unit 1 is at 100% power: • The 1A AF pump has been OOS for the last 68 hours. • The 1B AF pump was just declared INOPERABLE. 1. Unit 1 will ...

2. What is the reason for the required action?

A. 1. be in MODE 3 in 7 hours and MODE 4 within 13 hours

2. LCO 3.0.3 is not applicable because it could force the unit into a less safe condition.

B. 1. be in MODE 3 in 7 hours and MODE 4 within 13 hours 2. The condition of the unit is NOT specifically addressed by the associated ACTIONS.

C. 1. NOT ramp until ONE train of Aux. Feed is OPERABLE

2. The condition of the unit is NOT specifically addressed by the associated ACTIONS.

D. 1. NOT ramp until ONE train of Aux. Feed is OPERABLE 2. LCO 3.0.3 is not applicable because it could force the unit into a less safe condition.

Answer: D

Answer Explanation



A. Incorrect. See below. Plausible because many systems require entry into LCO 3.0.3 when both trains are inoperable. The second part is plausible because it is the bases for LCO 3.7.5 Required Action C.1. B. Incorrect. See below. Plausible because many systems require entry into LCO 3.0.3 when both trains are inoperable. The second part is plausible because it is the bases for LCO 3.0.3. C. Incorrect. See below. Plausible because the first part is the required action for LCO 3.7.5. The second part is plausible because it is the bases for many LCO’s. D. Correct. T.S. 3.7.5 Auxiliary Feedwater (AF) System is the applicable specification. Condition C applies as 2 AF pumps are inoperable. The ACTION is: Initiate action to restore one AF train to OPERABLE status. The bases states for Required Action C.1 is modified by a Note indicating that all required MODE changes or power reductions are suspended until one AF train is restored to OPERABLE status. In this case, LCO 3.0.3 is not applicable because it could force the unit into a less safe condition. This question meets the K/A by testing the examinee on T.S. Action Statements that are 1 hour or less. It is SRO due to the decision as to which action to take based on the RO recommendation. Further, it is SRO based on the T.S. Bases portion of the question. From the ES-401 Attachment 2 Figure 1: Screening for SRO-only linked to 10 CFR 55.43(b)(2) (Tech Specs): • Can NOT be answered by knowing less than 1 hour Tech Specs. NOT entirely. • Can NOT be answered by knowing information listed "above-the-line". NO • Can NOT be answered by knowing the TS Safety Limits. NO • Does involve one or more of the following for TS, TRM or ODCM: YES - Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1). NO - Application of generic LCO requirements (LCO 3.0.1 thru 3.0.7 and SR 4.0.1 thru 4.0.4). NO - Knowledge of TS bases that is required to analyze TS required actions and terminology. YES References: T.S. 3.7.5 and bases bases for T.S. 3.7.5




System ID: 1845750 User-Defined ID: 2017 NRC Q78A Cross Reference Number: S.AF1-1

Topic:

Unit 1 is at 100% power: The 1A AF pump has been OOS for the last 68 hours. The 1B AF pump was

RO Importance: SRO Importance: 4.5 K/A: 2.2.39 Comments: New Question

SRO Level High Cog Level due to performing more than 1 mental manipulation. K/A: Generic 2.2.39: Knowledge of less than or equal to one hour Technical Specification action statements for systems. Objective: S.AF1-16, Given a set of plant conditions, DETERMINE from memory, applicable Auxiliary Feedwater Tech Spec/TRM operability requirements.



79 ID: 2017 NRC Q79B Points: 1.00 The crew has entered 2BFR-S.1, RESPONSE TO NUCLEAR POWER GENERATION/ATWS, due to a Red First Out Annunciator for a Reactor Trip. 2B SG has been identified as faulted. Step 13 of 2BFR-S.1, ISOLATE FAULTED SG, is in progress. • 2FW520, 2B S/G FW Reg Valve, is showing dual indications. What action must be directed?

A. Close 2FW520A, 2B S/G FW Reg Bypass Valve.

B. Close 2FW006B, 2B S/G FW Reg Upstream Isolation Valve.

C. Close 2FW035B, 2B S/G FW Tempering Line Downstream Isolation Valve.

D. Close 2FW039B, 2B S/G FW Preheater Bypass Downstream Isolation Valve.

Answer: B

Answer Explanation



A. Incorrect. See below. Plausible because 2FW520B is a FW isolation valve which is verified closed in this step. It is also a flowpath around 2FW520. B. Correct. Step 13 of 2BFR-S.1, RESPONSE TO NUCLEAR POWER GENERATION/ATWS, directs the operator to isolate the faulted SG. 2FW520 is one of those isolation valves and is showing dual indication. This requires isolation per the RNO, which directs the operator to close 2FW006B. C. Incorrect. See above. Plausible because 2FW035B is a FW isolation valve which is verified closed in this step. Also, this is an additional FW flowpath. D. Incorrect. See above. Plausible because 2FW039B is a FW isolation valve which is verified closed in this step. Also, this is an additional FW flowpath. Meets the K/A, the question requires the examinee evaluate the position indication of 2FW520 and then determine which FW valve to close per 2BFR-S.1 step 13 RNO. Question is SRO Level due to required knowledge the actions needed to complete a step that is not an immediate action step. This question requires detailed knowledge of the ATWS procedure. From the ES-401 Attachment 2 Figure 2: Screening for SRO-only linked to 10CFR55.43(b)(5) (Assessment and selection of procedures): 1) assessing plant conditions and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. Using the flowchart, this question can: • NOT be answered solely by knowing “systems knowledge”,i.e., how the system works,

flowpath, logic, component location. NO • NOT be answered solely by knowing immediate operator actions. NO • NOT be answered solely by knowing entry conditions for AOPs or plant parameters that

require direct entry to major EOPs. NO • NOT be answered solely by knowing the purpose, overall sequence of events, or overall

mitigative strategy of a procedure. NO • CAN be answered with knowledge of ONE or MORE of the following: • Assessing plant conditions (normal, abnormal, or emergency) and then selecting a

procedure or section of a procedure to mitigate, recover, or with which to proceed. YES, requires the examinee to select 2BFR-S.1 step 13 RNO to determine which valve to close due to 2FW520 being open.

Reference: 1BFR-S.1 RESPONSE TO NUCLEAR POWER GENERATION/ATWS




System ID: 1845769 User-Defined ID: 2017 NRC Q79B Cross Reference Number: T.FR1-07

Topic: The crew has entered 2BFR-S.1, RESPONSE TO NUCLEAR POWER GENERATION/ATWS, due to a Red First Out


SRO Level High Cog Level due to making procedural decisions. K/A: EPE029: Anticipated Transient Without Scram (ATWS), EA2: Ability to determine or interpret the following as they apply to a ATWS, EA2.05: System component valve position indications Objective: ANALYZE a given set of conditions and determine the appropriate operator actions to respond to an ATWS event. (T.FR1-07)



80 ID: 2017 NRC Q80A Points: 1.00 Both Units are at 100% power. • An Instrument Air leak has occurred in the U-1 Aux Building. • The EO reports that the leak is at 364' elevation in the Aux Building. • U1 NSO observes CHARGING FLOW is rising and SEAL INJECTION FLOWS are rising. • Unit 1 Pressurizer level is 71%. While isolating the leak per 0BOA SEC-4, LOSS OF INSTRUMENT AIR, ___(1)___ per ___(2)___.

A. 1. bypass and isolate 1CV121, CV Pumps Discharge Header FCV,

2. 1BOA SEC-4, LOSS OF INSTRUMENT AIR

B. 1. shutdown the running charging pump(s) 2. 1BOA SEC-4, LOSS OF INSTRUMENT AIR

C. 1. bypass and isolate 1CV121, CV Pumps Discharge Header FCV,

2. 1BEP ES-0.1, REACTOR TRIP RESPONSE

D. 1. shutdown the running charging pump(s) 2. 1BEP ES-0.1, REACTOR TRIP RESPONSE

Answer: A

Answer Explanation



A. Correct. Isolating the leak per 0BOA SEC-4, LOSS OF INSTRUMENT, will result in the isolation of IA to 1CV121, CV Pumps Discharge Header FCV. This will cause the valve to fail open. 1BOA SEC-4, LOSS OF INSTRUMENT AIR step 3, directs the operator to bypass and isolate 1CV121 if it is failed open. B. Incorrect. See above. The CV system has many valves that fail closed on loss of IA. It is plausible that you would shut down the CV pumps if the FCV failed close. Part 2 adds to the plausibility due to it is the procedure used to mitigate this situation. C. Incorrect. See above. The CV system has many valves that fail closed on loss of IA. It is plausible that you would bypass 1CV121 to supply RCP seal injection. Also, with the loss of charging, it is plausible that reactor would trip. 1BEP ES-0.1, REACTOR TRIP RESPONSE would then be used to stabilize the plant. Pressurizer level of 72% lends plausibility to Reactor Trip Response procedure because at 70% the High Pressurizer Level will annunciate. The Reactor trip setpoint is 92%. D. Incorrect. See above. The CV system has many valves that fail closed on loss of IA. It is plausible that you would shut down the CV pumps if the FCV failed close. It is plausible that reactor would trip with no charging flow. 1BEP ES-0.1, REACTOR TRIP RESPONSE would then be used to stabilize the plant. Pressurizer level of 72% lends plausibility to Reactor Trip Response procedure because at 70% the High Pressurizer Level will annunciate. The Reactor trip setpoint is 92%. The question meets the K/A by placing the examinee in a situation in which a leak in the IA system is occurring. This requires the examinee to determine leak location and determine which procedure contains actions to mitigate the situation while isolating the leak. It is SRO level by requiring the examinee to evaluate the conditions and select the right procedure. From the ES-401 Attachment 2 Figure 2: Screening for SRO-only linked to 10CFR55.43(b)(5) (Assessment and selection of procedures): 1) assessing plant conditions and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. Using the flowchart, this question can: • NOT be answered solely by knowing “systems knowledge”, i.e., how the system works, flowpath, logic, component location. No • NOT be answered solely by knowing immediate operator actions. No • NOT be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs. No • NOT be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure. No • CAN be answered with knowledge of ONE or MORE of the following: - Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. Yes References: 0BOA SEC-4, LOSS OF INSTRUMENT AIR- UNIT 0 1BOA SEC-4, LOSS OF INSTRUMENT AIR- UNIT 1




System ID: 1799573 User-Defined ID: 2017 NRC Q80A Cross Reference Number: 3D.OA-25-A

Topic: Both Units are at 100% power. An Instrument Air leak has occurred in the U-1 Aux Building. The EO


SRO Level High Cog Level, requires multiple anaysis. K/A: APE065; Loss of Instrument Air, AA2; Ability to determine and interpret the following as they apply to the Loss of Instrument Air, .03; Location and isolation of leaks Objective: 3D.OA-25-A, DISCUSS the response of integrated plant and system critical parameters to a Loss of Instrument Air, with and without operator action



81 ID: 2017 NRC Q81B Points: 1.00 Unit 1 is operating at 100% power when a grid transient occurs. • Bus 141 voltage is 3850v • Bus 142 voltage is 3950v What is the required action, if any, and MAXIMUM completion time per Tech Specs?

A. Enter 3.0.3 Immediately.

B. No actions required.

C. Restore AC electrical power distribution subsystem to OPERABLE status within 8 hours

AND 16 hours from discovery.

D. Be in Mode 3 within 6 hours.

Answer: C

Answer Explanation A. Incorrect. See below. This is plausible as the T.S. addresses busses on BOTH Units because the units can be crosstied. The 1 ESF Bus is affected on 1 unit but then the crosstie capability comes into question; therefore, it is plausible to conclude that 1 bus irregularity can affect both units which could possibly put the unit into T.S. 3.0.3 applicability B. Incorrect. See below. Plausible because Bus 141 degraded voltage setpoint is < 3847.5 volts. The examinee may conclude the bus is operable when above that setpoint. C. Correct. With 1 ESF Bus voltage low, condition A applies which states Restore AC electrical power distribution subsystem to OPERABLE status within 8 hours AND 16 hours from discovery of failure to meet LCO. D. Incorrect. See above. This is plausible as this is an action contained within T.S. 3.8.9, and this would be applied if the action and associated completion time were not met from condition A above. Meets K/A, Requires examinee to apply Technical Specifications for the AC systems. SRO only, From the ES-401 Attachment 2 Figure 1: Screening for SRO-only linked to 10 CFR 55.43(b)(2) (Tech Specs): • Can NOT be answered by knowing less than 1 hour Tech Specs. No • Can NOT be answered by knowing information listed "above-the-line". No • Can NOT be answered by knowing the TS Safety Limits. No • Does involve one or more of the following for TS, TRM or ODCM: - Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1). Yes Technical References: LCO 3.8.9, Distribution Systems-Operating and 1BOSR 8.9.1-1, ESF ONSITE POWER DISTRIBUTION WEEKLY SURVEILLANCE DIVISION 11.




System ID: 2029203 User-Defined ID: 2017 NRC Q81B Cross Reference Number: 3D.OA-25-A

Topic: Unit 1 is operating at 100% power when a grid transient occurs. Bus 141 voltage is 3850v

RO Importance: SRO Importance: 2.9 K/A: APE065AA2.03 Comments: New Question

License Level: SRO Cognitive Level: High, multiple analysis required. APE: 077 Generator Voltage and Electric Grid Disturbances G2.2.40 Ability to apply Technical Specifications for a system. S.AP1-15, Given a set of plant conditions, DETERMINE applicable AC Electrical Distribution System Tech Spec/TRM operability requirements. PROVIDE REFERENCE 1BOSR 8.9.1-1



82 ID: 2017 NRC Q82A Points: 1.00 A Steam Generator Tube Leak is in progress. • 1BOA SEC-8, STEAM GENERATOR TUBE LEAK, has been entered. • Chemistry is in the process of pulling a grab sample in order to determine the leak rate. • 1PR27J, SJAE/GS Exhaust Process Radiation Monitor, is reading 2.25E-05 μCi/cc above background. Use the available control room indications to estimate the leak rate.

A. 5 GPD

B. 45 GPD

C. 80 GPD

D. 110 GPD

Answer: C

Answer Explanation



A. Incorrect. See below. Plausible because BCB-1 Figure 36 has a log scale on one axis only. When reading log scales, operators sometimes find part of the number that they are looking for and stop there. This would be the correct answer for a reading of 2.25E-6. B. Incorrect. See below. Plausible because BCB-1 Figure 36 has a log scale on one axis only. When reading log scales, operators sometimes find part of the number that they are looking for and stop there. This would be the correct answer for a reading of 1.25E-5. C. Correct. 1BOA SEC-8, STEAM GENERATOR TUBE LEAK step 5a directs estimation of SG tube leak rate. One method allowed is use of 1BOSR SG-1, STEAM GENERATOR PRIMARY TO SECONDARY LEAKAGE ESTIMATION. This procedure allows the use of BCB-1 figure 36. Per BCB-1 figure 36, the leak rate is approximately 80 GPD. D. Incorrect. See above. Plausible because BCB-1 Figure 36 has a log scale on one axis only. When reading log scales, operators sometimes find part of the number that they are looking for and stop there. This this would be the correct answer for a reading of 3.25E-6. Meets K/A, the question requires the examinee to know how to use condensate air ejector exhaust monitor readings to determine steam generator leak. From the ES-401 Attachment 2 Figure 2: Screening for SRO-only linked to 10CFR55.43(b)(5) (Assessment and selection of procedures): 1) assessing plant conditions and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. Using the flowchart, this question can: • NOT be answered solely by knowing “systems knowledge”, i.e., how the system works, flowpath, logic, component location. No • NOT be answered solely by knowing immediate operator actions. No • NOT be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs. No • NOT be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure. No • CAN be answered with knowledge of ONE or MORE of the following: - Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. Yes, the examinee needs to select BCB-1 Figure 36, then use the information in the stem to answer the question. Technical References: 1BOSR SG-1, STEAM GENERATOR PRIMARY TO SECONDARY LEAKAGE ESTIMATION and 1BOA SEC-8, STEAM GENERATOR TUBE LEAK, BCB-1 Figure 36





Topic:

A Steam Generator Tube Leak is in progress. 1BOA SEC-8, STEAM GENERATOR TUBE LEAK has been enter

RO Importance: SRO Importance: 3.4 K/A: 037AA2.09 Comments: New Question

License Level: SRO Cognitive Level: High Cog due to performing analysis APE: 037 Steam Generator (S/G) Tube Leak AA2. Ability to determine and interpret the following as they apply to the Steam Generator Tube Leak: AA2.09 System status, using independent readings from redundant Condensate air ejector exhaust monitor Objective: T.OA43-03, ANALYZE a given set of plant conditions and DETERMINE the required actions per 1/2BOA SEC-8, Steam Generator Tube Leak PROVIDE REFERENCE BCB-1 FIGURE 36



83 ID: 2017 NRC Q83A Points: 1.00 At 1000 A release of 0WX01T is initiated per BCP 400-TWX01, LIQUID RADWASTE RELEASE FORM FOR RELEASE TANK 0WX01T. The following setpoints were entered: HIGH Setpoint ALERT Setpoint • 0PR001, Liquid Radwaste Effluent

Monitor 1.55E-5 μCi/ml 1.29E-5 μCi/ml

• 0PR010, Station Blowdown Monitor 5.63E-6 μCi/ml 3.05E-6 μCi/ml At 1105 • 0PR001 is reading 3.45E-5 μCi/ml • 0PR010 is reading 5.85E-6 μCi/ml. • Both Monitor readings have been steady for the last 62 minutes. ___(1)___ should have terminated the release AUTOMATICALLY.

A NARS notification ____(2)____ be made.

A. 1. 0PR001, Liquid Radwaste Effluent Monitor

2. must

B. 1. 0PR001, Liquid Radwaste Effluent Monitor 2. need NOT

C. 1. 0PR010, Station Blowdown Monitor

2. must

D. 1. 0PR010, Station Blowdown Monitor 2. need NOT

Answer: A

Answer Explanation



The interlocks associated with 0PR001 will automatically terminate a liquid release. BCP 400-TWX01, LIQUID RADWASTE RELEASE FORM FOR RELEASE TANK 0WX01T requires the operator to manually terminate a release for Rad Monitor readings above the high alarm setpoint on 0PR010. EAL RU1 applies per EP-AA-1002 Addendum 3. RU1 Any release of gaseous or liquid radioactivity to the environment greater than 2 times the ODCM for 60 minutes or longer. Emergency Action Levels (EAL): Notes: 1. Reading on any of the following effluent monitors > 2 times high alarm setpoint established by a current radioactive release discharge permit for ≥ 60 minutes: 0PR001, Liquid Radwaste Effluent Monitor. Classification of RU1 requires the NARS notification. A. Correct. See above. B. Incorrect. See above. The first part is correct making that part of the distractor plausible. The second part is plausible because operators sometimes become focused on the document they are using, such as EP-AA-1002 Addendum 3 sheet BY 2-1. RU1 note states if the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer valid for classification purposes. The reading that they are using should have terminated the release. C. Incorrect. See above. The first part is plausible because BCP 400-TWX01, LIQUID RADWASTE RELEASE FORM FOR RELEASE TANK 0WX01T requires the operator to manually terminate a release for Rad Monitor readings above the high alarm setpoint on 0PR010. The second part is correct making that part of the distractor plausible. D. Incorrect. See above. The first part is plausible because BCP 400-TWX01, LIQUID RADWASTE RELEASE FORM FOR RELEASE TANK 0WX01T requires the operator to manually terminate a release for Rad Monitor readings above the high alarm setpoint on 0PR010. The second part is plausible because operators sometimes become focused on the document they are using, such as EP-AA-1002 Addendum 3 sheet BY 2-1. RU1 note states if the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer valid for classification purposes. The reading that they are using should have terminated the release. Meets K/A, the question requires the examinee to evaluate EP-AA-1002 Addendum 3 sheet BY 2-1 and determine that a NARS notification is required. SRO only: From the ES-401 Attachment 2: Screening for SRO-only linked to 10CFR55.43(b)(4) (Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions) SRO-only knowledge should not be claimed for questions that can be answered solely based on RO knowledge of radiological safety principles; e.g., RWP requirements, stay-time, DAC-hours, etc. Some examples of SRO exam items for this topic include:

- Analysis and interpretation of coolant activity, including comparison to emergency plan criteria and/or regulatory limits. YES Technical References: EP-AA-114, NOTIFICATIONS. EP-AA-114-F-01, PWR Release In Progress Determination Guidance, EP-AA-1002 Addendum 3 sheet BY 2-1.




System ID: 1743109 User-Defined ID: 2017 NRC Q83A Cross Reference Number: T.ZP-007

Topic: At 1000 A release of 0WX01T is initiated per BCP 400-TWX01, LIQUID RADWASTE RELEASE FORM FOR RELEASE

RO Importance: SRO Importance: 4.1 K/A: 059G2.4.30 Comments: New Question

License Level: SRO Cognitive Level: High, from multiple analysis required. APE: 059 Accidental Liquid Radwaste Release G2.4.30 Knowledge of events related to system operation/status that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator. Objective: T.ZP-007, ENSURE proper notification is made to Onsite and Offsite personnel. Provide EP-AA-1002 Addendum 3 sheet BY 2-1 to the candidates



84 ID: 2017 NRC Q84 Points: 1.00 There is a fire in the Upper Cable Spreading Room, • The control room has been evacuated. 1BOA PRI-5, CONTROL ROOM INACCESSIBILITY, has been implemented. • The RSP has been aligned per Attachment A of 1BOA PRI-5. There is NO indication for 1B Steam Generator level and pressure at the RSP. What actions are required per 1BOA PRI-5?

A. Dispatch an operator to Fire Hazards Panel to align needed indications.

B. Locally fail air and throttle 1AF005 flow control valves.

C. Feed affected steam generator between 60 GPM and 80 GPM.

D. Restore operation from the main control room per attachment D.

Answer: A

Answer Explanation



A. Correct.1BOA PRI-5 step 8 requires an operator be dispatched to the Fire Hazards Panel to align needed indications. B. Incorrect. 1BOA PRI-5 step 8 requires an operator be dispatched to the Fire Hazards Panel to align needed indications. Plausible because step 10 (Check SG Levels) RNO is locally fail air and throttle 1AF005 flow control valves if AF flow cannot be controlled. C. Incorrect. 1BOA PRI-5 step 8 requires an operator be dispatched to the Fire Hazards Panel to align needed indications. Plausible because step OAS page requires feeding affected steam generator between 60 GPM and 80 GPM when level is less than 45%. D. Incorrect. 1BOA PRI-5 step 8 requires an operator be dispatched to the Fire Hazards Panel to align needed indications. Plausible because OAS page does have a kick out to attachment D. Meets K/A, Requires examinee to have knowledge of system function for Control Room Evacuation, requires knowledge of the function of the Fire Hazards Panel and the Remote Shutdown Panel. Per Braidwood Byron FPR amendment 27 dtd December 2016: Instruments at the remote shutdown panel may not be available if ventilation is lost to the AEER. In this case, the safe shutdown instruments at the fire hazards panel will remain available and be utilized. This is contained on page 2.4-30 of the UFSAR. From the ES-401 Attachment 2 Figure 2: Screening for SRO-only linked to 10CFR55.43(b)(5) (Assessment and selection of procedures): 1) assessing plant conditions and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. Using the flowchart, this question can: • NOT be answered solely by knowing “systems knowledge”, i.e., how the system works, flowpath, logic, component location. No • NOT be answered solely by knowing immediate operator actions. No • NOT be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs. No • NOT be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure. No • CAN be answered with knowledge of ONE or MORE of the following: - Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. Yes Technical References: 1BOA PRI-5, CONTROL ROOM INACCESSIBILITY and 1BOSR XFP-R1, FIRE HAZARDS PANEL INSTRUMENTATION 18 MONTH SURVEILLANCE





Topic: There is a fire in the Upper Cable Spreading Room, The control room has been evacuated.


License Level: SRO Cognitive Level: High, requires multiple analysis. APE: 068 Control Room Evacuation G2.1.27 Knowledge of system purpose and/or function. Objective: T.OA16-03, ANALYZE a given set of plant conditions and DETERMINE the required actions per 0/1/2BOA PRI-5, Control Room Inaccessibility



85 ID: 2017 NRC Q85C Points: 1.00 Unit 2 has entered 2BFR-C.1, RESPONSE TO INADEQUATE CORE COOLING, and is currently attempting to restore core cooling. • The CETCs are reading 1205°F and slowly RISING. • RCS pressure is 1600 psig. • NO CV pumps can be started. • SI pumps are running aligned to the RWST. • PZR level is off-scale LOW. • NO Auxiliary Feedwater pumps can be started. • All S/G narrow range levels are reading 3%. 1. Per 2BFR-C.1, what action must be directed?

2. When taking the action required in 2BFR-C.1, Pressurizer level will ______.

A. 1. START available RCPs.

2. STABILIZE

B. 1. START available RCPs. 2. RISE

C. 1. OPEN all PZR PORVs and isolation valves.

2. STABLIZE

D. 1. OPEN all PZR PORVs and isolation valves. 2. RISE

Answer: D

Answer Explanation



Per BFR-C.1 step 16, the RCPs will be started ONLY if the S/Gs are providing an adequate heat sink for the RCS. With NO AF pumps running and S/G levels less than 14%, the RCPs will NOT be started. The RNO column of step 16 then directs that ALL PZR PORVs be opened. Opening the PORVs will release adjoin energy from the RCS which will lower RCS pressure. As RCS pressure is lowered, the SI pumps will eventually inject into the RCS (when pressure lowers to approximately 1500 psig). This will cause PZR level to RISE. Initially, pressurizer level will remain stable, off-scale low, because RCS pressure as listed in the question stem is higher than the shutoff head of the SI pumps. A. Incorrect See below. Starting RCPs is plausible because the note prior to step 16 states, Normal conditions are desired but NOT required for starting RCPs. The second part is plausible because the candidate may believe that at 1600 psig the SI pumps would inject into the RCS, as the CV pumps will. Plausibility is further enhanced because, if the crew started the RCP, the forced flow would remove core heat and could result in PZR level STABILIZING. This would occur when break flow would equal make-up flow since they believe that the SI pumps will be injecting into the RCS and the running RCPs would be pumping mass out of the core. B. Incorrect. See below. Plausible because the note prior to step 16 states, Normal conditions are desired but NOT required for starting RCPs. The second part is correct as initially SI pump flow will not be injected and when core cooling is restored and RCS pressure lowers, ECCS pumps would begin to deliver more flow to the RCS. C. Incorrect. See below. Plausible because the conditions for an RCP start are NOT MET due to SG level being less than 14% NR indication. The RNO portion of that step has the Operator then open the PZR PORVs to depressurize the RCS. The second part is plausible because the initial opening of the PORVs will result in a more break flow. This is caused by the initial break flow plus the additional flow caused by the opening of the PORV and PZR level could stabilize when break flow equals make-up flow. D. Correct. With the average of the10 highest CETCs reading 1205°F and SG levels at 3%, the SRO would transition to 2BFR-C.1 step 16, Check If RCPs Should Be Started. The conditions for an RCP start are NOT MET due to SG level being less than 14% NR indication. The RNO portion of that step has the Operator then open the PZR PORVs to depressurize the RCS. Since initially RCS pressure is higher than SOH of the SI pumps, Pzr level will remain stable initially until RCS pressure is reduced by the PORV opening, which is the second leak or the "varying leak size". As RCS pressure is lowered, the ECCS pumps begin to deliver more flow, inventory will begin to increase and will result in a rising pressurizer level. As pressure continues to lower the RH pumps will deliver flow at up to about 4000 gpm per pump. Meets the K/A, the question requires the examinee to trend PZR level when the leak size is increased by opening the PZR PORVs during 2BFR-C.1, RESPONSE TO INADEQUATE CORE COOLING. This is similar to the event at TMI, which is a safety significant event. Training on this is required yearly. From the ES-401 Attachment 2 Figure 2: Screening for SRO-only linked to 10CFR55.43(b)(5) (Assessment and selection of procedures): 1) assessing plant conditions and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed.

Using the flowchart, this question can: YES • NOT be answered solely by knowing “systems knowledge”,i.e., how the system works, flowpath, logic, component location. NO, procedural step question • NOT be answered solely by knowing immediate operator actions. NO, step is not immediate action • NOT be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs. NO • NOT be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure. NO, question asks detailed information on 1 procedure step • CAN be answered with knowledge of ONE or MORE of the following: - Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. YES, the candidate must assess plant conditions and then determine which portion of the procedure to use. Reference: 2BFR-C.1, WOG Background document FC-C




System ID: 2025440 User-Defined ID: 2017 NRC Q85C Cross Reference Number: T.FR02-03

Topic: Unit 2 has entered 2BFR-C.1, RESPONSE TO INADEQUATE CORE COOLING, and is currently attempting to


SRO Level High Cog Level due to analysis K/A: EPE074: Inadequate Core Cooling, EA2: Ability to determine or interpret the following as they apply to a Inadequate Core Cooling, .05: Trends in water levels of PZR and makeup storage tank caused by various sized leaks in the RCS Objective: T.FR02-03, Without the use of the procedure, STATE the basis for the actions described in steps, notes, and cautions for the C-Series Procedures.



86 ID: 2017 NRC Q86 Points: 1.00 The Unit is in MODE 1. SI Accumulator 'A' conditions: • Level: 46% • Pressure: 630 psig • Boron concentration: 2170 ppm Per Technical Specification 3.5.1, ACCUMULATORS, 1. Which procedure will be implemented?

2. What is the basis for this action?

A. 1. BOP SI-22, Raising SI Accumulator Level

2. Ensure ECCS criteria contained in 10 CFR 50.46 will be met following a LOCA.

B. 1. BOP SI-22, Raising SI Accumulator Level 2. Ensure subcriticality in a post LOCA environment.

C. 1. BOP SI-21, Changing SI Accumulator Boron Concentration

2. Ensure ECCS criteria contained in 10 CFR 50.46 will be met following a LOCA.

D. 1. BOP SI-21, Changing SI Accumulator Boron Concentration. 2. Ensure subcriticality in a post LOCA environment.

Answer: D

Answer Explanation



Per T.S. 3.5.1, The LCO on level is 31% to 63%. The LCO on pressure is 602 to 647 psig. The LCO on boron concentration is 2200 to 2400ppm. The accumulator is in LCO 3.5.1 Condition A, Boron concentration NOT within limits. Per the T.S. bases the reason for the boron concentration is to "assure reactor subcriticality in a post LOCA environment" This is contained on page 5 of B 3.5.1. A. Incorrect: The first part is incorrect but plausible because the specification states accumulator

level must be greater than or equal to 31%. With the level at 46%, the T.S. requirement is met. This distractor is plausible because when RWST level lowers to 46%, the operator switches to "recirc" mode of ECCS following an accident. The second part is incorrect but plausible as this is the bases for the ECCS system minimum water volume (i.e. level), on page 4.

B. Incorrect: The first part is incorrect but plausible because the specification states accumulator level must be greater than or equal to 31%. With the level at 46%, the T.S. requirement is met. This distractor is plausible because when RWST level lowers to 46% the operator switches to "recirc" mode of ECCS following an accident. The second part is plausible because it is correct per the conditions in the stem.

C. Incorrect: The first part is correct as the boron concentration is outside of the limits. The second part is incorrect but it is plausible because the bases document states the overall LCO helps to ensure that ECCS criteria as specified by 10 CFR 50.46 are met. The bases document specifically mentions the minimum boron concentration and that it assures subcriticality.

D. Correct as stated in the explanation. The question meets the K/A by testing the candidate on low boron concentration in the Safety Injection System and has them selecting a procedure to correct the condition. From the ES-401 Attachment 2 Figure 2: Screening for SRO-only linked to 10CFR55.43(b)(5) (Assessment and selection of procedures): 1) assessing plant conditions and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. Using the flowchart, this question can: YES • NOT be answered solely by knowing “systems knowledge”,i.e., how the system works, flowpath, logic, component location. NO, • NOT be answered solely by knowing immediate operator actions. NO • NOT be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs. NO • NOT be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure. NO • CAN be answered with knowledge of ONE or MORE of the following: - Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. YES, examinee must assess the SI accumulators and determine which procedure will correct the out of specification condition. From the ES-401 Attachment 2 Figure 1: Screening for SRO-only linked to 10 CFR 55.43(b)(2) (Tech Specs): • Can NOT be answered by knowing less than 1 hour Tech Specs. • Can NOT be answered by knowing information listed "above-the-line". • Can NOT be answered by knowing the TS Safety Limits. • Does involve one or more of the following for TS, TRM or ODCM: - Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1). - Application of generic LCO requirements (LCO 3.0.1 thru 3.0.7 and SR 4.0.1 thru 4.0.4). - Knowledge of TS bases that is required to analyze TS required actions and terminology. YES Technical References: Tech Spec 3.5.1 and Bases 3.5.1




System ID: 1742771 User-Defined ID: 2017 NRC Q86 Cross Reference Number: S.EC1-16

Topic: The Unit is in MODE 1. SI Accumulator 'A' conditions:

RO Importance: SRO Importance: 3.9 K/A: 006A2.10 Comments: New Question

SRO Level Low Cog Level (Memory) K/A: 006 Emergency Core Cooling System (ECCS), A2: Ability to (a) predict the impacts of the following malfunctions or operations on the ECCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations, 10: Low boron concentration in the SIS Objective: From ECCS Lesson Plan (S-58) ANALYZE a given set of plant conditions and DETERMINE ECCS Tech Spec Operability requirements. (S.EC1-16) AND DISCUSS the applicable ECCS Technical Specifications Bases. (S.EC1-12)



87 ID: 2017 NRC Q87 Points: 1.00 Unit 1 is in MODE 4. • 1B RCFC is INOPERABLE due to motor failure. • It will take 15 days for the new motor to be delivered. Unit 1 must be in MODE 5 within a MAXIMUM of ...

A. 7 days

B. 7 days and 6 hours

C. 8 days and 12 hours

D. 14 days

Answer: C

Answer Explanation A. Incorrect. See below. Plausible because 7 days is the completion time for condition C. B. Incorrect. See below. Plausible because 7 days and 6 hours is the completion time for condition D1. C. Correct. LCO 3.6.6, Two containment spray trains and two containment cooling trains shall be OPERABLE. Condition C requires the restoration of the inoperable train in 7 days. If the completion time for condition C is not met condition D is entered. Condition D requires the unit to be in Mode 3 in 6 hours and Mode 5 in 36 hours for a total time to Mode 5 of 8 days and 12 hours. D. Incorrect. See above. Plausible because 14 days is the second part of the completion time for condition C. Meets K/A, examinee must have knowledge of limiting conditions for operations for the Containment Cooling System. But, safety limits are required RO level knowledge. Strong tie to LCO knowledge. SRO level: From the ES-401 Attachment 2 Figure 1: Screening for SRO-only linked to 10 CFR 55.43(b)(2) (Tech Specs): • Can NOT be answered by knowing less than 1 hour Tech Specs. No • Can NOT be answered by knowing information listed "above-the-line". No • Can NOT be answered by knowing the TS Safety Limits. No • Does involve one or more of the following for TS, TRM or ODCM: - Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1). Yes Technical References: LCO 3.6.6, Two containment spray trains and two containment cooling trains shall be OPERABLE. B 3.6.6 Containment Spray and Cooling Systems.




System ID: 1742784 User-Defined ID: 2017 NRC Q87 Cross Reference Number: S.VP1-11

Topic: Unit 1 is in Mode 4. 1B RCFC is INOPERABLE due to motor failure. It will take 15 days for the new


License Level: SRO Cognitive Level: High, multiple analysis required. 022 Containment Cooling System (CCS) G2.2.22 Knowledge of limiting conditions for operations and safety limits. Objective: S.VP1-11, Given a set of plant conditions, Given applicable reference material, ANALYZE a given set of plant conditions and DETERMINE Containment Ventilation Tech Spec/TRM operability requirements. PROVIDE T.S. 3.6.6 IN THE CANDIDATE REFERENCE PACKAGE



88 ID: 2017 NRC Q88A Points: 1.00 Unit 1 is at 100% • CC surge tank is 70% and rising. • The crew has entered 1BOA PRI-6, COMPONENT COOLING MALFUNCTION. • Both 1PR09J and 0PR09J, Unit 1 and Unit 0 CC Heat Exchanger Outlet Process Radiation Monitors,

are at the high alarm setpoint and RISING. 1. Per 1BOA PRI-6, the local operator will isolate the ...

2. This action could limit the ability to ...

A. 1. High Radiation Sample Sink (HRSS) sample coolers.

2. measure RWST Boron Concentration.

B. 1. High Radiation Sample Sink (HRSS) sample coolers. 2. measure Reactor Coolant System Boron Concentration.

C. 1. Seal Water Heat Exchanger.

2. maintain VCT level.

D. 1. Seal Water Heat Exchanger. 2. prevent CV pump cavitation.

Answer: B

Answer Explanation



A. Incorrect. The first part is correct making this part of the distractor plausible. The second part is incorrect but plausible as the SFP contains radioactive water that is sampled by the Chemistry department.

B. Correct. Step 6 of attachment B directs locating and isolating leakage from HRSS sample

coolers. If the HRSS sample sink is isolated RCS samples could no longer be taken or analyzed.

C. Incorrect. The first part is plausible as this is a source of radioactive RCS and this heat

exchanger is cooled by CC. Seal water HX pressure is essentially at VCT pressure so this would not leak into the CC system which is about 110 psig. The absence of seal water return flow can have an effect on VCT level as flow from the heat exchanger is direct back to the VCT, making this distractor plausible.

D. Incorrect. The first part is plausible as this is a source of radioactive RCS. Seal water HX

pressure is essentially at VCT pressure so this would not leak into the CC system which is about 110 psig. The absence of seal water return flow can have an effect on VCT level as flow from the heat exchanger is direct back to the VCT, making this distractor plausible. Lowering VCT level could result in CV pump cavitation, making this distractor plausible.

Meets K/A, the question requires the examinee to know the local auxiliary operator task taken per 1BOA PRI-6 for the high rad condition on 1PR09J and 0PR09J. Also, affects this action will have on the plant. SRO only: From the ES-401 Attachment 2 Figure 2: Screening for SRO-only linked to 10CFR55.43(b)(5) (Assessment and selection of procedures): 1) assessing plant conditions and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. Using the flowchart, this question can: • NOT be answered solely by knowing “systems knowledge”, i.e., how the system works, flowpath, logic, component location. No • NOT be answered solely by knowing immediate operator actions. No • NOT be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs. No • NOT be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure. No • CAN be answered with knowledge of ONE or MORE of the following: - Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. Yes, the examinee will need to evaluate the high rad condition and select the appropriate actions to take per 1BOA PRI-6. Technical References: 1BOA PRI-6, COMPONENT COOLING MALFUNCTION.





Topic: Unit 1 is at 100% CC surge tank is 70% and rising. The crew has entered 1BOA PRI-6, COMPONENT

RO Importance: SRO Importance: 4.0 K/A: 073G2.4.35 Comments: New question

License Level: SRO Cognitive Level: High, multiple analysis. 073 Process Radiation Monitoring (PRM) System 2.4.35 Knowledge of local auxiliary operator tasks during an emergency and the resultant operational effects. Objective: T.OA17-03, ANALYZE a given set of plant conditions and DETERMINE the required actions per 1/2BOA PRI-6, CCW Malfunction



89 ID: 2017 NRC Q89A Points: 1.00 On Unit 1: • 1A CV pump is RUNNING. • 1B SX pump discharge pressure has LOWERED to 82 psig. • EO reports a large amount of water leaking in the Aux Bldg. • Floor drains are keeping up with the leakage. • SX basin level is currently 85% and stable with all available makeup paths operating. • 1A CV pump bearing temperature indications on PPC are in alarm and RISING. Which of the following must be directed?

A. Shutdown 1B SX pump.

B. Manually TRIP the Unit 1 Reactor.

C. Dispatch an operator to close 0SX007 and 1SX007, CC HX SX Outlet valves.

D. Dispatch an operator to align FP cooling to the 1A CV pump.

Answer: D

Answer Explanation



A. Incorrect. 0BOA PRI-8 would only direct shutting down SX if there was imminent danger to vital equipment. Plausible because 0BOA PRI-8 does direct shutting down all SX when there is imminent danger of flooding vital equipment. B. Incorrect. Tripping the reactor is plausible as that action is performed in 0BOA PRI-7, step 1 when no SX pumps are operating. C. Incorrect. 1BOA PRI-7 step 2 RNO directs throttling 0SX007 and 1SX007 while maintaining system temperatures. Plausible because 1BOA PRI-7 step 2 RNO does direct manipulating 0SX007 and 1SX007. D. Correct. With the given conditions, the CV pump needs to have an alternate source of cooling water for the oil coolers. 1BOA PRI-7 step 2 RNO directs FP to be lined up per Attachment B. Also, 0BOA PRI-8 would direct FP to be lined up per Attachment B of 1BOA PRI-7. Meets K/A, examinee must have the ability to predict the impacts of a SX leak, as indicated by low SX header pressure, on the SX system. Then, based on those predictions, use procedures to mitigate the consequences of the leak. SRO Only: From the ES-401 Attachment 2 Figure 2: Screening for SRO-only linked to 10CFR55.43(b)(5) (Assessment and selection of procedures): 1) assessing plant conditions and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. Using the flowchart, this question can: • NOT be answered solely by knowing “systems knowledge”, i.e., how the system works, flowpath, logic, component location. No • NOT be answered solely by knowing immediate operator actions. No • NOT be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs. No • NOT be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure. No • CAN be answered with knowledge of ONE or MORE of the following: - Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. Yes, the examinee will need to evaluate the conditions in the stem and then determine the actions required per 1BOA PRI-7. Technical References: 1BOA PRI-7, ESSENTIAL SERVICE WATER MALFUNCTION. 0BOA PRI-8, AUXILIARY BUILDING FLOODING.




System ID: 1845771 User-Defined ID: 2017 NRC Q89A Cross Reference Number: S.SX1-07-A

Topic:

On Unit 1: 1A CV pump RUNNING 1B SX pump discharge pressure has LOWERED to 82 psig. EO reports

RO Importance: SRO Importance: 3.1 K/A: 076A2.02 Comments: Modified LORT Bank question (1599723)

License Level: SRO Cognitive Level: High, multiple analysis required. 076 Service Water System (SWS) A2 Ability to (a) predict the impacts of the following malfunctions or operations on the SWS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: A2.02 Service water header pressure Objective: T.OA18-03, ANALYZE a given set of plant conditions and DETERMINE the required actions per 1/2BOA PRI-7/0BOA PRI-8, Essential Service Water Malfunction.



90 ID: 2017 NRC Q90 Points: 1.00 At 1000 Both units at 100% power. • 1A, 1B and 1C RCFCs are running in HIGH Speed. A fire breaks out in the Main Control Room. • After performing the applicable steps of 1BOA PRI-5, CONTROL ROOM INACCESSIBILITY, the crew

exits the MCR. • BOTH Units' Remote Shutdown Panels are activated. At 1030 • The TSC directs starting the 1D RCFC in HIGH Speed. The US will direct _________ (1) _________ to perform this action.

This action will ______________ (2) ______________.

A. 1. the Equipment Operator

2. lower containment airborne activity.

B. 1. the Equipment Operator 2. lower containment temperature.

C. 1. the Nuclear Station Operator

2. lower containment airborne activity.

D. 1. the Nuclear Station Operator 2. lower containment temperature.

Answer: D

Answer Explanation



Per 1BOA PRI-5, Attachment A step 1 for 1PL04J the control switches for the RCFC 1A and 1C HIGH SPEED are placed in LOCAL. The same action is performed on RCFC 1B and 1D HIGH SPEED per step 2. There is a start/stop C/S that may be manipulated to start and stop each of these fans (in high speed only). This action will be performed by the Unit NSO assigned to that panel. The slow speed RCFC breakers have no controls at the remote S/D panels. Per T.S. Bases 3.6.5 (page 1): The purpose of the containment cooling system is to limit temperature in the containment to less than design values. A. Incorrect: The Equipment Operator is plausible because the EO is dispatched during this

procedure for several in-plant functions. This is also plausible as the procedure directs the EO to LOCALLY operate several circuit breakers, if required, using procedure BOA ELEC-5. If this were the Low Speed RCFC breaker, this procedure would be utilized. The second part is incorrect but plausible as this system along with Containment Spray is designed to reduce the release of fission product radioactivity from the containment to the environment. This is also plausible as within the containment the Post LOCA Filter units also circulate air and do remove radioactivity from the atmosphere.

B. Incorrect: The Equipment Operator is plausible because the EO is dispatched during this procedure for several in-plant functions. This is also plausible as the procedure directs the EO to LOCALLY operate several circuit breakers, if required, using procedure BOA ELEC-5. If this were the Low Speed RCFC breaker, this procedure would be utilized. The second part is correct which makes this part of the distractor plausible.

C. Incorrect: The first part is correct, making this part of the distractor plausible. The second part is incorrect but plausible as this system along with Containment Spray is designed to reduce the release of fission product radioactivity from the containment to the environment. This is also plausible because the containment the Post LOCA Filter units also circulate air and do remove radioactivity from the atmosphere.

D. Correct as described above. Question meets the K/A by placing the examinee in a situation where the E-plan is activated because of a loss of control room indications of 15 minutes or longer and deciding which person will perform the task of increasing the effects of containment cooling. Starting another RCFC will increase containment cooling. Question is SRO level (see below) and because SRO is responsible for assigning tasks to shift personnel. From the ES-401 Attachment 2 Figure 2: Screening for SRO-only linked to 10CFR55.43(b)(5) (Assessment and selection of procedures): 1) assessing plant conditions and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. Using the flowchart, this question can: • NOT be answered solely by knowing “systems knowledge”, i.e., how the system works, flowpath, logic, component location. No • NOT be answered solely by knowing immediate operator actions. No • NOT be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs. No • NOT be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure. No • CAN be answered with knowledge of ONE or MORE of the following: - Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures. YES References: 1BOA PRI-5, B 3.6.5, 6E-1-4030 VP08




System ID: 1743059 User-Defined ID: 2017 NRC Q90 Cross Reference Number: S.VP1-03

Topic:

At 1000 Both units at 100% power. 1A, 1B and 1C RCFCs are running in HIGH Speed. A fire breaks out


SRO Level Low Cog Level (Memory) K/A: Sys103: Containment, Generic 2.4: Emergency Procedures/Plan, 2.4.34 Knowledge of RO tasks performed outside the main control room during and emergency and the resultant operational effects. Objective: State the design bases of the Reactor Containment Fan Cooler System (S.VP1-03)



91 ID: 2017 NRC Q91 Points: 1.00 The Unit is at 100% power. • Pressurizer Level, LVL CONT CH SELECT, switch is in CH459-460 position • Annunciator 1-9-D3, CHG LINE FLOW HIGH LOW, is in alarm and Charging flow is HIGH. A scan of the control boards reveals the following:

1. To mitigate the event enter ...

2. Direct action to ...

A. 1. 1BOA INST-2, OPERATION WITH A FAILED INSTRUMENT CHANNEL.

2. Trip bistables for OT∆T, reactor trip and runback.

B. 1. 1BOA INST-2, OPERATION WITH A FAILED INSTRUMENT CHANNEL. 2. establish letdown per 1BOA ESP-2, RE-ESTABLISHMENT OF CV LETDOWN.

C. 1. 1BOA PRI-1, EXCESSIVE PRIMARY PLANT LEAKAGE.

2. verify leak is NOT in charging header downstream of CV121, Centrifugal Charging Pumps Flow Control Valve.

D. 1. 1BOA PRI-1, EXCESSIVE PRIMARY PLANT LEAKAGE.

2. establish 75 gpm letdown.

Answer: B

Answer Explanation



The correct answer is based on a failure of the controlling PZR Level channel 459 failing LOW. When this occurs, the Master PZR LVL CONT, 1LK-459, sensing a low level will raise the output signal to open 1CV-121, CENT CHG PMPS FLOW CONTROL VLV. This will cause the CHG LINE FLOW HIGH LOW annunciator to light when greater than 150 gpm is achieved. BAR 1-9-D3 will be referred to which directs entry to 1BOA PRI-1 or to 1BOA INST-2. Instrument 2 would be the correct choice based on the other 2 NOT/NOP calibrated level channels being at the normal level for this power level. Letdown will have isolated because the controlling level channel is < 17%. BOA INST-2 will direct letdown to be re-established. A. Incorrect: The first part is correct thereby making this distractor plausible. The second part is incorrect but is plausible because these bistables would be tripped if the failure were a PZR pressure instrument failure. This is also plausible because you would enter the same procedure (BOA INST-2) but a different attachment. B. Correct as described above. C. Incorrect: The first part is incorrect but is plausible as actions contained in the referenced Annunciator Response procedure direct the operator to this procedure. The second part is incorrect but is plausible because this action would be carried out if BOA PRI-1 had been entered. D. Incorrect: The first part is incorrect but is plausible as actions contained in the referenced Annunciator Response procedure direct the operator to this procedure. The second part is incorrect, but distractor plausible because this action is contained in the procedure to address a primary to secondary leak, which could be indicated by lowering PZR level. This is also plausible because the correct answer and this distractor both have letdown manipulation evolutions. The question meets the K/A by placing the candidate in a condition in which excessive charging flow has occurred. This requires the candidate to make a distinction between 2 possible causes of the excessive charging flow and then choose a procedure which will mitigate the event. From the ES-401 Attachment 2 Figure 2: Screening for SRO-only linked to 10CFR55.43(b)(5) (Assessment and selection of procedures): 1) assessing plant conditions and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. Using the flowchart, this question can: • NOT be answered solely by knowing “systems knowledge”, i.e., how the system works, flowpath, logic, component location. No • NOT be answered solely by knowing immediate operator actions. No • NOT be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs. No • NOT be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure. No • CAN be answered with knowledge of ONE or MORE of the following: - Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. Yes From the ES-401 Attachment 2 Figure 1: Screening for SRO-only linked to 10 CFR 55.43(b)(2) (Tech Specs): • Can NOT be answered by knowing less than 1 hour Tech Specs. No • Can NOT be answered by knowing information listed "above-the-line". No • Can NOT be answered by knowing the TS Safety Limits. No • Does involve one or more of the following for TS, TRM or ODCM: Yes - Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1). No - Application of generic LCO requirements (LCO 3.0.1 thru 3.0.7 and SR 4.0.1 thru 4.0.4). No - Knowledge of TS bases that is required to analyze TS required actions and terminology. Yes References: BAR 1-9-D3, T.S. 3.4.9 (PZR level) and Bases





Topic: The Unit is at 100% power. Pressurizer Level, LVL CONT CH SELECT, switch is in CH459-460 position

RO Importance: SRO Importance: 3.2 K/A: 011A2.02 Comments: New Question

SRO Level High Cog Level K/A: 011 Pressurizer Level Control System, A2: Ability to (a) predict the impacts of the following malfunctions or operations on the PZR LCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations, .02 Excessive charging Objective: DISCUSS the bases for the pressurizer Tech. Specs. (S.RY1-28)



92 ID: 2017 NRC Q92B Points: 1.00 The 0E Waste Gas Decay Tank was recently placed in Storage and was sampled by Chemistry for a pending release. • The tank pressure is currently 92 psig The following parameters were reported by Chemistry: • Activity level of 9.0 E+4 Curies • Oxygen level of 1.5% • Hydrogen level of 0.7% What are the SRO's required mitigating strategy and impact on the plant based on the above parameters?

Mitigating Strategy Impact

A. Purge or Dilute the 0E GDT Reduces explosive mixture of the 0E GDT. per BCP 400-TWASTE GAS, GASEOUS EFFLUENT RELEASE FORM TYPE: WASTE GAS DECAY TANK

B. Release 0E GDT per Lowers background levels in the plant

BOP GW-6, REALIGNMENT and lowers the curie content of 0E GDT. OF GAS DECAY TANKS

C. Transfer some of the Lowers the curie content of 0E GDT

0E GDT contents to while raising the curie content of the another GDT per 0BOA- other tank. RAD-3, DECAY TANK HIGH ACTIVITY

D. Verify Oxygen and Hydrogen Determines valid explosive mixture

readings on 0GW01J per exists in 0E GDT. 0BOA-PRI-9, OXYGEN/ HYDROGEN EXPLOSIVE MIXTURE

Answer: C

Answer Explanation



A. Incorrect. See below. Plausible because diluting the tank contents would be performed as a response to a High Explosive concentration of Hydrogen and Oxygen levels per 0BOA PRI-9. Also, BCP 400-TWASTE GAS, GASEOUS EFFLUENT RELEASE FORM TYPE: WASTE GAS DECAY TANK, would be utilized to perform this evolution. B. Incorrect. See below. Plausible because releasing the contents of the GDT per BOP GW-6 would reduce the curie content of the tank. Also, this would lower the background levels in the plant. C. CORRECT. Waste Gas Decay Tank curie content of greater than 5.0 E+4 Curies requires entry into 0BOA RAD-3, DECAY TANK HIGH ACTIVITY. Step 2.b lowers the activity by transferring some of the contents to another GDT. The impact of this transfer would be to lower the curie content of 0E GDT while raising another tanks curie content. Once the curie content is lowered to less than 5.0 E+4 curies then the tank is place into storage in preps for release. D. Incorrect. See above. Plausible because Oxygen and Hydrogen levels are monitored on 0GW01J, by Chemistry, for the in-service GDT and while releasing a GDT. However, the GDT Oxygen and Hydrogen levels are below the Explosive mixture limits and do not require any validation. Meets K/A – Examinee must have knowledge of the procedures for Waste Gas Disposal System (0BOA RAD-3, BOP GW-6, 0BOA PRI-9 and BCP 400-TWASTE GAS, GASEOUS EFFLUENT RELEASE FORM TYPE: WASTE GAS DECAY TANK ). They must have knowledge of the entry conditions and the mitigating strategy’s contained in the procedure and what the impact to the plant for taking those actions. The candidate must assess Rad levels and H2 and 02 levels based chemistry reports of station monitors readings. With this info the candidate must assess what conditions exist and what procedure to use to mitigate the issue. SRO only: From the ES-401 Attachment 2 Figure 2: Screening for SRO-only linked to 10CFR55.43(b)(5) (Assessment and selection of procedures): 1) assessing plant conditions and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. Using the flowchart, this question can: • NOT be answered solely by knowing “systems knowledge”, i.e., how the system works, flowpath, logic, component location. No • NOT be answered solely by knowing immediate operator actions. No • NOT be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs. No • NOT be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure. No • CAN be answered with knowledge of ONE or MORE of the following: - Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. Yes, the examinee will need to know what the entry conditions are as well as what each step is doing to be able to determine which procedure to use and what the mitigative strategy and impact on the plant are. Technical References: 0BOA RAD-3, DECAY TANK HIGH ACTIVITY BOP GW-6, REALAIGNMENT OF GAS DECAY TANKS 0BOA-PRI-9, OXYGEN/HYDROGEN EXPLOSIVE MIXTURE BCP 400-TWASTE GAS, GASEOUS EFFLUENT RELEASE FORM TYPE: WASTE GAS DECAY TANK




System ID: 2025430 User-Defined ID: 2017 NRC Q92B Cross Reference Number: 7D.OA-025-A

Topic: The 0E Waste Gas Decay Tank was recently placed in Storage and was sampled by Chemistry


License Level: SRO Cognitive Level: High, multiple analysis required. 071 Waste Gas Disposal System (WGDS) A2 Ability to (a) predict the impacts of the following malfunctions or operations on the Waste Gas Disposal System ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: A2.02 Use of waste gas release monitors, radiation, gas flow rate, and totalizer. Objective: 7D.OA-025-A Given a set of plant conditions indicating a High Gas Decay Tank Activity condition, EVALUATE operator response and DETERMINE appropriate actions.



93 ID: 2017 NRC Q93A Points: 1.00 The Unit is at 100% power. Natural Draft Cooling Tower Flume temperature has reached 96°F. BOP CW-25, NATURAL DRAFT COOLING TOWER OPERATION, has been entered. Per BOP CW-25, which of the following Abnormal Operating Procedures must be referenced if HIGH component temperatures occur?

A. BOA PRI-6, COMPONENT COOLING MALFUNCTION.

B. BOA PRI-7, ESSENTIAL SERVICE WATER MALFUNCTION.

C. BOA TG-1, TURBINE HIGH VIBRATION, ECCENTRICITY, OR DIFFERENTIAL

EXPANSION.

D. BOA TG-4, STATOR WATER HIGH CONDUCTIVITY.

Answer: C

Answer Explanation



Per BOP CW-25, NATURAL DRAFT COOLING TOWER OPERATION, Limitations and Action section informs the reader of actions to take based on high CW temperatures. This is included in section E.6. The main body of the procedure step F.1, warm weather operations sub step c. directs the operator to refer to BOA TG-1 for temperature limits. This is the tie to "explain and apply system limits and precautions". Step 6 of the TG-1 procedure has the operator check all TG support system temperatures "normal". A. Incorrect: BOA PRI-6 is plausible as CC is cooled by SX and SX is supplied by make-up from

the same source as Circ Water. B. Incorrect: BOA PRI-7 is plausible as SX receives makeup water from the source as the Circ.

Water system. C. Correct as described above. D. Incorrect: BOA TG-4 is plausible as WS cools stator water. Meets the K/A, the question requires the examinee to know, with a flume temperature of 96°F, BOP CW-25 will direct the operator to refer to BOA TG-1. Question is SRO Level based on requiring detailed information contained in the main body of the procedure. From the ES-401 Attachment 2 Figure 2: Screening for SRO-only linked to 10CFR55.43(b)(5) (Assessment and selection of procedures): 1) assessing plant conditions and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. Using the flowchart, this question can: YES • NOT be answered solely by knowing “systems knowledge”,i.e., how the system works, flowpath, logic, component location. NO, procedural step question • NOT be answered solely by knowing immediate operator actions. NO, step is not immediate action • NOT be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs. NO • NOT be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure. NO, question asks detailed information on 1 procedure step • CAN be answered with knowledge of ONE or MORE of the following: - Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. YES, the candidate must assess plant conditions and then determine which portion of the procedure to use. Reference: BOP CW-25 NATURAL DRAFT COOLING TOWER OPERATION




System ID: 1799111 User-Defined ID: 2017 NRC Q93A Cross Reference Number: S.CW1-10 C B

Topic:

The Unit is at 100% power. Natural Draft Cooling Tower Flume temperature has reached 96°F.


SRO Level High Cog Level due to multiple analysis. K/A: 075 Circulating Water System, G.2.1: Conduct of Operations, .32 Ability to explain and apply system limits and precautions. Objective: DESCRIBE the operation of the Natural Draft Cooling Towers including: c. Temperature limitations. (S.CW1-10-C B)



94 ID: 2017 NRC Q94A Points: 1.00 Per BOP AP-13, PLACING A GROUND TEST DEVICE IN A 4160V OR 6900V CUBICLE AND/OR GROUNDING A GROUND TEST DEVICE IN A 4160V OR 6900V CUBICLE, 1. __________ (1)__________ is required to be present during the Ground Test Device manipulation.

AND

2. An HVS qualified management individual ___ (2) ___ required to be present during the Ground Test Device manipulation.

A. 1. An individual qualified in CPR and First Aid

2. is

B. 1. An individual qualified in CPR and First Aid 2. is NOT

C. 1. A EMD Technician qualified on Ground Test Device operation,

2. is

D. 1. A EMD Technician qualified on Ground Test Device operation, 2. is NOT

Answer: A

Answer Explanation



Per BOP AP-13, PLACING A GROUND TEST DEVICE IN A 4160V OR 6900V CUBICLE AND/OR GROUNDING A GROUND TEST DEVICE IN A 4160V OR 6900V CUBICLE, Prerequisites section, #1: A management individual, qualified in High Voltage Switching, shall be present to verify proper GTD Operations. In the precautions section of the procedure, #10: A second person qualified in CPR and First Aid shall be present or within 4 minutes access time when employees are performing work on or near exposed lines or equipment energized at 50 volts or more. A. Correct as described above. B. Incorrect. The first part is correct making this part of the distractor plausible. The second part is incorrect but plausible because the examinee may conclude that with an individual qualified in CPR and First Aid present the HVS qualified management individual is not required. C. Incorrect: The first part is incorrect but plausible because there are several instances in which maintenance works alongside operating in the day to day operation of the plant. The second part is correct making this part of the distractor plausible. D. Incorrect: The first part is incorrect but plausible because there are several instances in which maintenance works alongside operating in the day to day operation of the plant. The second part is incorrect but plausible because the examinee may conclude that with an EMD Technician qualified on Ground Test Device present the HVS qualified management individual is not required. Question meets the K/A by testing the candidate on electrical safety practices to be used when racking manipulating a ground test device, an electrical device. The question is SRO Level based on responsibilities of an SRO, i.e. the SROs job. From the ES-401 Attachment 2 Figure 2: Screening for SRO-only linked to 10CFR55.43(b)(5) (Assessment and selection of procedures): 1) assessing plant conditions and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. Using the flowchart, this question can: YES • NOT be answered solely by knowing “systems knowledge”,i.e., how the system works, flowpath, logic, component location. NO, procedural step question • NOT be answered solely by knowing immediate operator actions. NO, step is not immediate action • NOT be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs. NO • NOT be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure. NO, question asks detailed information on 1 procedure step • CAN be answered with knowledge of ONE or MORE of the following: - Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. YES, the candidate must assess plant conditions and then determine which portion of the procedure to use. Reference: BOP AP-13




System ID: 1799129 User-Defined ID: 2017 NRC Q94A Cross Reference Number: PA-21

Topic: Per BOP AP-13, PLACING A GROUND TEST DEVICE IN A 4160V OR 6900V CUBICLE AND/OR GROUNDING A GROUND

RO Importance: SRO Importance: 3.6 K/A: G2.1.26 Comments: New question

SRO Level Low Cog Level G2: Generic, .1: Conduct of Operations, .26: Knowledge of industrial safety procedures (such as rotating equipment, electrical, high temperature, high pressure, caustic, chlorine, oxygen, and hydrogen). Objective: HVS: PA-21, Given a set of conditions regarding 4.16 or 6.9 KV breakers or ground/test devices, ANALYZE the situation and DETERMINE the required actions



95 ID: 2017 NRC Q95A Points: 1.00 Which of the following evolutions requires a Licensed Supervisor with an active SRO License to be present at the refuel cavity with NO other concurrent responsibilities?

A. Unlatching the control rod drive shafts.

B. Refueling Machine operability check outs.

C. Withdrawing the MIDS thimble tubes.

D. Lifting the reactor vessel head.

Answer: A

Answer Explanation A. Correct. Per OU-AP-200 the Licensed Supervisor with an active SRO License must be present at the refuel cavity with no other concurrent responsibilities whenever a core alteration is in progress. Per 1BGP 100-6 it states that unlatching the control rod drive shaft is considered to be a core alteration. B. Incorrect. Plausible because 1BOSR 9.c.1-2, Unit One Refueling Machine (100 Hours Prior) Operability Surveillance, is directed to be performed per 1BGP 100-6T4A. Also, to successfully complete this surveillance, a load test is performed. This may lead the examinee to conclude that a Licensed Supervisor with an active SRO License needs to be present at the refuel cavity with NO other concurrent responsibilities. C. Incorrect. Plausible because withdrawing the MIDS thimble tubes is directed to be performed per 1BGP 100-6. Also, the MIDS thimble tubes do extend up into the which could lead the examinee to concluded that a Licensed Supervisor with an active SRO License needs to be present at the refuel cavity with NO other concurrent responsibilities. D. Incorrect. Plausible because is directed to be performed per 1BGP 100-6. Also, the examinee may conclude disassembling the reactor is an actual core alteration, which would require a Licensed Supervisor with an active SRO License to be present at the refuel cavity with NO other concurrent responsibilities. Meets K/A, examinee must have knowledge of procedures and limitations involved in core alterations. SRO level because the SRO is the one responsible for having an understanding of what procedure requirements and limitations are required with core alterations, and also when a core alteration is occurring. From the ES-401 Attachment 2: Screening for SRO-only linked to 10CFR55.43(b)(6) (Procedures and limitations involved in initial core loading, alterations in core configuration, control rod programming, and determination of various internal and external effects on core reactivity) YES, requires the examinee to know the requirement for core alts. Technical References: 1BGP 100-6, Refueling Outage, OU-AP-200, Administrative Controls during Fuel Handling Activities For Byron and Braidwood.




System ID: 1845762 User-Defined ID: 2017 NRC Q95A Cross Reference Number: T.GP06-05

Topic: Knowledge of procedures and limitations involved in core alterations

RO Importance: SRO Importance: 4.1 K/A: G2.1.36 Comments: <QQ 1138040(1412)><<Bank Question from BWD 2014 NRC

Exam Q 94 (ID: SG1027-N14-94) License Level: SRO Cognitive Level: High, multiple analysis required 2.0 GENERIC KNOWLEDGES AND ABILITIES 2.1.36 Knowledge of procedures and limitations involved in core alterations. Objective: T.GP06-05, ANALYZE a given set of plant conditions and determine the required operator actions per GP 100-6, Refueling Outage



96 ID: 2017 NRC Q96C Points: 1.00 Per WC-AA-101, ONLINE WORK CONTROL PROCESS, which of the following are responsibilities of Operations Shift Management? 1. Audits the work planning and scheduling processes to ensure maintenance is completed. 2. Ensures appropriate actions are taken to mitigate online risk. 3. Responsible for declaring Structure, System, and Component (SSC) functionality/availability. 4. Ensures project critiques are performed for projects requiring a fragnet.

A. 1 and 2

B. 2 and 3

C. 3 and 4

D. 4 and 1

Answer: B

Answer Explanation



Per WC-AA-101, ONLINE WORK CONTROL PROCESS, Operations shift management is responsible for (among other items), ensuring appropriate actions are taken to mitigate online risk and for declaring Structure, System, and Component (SSC) functionality/availability. A. Incorrect: The first part is incorrect but plausible because this action is required to be

performed per the WC-AA-101-109. The second part is correct making this part of the distractor plausible.

B. Correct as described above. Before any work takes place on any SSC's at the station the

people performing the work, as a minimum, must check in with WEC and then the WEC will direct the worker(s) into the Control Room, if they feel that it's appropriate. It is imperative that shift personnel are aware of the status of their unit at all times. This answer is correct because Shift personnel determine OPERABILITY of a component, if applicable. Shift personnel will also determine the "risk" to the unit by determining the effects of a loss of a piece of equipment to the safety functions when taken in combination with other equipment that is out of service.

C. Incorrect: The first part is correct making this part of the distractor plausible. The second part

is incorrect but plausible as this action is performed by the procedure, but it is performed by the Cycle or Work Week Manager.

D. Incorrect: The first part is incorrect but plausible as this action is performed by the procedure

but it is performed by the Cycle or Work Week Manager. The second part is incorrect but plausible as this action is performed per the procedure, but it is performed by the Site Maintenance Rule Coordinator. Plausibility is further enhanced because the System Health Report is a document on shift which annotates (a)(1) systems.

Meets the K/A, the question requires the examinee to know operations management responsibilities as it applies to performing maintenance at the station. Frequently an Out of Service is performed for maintenance activities on plant equipment. The question is SRO Level because per 10 CFR 55.43(b)(3) and Per ES-401 Attachment 2, some examples of SRO exam items for that topic include: • Administrative processes for temporary modifications. (YES), When an O.O.S is hung in support of a maintenance activity that is a temporary modification to the plant. • Administrative processes for disabling annunciators. • Administrative processes for the installation of temporary instrumentation. • Processes for changing the plant or plant procedures. Reference: WC-AA-101




System ID: 2029204 User-Defined ID: 2017 NRC Q96C Cross Reference Number: 8E.AM-137

Topic: Per WC-AA-101, ONLINE WORK CONTROL PROCESS, which of the following are responsibilities of Operation

RO Importance: SRO Importance: 4.1 K/A: G2.1.36 Comments: New Question

License Level: SRO Cognitive Level: Low, from memory 2.0 GENERIC KNOWLEDGES AND ABILITIES 2.2.19 Knowledge of maintenance work order requirements Objective: 8E.AM-137, AUTHORIZE the performance of maintenance on shift



97 ID: 2017 NRC Q97A Points: 1.00 An emergent priority troubleshooting work activity is in progress. The initial troubleshooting did not identify the problem. The maintenance crew has returned to the control room with new troubleshooting work instructions to further investigate the malfunction. What is the Unit Supervisor's responsibility for this activity per MA-AA-716-004, CONDUCT OF TROUBLESHOOTING?

A. Leads the troubleshooting process.

B. Update the Maintenance Alteration Log.

C. Review and approve the new troubleshooting work instructions and evaluate the new

activity for risk.

D. Reviews the new troubleshooting work instructions and documents the troubleshooting results in the Troubleshooting Log.

Answer: C

Answer Explanation



Per MA-AA-716-004, (step 3.8.1) the Unit Supervisors responsibilities are to: Authorizes field troubleshooting activities and ensures adequate bounds have been established to limit plant impact and/or cause a change from previous risk assessment values by review and approval of each applicable Attachment 1, Troubleshooting Log. A. Incorrect: This activity is plausible as it can fall under the SM/Designee responsibility but is not

listed as a US responsibility. (Reference step 3.7.1) B. Incorrect: This is plausible as the maintenance alteration log will be updated though out the

troubleshooting process but that update will made by the "field work group". This is annotated on step 4.1.10.3 of procedure MA-AA-716-004

C. Correct as stated above. D. Incorrect: The first part is partially correct (reviewing the new work instructions), which makes

this part of the distractor plausible. The second part is normally performed by the maintenance crew but could be performed by operating, making this choice plausible.

Question meets the K/A by testing the candidate on (1) the troubleshooting process and on (2) on what to do to manage a change to the troubleshooting field activities. Question is SRO because: Per ES-401 Attachment 2, some examples of SRO exam items for that topic include: • 10 CFR 50.59 screening and evaluation processes. • Administrative processes for temporary modifications. YES • Administrative processes for disabling annunciators. • Administrative processes for the installation of temporary instrumentation. • Processes for changing the plant or plant procedures Reference: MA-AA-716-004, CONDUCT OF TROUBLESHOOTING




System ID: 1744291 User-Defined ID: 2017 NRC Q97A Cross Reference Number: 7E.AM-137-A

Topic: An emergent priority troubleshooting work activity is in progress. The initial troubleshooting

RO Importance: SRO Importance: 3.8 K/A: 2.2.20 Comments: New question

SRO level Low Cog Level 2.2 Equipment Control 2.2.20 Knowledge of the process for managing troubleshooting activities. Objective: 7E.AM-137-A, DESCRIBE the requirements for authorization to perform maintenance on shift



98 ID: 2017 NRC Q98C Points: 1.00 1000 The Unit is in MODE 1 • 1AR020, Containment High Range Area Radiation, is declared INOPERABLE. At 1100 • 1AR021, Containment High Range Area Radiation, is declared INOPERABLE. 1. What Tech. Spec. action is required?

2. What is the Tech. Spec. Bases for 1AR020 and 1AR021 operability?

A. 1. Verify 1PR011, Containment Atmosphere Rad Monitor, is operable.

2. Monitor for the potential of significant radiation releases and provide release assessment for Emergency Plan.

B. 1. Restore 1 channel to OPERABLE within 7 days.

2. Monitor for the potential of significant radiation releases and provide release assessment for Emergency Plan.

C. 1. Verify 1PR011, Containment Atmosphere Rad Monitor, is operable.

2. Isolate the containment purge flowpaths which ensures the containment leakage rate assumptions of the safety analysis remain valid.

D. 1. Restore 1 channel to OPERABLE within 7 days.

2. Isolate the containment purge flowpaths which ensures the containment leakage rate assumptions of the safety analysis remain valid.

Answer: B

Answer Explanation



Per T.S. 3.3.3 P.A.M. Instrumentation the rad monitors apply Condition D by Table 3.3.3-1. Condition D of 3.3.3 requires the restoration of 1 channel within 7 days. The bases for these rad monitors per B.3.3.3 is to monitor for the potential of significant radiation releases and to provide release assessment for use by operators in determining the need to invoke site emergency plans.. A. Incorrect. Plausible because 1PR11, Containment Atmosphere Rad Monitor serve the containment structure. The examinee may conclude it can provide backup to the 1AR020 & 1AR021, Containment High Range Area Radiation. The second part is correct, making this part of the distractor plausible. B. Correct as stated above. C. Incorrect. Plausible because 1PR11, Containment Atmosphere Rad Monitor serve the containment structure. The examinee may conclude it can provide backup to the 1AR020 & 1AR021, Containment High Range Area Radiation. The second part is incorrect but is plausible as 1AR011 & 1AR012, Containment Fuel Handling Incident monitors, are in the containment and are Safety Related and provide an interlock function (for containment fuel handling incident). The basis for the containment fuel handling Incident monitors is plausible because again, it serves the containment structure. D. Incorrect. The first part is correct, making this part of the distractor plausible. The second part is incorrect but is plausible as 1AR011 & 1AR012, Containment Fuel Handling Incident monitors, are in the containment and are Safety Related and provide an interlock function (for containment fuel handling incident). The basis for the containment fuel handling Incident monitors is plausible because again, it serves the containment structure. Reference: T.S. 3.3.3 (PAM instrumentation) and Bases document. The question meets the K/A by examining the candidate about knowledge of radiation monitors systems, specifically the Containment High Range rad monitors. The question is SRO Only based on knowledge of Tech Spec actions greater than 1 hour and the bases of the specification. Question is SRO level. From the ES-401 Attachment 2 Figure 1: Screening for SRO-only linked to 10 CFR 55.43(b)(2) (Tech Specs): • Can NOT be answered by knowing less than 1 hour Tech Specs. No • Can NOT be answered by knowing information listed "above-the-line". No • Can NOT be answered by knowing the TS Safety Limits. No • Does involve one or more of the following for TS, TRM or ODCM: Yes - Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1). No - Application of generic LCO requirements (LCO 3.0.1 thru 3.0.7 and SR 4.0.1 thru 4.0.4). No - Knowledge of TS bases that is required to analyze TS required actions and terminology. Yes References: TS 3.3.3 and bases B3.3.3




System ID: 2029506 User-Defined ID: 2017 NRC Q98C Cross Reference Number: S.AR1-15

Topic:

1000 The Unit is in MODE 1 1AR021 Containment High Range Area Radiation is declared INOPERABLE.

RO Importance: SRO Importance: 3.1 K/A: G2.3.15 Comments: New Question

SRO Level High Cog Level due to multiple analysis required K/A: 2: Generic, .3: Radiation Control, .15:Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring, etc. Objective: S.AR1-15, Given a set of plant conditions, DETERMINE applicable Radiation Monitoring Tech Spec/TRM operability requirements.



99 ID: 2017 NRC Q99 Points: 1.00 A fire has been reported. Per BAP 1100-10, RESPONSE PROCEDURE FOR FIRE, notify the ...

A. FIRE Marshall and begin E-Plan evaluation.

B. SHIFT Manager and begin E-Plan evaluation.

C. SHIFT Manager and begin Reportability Manual evaluation.

D. FIRE Marshall and begin Reportability Manual evaluation.

Answer: B

Answer Explanation



Per BAP 1100-10, RESPONSE PROCEDURE FOR FIRE, page 3, ...The Unit Supervisor in turn notifies the Shift Manager and begins Emergency Plan evaluation. A. Incorrect: This distractor is plausible as the Fire Marshall is notified that a fire alarm has been

received. The second part of the question "begin E-Plan evaluation, is correct making this distractor correct.

B. Correct as described above. C. Incorrect: The first part is correct making this distractor plausible. The second part is incorrect

but plausible as a review of the Reportability Manual, which is an SRO function may be performed but this is not specified in BAP 1100-10.

D. Incorrect: This distractor is plausible as the Fire Marshall is notified that a fire alarm has

been received. The second part is incorrect but plausible as a review of the Reportability Manual, which is an SRO function may be performed but this is not specified in BAP 1100-10.

The question meets the K/A by examining the candidate on fire protection procedures, specifically BAP 1100-10, RESPONSE PROCEDURE FOR FIRE. Question is SRO Level by asking the SRO responsibilities during a fire event. From the ES-401 Attachment 2 Figure 2: Screening for SRO-only linked to 10CFR55.43(b)(5) (Assessment and selection of procedures): 1) assessing plant conditions and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. Using the flowchart, this question can: • NOT be answered solely by knowing “systems knowledge”, i.e., how the system works, flowpath, logic, component location.NO • NOT be answered solely by knowing immediate operator actions. NO • NOT be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs.NO • NOT be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure. NO • CAN be answered with knowledge of ONE or MORE of the following: - Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. - Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps. - Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures. - Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures. YES Reference: BAP 1100-10




System ID: 1744135 User-Defined ID: 2017 NRC Q99 Cross Reference Number: 7E.AM-200-A

Topic:

A fire has been reported.

Per BAP 1100-10, RESPONSE PROCEDURE FOR FIRE, notify the ...

RO Importance: SRO Importance: 3.7 K/A: G2.4.25 Comments: Question 92 of the 2013 Byron NRC exam

SRO Level Low Cog Level

K/A: 2 Generic , 4: Emergency Procedures/Plan, 25: Knowledge of fire protection procedures

Objective: Given the appropriate procedure, describe consideration and requirements for supervising actions to combat a plant fire (7E.AM-200-A)



100 ID: 2017 NRC Q100 Points: 1.00

(Question 100 withheld from public disclosure due to security-related content.)

examination answer key · 2019-12-16 · main turbine bypass steam (40% of main steam design flow...

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