ML14080A033

Initial Exam 2013-302 Draft SRO Written Exam

ML14080A033
Person / Time
Site:	Sequoyah
Issue date:	03/21/2014
From:	NRC/RGN-II
To:	Tennessee Valley Authority
Shared Package
ML14080A058	List: ML14080A029 ML14080A032 ML14080A033 ML14080A036 ML14080A038 ML14080A039 ML14080A042 ML14080A044 ML14080A046 ML14080A050 ML14080A052 ML14080A055
References
50-327/13-302, 50-328/13-302
Download: ML14080A033 (69)

Text

SRO

1. 76. Given the following plant conditions:

- Unit 1 is at 100% power when a pressurizer safety valve failed open.

- The operator manually tripped the reactor and initiated a safety injection.

- While performing the step to determine if the RHR spray should be placed in service in accordance with E-1, "Loss of Secondary or Reactor Coolant", the crew determines the following:

- When pressurizer pressure dropped to 1280 psig, the safety valve reclosed and pressurizer pressure started to rise.

- Containment pressure rose to a maximum of 2.6 psig, and began trending down.

- Current plant conditions are:

- Pressurizer level is 100%.

- PZR pressure 1390 psig

- RCS subcooling is 42oF.

- All four SG levels at 33% narrow range.

Which ONE of the following identifies the required procedure implementation and operation of the RCPs for the above conditions?

A. Transition from E-1 to ES-1.1, SI Termination; The RCPs will have remained running throughout the event.

B. Transition from E-1 to ES-1.1, SI Termination; The RCPs would have been shutdown, but will be restarted in ES-1.1, SI Termination.

C. Continue E-1 until a transition is directed to ES-1.2, Post LOCA Cooldown; The RCPs will have remained running throughout the event.

D. Continue E-1 until a transition is directed to ES-1.2, Post LOCA Cooldown; The RCPs would have been shutdown, but will be restarted in ES-1.2, Post LOCA Cooldown.

Answer: A

DISTRACTOR ANALYSIS:

A. Correct, with the safety valve reclosed and the conditions as identified in the stem, SI termination criteria is met. While the crew would be beyond the step in E-1 that first checks for SI termination and beyond the followup step for checking the criteria, the SI termination step is a continuous action step and if the criteria is met the transition is to be made. Subcooling is greater than the 40oF setpoint, pressurizer level is above the 10% setpoint, heat sink is established and RCS pressure rising meet the entry conditions for ES-1.1. Containment pressure did not rise to the automatic initiation setpoint of 2.8 psig (Phase B) nor did the RCS pressure drop to the 1250 psig setpoint, so the RCP trip criteria was not met and the pumps remained in service.

B. Incorrect, With the conditions identified in the stem, the SI termination criteria is met and a transition to ES-1.1 is required. The RCP trip criteria was not met and the pumps would have remained in service throughout the event. Plausible because the transition to ES-1.1 is the correct transition and if the RCPs had been stopped they would be restarted in ES-1.1.

C. Incorrect, While ES-1.2 would be entered if E-1 was continued, the conditions identified in the stem indicate SI termination criteria is met and a transition to ES-1.1 is required. The RCP trip criteria was not met and the pumps would have remained in service throughout the event. Plausible because the transition to ES-1.2 would be the correct transition if the SI could not be terminated and the RCPs remaining in service throughout the event is correct.

D. Incorrect, While ES-1.2 would be entered if E-1 was continued, the conditions identified in the stem indicate SI termination criteria is met and a transition to ES-1.1 is required. Because the RCP trip criteria was not met, the pumps would have remained in service throughout the event. Plausible because the transition to ES-1.2 would be the correct transition if the SI could not be terminated and if the RCPs had been stopped they would be restarted in ES-1.2.

Question Number: 76 Tier: 1 Group: 1 K/A: 008 Pressurizer Vapor Space Accident AG2.1.7 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretations.

Importance Rating: 4.4 / 4.7 10 CFR Part 55: 41.5 10CFR55.43.b: 5 K/A Match: This question matches the K/A by having the candidate evaluate the plant conditions that are presented and make an operational decision as to the next actions that will be required. SRO due to evaluating plant conditions and selecting the appropriate procedures to mitigate the event.

Technical

Reference:

E-1, Loss of Reactor Or Secondary Coolant, Rev 23 ES-1.1, SI Termination, Rev 10 ES-1.2, Post LOCA Cooldown, Rev 17 Proposed references None to be provided:

Learning Objective: OPL271E-1 rev 3 obj. 12. Describe the conditions and reason for transitions within E-1.

Question Source:

New Modified Bank Bank X Question History: SQN bank question used on 01/2009 NRC exam.

Comments:

1. 77. Given the following plant conditions:

- Unit 1 was operating 100% RTP.

- The 1A-A CCP tagged for maintenance due to an oil leak.

- A LOCA occurred an hour ago.

- During the performance of ES-1.3, "Transfer to RHR Containment Sump,"

containment sump valves 1-FCV-63-72 and 1-FCV-63-73 could not be opened.

- The crew entered ECA-1.1, "Loss of RHR Sump Recirculation."

- While performing step 20 "Monitor if ECCS flow should be terminated," the crew observes the following parameters:

- All RCPs are OFF

- RVLIS Low Range is 70% and stable

- Subcooling is 50°F and stable

- RWST level is 20% and lowering.

- CTMT pressure is 3.0 psig and slowly lowering.

Which ONE of the following identifies:

(1) in accordance with ECA-1.1, an action that you will direct the crew to take and (2) the operational impact of 1B-B CCP tripping on over current?

A. (1) Stop and Start ECCS pumps as necessary to establish minimum ECCS flow.

(2) the RCP's #1 seal could be adversely affected.

B. (1) Reset Phase A and Phase B, and stop RHR pumps, SI pumps and all but one CCP.

(2) the RCP's #1 seal could be adversely affected.

C. (1) Stop and Start ECCS pumps as necessary to establish minimum ECCS flow.

(2) the RCS loops may not refill to ensure secondary heat transfer is available.

D. (1) Reset Phase A and Phase B, and stop RHR pumps, SI pumps and all but one CCP.

(2) the RCS loops may not refill to ensure secondary heat transfer is available.

Answer: A

DISTRACTOR ANALYSIS:

A. Correct, Within the guidance of ECA-1.1, the normal SI termination criteria is relaxed to allow for a reduction in ECCS flow to extend the time of depletion of the RWST. However, the criteria to terminate ECCS flow is 90°F verse the normal 40°F.

With only 50°F subcooling given in the stem, the operators would be directed to either open and close CCPIT valves or start and stop ECCS pumps as necessary to maintain the minimum amount of SI flow to keep the core cooled. Also in accordance with EPM-3-ECA-1.1, Basis document for ECA-1.1, with a phase B initiated the only cooling being provided to the RCP seals is through seal injection flow. Thus if the only running CCP trips the RCP #1 seals could be adversely affected. .

B. Incorrect, Plausible since there is more than enough RCS subcooling to terminate SI, however for conditions of ECA-1.1 the minimum amount of subcooling is 90°F to reset phase A & B, and terminate ECCS flow. Also plausible since the second part is correct.

C. Incorrect, Plausible since the second part is correct. Also plausible since the main goal of SI flow is to fill or keep full the RCS loops to provide for the use of SGs as the primary heat removal source. However during times of a Large Break LOCA, the ECCS flow is the primary heat removal tool with the ability to maintain RCS inventory more of a goal during Small Break LOCAs.

D. Incorrect, Plausible since there is more than enough RCS subcooling to terminate SI, however for conditions of ECA-1.1 the minimum amount of subcooling is 90°F to reset phase A & B, and terminate ECCS flow. Also plausible since the main goal of SI flow is to fill or keep full the RCS loops to provide for the use of SGs as the primary heat removal source. However during times of a Large Break LOCA, the ECCS flow is the primary heat removal tool with the ability to maintain RCS inventory more of a goal during Small Break LOCAs.

Question Number: 77 Tier: 1 Group: 1 K/A: 011 Large Break LOCA EA2.05 Ability to determine or interpret the following as they apply to a Large Break LOCA:

Significance of charging pump operation.

Importance Rating: 3.3 / 3.7 10 CFR Part 55: n/a 10CFR55.43.b: 5

K/A Match: This question matches the K/A by having the candidate recall the most significant reason to keep a CCP running during the recovery phase of a Large Break LOCA. SRO by having the candidate recall specific basis information for a possible equipment failure. Also SRO by having the candidate recall specific information from the procedure basis document.

Technical

Reference:

ECA-1.1 Loss of RHR Sump Recirculation, rev 11 EPM-3-ECA-1.1, Basis Document for ECA-1.1, rev 5 Proposed references None to be provided:

Learning Objective: OPL271ECA-1.1 rev 3 obj. 6. Given the procedure and a set of initial conditions, determine actions required to mitigate the event in progress.

Question Source:

New X Modified Bank Bank Question History: New question written for 1311 ILT exam Comments:

1. 78. Given the following plant conditions:

- Unit 1 is in Mode 6 with the reactor vessel head being detensioned.

- The following plant parameters:

- RHR pump 1A-A is in service with flow at 3000 gpm.

- 1A-A RHR pump amps = 20 amps

- RHR discharge pressure = 80 psig

- RCS pressure 0 psig.

- An equipment malfunction occurs which results in the following plant parameters:

- 1A-A pump flow = 0 gpm

- 1A-A pump amps = 3 amps

- RHR discharge pressure = 20 psig Which ONE of the following identifies both; (1) the equipment malfunction that occurred and (2) in accordance with T.S. 3.9.8.1, Residual Heat Removal and Coolant Circulation, bases, the requirement for having one loop of RHR in operation?

(1) (2)

A. 1A-A RHR pump shaft shear to ensure sufficient coolant circulation to minimize the effects of a boron dilution incident.

B. 1A-A RHR pump shaft shear to ensure sufficient coolant flow through the reactor vessel to prevent reactor vessel thermal stress.

C. FCV-63-93, RHR Cold leg to ensure sufficient coolant circulation to minimize injection to Loops 2 & 3 closed the effects of a boron dilution incident.

D. FCV-63-93, RHR Cold leg to ensure sufficient coolant flow through the injection to Loops 2 & 3 closed reactor vessel to prevent reactor vessel thermal stress.

Answer: A

DISTRACTOR ANALYSIS:

A. Correct, Amps will lower for a shaft shear and will be much less than normal running amps. In accordance with T.S. 3.9.8.1 bases part 2) sufficient coolant circulation is maintained through the reactor core to minimize the effects of a boron dilution incident and prevent boron stratification.

B. Incorrect, Plausible since amps will lower for a shaft shear and will be much less than normal running amps. Also plausible since during plant cooldown thermal stress on the reactor vessel is a major concern, however with the RCS pressure at "0" and the vessel head being de-tensioned, thermal stress is no longer a major concern.

C. Incorrect, Plausible if the candidate determines that lowered amps would be an indication of lowered flow caused by closing the loop injection valve, also plausible if the candidate does not recall where the loop flow is measured and determines that the flow is downstream of the injection valve. Also plausible since the second part is correct.

D. Incorrect, Plausible if the candidate determines that lowered amps would be an indication of lowered flow caused by closing the loop injection valve, also plausible if the candidate does not recall where the loop flow is measured and determines that the flow is downstream of the injection valve. Also plausible since during plant cooldown thermal stress on the reactor vessel is a major concern, however with the RCS pressure at "0" and the vessel head being de-tensioned, thermal stress is no longer a major concern.

Question Number: 78 Tier: 1 Group: 1 K/A: 025 Loss of Residual Heat Removal (RHR) System AA2.01 Ability to determine and interpret the following as they apply to the Loss of Residual Heat Removal System:

Proper amperage of running LPI/decay heat removal/RHR pump(s).

Importance Rating: 2.7 / 2.9 10 CFR Part 55: 43.5 / 45.13 10CFR55.43.b: 5 K/A Match: KA is matched because the questions requires ability to identify off normal conditions of RHR pump amperes and the cause of those conditions. Also is SRO because it requires the knowledge of Tech

Spec bases for RHR loop operation during refueling conditions.

Technical

Reference:

Tech Spec. 3.9.8.1, Refueling Operations - Residual Heat Removal and Coolant Circulation amen 305 Proposed references None to be provided:

Learning Objective: OPTSTG200.RHR rev 8 Obj. 11 Obj. 13 b Question Source:

New X Modified Bank Bank Question History: New question for ILT 1311 exam.

Comments:

1. 79. Given the following plant conditions:

- Unit 2 is at 100% RTP.

- Unit 1 is at 547°F and 2250 psig, reactor trip breakers are OPEN.

- The A-A EGTS fan is tagged out for maintenance.

- CSST "B" is tagged out for tap changer maintenance.

- Due to a malfunction in the switchyard, off-site power is lost to CSST "A."

If the above conditions associated with CSST "A" remain unchanged for 24 hrs, in accordance with Tech Specs, which ONE of the following identifies the operational implications on Unit 2?

Reference Provided A. No shutdown is required.

B. Unit 2 must be in Hot Shutdown within 12 hrs.

C. Unit 2 must be in Hot Shutdown within 13 hrs.

D. Unit 2 must be in Hot Shutdown within 14 hrs.

Answer: D DISTRACTOR ANALYSIS:

A. Incorrect. Plausible if the candidate does not determine that EGTS is a shared system and thus applies T.S. 3.6.1.8 guidance which allows 7 days to return both trains of EGTS to operable status. Thus within 24 hrs no action would be required for Unit 2.

B. Incorrect. Plausible if the candidate applies Tech Spec 3.0.3 to this condition. With the second train of EGTS inoperable the candidate may logically conclude that the unit needs to be shut down immediately and does not apply the 1 hr preparation time allowed in T.S. 3.0.3.

C. Incorrect. Plausible if the candidate thinks that since both trains of EGTS are inoperable and applies the time limit of Tech Spec 3.0.3 (1 hr + 6 hrs + 6 hrs = 13 hrs)

D. Correct. With the "A" CSST de-energized, Start Bus 1A is de-energized. This will cause a loss of normal power to Shutdown board 1B-B. DG 1B-B will start and re-energize the shutdown board. This failure will only initially affect Unit 1. However with the normal power supply inoperable for 1B1 SD Board, the B-B EGTS fan is inoperable per T.S. 3.0.5. Since EGTS is a shared system then Unit 2 will apply T.S. 3.0.5 and needs to be in Hot Shutdown in 14 hrs. (2 hrs + 6 hrs + 6 hrs = 14

hrs)

Question Number: 79 Tier: 1 Group: 1 K/A: 056 Loss of Offsite Power AG2.2.40 Ability to apply Technical Specifications for a system.

Importance Rating: 3.4 / 4.7 10 CFR Part 55: 41.10 / 43.2 / 43.5 / 45.3 10CFR55.43.b: 2 K/A Match: Question provides a loss of Off site power and the allowances for applying Tech Specs.

This question tests the SRO knowledge of TS requiring evaluation of operability of EGTS system and application of Tech Spec 3.6.1.8 and generic Tech Spec 3.0.5 Technical

Reference:

TS SR 4.0.2 TS LCO 3.6.1.8 Emergency Gas Treatment System -

EGTS- Cleanup System Proposed references Tech Spec 3.6.1.8 "Emergency Gas Treatment System -

to be provided: EGTS- Cleanup System" amend 263 page 3/4 6-13 Learning Objective: OPT200.TS-APP Obj 3 Question Source:

New X Modified Bank Bank Question History: New for SQN ILT 1311 Comments:

1. 80. Given the following plant conditions:

- Unit 1 is at 12% power during a plant startup.

- A loss of Essential and Non-Essential Control Air has occurred.

- The crew is attempting to restore Control Air in accordance with AOP-M.02, "Loss of Control Air."

- RCS temperature is 555°F and rising.

- PZR level is 72% and rising.

- SG NR levels are 15% and lowering.

Based on the current plant conditions, in accordance with AOP-M.02, which ONE of the following indentifies:

(1) the reason a reactor trip is required, and (2) the action(s) required to control RCS temperature _____ be performed concurrently with E-0?

A. (1) loss of PZR level control (2) can B. (1) loss of PZR level control (2) can NOT C. (1) loss of SG level control (2) can D. (1) loss of SG level control (2) can NOT Answer: C

DISTRACTOR ANALYSIS:

A. Incorrect, Plausible since the guidance in accordance with AOP-M.02, if PZR level is greater than 70% then action is to evaluate a plant shutdown and if level approaches 92% then manually trip reactor. (sect 2.2 step 20) The candidates could confused the guidance for controlled shutdown vs reactor trip. Also AOP-M.02 is to be performed concurrently with E-0.

B. Incorrect, Plausible since the guidance in accordance with AOP-M.02, if PZR level is greater than 70% then action is to evaluate a plant shutdown and if level approaches 92% then manually trip reactor. (sect 2.2 step 20) The candidates could confused the guidance for controlled shutdown vs reactor trip. Also plausible if the candidate thinks that they are to GO TO E-0 and not complete the actions of AOP.M.02 concurrently.

C. Correct, In accordance with AOP-M.02, if SG levels are not being maintained on program, the operators are directed that if SG levels are approaching an auto trip setpoint then they are to trip the reactor and go to E-0 while continuing with this procedure. (sect 2.2 step 3) Thus E-0 and AOP-M.02 are to be implemented concurrently.

D. Incorrect, Plausible since the first part is correct. Also plausible if the candidate thinks that they are to GO TO E-0 and not complete the actions of AOP.M.02 concurrently.

Question Number: 80 Tier: 1 Group: 1 K/A: 065 Loss of Instrument Air AA2.06 Ability to determine or interpret the following as they apply to a Loss of Instrument Air; When to trip the reactor if instrument air pressure is de-creasing.

Importance Rating: 3.6 / 4.2 10 CFR Part 55: n/a 10CFR55.43.b: 5 K/A Match: This question matches the K/A by having the candidates determine when the Reactor would be tripped on a loss of instrument air. Also SRO by having the candidate analyze the conditions in the stem and determine the correct procedures to use to mitigate the Loss of

Instrument Air.

Technical

Reference:

AOP-M.02, Loss of Inst Air, rev 19 E-0 Proposed references None to be provided:

Learning Objective: OPL271AOP-M.02, obj. 7 Question Source:

New Modified Bank Bank X Question History: NRC question used on 04/2007 NRC exam Comments:

1. 81. Given the following plant conditions:

- The Unit 1 crew enters ECA-1.2, "LOCA Outside Containment."

- When ECA-1.2 is complete the following conditions exist:

- CNMT pressure is 0.1 psig and stable.

- RWST level is 70% and lowering.

- CETs are indicating 510°F.

- RCS pressure is 800 psig and lowering.

- RVLIS is 46% and slowly rising.

Which ONE of the following completes the statements below?

The crew is required to transition to ___(1)___ .

The Emergency Plan Classification for this event is a(n) ___(2)___ .

REFERENCE PROVIDED A. (1)E-1, "Loss of Reactor or Secondary Coolant" (2)Site Area Emergency B. (1)ECA-1.1, "Loss of RHR Sump Recirculation" (2) Alert C. (1)E-1, "Loss of Reactor or Secondary Coolant" (2)Alert D. (1)ECA-1.1, "Loss of RHR Sump Recirculation" (2) Site Area Emergency Answer: D DISTRACTOR ANALYSIS:

A. Incorrect, Plausible because ensuring RHR is isolated first is correct and if the RCS pressure had been rising instead of lowering, the correct transition would be to E-1.

Also the second part is correct.

B. Incorrect, Plausible since EPM-3-ECA-1.2 identifies a rupture or break outside containment is most probable to occur in the Low pressure RHR System piping.

With the RCS pressure dropping, the break has not been isolated and the procedure will direct a transition to ECA-1.1. Also plausible if the candidate misses the Loss of Containment 1.3.2 due to missing the criteria of no changing CNMT parameters then they would choose an Alert based on Loss of RCS barrier.

C. Incorrect, Plausible because sequentially isolating the RHR cold leg injection paths is performed in the procedure but after the RHR suction is isolated and if the RCS pressure had been rising, the correct transition would be to E-1. Also plausible if the candidate misses the Loss of Containment 1.3.2 due to missing the criteria of no changing CNMT parameters then they would choose an Alert based on Loss of RCS barrier.

D. Correct, EPM-3-ECA-1.2 identifies a rupture or break outside containment is most probable to occur in the Low pressure RHR System piping. With the RCS pressure dropping, the break has not been isolated and the correct procedure transition would be to ECA-1.1. Also with a LOCA is progress and no change in CNMT pressure or sump level it is a Loss of Containment Barrier 1.3.2 and with a RCS leak which results < 40°F subcooling, that would be a Loss of RCS Barrier 1.2.2.

With the Loss of 2 Barriers a Site Area Emergency would be declared.

Question Number: 81 Tier: 1 Group: 1 K/A: W/E04 LOCA Outside Containment G 2.4.18 Knowledge of specific bases for EOPs.

Importance Rating: 3.3 / 4.0 10 CFR Part 55: 41.10 / 43.5 / 45.13 10CFR55.43.b: 5 K/A Match: KA is matched because the question requires knowledge of the mitigation strategies of the procedure for responding to a LOCA outside containment and the ability to assess conditions and make the correct transition from the procedure. Also SRO due to assessing plant conditions and selection of appropriate procedure to mitigate the event and the EPIP classification of this event.

Technical

Reference:

ECA-1.2, "LOCA Outside Containment, Revision 10 EPM-3-ECA-1.2 Basis document for ECA-1.2 LOCA Outside Containment. rev 2 EPIP-1, Emergency Plan Classification Matrix, rev 49 Proposed references EPIP-1 Classification Matrix pg 11 & 12 to be provided:

Learning Objective: OPL271ECA-2.1 obj 5 & 7

Question Source:

New Modified Bank Bank X Question History: SQN bank exam question which has been rewritten to include EPIP classification of the event. (Used on 1201 exam)

Comments:

1. 82. Given the following plant conditions:

- Unit 1 is at 100% power.

- The local leak rate test on the lower containment air lock test performed following a containment entry was NOT satisfactory.

- Plant Engineering has just reported that 0-SI-SLT-000-160, "Primary Containment Total Leak Rate," has just been completed.

- System Engineering reports that U1 overall containment leakage rate Acceptance Criteria has been exceeded.

Which ONE of the following identifies:

(1) the most limiting Tech Spec LCO action time for the condition described and (2) the Tech Spec basis for the maximum allowed total containment leakage (La)?

A. (1) 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> (2) To prevent exceeding the design capability of the EGTS B. (1) 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (2) To prevent exceeding the design capability of the EGTS C. (1) 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> (2) To prevent exceeding 10CFR100 limits at the site boundary D. (1) 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (2) To prevent exceeding 10CFR100 limits at the site boundary Answer: C

DISTRACTOR ANALYSIS:

A. Incorrect, Plausible since the 1 hr LCO is correct, and leakage from the containment into the annulus is processed by the EGTS trains before release.

B. Incorrect, Plausible if the candidate determined that the 24 hr LCO action of T.S.

3.6.1.3.b applied to an inoperable air lock and and leakage from the containment into the annulus is processed by the EGTS trains before release.

C. Correct, With the overall containment leakage greater than allowable T.S. 3.6.1.3 footnote (2) states that T.S. 3.6.1.1 Containment Integrity is to be applied which has a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> action statement. In accordance with Tech Spec basis, the containment allowable leakage limit prevents exceeding 10CFR100 limits.

D. Incorrect, Plausible if candidate determined that the 24 hr LCO action of T.S.

3.6.1.3.b is the only applicable specification and the containment allowable leakage limit prevents exceeding 10CFR100 limits..

Question Number: 85 Tier: 1 Group: 2 K/A: 069 Loss of Containment Integrity AG 2.1.23 Ability to perform specific system and integrated plant procedures during all modes of plant operation.

Importance Rating: 4.3 / 4.4 10 CFR Part 55: n/a 10CFR55.43.b: 5 K/A Match: This question matches the K/A by having the candidate integrate procedure data and determine that Tech Spec action is required ( this has a 1 hr LCO action statement). The question is SRO because it requires knowlege of Tech Spec 3.6.1.3 and Tech Spec 3.6.1.1 basis.

Technical

Reference:

T.S. 3.6.1.3, T.S. 3.6.1.1 and Basis 0-SI-SLT-000-160.0 rev 0004 Proposed references None to be provided:

Learning Objective: OPTSTG200.CNTMTSTRUCTURE obj 10 Using Technical Specifications, Technical Requirements Manual, and the ODCM,

a. List from memory, Containment Structure LCOs and/or Technical Requirements having action times <

one hour.

b. Explain applicable Containment Structure LCO, Technical Requirements and ODCM bases.

Cognitive Level:

Higher X Lower Question Source:

New Modified Bank Bank X Question History: SQN bank question written for 1009 SRO NRC exam Comments:

1. 83. Given the following plant conditions:

- Unit 1 is at 100% RTP.

- At 0800 a dropped rod occurred.

- At 1200 the rod was recovered and power was restored to 100%.

- At 1215 the following alarms were received panel 1-M12-A:

- 1-RA-90-1A, "AUX BLDG AREA RAD MONITOR HIGH," (A-7)

- 1-RA-90-59A, "RX BLDG AREA RAD MON HIGH," (B-3)

- At 1300 on 11/25 RCS Activity was determined to be 28 microcuries/gram DOSE EQUIVALENT I-131.

Which ONE of the following completes the statements below?

(1) In accordance with AOP-R.06, "High RCS Activity," the concurrent rad monitor alarms ________ consistent with the given plant conditions.

(2) In accordance with Tech Spec 3.4.8, "Specific Activity," the unit is required to be placed in Mode 3, with Tave < 500oF, NO later than________ .

Reference Provided A. (1) are (2) 1900 on 11/27 B. (1) are (2) 1900 on 11/25 C. (1) are NOT (2) 1900 on 11/27 D. (1) are NOT (2) 1900 on 11/25 Answer: B.

DISTRACTOR ANALYSIS:

A. Incorrect, Plausible since the first part is correct, In accordance with AOP-R.06, both the 1-RA-90-1A and 1-RA-90-59A are listed as probable symptoms of high RCS activity. Also plausible if the candidate applies the requirement of >.35 microcuries/gram for 48 hrs. Thus 1300 + 48 hrs + 6 hrs = 1900 hrs on 9/27 vs the same day.

B. Correct, In accordance with AOP-R.06, both the 1-RA-90-1A and 1-RA-90-59A are listed as probable symptoms of high RCS activity, thus consistent with plant conditions. Also since the RCS activity has exceeded the limit line of Tech Spec figure 3.4-1, then in accordance with action "a" the reactor is to be in at least HOT STANDBY with Tavg less than 500oF within 6 hrs. Thus (1300 + 6 hrs = 1900) on the same day.

C. Incorrect, Plausible if the candidate does not think that the Hi rad alarm on 1-RA 1A (an Aux Bldg alarm) would not be present if the hi activity is only in the RCS (thus only the Rx Bldg would be affected). The candidate could also think that 1-RA-90-106 or 90-112 should be in alarm instead. Also plausible since the second part is correct.

D. Incorrect, Plausible if the candidate does not think that the Hi rad alarm on 1-RA 1A (an Aux Bldg alarm) would be present if the hi activity is only in the RCS (thus only the Rx Bldg would be affected). The candidate could also think that 1-RA 106 or 90-112 should be in alarm instead. Also plausible if the candidate applies the requirement of >.35 microcuries/gram for 48 hrs. Thus 1300 + 48 hrs + 6 hrs = 1900 hrs on 9/27 vs the same day.

Question Number: 83 Tier: 1 Group: 2 K/A: 076 High Reactor Coolant Activity AG2.4.46 Ability to verify that the alarms are consistent with the plant conditions.

Importance Rating: 4.2 / 4.2 10 CFR Part 55: 41.10 10CFR55.43.b: 2 K/A Match: This question matches the K/A by having the candidate analyze the alarms given and determine if the alarms are consistent with the plant conditions. Also SRO by testing the candidates knowledge of Tech Spec and bases for high RCS activity.

Technical

Reference:

AOP-R.06 High RCS Activity, rev 11 Tech Spec 3.4.8, Specific Activity Amend 301 Proposed references Tech Spec 3.4.8 and Figure 3.4-1 to be provided:

Learning Objective: OPL271AOP-R.06 rev 2 Obj. 7 Describe the Tech Spec and TRM actions applicable during the performance of AOP-R.06.

Question Source:

New X Modified Bank Bank Question History: New question developed for 1311 exam Comments:

1. 84. Given the following plant conditions:

- Unit 1 was tripped from 100% power due to a Small Break LOCA.

- Containment pressure peaked at 2.9 psig.

- The crew is performing ES-1.2, "Post LOCA Cooldown and Depressurization."

- The crew is evaluating step 37, "Determine if RHR should be placed in service:"

- The following parameters exist:

- Both RHR pumps are stopped.

- Normal charging has been established.

- PZR level is 28% and stable.

- RCP #2 is the only RCP running.

- RCS pressure - 290 psig and slowly trending down.

- RCS temperature - 340°F and slowly trending down.

- Containment pressure - 1.1 psig and slowly trending down Which ONE of the following complete the statements below?

(1) The maximum cooldown rate allowed during the performance of ES-1.2 is _____.

and (2) The SRO will direct the crew to ______.

A. (1) 50°F/hr (2)transfer to ES-1.4, "Transfer to Hot Leg Recirculation", with TSC concurrence.

B. (1) 50°F/hr (2) place RHR in service in accordance with EA-74-1, "Placing RHR Shutdown Cooling in Service," with TSC concurrence.

C. (1) 100°F/hr (2) transfer to ES-1.4, "Transfer to Hot Leg Recirculation", with TSC concurrence.

D. (1) 100°F/hr (2) place RHR in service in accordance with EA-74-1, "Placing RHR Shutdown Cooling in Service," with TSC concurrence.

Answer: D

DISTRACTOR ANALYSIS:

A. Incorrect, Plausible since 50 °F/hr is the cooldown limit in certain conditions within the EOP network. The second part is plausible as the next step if RHR conditions are not established and since the event is deep into the recovery is to now consider (step 38) if hot leg recirculation is required with consultation with the TSC.

B. Incorrect, Plausible as 50 °F/hr is the cooldown limit in certain conditions within the EOP network. Also plausible since the second part is correct.

C. Incorrect, Plausible since the maximum cooldown rate is 100°F/hr based on guidance from ES-1.2 and is correct. The second part is plausible as the next step if RHR conditions are not established and since the event is deep into the recovery is to now consider (step 38) if hot leg recirculation is required with consultation with the TSC.

D. Correct, In accordance with ES-1.2, the Maximum cooldown rate is 100°F/hr and the necessary conditions to establish RHR have been established.

Question Number: 84 Tier: 1 Group: 2 K/A: WE 03 LOCA Cooldown and Depressurization EA 2.1 Ability to determine and interpret the following as they apply to the (LOCA Cooldown and Depressurization):

Facility conditions and selection of appropriate procedures during abnormal and emergency operations.

Importance Rating: 3.4 / 4.2 10 CFR Part 55: n/a 10CFR55.43.b: 5 K/A Match: This question matches the K/A by requiring the candidate to analyze the data presented during the LOCA Cooldown and Depressurization to determine if proper cooldown rate has been established and whether or not RHR can be placed in service per EA-74-1. SRO by requiring the candidate to assess plant conditions and select the appropriate procedure to mitigate the accident.

Technical

Reference:

ES-1.2, Post LOCA Cooldown and Depressurization, Rev 18

Proposed references None to be provided:

Learning Objective: OPL271ES-1.2 rev 2 Obj 7 Describe the conditions and reason for transitions within this procedure and transitions to other procedures.

Question Source:

New X Modified Bank Bank Question History: New question written for 1311 ILT exam Comments:

1. 85. Given the following plant conditions:

- While operating at 100% RTP, a small break LOCA occurred on Unit 1.

- The operating crew has progressed through the emergency instructions and has placed RHR Train A in service in accordance with 0-SO-74-1, "Residual Heat Removal System," during performance of ES-1.2, "Post LOCA Cooldown and Depressurization."

- The STA identifies a YELLOW path to FR-C.3, "Saturated Core Cooling,"

and the SRO makes the transition.

Which ONE of the following completes the statements below?

The basis of FR-C.3 is to __(1)__.

After entering FR-C.3, a transition to AOP-R.03, "RHR System Malfunctions,"

__(2)__ be directed.

(1) (2)

A. establish RCS cooling via SGs will B. establish RCS cooling via SGs will NOT C. restore subcooling using ECCS will D. restore subcooling using ECCS will NOT Answer: C.

DISTRACTOR ANALYSIS:

A. Incorrect, Plausible because FR-C.2, "Degraded Core Cooling," does use the secondary to cool the RCS and the transition to AOP-R.03 is correct.

B. Incorrect, Plausible because FR-C.2, "Degraded Core Cooling," does use the secondary to cool the RCS and because normally an FRG will transition to another procedure in the EOP versus a transition to an AOP.

C. Correct, FR-C.3 attempts to increase the ECCS flow and reduce the loss of primary coolant. Both of these actions will result in restoring a minimum of RCS subcooling.

The WOG background description states "Primarily, concern for core cooling arises when the RCS reaches saturation due to a loss of RCS inventory. Without adequate makeup, the continued loss of inventory will cause the core to partially uncover.

Guideline FR-C.3 has been developed to address this concern. The operator is instructed to begin safety injection and check for any open RCS vent path in an attempt to stop the loss of RCS inventory." The first step in FR-C.3 directs "IF RHR in shutdown cooling mode, THEN ** GO TO AOP-R.03, RHR System Malfunctions.

D. Incorrect, Plausible because the purpose of FR-C.3 is to restore RCS subcooling using the ECCS. Also, because normally an FRG will transition to another procedure in the EOP versus a transition to an AOP.

Question Number: 85 Tier: 1 Group: 2 K/A: W/E07 Saturated Core Cooling EA2.2 Ability to determine and interpret the following as they apply to the (Saturated Core Cooling):

Adherence to appropriate procedures and operation within the limitations in the facility's license and amendments.

Importance Rating: 3.3 / 3.9 10 CFR Part 55: na 10CFR55.43.b: 5 K/A Match: K/A is matched because the question requires knowledge of the specific bases of EOP FR-C.3. SRO because question requires candidate to assess plant conditions and select appropriate procedures to mitigate the accident.

SRO ONLY: The question requires knowledge of diagnostic steps and decision

points in the emergency operating procedures (EOP) that involve transitions to event specific subprocedures or emergency contingency procedures. It requires assessing plant conditions and then selecting a procedure to mitigate, recover, or with which to proceed. One area of SRO level knowledge (with respect to selecting a procedure) is knowledge of the content of the procedure versus knowledge of the procedures overall mitigative strategy or purpose.

Technical

Reference:

FR-C.3, Saturated Core Cooling, Revision 0009 EPM-3-FR-C.3, Basis Document for FR-C.3 Saturated Core Cooling, Rev. 3 Proposed references None to be provided:

Learning Objective: OPL271FR-C.3 rev 2

5. Descibe the bases for all limits, notes, cautions and steps of FR-C.3.

Cognitive Level:

Higher X Lower Question Source:

New Modified Bank Bank X Question History: Bank question written for the WBN 03/2013 NRC exam.

Comments:

1. 86. Given the following plant conditions:

- Unit 1 is performing a plant shutdown in accordance with 0-GO-7, "Unit Shutdown From Hot Standby To Cold Shutdown."

- Following a scheduled work week train swap, the 1B-B RHR pump is tagged out of service.

- Current plant conditions are:

- RCS temperature = 260°F

- RCS pressure = 290 psig

- SG conditions:

1. 1 #2 #3 #4

_5% Wide Range _ 20% Narrow Range 8% Narrow Range_ 15% Wide Range_

RCP Tagged RCP Running RCP available RCP Tagged

- During post work review, I&C technicians report to the WCC SRO that the flow transmitter for 1-FCV-74-12A, RHR Pump 1A Mini-flow valve, has failed high.

Which ONE of the following completes the statements below?

(1) Prior to taking any operator action the 1A-A RHR pump is _______

and (2) in accordance with Tech Spec 3.4.1.3, "Reactor Coolant System - Shutdown,"

bases, the requirements for this specification _____ being met.

A. (1) OPERABLE (2) are B. (1) OPERABLE (2) are NOT C. (1) INOPERABLE (2) are D. (1) INOPERABLE (2) are NOT Answer: C.

DISTRACTOR ANALYSIS:

A. Incorrect, Plausible if the candidate does not recognize that the miniflow valve will not open on a low flow condition thus may cause excessive pump heating and damage. Also plausible since the second part is correct.

B. Incorrect, Plausible if the candidate does not recognize that the miniflow valve will not open on a low flow condition thus may cause excessive pump heating and damage. Also plausible if the candidate does not recognize that SG #3 is above 10% WR (but thinks it needs to be above 10% NR) thus there are two remaining cooling loops. According to the basis to meet single failure criteria there needs to be 2 operable loops for decay heat removal. With two SGs being above the required level (10% WR) and having an operable RCP then criteria is being met.

C. Correct, 1-FIS-74-12A, is the flow transmitter for the miniflow valve 1-FCV-74-12.

This valve opens during low flow conditions to ensure that at least 500 gpm flow passes through the RHR pump to prevent pump damage. With this flow transmitter failed high, the recirc valve would not open automatically during low flow conditions and could result in RHR pump damage from overheating. Also in accordance with T.S. 3.4.1.3 during mode 4 at least 2 heat removal loops must be operable. To make an RCS loop operable in accordance with the surveillance 4.4.1.3.2 a minimum of 10% WR level is required for a loop to be considered operable. There are currently 2 RCS loops Operable. Thus with the 1A-A RHR pump inoperable then the requirements are being met.

D. Incorrect, Plausible since the first part is correct. Also plausible if the candidate does not recognize that SG #3 is above 10% WR (but thinks it needs to be above 10% NR) thus there are two remaining cooling loops. According to the basis to meet single failure criteria there needs to be 2 operable loops for decay heat removal. With two SGs being above the required level (10% WR) and having an operable RCP then criteria is being met.

Question Number: 86 Tier: 2 Group: 1 K/A: 005 Residual Heat Removal System (RHR)

A2.01 Ability to (a) predict the impacts of the following malfunctions or operations on the RHR system, and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Failure modes for pressure, flow, pump motor amps, motor temperature, and tank level instrumentation.

Importance Rating: 2.7 / 2.9 10 CFR Part 55: 41.5

10CFR55.43.b: 2, 5 K/A Match: This question matches the K/A by having the candidate determine the effects of an instrument failure on the RHR system and based on those effects determine if the minimum Tech Spec requirements are being met. SRO by determining equipment operability and knowledge of Tech Spec basis to determine if intent of the specification is being met.

Technical

Reference:

Tech Spec 3.4.1.3, Reactor Coolant System - Shutdown Proposed references None to be provided:

Learning Objective: OPTSTG200.RHR rev 8 Obj. 8.c.

Obj.13.b Question Source:

New X Modified Bank Bank Question History: New question written for 1311 ILT exam Comments:

1. 87. Given the following plant conditions:

- Unit 1 is shutdown for refueling.

- Unit 2 is at 100% power.

- The C-S CCS pump has been powered from its alternate (Train A) power supply for the last 72 hrs.

- While making preparations to comply with Tech Spec 3.7.3, "Component Cooling Water System," an automatic reactor trip occurs on Unit 2.

- All equipment operates as designed and the crew transitions to ES-0.1, "Reactor Trip Response."

In accordance with SPP-3.5, "Regulatory Reporting Requirements," which ONE of the following identifies both..

(1) the TVA internal notification(s) required to be made directly by the Shift Manager and (2) the maximum time allowed for making the first required notification to the NRC?

REFERENCE PROVIDED A. (1) Site Operations Management ONLY (2) 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> B. (1) Site Operations Management ONLY (2) 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> C. (1) Site Operations Management and Duty Plant Manager (2) 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> D. (1) Site Operations Management and Duty Plant Manager (2) 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Answer: C.

DISTRACTOR ANALYSIS:

A. Incorrect, Plausible since the Shift Manager is responsible for contacting the Site Operations Management, but is also directly responsible for contacting the Duty Plant Manager. Also plausible since a 4 hr report would be required per 10CFR50.72(b)(2).

B. Incorrect, Plausible since the Shift Manager is responsible for contacting the Site Operations Management, but is also directly responsible for contacting the Duty Plant Manager. Also plausible since an automatic reactor trip is reportable within 8 hrs per 10CFR50.72(b)(3), however the latest reporting time would be 4 hrs based on a Tech Spec required shutdown.

C. Correct, In accordance with NPG-SPP-3.5, the shift manager is directly responsible to contact the Site Operations Management and Duty Plant Manager, it is then the Duty Plant Manager's responsibility to make other contacts as directed by NPG-SPP-3.5, Appendix D. Also since this was a plant shutdown required by Tech Specs, a 4 hr report is required in accordance with 10CFR50.72(b)(2).

D. Incorrect, Plausible since in accordance with NPG-SPP-3.5, the shift manager is directly responsible to contact the Site Operations Management and Duty Plant Manager, it is then the Duty Plant Manager's responsibility to make other contacts as directed by NPG-SPP-3.5, Appendix D. Also plausible since an automatic reactor trip is reportable within 8 hrs per 10CFR50.72(b)(3), however the latest reporting time would be 4 hrs based on a Tech Spec required shutdown.

Question Number: 87 Tier: 2 Group: 1 K/A: 008 Component Cooling Water (CCW) System 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.

Importance Rating: 2.7 / 4.1 10 CFR Part 55: (CFR: 41.10 / 43.5 / 45.11) 10CFR55.43.b: 5 K/A Match: This question matches the K/A by requiring the knowledge of internal and extenal notifications as a result of a required plant shutdown and an automatic reactor trip. SRO because internal and Regulatory notifications are task specific to a Senior Reactor Operator

Technical

Reference:

NPG-SPP-3.5, Regulatory Reporting Requirements, Appendix A, rev 0008 NPG-SPP-3.5, Appenix D Proposed references NPG-SPP-3.5 Appenix A, rev 0008 (redacted) to be provided:

Learning Objective: OPL271SPP-3.5 obj. B.3 Question Source:

New X Modified Bank Bank Question History: Question written for 1311 ILT exam Comments:

1. 88. Given the following plant conditions:

- Unit 1 is being shut down for refueling.

- RCS Tavg = 250°F

- PZR pressure = 290 psig

- #2 RCP is running

- The "1B" train of RHR is being placed in service per 0-SO-74-1, "Residual Heat Removal System."

In accordance with Tech Spec 3.6.2.1, "Containment Spray Subsystems," which ONE of the following identifies the:

(1) RHR spray operability ______ required for this plant condition and (2) the local operator action that may be required, in accordance with Tech Spec 3.6.2.1, bases?

A. (1) is NOT (2) cool common suction piping during sump recirculation phase of a LOCA.

B. (1) is (2) realign ECCS components that were disabled to comply with LCO 3.4.12 (LTOP System).

C. (1) is (2) cool common suction piping during sump recirculation phase of a LOCA.

D. (1) is NOT (2) realign ECCS components that were disabled to comply with LCO 3.4.12 (LTOP System).

Answer: A.

DISTRACTOR ANALYSIS:

A. Incorrect, Plausible if the candidate does not recognize that although Containment spray is required for Modes 1,2,3 and 4, RHR spray is NOT required in Mode 4.

Also plausible since the second part is correct.

B. Incorrect, Plausible if the candidate does not recognize that although Containment spray is required for Modes 1,2,3 and 4, RHR spray is NOT required in Mode 4.

Also plausible since LTOP actions would disable parts of the ECCS system, however local operator action is not needed for this action in accordance with Tech Spec bases.

C. Correct, In accordance with T.S 3.6.2.1 the RHR spray trains are NOT required to be operable in Mode 4, thus no action required. Also in accordance with T.S. bases, local operator action may be required to cool the common suction line for Containment Spray and RHR if RHR was in service at the time of a LOCA, due to RCS temperatures greater than 190 °F. This elevated temperature may result in the CNMT spray pumps not having enough NPSH during startup if a LOCA was to occur.

D. Incorrect, Plausible since the first part is correct. Also plausible since LTOP actions would disable parts of the ECCS system, however local operator action is not needed for this action in accordance with Tech Spec bases.

Question Number: 88 Tier: 2 Group: 1 K/A: 026 Containment Spray System (CSS)

G2.2.25 Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits.

Importance Rating: 3.2 / 4.2 10 CFR Part 55: 41.5/41.7 10CFR55.43.b: 2 K/A Match: This question matches the K/A by having the candidate recall actions that are addressed in the Containment Spray System Tech Spec bases. Also SRO due to requiring specific knowledge of actions addressed by T. S. 3.6.2.1 bases.

Technical

Reference:

Tech Spec 3.6.2.1, Containment Spray Subsystems, bases ammend 323.

Proposed references None to be provided:

Learning Objective: OPL200CS rev 9 Obj. 11. b & c Question Source:

New X Modified Bank Bank Question History: New question written for the 1311 ILT NRC exam Comments:

1. 89. Given the following plant conditions:

- Units 1 and 2 are operating at 100% power.

- The intertie transformer is OOS.

- At 0800, during the monthly surveillance test, the 1A-A DG suffers a catastrophic failure during startup.

- At 0830, due to a switching error, a disconnect in the 500kV switchyard is opened resulting in a loss of 500kV power.

Which ONE of the following the procedures will be used to mitigate the event on Unit 1?

A. AOP-P.01, "Loss of Offsite Power."

B. ES-0.1, "Reactor Trip Response."

C. ES-0.2, "Natural Circulation Cooldown."

D. AOP-P.05, "Loss of Unit 1 Electrical Shutdown Boards."

Answer: B.

DISTRACTOR ANALYSIS:

A. Incorrect, Plausible if the candidate correctly analyzes that the start buses will remain energized. Also plausible if the candidate determines that AOP-P.07 would be used to restore 500 kV power, however AOP-P.07 provides actions for a loss of all Off-site which would mean 161 kV power since that is power supply to the start buses.

B. Correct, With the intertie breaker OOS then there is no direct tie between 500 kV switchyard and the 161 kV switchyard. Thus if 500 kV switchyard is lost, then only Unit 1 is effected in that there is no output for unit 1 main generator and the unit will trip. Due to the electrical lineup with 161 kV still available then the Start buses are still energized. Since the start bused provide power to the shutdown boards all shutdown boards will remain energized even though 1A-A EDG engine failed. After the Unit 1 trip, the crew would proceed to ES-0.1 for plant stabilization.

C. Incorrect, Plausible if the candidate determines that the start buses would be de-energized for this condition. Thus with the start buses de-energized no RCPs would be running and ES-0.2 would be used to mitigate.

D. Incorrect, Plausible if the candidate determines that the start buses would be de-energized. Also plausible since EDG 1A-A has failed, then shutdown board 1A-A would be de-energized and AOP-P.05 would be used to re-energize the bus.

Question Number: 89 Tier: 2 Group: 1 K/A: 062 AC Electrical Distribution A2.07 Ability to (a) predict the impacts of the following malfunctions or operations on the AC Electrical Distribution System, and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Consequences of opening a disconnect under load.

Importance Rating: 3.0 / 3.4 10 CFR Part 55:

10CFR55.43.b: 2, 5 K/A Match: This question matches the K/A by having the candidate determine the consequences of operating a disconnect under load in the 500kV switchyard. SRO by having the candidate analyze the conditions and determine if the appropriate procedure to mitgate the event .

Technical

Reference:

Tech Spec 3.8.1.1, AC Sources - Operating Proposed references None to be provided:

Learning Objective: OPT200.SWYD rev 3 Obj. 8. Given specific plant conditions, Analyze the effect that a loss or malfunction of the following will have on the Switchyard.

c. Transmission lines Question Source:

New X Modified Bank Bank Question History: New question written for 1311 ILT exam Comments:

1. 90. Given the following plant conditions:

- Unit 1 is in Mode 3.

- The following alarms are received in the control room panel 1M-C:

- "125V DC VITAL CHGR II FAILURE OR VITAL BAT II DISCHARGE" (B4)

- "125V DC VITAL BAT BD II ABNORMAL" (B5)

- Battery Board II Voltage indicates 119 VDC and lowering slowly.

Which ONE of the following completes the statement below?

(1) In accordance with Tech Spec 3.8.2.3, "D.C. Distribution - Operating," Battery Board II is ____________ ,

and (2) the action required?

A. (1) INOPERABLE (2) Align the Spare Charger in accordance with 0-SO-250-1, "125 Volt dc Vital Battery Boards."

B. (1) INOPERABLE (2) Reduce Battery loading as necessary in accordance with 0-SO-250-1, "125 Volt dc Vital Battery Boards."

C. (1) OPERABLE (2) Align the Spare Charger in accordance with 0-SO-250-1, "125 Volt dc Vital Battery Boards."

D. (1) OPERABLE (2) Reduce Battery loading as necessary in accordance with 0-SO-250-1, "125 Volt dc Vital Battery Boards."

Answer: A.

DISTRACTOR ANALYSIS:

A. Correct, In accordance with T.S. 3.8.2.3, surveillance with bus voltage less than 125 VDC, thus the batt board is INOPERABLE. Also the guidance provided in 0-SO-250-1 is to align the spare charger to the bus to correct the low voltage condition.

B. Incorrect, Plausible because the first part is correct. In accordance with T.S. 3.8.2.3, surveillance with bus voltage less than 125 VDC, thus the batt board is INOPERABLE. Also plausible if the candidates did not recognize that the charger output breaker is open and determine that by lowering the load on the bus they would raise the voltage.

C. Incorrect, Plausible since the bus is not inoperable just because the charger is disconnected, however it is inoperable because the voltage is below 125 Vdc. Also plausible since the second part is correct.

D. Incorrect, Plausible since the bus is not inoperable just because the charger is disconnected, however it is inoperable because the voltage is below 125 Vdc. Also plausible if the candidates did not recognize that the charger output breaker is open and determine that by lowering the load on the bus they would raise the voltage Question Number: 90 Tier: 2 Group: 1 K/A: 063 DC Electrical Distribution G2.4.31 Knowledge of annunciator alarms, indications, or response procedures.

Importance Rating: 3.6 / 4.0 10 CFR Part 55: 41.10 10CFR55.43.b: 5 K/A Match: This question matches the K/A by having the candidate assess the alarms provided in the stem and determine what type of failure/malfunction has occurred and determine what type of action is required by the alarm procedures. SRO by assessing the plant conditions and selecting the appropriate section of a procedure which will correct the conditions. Also requires detailed knowledge of the selected procedure to be able to determine the correct actions.

Technical

Reference:

1-AR-M1-C (B-4) rev 0045 1-AR-M1-C (B-5) rev 0045

0-SO-250-1, 125 Volt DC Vital Power System. rev 54 Proposed references None to be provided:

Learning Objective: OPL271AOP-P.02 rev 2 Obj. 13 Question Source:

New Modified Bank Bank X Question History: SQN bank question Comments:

1. 91. Given the following plant conditions:

- Unit 1 is in Mode 3 following a planned shutdown for maintenance.

- The unit is being cooled down in accordance with GO-7, "Unit Cooldown from Hot Standby to Cold Shutdown."

- Current RCS temperature is 465oF and pressurizer pressure 1535 psig.

A pressurizer PORV inadvertently opened and was closed by placing its control switch to manual.

Which ONE of the following identifies:

(1) the procedure required to be entered to address the failure, and, (2) in accordance with Tech Spec 3.4.3 "Safety and Relief Valves - Operating,"

bases, if the problem was only associated with the PORV's automatic circuit, the PORV is _______?

(1) (2)

Procedure PORV Status A. AOP-R.05, "RCS Leak and Operable Leak Source Identification" B. AOP-R.05, "RCS Leak and Inoperable Leak Source Identification" C. AOP-I.04, "Pressurizer Instrument Operable and Control Malfunctions" D. AOP-I.04, "Pressurizer Instrument Inoperable and Control Malfunctions" Answer: C.

DISTRACTOR ANALYSIS:

A. Incorrect, Plausible because a PORV opened would be an RCS leak but AOP-R.05 is not the procedure with a section containing the Immediate Operator Actions required for addressing a failed open PORV, AOP-I.04 does have the section with the actions. The PORV remains operable if only the automatic circuit is affected.

Plausible because AOP-R.05 is the procedure for an RCS leak and the PORV being operable is correct.

B. Incorrect, Plausible because a PORV opened would be an RCS leak but AOP-R.05 is not the procedure with a section containing the Immediate Operator Actions required for addressing a failed open PORV, AOP-I.04 does have the section with the actions. The PORV is not inoperable if only the automatic circuit is affected.

Plausible because a failed open PORV is an RCS leak and in other systems failure of an automatic function will cause a component to be inoperable.

C. Correct, AOP-I.04 contains the Immediate Operator Actions in a section for response to a failed open PORV and in accordance with the Tech Spec Bases, the PORV is operable if it is capable of being manually operated to control RCS pressure (automatic operation not required).

D. Incorrect, Plausible because AOP-I.04 contains the Immediate Operator Actions in a section for response to a failed open PORV but the PORV is not inoperable if only the automatic control circuit has a problem. Plausible because AOP-I.04 containing the Immediate Operator Actions in a section for response to a failed open PORV is correct and in other systems failure of an automatic function will cause a component to be inoperable.

Question Number: 91 Tier: 2 Group: 2 K/A: 002 Reactor Coolant System G2.2.37 Ability to determine operability and/or availability of safety related equipment.

Importance Rating: 3.6 / 4.6 10 CFR Part 55: 41.7 10CFR55.43.b: 5 K/A Match: This question matches the K/A by having the candidate determine the operability of an RCS safety related piece of equipment (PORV).

Also SRO due to requiring the candidate to use knowledge from the Tech Spec basis to make the operability determination.

Technical

Reference:

AOP-R.05, RCS Leak and Leak Source Identification, rev 18 AOP-I.04, Pressurizer Instrument Malfunctions, rev 12 Tech Spec 3.4.3 bases, Ammend 308 Proposed references None to be provided:

Learning Objective: OPL271AOP-I.04 B.5 Describe the mitigating strategy for the failure that initiated entry into AOP-I.04.

Question Source:

New Modified Bank Bank X Question History: SQN bank question used on Sept 2009 NRC exam Comments:

1. 92. Given the following plant conditions:

- A vapor space LOCA occurred on Unit 1.

- Some fuel damage was experienced during the accident.

- The LOCA has been stopped and the Safety Injection has been terminated.

The operating crew has transitioned to FR-I.3, "Voids in Reactor Vessel," and is currently addressing the steps to prepare for and execute venting the reactor vessel.

The following conditions are reported to exist:

- PRT pressure is 5 psig.

- Containment hydrogen concentration is 6.3%.

- Containment Pressure is 0.29 psid.

- Neither Air Return Fan is available.

Which ONE of the following identifies ...

(1) the actions directed by the procedure relative to Hydrogen Recombiner operation and (2) the correct action associated with vessel venting?

Hydrogen Reactor Vessel Recombiner Venting A. Would be placed Vent the vessel before transitioning in service. from FR-I.3.

B. Would be placed Make transition from FR-I.3 without in service. venting the reactor vessel.

C. Would NOT be Vent the vessel before transitioning placed in service. from FR-I.3.

D. Would NOT be Make transition from FR-I.3 without placed in service. venting the reactor vessel.

Answer: D

DISTRACTOR ANALYSIS A. Incorrect, the hydrogen concentration is above the limit for placing the recombiner in service for venting the vessel. Plausible because the recombiner would be place is service if the hydrogen concentration had been between 3.0% and 6.0% and with the PRT not ruptured the vessel venting would be to the PRT, not to the containment atmosphere.

B. Incorrect, the hydrogen concentration is above the limit for placing the recombiner in service for venting the vessel. Plausible because the recombiner would be place is service if the hydrogen concentration had been between 3.0% and 6.0% and not venting the reactor vessel correct C. Incorrect, the recombiner would not be placed in service but the vessel would not be vented. Plausible because not placing the recombiner in service is correct and with the PRT not ruptured the vessel venting would be to the PRT and not to the containment atmosphere.

D. Correct, the recombiner would not be placed in service and FR1.3 would be transitioned from without venting the vessel. The hydrogen concentration in containment is above the maximum allowed to place the recombiners in service and also above the maximum allowed to vent the reactor vessel.

Question Number: 92 Tier: 2 Group: 2 K/A: 028 Hydrogen Recombiner and Purge Control System (HRPS)

G2.1.32 Ability to explain and apply system limits and precautions.

Importance Rating: 3.8 / 4.0 10 CFR Part 55: 41.10 / 43.2 / 45.12 10CFR55.42.b: 2, 5 K/A Match: Question requires the applicant to recognize that the hydrogen concentration is high enough to cause detonation with resulting equipment damage in containment if the Recombiner were to be placed in service when the procedure addresses the recombiner operation and recognize the procedure requirements will prevent placing of the recombiner in operation.

Technical

Reference:

FR-I.3, Voids in Reactor Vessel, Rev 11 E-0, Reactor Trip or Safety Injection, rev 33

ES-0.5 Appendix D Hydrogen Mitigation Actions, rev 4 Proposed references None to be provided:

Learning Objective: OPL271FR-I.3

4. Describe the bases for all limits, notes, cautions, and steps of FR-I.3.

5. Describe the conditions and reason for transitions within this procedure and transitions to other procedures.

Question Source:

New Modified Bank Bank X Question History: SQN bank question used on 2010 ILT exam.

Comments:

1. 93. Given the following plant conditions:

- Unit 1 in Mode 3 with the RCS at normal operating temperature and pressure preparing for reactor startup.

- Alarm "PS-32-104 TRAIN A AUX CONTROL AIR PRESS LOW" (M15-B) is actuated.

- Control Air pressure indications are:

- PI-32-200 Control Air Header pressure is 65 psig and lowering.

- PI-32-104 Aux Bldg Control air header A pressure is 65 psig and lowering.

- PI-32-105 Aux Bldg Control air header B pressure is 82 psig and rising.

Which ONE of the following identifies both...

1) the direction given the AUO and

2) When reviewing Tech Specs in accordance with direction in AOP-M.02, "Loss of Control Air," which train(s) of AFW will be INOPERABLE on Unit 1?

Note:

0-FCV-32-82, Aux. Compsr. A-A Aux. Bldg Isol.

0-FCV-32-85, Aux. Compsr. B-B Aux. Bldg Isol.

1-FCV-32-80, Unit 1 Train A Rx Bldg Isol.

1-FCV-32-102, Unit 1 Train B Rx Bldg Isol.

Direction U-1 AFW Trains Inoperable A. Ensure 0-FCV-32-82, Motor Driven Train A only and 0-FCV-32-85 are closed.

B. Ensure 1-FCV-32-80 Motor Driven Train A and Turbine Driven Train and 1-FCV-32-102 are closed.

C. Ensure 0-FCV-32-82, Motor Driven Train A and Turbine Driven Train and 0-FCV-32-85 are closed.

D. Ensure 1-FCV-32-80 Motor Driven Train A only and 1-FCV-32-102 are closed.

Answer: A

DISTRACTOR ANALYSIS:

A. Correct, The isolation valves should automatically close at 69 psig. With the control header pressure and Aux Bldg control header A at the same pressure with Aux Bldg B header rising it would indicate that at least one of the isolation valves did not go closed. AOP-M02 would have the operators ensure that the 0-FCV-32-82 & 85 are closed to isolate the leak. The Aux Bldg A header pressure is less than 70 psig.

According to the notes in alarm response procedure, the A header would be considered Inoperable. In accordance with the T/S Bases, the loss of a single train of essential air is no more severe than the loss of one of the motor driven trains because the TD Pump can still supply to steam generators, thus the A MD train is inoperable, however the TD pump is not.

B. Incorrect, Plausible if the candidate thinks that 1-FCV-32-102 is also isolated by the AOP-M.02 for leak isolation. This valve would automatically isolate at 50 psig. Also Only the Train A is inoperable due to loss of air to the A MD driven LCVs. The T-D AFW pump can still supply 2 steam generators due to B train of air available, thus the TD pump is not inoperable.

C. Incorrect, Plausible since the first part of is correct. Also the candidate may get confused on whether the TD AFW train is also inoperable, however in accordance with the T/S Bases, the loss of a single train of essential air is no more severe than the loss of one of the motor driven trains because the TD Pump can still supply two steam generators.

D. Incorrect, Plausible if the candidate thinks that 1-FCV-32-102 is also isolated by the AOP-M.02 for leak isolation. This valve would automatically isolate at 50 psig. Also the second part is correct, the TD pump is not inoperable.

Question Number: 93 Tier: 2 Group: 2 K/A: 079 Station Air A2.01 Ability to (a) predict the impacts of the following malfunctions or operations on the SAS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Cross-connection with IAS Importance Rating: 2.9 / 3.2 10 CFR Part 55: 41.5 / 43.5 / 45.3 / 45.13 10CFR55.43.b: 2

K/A Match: KA is matched because the question requires the ability to predict the impact a loss of air has on the operability of the AFW system and requires using the information in the bases of the Tech Spec in determining the impact.

Technical

Reference:

Technical Specification Bases AOP-M.02, Loss of Control Air, Rev.

Proposed references None to be provided:

Learning Objective: OPL271AOP-M.02 obj 7 & 8 Question Source:

New Modified Bank Bank X Question History: SQN bank question with the stem and choices modified Comments:

1. 94. In accordance with FHI-3, "Movement of Fuel," which ONE of the following identifies the maximum number of NEW and IRRADIATED fuel assemblies within the areas listed below that can be located out of approved storage locations?

New Fuel Assemblies Irradiated Fuel Assemblies within the fuel-handling area within the refueling canal A. 1 2 B. 1 3 C. 2 2 D. 2 3 Answer: B.

DISTRACTOR ANALYSIS:

A. Incorrect, FHI-3 Limitation B allows one non-irradiated nuclear fuel assembly within the fuel-handling area and three ( not 2) nuclear fuel assemblies within the refueling canal to be out of approved storage locations.

B. Correct, FHI-3 Limitation B allows one non-irradiated nuclear fuel assembly within the fuel-handling area and three nuclear fuel assemblies within the refueling canal to be out of approved storage locations.

C. Incorrect, FHI-3 Limitation B allows one (not 2) non-irradiated nuclear fuel assembly within the fuel-handling area and three (not 2) nuclear fuel assemblies within the refueling canal to be out of approved storage locations.

D. Incorrect, FHI-3 Limitation B allows one (not 2) non-irradiated nuclear fuel assembly within the fuel-handling area and 3 nuclear fuel assemblies within the refueling canal to be out of approved storage locations.

Question Number: 94 Tier: 3 Group: n/a K/A: G 2.1.42 Conduct of Operations Knowledge of new and spent fuel movement procedures Importance Rating: 2.5 / 3.4 10 CFR Part 55: 41.10 10CFR55.43.b: 7 K/A Match: This question matches the K/A by testing the candidates knowledge of Fuel movement procedure requirements for new and irradiated fuel. SRO for Sequoyah Station due to the fact that fuel movement requirements is and SRO specific task.

Technical

Reference:

FHI-3, Movement of Fuel, Rev 70 Proposed references None to be provided:

Learning Objective: OPT200.FH rev 6 Obj. 7.c Cognitive Level:

Higher Lower X Question Source:

New Modified Bank Bank X Question History: SQN bank question used on SRO Audit Exam 1/2009, and Sept 2010 exam.

Comments:

1. 95. Which ONE of the following completes the statements below?

In accordance with NPG-SPP-10.1, "System Status Control," the ___(1)_____

is responsible for authorizing the relaxation of status control.

And Maintenance requests that an additional drain valve be opened, located inside an isolation boundary, for a system where status control has been relaxed. The valve currently is controlled by a Work Order. The US __(2)____ required to log the change in position on NPG-SPP-10.1, Attachment 3, "Deviation Tracking Sheet."

A. (1) Shift Manager (2) is B. (1) Shift Manager (2) is NOT C. (1) Operations Superintendent (2) is D. (1) Operations Superintendent (2) is NOT Answer: D DISTRACTOR ANALYSIS:

A. Incorrect, Plausible since the shift manager is normally responsible for the control of all shiftly functions and duties. The second part is plausible since normally any valve manipulation that occurs inside a relaxed status control area not controlled by a hold or work order would be logged in attachment 3.

B. Incorrect, Plausible since the shift manager is normally responsible for the control of all shiftly functions and duties. The second part is correct.

C. Incorrect, The first part is correct. The second part is plausible since normally any valve manipulation that occurs inside a relaxed status control area not controlled by a hold or work order would be logged in attachment 3.

D. Correct, In accordance with NPG-SPP-10.1, the Operations Superintendent is responsible for relaxing system status control. Also for valve manipulations which may occur inside a previously established system boundary and/or controlled by a work order, the US is NOT required to log the item. It will be re-aligned to the correct position post maintenance when doing return to service lineups.

Question Number: 95 Tier: 3 Group:

K/A: G 2.2.15 Equipment Control Ability to determine the expected plant configuration using design and configuration control documentation, such as drawings, line-ups, tag-outs, etc.

Importance Rating: 3.9 / 4.3 10 CFR Part 55: 41.10 10CFR55.43.b: 3 K/A Match: This question matches the K/A by having the candidate identify the requirements for maintaining configuration control during shutdown conditions. SRO level by asking the candidate the SRO levels of authority required to relax configuration control.

Technical

Reference:

NPG-SPP-10.1, System Status Control, rev 0003 Proposed references None to be provided:

Learning Objective: OPL271NPG-SPP-10.1 rev 0 Obj. 1.b & c Question Source:

New X Modified Bank Bank Question History:

Comments:

1. 96. Given the following conditions:

- Unit 2 is coming out of a refueling outage.

- 0-SI-SXV-074-203.2, "Full Stroking of RHR Valves 2-FCV-74-1 and 2-FCV-74-2" is being performed.

- Personnel are in position and monitoring valve operation locally.

- 2-FCV-74-1 took 118.8 seconds to open on first cycle.

- 2-FCV-74-1 took 87.5 seconds to close on first cycle.

- Valve operated smoothly and all indication worked properly.

Which ONE of the following identifies the current status of the valve and the correct procedural action to take under these conditions?

Reference Provided A. Valve opening time is in the ALERT range.

Cycle the valve two more times recording the opening time.

Notify System Engineering representative.

Valve is INOPERABLE until System Engineering evaluates test data.

B. Valve opening time is in the ALERT range.

Cycle the valve two more times recording the opening time.

Notify System Engineering representative.

Valve is OPERABLE until System Engineering evaluates test data.

C. Valve closing time is in the ALERT range.

Cycle the valve two more times recording the closing time.

Notify System Engineering representative.

Valve is OPERABLE until System Engineering evaluates test data.

D. Valve closing time is in the ALERT range.

Cycle the valve two more times recording the closing time.

Notify System Engineering representative.

Valve is INOPERABLE until System Engineering evaluates test data.

Answer: B.

DISTRACTOR ANALYSIS:

A. Incorrect. Plausible since the first part is correct for Unit 2. Closing time is in the ALERT range for Unit 1, but is Acceptable for Unit 2. Also plausible since the candidate could decide that the valve is inoperable, however the valve is still operable until such time as a System Engineering Evaluation determines the stroke time to be unacceptable..

B. Correct. Per 0-SI-SXV-074-266.0 Appendix A the valve Opening time for Unit 2 is in the Alert range and the closing time is acceptable. This would require System Engineering to Evaluate the operability of the valve.

C. Incorrect. Plausible if the candidate looks at the wrong line in Appendix A. Closing time is in the ALERT range for Unit 1, but is Acceptable for Unit 2. The valve remains operable pending System Engineering evaluation.

D. Incorrect. Plausible if the candidate looks at the wrong line in Appendix A. Also the valve is still operable until such time the System Engineering Evaluation determines the stroke time to be unacceptable.

Question Number: 96 Tier: 3 Group:

K/A: Equipment Control G 2.2.20 Knowledge of the process for managing troubleshooting activities.

Importance Rating: 2.6 / 3.8 10 CFR Part 55: N\A 10CFR55.43.b: 5 K/A Match: Question meets K/A criteria for managing a troubleshooting activity for a RHR valve stroke in which the closing time falls into the Alert Range. The SRO candidate must know the proper procedure and the requirements therin for the direction to procede.

Technical

Reference:

0-SI-SXV-074-266.0, Appendix A 0-SI-SXV-074-203.2 Proposed references 0-SI-SXV-074-266.0 pages 1-19 Rev 0018 to be provided:

Learning Objective: OPL271OPDP-8, rev 2 Obj. 10

Question Source:

New Modified Bank X Bank Question History: SQN bank question that was modified to increase the plausibility of the distractors and Modified stem data to change answer from "C" to "B" Comments:

1. 97. Given the following plant conditions:

- A LOCA occurred on Unit 1.

- A Site Area Emergency has been declared.

- The Containment Barrier is intact (i.e. no loss or potential loss of containment).

- The Containment Critical Safety Function Status Tree (FR-Z) is Yellow due to high radiation in containment.

- Containment pressure is 2.1 psig and decreasing.

- There is no release to the environment in progress.

- The "A" train of containment spray is operating normally for plant conditions.

- The "B" containment spray pump tripped, 25 minutes ago, after pump amps were observed to be oscillating.

- Authorization has been given for an emergency responder to receive an emergency exposure of 11 Rem TEDE in order to restore "B" train containment spray.

Which ONE of the following completes the statement below?

The decision __(1)___ the requirements of EPIP-15, "Emergency Exposure Guidelines," because ___(2)_____.

(1) (2)

A. met emergency exposure limits apply during any REP classification B. violated emergency exposure limits only apply during a General Emergency C. met emergency exposure is necessary to maintain critical safety functions D. violated emergency exposure is NOT necessary to maintain critical safety functions Answer: D.

DISTRACTOR ANALYSIS:

A. Incorrect, Plausible if applicant believes that B train CS is necessary to prevent a release and emergency exposure limits always apply during REP to correct any situation. However per EPM-4 Yellow paths are OPTIONAL to the operator therefore not required to maintain critical safety functions. The REP does not recognize yellow paths for REP criteria to protect public health.

B. Incorrect. Plausible since the action is correct, however the reason is incorrect.

Also plausible because there are different restrictions depending on REP classifications and in this case there is no direct threat to public heath.

C. Incorrect, Plausible if applicant believes that B train CS is necessary to prevent a release or to maintain FR-Z yellow or restore to green. However "B" train CS is not necessary to maintaining critical safety functions or to protect the public since "A" train is reducing CNMT pressure without it. Per EPM-4 Yellow paths are OPTIONAL to the operator therefore not required to maintain critical safety functions. The REP does not recognize yellow paths for REP criteria to protect public health.

D. Correct, "B" train CS is not necessary to maintain critical safety functions or to protect the public. Per EPM-4 Yellow paths are OPTIONAL to the operator therefore not required to maintain critical safety functions. The Radiological Emergency Plan (REP) does not recognize yellow paths for REP criteria to protect public health, thus does not meet requirement for emergency exposures to maintain critical safety functions per EPIP-15.

Question Number: 97 Tier: 3 Group:

K/A: G 2.3.4 Knowledge of radiation exposure limits under normal or emergency conditions.

Importance Rating: 3.2 / 3.7 10 CFR Part 55: 41.10 10CFR55.43.b: NA K/A Match: KA is matched because the question tests the knowledge of radiation contol limits and how to apply them during implimentation of the Radiological Emergency Plan.

Technical

Reference:

EPIP-15, Emergency Exposure Guidelines, Revision 9

Proposed references None to be provided:

Learning Objective: OPL271REP Discuss the Radiological Emergency Plan

f. Describe the process of authorizing Emergency Radiological Exposures in accordance with EPIP-15 Question Source:

New Modified Bank Bank X Question History: SQN ILT 1201 NRC Exam Comments: Reordered answer and distractors.

1. 98. Given the following plant conditions:

- A diving operation in the Spent Fuel Pit is planned to commence later in the shift.

Which ONE of the following completes the statements below?

The dive __(1)__ the requirements of NPG-SPP-07.3, " Work Activity Risk Management Process," to be classified as a HIGH risk activity.

If the rad level in the area of the dive is estimated at 58 Rad/hr, in accordance with RCI-14, "Radiation Work Permit (RWP) Program," the RWP would be required to be approved by __(2)___ .

(1) (2)

A. meets Radiation Protection Manager only B. meets Radiation Protection Manager and Plant Manager C. does NOT meet Radiation Protection Manager only D. does NOT meet Radiation Protection Manager and Plant Manager Answer: B.

DISTRACTOR ANALYSIS:

A. Incorrect, Plausible, because the diving operation being a HIGH risk activity is correct and for values of < 50 rad/hr the Radiation Protection manager would be the only approval that is required.

B. Correct, NPG -SPP-07.3, Work Activity Risk Management Process, Attachment 2 Initial Risk Characterization identifies that 'All activities associated with Diving Operations, due to industry operating experience' are to be characterized as HIGH risk and RCI-14, RADIATION WORK PERMIT (RWP) PROGRAM identifies 'The Plant Manager and Radiation Protection Manager must approve in writing entries into areas where whole body dose rates are = 50 Rad/hour.

C. Incorrect, Plausible, because the applicant may recall the criteria on the Site High Focus Risk and Attachment 5 'PWR Operational Risk Review' sheet (RED SHEET) and identify that diving is not on the sheet and for values of < 50 rad/hr the Radiation Protection manager would be the only approval that is required.

D. Incorrect, Plausible, because the applicant may recall the criteria on the Site High Focus Risk and Attachment 5 'PWR Operational Risk Review' sheet (RED SHEET) and identify that diving is not on the sheet. Also plausible because the Plant Manager and Radiation Protection manager being required to approve in writing entries into areas where whole body dose rates are = 50 Rad/hour is correct.

Question Number: 98 Tier: 3 Group: n/a K/A: G2.3.7 Radiation Control Ability to comply with radiation work permit requirements during normal or abnormal conditions.

Importance Rating: 3.5 / 3.6 10 CFR Part 55: 41.12 / 45.10 10CFR55.43.b: Related to item 7.

K/A Match: This question matches the K/A by testing the candidates knowledge of the RWP requirements for entry into high radiation areas and SRO level by testing the candidates knowledge of the risk factor associated with diving operations.

Technical

Reference:

NPG-SPP-07.3, Work Activity Risk Management Process, Rev. 0005 RCI-14, Radiation Work Permit (RWP) Program, Revision 0052 Proposed references None to be provided:

Learning Objective: OPL271SPP-7.1 obj. 4 & 9 Question Source:

New Modified Bank Bank X Question History: SQN ILT 1201 Exam, SQN ILT 1311 NRC Exam Comments:

1. 99. Given the following plant conditions:

0200 - Unit 2 is operating at 100% power when implementation of AOP-C.04, "Shutdown from Auxiliary Control Room," was required.

0204 - All main control room actions have been completed and the main control room is evacuated.

0212 - All transfer switches are placed in the required position in accordance with AOP-C.04, Checklist 2, Unit 2 Auxiliary Control Room.

0218 - AUO reports 6.9Kv Shutdown Board equipment is configured in accordance with Checklist 4.

Which ONE of the following completes the statements below?

The conditions require the declaration of a/an ___(1)___ in accordance with the Radiological Emergency Plan.

___(2)____ control is available for Pressurizer Backup Heaters 2A-A.

REFERENCE PROVIDED A. (1) Alert (2) ONLY manual B. (1) Alert (2) BOTH manual and automatic C. (1) Site Area Emergency (2) ONLY manual D. (1) Site Area Emergency (2) BOTH manual and automatic Answer: B.

DISTRACTOR ANALYSIS:

A. Incorrect, Plausible because the declaration of an ALERT is correct. Also plausible since most automatic functions are disabled when the components transfer switch is in the AUX position however the PZR B/U heaters can be operated either manually or in automatic.

B. Correct, Because control was transferred (switches on L11A and L11B placed in AUX position as reported by the completion of checklist 2) and control established within the 15 minutes, the required declaration is an ALERT. Also, after 6.9kV Shutdown Board equipment has been configured per Attach 4, the Pressurizer Backup Heaters 1A-A can be operated in either automatic or manual from the control on the breaker door.

C. Incorrect, Plausible because the declaration of an SAE would be correct if the candidate thought that control had not been established within the 15 minute allowance and needed to wait for the report from the AUO that checklist 4 had been completed. However by reporting the completion of checklist 2 the AUO has signified that all switches on L11A and L11B have been placed in AUX. Also plausibe since most automatic functions are disabled when the components transfer switch is in the AUX position.

D. Incorrect, Plausible because the declaration of an SAE would be correct if candidate thought that control had not been established within the 15 minute allowance and needed to wait for the report from the AUO that checklist 4 had been completed. However by reporting the completion of checklist 2 the AUO has signified that all switches on L11A and L11B have been placed in AUX. Also the heaters having both manual and automatic control after the transfer switch is in the AUX position is correct.

Question Number: 99 Tier: 3 Group: n/a K/A: G 2.4.35 Knowledge of local auxiliary operator tasks during an emergency and the resultant operational effects.

Importance Rating: 3.8 / 4.0 10 CFR Part 55: 41.10 /43.5 /45.13 10CFR55.43.b: 5 K/A Match: This question matches K/A by requiring candidate to know the operational effect of placing 6.9Kv breakers to the aux position. SRO

by determining the EPIP Classification associated with abondoning the Main Control Room.

Technical

Reference:

AOP-C.04, "Shutdown from the Auxiliary Control Room, Revision 18 EPIP-1, Emergency Plan Classification Matrix, Revision 44 Proposed references EPIP-1 Section 4 (pgs 25-32) to be provided:

Learning Objective: OPL271AOP-C.04 obj. 3 & 12 OPL271REP obj. 3 Question Source:

New Modified Bank Bank X Question History: SQN bank question (originally DC Cook question) Also WBN question 034 K6.02 stem modified Comments: This question was on 1202 NRC exam

1. 100. Given the following plant conditions:

- A General Emergency has been declared due to a plant event that resulted in a radiation release.

- The duration of the release is estimated to be 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

- Wind Direction is from 181 degrees @ 10 mph.

- Dose projections and surveys have not yet been performed Which ONE of the following identifies the Protective Action Recommendations required in accordance with EPIP-5, "General Emergency?"

REFERENCE PROVIDED A. Shelter the entire 10 mile EPZ, consider issuance of KI in accordance with the State Plan.

B. Evacuate A1-B1-C1-D1, A2, A3 and B2. Shelter the remainder of the 10 mile EPZ.

C. Evacuate A1-B1-C1-D1, A2 and B2. Shelter the remainder of the 10 mile EPZ.

D. Evacuate A1-B1-C1-D1, A2, -5, -6, B2, -3, -4. Shelter the remainder of the 10 mile EPZ.

Answer: C.

DISTRACTOR ANALYSIS:

A. Incorrect, Plausible, since this would be selected if the applicant determines that a short duration release where plant areas cannot be evacuated prior to plume arrival applies.

B. Incorrect, Plausible, since this could be selected if the applicant makes an error selecting the appropriate wind direction limit.

C. Correct, this is the required sectors per Appendix A of EPIP-5 for the conditions given.

D. Incorrect, Plausible since this could be selected by the applicant if they make an error when determining whether or not dose will be exceeded at and beyond the 5 mile radius.

Question Number: 100 Tier: 3 Group: n/a K/A: G 2.4 Emergency Procedures/Plan:

2.4.38 Ability to take actions called for in the facility emergency plan, including supporting or acting as emergency coordinator if required.

Importance Rating: 2.4 / 4.4 10 CFR Part 55: 41.10 / 43.5 / 45.11 10CFR55.43.b: 3, 5, SRO task K/A Match: Question requires the ability to determine the proper Protective Action Recommendation during a General Emergency and is SRO because the SRO (SM) can be the Site Emergency Director.

Technical

Reference:

EPIP-5, General Emergency, Revision 44 Proposed references Appendix A and H of EPIP-5 (3 pages) to be provided:

Learning Objective: OPL271REP

4. DETERMINE protective action recommendations using appropriate procedures.

Question Source:

New Modified Bank X Bank Question History: SQN bank question G 2.4.38 written for a 2009 audit exam with minor wording changes due to procedure revisions., SQN ILT 1211, SQN ILT 1305 Audit, SQN ILT 1311 Comments: Modified question stem by changing wind direction which changes the sectors and moved correct answer.