ML17172A703

Section 3 - North Anna 2016301 SRO Written Draft - Delay Release 2 Yrs

ML17172A703
Person / Time
Site:	North Anna
Issue date:	06/21/2017
From:	NRC/RGN-II
To:	Virginia Electric & Power Co (VEPCO)
References
Download: ML17172A703 (96)

Text

QUESTIONS REPORTfor SRO Exam (2-28)076 - 008AG2.4.6 001/SRO/T1/G1/3.7/4.7/NEW//Unit-1 was operating at 100% power when a SGTR occurred on "B" S/G.* The crew manually tripped the Reactor and completed 1-E-0 "Reactor Trip or Safety Injection"* Transition was made to 1-E-3 "Steam Generator Tube Rupture"

• RCPs are secured.* Both PRZR PORV's are available.

• Crew is currently depressurizing the RCS to minimize break flow and refill PRZR using 1-RC-PCV-1456 "PRZR PORV".

• Ruptured S/G pressure is 1035 psig* TSC has been activated and has directed using 1-ES-3.1 "Post-SGTR cooldown using backfill" for recoveryCurrent conditions:* RWST level = 93% * "B" S/G level = 86% NR* PRZR level is 29% and increasing.* RCS Pressure is 1034 psig.* Crew is stopping the RCS Depressurization when 1-RC-PCV-1456 will not close.

• While attempting to close 1-RC-MOV-1535, the breaker trips. Based on Current Conditions the proper procedure transition will be: 1-ECA-3.1 "SGTR with Loss of Reactor Coolant Subcooled Recovery Desired" 1-ECA-3.2 "SGTR with Loss of Reactor Coolant Saturated Recovery Desired" 1-E-1 "Loss of Reactor or Secondary Coolant" 1-ES-3.1 "Post-SGTR Cooldown using backfill" A.B.C.D.

Distractor Analysis:This requires the examinee to have the knowledge of EOP mitigation strategies as theapply to Pressurizer Vapor Space Accident CORRECT A. 1-ECA-3.1 "SGTR with Loss of Reactor Coolant Subcooled Recovery Desired"Transition to 1-ECA-3.1 from 1-E-3 is made due to the following: Ruptured S/G not isolated from one intact S/G Ruptured S/G < 350 psig No Intact S/G

< 250 psid between faulted S/G and highest intact S/G RCS subcooling < 45 DEFG PRZR level < 21% with uncontrolled decrease RCS Press Decreasing Nonisolable PZR PORV stuck openINCORRECT B. 1-ECA-3.2 "SGTR with Loss of Reactor Coolant Saturated Recovery Desired"Incorrect but plausible if the SRO candidate is unaware that the only transition to1-ECA-3.2 "SGTR with Loss of Reactor Coolant Saturated Recovery Desired is from 1-ECA-3.1. There are no direct transitions from 1-E-3. The transition entry conditionsfrom ECA-3.1 is RWST level < 58% or Containment sump level is lower than expected.or Ruptured S/G level > 90% [85%] INCORRECT C. 1-E-1 "Loss of Reactor or Secondary Coolant"Incorrect but plausible because the SRO candidate knows that E-1 provides guidancefor operating personnel to recover from a loss of reactor or secondary coolant. Thequestion states that the crew went directly from E-0 to E-3 and then the PORV andblock valve fails open. The candidate may think the proper transition would be to E-1to combat the loss of reactor coolant prior to transitioning to an ECA procedure. INCORRECT D. 1-ES-3.1 "Post-SGTR Cooldown using backfill"Incorrect but plausible if the SRO candidate focus in on the TSC directive to recoverusing 1-ES-3.1. 1-ES-3.1, 1-ES-3.2, and 1-ES-3.3 are recovery procedures and onlytransitioned to after 1-E-3 is completed.

K/A:008AG2.4.6 Pressurizer (PZR) Vapor Space Accident (Relief Valve Stuck Open)Knowledge of EOP mitigation strategies.Technical

References:

1-E-3 "Steam Generator Tube Rupture"Westinghouse background for E-31-ECA-3.1 "SGTR with Loss of Reactor Coolant Subcooled Recovery Desired"Westinghouse background for ECA-3.11-ECA-3.2 "SGTR with Loss of Reactor Coolant Saturated Recovery Desired"Westinghouse background for ECA-3.2 1-ECA-3.3 "SGTR without Pressurizer Pressure Control" Westinghouse background for ECA-3.3 1-E-1 "Loss of Reactor or Secondary Coolant" Westinghouse background for E-111715-FM-093B (Sh 1 of 3) Reactor Coolant system References provided to applicants: NoneLearning Objective:U 13844Entry conditions for 1-ECA-3.1 "SGTR with Loss of Reactor Coolant SubcooledRecovery Desired"Conditions that result in leaving 1-ECA-3.1 "SGTR with Loss of Reactor CoolantSubcooled Recovery Desired"U 9594Explain how it is determined that a transition from ECA-3.1 "SGTR with Loss of ReactorCoolant Subcooled Recovery Desired" to ECA-3.2 "SGTR with Loss of Reactor CoolantSaturated Recovery Desired" is appropriateU 13845List the purpose of 1-ECA-3.2 "SGTR with Loss of Reactor Coolant SaturatedRecovery Desired"U 13846List the purpose of 1-ECA-3.3 "SGTR Without Pressurizer Pressure Control"List the entry Conditions of 1-ECA-3.3U 13882Explain the purpose of post SGTR cooldown procedures 1-ES-3.1, 3.2, 3.3.U 13683List various information associated with E-1 "Loss of Reactor or Secondary Coolant" Question Source: NEWQuestion History: None Question Cognitive Level: Comprehension/Analysis 10 CFR Part 55 Content:SRO only 10 CFR-55.43(b)(5)

Assessment of facility conditions and selection of appropriate procedures duringnormal, abnormal. and emergency situations. 10 CFR-55.41(a)(b)(10) Administrative, normal, abnormal, and emergency operating procedures at the facility.Comments:

This question matches the K/A statement by requiring the SRO applicant to have theknowledge of EOP mitigation strategies as they apply to Pressurizer Vapor SpaceAccident. To arrive at the correct answer the SRO applicant must recognize the properprocedure during a emergency situation that would be required to mitigate from a pressurizer vapor space accident when the PORV fails open and can not be isolatedduring the performance of 1-E-3 "Steam Generator Tube Rupture".

077 - 054AG2.1.19 001/SRO/T1/G1/3.9/3.8/NEW//NOGiven the following conditions:Unit 1 was at 100% power when a Loss of all Main Feed Water occurred The crew has just transitioned to 1-ES-0.1, Reactor Trip ResponseThe following indications are noted on PCS:RCS Tave = 548° F and stable Based on the indications of the attached PCS trend which ONE of the following will theSRO direct the crew to perform?(Reference provided)Transition to 1-FR-H.1, Response to Loss of Secondary Heat SinkRemain in 1-ES-0.1 and perform 1-AP-22.4, Loss of Both Motor-Driven AFWPumpsTransition to 1-FR-H.5, Response to Steam Generator Low LevelReturn to 1-ES-0.0, Re-DiagnosisA.B.C.D.

Distractor Analysis:This requires the examinee to have the ability to use plant computer to evaluate systemor component status associated with a Loss of Main Feedwater A. CORRECT Transition to 1-FR-H.1, Response to Loss of Secondary Heat SinkGiven the indications showing on the PCS graph, ( AFW Flow to "A" SG At 500 GPM &"A" SG WR Level Decreasing) AFW is not reaching the A SG and therefore a loss ofheat sink is indicated.B. INCORRECT Remain in 1-ES-0.1 and perform 1-AP-22.4, Loss of Both Motor-Driven AFW Pumps Plausible because if the candidate thinks that the indicated AFW flow is reaching theSG (Based on AFW Flow to "A" SG At 500 GPM) then this would be the requiredaction.C. INCORRECT Transition to 1-FR-H.5, Response to Steam Generator Low LevelPlausible because FR-H.5 conditions are met since all narrow range levels are below11%.D. INCORRECT Return to 1-ES-0.0, Re-DiagnosisPlausible because re-diagnosis can be used in many situations after E-0 has beenexited but the candidate must know that SI should be in service or required to enterES-0.0.

K/A:054AG2.1.19Loss of Main FeedwaterAbility to use plant computer to evaluate system or component statusTechnical

References:

PCS1-F-0References provided to applicants: PCS Printout Question History: None Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content:SRO only 10 CFR-55.43(b)(5)Assessment of facility conditions and selection of appropriate procedures duringnormal, abnormal and emergency situations 10 CFR-55.41(10)Administrative, normal, abnormal, and emergency operating procedures for the facilityComments:

This question matches the K/A statement by requiring the SRO applicant show the ability to use plant computer to evaluate system or component status associated with aLoss of Main Feedwater. To arrive at the correct answer the SRO applicant mustrecognize what kind of trend is associated with feed water flow going to A S/G, andthen understand what affect this flow will have on indicated WR "A" S/G wide rangelevel, then selecting the proper procedural flowpath.

QUESTIONS REPORTfor SRO Exam Jan Submittal

1. 078 - 055EA2.03 001/SRO/T1/G1/3.9/4.7/MODIFIED//Unit 1 was operating at 75% power when a loss of AC power occurred. The followingplant conditions exist: The crew is performing the actions of 1-ECA-0.0 "Loss of All AC Power"* The operators completed Step 10 placing various control switches in thePull-To-Lock position, including the charging pump switches.* E Transfer bus is being supplied by the SBO* 1J 4160 bus has been restored* The Shift Technical Advisor reports a red path still exist on heat sink.Based on the above information, the crew shoul d ____________ .Proceed to step 31 of 1-ECA-0.0, to facilitate recovery actions.Continue on sequentially with 1-ECA-0.0 from step 11 until transition, to appropriaterecovery guildelines.Transition to 1-E-0 "Reactor Trip or Safety Injection", then when directed, transitionto 1-FR-H.1 "Response to Loss of Secondary Heat Sink".Complete 1-ECA-0.0, then immediately transition to 1-FR-H.1 "Response to Loss ofSecondary Heat Sink".

A.B.C.D.Distractor Analysis:This requires the examinee to have the ability to determine or interpret actions necessary to restore power as they apply to a Station Blackout. A. CORRECT Proceed to step 31 of 1-ECA-0.0, to facilitate recovery actions.Correct as per the Westinghouse background documents for this caution in ECA-0.0,which is prior to step 10, "To minimize the deterioration of plant conditions, recovery actions should be started as soon as AC power is restored". ECA-0.0 is written suchthat step 31 can be entered from any step that follows this caution. B. INCORRECT Continue on sequentially with 1-ECA-0.0 from step 11 until transition to appropriate recovery guildelines.Incorrect but plausible if the candidate is not familiar with the Caution in 1-ECA-0.0"When power is restored to any AC emergency bus, then, to facilitate recovery actions,recovery should continue with Step 31. Also, at the beginning of ECA-0.0, there is aWednesday, January 27, 2016 1:57:56 PM 1

QUESTIONS REPORTfor SRO Exam Jan Submittalnote "CSF Status Trees should be monitored for information only. FRs should not beimplemented". C. INCORRECT Transition to 1-E-0 "Reactor Trip or Safety Injection", then when directed, transition to 1-FR-H.1 "Response to Loss of Secondary Heat Sink".Incorrect but plausible if the candidate feels the need to verify equipment and ESFactuation per 1-E-0 after the AC bus is returned to service.D. INCORRECT Complete 1-ECA-0.0, then immediatly transition to 1-FR-H.1 "Response to Loss of Secondary Heat Sink".Incorrect but plausible Note in ECA-0.0 "CSF Status Trees should be monitored forinformation only. FRs should not be implemented". This priority is necessary since all FRGs are written on the premise that at least one ac emergency bus is energized. Butin this case the recovery procedure from 1-ECA-0.0 will be either 1-ECA-0.1 "Loss of allAC power Recovery without SI required" which contains a note at the beginning of theprocedure "CSF status trees should be monitored for information only. FRs should notbe initiated before completion of Step 9" or ECA-0.2 "Loss of all AC power Recoverywith SI required" which contains an note at the beginning of the procedure "CSF statustrees should be monitored for information only. FRs should not be initiated beforecompletion of Step 11". Actions in guideline ECA-0.0 defeat automatic loading of majorplant equipment on the energized AC emergency bus. Steps 1 through 9 of ECA-0.1starts normal operational equipment as necessary based on plant conditions, andSteps 1 through 11 of ECA-0.2 start safeguards equipment as necessary based onplant conditions. Both of these procedures have priority over the FRGs. K/A: 055EA2.03 Loss of Offsite and Onsite Power (Station Blackout)

Ability to determine or interpret the following as they apply to a Station Blackout:Actions necessary to restore powerTechnical

References:

1-ECA-0.0 "Loss of all AC Power" Westinghouse background information ECA-0.0 "Loss of all AC Power" 0-OP-6.4 "Operation of the SBO Diesel (SBO event)"DWG 11715-FE-1BB.

DWG 11715-FE-21M NAPs UFSAR 1-F-0 "Critical Safety Function Status Trees" Wednesday, January 27, 2016 1:57:56 PM 2

QUESTIONS REPORTfor SRO Exam Jan SubmittalReferences provided to applicants: NoneLearning Objective:U 13830 List the following information associated with 1-ECA-0.0 "Loss of All AC Power" Purpose of the procedureModes of applicabilityEntry Conditions Major action categories Conditions that result in leaving the procedure U 13831Explain why the FR procedures are not implemented during the performance of1-ECA-0.0 Question Source: Modified NAPS Vision Exam Bank Question History: Unit 1 was operating at 60% power when a loss of all AC power occurred. The followingplant conditions exist:

The crew is performing the actions of 1-ECA-0.0 "Loss of All AC Power" The operators have just completed placing various control switches in the PULL-TO-LOCK position, including the charging pump switches The "J" EDG has been started locally and 1J bus has been restored The STA reports a red path still exists on heat sinkBased on the above information, the crew should _________.A. Proceed to step 29 of 1-ECA-0.0 (CORRECT) B. Continue on with 1-ECA-0.0 from step in effect until transition to appropriate recovery procedureC. Transition to 1-E-0 "Rx Trip or SI" then when directed transition to 1-FR-H.1 "Response to Loss of Secondary Heat Sink" D. Transition directly to 1-FR-H.1 "Response to Loss of Secondary Heat Sink"Question Cognitive Level: Comprehension or Analysis10 CFR Part 55 Content:SRO only - 10 CFR-55.43(b)(5)

Assessment of facility conditions and selection of appropriate procedures, duringnormal, and emergency situations.

10 CFR-55.45(13)

Demonstrate the applicants ability to function within the control room team as appropriate to the assigned position, in such a way that the facility licensee'sprocedures are adhered to and that the limitations in its license and amendments arenot violated. Wednesday, January 27, 2016 1:57:56 PM 3

QUESTIONS REPORTfor SRO Exam Jan SubmittalComments:This question matches the K/A statement by requiring the SRO applicant to correctlyinterpret a set of given plant conditions during a Station Blackout (Loss of All ACPower) casualty, and then determine the appropriate procedure to take actionsnecessary to mitigate the accident. To arrive at the correct answer, the SRO applicantmust recognize that in ECA-0.0 scenario FR procedures are not implemented due toinadequate electrical power situation, and also to recognize the need to jump ahead inthe procedure to steps that facilitate recovery actions after power is restored to any AC emergency bus. Also, have knowledge that FRs will be monitored until equipment isstarted based on plant conditions before FRs become applicable in the recoveryprocedures that are transitioned to from ECA-0.0 Wednesday, January 27, 2016 1:57:56 PM 4

QUESTIONS REPORTfor SRO Exam Jan Submittal

2. 079 - 058AG2.1.20 001/SRO/T1/G1/4.6/4.6/MODIFIED//Unit-2 in operating at 100% power when DC bus 2-III is lost.

• The crew is performing the steps of attachement 9 "Loss of DC Bus 2-III of 0-AP-10 "Loss of Elect Power"* The operating crew can not re-energize 125-volt DC bus* VCT level = 20%

• Pressurizer level > 28% The SRO will initiate ______(1)_______, and direct the crew to _____(2)_______. (1) 2-AP-49.1 "Loss of Normal and Excess Letdown" (2) Swap Charging suction to RWST (1) 2-AP-49 "Loss of Normal Charging" (2) Isolate letdown due to 2-CH-FCV-2122 being closed (1) 2-AP-49 "Loss of Normal Charging" (2) Place 2-CH-FCV-2122 in local control at Aux Shutdown Panel and control flow (1) 2-AP-49.1 "Loss of Normal and Excess Letdown" (2) Reduce Makeup flow and Maximize RCS sample flow A.B.C.D.Distractor Analysis:This requires the examinee to have the ability to interpret and execute procedure steps for Loss of DC power A. INCORRECT (1) 2-AP-49.1 "Loss of Normal and Excess Letdown" (2) Swap Charging suction to RWSTIncorrect but plausible, first part is correct, 0-AP-10 "Loss of Power", attachment 19 willdirect the crew to initiate 2-AP-49.1 "Loss of Normal and Excess Letdown" If theoperating crew can not re-energize 125-volt DC bus 2-III and pressurizer level isgreater than 28%, 2-AP-49.1 will then direct the crew to minimize Reactor CoolantSystem (RCS) makeup flow and to maximize RCS sample flow, to reduce the rate ofPressurizer level increase, because of the Loss of Normal and Excess Letdown. Second part is incorrect but plausible if the candidate associates the loss of normal andexcess letdown to a reduction is VCT level, then swapping to the RWST could beplausible, also the step prior to checking Pressurizer level in 0-AP-10, attachment 9,checks VCT level and if < 15% requires transferring charging pump suction to theRWST. B. INCORRECT (1) 2-AP-49 "Loss of Normal Charging" (2) Isolate letdown due to 2-CH-FCV-2122 being closed Incorrect but plausible, first part is incorrect; there is not any transitions to 2-AP-49Wednesday, January 27, 2016 1:57:56 PM 5

QUESTIONS REPORTfor SRO Exam Jan Submittal"Loss of Normal Charging" from 0-AP-10. But the candidate may think normal chargingis lost due to the DC bus loss and assume that it is the correct procedure for this event.

Second part is incorrect but could be plausible if the candidate does not understandthat the signal generated to the E/P for 2-CH-FCV-2122 comes from the process plantcabinet 6, which has two separate power supplies. C. INCORRECT (1) 2-AP-49 "Loss of Normal Charging" (2) Place 2-CH-FCV-2122 in local control at Aux Shutdown Panel and control flowIncorrect but plausible, first part is incorrect; there are no transitions to 2-AP-49 "Lossof Normal Charging" from 0-AP-10. But the candidate may think normal charging is lostdue to the DC bus loss and assumes that it is the correct procedure for this event.

Second part is incorrect but could be plausible if the candidate knows that2-CH-FCV-2122 can be controlled from the Aux Shutdown Panel and 2-AP-49 doeshave a step for the operators to use 2-AP-20, Operation from the Auxiliary ShutdownPanel to shift 2-CH-FCV-2122 to local control in the ASP and control charging flow if2-CH-FCV-2122 demand in not indicated in the control room. D. CORRECT (1) 2-AP-49.1 "Loss of Normal and Excess Letdown" (2) Reduce Makeup flow and Maximize RCS sample flow Both the first part and second part are correct. 0-AP-10 "Loss of Power", attachment 19will direct the crew to initiate 2-AP-49.1 "Loss of Normal and Excess Letdown" If theoperating crew can not re-energize 125-volt DC bus 2-III and pressurizer level isgreater than 28%, 2-AP-49.1 will then direct the crew to minimize Reactor CoolantSystem (RCS) makeup flow and to maximize RCS sample flow, to Reduce the rate ofPressurizer level increase because of the Loss of Normal and Excess Letdown. K/A: 058AG2.1.20 Loss of DC Power Ability to interpret and execute procedure steps.Technical

References:

0-AP-10 "Loss of Electrical Power" 1-AP-49.1 "Loss of Normal and Excess Letdown" 1-AP-49 "Loss of Normal Charging" UFSAR for NAPS Unit 1 and 2 Section 8.3.2 Direct Current Power System12050-CH-001 "CVCS Charginig flow control to the Regen Hx 2-CH-E-3 References provided to applicants: NoneWednesday, January 27, 2016 1:57:56 PM 6

QUESTIONS REPORTfor SRO Exam Jan SubmittalLearning Objective:U 11902 Given the results of the electrical power system diagnostic (0-AP-10),sequence in order of priority the actions necessary to restore the Electrical Distribution System (Dealing with the effects of the loss of power) U 11549 Explain concepts associated with responding to the loss of 125-volt DC bus 1-III(why RCS makeup flow is minimized and RCS sample flow is maximized, if the DC buscan not be re-energized. with PZR level >28%)U 18050Recognize plant conditions that result in a transition to or from 1-AP-49.1 "Loss ofNormal and Excess Letdown"Explain the high level actions, major action categories, key mitigating strategies, andtheir basis. Question Source: Modified - NAPs Vision BANK (60591)Question History: Unit-2 in operating at 100% power when DC bus 2-III is lost.

• The crew is performing the steps of attachement 9 "Loss of DC Bus 2-III of 0-AP-10 "Loss of Elect Power"If the operating crew can not re-energize 125-volt DC bus 2-III and pressurizer level isgreater than 28%, the crew is directed to minimize Reactor Coolant System (RCS)makeup flow and to maximixe RCS sample flow.This action is taken in order to _______(a)________ due to ________(b)_________.A. (a) Increase the rate of RCS inventory turnover (b) Loss of blender flow path B. (a) Reduce the rate or Pressurizer level increase (b) Loss of normal and excess letdown (CORRECT)C. (a) Reduce the rate of Volume Control Tank level decrease (b) Loss of normal and excess letdownD. (a) Reduce the rate of RCS inventory turnover (b) Loss of blender flow path Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content:SRO only - 10 CFR-55.43 (b)(5)

Assessment of facility conditions and selection of appropriate procedures duringnormal, abnormal, and emergency situations 10 CFR-55.41 (a)(b)10 Administrative, normal, abnormal, and emergency operating procedures for the facility,Wednesday, January 27, 2016 1:57:56 PM 7

QUESTIONS REPORTfor SRO Exam Jan SubmittalComments:This question matches the K/A statement by requiring the SRO applicant to have theability to interpret and execute procedure steps for Loss of DC power. To arrive at thecorrect answer the SRO applicant must recognize the proper procedure and proceduresteps that would be required to mitigate the given DC bus conditions. Also theapplicant will be required to have the knowledge of various system lineups associatedwith the stated DC bus loss.

Wednesday, January 27, 2016 1:57:56 PM 8

QUESTIONS REPORTfor SRO Exam Jan Submittal

3. 080 - 077AA2.05 001/SRO/T1/G1/3.2/3.8/NEW//Unit-1 and Unit-2 are both operating at 100% power.Shortly after midnight, both units are experiencing mega-watt swings along with 500kvbus, voltage and frequency fluctuations. As per 0-AP-8, Response to Grid Instability, describe the operation of the voltageregulator and when offsite power should be declared inoperable. Place voltage regulators in MANUAL; grid voltage increases to 550 KVMaintain voltage regulators in AUTO; grid voltage decreases to 504 KVMaintain voltage regulators in AUTO; grid voltage increases to 550 KV Place voltage regulators in MANUAL; grid voltage decreases to 504 KV A.B.C.D.Wednesday, January 27, 2016 1:57:56 PM 9

QUESTIONS REPORTfor SRO Exam Jan SubmittalDistractor Analysis:This requires the examinee to have the ability to determine and interpret operationalstatus of offsite circuit as they apply to Generator Voltage and Electric Grid Disturbances. A) INCORRECT Place voltage regulators in MANUAL; grid voltage increases to 550 KV First part is incorrect but plausible if operator associates mega watt, voltage andfrequency swing as a requirement for placing Voltage Regulator in Manual. Note in0-AP-8 = Fluctuations in Mega-watts indicate grid disturbance, Voltage Regulator failures have little to no affect on mega-watt output. The Voltage Regulators should bekept in AUTO during grid disturbances and has the operator verify Voltage Regulator is in AUTO or place in AUTO if available. Second part is incorrect because high switchyard voltage is NOT an immediate operability concern. The concern is more of a long-term degradation issue for energized equipment, which is covered by a note in0-AP-8. This is plausible because grid voltage of 550 KV is used as a trigger point in0-AP-8 to contact System engineer as soon as possible to evaluate GDC-17requirements and Offsite power source operability. B) CORRECT Maintain voltage regulators in AUTO; grid voltage decreases to 504 KVFirst part is correct: Note in 0-AP-8, Fluctuations in Mega-watts indicate griddisturbance, Voltage Regulator failures have little to no affect on mega-watt output. The Voltage Regulators should be kept in AUTO during grid disturbances and has theoperator verify Voltage Regulator is in AUTO or place in AUTO if available. Should the candidate apply 1-AP-26 Failure of Main Generator Voltage Regulator base on voltageswings, then Mega-watts are verified to be stable before the voltage regulator can beplaced in Manual. Second part is correct: < 505 KV is the trigger point in 0-AP-8 and1-LOG-4 to declare offsite power inoperable and enter the applicable actions of T.S.3.8.1 C) INCORRECT Maintain voltage regulators in AUTO; grid voltage increases to 550 KVPlausible because first part is correct, second part is plausible (see above) D) INCORRECT Place voltage regulators in MANUAL; grid voltage decreases to 504 KVFirst part incorrect but plausible (see above), Second part is correct K/A:

077AA2.05 Generator Voltage and Electric Grid DisturbancesAbility to determine and interpret the following as they apply to l Generator Voltage andElectric Grid Disturbances:Operational status of offsite circuitWednesday, January 27, 2016 1:57:56 PM 10 QUESTIONS REPORTfor SRO Exam Jan SubmittalTechnical

References:

0-AP-8, Response to Grid Instability 1-OP-26.8, 500 KV Switchyard Voltage 1-AP-26 Failure of Main Generator Voltage Regulator High2-AP-26 Failure of Main Generator Voltage Regulator High T.S. 3.8.1 "AC Sources-Operating" T.S. 3.8.1 Bases 1-LOG-4 "Unit-1 Control Board (Modes 1-4)"

1-LOG-4A "Unit-1 Control Board (Modes 5 & 6)" 2-LOG-4 "Unit-2 Control Board (Modes 1-4)"2-LOG-4A "Unit-2 Control Board (Modes 5 & 6)" References provided to applicants: NoneLearning Objective:U 11991 Given a set of plant conditions, evaluate Main Generator Control and ProtectionSystem operations in light of the following issues:

• Effect of a failure, malfunction, or loss of a related system or component on a system.

• Effect of a failure, malfunction, or loss of components in a system or related systems

• Impact on Technical Specifications

• Response if limits or setpoints associated with a system or its components have been exceeded

• Proper operator response to the condition as statedU 18006Perform action of 0-AP-8, "Response to Grid Instabiltiy" Question Source: New Question History:Question Cognitive Level: Memory or Fundamental knowledge 10 CFR Part 55 Content:SRO only - 10 CFR-55.43 (b)(5)

Assessment of facility conditions and selection of appropriate procedures duringnormal, abnormal, and emergency situations. 10 CFR-55.41 (a)(b)(5)

Facility operating characteristics during steady state and transient conditions, includingWednesday, January 27, 2016 1:57:56 PM 11 QUESTIONS REPORTfor SRO Exam Jan Submittalcoolant chemistry, causes and effects of temperature, pressure and reactivity changes,effects of load changes, and operating linitations and reasons for these operatingcharacteristics. Comments:This question matches the K/A statement by requiring the SRO applicant to analysisand interpretation of operational status of offsite circuit as they apply to GeneratorVoltage and Electric Grid Disturbances. To arrive at the correct answer the SROapplicant must recognize if the voltage regulator is functioning properly and thecondition is due to a perturbation on the grid (MW swing), then the best course ofaction is to leave the voltage regulator in automatic and allow it to control voltage. Alsoto possess knowledge of the setpoint that requires declaring the offsite powerinoperable.

Wednesday, January 27, 2016 1:57:56 PM 12 081 - WE05EA2.2 001/SRO/T1/G1/3.7/4.3/BANK//Given the following conditions:Unit-2 was at 100% when a total loss of S/G Feedwater occurred.* "A" S/G wide-range level is 32%* "B" S/G wide-range level is 12%* "C" S/G wide-range level is 12%* Containment pressure is 11 psia

• Containment radiation peaked at 1.0E1 R/hr* The Crew has just entered 2-FR-H.1 "Response to Loss of Secondary Heat Sink* Secondary Heat Sink is required For the given plant conditions, which one of the following, list the proper order ofpriority for cooling restoration per 2-FR-H.1?1) AFW, 2) MFW, 3) RCS bleed and feed, 4) Condensate, 5) Fire Protection orService Water1) RCS bleed and feed, 2) MFW, 3) Condensate, 4) AFW, 5) Fire Protection orService Water 1) AFW, 2) MFW, 3) Condensate, 4) Fire Protection or Service Water, 5) RCSbleed and feed.

1) RCS bleed and feed, 2) AFW, 3) MFW, 4) Condensate, 5) Fire Protection orService WaterA.B.C.D.

Distractor Analysis:This requires the examinee to have the knowledge for adherence to appropriateprocedures and operation within the limitations in the facility license and amendmentsas they apply to the Loss of Secondary Heat Sink. A) INCORRECT 1) AFW, 2) MFW, 3) RCS bleed and feed, 4) Condensate, 5) Fire Protection or Service WaterThis could be plausible if the candidate is not familiar with the step that checks if RCSbleed and feed is required, the candidate may think it's appropriate to try and recoverflow to S/G using less intrusive sources AFW then MFW before the bleed and feed is started, Condensate and Fire Protection/Service water requires lowering S/G pressureso they would be last. B) INCORRECT 1) RCS bleed and feed, 2) MFW, 3) Condensate, 4) AFW, 5) Fire Protection or Service WaterThis could be plausible if the candidate knows that the RCS bleed and feed is required:(Any 2 S/G wide range level < 14% [24%]). but is not familiar with the correct sequencethat the procedure, has AFW, falling into the recovery plan. C) INCORRECT 1) AFW, 2) MFW, 3) Condensate, 4) Fire Protection or Service Water, 5) RCS bleed and feed. This could be plausible if the candidate is not familiar with the urgency to start a RCSbleed and feed based on loss of heat sink but is familiar with the order in which theoperator attempts to establish flow. (AFW flow, MFW flow, Condensate flow then low pressure water source (FP or SW) in that order) D) CORRECT 1) RCS bleed and feed, 2) AFW, 3) MFW, 4) Condensate, 5) Fire Protection or Service WaterBased on the units stated conditions the following can be established.1) RCS bleed and feed is required: (Any 2 S/G wide range level < 14%). FR-H.1 is torestore and/or maintain adequate secondary heat removal capability and to establishRCS bleed and feed heat removal if secondary heat removal capability cannot bemaintained. This is a continous action step within the procedure and the operators willtransition when the proper criteria is met. Then the operator attempts to restore or establish AFW flow, MFW flow, Condensate flow then low-pressure water source (FPor SW) in that order,K/A:WE05EA2.2Loss of Secondary Heat Sink Adherence to appropriate procedures and operation within the limitations in thefacility*s license and amendments.Technical

References:

2-FR-H.1, "Response to loss of secondary heat sink" Background information for Westinghouse owners group Emergency ResponseGuidelines (HP-Rev. 2) FR-H.1 Response to Loss of Secondary Heat Sink References provided to applicants: NoneLearning Objective:U 11254Given a set of plant conditions, evaluate criteria that require bleed and feed to beinitiated U 11304Why AFW is the preferred source of steam generator feedwater versus MFWQuestion Source: Bank (NAPs Vision Bank) Question History:Given the following conditions:Unit-1 was at 100% when a loss of all S/G feedwater occurredA S/G wide range level is 30%B and C S/G wide rage levels are 12%

Containment pressure is 13 psia Containment radiation peaked at 1.0E1 R/hr The crew has just entered 1-FR-H.1 "Response to Loss of Secondary Heat Sink, andhas completed Step 1, CHECK IF SECONDARY HEAT SINK IS REQUIRED (answerYes) For the Stated Plant conditions, which ONE of the following lists the proper order ofpriority of cooling restoration per 1-FR-H.1? A. 1) RCS bleed and feed; 2) AFW; 3) MFW; 4) Condensate 5) FP/SWB. 1) RCS bleed and feed; 2) MFW; 3) Condensate; 4) AFW; 5) FP/SW C. 1) AFW; 2) MFW; 3) Condensate; 4) RCS bleed and feed; 5) FP/SW D. 1) AFW; 2) MFW; 3) Condensate; 4) FP/SW; 5) RCS bleed and feed Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content:SRO only - 10 CFR-55.43 (b)(5)Assessment of facility conditions and selection of appropriate procedures duringnormal, abnormal, and emergency situations.10 CFR-55.45(a)(13)Demonstrate the applicant's ability to function within the control room team asappropriate to the assigned position, in such a way that the facility licensee's procedures are adhered to and that the limitatioons in its license and amendments arenot violated. Comments:This question matches the K/A statement by requiring the SRO applicant to have theability to determine and interpret plant conditions, and maintain adherence to appropriate procedures, and operation within the limitations in the facility license, andamendments, as they apply to the Loss of Secondary Heat Sink. To arrive at thecorrect answer, the SRO applicant must recognize the proper procedure sequence, byassessing plant conditions, and then, selecting a flow path used by 2-FR-H.1. This alsorequires knowledge of a decision point, that envolves transition, within 2-FR-H.1, tomitigate recovery.

QUESTIONS REPORTfor SRO Exam Jan Submittal

4. 082 - 001AA2.04 001/SRO/T1/G2/4.2/4.3/NEW//Unit 1 is being returned to 100% power following a scheduled refueling outage. Reactor power has been stabilized at 50% for shift turnover.

• RCS Pressure 2235 psia

• Tavg and Tref are matched* Control Rods are in Auto After assuming the duties as the Unit-1 Unit Supervisor, the selected First StagePressure channel fails High.Which one of the following answers list the proper procedure that addresses all thefailures of this event and power response to complete the statement?Direct the operator to take immediate actions as per _____(1)______ and initial reactorpower response will _____

(2)______. (1) 1-AP-1.1 "Uncontrolled Rod Motion" (2) Decrease (1) 1-AP-1.1 "Uncontrolled Rod Motion" (2) Increase (1) 1-AP-3 "Loss of Vital Instrumentation" (2) Increase (1) 1-AP-3 "Loss of Vital Instrumentation" (2) Decrease A.B.C.D.Distractor Analysis:This requires the examinee to have the ability to determine and interpret reactor powerand its trend as they apply to continuous rod withdraw.

A) INCORRECT (1) 1-AP-1.1 "Uncontrolled Rod Motion" (2) DecreaseFirst part is incorrect but plausible because when impulse (1st stage pressure) failshigh rate of change is sensed by the power mismatch unit and a large Delta is seenbetween Tavg and Tref. This large temperature error signal and power mismatch errorsignal are summed and rods will begin to withdraw at maximum speed. At this pointcandidate may select 1-AP-1.1 because it will address rod motion from a failure within

the rod control circuitry. But will not address other parameters controlled by first stagepressure instrumentation. The operators should use 1-AP-3 as the controllingprocedure for the loss of vital instrumentation, This will provide the correct response toplace rods in manual, control Steam Generator level, and address steam dumps. It alsostabilizes, defeats, and corrects the failed indication along with directing the proper T.Srequirements. Second part is incorrect but plausible if the candidate thinks rods will insert, attempting to drive first stage pressure down to match Tavg. Wednesday, January 27, 2016 1:57:56 PM 13 QUESTIONS REPORTfor SRO Exam Jan SubmittalB) INCORRECT (1) 1-AP-1.1 "Uncontrolled Rod Motion" (2) IncreaseFirst part is incorrect but plausible because when impulse (1st stage pressure) failshigh rate of change is sensed by the power mismatch unit and a large Delta is seenbetween Tavg and Tref. This large temperature error signal and power mismatch errorsignal are summed and rods will begin to withdraw at maximum speed. At this pointcandidate may select 1-AP-1.1 because it will address rod motion from a failure within

the rod control circuitry. But will not address other parameters controlled by first stagepressure instrumentation. The operators should use 1-AP-3 as the controlling procedure for the loss of vital instrumentation, This will provide the correct response toplace rods in manual, control Steam Generator level, and address steam dumps. It alsostabilizes, defeats, and corrects the failed indication along with directing the proper T.Srequirements. Second part is correct, when First stage pressure fails HIGH, a large rateof change in the power mismatch circuit is sensed, also a large difference is seenbetween Tavg and Tref indicating turbine power increasing greater than reactor powerand rods will begin to withdraw at maximum speed. and reactor power will increase C) CORRECT (1) 1-AP-3 "Loss of Vital Instrumentation" (2) Increase1-AP-3 "Loss of Vital Instrumentation is the correct procedure to stabilize, defeat, andcorrect the failed indication. Rods are placed in manual, steam generator levels aremaintained, steam dumps are transferred to steam pressure mode, the controllingchannel is then swapped and the appropriate MOP is entered for repairs. When First stage pressure fails HIGH, a large rate of change in the power mismatchcircuit is sensed, also a large difference is seen between Tavg and Tref indicatingturbine power increasing greater than reactor power and rods will begin to withdraw at maximum speed. and reactor power will increaseD) INCORRECT (1) 1-AP-3 "Loss of Vital Instrumentation" (2) Decrease First part is correct, 1-AP-3 is the correct procedure to cover all issues of concern. Second part is incorrect but plausible, if the candidate thinks rods will insert, attemptingto drive first stage pressure down to match Tavg.

K/A: 001AA2.04 Continuous Rod WithdrawalWednesday, January 27, 2016 1:57:57 PM 14 QUESTIONS REPORTfor SRO Exam Jan SubmittalWithdrawal :Reactor power and its trend Technical

References:

1-AP-3 "Loss of vital instrumentation1-AP-1.1 "Continuous uncontrolled rod motion"1-AR-B-A7 "Median/HI TAVG <> TREF Deviation"References provided to applicants: NoneLearning Objective:U 1739 Descripe the response of the Rod Control System to a failure of the controlling turbine impulse pressure channel

• Response if limits or setpoints associated with a system or its components have been exceeded

• Proper operator response to the condition as statedU 17998 Explain the purpose of 1-AP-3 "Loss of Vital Instrumentation" Explain the symptoms and entry conditions Explain applicable TS/TRM/Reportability Explain High Level actionsRecognize plant conditions that result in transition to 1-AP-3 U 17995Explain the purpose of 1-AP-1.1 "Continuous Uncontrolled Rod Motion" Explain the purpose Recognize plant conditions that result in transition to 1-AP-1.1Explain High Level actionsQuestion Source: New Question History:Question Cognitive Level: Comprehension or Analysis10 CFR Part 55 Content:82 - 10 CFR-55.43 (b)(5)

Assessment of facility conditions and selection of appropriate procedures duringnormal, abnormal, and emergency situations Comments:

This question matches the K/A statement by requiring the SRO applicant to have aability to determine and interpret reactor power and its trend as they apply tocontinuous rod withdraw. To arrive at the correct answer the SRO applicant must havethe systems knowledge to assess plant conditions by understanding the effect aparticular failure of the selected channel of 1st stage pressure will have on Rod ControlWednesday, January 27, 2016 1:57:57 PM 15 QUESTIONS REPORTfor SRO Exam Jan SubmittalSystem. Then the candidate must choose a procedure that has actions to stabilize andrecover the plant that address all associated failures.

Wednesday, January 27, 2016 1:57:57 PM

16 083 - 060AG2.4.30 001/NEW/1/2/2.7/4.1/NEW//NOBoth Units are at 100% power. Chemistry is in the process of sampling "A" WGDT(1-GW-TK-1A) when the following occurs:22:04 Unit 2 receives annunciator 2B-B5 PROCESS VENT VNT STACK A&B HI HI RADIATION.22:05 1-VG-RI-180-2, Vent Stack B, is reading 4E+5 uCi/sec 22:08 Chemistry reports diaphragm on 1-GW-TK-1A sample isolation valve is severely cracked and cannot be isolated.22:09 HP is notified to sample and survey release at 1-GW-TK-1A22:10 Mech notified to determine time required to repair/terminate leak. 22:15 1-VG-RI-180-2, Vent Stack B, peaks at 4.06E+6 uCi/sec 22:17 1-VG-RI-180-2, Vent Stack B, is reading 3E+6 uCi/sec and is decreasing22:21 1-VG-RI-180-2, Vent Stack B, is reading a steady 2E+5 uCi/sec22:23 HP takes two samples and confirms the release rate from 1-GW-TK-1A sample isolation valve is three times ODCM limit22:45 Mechanics report it will take them 30 minutes to isolate the leak. The highest required emergency classification is ___(1)___ and the NRC is required tobe notified within ___(2)___ of event declaration. REFERENCE PROVIDED(1) NOUE(2) 15 minutes(1) NOUE(2) 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />(1) ALERT(2) 15 minutes(1) ALERT(2) 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> A.B.C.D.Distractor Analysis:Abnormal plant evolution associated with an accidental gaseous radwaste releaseevent, Knowledge of events related to system operations/status that must be reportedto internal organizations or external agencies such as the State, the NRC, or thetransmission system operator. A) INCORRECT NOUE; 15 minutes First part is correct SM will declare NOUE, 19 min into the release HP notifiesoperations that sample results at 1-GW-TK-1A is 3 time ODCM limit, this meets the limits of RU1.4, the 60 min time frame is met because of Note 2 of the EAL's whichstates the SEM should not wait until the applicable time has elapsed, but shoulddeclare the event as soon as it is determined that the conditon will likely exceed theapplicable time. This occurs at 41 minutes into the event when Mech report back theywill need an additional 30 minutes to repair/terminate the leak. Second part is incorrectbut plausible because 15 minutes is the required time the SM has to notified the stateand local Emergency Operations Centers (EOC) of declaring the emergency class. B) CORRECT NOUE; 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> First part is correct, SM will declare NOUE, 19 min into the release HP notifiesoperations that sample results at 1-GW-TK-1A is 3 time ODCM limit, this meets thelimits of RU1.4, the 60 min time frame is met because of Note 2 of the EAL's whichstates the SEM should not wait until the applicable time has elapsed, but shoulddeclare the event as soon as it is determined that the conditon will likely exceed theapplicable time. This occurs at 41 minutes into the event when Mech report back theywill need an additional 30 minutes to repair/terminate the leak. Second part is correctand can be confirmed by the Note in EPIP-2.02 Notification of NRC "NRC notificationshall be made immediately after notification of State and local governments and in allcases within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> from the time of event declaration. C) INCORRECT ALERT; 15 minutes First part in incorrect but plausible, candidate could pick this because EAL RA1.2threshold set point for Vent stack B 1-VG-RI-180-2 is > 4.07E+6, and reading onmonitor at 22:15 peaks at 4.06E+6 and never is greater than and only maintained thispeak value for 2 min of the required

> 15 min., the value is well below the thresholdwhen mechanics inform the SEM the repair of the leak will take another 30 min. Ifcandidate does not notice that readings are below the set point then at 22:30 the Alertclassification would be legitimate. Second part is incorrect but plausible because 15minutes is the required time the SM has to notified the state and local EmergencyOperations Centers (EOC) of declaring the emergency class. D) INCORRECT ALERT; 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> First part in incorrect but plausible, candidate could pick this because EAL RA1.2threshold set point for Vent stack B 1-VG-RI-180-2 is > 4.07E+6, and reading onmonitor at 22:15 peaks at 4.06E+6 and never is greater than and only maintained thispeak value for 2 min of the required

> 15 min, the value is well below the thresholdwhen mechanics inform the SEM the repair of the leak will take another 30 min. Ifcandidate does not notice that readings are below the set point then at 22:30 the Alertclassification would be legitimate. Second part is correct and can be confirmed by theNote in EPIP-2.02 Notification of NRC "NRC notification shall be made immediatelyafter notification of State and local governments and in all cases within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> from the time of event declaration.

K/A:060AG2.4.30 Accidental Gaseous Radwaste ReleaseKnowledge of events related to system operation/status that must be reported tointernal organizations or external agencies, such as the State, the NRC, or thetransmission system operator.Technical

References:

EAL chart EAL bases for RU1.3EAL bases for RU1.4EAL bases for RA1.2 2-AR-B-B5 PROCESS VENT VNT STACK A&B HI HI RADIATION 0-AP-5.2 "MGP Radiations Monitoring System" EPIP-2.02 "Notification Of NRC" EPIP-1.02 "Response To Notification Of Unusual Event" EPIP-2.01 "Notification Of State And Local Governments" 0-AP-54 "Accidental, unplanned, or uncontrolled radioactive gaseous waste release. References provided to applicants: EAL chart Learning Objective:

U 14319 Evaluate a set of plant conditions associated with the emergency plan implementing procedures. Question Source: New Question History:Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content:SRO only - 10 CFR 55.43 (b)(5)Assessment of facility conditions and selection of appropriate procedures duringnormal, abnormal and emergency situations. 10 CFR 55.41.10(a)(b)(10)

Administrative, normal, abnormal, and emergency operating procedures for the facilityComments:

This question matches the K/A statement by requiring the SRO applicant to have theKnowledge of events related to system operations/status that must be reported tointernal organizations or external agencies such as the State, the NRC, or thetransmission system operator. Under an abnormal plant evolution associated with anaccidental gaseous radwaste release event. This question requires the candidate toassess various parameters and make decisions and then selecting a notification basedon the diagnostics. To arrive at the correct answer the SRO applicant must have the use of EAL basis, that provides guidance for EAL decision points. Along with theprocedure knowledge, required to make notification within the prescribed time frame.

QUESTIONS REPORTfor SRO Exam Jan Submittal

5. 084 - 061AA2.03 001/SRO/T1/G2/3.0/3.3/NEW//Unit-2 is defueled and fuel assembly insert shuffle is in progress with the Fuel BuildingRadiation Automatic Interlock key switch in enabled.

• Annunciator 1K-D2 RAD MONITOR SYSTEM HI RAD LEVEL, actuates.* Annunciator 1K-D4 RAD MONITOR SYSTEM HI-HI RAD LEVEL actuates.1-RMS-RM-153 "Fuel Pit Bridge Radiation Monitor" is noted to be pegged high andunresponsive to a source check or reset. 1-RMS-RM-152 "New Fuel Storage Radiation

Monitor" indication has not ch anged and Health Physics reports radiation levels in thefuel building have not changed. Which of the following actions will be required? Place the fuel building radiation automatic interlock switch in Disable AND Enter (CR) Condition Report; No other action required.

Leave the fuel building radiation automatic interlock switch in Enable AND Enter (CR) Condition Report: No other action required. Place the fuel building radiation automatic interlock switch in Disable AND Enter action of TR 3.3.7(c) Leave the fuel building radiation automatic interlock switch in Enable AND Enter action of TR 3.3.7(c)

A.B.C.D.Distractor Analysis:The candidate requires the ability to determine and interpret setpoints for alert and highalarms as they apply to the Area Radiation Monitoring (ARM) system alarms.CORRECT A. Place the fuel building radiation automatic interlock switch in Disable AND Enter (CR) Condition Report; No other action required.On a HI-HI alarm on 1-RM-RMS-152 or 153 and the key switch in ENABLE position thefollowing will occur: After a 2 minute time delay will automatically dump the MCRbottled air, close the MCR dampers and start the MCR emergency ventilation fans.Based on 1-RMS-RM-153 high indications along with 1-RMS-RM-152 indication beingnormal, and local reports of normal radiations levels in the fuel Building, the crew will bein 0-AP-5.1 "Common Radiation Monitors System" and would make the assessment,1-RMS-RM-153 has malfunctioned and the SRO will direct the key switch to be placedin DISABLE this is a critical decision because going to Disable is required to be donewithin 2 minutes of receiving the alarm. Submitting the CR will be the only other actionrequired because, both 1-RMS-RM-152 and 1-RMS-RM-153 are located in the fuelbuilding but TRM 3.3.7 "Radiation Monitoring Instrumentation" Table 3.3.7-1 list the required area monitor as "Fuel Storage Pool Area Criticality Monitor" and only if you goto the bases for the TRM does it specifically call out 1-RMS-RM-152 as the requiredmonitor. Wednesday, January 27, 2016 1:57:57 PM 17 QUESTIONS REPORTfor SRO Exam Jan SubmittalINCORRECT B. Leave the fuel building radiation automatic interlock switch in Enable AND Enter (CR) Condition Report: No other action required.Plausible selection for leaving the fuel building radiation automatic interlock switch inenable is because the candidate may assume that the only RM that feeds into the fuelbuilding radiation automatic interlock switch comes from 1-RMS-RM-152. 0-AP-5.1 listthe actions for 1-RMS-RM-152 first, second possibility is the fact that 1-RMS-RM-152 isassociated with TRM 3.3.7 "Radiation Monitoring Instrumentation" and TRM 3.3.9"Regulatory Guide (RG) 1.97 Instrumentation", 1-RM-RMS-153 is not listed in eitherTRM. Second part is correct, submitting the CR will be the only other action required because, both 1-RMS-RM-152 and 1-RMS-RM-153 are located in the fuel building butTRM 3.3.7 "Radiation Monitoring Instrumentation" Table 3.3.7-1 list the required area monitor as "Fuel Storage Pool Area Criticality Monitor" and only if you go to the basesfor the TRM does it specifically call out 1-RMS-RM-152 as the required monitor. INCORRECT C. Place the fuel building radiation automatic interlock switch in Disable AND Enter action of TR 3.3.7(c) Plausible selection because first part is correct, based on 1-RMS-RM-153 highindications along with 1-RMS-RM-152 indication being normal, and local reports ofnormal radiations levels in the fuel Building, the crew will be in 0-AP-5.1 "CommonRadiation Monitors System" and would make the assessment, 1-RMS-RM-153 hasmalfunctioned and the SRO will direct the key switch to be placed in DISABLE, this is a critical decision because going to Disable is required to be done within 2 minutes ofreceiving the alarm. On a HI-HI alarm 1-RM-RMS-152 or 153 and the key switch inENABLE position the following will occur: After a 2 minute time delay will automaticallydump the MCR bottled air, close the MCR dampers and start the MCR emergency ventilation fans. Second part is incorrect but plausible because, TRM 3.3.7 "Radiation Monitoring Instrumentation" Table 3.3.7-1 list the required area monitor as "Fuel Storage Pool Area Criticality Monitor" and only if you go to the bases for the TRM doesit specifically call out 1-RMS-RM-152 as the required monitor the candidate mayassume the TRM is referring to 1-RMS-RM-153 . INCORRECT D. Leave the fuel building radiation automatic interlock switch Interlock switch in Enable AND Enter action of TR 3.3.7(c). Plausible selection for leaving the fuel building radiation automatic interlock switch in enable is because the candidate may assume that the only RM that feeds into the fuelbuilding radiation automatic interlock switch comes from 1-RMS-RM-152. 0-AP-5.1 listthe actions for 1-RMS-RM-152 first, second possibility is the fact that 1-RMS-RM-152 isassociated with TRM 3.3.7 "Radiation Monitoring Instrumentation" and TRM 3.3.9"Regulatory Guide (RG) 1.97 Instrumentation", 1-RM-RMS-153 is not listed in eitherTRM. Second part is incorrect but plausible because, TRM 3.3.7 "Radiation MonitoringInstrumentation" Table 3.3.7-1 list the required area monitor as "Fuel Storage PoolWednesday, January 27, 2016 1:57:57 PM 18 QUESTIONS REPORTfor SRO Exam Jan SubmittalArea Criticality Monitor" and only if you go to the bases for the TRM does it specificallycall out 1-RMS-RM-152 as the required monitor the candidate may assume the TRM isreferring to 1-RMS-RM-153 . K/A: 061AA2.03 Area Radiation Monitoring (ARM) System Alarms Ability to determine and interpret the following as they apply to the Area RadiationMonitoring (ARM) System Alarms:Setpoints for alert and high alarmsTechnical

References:

0-OP-21.11 "Fuel Bldg ventilation lineup for evolutions over spent fuel pool0-AP-5.1 "Common unit radiation monitoring system" 1-AR-K-D4 (1K-D4) "RAD MONITOR SYST HI-HI RAD LEVEL"1-AR-K-D2 (1K-D2) "RAD MONITOR SYSTEM HI RAD LEVEL" TR 3.3.7 "Radiation Monitoring Instrumentation" and bases TR 3.3.9 Regulatory guide (RG) 1.97 instrumentation 11715-RM-035 (1-RMS-RM-152) 11715-RM-036 (1-RM-RMS-153) 0-LOG-6A, Backboards Logs References provided to applicants: TR 3.3.7Learning Objective:U 5241 List means provided to locally determine high radiation in a monitored area as it appliesto the area radiation monitors.U 5242Explain the response of all equipment which is directly affected by HIGH-HIGH alarmsassociated with the New fuel storage area and Fuel pit bridge area Westinghouse arearadiation monitorsU 17486Explain the concepts associated with the Radiation Monitoring Instrumentationtechnical requirement and bases (TR-3.3.7) U 18003Perform the following actions of 0-AP-5.1 "Common Unit Radiation Monitoring Sys"*Recognize the symptoms and entry conditionsWednesday, January 27, 2016 1:57:57 PM 19 QUESTIONS REPORTfor SRO Exam Jan Submittal*Explain high level actions, key mitigating strategies, and their basis. Question Source: NEWQuestion History: Question Cognitive Level: 10 CFR Part 55 Content:SRO only 10 CFR-55.43 (b)(4)

Radiation hazards that may arise during normal and abnormal si tuations, includingmaintenance activities and various contamination conditions. 10 CFR-55.45 (a)(13)

Demonstrate the applicants ability to function within the control room team as appropriate to the assigned position, in such a way that the facility licensee'sprocedures are adhered to and that the limitations in its license and amendments arenot violated. Comments:This question matches the K/A statement by requiring the SRO applicant to have aability to determine and interpret setpoints for alert and high alarms as they apply to theArea Radiation Monitoring (ARM) system alarms, in particular 1-RMS-152 "New Fuel Storage Area RM and 1-RM-RMS-153 "Fuel Pit Bridge RM" have an understanding ofthe bases behind the limits set forth in technical requirement manual. To arrive at thecorrect answer the SRO applicant must recognize normal and abnormal situation basedon indication and field personnel feedback. Have a systems knowledge of applicableactions performed by the RMs. Have a knowledge of steps within abnormal procedurethat correspond to this particular failure and then to apply required TRM actions.

Wednesday, January 27, 2016 1:57:57 PM 20 QUESTIONS REPORTfor SRO Exam Jan Submittal

6. 085 - WE14EG2.4.21 001/SRO/T1/G2/4.0/4.6/MODIFIED//Unit-1 experienced a reactor trip and safety injection from 100% power. The followingconditions exist:* The operating crew just entered 1-ECA-1.1 "Loss of Emergency Coolant Recirculation"* No Containment sump blockage exists

• Containment pressure is 61 psia

• Two recirculation spray pumps are available (Not Running)* Recirculation spray sump level is 3 feet 0 inches

• Two quench spray pumps are runningWhich one of the following descr ibes the proper sequence? Remain in 1-ECA-1.1 "Loss of Emergency Coolant Recirculation" FRs not applicable during 1-ECA-1.1. Transition to 1-FR-Z.1 "Response to High Containment Pressure" Do not start any recirculation spray pumps. Remain in 1-ECA-1.1 "Loss of Emergency Coolant Recirculation" FRs should not be initiated before completion of step 9. Transition to 1-FR-Z.1 "Response to High Containment Pressure" Start available recirculation spray pumps. A.B.C.D.Distractor Analysis:Based on Loss of Containment Integrity it requires the candidate to have knowledge ofthe parameters and logic used to assess the status of safety functions, such asreactivity control, core cooling and heat removal, reactor coolant system integrity,containment conditions, radioactivity release control, etc.. A. INCORRECT Remain in 1-ECA-1.1 "Loss of Emergency Coolant Recirculation" FRs not applicable during 1-ECA-1.1.Incorrect but plausible because of the beginning note in 1-ECA-1.1 IF ContainmentSump Blockage has occurred, THEN FRs should not be implemented until directed inthis procedure. If the candidate is not sure of the details, mitigative strategy or hasknowledge of this step then this answer may be chosen.

B. CORRECT Transition to 1-FR-Z.1 "Response to High Containment Pressure" Do not start any recirculation spray pumps. Correct, under the questions given conditions Containment pressure is in the required red path transition path for 1-FR-Z.1 (Containment Pressure

> 60 psia andWednesday, January 27, 2016 1:57:57 PM 21 QUESTIONS REPORTfor SRO Exam Jan SubmittalContainment sump blockage does not exists). However there is a Caution in 1-FR-Z.1"If 1-ECA-1.1 Loss of Emergency coolant recirculation, is in effect then to preserveRWST inventory, Step 2 through Step 5 of this procedure should not be performed.This caution warns the operator that the operation of the containment spray pumpsindicated in guideline 1-ECA-1.1 takes precedence. Step 2 through Step 5 of 1-FR-Z.1 Checks if CDA is required, verifies proper operation of Containment quench spraysystems, verifies proper operation of the SW system and verifies proper operation of the Containment Recirc spray system. These steps will not be performed based on the Caution. C. INCORRECT Remain in 1-ECA-1.1 "Loss of Emergency Coolant Recirculation" FRs should not be initiated before completion of step 9Incorrect but plausible because of the beginning note in 1-ECA-1.1 IF ContainmentSump Blockage has occurred, THEN FRs should not be implemented until directed inthis procedure. If the candidate is not sure of the details or has knowledge of this stepthen this answer may be chosen. The statement FRs should not be initiated beforecompletion of step 9 is a requirement listed by a note in 1-ECA-0.1 and is also repeatedclosely in 1-ECA-0.2 D. INCORRECT Transition to 1-FR-Z.1 "Response to High Containment Pressure" Start available recirculation spray pumps.Incorrect but plausible because, under the questions given conditions, Containmentpressure is in the required red path transition path for 1-FR-Z.1 (Containment Pressure

> 60 psia and Containment Containment sump blockage does not exists). If thecandidate does not possess the knowledge of the procedure content in the Caution atthe beginning of the procedure "If 1-ECA-1.1 Loss of Emergency coolant recirculation,is in effect then to preserve RWST inventory, Step 2 through Step 5 of this procedureshould not be performed", and the candidate is only familiar with the overall mitigative strategy, then this answer could be chosen.

K/A:

WE14EG2.4.21 High Containment Pressure Knowledge of the parameters and logic used to assess the status of safety functions,such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.Technical

References:

1-ECA-1.1 "Loss of Emergency Coolant Recirculation"Westinghouse Owners Group background information ECA-1.11-FR-Z.1 "Response to High Containment Pressure"Wednesday, January 27, 2016 1:57:57 PM 22 QUESTIONS REPORTfor SRO Exam Jan Submittal1-F-0 "Critical Safety Function Status Tree" References provided to applicants: NoneLearning Objective:U 13008 Explain why if 1-ECA-1.1 is in effect, the steps which place quench spray in serviceshould not be performed in 1-FR-Z.1U 5829Explain how ECA-1.1 is used when containment sump blockage occurs duringrecirculation mode, including use of FRs Question Source: Callaway 2007 SRO question 85 Question History: ModifiedGiven the following conditions:

• A LOCA has occurred.*Due to several component failures, the crew was required to perform ECA-1.1 "Loss of emergency coolant recirculation.

• The crew is now entering FR-Z.1 "Response to high containment pressure*Containment pressure is 61 psig and stable*Both Containment spray pumps are off

• RWST level is 8%Which (1) ONE of the following describes the strategy for reducing containment pressure?A. Start both containment spray pumps in accordance with FR-Z.1 Red path takeprecedence over ECA actionsB. Operate Containment spray pumps in accordance with the guidance in ECA-1.1 asdirected by FR-Z.1. Continue in FR-Z.1 until exit criteria is met. (CORRECT) C. Perform ONLY the FR-Z.1 actions that do NOT conflict with or undo the action takenin ECA-1.1, Two containment Coolers will provide adequate depressurization to meet the containment safety function requirements D. Do NOT perform actions of FR-Z.1 until the RWST LO LO Level alarm is clear andContainment Spray Pumps may be restarted. Ensure all other automatic actionsrelated to containment isolation have occurred as required. Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content:SRO only 10 CFR-55.43 (b)(5)Assessment of facility conditions and selection of appropriate procedures duringnormal, abnormal, and emergency situations.10 CFR-55.41 (a)(b)(7)Wednesday, January 27, 2016 1:57:57 PM 23 QUESTIONS REPORTfor SRO Exam Jan SubmittalDesign, components, and functions of control and safety systems includinginstrumentation, signals, interlocks, failure modes, and automatic and manual featuresComments: Question requires the SRO candidate to possess knowledge of diagnostic steps and decision points in the emergency operating procedures (EOP) that involve transitions toevent specific sub-procedures or emergency contingency procedures. Needs to knowcontent of the procedure verses knowledge of the procedures overall stategy orpurpose. Wednesday, January 27, 2016 1:57:57 PM 24 086 - 012A2.05 001/SRO/T2/G1/2/3.2/MODIFIED//Given the following:* Unit-1 startup has commenced following welding repairs to S/G blowdown line in containment.* Reactor Power is at 4% of rated thermal power

• Tavg is 550 degrees being maintained by steam dumps* Pressurizer Pressure Control is in Automatic and controlling pressure at 2235 psig

• 1-RC-PT-1456 Pressure Protection Channel II fails Hi

• The crew is in 1-AP-3 and is directed to completed 1-MOP-55.73 Pressurizer Pressure Protection Instrumentation.* 1-MOP-55.73 directs the Unit Supervisor to refer to TS 3.3.1 "Rx trip instrumentation and TS 3.3.2 "Engineered Safety Feature Actuation System instrumentation"Which instrument FUNCTION requires entry into a LCO Condition AND what is thePROTECTION provided by this function? Function ProtectionOverpower Delta-T Reactor Trip Ensures the design limit DNBR is met Overtemperature Delta-T Reactor Trip Ensures the integrity of the fuel Overtemperature Delta -T Reactor Trip Ensures the design limit DNBR is met Overpower Delta-T Reactor Trip Ensures the integrity of the fuel A.B.C.D.

Distractor Analysis:This requires the examinee to have the ability to (a) predict the impacts of faulty orerratic operation of detectors and function generators on the RPS, and (b) based onthose predictions, use procedures to correct, control, or mitigate the consequences ofthose malfunctions or operations:A) INCORRECT Function ProtectionOverpower Delta-T Reactor Trip Ensures the design limit DNBR is met Function is incorrect but plausible because, TS 3.3.1 Overpower Delta-T is applicablein Modes 1,2 but it does not use Pressurizer Pressure in its calculation like Overtemperature Delta-T does, so it does not require LCO entry Protection is alsoincorrect but plausible because it would be correct for Overtemperature Delta-T reactortrip B) INCORRECT Function Protection Overtemperature Delta-T Reactor Trip Ensures the integrity of the fuel Function is correct TS 3.3.1 Overtemperature Delta-T setpoints are impacted by RCSpressure and is applicable in Mode 1,2 to place the channel in trip within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.Protection is incorrect but plausible because it would be correct for Overpower Delta-Treactor trip C) CORRECT Function ProtectionOvertemperature Delta -T Reactor Trip Ensures the design limit DNBR is met TS 3.3.1 Overtemperature Delta-T setpoints are impacted by RCS pressure and isapplicable in Mode 1,2 to place the channel in trip within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and ensures protection from violating the DNBR limit is met. D) INCORRECT Function Protection Overpower Delta-T Reactor Trip Ensures the integrity of the fuel Function is incorrect but plausible because, TS 3.3.1 Overpower Delta-T is applicablein Modes 1,2 but it does not use Pressurizer Pressure in its calculation likeOvertemperature Delta-T does, so it does not require LCO entry. Also OverpowerDelta-T limits the required range of Overtemperature Delta-T. Protection is correct,Overpower Delta-T ensures the integrity of the fuel, no fuel pellet melting and < 1%cladding strain. K/A:012A2.05Reactor Protection System (RPS)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 theconsequences of those malfunctions or operations:Faulty or erratic operation of detectors and function generators Technical

References:

1-AP-3 "Loss of Vital Instrumentation"1-MOP-55.73 "Pressurizer Pressure Protection Instrumentation"TS 3.3.1 & Bases "Rx Trip System Instrumentation" TS 3.3.2 & Bases "Engineered Safety Feature Actuation System Instrumentation"References provided to applicants: None Learning Objective:U 9689 Explain the concepts associated with the low pressurizer pressure function of the reactor trip system instrumentation TS and bases U 9690Explain the concepts associated with the high pressure function of the reactor tripsystem instrumentation TS and bases U 16192Explain the concepts associated with the low-low pressurizer safety injection function ofthe reactor trip system instrumentation TS and basesU 16221Explain the concepts associated with the low power reactor trips block P-11 Question Source:U 9687Explain the concepts associated with the overtemperature delta-T function of thereactor trip system instrumentation TS and bases U 9688Explain the concepts associated with the overpower delta-T function of the reactor tripsystem instrumentation TS and basesQuestion Source: Modification of 2011 Kewanee SRO ILT Exam question. (K/A 012A2.05) Given the following: Unit restart has commenced 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> after a spurious reactor trip Rx Pwr is at 4% of rated thermal power Tave is 547 degrees maintained by the steam dumps Pressurizer Pressure Control is in auto controlling at 2235 psig PT-429, Red Channel Pressurizer Pressure, has FAILED HIGH The crew has implemented AOP-MISC-001, "Response to Instrument Failure" AOP-MISC-001 directs the unit supervisor to refer to TS 3.3.1 and TS 3.3.2 LCO 3.3.1 Ther RPS instrumentation for each function in table 3.3.1-1 shall be operable LCO 3.3.2 The ESFAS instrumentation for each function in table 3.3.2-1 shall be operableWhich instrumentation function requires ENTRY into a LCO CONDITION AND what is the protection provided by this function? FUNCTION PROTECTIONA. Low pressure safety injection function Prevents exceeding RCS pressure safety limit B. Overtemperature Delta-T reactor trip function Prevents exceeding DNBR limits C. Pressurizer Low Pressure reactor trip function Prevents exceeding subcooling limits during Mode 1 operation D. Pressurizer High Pressure reactor trip function Prevents exceeding the peak centerline fue temperature limit specification in TSQuestion History: Kewaunee 2011Question Cognitive Level: Memory and Fundamental Knowledge 10 CFR Part 55 Content:SRO only - 10 CFR-55.43 (b)(5)Assessment fo facility conditions and selection of appropriate procedures duringnormal, abnormal, and emergency situations 10 CFR-55.41 (a)(b)(5)Facility operating characteristics during steady state and transient conditions, includingcoolant chemistry, causes and effects of temperature, pressure and ractivity changes,effects of load changes, and operating linitations and reasons for these operating characteristics. Comments:This question matches the K/A statement by requiring the SRO applicant to analysisand predict the impacts of faulty or erratic operation of Pressure Protection channelfailure on the RPS. To arrive at the correct answer the SRO applicant must recognizewhich affected protection requires entry into a LCO based on the questions givenconditions and have knowledge of the Technical Specification basis for each functions protection provided.

QUESTIONS REPORTfor SRO Exam Jan Submittal

7. 087 - 013A2.04 001/SRO/T2/G1/3.6/4.2/BANK//The actions of 0-AP-41 "Severe Weather Conditions" are being performed due to athunderstorm warning, when lightning strikes the SWYD and a loss of offsite poweroccurs.

• 1-ECA-0.0 "Loss of all AC power" was entered.

• Operations have completed all actions up to determining the correct recoverymethod and procedure.* One AC emergency bus has been energized.* RCS subcooling is 35 DEGF.

• Pressurizer level is 5%.

• Adverse Containment Conditions do not exist.

• Safeguards equipment is not operating. Which answer will complete the following statement? SRO will direct procedure transition to __________ . 1-ES-0.0 "Re-Diagnosis", to determine or confirm the most appropriatepost-accident recovery procedure.1-ECA-0.1 "Loss of all AC power without SI required", recover the plant based onstable RCS conditions following a lose of all AC power. 1-E-0 "Reactor trip or safety injection", verify proper response of Reactor Protectionand safety injection following restoration of at least one emergency bus.1-ECA-0.2 "Loss of all AC power recovery with SI required" to manually loadsafeguards equipment as appropriate.

A.B.C.D.Distractor Analysis:

This requires the examinee to have the ability to (a) predict the impacts of the loss ofInstrument bus on the Engineered Safety Features Actuation System (ESFAS); and (b)based on those predictions. use procedures to correct, control, or mitigate theconsequences of those malfunction or operations. A. INCORRECT 1-ES-0.0 "Re-Diagnosis", to determine or confirm the most appropriate post-accident recovery procedure.Incorrect but plausible because the candidate may decide, since one emergency bushas been recovered, the crew no longer needs to be in the 1-ECA-0 series procedures,and with a low pressurizer level, 1-E-1 should be your recovery procedure and thequickest way is through 1-ES-0.0 "Re-Diagnosis" or 1-ES-0.0 could be used todetermine/confirm the most appropriate post-accident recovery procedure the crew should be in. 1-ES-0.0 should only be used when SI is in service or is required and1-E-0 Reactor Trip or Safety injection has been exited. Based on the given conditionentry into the EOP was directly through 1-ECA-0.0 so 1-ES-0.0 can not be used. Wednesday, January 27, 2016 1:57:57 PM 25 QUESTIONS REPORTfor SRO Exam Jan SubmittalB. INCORRECT 1-ECA-0.1 "Loss of all AC power without SI required", recover the plant based on stable RCS conditions following a lose of all AC power. Incorrect but plausible if the candidate reflects only on Subcooling and the fact that oneemergency bus has been recovered and safeguards equipment is not operating, butfollowing the restoration of AC power, selection depends on the existence of RCSsubcooling 25DEGF [75], existence of pressurizer level > 21% [26], and verification thatSI equipment has not automatically actuated upon power restoration. If RCS conditions have not deteriorated significantly (i.e. all criteria satisfied), the operator isdirected to guidline ECA-0.1 to recover the plant using normal operational systems. C. INCORRECT 1-E-0 "Reactor trip or safety injection", verify proper response of Reactor Protection and emergency core cooling systems following restoration of at least one emergency bus.Incorrect but plausible because the candidate may think that since one emergency bushas been recovered, the crew no longer needs to be in the 1-ECA-0 series proceduresand there is a need to verify that the Rx is tripped and initiate SI if required. D. CORRECT 1-ECA-0.2 "Loss of all AC power recovery with SI required" to manually load safeguards equipment as appropriate.Correct, 1-ECA-0.0 provides recovery guidelines based on existing RCS conditions. The criteria for recovery selection includes (1) existence of RCS subcooling (>25DEGF), (2) existence of Pressurizer level (>21%) and (3) verification that SI equipmenthas not automatically actuated upon ac power restoration, which would occur if powerwas restored after SI signal actuation but before SI signal was reset. If any of thecriteria is not satisfied the operator is directed to 1-ECA-0.2. This procedure will startsafeguards equipment as appropriate and then direct the operator to guideline 1-E-1"Loss of Reactor or Secondary Coolant, for subsequent recovery actions.

K/A: 013A2.04 Engineered Safety Features Actuation System (ESFAS)Ability to (a) predict the impacts of the following malfunctions or operations on theESFAS; and (b) based Ability on those predictions, use procedures to correct, control,or mitigate the consequences of those malfunctions or operations;Loss of instrument busTechnical

References:

1-ECA-0.0 "Loss of all AC power"1-ECA-0.0 Westinghouse owners group background document.1-ECA-0.1 "Loss of all AC power recovery without SI required".

1-ECA-0.1 Westinghouse owners group background document.

1-ECA-0.2 "Loss of all AC power recovery with SI required".

1-ECA-0.2 Westinghouse owners group background document. Wednesday, January 27, 2016 1:57:57 PM 26 QUESTIONS REPORTfor SRO Exam Jan SubmittalReferences provided to applicants: NoneLearning Objective:U 13830 List conditions that result in leaving 1-ECA-0.0 "Loss of all ac power" U 9630 Explain why SI is reset after depressurization of S/G'sU 13837 List concepts of 1-ECA-0.1 "Loss of all ac power without SI required"U 13838 List concepts of 1-ECA-0.2 "Loss of all ac power with SI required" Question Source: NAPS Vision Bank (7999)Question History:

Assume the following plant conditions:

• ECA-0.0 is in effect due to a loss of all AC power *Operators have completed all actions up to determining the correct recovery method and procedure

• One AC emergency bus is energized *The RCS is 44DEGF subcooled *PZR level is 50%

• SFGD's equipment is not operatingThe operator should go to procedure _______________. A. ECA-0.1, Loss of all ac power recovery W/O SI required, because all conditionsrequired are met B. ES-0.0, Re-Diagnosis, to determine or confirm the most appropriate post-accident recovery procedure C. ECA-0.2, Loss of all ac power recovery with SI required, to use engineered safeguards equipment to recover plant conditions following restoration of ACemergency power to at least one bus D. E-0, Rx trip or SI, to verify proper response of the reactor protection and emergencycore cooling systems following restoration of AC emergency power to at least on bus. Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content:SRO only - 10 CFR-55.43 (b)(5)

Assessment of facility conditions and selection of appropriate procedures duringnormal, abnormal, and emergency situations. 10 CFR-55.41 (a)(b)(5)

Facility operating characteristics during steady state and transient conditions, includingcoolant chemistry, causes and effects of temperature, pressure and reactivity changes,Wednesday, January 27, 2016 1:57:57 PM 27 QUESTIONS REPORTfor SRO Exam Jan Submittaleffects of load changes, and operating limitations and reasons for these characteristicsComments:This question matches the K/A statement by requiring the SRO applicant to have theability to (a) predict the impacts of the loss of Instrument bus on the Engineered SafetyFeatures Actuation System (ESFAS); and (b) based on those predictions, useprocedures to correct, control, or mitigate the consequences of those malfunction or operations. To arrive at the correct answer the SRO applicant must demonstrate theability to make the proper diagnostics bases on given conditions, and then use theinformation to transition to an event specific procedure that address the plantconfiguration based on ESFAS systems and electrical availability. Wednesday, January 27, 2016 1:57:57 PM 28 88 - 039A2.03 001/NEW/2/1/3.4/3.7/MODIFIED//NOUnit-1 was operating at 100% power when the following sequence of events occurred:

• Annuciator 1K-G6, N-16 RAD DET, alarms.

• N-16 indication on "B" S/G AND Main Steam Header are in alarm and trending up.* 1-RM-SV-121, Condenser Air Ejector Radiation Monitor indication has not changed.* PRZR level is rapidly decreasing.

• Controller demand on 1-FW-FCV-1488 "B MFRV" shows a slight decrease.

• The crew trips the Reactor and initiates SI. Current Status* The immediate actions of 1-E-0, Reactor Trip or Safety Injection have just been completed,

• All S/G are below Narrow Range level indication

• The BOP recommends isolating AFW flow to "B" S/GWhich ONE of the choices below states the actions that the SRO will direct the crew toperform? Isolate AFW to "B" S/G; Contact chemistry and HP for confirmation of the affected S/G prior to initiating1-E-0, Attachment 8 "Ruptured S/G isolation. Isolate AFW to "B" S/G; Initiate 1-E-0, attachment 8, Ruptured S/G isolationDo not isolate AFW to "B" S/G; Contact chemistry and HP for confirmation of the affected S/G prior to initiating1-E-0, Attachment 8 "Ruptured S/G isolation

Do not isolate AFW to "B" S/G; Initiate 1-E-0, Attachment 8, "Ruptured S/G isolation.

A.B.C.D.Distractor Analysis:

Ability to (a) predict the impacts of indications and alarms for main steam and arearadiation monitors (during SGTR) on the MRSS and (b) based on predictions, useprocedures to correct, control, or mitigate the consequences of those malfunctions oroperations. A INCORRECT Isolate AFW to "B" S/G; Contact chemistry and HP for confirmation of the affected S/G prior to initiating 1-E-0, Attachment 8 "Ruptured S/G isolation. First part is plausible because reducing AFW flow could confirm to the crew that indeed"B" S/G has a tube rupture. The Westinghouse background document mentions "Sinceprimary-to-secondary leakage adds additional inventory which accumulates in theruptured steam generator, level should return to the narrow range in that steamgenerator(s) significantly earlier and will continue to increase more rapidly. This response provides confirmation of a steam generator tube ru pture event and alsoidentifies the affected steam generator(s)". Second part is also plausible because1-RM-SV-121 should be showing something and it is not (malfunction); the candidatewho does not feel N-16 indication is enough to go on may delay action that areimportant to limiting offsite dose (attachment 8 isolates steam supply to AFW terryturbine, which is a local action). Also the continuous action page for 1-E-0 (step 7)states Attachement 8 MAY be used, it does not directly instruct the operator to use it. B INCORRECT Isolate AFW to "B" S/G; Initiate 1-E-0, attachment 8, Ruptured S/G isolation. First part is plausible because reducing AFW flow could provide the crew with more time before over filling the S/G, but knowledge of Westinghouse background guidelinesprovides the importance of maintaining the water level above the top of the U-tubes. 1)Provides Insulation between the ruptured steam generators steam space and the U-tubes. 2) Ensures a heat sink is available if no other intact steam generator isavailable. 3) Prevents misdiagnosis of the ruptured steam generator due to imbalancesin feedwater flow. Second part is correc t, attachment 8 should be implemented foridentification (which is confirmed by both "B" S/G and Main Steam header N-16Radiation monitor indication along with a slight decrease of demand on1-FW-FCV-1488) and isolation which is important to limiting offsite dose. C INCORRECT

Do not isolate AFW to "B" S/G; Contact chemistry and HP for confirmation of the affected S/G prior to initiating 1-E-0, Attachment 8 "Ruptured S/G isolation.First part is correct need U-tubes covered before isolation. Second part is incorrect butplausible because 1-RM-SV-121should be showing something and it is not(malfunction); the candidate who does not feel N-16 indication is enough to go on maydelay action that are important to limiting offsite dose (attachment 8 isolates steamsupply to AFW terry turbine, which is a local action. D CORRECT Do not isolate AFW to "B" S/G; Initiate 1-E-0, Attachment 8, "Ruptured S/G isolation".First part is correct the Westinghouse background documents Ba sis describes therequirement to maintain the S/G U-tubes covered before isolation 1) ProvidesInsulation between the ruptured steam generators steam space and the U-tubes. 2)Ensures a heat sink is available if no other intact steam generator is available. 3)Prevents misdiagnosis of the ruptured steam generator due to imbalances in feedwaterflow. and second part is correct because attachment 8 should be implemented foridentification (which is confirmed by both "B" S/G and Main Steam header N-16Radiation monitor indication along with a slight decrease of demand on1-FW-FCV-1488) and isolation which is important to limiting offsite dose.

039A2.03 Main and Reheat Steam System (MRSS)Ability to (a) predict the impacts of the following malfunctions or operations on theMRSS; and (b) based on predictions, use procedures to correct, control, ormitigate the consequences of those malfunctions or operations:Indications and alarms for main steam and area radiation monitors (during SGTR)Technical

References:

1-AP-24 "S/G tube leak"1-AP-5 "Unit-1 Radiation Monitoring System 1-E-0 and Westinghouse Owners Group Background document 1-E-3 and Westinghouse Owners Group Background document 1-EI-CB-21K Annunciator G6 TR 3.4.4 Primary to Secondary Leakage TR 3.4.4 Primary to Secondary Leakage Bases References provided to applicants: NoneLearning Objective:U 18027 Perform actions of 1-AP-24 "Steam Generator Tube leak"U 18002 Perform actions of 1-AP-5 "Unit-1 Radiation Monitoring System"U 12174Given a set of plant conditions, determine the appropriate recovery procedureU 12677Explain the concepts associated with identifying and isolating a ruptured steamgenerator in accordance with 1-E-3Question Source: Modified Bank (2012 North Anna SRO written Exam, Question #82)

Question History:Unit 1 was operating at 100% power when the following sequence of events occurred

• Annunciator K-G6, N-16 RAD DET, alarms

• RO reports N-16 modules for "C" SG AND Main Steam Header are in alarm and trending up

• BOP reports 1-RM-SV-121, Condenser Air Ejector Radiation Monitor indication has not changed

• RO reports PRZR level rapidly decreasing* The crew trips the reactor and initiates SICurrent Status

• The immediate actions of 1-E-0 Reactor Trip of Safety Injection, have just been completed.* All SG's are below the Narrow Range.* The BOP recommends isolating AFW flow to the "C" SG.

Based on the above, which ONE of the following identifies how the US shouldproceed?A. allow the BOP to isolate AFW to "C" SG; contact HP & Chemistry for confirmation ofthe affected SG prior to initiating 1-E-0, Attachment 8, Ruptured SG Isolation.B. allow the BOP to isolate AFW to "C" SG; have the RO initiate 1-E-0, attachment 8,Ruptured SG Isolation.

C. DO NOT allow the BOP to isolate AFW to "C" SG; contact HP & Chemistry forconfirmation of the affected SG prior to initiating 1-E-0, Attachment 8, Ruptured SGIsolation.

D. DO NOT allow the BOP to isolate AFW to "C" SG; have the RO initiate 1-E-0,attachment 8, Ruptured SG Isolation.Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content:SRO only - 10 CFR-55.43 (b)(5)

Assessment of facility conditions and selection of appropriate procedures duringnormal, abnormal, and emergency situations. 10 CFR-55.41 (a)(b)(5)

Facility operating characteristics during steady state and transient conditions, includingcoolant chemistry, causes and effects of temperature, pressure, and reactivity changes,effects of load changes, and operating limitations and reasons for these operating characteristics.Comments:This question matches the K/A statement by requiring the SRO applicant to possessthe ability to predict the impacts of indications and alarms for main steam and arearadiation monitors (during SGTR). To arrive at the correct answer the SRO must be aware of the radiation monitor flowpath within the Main Steam system. The SRO mustalso show knowledge of diagnostic steps, and decision points, based on indications thecandidate is given by plant conditions, and then use this information in deciding whenand how to implement attachments of the Emergency Operating Procedures.

089 - 076G2.1.25 001/SRO/T2/G1/3.9/4.7/MODIFIED//Unit-2 has been at 100% power for 90 days following a 26 day refueling outageUnit-1 is currently in Day 2 of a scheduled refueling outage, with the followingconditions:

• RHR discharge Temp = 115 DEGF

• PZR level is at 28% and on VCT float

• RCS loop stop valves are all open At 01:30 due to an Electrical fault a fire starts in the SW pump house. The followingconditions exist;* The crew is performing actions of 0-FCA-9 "Service Water Pump House Fire"

• All SW pumps have been secured.* Component Cooling temperatures are rapidly increasing.

• Annunciator SFP Hi/Hi-Hi Temp has illuminated.

• Spent fuel pool level is +1 inches.

• 01:55 Fire Brigade Scene Leader reports that the fire has been extinguished at the SWPH. Per Attachment 7 of 0-AP-27 "Malfunction of the Spent Fuel Pit System" the expectedheat up rate for the spent fuel pit under these conditions is ____(1)_____ DEGF perhourAnd Based on conditions and not on Shift Manager judgment what is the highest EALclassification that will be entered ___________ ? REFERENCE PROVIDED (1) 3.20(2) Alert(1) 3.99(2) NOUE(1) 3.20(2) NOUE(1) 3.99(2) Alert A.B.C.D.

Distractor Analysis:The candidate is required to have the ability to interpret reference materials, such asgraphs, curves, tables, etc.. that would be associated with the Service Water system. A. CORRECT (1) 3.20 (2) AlertFirst part is correct 3.20 heat up rate is correct using the table Conditions for Back toBack Refuelings, which is correct for less than 120 days between refuelings. Secondpart is also correct the highest EAL classification from the table will be HA2.1 Fire orexplosion in table H-1 area AND Plant personnel report visible damage to any safety related structure, system, or component within the area OR Affected system parameterindications show degraded performance As per the Bases for the EAL the wording ofthis EAL does not imply that an assessment of safety related structure system andcomponent performance should be performed; rather that safety related structuresystem and component are degraded as a result of the event. In this case all four SWpumps have been affected. Or CA3.1 An unplanned event results in RCS temperature> 200 DEGF for > Table C-3 duration (Note 3 = The SEM should not wait until the applicable time has elapsed, but should declare the event as soon as it is determinedthat the condition will likely exceed the applicable time.) B. INCORRECT (1) 3.99 (2) NOUEThe first part is incorrect but plausible if the candidate incorrectly uses the table basedon conditions for Non-Back to Back Refueling outages, then the expected heat up ratewould be 3.99 DEGF/hour. Second part is also incorrect but plausible if the candidateselects EAL HU2.1 from the table due to the fire at the SWPH not extinguished within 15 min. or CU3.1 An unplanned event results in RCS temperature > 200 DEGFC. INCORRECT (1) 3.20 (2) NOUEFirst part is correct See above B answer. Second part is incorrect See above answer A D. INCORRECT (1) 3.99 (2) AlertFirst part is incorrect See above answer A. Second part is correct See above answer B K/A:076G2.1.25 Service Water System (SWS)Ability to interpret reference materials, such as graphs, curves, tables, etc.Technical

References:

EAL BasesEAL Table 0-AP-27 "Malfunction of spent fuel pit system"0-AP-12 "Loss of Service Water"0-FCA-9 "Service Water Pump House Fire"1-EI-CB-21E Annuciator C5 SFP HI/HI-HI TEMP (1-AR-E-C5)11715-FM-79C (sh 3 of 5)

UFSAR 9.1.3.3.1References provided to applicants:

• 0-AP-27 Attachment 7 "Spent Fuel Pool Heat-Up rate following loss of cooling"* EAL Tables Learning Objective:U 18030Perform various actions of 0-AP-27 "Malfunction of spent fuel pit system"U 18011Perform the actions of 0-AP-12 "Loss of Service Water"U 13910Explain the concepts associated with responding to a fire in the service water pumphouse in accordance with 0-FCA-9 U 13699Explain importance of correctly applying classification criteria to actual plant conditionswhen classifying an emergency.Question Source: Modified NAPS Vision Bank The following conditions existUnit-1 is currently in a scheduled refueling outageThe refueling team is currently latching control rods in the unit one containment Unit 2 has been at 100% power for 3 month following a 26 day refueling outage Both SFP cooling pumps are currently isolated due to a leak on a common lineAnnunciator E-C5, SFP Hi/Hi-Hi TEMP has illuminatedPer 0-AP-27 "Malfunction of Spent Fuel Pit System" the expected heat-up rate for the spent fuel pit under these conditions is ________ degrees Fahrenheit per hour (F/hr)A. 3.99B. 4.29 (Correct)C 9.91D 11.40 Question History:NoneQuestion Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content:SRO only - 10 CFR-55.43.(b)(5)Assessment of facility conditions and selection of appropriate procedures duringnormal, abnormal, and emergency situations.

10 CFR-55.41.(a)(b)(10)Administrative, normal, abnormal, and emergency operating procedures for the facility.Comments:

This question matches the K/A statement by requiring the SRO applicant to have the ability to interpret reference materials, such as graphs, curves, tables, etc. associatedwith the SW system. On a loss of all service water, component cooling temperaturesare expected to increase, with one unit having just been shutdown, RHR in service andthe large Decay Heat factor, this will have a large effect on the Spent Fuel Pittemperatures. To arrive at the correct answer the SRO applicant must posses the ability to determine the proper heat up rate of the Spent Fuel Pool based on the abilityto use the correct table for the questions given conditions, and then determine thecorrect heat up rate based on the status of the core. Second the SRO candidate mustbe able to demonstrate the correct use of the EAL tables to determine the highest EALclassification for the stated conditions.

QUESTIONS REPORTfor SRO Exam Jan Submittal

8. 090 - 103G2.2.38 001/SRO/T2/G1/3.6/4.5/NEW//Unit-1 is at 100%, when Annuciator 1J-G1 "Containment Partial Pressure +0.25 PSICH I-II", locks in with the following conditions:

• RWST = 48 DEGF* Service Water Temp = 83 DEGF* Partial Pressure Setpoint is set at 10.85 psia on Ch I and II

• Containment Temperature = 96.0 DEGF Which of the following completes the statement?Based on these conditions, containment air partial pressure is ____________ foroperations AND the limits on pressure and temp ensures that the ___________.Acceptable, Peak containment pressure will be limited to the upper containment pressure of 45psigUnacceptable,Containment pressure will not reach the lower design pressure of 8.6 psia for aninadvertent containment spray actuation. Unacceptable, Peak containment pressure will be limited to the upper containment pressure of 45psigAcceptable, Containment pressure will not reach the lower design pressure of 8.6 psia for aninadvertent containment spray actuation.

A.B.C.D.Distractor Analysis:

The candidate is required to have the knowledge of conditions and limitations in thefacility license as they relate to containment. A. INCORRECT Acceptable, Peak containment pressure will be limited to the upper containment pressure of 45 psig.First part is incorrect but is plausible because the partial pressure is given to thecandidate as the alarm which is +0.25 PSI greater than the setpoint which is given as10.85 psia, so the sum of the two (11.1 psia) is the point used on the graph. if thecandidate does not add the two then the point will fall in the acceptable region. Secondpart is the correct limit as per the bases of TS 3.6.4 "Containment Pressure". If thecandidate is not familiar with the bases value, it can also be found in section 5.5.15(b)

Administrative Controls for Containment Leakage Rate Testing Program. B. INCORRECT Unacceptable, Containment pressure will not reach the lower design pressure of 8.6 psia for an inadvertent containment spray actuation. Wednesday, January 27, 2016 1:57:57 PM 29 QUESTIONS REPORTfor SRO Exam Jan SubmittalFirst part is correct the pressure falls in the unacceptable region of the gragh Figure3.6.4-1 (alarm which is +0.25 PSI greater than the setpoint which is given as 10.85psia, so the sum of the two (11.1 psia) is the point used on the graph). and the secondpart in incorrect, but plausible because, the design pressure is 5.5 psia, but theanalysis calculation used in the TS bases shows the minimum pressure that should bereached inside of containment would be 8.6 psia. C. CORRECT Unacceptable, Peak containment pressure will be limited to the uppercontainment pressure of 45 psigFirst part is correct the pressure falls in the unacceptable region of the gragh Figure3.6.4-1 (alarm which is +0.25 PSI greater than the setpoint which is given as 10.85psia, so the sum of the two (11.1 psia) is the point used on the graph). and the Secondpart is the correct limit as per the bases of TS 3.6.4 "Containment Pressure". If thecandidate is not familiar with the bases value, it can also be found in section 5.5.15(b)Administrative Controls for Containment Leakage Rate Testing Program.D. INCORRECT Acceptable, Containment pressure will not reach the lower design pressure of 8.6 psia for an inadvertent containment spray actuation.First part is incorrect but is plausible because the partial pressure is given to thecandidate as the alarm which is +0.25 PSI greater than the setpoint which is given as10.85 psia, so the sum of the two (11.1 psia) is the point used on the graph. if thecandidate does not add the two, the point will fall in the acceptable region. The second part is incorrect, but plausible because, the design pressure is 5.5 psia, but the analysiscalculation used in the TS bases shows the minimum pressure that should be reachedinside of containment would be 8.6 psia.

K/A:

103G2.2.38 Containment System Knowledge of conditions and limitations in the facility license. Technical

References:

1-AR-J-G1 "Containment Partial Press +0.25 PSI CH I-II" 11715-LM-013 11715-LM-014 1-AP-18 "Increasing containment pressure" TS 3.6.4 Containment Pressure and BasesTS 3.6.4 Figure 3.6.4-1 References provided to applicants: Figure 3.6.4-1 of TS 3.6.4 Learning Objective:U 5756 List the following information concerning the containment stucture.Wednesday, January 27, 2016 1:57:57 PM 30 QUESTIONS REPORTfor SRO Exam Jan Submittal* Design PressureU 16273Explain the concepts associated with the containment pre ssure technical specificationand bases U 16274Explain the concepts associated with the contaiment air temperature technicalspecification and bases U 18017Perform actions of 1-AP-18 "Increasing Containment Pressure" U 4375 List all indications available to the control room operator for abnormal containmentpressure, including all annunciator setpoints. Question Source: NewQuestion History: Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content:SRO only 10 CFR 55.43 (b)(1)

Conditions and limitations in the facility license 10 CFR 55.45 (a)(13)Demostrate the applicant's ability to function within the control room team asappropriate to the assigned position, in such a way that the facility licensee'sprocedures are adhered to and that the limitations in its license and amendments arenot violated 10 CFR 55.41 (a)(b)(10)

Administrative, normal, abnormal, and emergency procedures for the facility10 CFR 55.41 (a)(b)(7)

Design, components, and functions of control and safety systems, includinginstrumentation, signals, interlocks, failures modes , and automatic and manualfeatures. Comments:This question matches the K/A statement by requiring the SRO applicant to have theknowledge of conditions and limitations in the facility license as they relate tocontainment. To arrive at the correct answer the SRO applicant must have theknowledge to correctly assess plant conditions, and interpret those conditions, todetermine the proper technical specification requirement that needs to be applied. Alsothe applicant is asked to possess the knowledge of the importance behind therequirement as outlined in the technical specifications bases document. Wednesday, January 27, 2016 1:57:57 PM 31 091 - 015G2.2.25 001/SRO/T2/G2/3.2/4.2/MODIFIED//Unit-1 status is as follows:* A reactor startup is in progress.* Both Intermediate Range channels indicate approximately 5 E -11 amps.

• Source Range Channel N-31 fails low.

• The Crew initiates 1-AP-4.1 "Malfunction of Source Range Nuclear Instrumentation".Which One of the following describes (1) the 1-AP-4.1 required response and (2) theTechnical Specification Bases for the response? (1) Continue the reactor startup; (2) With only one source range channel operable 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> is allowed to restore two channels to service, Protection is provided by the logic of 1/2 SR channels > 10 E 5 cps.(1) Suspend the reactor startup; (2) One decade of overlap between both SR and IR channels cannot be verified. (1) Suspend the reactor startup; (2) With only one source range channel operable, a single random failure will disable the SR Trip function.

(1) Continue the reactor startup; (2) Intermediate Range and Power Range low setpoint trip functions provide the necessary redundant protection. A.B.C.D.Distractor Analysis:This requires the examinee to have the knowledge of the bases in TechnicalSpecifications for limiting conditions for operations and safety limits. A. INCORRECT (1) Continue the reactor startup; (2) With only one source range channel operable 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> is allowed to restore two channels to service. Protection is provided by the logic of 1/2 SR channels > 10 E 5 cps.The first part is incorrect but plausible if the operator is unfamiliar with the abnormalprocdedure guidance associated with the SR failing Low. The Trip Logic is 1/2 SR channels > 10 E 5 cps, so a Rx trip will not occur. The fact that this occurs at < P-6value (1 X 10 E -10 amps) then the AP checks Reactor power constant, and thenrestores affected channel to service before exceeding P-6 permissive. Because theprocedure does not required inserting rods or opening the Rx trip breakers, it couldmake this a plausible choice for the candidate. The second part is also incorrect but aplausible choice matched with the first answer because T.S. 3.3.1, Table 3.3.1-1 alsolist action J as a condition associated with one source range channel inoperable. If the candidate associates this completion time of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> with a Mode 2 requirement thenthis answer could be matched together with the first part and be chosen. Alsocombining in the idea that protection is still being provided with the Rx trip logic alsomakes this a plausble Tech Spec bases choice.

B. INCORRECT (1) Suspend the reactor startup; (2) One decade of overlap between both SR and IR channels cannot be verified. The first part is correct. The source range trip logic is 1/2 SR channels > 10 E 5 cps, soa Rx trip will not occur. The fact that this occurs at < P-6 value (1 X 10 E -10 amps)then the AP checks Reactor power constant and then restores affected channel to service before exceeding P-6. The second part is incorret but plausible because thestart up procedures (OP-1.5 / 1.7) has the operators verify that proper overlap isoccuring between the SR and IR instrumentation before maually blocking the SourceRange Neutron Flux trip at the P-6 permissive. The candidate may assume this is listedin the bases for the SR intrumentation. C. CORRECT (1) Suspend the reactor startup; (2) With only one source range channel operable, a single random failure will disable the SR Trip function.The first part is correct. The source range trip logic is 1/2 SR channels > 10 E 5 cps, soa Rx trip will not occur. The fact that this occurs at < P-6 value (1 X 10 E -10 amps)then the AP checks Reactor power constant and then restores affected channel toservice before exceeding P-6. The second part is also correct, The Tech Spec 3.3.1bases for SR specifies " Two Operable channels are sufficient to ensure no singlerandom failure will disable this Trip function. D. INCORRECT (1) Continue the reactor startup; (2) Intermediate Range and Power Range low setpoint trip functions provide the necessary redundant protection. The first part is incorrect but plausible if the operator is unfamiliar with the abnormalprocdedure guidance associated with the SR failing Low. The Trip Logic is 1/2 SRchannels > 10 E 5 cps, so a Rx trip will not occur. The fact that this occurs at < P-6value (1 X 10 E -10 amps) then the AP checks Reactor power constant, and thenrestores affected channel to service before exceeding P-6 permissive. Because the procedure does not required inserting rods or opening the Rx trip breakers, it couldmake this a plausible choice for the candidate. Second part is incorrect but plausibleand could be matched with the first part because the basis associated with SR neutronflux mentions "Above P-6 setpoint, the IR and PR low setpoint trip will provide coreprotection for reactivity accidents. K/A:015G2.2.25Nuclear Instrumentation SystemKnowledge of the bases in Technical Specifications for limiting conditions foroperations and safety limits.Technical

References:

Technical specification basis 3.3.1 "Reactor Trip System (RTS) Instrumentation" 1-AP-4.1 "Malfunction of Source Range Nuclear Instrumentation"1-OP-1.7 "Unit Startup from Mode 3 to Mode 2 following refueling"1-OP-1.5 "Unit Startup from Mode 3 to Mode 2" References provided to applicants: NoneLearning Objective:Question Source: Modified Bank Wolf Creek 2006 SRO ILT exam #96 (015G2.2.25)Given the following conditions:A Rx startup is in progressBoth IR channels indicate approximately 5 E -11SR channel N-31 fails DOWNSCALE Which ONE (1) on the following describes the required operator response and theTechnical Specification basis for the response?A. Continue the Rx startup; with only one source range channel operable; 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> isallowed to restore two channels to serviceB. Suspend the reactor startup; SR channels are not required to trip the reactorhowever the SR monitoring functions must be available.C. Continue the reactor startup; the IR flux trip and the PR flux low trp provide thenecessary core protectionD. Suspend the Rx startup; with only one SR channel operable, the minimum required SR high Flux Trip protection is not met. (CORRECT) Question History: None Question Cognitive Level: Memory 10 CFR Part 55 Content:SRO only 10 CFR-55.43(b)(2)Facility operating limitations in the technical specifications and their bases10 CFR-55.41(b)(5) Facility operating characteristics during steady state and transient conditions, includingcoolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operatingcharacteristicsComments:This question matches the K/A statement by requiring the SRO applicant todemonstrate a knowledge of the bases in technical specifications for limiting conditionsfor operations and safety limits. To arrive at the correct answer the SRO applicant must recognize the operational requirements, associated with a loss of the sourcerange nuclear instrumentation, as they would apply to a Reactor Startup. Along withthe technical specification bases behind the limit imposed.

092 - 028A2.02 001/NEW/2/2/3.5/3.9/MODIFIED//NOA LOCA has occurred on Unit-1 with the following conditions:

• Crew is in 1-E-1 "Loss of Reactor or Secondary Coolant".

• The crew is at Step 19 to check containment Hydrogen Concentration.

• Containment Pressure is 22 psia

• Containment Hydrogen Concentration is 3.5%

• The Technical Support Center is fully manned and activated. Which of the following answers completes the statement? The Analyzer will provide an accurate reading _____(a)_____ minutes after beingplaced in service and based on containment Hydrogen concentration the Hydrogen recombiner will _____(b)______ . (a) 15 (b) be placed in service, using 1-OP-63.1 "Post Accident thermal H2 Recombiner"(a) 15(b) not be placed in service, Consult TSC for additional recovery actions(a) 5(b) be placed in service, using 1-OP-63.1 "Post Accident thermal H2 Recombiner" (a) 5(b) not be placed in service, Consult TSC for additional recovery actionsA.B.C.D.Distractor Analysis:Malfunctions or operations on the HRPS; and (b) based on those predictions, use procedures to correct, control or mitigate the consequences of those malfunctions oroperations: LOCA condition and related concern over hydrogen. A INCORRECT (a) 15 (b) be placed in service, using 1-OP-63.1 "Post Accident thermal H2 Recombiner".

First part is incorrect but plausible if the candidate in unfamiliar with the Note in 1-E-1just prior to step 19 to place the analyzer in service "There is a delay time of 5 minutesbefore the containment Hydrogen Analyzer will provide an accurate reading. Secondpart is correct concentration set point for placing the Hydrogen Recombiner in service is between 0.5% and 4%, for any concentration above or below this band, the proceduresrequire the operators to contact the TSC/Plant staff for recovery actions. Even thoughthe question provides a containment pressure of 22 psia, there are no alternate maxand min setpoints associated with placing the recombiner in service under adversecontainment conditions. B INCORRECT (a) 15 (b) not be placed in service, Consult TSC for additional recovery actions.First part is incorrect but plausible if the candidate in unfamiliar with the Note in 1-E-1just prior to step 19 to place the analyzer in service "There is a delay time of 5 minutesbefore the containment Hydrogen Analyzer will provide an accurate reading. Secondpart is incorrect but plausible because the recombiner can be placed in service at aconcentration between 0.5% and 4%, but the given containment pressure of 22 psia isequal to an adverse containment condition, this could draw in the candidate, andprovides a test on their knowledge of the adverse containment requirements. Olderrevisions of the emergency procedures had an adverse containment pressuremaximum concentration value of 1.6% for placing the Recombiner in service. C INCORRECT (a) 5 (b) be placed in service, using 1-OP-63.1 "Post Accident thermal H2 Recombiner".First part is correct Note in 1-E-1 just prior to step 19 to place the analyzer in service"There is a delay time of 5 minutes before the containment Hydrogen Analyzer willprovide an accurate reading. Second part is correct concentration set point for placingthe Hydrogen Recombiner in service is between 0.5% and 4%. Even though thequestion provides a containment pressure of 22 psia, there are no alternate max andmin setpoints associated with placing the recombiner in service under adversecontainment conditions. D CORRECT (a) 5 (b) not be placed in service, Consult TSC for additional recovery actions.First part is correct Note in 1-E-1 just prior to step 19 to place the analyzer in service"There is a delay time of 5 minutes before the containment Hydrogen Analyzer willprovide an accurate reading. Second part is incorrect but plausible because the recombiner can be placed in service at a concentration between 0.5% and 4%, but thegiven containment pressure of 22 psia is equal to an adverse containment condition,this could draw in the candidate, and provides a test on their knowledge of the adversecontainment requirements. Older revisions of the emergency procedures had an adverse containment pressure maximum concentration value of 1.6% for placing therecombiner in service. K/A: 092 - 028A2.02 Technical

References:

1-E-1 "Loss of Reactor and Secondary Coolant" and Westinghouse backgroundinformation1-ECA-1.1"Loss of Emergency Coolant Recirculation" and Westinghouse backgroundinformation 1-ECA-3.1"SGTR with Loss of Reactor Coolant Subcooled Recovery Desired" andWestinghouse background information1-ECA-3.2 "SGTR with Loss of Reactor Coolant Saturated Recovery Desired" andWestinghouse background information1-ES-1.2 "Post LOCA cooldown and Depressurization" and Westinghouse background information1-FR-C.1 "Response Inadequate Core Cooling" and Westinghouse backgroundinformation1-FR-I.3 "Responds to voids in the Reactor Vessel" and Westinghouse backgroundinformation1-OP-63.1 "Post Accident Thermal Hydrogen Recombiner"1-OP-63.2 "Containment Hydrogen Analyzer" 11715-FM-106A Sht 4 References provided to applicants: NoneLearning Objective:U 5451List the design basis of the Containment Atmosphere Cleanup SystemU 5340Assuming that the purge blowers and hydrogen recombiner are operable, explain theconditions under which the purge blowers and not the recombiner would be used forremoving hydrogen from containment during a LOCAU 5450List information associated with hydrogen in containment U 17487Explain the concepts associated with the Containment Hydrogen Analyzers technical Question Source: Modified St. Lucie 2012 SRO ILT exam bank (K/A 028A2.02) Question History: Unit 1 has entered 1-EOP-03, LOCA with the following conditions:* Containment pressure is 12 psig* All containment coolers and Containment spray pumps are operating

• Containment Hydrogen concentration is 4.5%The TSC is fully manned and plant conditions have been communicated to the TSC Which ONE of the following states the status of the Containment Temperature &Pressure Safety function AND when should the H2 purge system or HydrogenRecombiners be placed in service?1) Containment Temperature & Pressure Safety function is:2) The H2 purge system or Hydrogen Recombiner should be placed in service:A. 1) Met 2) only as directed by the Technical Support CenterB. 1) Met 2) when the Unit Supervisor deems it necessary based on Containment H2 concentrationC. 1) Not Met (CORRECT) 2) only as directed by the Technical Support CenterD. 1) Not Met

2) when the Unit Supervisor deems it necessary based on Containment H2 concentrationQuestion Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content:SRO only - 10 CFR-55.43 (b)(5)Assessment of facility conditions and selection of appropriate procedures during normalabnormal, and emergency situations. 10 CFR-55.41 (a)(b)(5)Facility operating characteristics during steady state and transient conditions, includingcoolant chemistry, causes and effects of temperature, pressure and reactivity changes and operating limitations and reasons for these operating characteristics. Comments:This question matches the K/A statement by requiring the SRO applicant to have theability to correctly direct operation of the Hydrogen Recombiner under LOCA conditions. To arrive at the correct answer the SRO applicant must have the knowledgeto determine when accurate and reliable information is being obtained by availableindications, which will be used to coordinate the crew to properly make the decisionconcerning operation of the Hydrogen Recombiner and coordination with supportingstaff members.

QUESTIONS REPORTfor SRO Exam Jan Submittal

9. 093 - 071G2.4.21 001/SRO/T2/G2/4.0/4.6/BANK/NAPS 2010/The Waste Gas Decay Tanks (WGDT) was sampled to determine the quantity ofradioactive material contained in each gas storage tank.Confirmed sample results are as follows:* "A" WGDT - 7,000 curies of noble gas* "B" WGDT - 16,000 curies of noble gas Based on these sample results, which ONE of the following identifies the implicationsper TR 3.10.3 Gas Storage Tanks? The quantity of radioactive material in the WGDTs ____(1)______. The TRM bases forrestricting the quantity of radioactive material is to limit dose at the exclusion boundaryto _____(2)______. (1) is NOT within limits, enter actionTR 3.10.3 action (2) 2.0 rem (1) is within limits, entry into TR 3.10.3 in NOT required (2) 0.5 rem (1) is NOT within limits, enter actionTR 3.10.3 action (2) 0.5 rem (1) is within limits, entry into TR 3.10.3 in NOT required (2) 2.0 rem A.B.C.D.Wednesday, January 27, 2016 1:57:57 PM 32 QUESTIONS REPORTfor SRO Exam Jan SubmittalDistractor Analysis:This requires the examinee to have knowledge of the Waste Gas Disposal parametersand logic used to assess the status of safety functions, such as reactivity control, corecooling and heat removal, reactor coolant system integrity, containment conditions,radioactivity release control, etc..A INCORRECT (1) is NOT within limits, enter actionTR 3.10.3 action (2) 2.0 remThe first part is incorrect by plausible since the value of "B" tank is well in excess ofthose that would normally ever be encountered and the candidate may be aware thatthere is a limit, but if they are not sure of the specifiec value they may default to thisdistractor based on the large values given. Second part is also incorrect but plausiblebecause the TRM bases states for an uncontrolled release the total body exposure toan individual at the nearest exclusion boundary will not exceed 0.5 rem in an event of 2hours. B CORRECT (1) is within limits, entry into TR 3.10.3 in NOT required (2) 0.5 rem First part is correct 25,000 is the TR Limit so we are within it. The second part is alsocorrect per the TR bases.C INCORRECT (1) is NOT within limits, enter actionTR 3.10.3 action (2) 0.5 rem The first part is incorrect by plausible since the value of "B" tank is well in excess ofthose that would normally ever be encountered and the candidate may be aware thatthere is a limit, but if they are not sure of the specifiec value they may default to thisdistractor based on the large values given. The second part is correct.D INCORRECT (1) is within limits, entry into TR 3.10.3 in NOT required (2) 2 rem The first part is correct. Second part is incorrect but plausible because the TRM basesstates for an uncontrolled release the total body exposure to an individual at the nearest exclusion boundary will not exceed 0.5 rem in an event of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

K/A:

071G2.4.21 Waste Gas Disposal System (WGDS)

Knowledge of the parameters and logic used to assess the status of safety functions,such as reactivity control, core cooling and heat removal, reactor coolant systemWednesday, January 27, 2016 1:57:57 PM 33 QUESTIONS REPORTfor SRO Exam Jan Submittalintegrity, containment conditions, radioactivity release control, etc.Technical

References:

TR 3.10.3 "Gas Storage Tanks" TR 3.10.3 Bases References provided to applicants: NoneLearning Objective:

U 17527 Explain the concepts associated with the Gas Storage Tanks technical requirementsand bases (TR-3.10.3)Question Source: 2010 NAPs ILT SRO test Question 97 (G2.3.6)Question History: The Waste Gas Decay Tanks (WGDT) was sampled to determine the quantity ofradioactive material contained in each gas storage tank.Confirmed sample results are as follows:* "A" WGDT - 15,000 curies of noble gas* "B" WGDT - 6,000 curies of noble gas Based on these sample results, which ONE of the following identifies the implicationsper TR 3.10.3 Gas Storage Tanks AND includes the associated TRM BASES? The quantity of radioactive material in the WGDTs ____(1)______. The TRM base for restricting the quantity of radioactive material is to limit dose at the exclusion boundaryto _____(2)______. A. is within limits, entry into TR 3.10.3 action is NOT required; 0.5 rem (CORRECT)B. is within limits, entry into TR 3.10.3 action is NOT required; 0.1 remC. is NOT within limits, enter TR 3.10.3 action; 0.5 rem D. is NOT within limits, enter TR 3.10.3 action; 0.1 remQuestion Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content:SRO only - 10 CFR 55.43 (b)(5)Assessment of facility conditons and selection of appropriate procedures duringnormal, abnormal, and emergency situations 10 CFR-55.41 (a)(b)(7)Design, components, and functions of control and sagety systems, includinginstrumentation, signals, interlocks, failure modes, and automatic and manual features. Comments:

This question matches the K/A statement by requiring the SRO applicant to have aWednesday, January 27, 2016 1:57:57 PM 34 QUESTIONS REPORTfor SRO Exam Jan Submittalknowledge of limits placed on the Gas storage tanks and to have an understanding ofthe bases behind the limits set forth in technical specifications. To arrive at the correctanswer the SRO applicant must recognize what value the quantity of radioactivematerial contained in each storage tank is limited to AND why that limit providesprotection. Wednesday, January 27, 2016 1:57:57 PM 35 094 - G2.1.40 001/SRO/T3//2.8/3.9/NEW//Unit-1 has been shutdown from 100% power for a scheduled refueling outage.In accordance with OP-AA-100 "Conduct of Operations", which one of the followingchoices answers the statement? The _____________ is responsible for overall supervision and coordination of refuelingoperations, including fuel movement. Operations Manager Fuel Handling Supervisor Shift Manager Refueling SROA.B.C.D.

Distractor Analysis:This requires the examinee to have the knowledge of refueling administrativerequirements. A. INCORRECT Operations ManagerIncorrect but plausible, during refueling operation at NAPS there is a OperationsManager assigned to be at the station during dayshift and nightshift activities. Thecandidate may assume that overall responsibility is assigned to this individualB. INCORRECT Fuel Handling Supervisor Incorrect but plausible, 1-OP-4.1 "Controlling Procedure for Refueling" and other OP-4series procedures associated with refueling includes various steps where the FuelHanding Supervisor authorization is required, and if the candidate associates overallsupervision of refueling operations as his responsibility this distractor may be chosen.C. INCORRECT Shift ManagerIncorrect but plausible because the Shift Manger has control room oversight and at alltimes the Shift Manger is in concurrence with the outage unit acitivities. The candidate may assume that overall responsibility is assigned to this individual. D. CORRECT Refueling SROIn accordance with OP-AA-100 "Conduct of Operations" Attachment 6, The RefuelingSRO is responsible for overall supervision and coordination of refueling operations,including fuel movement. When assigned, this refueling SRO has NO other concurrentresponsibilities.

K/A:G2.1.40Knowledge of refueling administrative requirements.Technical

References:

1-OP-4.1 "Controlling Procedure for refueling"OP-AA-100 "Conduct of Operations"References provided to applicants: NoneLearning Objective:U 13561Explain Personnel requirements associated with core alterations (OP-AA-100) Question Source: NEW Question History: Question Cognitive Level: Memory or Fundamental Knowledge10 CFR Part 55 Content:SRO only 10 CFR-55.43(b)(5)Assessment of facility conditions and selection of appropriate procedures duringnormal, and emergency situations.

10 CFR-55.41(a)(b)(10)Administrative, normal, abnormal, and emergency operating procedures for the facilityComments:This question matches the K/A statement by requiring the SRO applicant to have the knowledge of refueling administrative requirements. To arrive at the correct answer theSRO applicant must have the knowledge of OP-AA-100 "Conduct of Operations", whichlists the duties and responsibilities of the Refueling SRO.

095 - G2.2.14 001/NEW/3//3.9/4.7/3.3//NOMechanics have a work order to replace the cable spreading to control room stairwell door due to a broken hinge.

• The crew has entered action of T.S. 3.7.10 "Main Control Room/Emergency Switchgear Room (MCR/ESGR) Emergency Ventilation System (EVS).

• 1-LOG-17 "Unit 1 & 2 Control Room Boundary Breaching Log" has been initiated. The control room boundary can NOT be maintained operable by use of administrative control. Which one of the following will complete the statement?In accordance with T.S. bases, which Emergency Switchgear Emergency Filtered AirSupply Fan, can NOT be used to supply the main control room and relay rooms in thePressurization Mode. 2-HV-F-41 1-HV-F-41 1-HV-F-42 2-HV-F-42 A.B.C.D.

Distractor Analysis:This requires the examinee to have the knowledge of the process for controllingequipment configuration or status. In accordance with ER-NA-CRH-100 "Control RoomHabitability Program" Control room boundary breeches shall be controlled byprocedurally administrative controls, NAPS has adopted the use of 1-LOG-17 "Unit 1 &2 Control Room Boundary Breaching Log" as the procedure to establish an equipmentconfiguration for appropriate compensatory actions depending on the nature of thebreach.A. INCORRECT 2-HV-F-41As per the Bases associated with T.S. 3.7.10 "Main Control Room/EmergencySwitchgear Room (MCR/ESGR) Emergency Ventilation System (EVS) 3 of the 4 Emergency Switchgear Emergency Filtered Air Supply Fans can be used to supportcontrol room habitability. 1-HV-F-41 CAN NOT be used due to the location of its airintake with respect to Vent Stack B B. CORRECT 1-HV-F-41See above explanation of the bases associated with T.S. 3.7.10 C. INCORRECT 1-HV-F-42 See Above explanation of the bases associated with T.S. 3.7.10 D. INCORRECT 2-HV-F-42 See Above explanation of the bases associated with T.S. 3.7.10 K/A:095 - G2.2.14Technical

References:

1-LOG-17 "Unit 1 & 2 Control Room Boundary Breaching Log" T.S. 3.7.10 "Main Control Room/Emergency Switchgear Room (MCR/ESGR)Emergency Ventilation System (EVS)

Bases of T.S. 3.7.10ER-NA-CRH-100 "Control Room Habitability Program"References provided to applicants: NoneLearning Objective:U 16288Explain the various concepts associated with the main control room/emergencyswitchgear room (MCR/ESGR) emergency ventilation system (EVS)U 5782Information associated with the control room ventilation fans and filtersU 5784Explain the purpose of the control room emergency fans Question Source: NEWQuestion History:

Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content:SRO only - 10 CFR-55.43 (b)(3)Facility licensee procedures required to obtain authority for design and operatingchanges in the facility. 10 CFR-55.41 (a)(b)(10)Administrative, normal, abnormal, and emergency operating procedures for the facilityComments:This question matches the K/A statement by requiring the SRO applicant to possessthe knowledge for controlling equipment configuration to maintain the Unit 1 & Unit 2Control Room Pressure Boundary. In accordance with ER-NA-CRH-100 "Control RoomHabitability Program" Control room boundary breeches shall be controlled byproceduralized administrative controls, NAPS has adopted the use of 1-LOG-17 "Unit 1

& 2 Control Room Boundary Breaching Log" as the procedure to establish anequipment configuration for appropriate compensatory actions depending on the natureof the breach. To arrive at the correct answer the SRO must know the requirementsstated in T.S. 3.7.10 bases pertaining to the ventilation fans and their use, and also beaware of the administrative procedure requirement of 1-LOG-17.

096 - G2.2.21 001/SRO/3//3.9/4.3/NEW//Unit-2 Mode 1, 100% power.The safeguard operator discovers a small packing leak on 2-MS-TV-201B ("B" MainStream Trip Valve). The leak comes in surges but is not steaming. The leak has beenquantified at 30 dpm. CR was submitted and WO is issued to adjust packing to the lastknown torque value IAW component engineering recommendation.The WOs Post Maintenance Testing sheet requires the following:

• Valve timed stroke according to 2-PT-212.9 "Valve inservice inspection"

• External leak checkBased on plant conditions, the PMT requirement to perform 2-PT-212.9 will need to be ____(1)______, and at the minimum, justification need to be provided by ____(2)____.(1) Deferred (2) Engineering(1) Waived (2) FSRC (1) Waived(2) Engineering(1) Deferred(2) FSRC A.B.C.D.Distractor Analysis:The candidate is required to have the knowledge of pre- and post-maintenanceoperability requirements A. INCORRECT (1) Deferred (2) EngineeringFirst part is incorrect but plausible if the candidate is not familiar with the definitionsection of VPAP- 2003 "Post Maintenance Testing Program" which defines DeferredTest as a test requirement that will not be performed by the time the test data sheet is closed. Test deferrals still means that the test requirements need to be completed, theVPAP also states "All post maintenance testing requirements normally should becompleted prior to declaring the equipment operable. 2-MS-TV-201B is required to beoperable in Modes 1, 2 and 3 as per TS 3.7.2 "Main Steam Trip Valves (MSTVs)". Second part is correct, at the minimum, justification shall be provided by engineering. VPAP-2003 includes "Providing written justification when a test listed in the PMTrequirements section of the Test Data Sheet is to be waived, as a responsibility of Engineering. Engineering would also determine if a higher level of justification isrequired by FSRC B. INCORRECT (1) Waived (2) FSRC First part is correct, VPAP-2003 defines a waived test as " a test which appears on theTest Data Sheet but will not be completed as specified or performed at all based on aspecial set of plant or maintenance conditions. Tests may only be waived prior to thetest being performed; it is not permissible to disposition a failed test by waiving it.

Second part is incorrect but plausible because, at the minimum, justification shall beprovided by engineering. VPAP-2003 includes "Providing written justification when atest listed in the PMT requirements section of the Test Data Sheet is to be waived, as aresponsibility of Engineering. Engineering could then determine if a higher level ofjustification is required by FSRC C. CORRECT (1) Waived (2) EngineeringVPAP-2003 defines a waived test as " a test which appears on the Test Data Sheet butwill not be completed as specified or performed at all based on a special set of plant ormaintenance conditions. Tests may only be waived prior to the test being performed; it is not permissible to disposition a failed test by waiving it. The minimum, justificationshall be provided by engineering. VPAP-2003 includes "Providing written justificationwhen a test listed in the PMT requirements section of the Test Data Sheet is to bewaived, as a responsibility of Engineering. D. INCORRECT (1) Deferred (2) FSRC Both sections are incorrect. First part is incorrect but plausible if the candidate is notfamiliar with the definition section of VPAP- 2003 "Post Maintenance Testing Program"which defines Deferred Test as a test requirement that will not be performed by thetime the test data sheet is closed. Test deferrals still means that the test requirements need to be completed, the VPAP also states "All post maintenance testingrequirements normally should be completed prior to declaring the equipment operable.2-MS-TV-201B is required to be operable in Modes 1, 2 and 3 as per TS 3.7.2 "MainSteam Trip Valves (MSTVs)". Second part is incorrect but plausible because, at theminimum, justification shall be provided by engineering. VPAP-2003 includes"Providing written justification when a test listed in the PMT requirements section of theTest Data Sheet is to be waived, as a responsibility of Engineering. Engineering could then determine if a higher level of justification is required by FSRC K/A:

Knowledge of pre- and post-maintenance operability requirements.Technical

References:

Operations Site Specific Instructions OP-AA-100-1001 "Operations DepartmentInstructions"(Title: Operations PMT Review) VPAP-2003 "Post Maintenance Testing Program" OP-AA-200 "Equipment Clearance"WM-AA-100 "Work Management"North Anna PMT Test Data Sheet Work Order #59102710332References provided to applicants: NoneLearning Objective:U 13130 List the conditions that are acknowledged and approved by the shift manager upon approval of a work order.U 13634Explain the Operations Department's responsibilities associated with post-maintenancetesting) U 13120Explain the how to determine whether testing is required prior to restoringsafety-related equipment to service. Question Source: NEWQuestion History:

Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content:SRO only - 10 CFR-55.43(b)(2)

Facility operating limitations in the technical specifications and their bases.10 CFR-44.41(a)(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facilityComments:This question matches the K/A statement by requiring the SRO applicant to have theknowledge of pre- and post-maintenance operability requirements. To arrive at thecorrect answer the SRO applicant must have the knowledge to correctly assess plantconditions, and interpret those conditions, during the review and authorization of a WO.Specifically, the candidate needs to know that 2-MS-TV-201B is required under the given plant conditions. Also, the candidate needs the knowledge of VPAP-2003 toprocess the proper course of actions that will allow the required maintenance to be performed, and maintain operability of the system.

QUESTIONS REPORTfor SRO Exam Jan Submittal

10. 097 - G2.3.14 001/SRO/T3//3.4/3.8/MODIFIED/SURRY 2010/Unit-1 is at 100% full power.* 1-GM-F-1 "Isophase bus duct cooling fan No 1 is tagged out due to bad contactor,and is currently being worked by Electricians.

• 1-GM-F-2 "Isophase bus duct cooling fan No 2" suddenly trips.* Crew has entered 1-AP-2.2 "Fast Load Reduction".* Chemistry has been notified to perform and Isotopic Analysis for power reductionsgeater than 15 percent in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. Which one of the following completes the following statement? The RCS activity must be limited to _____(1)_______ DOSE EQUIVALENT IODINE-131. Which ensures the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> doses at the site boundary will not exceedlimits following _______(2)_____ accidents. (1) Less than or equal to 0.1 uCi/gm (2) LOCA (1) Less than or equal to 1.0 uCi/gm (2) SGTR (1) Less than or equal to 1.0 uCi/gm (2) LOCA (1) Less than or equal to 0.1 uCi/gm (2) SGTR A.B.C.D.Wednesday, January 27, 2016 1:57:57 PM 36 QUESTIONS REPORTfor SRO Exam Jan SubmittalDistractor Analysis:This requires the examinee to have the knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities. A INCORRECT (1) Less than or equal to 0.1 uCi/gm (2) LOCA First part is incorrect but plausible because 0.1 uCi/gm is the activity limit for theSecondary Specific Activity per TS 3.7.7, Second part is incorrect but plausible if thecandidate is unfamiliar with the TS bases background or applicable safety analysis,which includes the SGTR but not a LOCA. B CORRECT (1) Less than or equal to 1.0 uCi/gm (2) SGTR First part is correct Reactor Coolant DOSE EQUIVALENT I-131 specific activity verifiedto be less than or equal to 1.0 uCi/gm, and the second part is correct the LCO limits areestablished to minimize the dose consequences in the event of a steam generator tuberupture (SGTR)C INCORRECT (1) Less than or equal to 1.0 uCi/gm (2) LOCA Incorrect but plausible because the first part is correct Reactor Coolant DOSEEQUIVALENT I-131 specific activity verified to be less than or equal to 1.0 uCi/gmSecond part is incorrect but plausible if the candidate is unfamiliar with the TS basesbackground or applicable safety analysis, which includes the SGTR but not a LOCA. D INCORRECT (1) Less than or equal to 0.1 uCi/gm (2) SGTR or SLBFirst part is incorrect but plausible because 0.1 uCi/gm is the activity limit for theSecondary Specific Activity per TS 3.7.7, and the second part is correct the LCO limitsare established to minimize the dose consequences in the event of a steam generatortube rupture (SGTR)

K/A: G2.3.14 Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities.Wednesday, January 27, 2016 1:57:57 PM 37 QUESTIONS REPORTfor SRO Exam Jan SubmittalTechnical

References:

TS 3.4.16 TS Bases 3.4.16 1-PT-53.5 "RCS Specific Activity - Iodine Isotopic Analysis References provided to applicants: NoneLearning Objective:U 16260 Explain the concepts associated with the RCS specific activity technical specificationand bases (TS-3.4.16) U 16285Explain the concepts associated with the secondary specific activity technicalspecification and bases (TS-3.7.7) Question Source: Modified Surry 2010 ILT SRO exam (G2.3.14)Question History: ModifiedCurrent Conditions* Unit-1 is at full powerWhich one of the following completes the statement conserning:1) the DOSE EQUIVALENT IODINE-131 limit for RCS activity in accordance with TS3.1.D Maximum Reactor Coolant Activity AND 2) the assumed release duration throughthe main steam safety valves and atmospheric relief valves in accordance with TS bases?The RCS activity must be limited to _____________ DOSE EQUIVALENTIODINE-131. Primary water assumed to enter the secondary system and be release fora period of

_______________

A. < to 0.1uCi/cc: 60 minutes B. < to 0.1uCi/cc: 30 minutes C. < to 1.0uCi/cc: 60 minutes D. < to 1.0uCi/cc: 30 minutes (CORRECT) Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content:97 - 10 CFR 55.43 (b)(4)

Radiation hazards that may arise during normal and abnormal si tuations, includingmaintenance activities and various contamination conditions.10 CFR 55.41 (a)(b)(12)Radiological safety principles and procedures Wednesday, January 27, 2016 1:57:57 PM 38 QUESTIONS REPORTfor SRO Exam Jan SubmittalComments:This question matches the K/A statement by requiring the SRO applicant to analysisand interpretation of coolant activity, including comparison to emergency plan criteriaand/or regulatory limits, by understanding the bases behind the limits set forth intechnical specifications. To arrive at the correct answer the SRO applicant mustrecognize what value RCS activity is limited to AND under which accident the limitprovides protection. Wednesday, January 27, 2016 1:57:57 PM 39 098 - G2.3.15 001/SRO/3//2.9/3.1/NEW//Due to elevated turbine vibrations, the night shift crew is ramping Unit-1 to Mode 2 fora Turbine Balance Shot. At 20:30 and 79% power, annuciator 1K-G6 "N-16 Rad Det" alarms. 1-MS-RI-191 "B S/G N-16 and 1-MS-RI-193 "MS Header" are in Alert at 5 GPD, thecrew enters 1-AP-5, " Unit 1 Radiation Monitoring Systems." During the ramp the following occurs:

1-MS-RI-191 and 1-MS-RI-193 indications:

• 21:30 and 61% power = 9 GPD

• 22:30 and 43% power = 14 GPD

• 23:30 and 26% power = 20 GPD1-SV-RM-121 "Condenser Air Ejector RM" indicated a consistent but slowly increasingcount rate.The shift STA has confirmed leakage rate trend data is correct. During the ramp, TRM 3.4.5 Primary to Secondary Leakage Detection SystemCondition A will be entered when power goes below ____(1)______ % power, andMandatory frequency of grab samples is _____(2)____ hours. REFERENCE PROVIDED(1) 30(2) 12(1) 30(2) 4(1) 25(2) 12(1) 25(2) 4A.B.C.D.Distractor Analysis:This requires the examinee to have the knowledge of radiation monitoring systems,such as fixed radiation monitors and alarms, portable survey instruments, personnelmonitoring equipment, etc. A. INCORRECT (1) 30 (2) 12First part is incorrect but plausible, If the candidate would consider both N-16 and condenser air ejector radiation monitors non functional at 30% power. As the unit isramping down, at 30% power 1-SV-RM-121 "Condenser Air Ejector RM" is declared Non-functional, this is described in TRM 3.4.5 bases, "RCS Ar-41 activity is expected tobe sufficient to support functionality of the condenser air ejector exhaust system shortlyafter 30% power is reached. At this point N-16 is still functional and meeting therequirement for the continuous readout radiation monitoring systems. Second part isalso incorrect but plausible if the candidate incorrectly reads table 3.4.5-1 of the TRM. When leakage is > 5 GPD and < 30 GPD grab samples shall be once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, butif leakage is > 5 GPD and < 30 GPD AND rate of increase is > 5 GPD then grabsamples will be taken once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.B. INCORRECT (1) 30 (2) 4First part is incorrect but plausible, if the candidate would consider both N-16 andcondenser air ejector radiation monitors not functional at 30% power (SEE plausibilityexplanation for answer A). Second part is correct. Because based on N-16 trends therate of increase is > 5 GPD and grab samples should be taken once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> C. INCORRECT (1) 25 (2) 12First part is correct, During a down power ramp at 30% power the condenser air ejectorRM is declared non-functional, and as per the 3.4.5 TRM bases "RCS N-16 activity will be sufficient to support FUNCTIONALITY of the N-16 system with reactor power at orabove 25% power. At NAPs the shutdown procedures then declare N-16 non-functionalat 25% power. At that point we do not have a continuous readout radiation monitoringsystem functional and Condition A will be entered. The second part is incorrect asstated above (SEE plausibility explanation for answer B)D. CORRECT (1) 25 (2) 4First part is correct, As the unit is ramping down, at 30% power 1-SV-RM-121"Condenser Air Ejector RM" is declared Non-functional, this is described in TRM 3.4.5 bases, "RCS Ar-41 activity is expected to be sufficient to support functionality of thecondenser air ejector exhaust system shortly after 30% power is reached. At 25%power N-16 is then also declared Non-functional, as per the 3.4.5 TRM bases "RCSN-16 activity will be sufficient to support FUNCTIONALITY of the N-16 system withreactor power at or above 25% power. Second part is also correct, because based onN-16 trends, the rate of increase is > 5 GPD. Between 61% and 43% power the rate ofincrease was at 5 gpd and then between 43% and 26% the rate of increase was 6 gpd, so based on Note (b) of table 3.4.5-1 grab samples should be taken once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />,and the STA has confirmed the leakage rate trends are correct.

K/A:G2.3.15Knowledge of radiation monitoring systems, such as fixed radiation monitors andalarms, portable survey instruments, personnel monitoring equipment, etc.Technical

References:

TRM 3.4.5 "Primary to Secondary Leakage detection systems"TRM 3.4.5 Bases 1-OP-2.2 "Unit Power Operation From Mode 1 to Mode 2"References provided to applicants: TRM 3.4.5 "Primary to Secondary Leakage Detection Systems Page 3.4.5-1Page 3.4.5-2Page 3.4.5-4 (Table 3.4.5-1)NOTE: Page 3.4.5-3 TRM Surveillance Requirements WILL NOT BE INCLUDED Learning Objective:U 18002 Explain High Level Actions of 1-AP-5 "Unit-1 Radiation Monitoring System"Explain applicable TRMs U 5263List the means provided to the Control Room to determine abnormal conditions as they apply to the N-16 main steam radiation monitor.U 5264Explain Relationship between indicated Rx power and N-16 RM readingsWhy the N-16 MS RM indication is invalid below 25% Rx powerQuestion Source: NEWQuestion History: Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content:SRO only - 10 CFR 55.43 (b)(4)Radiation hazards that may arise during normal and abnormal situations, includingmaintenance activities and various contamination conditions.10 CFR 55.41 (a)(b)(12)

Radiological safety principles and procedures Comments:This question matches the K/A statement by requiring the SRO applicant to analysisand interpret Main Steam RM readings, including comparison to operational procedures and then regulatory limits, by understanding the bases behind the limits set applicant must recognize what kind of trend is associated with the N-16 radiationmonitors. Having a backup confirmation of that trend. Then applying TRM actions,based on unit conditions outlined in the bases, that support functionality of themonitoring equipment.

099 - G2.4.27 001/SRO/3/H/3.4/3.9/NEW//NOUnit-1 is at 100% power when a fire occurs in the Unit-1 Emergency Switchgear Room.The Control Room crew enters 1-FCA-2 "Emergency Switchgear Room Fire".

• The operating crew is ready to establish RCS cooldown.

• H emergency bus is operable.* J emergency bus is de-energized.* Two CRDM fans are available and running.

• Charging is aligned through the BIT.

• An operator has been dispatched to perform attachment 15 "Fuel Building and Mitigating Spurious Valve Operations" Which one of the following completes the statement below?

The operator assigned to establish communications with the control room from theremote monitoring panel in the fuel building shall have a minimum qualification as a_____(1)______ and the cooldown rate limit will be _____(2)______. (1) Licensed Operator (2) < 15 DEGF/HR (1) Non-Licensed operator who has completed step 7 (2) < 15 DEGF/HR (1) Licensed Operator (2) < 25 DEGF/HR (1) Non-Licensed operator who has completed step 7 (2) < 25 DEGF/HR A.B.C.D.

Distractor Analysis:This requires the examinee to have knowledge of "fire in the plant" procedures.A. CORRECT (a) Licensed Operator (b) < 15 DEGF/HR(a) Is correct because 1-FCA-2 procedure require the assignment of a LicensedOperator for performing Remote Monitoring operations. (b) is correct, the RCScooldown limits are posted frequently through out the cooldown sequence and also asa CAUTION of attachment 15 in 1-FCA-2 "Emergency Switchgear Room". 1) With three CRDM fans running cooldown rate is limited to < 25 DEGF/HR 2) With less thanthree CRDM fans running cooldown rate is limited to < 15 DEGF/HR. B. INCORRECT (a) Non-Licensed operator who has completed step 7 (b) < 15 DEGF/HR(a) Is incorrect but plausible if the candidate is not sure of the procedure qualificationrequirements, it would be very likely the candidate would consider a Non-Licensedoperator who has completed step 7 as having the qualifications since the operator isnot operating equipment and only providing indicated readouts to the control room and also the Fuel Building is a watchstation of a Non-Licensed operators who hascompleted step 7. Also, a watchstation qualified operator is assigned to perform actionswithin other FCA series procedures/attachments, example is controlling S/G levelslocally in 0-FCA-1 and if remote control of PORV's is not available they can be locallyoperator by a operator qualified for the watchstation but with constant communicationand direction from the MCR such as in attachment 6 in 1-FCA-2. Attachment 15 is theonly attachment in 1-FCA-2 out of 17 attachments that requires actions to be performed by a Licensed operator (b) is correct for the questions stated condition that only twoCRDM fans are available and running so the cooldown graph that would be chosen foruse would be limited to <15 DEGF/HR which is explained by a CAUTION in attachment15 and posted frequently through out the cooldown sequence of 1-FCA-2. C. INCORRECT (a) Licensed Operator (b) < 25 DEGF/HR (a) Is correct, 1-FCA-2 procedure requires the assignment of a Licensed Operator forperforming remote monitoring operations. (b) is incorrect but plausible if candidate is not sure of FCA requirements vs EOP procedure requirements to determine whichcooldown graph to use based on CRDM fans in service, of either < 25 or < 15DEGF/HR. D. INCORRECT (a) Non-Licensed operator who has completed step 7 (b) < 25 DEGF/HR (a) Is incorrect but plausible if the candidate is not sure of the procedure qualificationrequirements, it would be very likely the candidate would consider a Non-Licensedoperator who has completed step 7 as having the qualifications since the operator is not operating equipment and only providing indicated readouts to the control room andalso the Fuel Building is a watchstation of a Non-Licensed operators who hascompleted step 7. Also, a watchstation qualified operator is assigned to perform actionswithin other FCA series procedures/attachments, example is controlling S/G levelslocally in 0-FCA-1 and if remote control of PORV's is not available they can be locally operator by a operator qualified for the watchstation but with constant communicationand direction from the MCR such as in 1-FCA-3 attachment 10 (b) is incorrect butplausible if candidate is not sure of FCA requirements vs EOP procedure requirements,to determine which cooldown graph to use based on CRDM fans in service, of either <25 or < 15 DEGF/HR.

K/A:099 - G2.4.27Technical

References:

1-FCA-2 "Emergency Switchgear Room Fire" References provided to applicants: None Learning Objective:U 13906Determine the maximum Reactor Coolant System cooldown rate allowed whileresponding to a fire in the emergency switchgear room in accordance with 1-FCA-2U 9113Explain the qualifications of the operator assigned to conduct remote monitoringoperations in the fuel building. Question Source: New Question History:

Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content:SRO only - 10 CFR-55.43 (b)(5)Assessment of facility conditions and selection of appropriate procedures duringnormal, abnormal, and emergency situations. 10 CFR-55.41 (a)(b)(10)Administrative, normal, abnormal, and emergency operating procedures for the facility. Comments:This question matches the K/A statement by requiring the SRO applicant to possess aworking knowledge of NAPs Fire contingency actions (FCA). To arrive at the correctanswer the SRO must be aware of the procedure administrative requirement to send aLicense operator to the fuel building, who in turn will provide remote monitoring to the operators in the MCR or Auxiliary shutdown panel. The SRO must also showknowledge of diagnostic steps and decision point to obtain the proper sub-procedure interms of selecting the proper cooldown graph.

100 - G2.4.30 001/SRO/3//2.7/4.1/BANK//Unit-2 is defueled with maintenance activities in progress in the Containment Building.A call comes into the control room that a worker has fallen from a scaffold platform incontainment and is unconscious.

• The crew enters 0-AP-51 "Personnel Injury - Operations Response"

• The injured worker has various lacerations and abrasions.

• The First Aid Team determines the worker must be transported to a hospital for treatment.

• HP survey of the fall area and the workers DAD, indicates no radiological overexposure exist.

• Some of the workers wounds are contaminated and initial decon efforts have been unsuccessful.Due to the nature of the injuries a decision is made to immediately transport the workerto the appropriate hospital.Assuming no news releases or notification to other government agencies are planned,a ___________ notification to the NRC is required in accordance with VPAP-2802,Notifications and Reports, and 10 CFR-50.72. REFERENCE PROVIDEDImmediate One hourFour hourEight hour A.B.C.D.

Distractor Analysis:The candidate is required to have the knowledge of events related to systemoperation/status that must be reported to internal organizations or external agencies,such as the State, the NRC, or the transmission system operator. INCORRECT A. ImmediateIncorrect but plausible if the candidate is unfamiliar with using VPAP-2802, and isdirected to section 6.3.2 which is plausible if the candidate uses the topic of Personneloverexposures under Radiation or Exposure Events or goes to section 6.28.2 (seeabove) also the candidate could be directed to section 6.3.2 using Miscellaneous Events or conditions under injuries. INCORRECT B. One hourIncorrect but plausible if the candidate is unfamiliar with using VPAP-2802, and isdirected to section 6.3.3, or goes to section 6.28.2 for internal (Dominion) notification.

This has a step requiring the Manager of Nuclear Operations or Operations Manageron Call to notify the Site Vice President or a director of a potentially media significantevent. This step is referenced in section 6.3.5.a.6 and in section 6.2.1(i) bullet forInjuries or accidental deaths or Ambulance transport of personnel to an off-site medicalfacility. INCORRECT C. Four hourIncorrect but plausible if the candidate is unfamiliar with using VPAP-2802, and isdirected to section 6.3.4 which could be required if a news release is planned CORRECT D. Eight hourThe candidate can use section 6.2 of VPAP-2802 "Non-Scheduled Notification andReports" and search for the matching criteria. In this case section 6.2.1(i)

"Miscellaneous Events or Conditions" (*) Transport of contaminated injured person, andwill be directed to step 6.3.5.a.6. Section 6.3.5 is for eight hour notifications; a - Assoon as practical, but within eight hours, the Shift Manager shall notify the NRCOperations Center via the ENS of; 6 - Any event requiring the transport of aradioactively contaminated person to an off-site medical facility for treatment. K/A:100 - G2.4.30Knowledge of events related to system operation/status that must be reported tointernal organizations or external agencies, such as the State, the NRC, or thetransmission system operator.

Technical

References:

VPAP-2802 "Notifications and Reports" References provided to applicants: VPAP-2802 "Notifications and Reports" Learning Objective:U 9391 Define immediate notification and Reportable as they apply to reportabilityrequirements (VPAP-2802)U 9390List Documents to be used and purpose of worksheet as it applies to reportabilityrequirements (VPAP-2802)U 9389Explain the process for notifying NRC during Non-emergency and emergency events(VPAP-2802)U 14323Given a copy of VPAP-2802 "Plant Reporting Requirement" evaluate a set of plantconditions associated with reportability requirements in light of the following issues(SRO) Question Source: NAPS Vision Bank Question History: Unit-2 is defueled with maintenance activities in progress in the Containment Building.During one of these activities, a worker falls form a scaffold ladder. The worker isrendered unconscious and sustains various lacerations and abrasions. The first aid team is dispatched and determines that the worker must be transported to a hospitalfor treatment. Based on HP survey of the fall area and the workers DAD, noradiological overexposure is suspected, however, it is found that some of the workerswounds are contaminated. Initial efforts to decontaminate the worker on site are notsuccessful further attempts are suspended. Due to the nature of the injuries, a decisionis made to immediately transport the worker to the appropriate hospital. The workerarrives at the hospital by ambulance within the next hour to receive further medical treatment. Assuming no news releases or notifications to other government agencies are planned_____________ notification to the NRC is required in accordance with 10CFR50.72A. an eight hour (correct) B. no C. a four hourD. a one hour Question Cognitive Level: Memory or Fundamental Knowledge10 CFR Part 55 Content:SRO only 10 CFR-55.43(b)(5)

Assessment of facility conditions and selection of appropriate procedures duringnormal, abnormal, and emergency situations. 10 CFR-55.41(a)(b)(10)Administrative, normal abnormal, and emergency operating procedures for the facility. Comments:This question matches the K/A statement by requiring the SRO applicant to have theknowledge of events related to system operation/status that must be reported to internal organizations or external agencies, such as the State, the NRC, or thetransmission system operator. To arrive at the correct answer the SRO applicant musthave the knowledge to correctly assess plant conditions, and interpret those conditions,to determine the proper reporting requirement. Also the candidate must demonstrateproficiency negotiating through VPAP-2803 "Notifications and Reports" to determine the reporting requirements and limits that needs to be applied.