ML15125A468

Initial Exam 2014-301 Draft SRO Written Exam

ML15125A468
Person / Time
Site:	Mcguire, McGuire
Issue date:	05/05/2015
From:	NRC/RGN-II
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Download: ML15125A468 (206)

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ILT-30 MNS SRO NRC Examination QUESTION 75 GEN2.4 2.4.30 GENERIC Emergency Procedures / Plan Emergency Procedures I Plan Knowledge of events related to system operationlstatus that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator. (CFR: 41.10/43.5 /45.11)

Given the following sequence of events on Unit 1:

0200 A LOCA occurs on the unit 0205 The OSM declared an Alert 0210 The OSM designates the WCC SRO as the Control Room Communicator and directs him to prepare the Event Notification Form (ENF) for his review 0220 The Control Room Communicator completes the ENF Which ONE (I) of the following indicates the LATEST allowable notification times required by RP-29 (NOTIFICATIONS TO OFFSITE AGENCIES FROM THE CONTROL ROOM)?

State & Counties NRC A. 0220 0300 B. 0220 0305 C. 0225 0300 D. 0225 0305 Thursday, May 29, 2014 Page 222 of 298

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ILT-30 MNS SRO NRC Examination QUESTION 75 75 General Discussion In accordance with RP-002 (ALERT), the state and counties must be notified within 15 minutes of the declaration of an event.

In accordance with RP-002, the NRC must be notified as soon as possible but, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of event declaration.

Answer A Discussion INCORRECT: See explanation above.

PLAUSIBLE:

The first part is correct.

For the second part of the question, if the applicant concludes that the required notification time to the NRC was one hour from time of discovery, they would determine that 0300 was the correct notification time.

Answer B Discussion CORRECT: See explanation above.

Answer C Discussion INCORRECT: See explanation above.

PLAUSIBLE:

For the first part of the question, if the applicant concludes that the notification time was 15 minutes from assignment of an Emergency Communicator they would conclude that 0225 was the correct notification time for the state and counties.

For the second part of the question, if the applicant concludes that the required notification time to the NRC was one hour from time of discovery, they would determine that 0300 was the correct notification time.

Answer D Discussion INCORRECT: See explanation above.

PLAUSIBLE:

For the first part of the question, if the applicant concludes that the notification time was 15 minutes from assignment of an Emergency Communicator they would conclude that 0225 was the correct notification time for the state and counties.

The second part is correct.

Basis for meeting the K The KIA is matched because it requires the applicant to have knowledge of the reporting requirements to offsite agencies and the NRC related to an event that has effected the status of the unit.

Basis for Hi Cog This is a higher cognitive level question because it requires more than one mental step.

First, the applicant must recall from memory the reporting time requirements to the states and counties (15 minutes) and the NRC (1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />).

Next, the applicant must analyze the sequence of events to determine when the reporting clock starts, add the appropriate reporting time requirement to the start time, and determine from times in each answer which one corresponds to the calculated time.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension BANK 2007 MNS NRC SRO Retake Examination NRC Q75 (Bank 3496)

Development References Student References Provided

References:

RP-029 (NOTIFICATIONS TO OFFSITE AGENCIES FROM THE CONTROL ROOM)

Learning Objectives:

OP-MC-EP-EMP Objectives 10 & 14 Thursday, May 29, 2014 Page 223 of 298

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ILT-30 MNS SRO NRC Examination QUESTION 75 7s GEN2.4 2.4.30 GENERIC Emergency Procedures / Plan Emergency Procedures / Plan Knowledge of events related to system operation/status that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator. (CFR: 41.10 / 43.5 / 45.11) 401-9 Comments: RemarkslStatus Thursday, May 29, 2014 Page 224 of 298

Q75

References:

• Potential Loss of Either Fuel Clad or Nuclear Coolant System and Loss of Any Other Barrier.

• If a Steam Generator TLibe leak is in progress arid the air ejectors are in service, it is considered to be a secondary side release in progress. This is considered to be a LOSS of the CONTAINMENT barrier in addition to the LOSS or POTENTIAL LOSS of the NUCLEAR COOLANT barrier. This results in tile classification forthis event being a Site Area Emergency.

• Failure of Reactor Protection System Instrumentation to complete or initiate an automatic reactor trip once a Reactor Protection System trip setpoint has been exceeded ATWS) and manual trip WAS NOT successful.

Notify same agencies as for notification of Unusual Event orAlert Activate TSO!OSC/EOF and conduct a Site Assembly.

D. General Emerency Events are in progress or have occurred which involve an actual or imminent sLibstantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that resUlts in an actLial loss of physical control of the facility.

Releases can be reasonably expected to exceed EPA Protective Action GLiideline exposure levels offsite for more than the immediate site area.

This EAL will be declared if the point total from the Fission Product Barrier Matrix is 1 1-13 points. Note that the matrix is only applicable in ?odes 1 4.

Examples:

• Loss of All Three Fission Barriers (13 points)

• Loss of Any Two Fission Barriers with the Potential Loss of the Third.

• Failure of Reactor Protection System Instrumentation to complete or initiate an automatic reactor trip once a Reactor Protection System trip setpoint has been exceeded (ATWS) and manual trip WAS NOT successful and there is indication of all extreme challenge to the ability to cool the core.

DUring General Emergency, TSC. OSC, and EOF activation required. Conduct site assembly and mandatory evacuation of non-essential personnel.

Recommend Protection Actions to Off-site Agencies per RP/O!A150001029 2.1.2 Initial Notifications Objective#14 Initial notifications to the State(s) and counties must be made within 15 minutes of the event declaration using the Emergency Notification form (ENF).

For an Lipgrade in classification prior to or while transmitting an initial message.

• The notification for the lesser eniergency classification must be made within 15 minutes of the lesser classification declaration time.

• The agencies must be informed that an Upgrade in classification will be coming

Q75

References:

• Department of Energy (DOE) - Savannah River.

• Federal Emergency Management Agency (FEMA) Offsite planning and support.

• Others INPO - Other utility contact&

• Oak Ridge REACTS HospitaL Of these, the NRC is the lead agency for power plant radiological emergencies Local Offsite Support agencies include:

• Gilead Volunteer Fire Department

• Cornelius Volunteer Fire Department

• MEDIC

• Carolinas Medical Center 2.5 Public AiertinglNotitication System 2.5.1 Notification Objective#12 The Emergency CoordinatorlEOF Director shall assure prompt notification of Federal, State, and Local off-site authorities. The Emergency Notification Form (ENF) is used to ensure the same information is communicated to the off-site agencies. The forms used at Duke Energy are generic forms for the states of North Carolina, South Carolina, Georgia, and Alabama They are intended for non-nuclear personnel, Le. 911 operators at each of the 6 county warning points. One to two hours after the initial messages are transmitted to the counties, the trained county emergency management personnel will have set up their Emergency Operation Centers (EOC), and they will start taking the messages. They have a little nuclear knowledge, but again, the ENF should be written far the lay person.

Timely and accurate completion off the ENF contributes to the NRC Performance indicators (Pt). Timely means the form is filled out and initial contact with the state and counties is made within 15 minutes of the EAL classification. Accurate means the ENF is filled out correctly. Refer to Attachment 5.1 for instructions on filling out the ENF using RP/O!N5700/029 (Notifications to Ofisite Agencies from the Control Room).

Duke Power must notify the NRC immediately but rio later than one (1) hour of declaring the emergency I The NRC Event Notification Worksheet is used when communicating with the NRC. A form similar to this is used by all nuclear facilities so that the information looks the same regardless of the plant. RPIOIAf5700IOIO, NRC Immediate Notification Requirements provides guidance on filling out the NRC Event Notification Worksheet. This is covered in lesson plan OP-MC-ADM-RPIQ.

2.5.2 Public Alerting Method A system of fixed sirens is used for alerting the population (resident and transient) within the Plume Exposure EPZ. These sirens are activated by the Counties. Route alerting supplements the sirens. This involves mobile sirenslpublic address units (police, fire and rescue) running specific routes within the EPZ to alert the populace.

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Q75 Parent Question:

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EXAM BANK QUESTION: 3496 MNS Unit I experienced a LOCA inskle containment at 0200.

Given the following sequence of events:

0200 LOCA starts 0205 The OSM dedared an Alert 0210 The OSM designated you as the Control Room Communicator and directed you to prepare the initial notification messages for his review.

0220 You complete fltng out the first notification form What are the maximum allowable notification times required by RP/0/N5700iV02?

States and Counties NRC A 0215 0300 B 0220 0305 C 0225 0300 ft 0225 0305 Monday, September 30. 2013 Page 30 of 9213

Q75 Parent Question:

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EXAM BANK QUESTION: 3496 MNS General Discussion Answer A Discussion Incorrect only 10 minnre!t from alert declaration time and less than I hourt for NRC Plauslhk 15 minuter and 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> from start of the ent An5Wer B Discussion Correct Answer Answer C Discussion Incorrect: 20 minutes and 55 minutes from time of alert declaration Plausible: if the candidate thinks that the 15 minute dock starts when he/she is told to complete the initial notidcation focus Answer [I Discussion Incorrect 20 minutes from time of aleil declaration Plausible: 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> NRC notication from the alert declaration is correct Basis formeeting the KA I I Basis for Hi Cog I I Basis for SRO only I I Job Level Cognitive Level QuestionType Question Source RD Comprehension BAI4K Development References Student References Provided KA IKA_desc GEN2.4 Emergency Procedures / PlnKnowia4ge ofRD responsibilities in emergency plan implementation. (CFR: 41.10/45,11) 2.4.39 Monday September 30, 2013 Page 531 of 9213

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ILT-30 MNS SRO NRC Examination QUESTION 76 L 76 SYSOO3 A2.O1 Reactor Coolant Pump System (RCPS)

Ability to (a) predict the impacts of the following malfunctions or operations on the RCPS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5 / 43.5/ 45.3 /45/13)

Problems with RCP seals, especially rates of seal leak-off Given the following conditions on Unit 1:

• The unit is at 100% RTP

• #1 Seal Leakoff on IA NC pump indicates 6.5 GPM

• AP-08 (MALFUNCTION OF NC PUMP) Case I (NC PUMP SEAL OR PUMP LOWER BEARING MALFUNCTION) has been implemented

• The crew has reached the steps in AP-08 to trip the Reactor and stop the IA NC pump Procedure Legend:

E-0 (REACTOR TRIP OR SAFETY INJECTION In accordance with AP-08, Enclosure 2 (NC PUMP POST TRIP ACTIONS FOR

1. 1 SEAL FAILURE),:

1) the actions contained in the enclosure must be performed within 3-5 minutes after stopping the 1A NC pump to prevent

2) the requirement to perform the enclosure actions is applicable Which ONE (1) of the following completes the statements above?

A. 1. damage to the IA NC pump #2 & #3 seals

2. even after transition from AP-08 to E-0 B. 1. damage to the 1A NC pump #2 & #3 seals

2. only while AP-08 is in effect C. I. the VCT from exceeding design temperature limits

2. even after transition from AP-08 to E-0 D. 1. the VCT from exceeding design temperature limits

2. only while AP-08 is in effect Thursday, May 29, 2014 Page 225 of 298

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ILT-30 MNS SRO NRC Examination QUESTION 76 76 General Discussion In accordance with AP-08, after the Reactor is tripped and the NC pump is stopped the seal return valve for the AFFECTED NC pump (1NV-34A) must be closed within 3-5 minutes. This action is contained in Enclosure 2 (NC Pump Post Trip Actions For #1 Seal Failure). In accordance with the AP-08 Background Document, the seal return line must be isolated to prevent damage to the #2 and #3 seals due to high temperature water flowing past the seals.

Per AP-08, the requirement to close 1NV-34A within 3-5 minutes after stopping the pump is applicable even after transition to the EPs.

Answer A Discussion

[CORRECT: See explanation above.

Answer B Discussion TNCORRECT: See explanation above.

PLAUSIBLE:

Part 1 is correct.

Part 2 is plausible because conditional steps in Aps are typically no longer applicable when transition is made to the Eps. In this particular case the applicant must recall the caution from AP-08 that states the post pump trip actions in the AP-08 enclosure are applicable even after transition to the Eps to arrive at the correct answer.

Answer C Discussion INCORRECT: See explanation above.

PLAUSIBLE:

Part 1 is plausible because this is the basis for closing 1NV-94AC and 95B (NC Pumps Seal Return Cont Isolations) in AP-08 when both seal injection and thermal barrier cooling are lost.

Part 2 is correct.

Answer D Discussion INCORRECT: See explanation above.

PLAUSIBLE:

Part 1 is plausible because this is the basis for closing INV-94AC and 95B (NC Pumps Seal Return Cont Isolations) in AP-08 when both seal injection and thermal barrier cooling are lost.

Part 2 is plausible because conditional steps in APs are typically no longer applicable when transition is made to the EPs. In this particular case the applicant must recall the caution from AP-08 that states the post pump trip actions in the AP-08 enclosure are applicable even after transition to the Eps to arrive at the correct answer.

Basis for meeting the K The applicant demonstrates the ability to predict the impact of the NC pump seal problem by demonstrating a knowledge of the consequences of not performing the actions of Enclosure 2 as directed. The applicant then demonstrates the ability to use procedures to correct, control, or mitigate the consequences of the malfunction by demonstrating the knowledge that the actions of the AP enclosure are still applicable after transition to E-0.

Basis for Hi Cog Basis for SRO only This question meets the following criteria for an SRO only question as described in the Clarification Guidance for SRO-only Questions (Rev 1 dated 03/11/2010) under the Screening Criteria for question linked to 10CFR55.43(b)(5) (Assessment and Selection of Procedures):

1) The question can NOT be answered by knowing systems knowledge alone. The basis for closing the seal return isolation valve for the affected pump within 3-5 minutes is not covered by the NCP system lesson plan. Therefore, this is not systems level knowledge.

2) The question can NOT be answered by knowing immediate Operator actions.

3) The question can NOT be answered by knowing AOP or EOP entry conditions.

4) The question can NOT be answered by knowing the purpose, overall sequence of events, or overall mitigative strategy of the procedure.

5) The question requires the applicant to recall procedure content from AP-08 (i.e. that the Post Pump Trip Actions must still be performed even after transition to the EPs). Additionally, the applicant must recall why the procedure steps must be performed from the AP-08 Background Document.

Thursday, May 29, 2014 Page 226 of 298

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ILT-30 MNS SRO NRC Examination QUESTION 76 Job Level Cognitive Level QuestionType Question Source SRO Memory BANK 2010 MNS NRC Exam NRC Q76 (Bank 2776)

Development References Student References Provided

References:

AP-08 (MALFUNCTION OF NC PUMP)

AP-08 Background Document Learning Objectives:

OP-MC-AP-008 Objective 4 SYSOO3 A2.01 Reactor Coolant Pump System (RCPS)

Ability to (a) predict the impacts of the following malfunctions or operations on the RCPS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5 / 43.5/ 45.3 / 45/13)

Problems with RCP seals, especially rates of seal leak-off 401-9 Comments: Remarks/Status Thursday, May 29, 2014 Page 227 of 298

Q76

References:

MNS MALFUNCTION OF NC PUMP PAGE NO.

AP111AJ5500108 Enclosure 2 Page 1 of 3 Rev 14 UNiT I NC Pump Post Trip Actions For #1 Seal Failure ACTION/EXPECTED RESPONSE 1 RESPONSE NOT OBTAINED CAUTION Failure of number two and three seals may occur unless the affected NC Pump Seal Return Valve is closed immediately after the pump has coasted down to zero speed (3-5 mm). This enclosure must be completed even after transition to EP5.I

1. Record time of NC Pump shutdown:

2. Check if seal cooling available to Perform the following:

affected pump as follows:

a. CLOSE the following valves:

. Seal injection established (Normal or SSF Supply) . INV-94AC (UI NC Pumps Seal Water Return Cent Inside 1501)

OR 1NV95B (UI NC Pumps Seal Water

• KC to thermal barrier established. Return Cont Outside Isol).

b. Exit this enclosure.

3. Check any NC Pump number one seal TO Step 5.

leakoff flow GREATER TNANOR EQUAL TO6GPM.

4. Maintain seal injection flow greater than 9 GPM to affected pump(s).

Q76

References:

RE F ER EN C ES:

T S. 3.5.5, Seal Injection Flow ENCLOSURE 2 STEPs 5&6 PURPOSE:

Provide a transition to the time-related specific guidance for isolating the affected N OP seal return which isolates the excessive Leakage of the failed #1 seal, transfers the primary pressure drop function to the #2 seal while attempting to minimize the damage to the #2 seaL DISCUSSION:

The NCP seal return valve needs to be closed as soon as the NCP is at zero speed to minimize the time that the #2 and #3 seals are exposed to potentially hot water and to limit the risk of debris damage to the #2 and #3 seals.I Per Westinghouse Tech Bulletin ESBU-TB-93-01 -Ri says [or immediate shutdown on seal failure: RCP shutdown arid isolation of No. I seal leakoff within 5 minutes of a high or[ow No.1 seal Ieakoff flow indication per Conditior is 1,3, or 5. (No.

1 seal leakoff isolation needs to occur promptly after the ROP has stopped for Conditions 1, 3, and 5. RCP coastdown is normally less than 2 minutes after tripping pump. Actual coastdown time is dependenton the plaritand numberof RCP5 operating. Assuming 3 minutes for coastdown, the No.1 seal leakoff is to be isolated between 3 and 5 minutes after tripping the RCP. The plant data documented in PIP M09-1 857 supports: a tripped NCP will be at zero speed in less than 3 minutes corisistent with the Tech Bulletin) as long as at least one other NCP is running (longest being 2 mm 5 sec when tripping one of two operating pumps) arid it can take 4 mm 20 seconds to coast to zero speed for the last pump stopped.

Waiting 3-5 minutes ensures the NCP has stopped rotating which should also help reduce the wear on the soft seated #2 seal face.

As discussed above: If any NGP continues to run, the affected NGPs seal return will be isolated when the NCP has been off for 3 minutes and 5 minutes it all NCPs are off The thermal barrier isolations are re-opened if necessary to establish KG as the backup seal package cooling method. This step is last to ensure any of the previous steps transients are completed so no remaining perturbations will reclose the valves after they are opened.

REFER E N C ES:

Westinghouse Tech Bulletin ES B U-TB-93-0 i-Ri PIP M-09-i 857

Q76

References:

MNS MALFUNCTION OF NC PUMP PAGE NO.

APIIIN55OCWO8 Case 5 of 26 UNIT 1 Rev. 14 NC Pump Seal or Pump Lower Bearing Malfunction ACT1ONJEXPECTED RESPONSE RESPONSE NOT OBTAINED Perform the followinq:

a,

/

b. jEINIiMEsealcoolinqis restored, THEN observe Note priorto Step 7 and Q IQ Step 7.

c. RETURN TO Step 2.

liQII Up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of NC Pump operation may be required before seals seat and operate normally after seal maintenance or startup.

7. Check any NC Pump number one seal Perform the followinq:

leakoff GREATER THAN OR EQUAL TO 6GPM.I NOTE 0P111N6200!OOIB (Chemical and Volume Control System Charging), Enclosure 4.10 (Maintaining NC Pump Seal Leakoff) gives guidance on actions used to change seal leakoff flow.

a. jEsealleakoffslowlyqoingup,JJj contact station management for further guidance.

b. Continue to monitor NC Pump seal leakoff flow.

c. lFAI.ff))ME seal leakoff flow goes up to 6 GPM, TI-lEN O TO Step 8.

d. QTOStep9.

Q76

References:

McGuire time critical action, but is a management expectation and prudent action to prevent damage to the 112 and#3 seals. PIP M-07-0310 ACA#4 documentsthe removal of this action from McGuires time critical action list.

There are different requirements for tripping a NC Pump in Mode 3, 4, or 5 than there is in Mode 1 or 2. The NC Pump is tripped per the guidance in the RNO if in Mode 3, 4, or 5. if in Mode 1 012, different guidance is given in the A/ER column to trip the NC Ps.

Guidance is given to wait until reactor power is less than 5% before stopping the NC pump. This will ensure the NC pump will provide adequate flow/core cooling until reactor power is sufficiently low enough to preclude a challenge to fuel integrity.

After the NC Pump is stopped in the RNO, direction is given to secure any boron dilution if all NC Pumps are off. This will preclude any diluted pockets of water migrating to the core without sufficient mixing with the borated NC System water. If all NC Pumps are off with the Unit in Mode 3, direction is given to immediately open the reactor trip breakers in compliance with Tech Spec 3.4.5 action statement If opening the reactor trip breakers results in E-O entry conditions, direction is given to go there while continuing with thisAP as quickly as time allows, If the crew doesnt go to E-O after stopping the NC Pump, they are directed to continue on with this case.

If in Mode 1 or2, direction is given to trip the reactor first, and then trip the NC Pump. This may be a conservative action (see discussion at beginning of basis document). As they go to E-O, they are directed to continue on with this AP as quickly as time allows.

REFERENCES:

PIPs M-03-1 992 and M-.07-0310 Tech Spec3.4.5 CASE I STEP6:

step wil.,. o prevent transition k

• next seven - - seal failures, and back to monitoring for NCP trip criteria. This has the benefit of skipping the steps that would close the individual seal return isolation valves. The individual seal return isolation valves need to remain open for loss of all seal cooling events. The following is an excerpt from DW-94-O1 1:

Isolation of the #1 seal leakofi line during a loss of all seal cooling event would force the 112 NCP seal into the high pressure mode of operation at high temperature. This is beyond the design basis of the #2 seal, and the response of the #2 seal to high pressure operation without

Q76 Parent Question:

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2010 MNS SRO NRC Examination QUESTION 76 I SYSDOS 2.1.20 RctorCoo!aitPuiup Stm(CPS)

SYS003 GENERIC Ability to intrprt nde cute procedure tep (CFR: 4L10 43.5 4512)

Given the following conditions on Unit 1:

• The unit was initially at 100% RTP

• #1 Seal Leakoff on 1 A NC pump indicates 6.5 GPM

• AP-08 (Malfunction of NC Pump) Case I (NC Pump Seal or Pump Lower Bearing Malfunction has been implemented

• The crew has reached the steps in AP-08 to trip the Reactor and stop the iANCpump In accordance with AP-08. Enclosure 2 (NC Pump Post Trip Actions For #1 Seal Failure) must be performed within 35 minutes after stopping the IA NC pump to prevent (1) . The requirement to perform these actions is applicable (2) -

Which ONE (1) of the following completes the statement above?

A. 1. damage to the 1A NC pump #2 & #3 seals 2, only while AP-08 is in effect B. 1. damage to the IA NC pump #2&#3 seals

2. even after transition from AP-08 to E-0 C. 1. the VCT from exceeding design temperature limits

2. only while AP-08 is in effect

0. 1. the VCT from exceeding design temperature limits

2. even after transition from AP-08 to E-0 Tuesday, September 14, 2010 Page 221 of 29

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2010 MNS SRO NRC Examination QUESTION 76 General Discussion n accordance with AP-OS. after the Reactor is tripped and the NC pump is stopped the seal return valve for the ATFECT3D NC pump (1NV-34A) insist be closed within 3-5 minutes. This action is contained in Enclosure 2 (NC Pump Post Trip Actions For l Seal Failure). Itt accordance with the AP-08 BackoundDocument, the seal return line mustbe isolated to prevent damage to the 2 and 3 seals due to high temperature water dotting past the seats.

er AP-OS, the requirement to close lNV-34A within 3-5 minutes after stopping the pump is applicable even after transition to the EPs.

1 Answer A Discussion NCORRECT: See ezplanation above,

>LAUSLE: Part 1 is correct.

art 2 is plausible because conditional steps in APt are typicallyno longer applicable when transition is made to the EP5, In this particular case he asiplicant insist recall the caution toni AP-G3 that states the post pump trip actions in the AP-0S enclosure are applicable even after transition o the EPs to arrive at the correct answer.

nswer B Discussion

[CORRECT: See erplanadon above.

Answer C Discussion NCORRECT: See ercplrnrtion above, LAUSLE: Part lit. plausible because this is the basis for closing 1NV-P4AC and P53 (NC Pumps Seal Return Cont Isolations) in AP-OS then bath seal injection and thennal barrier cooling are lost.

art 2 5 Plausible because conditional steps mAPs are typically no longer applicable when transition is made to the EP5. hi this particular case the applicant muse recall the caution froniAP$S that states thepost pump trip actions in the AP-OS enclosure are applicable even after transition o the EPs to arrive at the correct answer.

Answer D Discussion NCORRECT: See explanation above.

PLAUSLE: Part I is plausible because this is The basis for closing 1NV-P4AC and P53 (NC Pumps Seal Return Cone isolations) inAP-dS then bath seal injection and thermal barrier cooling are lost.

lan 2 is correct.

Basis for meeting the KA applicant demonstrates the ability to interpret procedure steps by demonstrating a knowledge of basis for performing the Post Pump Trip ctions of Enclosure 2 (specifically closing the NC pump seal return valves within 3-5 minutes). The applicant demonstrates the ability to xecute procedure steps by demonstrating the knowledge that the Al procedure steps must be performed even after transition to the EP5.

Basis for Hi Cog Basis for SRO only his question meets the following criteria for an SilO only question as described in the Clarification Guidance for SRO-onlv Questions (Rev I ated 03 11 20l01 under the Screenins Criteria fr question linked to 10CFR5S.43(b)(5) (Assessment and Selection of Procedures):

1) The question can NOT be answered by knowing systems knowledge alone. The basis for closing the seal return isolation valve for the affected ump within 3-5 minutes is not covered by the NCP system lesson plan. Therefore, this is not systems level knowledge.

2) The question can NOT be answered by knowing iuunediate Operator actions.

3) The question can NOT be answered by knotting AOP cc FOP entry conditions,

) The question can NOT be answered by knowing the purpose, overall sequence of events, or overall mitigative strategy of the procedure.

5) The question requires the applicant to recall procedure content from AP-dS (.e. that the Post Pump Trip Actions must still be performed even after transition to the EP5). Additionally. the applicant must recall why the procedure steps mint beperfbnned from the AP.-0S Background lacunient, Job Level Cognitive Level QuestionType Question Source SRO Memory NEW Tuerday, September 14, 2010 Page 222 of 29

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ILT-30 MNS SRO NRC Examination QUESTION 77 77 SYSO22 2.1.28 Containment Cooling System (CCS)

SYSO22 GENERIC Knowledge of the purpose and function of major system components and controls. (CFR: 41.7)

In accordance with Tech Spec 3.6.6 (CONTAINMENT SPRAY SYSTEM) Basis,:

1) the Containment Spray System is designed to work in conjunction with the Ice Condensers to limit peak Containment Pressure and Temperature during a Design-Basis AND

2) a Design-Basis LOCA will result in peak Containment Which ONE (1) of the following completes the statements above?

A. 1. LOCA OR Steam Line Break

2. pressure B. 1. LOCA OR Steam Line Break

2. temperature C. 1. LOCA AND Steam Line Break occurring simultaneously

2. pressure D. 1. LOCA AND Steam Line Break occurring simultaneously

2. temperature Thursday, May 29, 2014 Page 228 of 298

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ILT-30 MNS SRO NRC Examination QUESTION 77 77 General Discussion In accordance with the Containment Spray System Design Basis:

The Containment Spray System provides containment atmosphere cooling to limit post accident pressure and temperature in containment to less than the design values.

The Containment Spray System is by definition an integral part of the Containment Cooling System.

In accordance with TS 3.6.6 (Containment Spray System) basis, the Containment Spray System is designed to limit containment pressure during a DBA LOCA or SLB. However, no two DBAs are assumed to occur simultaneously or consecutively.

In accordance with TS 3.6.6 Basis, peak Containment pressure is reached on a Design-Basis LOCA and peak Containment temperature is reached on a Design-Basis Steam Break Accident.

Answer A Discussion CORRECT. See explanation above.

Answer B Discussion INCORRECT: See explanation above.

PLAUSIBLE:

Part 1 is correct.

Part 2 is plausible because peak containment temperature occurs during a DBA SLB.

Answer C Discussion INCORRECT: See explanation above.

PLAUSIBLE:

Part I is plausible because both DBAs are considered during the analysis. However, they are considered to occur separately, not simultaneously or consecutively.

Part 2 is correct.

Answer D Discussion INCORRECT: See explanation above.

PLAUSIBLE:

Part 1 is plausible because both DBAs are considered during the analysis. However, they are considered to occur separately, not simultaneously or consecutively.

Part 2 is plausible because peak containment temperature occurs during a DBA SLB.

Basis for meeting the K The K/A is matched because the applicant demonstrates a knowledge of the function of the purpose and function of the Containment Cooling System (i.e. Containment Spray System) by demonstrating a knowledge of the Containment Spray System Design-Basis capability.

Basis for Hi Cog Basis for SRO only This question meets the following criteria for an SRO only question as described in the Clarification Guidance for SRO-only Questions Rev 1 dated 03/11/2010 for screening questions linked to 10CFR55.43(b)(2) (Tech Specs):

1) This question can NOT be answered by knowing less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Tech Specs The question does not relate to less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Tech Specs

2) This question can NOT be answered by knowing information listed above-the-line.

This is not related to above the line knowledge in Tech Specs.

3) This question can NOT be answered by knowing the TS Safety Limits or their bases.

This question is not related to Tech Spec Safety Limits.

4) This question requires the applicant to have knowledge of the Tech Spec Basis. Specifically, it requires the applicant to have knowledge of the Design-Basis capability of the Containment Spray System related to the occurrence of specific Design-Basis accidents. It is therefore SRO-only knowledge.

Thursday, May 29, 2014 Page 229 of 298

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ILT-30 MNS SRO NRC Examination QUESTION 77 77 Job Level Cognitive Level Questionrype Question Source SRO Memory BANK 2010 MNS SRO Audit Examin Q90 (Bank 2990)

Development References Student References Provided

References:

Tech Spec 3.6.6 (CONTAINMENT SPRAY SYSTEM) Basis Learning Objectives:

OP-MC-ECC-NS Objective 13 SYSO22 2.1.28 Containment Cooling System (CCS)

SYS022 GENERIC Knowledge of the purpose and function of major system components and controls. (CFR: 41.7) 401-9 Comments: RemarkslStatus Thursday, May 29, 2014 Page 230 of 298

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References:

Containment Spray System B 3.6.6 B 3.6 CONTAINMENT SYSTEMS B 366 Containment Spray System BASES BACKGROUND The Containment Spray System provides containment atmosphere cooling to limit post accident pressure ard temperature in containment to less than the design values Reduction of containment pressure and the iodine removal capability of the spray reduce the release of fission product radioactivity from containment to the environment, in the event of a Design Basis Accident (DBA). The Containment Spray System is designed to meet the requirements of 10 CFR 50 Appendix A GDC 38 Containment Heat Removal. GDC 39, Inspection of Containment Heat Removal Systems.

GDC 40, Testing of Containment Heat Removal Systems GDC 41, Containment Atmosphere Cleanup, GDC 42. Inspection of Containment Atmosphere Cleanup Systems, and GDC 43, Testing of Containment Atmosphere Cleanup Systems (Ret 1)

The Containment Spray System consists of two separate trains of equal capacity, each capable of meeting the system design basis spray coverage.

Each train includes a conitairinierit spray pump, one containment spray heat exchanger, spray headers noles. valves, arid piping. Each train is powered from a separate Engineered Safety Feature (ESF) bus. One train of Containment Spray flow is manually initiated with sLiction on the Containment Sump after commencement ot the ECCS sump recirculation mode of operation The diversion of a portion of tlie recirculation flow froni each train of the Residual Heat Removal (RHR) System to additional redundant spray headers completes the Containment Spray System heat removal capability.

Each RHR train is capable of supplying spray coverage, if desired, to supplement the Containment Spray System.

The Containment Spray System provides a spray of cold or subcooled borated water into the upper containment volume to limit the containment pressure and temperature during a DBA. In the recirculation mode of operation. heat is removed from the containment sump water by the Containment Spray System and RH R heat exchangers Each train of the Containment Spray System provides adequate spray coverage to meet the system design requirements for containment heat removaL MeGune Unit I and 2 b2661

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References:

Ccntinment Surav System B 366 BASES BACKGROUND (continued)

For the hypothetical double-ended rupture of a Reactor Coolant System pipe, the pH of the sump solution (and, consequently, the spray solution) is raised to approximately 79 within one hour of the onset of the LOCA. The resultant pH of the sump solution is based on the mixing of the RCS fluids, ECCS injection fluid, and the melted ice which are combined in the sump.

The alkaline pH of the containment sump water minimizes the evolution of iodine and the occurrence of chloride and caustic stress corrosion on mechanical systems and components exposed to the fluid.

Containment Spray is manually initiated from the Control Room by opening the Containment Spray System (CSS) Pump discharge valves and starting the CSS Pump. The CSS is typically not activated until an RWST Low-Low level alarm is received. This alarm signals the operator to manually align the ECCS to the recirculation mode and manually initiate containment spray. The CSS maintains an equilibrium temperature between the containment atmosphere and the recirculated sump water. Operation of the CSS in the recirculation mode is controlled by the operator in accordance with emergency operation procedures.

The RHR spray operation is initiated manually, when required by the emergency operating procedures, after the Emergency Core Cooling System (ECCS) is operating in the recirculation mode. The RHR sprays are available to supplement the Containment Spray System, if desired, in limiting containment pressure. This additional spray capacity would typically be used after the ice bed has been depleted and in the event that containment pressure rises above a predetermined limit. The Containment Spray System is an ESF system. it is designed to ensure that the heat removal capability required during the post accident period can be attained.

The operation of the Containment Spray System, together with the ice condenser is adequate to assure pressure suppression subsequent to the initial jgfjJ of steam arid water from a DRA. During the post period, the Air Return System (ARS) is automatically started.

The ARS returns upper compartment air through the divider barrier to the lower compartment. This serves to equalize pressures in containment and to continue circulating heated air and steam through the ice condenser, where heat is removed by the remaining ice.

McGure Unit 1 and 2 B 366-2 Revision No. 122

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References:

Conta.nm:ent Spray System B 366 BACKGROUND (continued)

The Containment Spray System hrnits the temperature and pressure that could be expected following a DBA. Protection of containment integrity limits leakage of fission product radioactivity from containment to the environment APPLICABLE The limiting DBAs considered relative to containment OPERABILITY SAFETY ANALYSES are the loss of coolant accident (LOCA) and the steam line break (SLB)

The DBA LOCA and SLB are analyzed using computer codes designed to predict the resultant containment pressure and temperature transients. No two DBAs are assumed to occur simultaneously or consecutively. The postulated DBAs are analyzed, in regard to containment ESF systems, assuming the loss of one ESF bus, which is the worst case single active failure, resulting in one train of the Containment Spray System the RHR System, and the ARS being rendered inoperable (Ref 2).

of the containment environmental qu temperature is to ensure the OPERABILITY of safety related equipment inside containment (Ret. 3).

The Containment Spray System actuation modeled in the containment analysis is based on the time associated with reaching the RWST Low Level Setpoint and operator action prior to achieving full flow through the containment spray nozzles. A delayed response time initiation provides conservative analyses of peak calculated containment temperature and pressure responses. The Containment Spray System total response time is composed of operator action, system startup time, and time for the piping to fill.

For certain aspects of transient accident analyses, maximizing the calculated containment pressure is not conservative. In particular, the ECCS cooling effectiveness during the core Jjj phase of a LOCA analysis increases with increasing containment backpressure. For these calculations, the containment backpressure is calculated in a manner designed to conservatively minimize, rather than maximize, the calculated transient containment pressures in accordance with 10 CFR 50, Appendix K (Ref. 4).

Revision No, 122

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References:

,IO INTRODUCTION Operator Fundamental Focus; Knowledge Explain to the class that the material presented in this lesson p1 the Licensed Operato?s ability to perform in theirassigned role Fundamentals by increasing his or her knowledge of the NS sy including establishing an understanding of why and how the sy understanding system and component design, and understandi into an overall Integrated Plant design.

11. Purpose Obiective #1 The Containment Spray System is an engineered safeguard feature which serves to remove thermal energy from the containrrnt in the event of a Loss-Of-Coolant Accident (LOCA). It performs this function in conjunction with the Emergency Core Cooling System (ECCS, which the reactor by direct injection. The heat removal capability of the spray system keeps the Containment pressure below desig pressure of 15 psig after all the ice has melted (zI hour). while steam generation in the core continues to enter the Containment. The NS System also serves to remove fission product iodine from the post-accident Containment atmosphere.

12. General Description The Containment Spray System (NS) consists of two spray pumps and two spray heat exchangers in parallel, with associated piping, valves and spray headers per unit. These spray headers are located in the upper Containment volume. The system is supplemented with two Residual Heat Removal System (ND) pumps and heat exchangers in parallel, th associated piping, valves and individual spray headers perunit. These spray headers are also located in the upperCantainment volume.

One spray system is definedas one spray pumpwith spray heat exchanger and partial flow from one RHR pump with heat exchanger.

Adequate Containment cooling is provided by one spray system operating in the following modes:

I Recirculation of water from the Containment sump by the CantainmentSpray Pump through the spray heat exchanger and back to the Containment afterthe Refueling Water Storage Tank has been drained and the Unit is in cold leg recirculation and containment pressure is greaterthan 3 psig This spray is useful in reducingthe temperature of the waterthat has been collected in the lowercompartment ofthe Containment.

2. Diversion of a portion of the recirculation flow from the Residual Heat Removal System to additional spray headers completes the spray system heat removal capability. This operation is performed manually by the operatorand occurs afterSafety Injection Cold Leg Recirculation Switchover has been completed and Containment pressure has reached the containment vessel design pressure setpoint.

FOR TRAiFtIG PURPOSES OPLY Page cf

Q77 Parent Question:

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2010 MNS SRO Audit Examination QUESTION 90 SYSD26 2.2.42 -Continjnet SpraSvtm(cSS; SYSO25GENPJC Abthicoiz 1vi coLn forTchnic1 CFR. 4i 4]JG 45.2 45.3 4.3)

Per Tech Spec 3.&6 (Containment Spray System) basis, the Containment Spray System is designed to limit peak Containment Pressure and Temperature during a Design-Basis (1)

Peak Containment Pressure will occur during a Design-Basis f2 Which ONE (1) of the following completes the statements above?

A 1. LOCAQ Steam Line Break

2. Steam Line Break B. 1. LOCA OR Steam Line Break

2. LOCA C. 1. LOCA AND Steam Line Break occurring snultaneously

2. Steam Line Break D. 1. LOCA Steam Line Break occurring simultaneously

2. LOCA FridayOctber 1. 2010 Page 196 Gf 220

Q77 Parent Question:

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2010 MNS SRO Audit Examination QUESTION 90 24P01 General Discussion Itu accordance with TS 3.d,d (Containment Spray System) basis, the Containment Spray System is designed to limit cont muent pressure during Ia DBA LOCA or SLB. However, no txo OBAs are assumed to occur simultaneously or con ecutively, Answer A Discussion NCORRECT: See explanation above.

L4USLE: Part 1 is correct Part 2 is plausible becausepeak containment temperature occurs during a DBA SLB.

Answer B Discussion See explanation above.

Answer C Discussion NCORRECT: See exnlanatioa. above, L4US1BLE: Part I is plausible because both DBAs are considered duriur the analysis. However, they are considered to occur separately. not siniulaneously or consecutively.

art 2 is plausible because peak containment temperature occurs during a DBA SLB.

Answer D Discussion NCORRECT: See explanation above.

LAUSLE: Part 1 is plausible because both DBAs are considered during the analysis. However, they are considered to occur separately, not simutaaeously or con.secsstively.

lart 2 is cosrect.

Basis for meeting the KA Basis for Hi Cog Basis for SRO only Job Level Cognitive Level Questioniype Question Source SRO Memo NEW Development References Student References Provided SYSO26 2.2.42 containment Spray System (CSS) 5Y5026 GENERIC Ability to recognize systeniparameters that are entry-level conditions for Technical Specifications. (CFR 41.7 41.10 43,2 43.3 45.3) 01-9 Comments: RemarkslStatus Erida October 1& 2010 Page 197 of 220

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ILT-30 MNS SRO NRC Examination QUESTION 78 78 SYSO26 2.4.31 Containment Spray System (CSS)

SYSO26 GENERIC Knowledge of annunciator alarms, indications, or response procedures. (CFR: 41.10 / 45.3)

Given the following initial conditions on Unit I:

• A LOCA has occurred inside Containment

• Containment pressure is Ii PSIG

• ES-I .3 (TRANSFER TO COLD LEG RECIRC) has been implemented

• The IA NS pump is running Subsequently, the 1A NS pump trips.

PROCEDURE LEGEND:

FR-Z.l (RESPONSE TO HIGH CONTAINMENT PRESSURE)

Based on the conditions above,:

I) the current condition of the Containment Critical Safety Function is

2) the FIRST required action would be to attempt to restore NS Which ONE (1) of the following completes the statements above?

A. I. RED

2. while remaining in ES-I .3 B. 1. ORANGE

2. while remaining in ES-I .3 C. I. RED

2. after transition to FR-Z.I D. I. ORANGE

2. after transition to FR-Z.I Thursday, May 29, 2014 Page 231 of 298

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ILT-30 MNS SRO NRC Examination QUESTION 78 78 General Discussion In accordance with ES-I .3 (TRANSFER TO COLD LEG RECTRC), a failure of NS with Containment pressure greater than 3 PSIG will cause a Containment orange path. However, ES-i .3 directs the operators to attempt to restore NS by performing actions in ES-I .3 prior to implementing FR-Z.i.

Answer A Discussion INCORRECT: See explanation above.

PLAUSIBLE:

The first part is plausible if the applicant confuses the Containment Critical Safety Function for a RED and ORANGE path. They could conclude that a loss of NS with Containment pressure greater than 3 PSIG would result in a RED path instead of an ORANGE path. This is plausible since the Containment CSF is one of the cases where both a RED path and ORANGE path lead to the same Functional Restoration Procedure.

The second part is correct.

Answer B Discussion

[CORiCT: See explanation above.

Answer C Discussion ftNCORRECT: See explanation above.

PLAUSIBLE:

The first part is plausible if the applicant confuses the Containment Critical Safety Function for a RED and ORANGE path. They could conclude that a loss of NS with Containment pressure greater than 3 PSIG would result in a RED path instead of an ORANGE path. This is plausible since the Containment CSF is one of the cases where both a RED path and ORANGE path lead to the same Functional Restoration Procedure.

The second part is plausible because in most cases if a RED or ORANGE path is received, transition to FR-Z.l would be the correct response.

And, if the actions performed in ES-l.3 were unsuccessful, transition to FRZ.i would be required.

Answer D Discussion INCORRECT: See explanation above.

PLAUSIBLE:

The first part is correct.

The second part is plausible because in most cases if a RED or ORANGE path is received, transition to FR-Z.i would be the correct response.

And, if the actions performed in ES-l.3 were unsuccessful, transition to FR-Z.l would be required.

Basis for meeting the K The KJA is matched because it requires the applicant to have knowledge what type of alarm is going to be received (i.e. RED or ORANGE path on Containment) and the correct procedural response based on those conditions.

Basis for Hi Cog This is a higher cognitive level question because it requires more than one mental step.

First, the applicant must analyze the conditions given to determine the effect on the Containment Critical Safety Function.

Next, the applicant must recall from memory the required procedural actions based on the conditions given.

Basis for SRO only This question meets the following criteria for an SRO only question as described in the Clarification Guidance for SRO-only Questions Rev 1 dated 03/11/2010 for screening questions linked to I 0CFR5 5.43(b)(5) (Assessment and selection of procedures):

1) The question can NOT be answered solely by knowing systems knowledge.

This question requires detail procedure knowledge and assessment of plant conditions related to procedure selection.

2) The question can NOT be answered by knowing immediate operator actions.

This question is related to knowledge of procedure selection criteria within the body of the procedure.

3) The question can NOT be answered solely by knowing entry conditions for AOP or direct entry conditions for EOPs.

These are not related to entry conditions for an EOP. This is related to differentiation between two procedures which would perform the same recovery actions and selection of the appropriate procedure based on plant conditions and specific procedural guidance.

4) The question can NOT be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of the procedure.

This is detailed knowledge of procedure step guidance for appropriate procedure selection.

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ILT-30 MNS SRO NRC Examination QUESTION 78 78

5) The question requires detailed knowledge of procedure content. It requires the applicant to assess plant conditions based on given information and select the appropriate procedure based on guidance within the body of the proceudre. Therefore, it is SRO knowledge.

Job Level Cognitive Level QuestionType Question Source SRO Comprehension NEW Development References Student References Provided

References:

ES-l.3 (TRANSFER TO COLD LEG RECIRC)

Learning Objectives:

EPE1006 SYSO26 2.4.31 Containment Spray System (CSS)

SYSO26 GENERIC Knowledge of annunciator alarms, indications, or response procedures. (CFR: 41.10 / 45.3) 401-9 Comments: RemarkslStatus Thursday, May 29, 2014 Page 233 of 298

Q78

References:

MNS CRITICAL SAFETY FUNCTION STATUS TREES PAGE NO.

EPJ1IN5000JF-O Containment Page 1 of I Rev5 UNITI 00 TO FR-Z .1 GOTO FR-Z .1 GO TO FR-Z 2 GO TO FR-Z .1 0010 FR-Z 2 GOTO CSF&kT

MNS TRANSFER TO COLD LEG RECIRC PAGE NO.

EPI1IAI5000IES-1.3 13 of 25 Rev. 27 1

UNIT ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

8. (Continued)

NOTE A failure of NS with containment pressure greater than 3 PSIG will cause a containment orange path. The following step should be performed pjto implementing FR-Z.1.

• h. IF AT ANY TIME NS flow lost, OR RN is lost to operating train, THEN start available NS pump as follows:

1) Ensure affected NS pump is off.

NOTE If the following steps clear the containment orange path, FR-Z. 1 does not require performance as an orange path unless previously implemented.

2) Check at least one of the following 2) Perform the following:

alarms LIT:

a) WHEN either 3 ft sump alarm is

. CONT SUMP LEVEL GREATER lit, THEN align and start other THAN 3 FT on 1AD-14 LIT - NS pump R Steps 8.d through 8g.

OR b) RETURN TO procedure and

. CONT SUMP LEVEL GREATER step in effect.

THAN 3 FT on 1AD-15 LIT. -

3) Check containment pressure - 3) if NS stopped due to pressure GREATER THAN 1 PSIG. below CPCS interlock, THEN perform the following:

a) WHEN containment pressure is greater than 1 PSIG, THEN align and start either NS pump PER Steps 8.d through 8.g.

b) RETURN TO procedure and step in effect.

4) Perform Steps 8.d through 8.g.

Q78

References:

MNS TRANSFER TO COLD LEG RECIRC PAGE NO.

EPI1IN5000IESI .3 9 of 25 UNIT 1 Rev. 27 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

8. Align NSforrecircasfoHows:

a. CLOSE the following valves:

I

. 1NS-20A (IA NS Pump Suction From FWST Isol)

• 1NS-3B (lB NS Pump Suction From FWST Isol b Check containment pressure - b. Perform the following:

GREATER THAN 3 P51G.

I

1) Wait up to 30 seconds for I NS-20A and INS-3B to close.

2> OPEN INS-18A(1A NS Pump Suction From Cent Sump Isol).

_3) OPEN INS-lB (lB NSPump Suction From Cent Sump Isol).

4) [FAT ANY TIME containment pressure goes above 3 PSIG THEN perform Step 8.

_5) QjStep9.

c. Check at least one of the following c. Perform the following:

alarms LIT:

I) WHEN either 3 jsump alarm is lit, I

. CONT SUMP LEVEL GREATER THEN perform Step 8.

THAN 3 Fi on 1AD-14 LIT

_2) IQStep9.

OR

• CONT SUMP LEVEL GREATER THAN 3 FT on 1AD-15 LIT.

d Check 1 A NS pump AVAILABLE TO Q Step 8 f RUN.

I

Q78

References:

MNS TRANSFER TO COLD LEG RECIRC PAGE NO.

EPI1 !A150001E5-1 3 ii of 25 Rev. 27 UNiT 1 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

8. (Continued)

f. Align B Train NS to containment sump as follows.

1) Check 1NI-1848 (lB ND Pump 1) iQStep8.g.

Suction From Cent Sump Isol) -

OPEN.

2> Check lA NS pump OFF. - 2) jf 1A NS pump is running, THEN IQStep 8.g.

_3) Check lB RN pump - QN 3) IQ Step 8.g.

OPEN 1NS-126 (16 NS HxOuttet 4) I Step 8.g.

Cent Outside Isol).

t4>_5) OPEN 1NS-15B (lB NS HxOutlet Cont Outside soD.

_5) §IQStep8.g.

6) Check 1NS-3B (lB NS Pump 6) IQStep8.g.

Suction From FWST lsol. -

CLOSED.

7) OPEN iNS-lB (lB NS Pump 7) QiQStep8.g.

Suction From Cent Sump Isol)

8) Wait up to 30 seconds for the 8) any valve remains closed or following valves to open: intermediate for over 30 seconds THENjQ Step8.g.

I

. 1NS-l2B

• 1NS-15B

• iNS-lB.

_9) Start lB NS pump. 9) QiQStep8.g.

10) OPEN 1RN-235B (B NS HX Inlet 10) Perform the following:

Isol) a) Stop 16 NS pump.

b) Q 1Q Step 8.g.

Ii) WHEN 1RN-235Bggto open 11) IF RN flow cannot be established THEN THROTTLE OPEN to 16 NS Hx, THEN stop lB NS 1 RN-238B (B NS Hx Outlet Isol) to pump.

establish 3600 GPM tel B NS Hx.

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ILT-30 MNS SRO NRC Examination QUESTION 79 79 SYSO59 A2. 12 Main Feedwater (MFW) System Ability to (a) predict the impacts of the following malfunctions or operations on the MFW; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5 /43.5 / 45.3 / 45.13)

Failure of feedwater regulating valves Given the following initial conditions on Unit 1:

• The unit is in MODE 3 preparing for a unit startup

• The OSM has asked you to evaluate the following valve stroke time results:

VALVE STROKE TIME (sec)

ICF-32AB (1A SIG CF CONTROL) 9.2 1CF-IO4AB (IAS/G CF CONTROL BYPASS) 8.2 ICE-i 26B CiA SIG CF TO CA NOZZLE BYPASS) 11.2 1 CF-23AB (1 B S/G CF CONTROL) 11.0 ICE-I O5AB (lB S/G CF CONTROL BYPASS) 8.3 ICF-127B (lB S/G CF TO CA NOZZLE BYPASS) 11.2 ICF-2OAB (1C S/G CF CONTROL) 8.2 1CF-IO6AB (1C S/G CF CONTROL BYPASS) 8.3 1CF-128B (1C S/G CF TO CA NOZZLE BYPASS) 8.1 1 CF-i 7AB (ID S/G CF CONTROL) 11.3 1CF-IO7AB (ID S/G CF CONTROL BYPASS) 8.5 I CF-i 29B (1 D S/G CF TO CA NOZZLE BYPASS) 8.3 Based on the conditions above, which ONE (1) of the following indicates the applicable Action Statements of Tech Spec 3.7.3 (MFIVs, MFCVs, MFCVs Bypass Valves, and MFW/AFW NBVs)?

REFERENCE PROVIDED A. DONLY B. B and C ONLY C. B and D ONLY D. B,CANDD Thursday, May 29, 2014 Page 234 of 298

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ILT-30 MNS SRO NRC Examination QUESTION 79 General Discussion In accordance with Tech Spec 3.7.3 (MFIV5, MFCVs, MFCVs Bypass Valves, and MFW/AFW NBV5):

Condition B is applicable for one or more inoperable CF Control Valves.

Condition C is applicable for one or more CF Control Bypass or CF/CA Nozzle Bypass Valves.

Condition D is applicable for two inoperable valves in the same flow path.

For the conditions given, Condition B applies to the 1D S/G flowpath, Condition B, C, and D apply to the lB S/G flowpath, Condition C applies to the 1A S/G flowpath.

Therefore, Conditions B, C, and D are applicable.

Answer A Discussion INCORRECT: See explanation above.

PLAUSIBLE:

It is plausible for the applicant to conclude that Condition D bounds Conditions B and C and that, regardless of the number of valves that are INOPERABLE, separate entries for Conditions B and C are not required.

Answer B Discussion INCORRECT: See explanation above.

PLAUSIBLE:

It is common for operators to overlook the fact that the CF/CA Nozzle Bypass valves are in in the same flowpath with the CF Control and CF Control Bypass valves. Therefore, it is plausible for the applicant to overlook that Condition D applies to the lB S/G Flowpath and determine that ONLY Conditions B and C apply.

Answer C Discussion INCORRECT: See explanation above.

PLAUSIBLE:

If the applicant overlooks that Condition C is applicable for IA S/G Flowpath and concludes that since Conditions B and C are bounded by Condition D on the lB S/G Flowpath (and consequently did not require separate entries), they would conclude that Condition B was applicable for the ID S/G Flowpath and that Condition D was applicable for the lB S/G Flowpath.

Answer D Discussion LCo11CT See explanation above.

Basis for meeting the K The K/A is matched because the SRO applicant demonstrates the impact of the feed reg valve (control valve) failure and the ability to use procedures (Tech Specs) to correct, control, or mitigate by demonstrating a knowledge of how to apply Tech Spec 3.7.3 to the conditions given.

Basis for Hi Cog This question is a higher cognitive level question because it requires detailed analysis and application of Tech Specs.

Basis for SRO only This question meets the following criteria for an SRO only question as described in the Clarification Guidance for SRO-only Questions Rev I dated 03/11/2010 for screening questions linked to IOCFR55.43(b)(2) (Tech Specs):

1) This question can NOT be answered by knowing less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Tech Specs The potential Tech Spec action times for this condition are either 8 or 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

2) This question can NOT be answered by knowing information listed above-the-line.

The question can only be answered by have knowledge of LOC actions and action times which are below-the-line.

3) This question can NOT be answered by knowing the TS Safety Limits or their bases.

This question is associated with Tech Spec 3.7.3 (MFIV5, MFCVs, MFCVs Bypass Valves, and MFW/AFW NBV5) and NOT Tech Spec Safety Limits (TS 2.0) or their bases.

4) This question requires the applicant to have knowledge of below-the-line TS LCO Actions that are greater than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. As such, it is SRO level knowledge.

Thursday, May 29, 2014 Page 235 of 298

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ILT-30 MNS SRO NRC Examination QUESTION 79 79 Job Level Cognitive Level QuestionType Question Source SRO Comprehension NEW Development References Student References Provided

References:

Tech Spec 3.7.3 (MFIVs, MFCVs, MFCVs Bypass Tech Spec 3.7.3 (MFIVs, MFCVs, MFCVs Bypass Valves, and MFW/AFW NBVs) Valves, and MFW/AFW NBVs)

Learning Objectives:

OP-MC-CF-CF Objective 28 SYSO59 A2. 12 Main Feedwater (MFW) System Ability to (a) predict the impacts of the following malfunctions or operations on the MFW; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5 /43.5 / 45.3 / 45.13)

Failure of feedwater regulating valves 401-9 Comments: RemarkslStatus Thursday, May 29, 2014 Page 236 of 298

Q79

References:

11HJs IFC s r iFc a B,jass Va! es and MF V NEs 373 3.7 PLANT SYSTEMS 3.7.3 Main Feedwater Isolation Valves (MFIVs), Main Feedwater Control Valves (MFCV5),

MFCVs Bypass Valves and Main Feedwater(MFW)to Auxiliary Feedwater(AFW)

Nozzle Bypass Valves (MFWAFW NBVs)

LCO 3.7.3 Four MFIVs, four MFCVs, four MFCVs bypass valves, and four MFW/AFW NBVs shall be OPERABLE.

APPLICABILITY: MODES 1,2, and3 exceptwhen MFIV, MFCV, MFCVs bypass valve or MFW!AFW NBV is closed and de-activated or isolated by a closed manual valve.

ACTIONS Separate Condition entry is allowed for each valve.

CONDITION REQUIRED ACTION COMPLETiON lIME A. One or A.1 Close or isolate MFIV. 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> more MFIVs AND inoperable.

A.2 Verify MFIVis closed or Once per isolated. 7 days B. One or B.1 Close or isolate MFCV. 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> more MFCVs AND inoperable.

B.2 Verify MFCVis dosed or Once per isolated. 7 days (continuec rdcGuire Unis I and 2

Q79

References:

MFI7s FC* riFC:: 1 n

373 ACTIONS (ccntinued CONDITION REQUIRED ACTION COMPLETION TIME C. One or Cl Close orisolate MFCVs 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> more bypass valve or MFWAFW MFC\Ps NBV bypass valves or AND MFWIAFW NBVs C 2 Verify MFCVs bypass Once per inoperable valve orMFW!AFWNBV is 7 days closed or isolated.

D Twovalvesinthesame Dl Isolateaffectedflavpath 8hours flow path inoperable..

E Required Action and El Be in MODE 3, 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met AND E2 BeinMODE4 l2hours SURVEILLANCE REQUIREIvENTS SURVEILLANCE FREQUENCY SR 31.31 Verify the closure tirr of each MF[V, MFC, MFCVs In accordance with bypass valve and MFW!PFW NBV s 10 secoids on an the IDJ2LLt actual or simulated actuation signal. Testing Progr m LicCuire Unite I end 2 Amendment Noe. 184:163

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ILT-30 MNS SRO NRC Examination QUESTION 80 80 SYS062 A2. 10 AC Electrical Distribution System Ability to (a) predict the impacts of the following malfunctions or operations on the ac distribution system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5 / 43.5 / 45.3 / 45.13)

Effects of switching power supplies on instruments and controls Given the following on Unit I:

hri.iii

• Prior to transferring the effected bus to its backup power supply, Annunciator 1AD-7 / D3 (VCT ABNORMAL LEVEL) illuminates

• Actual VCT level is 14%

Related to the abnormal VCT level, the loss of power has (1)

In response to the abnormal VCT level, the crew will (2)

Which ONE (I) of the following completes the statements above?

Procedure Legend:

OP/11A16102/003 (DCS SYSTEM OPERATION)

AP-15 (LOSS OF VITAL OR AUX CONTROL POWER) Enclosure 31 (VCT LEVEL CONTROL)

Component Legend:

I NV-I 37A (UI NC FILTER OTLT TO VCT 3-WAY CONVERSION CNTRL)

HUT (HOLDUP TANK)

A. 1. disabled VCT AUTO Makeup AND caused I NV-I 37A to fail to he HUT

2. remove a VCT level channel from service using OP/I/A16102 /

B. 1. disabled VCT AUTO Makeup AND caused 1 NV-i 37A to fail to the HUT

2. swap NV pumps suction to the FWST using AP-15 Enclosure 31 C. i. disabled VCT AUTO AND MANUAL Makeup

2. remove a VCT level channel from service using OP/i /A161 02i D. 1. disabled VCT AUTO AND MANUAL Makeup

2. swap NV pumps suction to the FWST using AP-15 Enclosure 31 Thursday, May 29, 2014 Page 237 of 298

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ILT-30 MNS SRO NRC Examination QUESTION 80 8O General Discussion The loss of power to I EKVA has caused a failure of both AUTO and MANUAL Makeup capability which cannot be restored until I EKVA is re energized or transferred to a backup power source. Therefore, when I AD-7 / D3 (VCT ABNORMAL LEVEL) is received, the crew must procedurally swap the NV pumps suction to the FWST lAW AP-I 5 Enclosure 31. AP- 15 requires swapping NV pump suction from the VCT to the FWST anytime VCT level goes below 16.

Answer A Discussion INCORRECT. See explanation above.

PLAUSIBLE:

The first part is plausible if the applicant concludes that the loss of power has effected one of the VCT level channels. If so, this would result in a failure of AUTO Makeup and cause 1NV-137A to divert.

The second part is plausible because the correct response for a failed VCT level channel would be to remove the failed channel from service lAW the OP.

Answer B Discussion INCORRECT. See explanation above.

PLAUSIBLE:

The first part is plausible if the applicant concludes that the loss of power has effected one of the VCT level channels. If so, this would result in a failure of AUTO Makeup and cause 1NV-l37Ato divert.

The second part is correct.

Answer C Discussion INCORRECT. See explanation above.

PLAUSIBLE:

The first part iscorrect and therefore plausible.

The second part is plausible because the correct response for a failed VCT level channel would be to remove the failed channel from service lAW the OP.

Answer D Discussion CORRECT. See explanation above.

Basis for meeting the K The KA is matched because the applicant must predict the impacts of a loss of power to IEKVA while it is being transferred to IKRP and based on that prediction, determine the correct procedural action to correct the abnormal VCT level.

Basis for Hi Cog This is a higher cognitive level question because it requires more than one mental step.

First, the applicant must diagnose the given conditions to determine the cause of the abnormal VCT level condition.

Then from the diagnosis performed in the first step, the applicant must determine the correct procedure and action to correct the abnormal VCT level problem.

Basis for SRO only This question meets the following criteria for an SRO only question as described in the Clarification Guidance for SRO-only Questions Rev I dated 03/11/2010 for screening questions linked to IOCFR55.43(b)(5) (Assessment and selection of procedures):

1) The question can NOT be answered solely by knowing systems knowledge.

While there is a systems element to the first part of the question, the effect of losses of power on equipment or systems functions is not part of systems training. In this particular case, the only time applicants would exposed to the effect of a loss of EKVA on the Makeup system would be during a loss of EKVA during a simulator scenario where a NOTE in AP- 15 Step 6 RNO would indicate that Auto and manual VCT makeup capability is unavailable.

Even if the first part of the question is considered systems knowledge, the second part of the question is NOT systems knowledge.

2) The question can NOT be answered by knowing immediate operator actions.

There are NO immediate actions associated with AP- 15.

3) The question can NOT be answered solely by knowing entry conditions for AOP or direct entry conditions for EOPs.

This is NOT related to entry conditions for AP-15.

Thursday, May 29, 2014 Page 238 of 298

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ILT-30 MNS SRO NRC Examination QUESTION 80 80D

4) The question can NOT be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of the procedure.

This question is related to detailed procedure step knowledge.

5) The question requires detailed knowledge of procedure content. Therefore, it is SRO knowledge.

Job Level Cognitive Level QuestionType Question Source SRO Comprehension NEW Development References Student References Provided

References:

AP-l5 (Loss of Vital or Aux Control Power)

Learning Objectives:

NONE SYSO62 A2. 10 AC Electrical Distribution System Ability to (a) predict the impacts of the following malfunctions or operations on the ac distribution system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5 /43.5 / 45.3 /45.13)

Effects of switching power supplies on instruments and controls 401-9 Comments: RemarkslStatus Thursday, May 29, 2014 Page 239 of 298

Q80

References:

M NS LOSS OF VITAL OR AUX CONTROL POWER PAGE NO.

AP/1 IAf55O 0/15 6 of 279 Rev. 25 UNIT 1 RESPOSE RESSE

6. Check Vital AC panelboard IEKVA Perform the following:

energized as follows:

a. IF letdown isolated, THEN have e Top row of channel status lights - available operator restore letdown PER NORMAL. one of the following while continuing with this procedure

• LF letdown orifice valves have power, THEN establish normal letdown PER EPItA5000!G-1 (Generic Enclosures) Enclosure 1 Estabishing Normal Letdown).

OR

. IF unable to restore normal letdown, THEN establish excess letdown PER EPIIIAJ5000IG-1 (Generic Enclosures), Enclosure 2 (Establishing Excess Letdown).

NOTE Auto and manual VCT makeup capability is unavailable

b. AT ANY TIME VOT level goes below 16% (VCT ABNORMAL LEVEL alarm (IAD-7, D-3 low setpoint), THEN align NV pump suction to FWST PER Enclosure 31 (VCT Level Control).

Q80

References:

OP!IIA/6100/OlO H Annunciator Response For Panel 1AD7 Page 27 of 53 Nomenclature: VCT ABNORMAL LVL Window Setpoint: Higir 96%

• Low: 16%

Origin:

Probable Cause: High: 1NV437A (NC Filters 0th 3WayCntr1) malfunction and continuous operation of Reactor Coolant Makeup System Low:

• Makeup system NOT operating properly and 1NV-137A (NC Filters Otit 34Vay Catri) malfunction

• NC System Leak Automatic Action: L Automatic makeup controlled between 4154%

2, At 4% level:

• INV-221 (NV Pump Suction from FWST) opens

• INV-222 (NV Pump Suction from FWST) opens

• 1NV442 (VCT Outlet Isol) closes

• lNV441 (VCT Outlet Isol) closes Continue On Next Page Unit I

Q80

References:

Enclosure 4.1 Opl!Aj1O2iOO3 RemovingiReturniug a VCT Level Page 1 of4 Channel FromiTo Service I. Limits and Precautions Li This procedure is Reactivity management related because it controls activities that can affect core reactivity by changing NC System boron concentration. (R.M)

2. Initial Conditions ii One of the following conditions exists:

OR L&E to calibrate a VCT Level transmitter

3. Procedure El D

U Unit I

Q80

References:

Enclosure 41 OPII/AJ61021003 Removing/Retunaing a VCT Level Page 2 of 4 Channel From/To Senice D

D o

0 0

13 Remove A icr Level Channel For JAE Calibration 111 On DCS Boric Acid Blender graphic, perform the following:

0 11 Li Select 2XS for VCT Level L 3.3.L2 Determine which level transmitter NOT to be calibrated:

0 NVAA 5760 (Transmitter A) 0 NVAA 5761 (Transmitter B)

O 13.L3 Select the level transmitter NOT to be calibrated for VCT level input (Transmitter A or Transmitter B).

0 3.3.L4 Select DEV MRE INH1B1T to block the input.

o 33.L5 Check MRE BLOCKEiY lit (blinking red).

Unit I

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ILT-30 MNS SRO NRC Examination QUESTION 81 8l SYSOO2 A2.04 Reactor Coolant System (RCS)

Ability to (a) predict the impacts of the following malfunctions or operations on the RCS; and (b) based on those predictions. use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5 / 43.5 I 45.3 / 45.5)

Loss of heat sinks Given the following initial conditions on Unit 1:

• The unit was at 100% RTP

• The TD CA pump is out-of-service for maintenance Subsequently:

• The unit is manually tripped when both CF pumps trip

• Both MD CA pumps fail to automatically start and cannot be started manually

• From E-0 (REACTOR TRIP OR SAFETY INJECTION) the crew transitions to FR-Hi (RESPONSE TO LOSS OF SECONDARY HEAT SINK)

• SG WR level is 40% in all SGs In accordance with FR-H.1,:

1) NC system Feed and Bleed must be initiated within a MAXIMUM of minutes after reaching Feed and Bleed initiation criteria.

2) after Feed and Bleed is initiated, efforts to restore feedwater flow to the S/Gs be terminated.

Which ONE (1) of the following completes the statements above?

A. 1.8

2. MAY B. 1.4

2. MAY C. 1.8

2. MAY NOT D. 1.4

2. MAY NOT Thursday, May 29, 2014 Page 240 of 298

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ILT-30 MNS SRO NRC Examination QUESTION 81 81 General Discussion In accordance with FR-H. 1 Background Document:

Feed and bleed should be initiated as quickly as possible after meeting criteria in FR-H. I, but must be initiated within 4 minutes of meeting criteria. Note that feed and bleed may need to be initiated within 8 minutes after reactor trips on low-low SG level, since the criteria will be quickly met. The NV recirc valve must be closed within 5 minutes of initiating feed and bleed. Although this action is NOT required for a design basis event, this item is included due to its PRA significance during a loss of secondary heat sink event.

In accordance with FR-H. 1, effort to restore feedwater flow to the S/Gs must continue until at least 450 GPM flow to one S/G is established (from CF or CA) orNR level in at least one intact S/G is greater than 11%.

Answer A Discussion INCORRECT: See explanation above.

PLAUSIBLE:

Part I is plausible because the basis document states that feed and bleed may need to be initiated within 8 minutes after the reactor trips on low-low SG level.

Part 2 is plausible if the applicant concludes that once some form of NC system heat removal is established that efforts to restore an adequate heat sink (via the S/Gs) may be terminated.

Answer B Discussion INCORRECT: See explanation above.

PLAUSIBLE:

Part I is correct.

Part 2 is plausible if the applicant concludes that once some form of NC system heat removal is established that efforts to restore an adequate heat sink (via the S/Gs) may be terminated.

Answer C Discussion INCORRECT: See explanation above.

PLAUSIBLE:

Part 1 is plausible because the basis document states that feed and bleed may need to be initiated within 8 minutes after the reactor trips on low-low SG level.

Part 2 is correct.

Answer D Discussion CORRECT: See explanation above.

Basis for meeting the K The K/A is matched because a loss of heat sink has occurred and the applicant must have knowledge of the impact on the RCS (i.e. how long they have after criteria has been met to initiate Feed and Bleed of the NC system). Also, they demonstrate the ability to use procedures to correct, control, and mitigate by having knowledge of when (procedurally) effort to restore feedwater can be terminated and by knowing the procedural time requirement (from the basis document) for initiating feed and bleed.

Basis for Hi Cog Basis for SRO only This question is SRO-only knowledge linked to 1 OCFR55.43(b)(5) (Assessment and Selection of Procedures) as described in the Clarification Guidance for SRO-only Question Rev 1 (dated 03/11/2010):

I) The question can NOT be answered solely by knowing systems knowledge. To answer the question the SRO applicant must have detailed knowledge of both procedure steps and the procedure background document.

2) The question can NOT be answered solely by knowing immediate operator actions. There are NO immediate actions associated with FR- H. 1 (Response to Loss of Secondary Heat Sink).

3) The question can NOT be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs.

The knowledge required is detailed procedure step knowledge and background document information.

4) The question can NOT be answered solely by knowing the purpose, overall sequence of events, or overall mitigate strategy of the procedure.

Thursday, May 29, 2014 Page 241 of 298

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ILT-30 MNS SRO NRC Examination QUESTION 81

5) The second part of the question related to when Feed and Bleed can be terminated is RO level knowledge. The first part of this question requires the applicant to have detailed knowledge of the procedure background document information (specifically time requirement to initiate Feed and Bleed). As such the first part of the question is SRO-level knowledge.

Job Level Cognitive Level QuestionType Question Source SRO Memory MODIFIED 2012 MNS NRC Q56 (Bank 4429)

Development References Student References Provided

References:

FR-H. 1 (Response to Loss of Secondary Heat Sink)

FR-H. 1 Background Document Learning Objectives:

EPFRH0O6 SYSOO2 A2.04 Reactor Coolant System (RCS)

Ability to (a) predict the impacts of the following malfunctions or operations on the RCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5 / 43.5 / 45.3 / 45.5)

Loss of heat sinks Comments: rksIStatus E9 Thursday, May 29, 2014 Page 242 of 298

Q81

References:

&O TIME CRITICAL TASKS

&1 Operator Action to initiate Feed and Bleed once criteria met:

Expectation: Feed and bleed should be initiated as quickly as possible after meeting criteria in FR-Hi but must be initiated within 4 minutes of rneetinq criteria.

The N\l recirc valve must be closed within 5 minutes of initiating feed and bleed. Although this action is NOT required for a design basis event, this item is included due to its P RA significance during a loss of secondary heat sink event.

&2 CA Control, Expectation:

Operations managementexpectation is that CA will be manually controlled without going p high in any SIG, Simulator performance has shown that operator performance is acceptable.

FORTRAINING PiRPOSES ONLY Pag 55 of 55

Q81

References:

MNS RESPONSE TO LOSS OF SECONDARY HEAT SINK PAGE NO.

EP/1IA/5000IFR-H.1 36 of 122 Revl7

.cT

r .E: cE:  :.r:

SE RESPONSE NOT OBTAINED

20. (Continued) z WHEN other feedwatersources available, THEN reviev Enclosure 14 (Aligning Additional Feedwaler Source After RY Feeds SJGs. prior to aligning other source

21. Check if NC System feed and bleed should be initiated:

a Check feed and bleed HAS BEEN

- a. 12 Step 21 .c.

PREVIOUSLY ESTABLISHED PER STEPS 23 throLigh 27

_b GOTOStep38 c CheckWfRlevelinatleast3 s- c RETURNTOStepI.

LESS THAN 24% (36% ACC)

22. Perform Steps 23 through 27 quicklyto establish NC heatremoval by NC feed and bleed.

23. Ensure all NC pumps -OFF.

24. Initiate S!l.

25. Check NVPMPSTC COLD LEG FLOWN Performthefollowing:

- INDICATING FLOW.

a. Start NV pumps.

b. Start NI pumps.

C. OPEN the following valves:

• 1NV22IA(U1 NVPumpFrom FWST Isol)

. INV222B(U1 NVPumpFrom FWST Isol).

d. CLOSE the following vaives:

. INV-141A(U1 VCTOuiletlsol)

INVI42B (UI VCT Outlet lsol).

(RNO continued on next oaae)

Q81

References:

MNS RESPONSE TO LOSS OF SECONDARY HEAT SINK PAGE NO.

EP/1/A/5000/FR-H.1 37 of 122 Rev. 17 UNIT 1 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

25. (Continued)

e. OPEN the following valves:

. 1 NI-QA (NC Cold Leg jjijFrom NV)

. INI-IOB(NCCold Leg jp.jFrornNV).

1. jf NV S/I flowpath is established, AND NV pump is on THEN GO TO Step 26.

g. IF both of the following conditions exist, THEN GO TO Step 26:

. AnyNlpumpison

• Reactor was tripped for at least gg minutes flQSoimlefflntin this ER

h. Continue attempts to restore ECCS flow.

. Continue attempts to restore feed flow to S/.

j. IE AT ANY.TiM containment pressure is greaterthan2o PSIG prior to reaching FWST LEVEL LO alarm setpoint (95 inches), THEN contact TSC to evaluate guidance to mitigate high containment pressure.

k. IF EP!1/N5000/FR-S.1 (Response To Nuclear Power GenerationfATWS)OR EP/1/A/5000/FR-C.1 (Response To Inadequate Core Cooling) has been previously implemented, THEN RETURN Step 6.

I. Energize H2 Igniters by depressing ON and OVERRIDE.

m. Dispatch operator to stop all Unit I NF AHUs PER EPIIIN5000IG-1 (Generic Enclosures), Enclosure 28 (De-energizing Ice Condenser AH Us).

n. TJf43P Step 6.

Q81

References:

MNS RESPONSE TO LOSS OF SECONDARY HEAT SINK PAGE NO.

EP/1/A5000/FR-H.1 38 of 122 UNIT 1 Rev. 17 SE RESPONSE NOT OBTAINED

26. Establish NC System bleed path as follows:

a. Check all Pzr PORV isolation valves - a. OPEN all Pzr PORV isolation valves.

OPEN.

b. Select OPEN on two Pzr PORVs that have an open Pzr PORV isolation valve.

c. Align N to Pzr PORVs by OPENING c. Perform the flowir:

the followinci valves

1) Ensure Phase B reset.

. 1 Nl-430A (eicj N2 From CLA To 1 NC-34A) 2) OPEN the following valves:

• iNl-$31B(Eeg.N2 From CLAT0 . 1VI129BVI SupplytoACont I NC-32B & 36B) J1J.Outside Isol)

• lVl-1608(VlSupplytoBCont aVI tjjLOutside Isol) 1VI-15OB(jy.ContNon-Cont Outside Isol).

3) LEVI header pressure is less than than 85 PSIG, THEN perform the following:

• Ensure Pzr PORVs with N2 aligned have been OPENED.

. Ensure only two Pzr PORV bleed paths are selected OPEN.

• Restore VI fRAP/1/N55OO/22 (Loss Of VI).

d. Check power to all Pzr PORV isolation d. Performtheftowir:

valves AVAILABLE.

1) Evaluate cause of power loss and initiate actions to restore power to affected isolation valve(s).

2) WHEN power is restored, THEN perform the following:

a) OPEN Pzr PORV isolation valves.

b) Ensure two Pzr PORVs that have an open Pzr PORV isolation valve are OPENED.

Q81

References:

MNS RESPONSE TO LOSS OF SECONDARY HEAT SINK PAGE NO EPJ1/A15000FR-Hi 39 of 122 Rev. 17 UNIT 1 c-:c. RESPONSE NOT CE

27. Check two PzrPORVs and associated Perform the following:

isolation valves OPEN.

a. Attempt to OPEN two Pzr PORVs and associated isolation valves.

b, IE two Pzr PORV flow paths are opened, THENjg Step 28.

c OPEN one train of head vent valves:

a TrainA

. I NC-272AC (Ui A Train Head Vent to PRT Isol)

. 1NC-273AC(Ui ATrain Head Vent to PRT Isol).

OR a Train B:

. INC-274B (UI B Train Head Vent to PRT 1501) a INC-275B(UI B Train Head Vent to PRT Isol).

28 Isolate NV Recirc flowpath as follows:

a CLOSE the following valves.

a INV-1508(Ll1 NV Pump Recirclso[l a INV-151A(U1 NVPunip Recirclsofl b Maintain NV recirc valves closed unless directed to open by subsequent steps.

Q81

References:

MNS RESPONSE TO LOSS OF SECONDARY HEAT SINK PAGE N0 EPIIIN5000IFR-Hi 40 of 122 urri Rev.17 c::: E:: E:Esc RESPONSE NOT OBTAINED 29 Establish containmerttH2 mitigation as follows:

a Check EP/1/A/5OOOFR-Si (Response a. ]Step29 c To Nuclear Power Generation!ATWS

- HAS PREVIOUSLY BEEN IMPLEMENTED b GOIQStep3O.

c. Check EP/1/AI50001FR-C.i (Response c QjgStep29e, To Inadequate Core conHr- HAS PREVIOIJSL( BEEN IMPLEMENTED d GOTOStep3O.

e Energize H: Igniters by depressing ON and OVERRlDE t Dispatch operator to stop aH Unit 1 NF AH Us PER EP!1 !A/50001G-1 (Generic Enclosures). Enclosure 28 (Dc-energizing Ice CondenserAHtJs).

30 Ensure Pzr heaters remain oft as follows:

. Place k B and D Pzr heaters in manual and off.

. Open C PZR HTR GRP SLIP BKR -

- 31. Have anotherlicensedoperatorcheck 511 equipment PER Enclosure 15 (Subsequent S1l Actions) while continuing with this procedure.

32. Maintain NC System heat removal by performing the following:

. Maintain S/I flow.

• Maintain two Pzr PORV flowpaths -

OPEN.

Q81

References:

MNS RESPONSE TO LOSS OF SECONDARY HEAT SINK PAGE NO.

EPIIIA/5000/FR-H.1 41 of 122 Rev.17 UNIT I ACTION!EXPECTED RESPONSE RESPONSE NOT OBTAINED 33 Reset the following

a. Sit, a. Reset S/t PER EPIIIA!5000IG-1 (Generic Enclosures), Enclosure 23 (Local Reset of S/I Signal).

b. Sequencers. b. Dispatch operator to open affected sequencer cortrol power breaker:

A Train 1 EVDA Breaker 6

• B Train I EVDD Breaker 8.

c. IF AT ANY TIME a B/O signal occurs THEN restart S/I equipment previously on-

Q81

References:

MNS RESPONSE TO LOSS OF SECONDARY HEAT SINK PAGE NO EP!1/A15000/FR-Hi 42 of 122 uirr 1 Rev 17 ACTIONJEXPECTED RESPONSE RESPONSE NOT OBTAINED

34. Check containment pressiwe:

a. Pressure- HAS REMAINED LESS a Perform the following:

THAN3PSIG

1) Check Monitor Light Panel Group 7 lit-

2) jf Group 7 window is dark on energized train(s), ThEN perform the following:

a) Initiate Phase B using PHASE r B&VX&CONT VENT ISOL TRAIN A(B pushbuttons.

b) Jf Group 7 window is still dark, THEN perform the following:

. Check OAC Monitor Light Program MONL) for Phase B, and align valves.

. OACisoutofservice, THEN ensure Phase B valves closed PER EP!1IN5aOOIG-1 (Generic Enclosures),

Enclosure 12 (Phase B Valve Checklist).

3) Reestablish Vito containment as follows:

a) Ensure Phase B reset.

b) OPEN the following valves:

a IVI-129B(VlSupplytoA Cont VI Jfj Outside Isol) a lVl-160B(VlSupplytoB ContVI tjjIJIOutside Isol) a 1Vl-15OB(jyLContNon-Cont Outside Isol).

c) VI header pressure is less than 85 PSIG, THEN restore VI PER AP/1 1A15500122 (Loss Of VI)

(RNO continued on next oaaei

Q81

References:

MNS RESPONSE TO LOSS OF SECONDARY HEAT SINK PAGE NO.

EP!IIN5000IFR-i-Ii 43 of 122 Rev17 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 34 (Continued)

4) iEATANYiiMwhileinthIs procedurea Phase B reoccurs, THEN perform the following:

a) Reset Phase B b) OPEN IVI-129B (VI Supply toA ContVI LOutside IsoI)

C) OPEN IVI-1 SOB (VI Supply to B ContVl JjOutside Isol).

d) OPEN IVI-150B (JCont Non-Cont Outside 1501).

5) Ensure all RV pumps are in manual and off.

6) Check Phase B HVAC equipment PER Enclosure 16 (Phase B HVAC Equipment).

7) iEJANYI1Mcontainment pressure is greater than 20 PSIG prior to reaching FWST LEVEL LO alarm setpoint (95 inches), TI-lEN contact TSC to evaluate guidance to mitigate high containment pressure.

8) GOTOStep36.

b IF AT ANY TIME while in this procedure containment pressure goes above 3 PSIG THEN perform Step 34.a.

c both of the follo4ng conditions exist, THEN perform Enclosure 17 (VX Manual Start And Isolating RV Cooling):

. Containment pressure has remained less than S PSIG

. Containment pressure is between 1 PSIG and 3 PSIG.

Q81

References:

MNS RESPONSE TO LOSS OF SECONDARY HEAT SINK PAGE NO EP/1/N5COO/FRHi 44 of 122 UNIT I Rev17 EEEE: :EzD SE RESPONSE NOT OBTAINED

35. Establish VI to coninmentas follows:

a check the following valves OPEN:

- a Perform the following:

. WI-i 29B (VI Supply to A Cont .Ess.Vl 1) Ensure Phase B reset

.Hclr Outside Isol)

2) OPEN vaive&

. 1VI-i 60B (VI Supply to B Cont Vl

.ljdtc Outside Isol)

. iVI-150B (.LwrContNon-s.sCont Outside Fsolj

b. Check VI header pressure GREATER

- b Restore VlAPI1IN55OOl22 (Loss THAN 85 PSIG Of VI)

UJ I

.2 C IL 0 c a)

CUE C ECU 02, z 0 C

. o C

ISV a) CUr-U. C,)

LJ P1 0

C, E

U.

2Q Cu P1 0

C, H

Ui I

3.3 ft If the applicant believes the Lump level is NOT adequate then implementing EC4.-l.l lwouldbe correct per ES-LI, Step 2, RNO column, step!. With this misconception it is also reasonable to believe that additional inventory is ieeded, and can be provided by initiating FWST makeup.

Answer B Discussion Answer C Discussion

[Plausible, since the pump and valve alignments are correct ECA-Li is plausible if the applicant confases the sunup level as being inadequate.

IThe sunup level given in the stemis 3.4 fret; the levellis.ted in ES-13, (Ttansfer to Cold LegRecirc.) as adequate is > 3.3 ft If the applicant elleves the sump level is NOT adequate, then implementing ECA-.L1 would be correct, per ES-LI, Step 2. RNO column, step!. With this Lmisconception it is also reasonable to believe that additional inventmy is needed, and can be provided by initiating FWST makeup.

Answer D Discussion

[Plausible, since the procedure selection (ECA-L3) is correct The pumps are cavitatiug and it is plausible to take action to protect the pumps.

,However the action in the second part f this distactor, though plausible, is only directed ECA-i.l, not in ECA-l.3.

Basis for meeting the KA The K/A is matched because the question involves a Large Break LOCA that has progressed to the point of a Transfer to Cold Leg Redrculation. The applicant is tested on the ability to determine if the safety related low head injection pumps are operable or available by anatyninz plant conditions invoking amps oscillation, and making a decision that the pumps are NOT available. With that determination, the applicant goes on to select a sub-procedure fin mitigation.

Basis for Hi Cog This is a high cognitive level question because the applicant must analyze a configuration of several operating pumps taking auction from, a source which may be degrading. The applicant is given conditions where pump amps begin oscillating, analyze those conditions, and draw a conclusion that will be used to select the appropriate procedure.

Basis for SRO only This question meets the following criteria fin an SIlO only question as described in the Clarification Guidance for SRO-only Questions (Tev I dated 03111/2010) under the Screening Criteria for questionlmnked to IOCFR55.43(bXS) (Assessment and Selection of Procedures):

1)The question can NOT be answered by knowing systems knowiedge alone.

4)DThe question can NOT be answered by knowing the purpose, overall sequence of events, or overall mitigative swategy of the procedure.

5)DThe question requires the auplicantto recall procedureconteut from several sub-procedures (EC.A-Ll. ECA-l.3, P5-1.3), assess plant conditions, and then select which sub-procedure contains the actions for mitigation. Specifically, the applicant is placed in a condition where ES 13, (Transfer to Cold Leg Resist.) is being implemented. Plant conditions then change which impact whether the actions in P5-1.3 are appropriate. The applicant has as determine that they are NOT, and then recall detailed, content from two additional sub-procedures, and make a determination on which sub-procedure contains the actions for fiarther mitigation of these new plant conditions.

Job Level Cognitive Level QuestionType Question Source SRO Comprehension NEW Development References Student References Provided CA-l. 1, (Loss of Emergency Coolant Recirculation), Rev, 034 ECA-l.3, (Containment Sunup Blockage), Rev, 007 S-l.3, Background Document Monday, September 30, 2013 Page 7300 of 9213

Q85

References:

MNS TRANSFER TO COLD LEG RECIRC PAGE NO.

EPfl/AISOOO/ESl 3 15 of 25 L UNf 1 Rev. 27 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

9. (Continued) d IF ND flow to NC loops is established THEN perform the following.

1) Monitor ND pumps for proper operatioft

2) iF AT ANY TIME loss of all ND flow to NC loops occurs. THEN perform the following:

a) IF sump blockage is suspected.

THEN GO TO EP/1!A/5000/EOA-1 .3 (Containment Sump Blockage).

b)

e. IF OAC available, THEN use AP AND EP DIAGNOSTICS pageCONT SUMP RECIRC PARAMS to assist in monitoring FWST level until NV and NI pumps are aligned for Cold Leg Reciro,

1. IF AT ANY TIME a B!O signal occurs, THEN restart S/I equipment previously on

g. EP/1/AJ5OOOIFO (Critical Safety Function Status Trees) may now be implemented,

h. RETURNIP procedure and step in effect, CAUflQ1 If a B!O occurs after aligning NV pumps to Cold Leg Recirc, NV pumps will auto sequence on without adequate suction. It is critical to start ND pumps to provide NV pump suction.

10. IF AT ANY TIME a B!O signal occurs, THEN restart SIL equipment previously on.

Q85

References:

MNS CONTAINMENT SUMP BLOCKAGE PAGE NO.

EPI1IN5000IECA-I .3 3 of 113 Rev,6 UNrrl ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED Ai,JTION To prevent damaging pumps, NV and NI pumps taking suction from ND pumps must be secured prior to stopping ND pumps.

4. IF AT ANy TIME ND pumps are cavitatirig, THEN perform the following:

a. Stop any NV or NI pumps taking suction from affected ND pump.

b. Stop NS pumps.

c. Check ND spray valves CLOSED:

- c. CLOSE valves.

a 1NS-43A (IA ND HxOutlet to NS Cent Outside Iso!)

a INS-388 (18 ND Hx Outlet to NS Cont Outside Iso!).

d. Check any ND pump ON. -

I Step 5.

e Reset modulating valves using reset buttons on RN control board.

C IF cavitation continues. THEN CLOSE the following valves a IND-29 (lA ND Hx Outlet Iso!)

a 1ND-14 (lB ND Hx Outlet lsol).

g IF IA ND pump continues to cavitate THEN CLOSE 1Nl-I73A1A FID to A &

B Cold Legs Cont Outside Isol).

h I 18 ND pump continues to cavitate.

THEN CLOSE lNl-178B (lB ND to C &

D Cold Legs ont Outside Iso!).

. Check INI-183B (Ui ND to B & CHot IF any ND pump is cavitating. THE.N Leg Cent Outside iso!) CLOSED

- CLOSE INI-1838.

i ND pumpis continue to cavitate.

THEN stop affected ND pump.

FOR REVIEW ONLY DO NOT DISTRIBUTE -

ILT-30 MNS SRO NRC Examination QUESTION 86 86 APEO27 2.4.31 Pressurizer Pressure Control System (PZR PCS) Malfunction APEO27 GENERIC Knowledge of annunciator alarms, indications, or response procedures. (CFR: 41.10 / 45.3)

Given the following conditions on Unit I:

• An NC system cooldown and depressurization is in progress in preparation for refueling

• NC system temperature is 260°F

• LTOP (LOW TEMPERATURE OVERPRESSURE PROTECTION) is in service

• I&E has just completed PT/i /A14150/014 (PZR LTOP PROTECTION ANALOG CHANNEL OPERATIONAL TEST) and reports the following automatic open setpoints:

o 1 NC-34A (Ui PZR PORV) 380 PSIG o 1 NC-32B (Ui PZR PORV) 390 PSIG

• Subsequently, the following are observed on Annunciator Panel IAD-6:

• The crew observes that 1 NI-430A (EMERG N 2 FROM CLA TO 1 NC-34A) is CLOSED Based on conditions above,:

1) 1NC-34A (Ui PZR PORV) is

2) 1NC-32B (UI PZR PORV) is Which ONE (1) of the following completes the statements above?

A. 1. OPERABLE

2. INOPERABLE B. I. INOPERABLE

2. INOPERABLE C. 1. OPERABLE

2. OPERABLE D. 1. INOPERABLE

2. OPERABLE Thursday, May 29, 2014 Page 255 of 298

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ILT-30 MNS SRO NRC Examination QUESTION 86 86 General Discussion When LTOP is placed in service (control switch taken from NORM to LOW PRESS), automatic opening of NC-34A and NC-32B at the LTOP setpoint is enabled and the nitrogen supply valves to the PORVs automatically open. At this point, motive force for the PORVs is supplied by both N2 and VI.

During LTOP operation, as long as one supply of motive force is available, the PORV remains operable.

Another condition of OPERABILITY for the PORVs in LTOP mode is that the PORV must be capable of automatically opening at 385 PSIG (JAW Tech Spec 3.4.12 LTOP).

With NC system temperature less than 300°F, LTOP is required to be OPERABLE.

Since 1NC-34A meets the required LTOP auto open setpoint AND has at least one source of motive force available, it is OPERABLE.

1NC-32B has both sources of motive force available. However, it does not meet the required Tech Spec LTOP auto open setpoint and is therefore INOPERABLE.

Answer A Discussion See explanation above.

Answer B Discussion INCORRECT: See explanation above.

PLAUSIBLE:

The first part is plausible if the applicant does not recall that one only source of motive force is required for the PORV to be OPERABLE or if the applicant concludes that both VI and N2 are not available to the PORV.

The second part is correct.

Answer C Discussion INCORRECT: See explanation above.

PLAUSIBLE:

The first part is correct.

The second part is plausible if the applicant incorrectly recalls the required LTOP auto open setpoint from TS 3.4.12 and concludes that the required setpoint is 390 PSIG or greater.

Answer D Discussion INCORRECT: See explanation above.

PLAUSIBLE:

The first part is plausible if the applicant does not recall that one only source of motive force is required for the PORV to be OPERABLE or if the applicant concludes that both VI and N2 are not available to the PORV.

The second part is plausible if the applicant incorrectly recalls the required LTOP auto open setpoint from TS 3.4.12 and concludes that the required setpoint is 390 PSIG or greater.

Basis for meeting the K The K/A is matched because it requires the applicant to have knowledge of the annunciator alarm associated with LTOP operation.

Basis for Hi Cog This is a higher cognitive level question because it requires more than one mental step.

First, the applicant must analyze the given conditions and determine from that analysis the possible cause of the annunciator alarm indications.

Next, the applicant must recall from memory the setpoint for LTOP system OPERABILITY from Tech Spec 3.4.12.

Next, the applicant must recall from memory the requirements for PORV OPERABILITY related to LTOP.

Finally, the applicant must associate all of the information above to determine the OPERABILTY of INC-34A and 1 NC-32B.

Basis for SRO only The question meets the following criteria for an SRO only question as described in the Clarification Guidance for SRO-only Questions Rev 1 dated 03/11/2010 for screening questions linked to IOCFR55.43(b)(2) (Tech Specs):

Thursday, May 29, 2014 Page 256 of 298

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86 ILT-30 MNS SRO NRC Examination QUESTION 86

1) This question can NOT be answered by knowing less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Tech Specs Question is related to knowledge of TS 3.4.12 Basis.

2) This question can NOT be answered by knowing information listed above-the-line.

The part of the question related to the LTOP setpoint is above-the-line knowledge. However, that knowledge along with the ability to determine the implications of the annunciator alarms (which requires knowledge of the TS basis) is required to determine the correct response.

3) This question can NOT be answered by knowing the TS Safety Limits or their bases.

Not related to TS Safety Limits.

4) This question requires the applicant to have knowledge of the basis of Tech Spec 3.4.12 and is therefore SRO knowledge.

Job Level Cognitive Level QuestionType Question Source SRO Comprehension NEW Development References Student References Provided

References:

Tech Spec 3.4.12 (LTOP)

Tech Spec 3.4.12 Basis Learning Objectives:

OP-MC-PS-NC Objective 24 APEO27 2.4.31 Pressurizer Pressure Control System (PZR PCS) Malfunction APEO27 GENERIC Knowledge of annunciator alarms, indications, or response procedures. (CFR: 41.10 / 45.3) 401-9 Comments: RemarkslStatus Thursday, May 29, 2014 Page 257 of 298

Q86

References:

From Tech Spec 3.4.12 (LTOP) Basis:

LTOP System B 3.412 B 4 REACTOR COOLANT SYSTEM (RCS)

B 3,4.12 Low Temperature Overpressure Protection (LTOP) System BASES BACKGROUND The LTOP System controls RCS pressure at low temperatures so the integrity of the reactor coolant pressure boundary (RCPB) is riot compromised by violating the pressure and temperature (P/T) limits of 10 CFR 50. Appendix G (Ref. 1). The reactor vessel is the limiting RCPB component for demonstrating such protection. This specification provides the maximum allowable actuation logic i fórthe power operated relief valves (PORVs) and LCO 3.4.3. RCS Pressure and TemperatLire (PIT) Limits, provides the maximum RCS pressure forthe existing RCS cold leg temperature during coolclown shutdown, arid h.ettip to meet the Reference I requirements during the LTOP MODES.

The reactor vessel material is Less tough at low temperatures than at normal operating temperature. As the vessel neutron exposure accumulates the material toughness decreases and becomes less resistant to pressure stress at low temperatures (Ref. 2). RCS pressure.

therefore, is maintained low at low temperatures and is increased only as temperature is increased.

The potential for vessel is most acute when the RCS is water solid, occurring only while shutdown; a pressure fluctuation can occur more quickly than an operator can react to relieve the condition.

Exceeding the RCS P/T limits by a significant amount could cause brittle cracking of the reactor vessel. LCO 3.4.3 requires administrative control of RCS pressure and temperature during attp and oooldown to prevent exceeding the specified limits.

This LCO provides RCS overpressure protection by having a rninimuni coolant input capability and having adequate pressure relief capacity.

Limiting coolant input capability requires all but one centrifugal charging pump or one safety injection pump incapable of injection into the RCS and isolating the accumulators. The pressure relief capacity reqLiires either two redundant PORVs or a depressurized RCS and an RCS vent ot sufficient size. One PORV or the open RCS vent is the overpressure protection device that acts to terminate an increasing pressure event.

With minimum coolant input capability, the ability to provide core coolant addition is restricted. The LCO does not require the makeup control McGure Uns 1 and 2 B 3*4121 Revision No. 115

Q86

References:

[TOP S.*ystem B 3412 BASES BACKGROUND (continued) system deactivated or the safety injection (SI) actuation circuits blocked.

Due to the lower pressures in the LTOP MODES and the expected core decay heat levels, the makeup system can provide adequate flow via the makeup control valve. If conditions require the use of more than one centrifugal charging pump for makeup in the event of loss of inventory, then pumps can be made available through manual actions.

PORV Requirements As designed for the LTOP System, each PORV is signaled to open if the RCS pressure reaches 385 psig when the PORVS are in the b-press mode of operatioa If the PORVs are being used to meetthe requirements of this specification, then RCS co[d leg temperature is limited in accordance with the LTOP analysis. For cases where no reactor coolant pumps are in operation, this temperature limit is met by monitoring of BOTH the Wide Range Cold Leg temperatures and Residual Heat Removal Heat Exchanger discharge temperature. These temperatures are the most representative of the fluid in the reactor vessel region. The LTOP actuation logic monitors both RCS temperature and RCS pressure. The signals used to generate the pressure pijJ originate from the safety related narrow range pressure transmitters. The signals used to generate the temperature originate from the wide range RTDs on cold leg C and hot leg D. Each signal is input to the appropriate NSSS protection system cabinet where it is converted to an intemal signal and then input to a comparator to generate an actuation signaL If the indicated pressure meets or exceeds the setpoint, a PORV is signaled to Open.

This Specification presents the PORVpgjt for LTOP. Having the tpgjp.t. of both valves within the limits ensures that the Reference I limits will not be exceeded in any analyzed event.

When a PORV is opened in an increasing pressure transient, the release of coolant will cause the pressure increase to slow and reverse. As the PORV releases coolant, the RCS pressure decreases until a reset pressure is reached and the valve is signaled to close. The pressure continues to decrease below the reset pressure as the valve closes.

RCS Vent Requirements Once the RCS is depressurized, a vent exposed to the containment atmosphere will maintain the RCS at containment ambient pressure in an RCS overpressure transient, if the relieving requirements of the transient do not exceed the capabilities of the vent. Thus, the vent path must be McGu[e Units I and 2 B 34122

Q86

References:

LTOP System B 3..4. 1.2 BASES BACKGROUND (continued) capable of relieving the flow resulting from the limiting LTOP mass or heat input transient, and maintaining pressure below the PIT limits. The required vent capacity may be provided by one or more vent paths.

The vent path(s) must be above the level of reactor coolant, so as not to drain the RCS when open.

APPLICABLE Safety analyses (Ref. 4) demonstrate that the reactor vessel is SAFETY ANALYSES adequately protected against exceeding the Reference 1 PIT limits. In MODES 1,2, and 3, and in MODE 4 with RCS cold leg temperature exceeding 300°F, the zer safety valves will prevent RCS pressure from exceeding the Reference I limits. At about 300°F and below, overpressure prevention falls to two OPERABLE PORVs or to a depressurized RCS and a sufficient sized RCS vent. Each of these means has a limited overpressure relief capability.

The actual temperature at which the pressure in the PIT limit curve fails below the safety valve setpoint increases as the reactor vessel material toughness decreases due to neutron Each time the PIT curves are revised, the LTOP System must be re-evaluated to ensure its functional requirements can still be met using the PORV method or the depressurized and vented RCS condition.

Any change to the RCS must be evaluated against the Reference 4 analyses to determine the impact of the change on the LTOP acceptance limits.

Transients that are capable ofy ijng the RCS are categorized as either mass or heat input transients, examples of which follow:

Mass Input Type Transients

a. Inadvertent safety injection; or

b. Charginglletdown flow mismatch.

Heat Input Type Transients

a. Inadvertent actuation ofj.j heaters; b Loss of RHR cooling; or McGuire Units 1 and 2 ReAsion No. 115

Q86

References:

BASES APPUCABLE SAFETY ANALYSES (continued) c Reactor coolant pump (RCP) startup with temperature asymmetry within the RCS or between the RCS and steam generators The following are required during the LTOP MODES to ensure that mass and heat input transients do not occur which either of the LTOP overpressure protection means cannot handle:

a Rendering all but one centrifugal charging pump or one safety injection pump incapable of injection;

b. Deactivating the accumulator discharge isolation valves in their closed positions; and

c. Disallowing start of an RCP if secondary temperature is more than 50°F above primary temperature in any one loop LCO 346, URCS LoopsMODE 4 and LCO 347, RCS LoopsMODE 5, Loops Filled, provide this protection The Reference 4 analyses demonstrate that either one PORV or the depressurized RCS and RCS vent can maintain RCS pressure below limits when only one centrifugal charging pump or one safety injection pump are actuated Thus, the LCO allows only one centrifugal charging pump or one safety injection pump OPERABLE during the LTOP MODES Since neither one PORV nor the RCS vent can handle the pressure transient from accumulator injection when RCS temperature is low the LCO also requires the accumulators isolation when accumulator pressure is greater than or equal to the maximum RCS pressure for the existing RCS cold leg temperature allowed in LCO 343.

The isolated accumulators must have their discharge valves closed and power removed.

Fracture mechanics analyses established the temperature of LTOP Applicabllity at 300°F The consequences of a small break loss of coolant accident (LOCA) in LTOP MODE 4 conform to I 0 GFR 50.46 and 10 CFR 50 Appendix K (Refs, 5 and 6), requirements by having a maximum of one centrifugal charging pump OPERABLE and SI actuation enabled.

MoGuire Units 1 and 2 B 3.4.124 Revision No. 115

Q86

References:

LTOP Systen B 5412 APPLICABLE SAFETY ANALYSES (continued)

PORV Performance The fracture mechanics analyses show that the vessel is protected when the PORVs are set to open at or below the specified limit. The fjj are derived by analyses that model the performance of the LTOP System, assuming the limiting LTOP transient of one centrifugal charging pump or one safety injection pump injecting into the RCS, These analyses consider pressure overshoot and undershoot beyond the PORV opening and closing, resulting from signal processing and valve stroke times. The PORVjgjj at or below the derived limit ensures the Reference I P/T limits will be met.

The PORVfppjj will be updated when the revised P/T limits conflict with the LTOP analysis limits. The PIT limits are periodically modified as the reactor vessel material toughness decreases due to neutron imt caused by neutron irradiation. Revised limits are determined using neutron projections and the results of examinations of the reactor vessel material irradiation surveillance specimens. The Bases for LCO 3.4.3, RCS Pressure and Temperature (P11) Limits, discuss these examinations.

The PORVs are considered active components. Thus, the failure of one PORV is assumed to represent the worst case, single active failure.

RCS Vent Performance With the RCSdepressurized, analyses show a vent size of 215 square inches is capable of mitigating the allowed LTOP overpressure transient The capacity of a vent this size is greater than the flow of the limiting transient for the LTOP configuration, one centrifugal charging pump or one safety injection pump OPERABLE, maintaining RCS pressure less than the maximum pressure on the P/T limit curve.

The RCS vent size will be re-evaluated for compliance each time the PIT limit curves are revised based on the results of the vessel material surveillance.

The RCS vent is passive and is not subject to active failure.

The LTOP System satisfies Criterion 2 of 10 CFR 50.36 (Ref. 7).

MoGuire Units 1 and2 B 5412-5 ReGsion No 115

Q86

References:

oP!2/A/dloofolo G Annunciator Response For Panel 2AD-6 Page 15 of 85 Nomenclature: PORV LO PRESS MODE NOT SELECTED Serpoint: NC temperature less than 320 T and lOW PRESS NOT selected oi PORV crverpress Protection Select 2NC 54A 2NU32B Oi igin PORV 1ut Relay GA Probable Cau se:

• NC temperature less than S2tr F and NORM selected

• Possible relay or circuit problem Automatic Action. None Immediate Action: Enure LOW PRESS selected for the following.

• PORV CIverp:e-s Protection Select 2NC34A

• PORV Oerpress Protection Select 2Nc3.2B Supplementary Action: None

References:

MCEE-250-0003-.Ol End Of Response Unit 2

Q86

References:

[TOP System 3.412 34 REACTOR COOLANT SYSTEM (RCS) 3.412 Low Temperature Overpressure Protection (LTOP) System LCO 3.412 An LTOP System shall be OPERABLE with a maximum of one centrifugal charging pump or one safety injection pump capable of injecting into the RS and the accumulators isolated and either a or b below:

a. Two power operated relief valves (PORVs) with lift setting <385 psig or

b. The RCS depressurized arid an RCS vent of> 2.75 square inches,

- NOTE--- - - -

APORVsec:urcclin the open position maybe used to rneetthe RCS vent requirement provided that its associated block valve is open and power removed.

APPLICABILITY: MODE 4 when any RCS cold leg temperature is <3OOF, MODE5 MODE 6 when the reactor vessel head is on.

Accumulator isolation is only required when accumulator pressure is greater than or equal to the maximum RCS pressure for the existing RCS cold leg temperature allowed by the PIT limit curves provided in Specification 3.4.3.

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ILT-30 MNS SRO NRC Examination QUESTION 87 871 EPEO38 2.1.20 Steam Generator Tube Rupture (SGTR)

EPEO38 GENERIC Ability to interpret and execute procedure steps. (CFR: 41.10 / 43.5 / 45.12)

Given the following conditions on Unit 1:

• A Reactor Trip and Safety Injection have occurred

• The crew is performing immediate actions contained in E-0 (REACTOR TRIP OR SAFETY INJECTION)

• IA Main Steam Line Radiation Monitor was increasing rapidly prior to the trip

• 1A SG pressure is dropping uncontrollably, and continues to drop to less than 200 PSIG after Main Steam Line Isolation

• 1B, IC, and 1D SG Main Steam Line pressures are stable Which ONE (1) of the following describes the procedure flowpath for this event when the crew makes a transition from E-0?

Procedure Legend:

E-1 (LOSS OF REACTOR OR SECONDARY COOLANT)

E-2 (FAULTED SG ISOLATION)

E-3 (STEAM GENERATOR TUBE RUPTURE)

ES-3.I (POST SGTR COOLDOWN USING BACKFILL)

ECA-3.I (SGTR WITH LOSS OF COOLANT ACCIDENT SUBCOOLED RECOVERY DESIRED)

ES-1.1 (SI TERMINATION)

A. E-3 ECA-3.1 B. E-2 == E-3 czz ES-3.1 C. E-2 E-3 ECA-3.1 D. E-3 E-2 ii E-1 ES-ti Thursday, May 29, 2014 Page 258 of 298

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ILT-30 MNS SRO NRC Examination QUESTION 87 L 87 General Discussion The first priority in E-0 is to check for a loss of coolant via a stuck open PZR PORV in which case a transition would be made to E- 1. Since one does not exist in this case, the next transition is to check all SGs/Main Steam lines intact. Since the IA SG is faulted, E-0 directs transition to E 2 for steps to isolate the faulted SG. When the isolation is complete, at the end of E-2 a check is made to see if the SG tubes are intact. Since there are indications of a SGTR, E-2 directs transition to E-3. Since the affected SG is both ruptured and faulted E-3 will eventually direct transition to ECA-3.l.

Answer A Discussion INCORRECT. See explanation above.

PLAUSIBLE:

It is plausible for the applicant to conclude that the tube rupture is more limiting that the fault and believe that E-3 should be entered first. Also, since E-3 contains actions for isolating the S/G, it is plausible for the appliant to conclude that E-2 need not be entered. From E-3 the correct transition is to ECA-l.3.

Answer B Discussion INCORRECT. See explanation above.

PLAUSIBLE:

The first two transitions are correct. It is plausible for the applicant to conclude that ES-3.l is correct since this would be the correct transition if the S/G was simply ruptured and not faulted or if it was rupture with steam pressure greater than 280 PSIG.

Answer C Discussion CORRECT. See explanation above.

Answer D Discussion INCORRECT. See explanation above.

PLAUSIBLE:

The transition to E-3 is plausible if the applicant concludes that the S/G tube rupture is most limiting. If that were the case the next logical transition would be to E-2 to isolate the steam leak. Once in E-2 if SI termination criteria is not met, the transition is to E-l and from E-l there is a transition to ES-l.l once SI termination criteria is met.

Basis for meeting the K The KIA is matched because it requires the appliant to have knowledge of the procedure transitions and flowpath related to a ruptured-faulted S/G.

Basis for Hi Cog This is a higher cognitive level question because it requires the applicant to perform a detailed analysis of given conditions to detemine a complex procedure transition flowpath.

Basis for SRO only This question meets the following criteria for an SRO only question as described in the Clarification Guidance for SRO-only Questions (Rev 1 dated 03/11/2010) under the Screening Criteria for question linked to 10CFR55.43(b)(5) (Assessment and Selection of Procedures):

1 )The question can NOT be answered by knowing systems knowledge alone.

4)The question can NOT be answered by knowing the purpose, overall sequence of events, or overall mitigative strategy of the procedure.

5)The question requires the applicant to recall procedure content from several sub-procedures, assess plant conditions, and then select which sub-procedure contains the actions for mitigation.

Job Level Cognitive Level QuestionType Question Source SRO Comprehension BANK 2011 MNS SRO Audit Examination Q86 (Bank 3395)

Development References Student References Provided E-l (LOSS OF REACTOR OR SECONDARY COOLANT)

E-2 (FAULTED SG ISOLATION)

E-3 (STEAM GENERATOR TUBE RUPTURE)

ES-3.l (POST SGTR COOLDOWN USiNG BACKFILL)

ECA-3.l (SGTR WITH LOSS OF COOLANT ACCIDENT SUBCOOLED RECOVERY DESIRED)

ES-l.l (SI TERMINATION)

Learning Objectives:

EPE3006 Thursday, May 29, 2014 Page 259 of 298

FOR REVIEW ONLY DO NOT DISTRIBUTE C

ILT-30 MNS SRO NRC Examination QUESTION 87 EPEO38 2.1.20 Steam Generator Tube Rupture (SGTR)

EPEO38 GENERIC Ability to interpret and execute procedure steps. (CFR: 41.10 / 43.5 / 45.12) 401-9 Comments: RemarkslStatus Thursday, May 29, 2014 Page 260 of 298

Q87

References:

MNS REACTOR TRIP OR SAFETY INJECTION PAGE NO.

EPI1!N50001E-0 15 of 38 Rev33 UNETI ACTION/EXPECTED RESPONSE RESPONSE NOT OETAINED

20. (Continued) c At [east one Pzr PORV isolation valve c power available, THEN OPEN one oPEN Pzr PORV isolation valve unless it was closed to isolate an open Pzr PORV 21 Check NC subcooling based on core exit at least one NV NI pump on, THEN TICs GREATER THAN OF.

- stop all NC pumps while maintaining seal injection flow.

22. Check if main Jjj intact: IF any SIG is faulted. THEN perform the following:

. All SIG pressures STABLE OR GOING UP a Implement EPJI/A15000!F-O (Critical Safety Function Status Trees).

. All Sf- PRESSURIZE[i b GO TO EP/lA5000/E-2 (Faulted Steam Generator Isolation).

23 Check if S!G tubes intact: IF SIG levels going up in an uncontrolled manner any EMF abnormal, TH

• The following secondary EMFs perform the following:

NORMAL:

a. Implement EP!1/A/5000/F-0 (Critical I EMF-33 (Condenser Air Ejector Safety Function Status Trees).

Exhaust) b GO TO EPII!A/5000/E-3 (Steam

. IEMF-34(L) (S!G Sample (Lo Range)) Generator Tube Rupture).

• I EMF-24 (SIG A)

. IEMF-25(SIGB)

. IEMF-26(SIG C)

. IEMF-27(S/G D)

. SIG levels STABLE OR GOING UP IN A CONTROLLED MANNER

Q87

References:

MNS FAULTED STEAM GENERATOR ISOLATION PAGE NO.

EPII!A15000IE-2 2 of 16 RevlQ UNITI ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED C. 0peratar Actions

1. Monitor Foldout page.

2 Maintain any faulted SIC or secondary break isolated during subsequent recovery actions unless needed for NC System cooldown.

3 Check the following valves- CLOSED: Perform the following:

. All MSIVs a. CLOSE valves.

. All MSIV bypass valves b IE any valve cannot be closed, THEN initiate Main Steam Isolation signal.

4 Check at least one SIC pressure - [Fall SifauIted, THEN GO TO STABLE OR GOING UP. EPIIIAI5000IECA-2.1 (Uncontrolledj Depressurization Of All Steam Generators).

5 Identify faulted SIGs): Perform the following:

. Any S/G pressure GOING DOWN IN a. Dispatch operators to search for AN UNCONTROLLED MANNER initiating break:

OR

. Maintg1ij:

. Any S;G DEPRESSURIZED e Main fj

. Other secondary piping

_b §QiQStep1O

6. Maintain at least one SIC available for NC System cooldown in subsequent steps.

Q87

References:

MNS FAULTED STEAM GENERATOR ISOLATION PAGE NO, EP,1!A5000E-2 3 of 16 UNIT 1 Rev. 10 El: EEo RESP.SE RESPONSE NOT OSTAINED T Check faulted SIGs SM PORV - PerfOrm the following:

CLO SED.

a. CLOSE faulted SIG(s) SM PORV.

b. IF SM PORV cannot be closed, THEN CLOSE SM PORV isolation valve,

c. IF SM PORV isolation valve cannot be closed, THEN dispatch operator to CLOSE SM PORV isolation valve, 8 Reset CA modulating valves

Q87

References:

MNS FAULTED STEAM GENERATOR ISOLATION PAGE NO.

EP!IIAI5000IE-2 4 of 16 UNril Rev. 10 ACTIONIEXPECTED RESPONSE RESPONSE HOT OBTAINED

9. Isolate faulted SIG(s) as follows:

a. For IASIG:

I) Check S!G A FDW ISOLATED 1> Perform the following:

status light (151-4) LIT a) Ensure the following valve(s) -

CLOSED:

. CLOSE 1CF-35AB (1A S/G CF Cont Outside 1501).

. CLOSE ICF-32AB (IA SIG CF Control>.

. CLOSE ICF-104A3 (IAS/G CF Control Bypass).

. CLOSE 1CF-126B (IAS/G CF To CA Nozzle Isol).

b) IF more than one Feedwater Isolation valve above is open, AND CM is still aligned to feed faulted SIG, THEN evaluate alternate means to stop CM flow to faulted S!G.

2) CLOSE I CA.-66AC (Ui TD CA 2> Perform the following Pump Dtsoh To 1A SIG Isol>

a> CLOSE ICA-64AB (UI TD CA Pump Disch To IA S/G Control).

b) Dispatch operator to CLOSE ICA-66AC (Unit 1 exterior doghouse, 750+8, FF-44, 4 j from inner waIl, 8 JI from column DD-44).

c) IF exterior doghouse not accessible, OR CA cannot be isolated, THEN dispatch operator to unlock and CLOSE 1CA-63 (Unit 1 TD CA Pump Disch To 1A S/G Control Inlet Isol) (Unit 1 CA pump 716÷9, BB-51, above door to TD CA pump, 4 JIsouth of IA CA pump).

Q87

References:

MNS FAULTED STEAM GENERATOR ISOLA11ON PAGE N0 EP1IIN5000IE-2 5 of 16 Rev.10 UNITI ACTIONIEXPECTED RESPONSE RESPONSE NOT OBTAINED

9. (Continued)

3) CLOSE ICA-62A(1A CA Pump 3) Perform the following Disch To IA SIG Isol) a) CLOSE ICA-SOA (IA CA Pump Disch To IA S/G Control).

b) Dispatch operator to CLOSE 1CA-62A (Unit I exterior doghouse, 750÷12, DD-44, southeast corner).

c) IF exterior doghouse not accessible, OR CA cannot be isolated, fliN dispatch operator to unlock and CLOSE ICA-59 (IA CA Pump Disch To IA S/G Control Inlet Isol) (Unit I CA pump 716÷10, CC50, above I B CA Pump).

4) Check BR valves CLOSED:

- 4) Perform the following:

a 1 RB-i B (IA SIG Jg ckjj Cont a) CLOSE valve(s)

Outside Isol Control) b) CLOSE IBB-123 (1A SIG a I BB-5A (A S/G BR Cont lnsde gyg Throttle Control) lsol)

5) CLOSE ISM-83 (A SM brie Drain lsol).

Q87

References:

MNS FAULTED STEAM GENERATOR ISOLAI1ON PAGE NO.

EPIIIAI5000!E-2 13 of 16 Rev.1O urri ExPEIJE: :rsE RESPONSE NOT OBTAINED 10 CLOSE IAS-12(Ul SM ToAS dr IF 1AS42 will not close, THEN perform ControHrilet Isol). the following:

a. IF controller for lAS-Il (Unit I Main Steam To Aux Steam JJt Control) (Unit I turbine çJg, 739, on column 1 F-34) is accessible, THEN dispatch operator to CLOSE lAS-Il Step 2 of EPI1IAI5000IG-i (Generic Enclosures), Enclosure 4 (Closing lAS-I 1 Using Local Controller) b IF controller for lAS-I I is not accessible, THEN dispatch operator to CLOSE 1AS-253 (Unit I Aux Steam

)J Isol) (service Lictg. 739+15, P-28, above overhead door to Unit I turbine bldg)

Q87

References:

MNS FAULTED STEAM GENERATOR ISOLATION PAGE NO.

EPI1IN5OOOIE2 14 of 16 Rev1O ACTIONIEXPECTED RESPONSE RESPONSE NOT OBTAINED

11. Check SIG tubes intact as follows:

a Check the foflowng EMFs NORMAL- a Q] EPI1IAI5000!E-3 (Steam Generator Tube Rupture)

. 1EMF-33(Condenser Air Ejector Exhaust)

. 1EMF-24(SIG A)

. 1EMF-25(S/G B)

. IEMF-26(S/G C)

. IEMF-27(S/G D) b if any S/G has previously been identified as ruptured THEN EPI1IAI5000JE-3 (Steam Generator Tube Rupture)

c. Notify RP to perform the following:

1) if S/G(s) fault known to be outside containment, THEN monitor area of steam fault for radiation.

2) Frisk all Unit I SIG cation columns to determine if activity level is significantly higher for any SIG

3) Notify Control Room of any abnormal radiation conditions.

d. WHEN activity results reported THEN notify station management to evaluate S/G activity

Q87

References:

MNS STEAM GENERATOR TUBE RUPTURE PAGE NO EP/1/AI5OOWE3 2 of 75 Rev. 23 UNIT ACTION/EXPECTED RESPONSE RESPONSE NOT OETAINED C. orAjjons I Monrtor Foldout page

2. Identify ruptured SIG(s: Perform the following:

. Any SIG NIR level GOING UP IN AN

- a. Continue to monitor S/G N/R levels and UNCONTROLLED MANNER tmtioe EMFs OR b. Dispatch operator and RP to check Unit 1 main eg)]ij in both exterior and

. Chemistry or RP has determined interior doghouses for activity to aid in ruptured SIG identifying ruptured SIG(s).

OR c. Notify RP to perform the following:

• Any of the following EMFs ABOVE

- . Frisk all Unit I S!G cation columns to NORMAL: determine if activity level is significantly higher for any SIG e IEMF-24(SIGA)

. Notify Control Room of survey

. IEMF-25(SIG B> results.

. IEMF-26 (SIG C) d WHEN wpturecl S/G(s) identified THEN immediately RETURNJD Step 3.

. 1EMF-27(SIG D)

_e iQStep1O 3 Check at least one S/G - AVAILABLE Maintain at least one S!G available for FOR NC SYSTEM COOLDOWN. NC System cooldown in subsequent steps.

Q87

References:

MNS STEAM GENERATOR TUBE RUPTURE PAGE NO.

EPI1IN5000IE-3 3 of 75 U]IT 1 Rev. 23 4.

r cz: E:iI :E::

Isolate flow from ruptured SIG(s) as RESPONSE NOT OBTAINED follows:

a. Check ruptured SIG(s) SM PC)RV - a WHEN wptured SIG pressure is less CLOSED than 1092 PSIG, THEN perform the following on affected SM PORV:

1) Check SM PORV closed.

2> ffi SM PORV is still open, fljN CLOSE its manual loader.

3) j SM PORV is still open, THEN perform the following:

a) CLOSE SM PORV isolation valve.

b) IF SM PORV isolation valve cannot be closed, THEN dispatch operator to CLOSE SM PORV isolation valve,

Q87

References:

MNS STEAM GENERATOR TUBE RUPTURE PAGE NO.

EPIl/N5000IE3 4 of 75 Rev. 23 UNiT 1 RESPONSE RESPONSE NOT OBTAINED ACTION/EXPECTED

4. (Continued) b Check SI lB and IC INTACT

- b Isolate TO CA pump steam supply from ruptured SIG as follows:

1) TO CA pump is the only source of feedwater, THEN maintain steam flowto it from at least one S/G.

2> Ensure operators dispatched in next step jediatel notify Control Room Supervisor when valves are closed.

3) Immediately dispatch two operators to concurrently verify (CV>, unlock and CLOSE valves on ruptured SIG(s):

• For1BS/G:

. ISA-78 (lB S/G SM Supply to Unit 1 TO CA Pump Turb Loop Seal Isol) (Unit 1 interior doghouse, 767+10, FF-53)

. ISA-2 (lB S/G SM Supply to Unit 1 TO CA Pump Turb Maint Isol) (Unit 1 interior doghouse, 767+12, FF-53>.

• For IC S/G:

• ISA-77(1CS/GSMSupplyto Unit 1 TO CA Pump Turb Loop Seal Isol) (Unit 1 interior doghouse, 767÷10, FF-53)

. iSA-I (IC S/G SM Supply to Unit 1 TO CA Pump Turb Maint Isol) (Unit I interior doghouse, 767+10, FF-53, above ladder).

4) IFATANYTlMElocalclosure of SA valves takes over 8 minutes, THEN isolate TO CA pump steam supply fj Enclosure 2 (Tripping TO CA Pump Stop Valve or Alternate Steam Isolation).

Q87

References:

MNS STEAM GENERATOR TUBE RUPTURE PAGE NO.

EPI1/A/50001E-3 5 of 75 Rev. 23 UNIT 1 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

4. (Continued)

c. Check toxdQwn isolation valves on ruptured S/Gs) CLOSED:

1) For 1AS/G 1) Perform the following:

IBB- lB (IA S/G alywr Cont a) CLOSE valve(s).

Outside lsol Control) b) CLOSE IBR-123(IAS/G

. 1BB-5A (AS/G BB Cont Inside JgwcLcwn Throttle Control).

Isol).

2) For1BS1G: 2) Perform the following:

. IBB-2B (lB SIG B[ovdown Cent a) CLOSE valve(s).

Outside Isol Control) b) CLOSE 1BB-124 (lB SIG

. IBB-BA (B SIG BB Corit Inside Throttle Control).

Isol).

3) ForICS/G: 3) Perform the following:

. 1BB-3B (IC S/G p.Cont a) CLOSE valve(s).

Outside Isol Control) b) CLOSE 1BB-125 (IC SIG

. 1BB-7A(CSIG BBContlnside Throttle Control).

Isol).

4) ForIDS!G: 4) Perform the following:

• IBB4B(IDS!GCont a) CLOSE valve(s).

Outside lsol Control) b) CLOSE IBB-126 (ID SIG

. IBB-8A(DSIGBBContlnside Throttle Control).

Isol).

d CLOSE steam drain on ruptured 516(s)

. ISM-83 (A SM Line Drain Isol)

• 1SM-89 (B SM Line Drain Isol)

. ISM-95 (C SM Line Drain Isol)

. 1SM-lol (D SM Line Drain Isol).

Q87

References:

MNS STEAM GENERATOR TUBE RUPTURE PAGE NO.

EPI1IN5000IE-3 6 of 75 Rev. 23 UNIT I ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

4. (Continued) e CLOSE the fol1owng valves Perform the foflowng ruptured S!G(s)

1) Place °STM PRESS 5lV CONTROLLER in manuaL

. MSIV bypass valve 2) Adjust STM PRESS CONTROLLER output to 0%.

3) Place STEAM DUMP SELECT in steam pressure mode.

4) Initiate Main Steam Isolation signal.

5) all SIG pressures are above 775 PSIG, THEN resetthe following to allow automatic SM PORV operation:

a MainIjglsolation a SMPORVs.

6) j ruptured SIG(s) MSIV and bypass valve are closed, THEN GO TO Step 5.

7) CLOSE the following valves on remaining S/

a MSIV a MSIV bypass valve.

8) any intact S/G MSIV and associated bypass valve closed, THEN QIQ Step 10) in this LQ.

9) §QjQ EP)1!AI5000IECA-3.1 (SGTR With Loss Of Reactor Coolant Subcooled Recovery Desired).

10) Select OFF RESET on steam dump interlock bypass switches.

11) Dispatch operator to atj CLOSE valves PER Enclosure 3 (Local Isolation SP Valves and Steam Drain Bypass Valves).

(RNO continued on next oaoe

Q87

References:

MNS STEAM GENERATOR TUBE RUPTURE PAGE NO.

EP/11A15000/E-3 8 of 75 Rev 23 UNIT ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

4. (Continued)

16) CLOSEfromControlRoomor dispatch operator to CLOSE the following valves:

. ISM-14(U1 Main Steam To CSAE 1501) (Unit I turbine 760+20, 1 [-1-31, west side of column)

. 1 TL-3 (SM To Steam Seal lsoi)

(Unit 1 turbine tLctg. 760+7, 1 D-33, northeast of DEH skid)

17) WHEN cooldown is initiated in subsequent steps, ]IiN use intact SIG(s) SM PORV for steam dump.

5. Control, ruptured SIG(s} level as fól Lows:

a Check ruptured SIG(s) NIR level - a Perform the followng GREATER THAN 11% (32% ACC)

1) !f any ruptured S/G is also faulted, THEN do not establish feed flow to the ruptured SIG unless needed for NC System cookiown.

2) IE any ruptured S/G is non-faulted is required for cooldown, THEN perform the following:

a) Establish and maintain feed flow to affected SIG(s).

b) WHEN affected S/G(s) N/R level greater than 11% (32% ACC),

THEN complete Step 5.b.

3) 12 Step 6.

Q87

References:

MNS STEAM GENERATOR TUBE RUPTURE PAGE NO.

EP!l/A15000!E-3 9 of 75 Rev. 23 UNIT 1 ACTIONJEXPECTED RESPONSE RESPONSE NOT OBTAINED

5. (Continued)

b. Isolate feed flow to ruptured SIG(s):

For 1ASIG:

1) CLOSE 1 CA-66AC (U 1 TD CA 1) Perform the following Pump Disch To IA SIG Isol) a) CLOSE ICA-64AB(UI TD CA Pump Disch To IA S/G Control).

b) Dispatch operator to CLOSE 1 CA-66AC (U 1 TD CA Pump Disch To 1A S/G Isol) (Unit I exterior doghouse, 750+8, FF-44, 4 .ftfrom inner wall, 8 it from column DD.-44).

c) IF exterior doghouse not accessible, OR CA cannot be isolated, THEN dispatch operator to unlock and CLOSE 1CA-63 (Unit I TD CA Pump Disch To IA S/G Control Inlet Isol) (Unit 1 CA pump jgj 716÷9, BB-51, above door to TD CA pump, 4 itsouth of 1A CA pump).

2) CLOSE 1CA62A(1A CA Pump 2) Performthe following Disch To IA S/G Isol) a) CLOSE ICA.-60A(IACA Pump Disch To IA SIG Control).

b) Dispatch operator to CLOSE 1CA-62A(1A CA Pump Disch To 1A SIG Isol) (Unit 1 exterior doghouse, 750+12, DD-44, southeast corner).

c) if exterior doghouse not accessible, OR CA cannot be isolated THEN dispatch operator to unlock and CLOSE I CA-59 (IA CA Pump Disch To IA S/G Control Inlet Isol) (Unit I CA pump fgj, 716+10, CC-SO, above lB CA Pump).

Q87

References:

MNS STEAM GENERATOR TUBE RUPTURE PAGE NO.

EPIIIAI5000IE-3 13 of 75 Rev. 23 UNIT I ACTIONIEX:EEED REsPE OBTAINED

6. Check ruptured SIG(s) pressure - Perform the following:

GREATER THAN 280 PSIG.

a. IF ruptured S!G SM PORV is open.

AND is in process of being locally isolated, THE perform the following:

1) WHEN selecting target temperature in Step 9, assume ruptured S!G pressure is between 280 PSIG and 299 PSIG, until ruptured S/G pressure recovers to greater than 300 PSIG.

_2) !FATANY.T!MEitis determined that ruptured S!G SM PORV cannot be isolated. OR pressure cannot be recovered to greater than 280 PSIG after isolation. THEN GO TO EP/1!A/5000!ECA-3 1 (SGTR With Loss Of Reactor Coolant -

Subcooled Recovery Desired)j

3) GQ TO Step T b GOTO EPII/A15000/ECA-31 (SGTR With Loss Of Reactor Coolant Subcooled Recovery Desired).

7. Check any NC pump RUNNING.

CAUTION NC T-Cold indication in the ruptured loop may cause an invalid Integrity Status Tree condition.

Disregard NC T-CoId indication in the ruptured loop, until directed by this EP or until this EP is exited.

Q87 Parent Question:

FOR REVIEW ONLY DO NOT DISTRIBUTE -

2011A MNS SRO Audit Examination QUESTION 86 I APEO4O AAIO1 Steam Line Rupture

)bthtw determine and tot elpret the following ac they apply to the Steam Line Rupture: (CFR: 415 4113)

OcculTence and location of a steam line rupture from pressure nd flow indications Given the thllov4ng conditions on Unit I:

A Reactor Trip and Safety tnjectbn have occurred

• The crew is perforning actions contained in E0, Reactor Trip or Safety Injection

• IA Main Steam Line Radiation Monitor was increasing rapidly prior to the tip

• 1A SG pressure is dropping uncontrollably, and continues to drop to less than 200 PSIG after Main Steam Line Isolation

• lB. 1C, and I D SG Main Steam Line pressures are stable Which ONE (I) of the following describes the procedure flowpath for this event when the crew makes a transition from E-0?

A. E-2 (Faulted SG lsolatbn) to E4 (Steam Generator Tube Rupture) to ES4.1 (Post SGTR Cooldown Using Bacilli)

B. E-2 (Faulted SG Isolation) to E4 (Steam Generator Tube Rupture> to ECA-31 (SGTR with Loss of Coolant Accident StcooIed Recovery Desired)

C E-3 (Steam Generator Tube Rupture) to E2 (Faulted SG Isolation) to E1 (Loss of Reactor or Secondary Coolant) to ES1 1 (SI Teimination)

0. E-3 (Steam Generator Tube Rupture> to ECA4.i (SGTR with Loss of Coolant Accident Subcooled Recovery Desired)

Thrada,Mcty 26, 2011 Page 1S4 of 2I

Q87 Parent Question:

FOR REVIEW ONLY DO NOT DISTRIBUTE -

2011A MNS SRO Audit Examination QUESTION 86 I General Discussion IThe fret priority in 5-0 is to check for a loss of coolant ia a stack open PZR PORV in which case a nonsition would be made to 5-1. Since one not exist in this case, the next wansition is to check all SOs Main Steam lines intact Since the IA 513 is ihahed. 5-0 directs nensition to 5-2 for steps to isolate the faulted SO. When the isolation is complete. at the end of 5-2 a check is made to see lithe SO ashes are intact Since there indications of a 513Th. 5-2 directs nansition to 5-3. Since the aflcted SOis both ruptured and faulted 5-3 will eventually direct u-ansition to CA-3.1.

Answer A Discussion IDICORB.ECT.

Answer B Discussion ICORRECT.

Answer C Discussion lCORRECT.

Answer D Discussion INCORRECT Basis for meeting the KA Basis for Hi Cog Basis for SRO only I I Job Level Cognitive Level QuestionType Question Source SRO Contarehension RANK 2008 5005 Audis Exam Q7d Development References Student References Provided APEO4O AA2.01 Steam Line Rupture Abilin to deteriinne and intetprer the following as they apply to the Steam Line Rupture: (CFR: 43.3 45.13)

Occun-etace and location of a steam line rupture from pressure and flats- inthcations 01-9 Comments: emarkslStatus Thursday, May 2 2011 Page 18 of 21

FOR REVIEW ONLY DO NOT DISTRIBUTE -

ILT-30 MNS SRO NRC Examination QUESTION 88 [ 88 APEO56 AA2.55 Loss of Offsite Power Ability to determine and interpret the following as they apply to the Loss of Offsite Power: (CFR: 43.5 / 45.13)

Subcooled margin monitors Given the following conditions:

• A Blackout has occurred on Unit I

• ECA-O.O, (LOSS OF ALL AC POWER) has been implemented

• The Subcooling Margin Monitor (SMM) indicates that NC Subcooling is (-)I °F

• A RED Path exists on the Heat Sink CSF Status Tree Subsequently, the following events occur:

• Power is restored to I ETA from SATA aligned to Unit 2

• The crew is preparing to transition to the appropriate recovery procedure Procedure Legend:

FR-H.1 (RESPONSE TO LOSS OF SECONDARY HEAT SINK)

ECA-O.1 (LOSS OF ALL AC POWER RECOVERY WITHOUT SI REQUIRED)

ECA-O.2 (LOSS OF ALL AC POWER RECOVERY WITH SI REQUIRED)

Which ONE (1) of the following describes the recovery strategy?

A. Transition to FR-H.1 upon exit from ECA-O.O and perform ECA-O.2 after FR-H.I is complete.

B. Transition to FR-H.1 upon exit from ECA-O.O and perform ECA-O.1 after FR-Hi is complete.

C. Transition to ECA-O.2 and enter FR-H.1 when directed by ECA-O.2.

D. Transition to ECA-O.1 and enter FR-H.1 when directed by ECA-O.1.

Thursday, May 29, 2014 Page 261 of 298

FOR REVIEW ONLY DO NOT DISTRIBUTE -

ILT-30 MNS SRO NRC Examination QUESTION 88 88 General Discussion Given the conditions in the stem of the question, the applicant should determine that monitoring of CSFSTs is for information only and that transition to Functional Restoration Procedures is not allowed until procedurally directed in either ECA-0. 1 or ECA-0.2.

Answer A Discussion INCORRECT: See explanation above.

PLAUSIBLE:

In most instances in the EOP network, FRPs take priority over all other procedures. And, since power has been restored to one emergency bus, it would be easy for an applicant to conclude that implementation of a RED Path FRP would take priority. ECA-0.2 is the correct recovery procedure so, if the applicant concluded that FR-H.l should be performed first, they would conclude that ECA-0.2 should be implemented next.

Answer B Discussion INCORRECT: See explanation above.

PLAUSIBLE:

In most instances in the EOP network, FRPs take priority over all other procedures. And, since power has been restored to one emergency bus, it would be easy for an applicant to conclude that implementation of a RED Path FRP would take priority. The difference between implementing ECA-0. 1 and ECA-0.2 upon transition from ECA-0.0 is a very common error among Licensed Operators. If is common for Licensed Operators to conclude that if SI has not been initiated prior to exit from ECA-0.0 that Safety Injection is not required and they transition to ECA-0. 1. It is plausible for a License Applicant to make the same mistake.

Answer C Discussion CORRECT: See explanation above.

Answer D Discussion INCORRECT: See explanation above.

PLAUSIBLE:

This answer is plausible because transition to one of the ECA procedures is the correct action. Additionally, the difference between implementing ECA-0.1 and ECA-0.2 upon transition from ECA-0.0 is a very common error among Licensed Operators. If is common for Licensed Operators to conclude that if SI has not been initiated prior to exit from ECA-0.0 that Safety Injection is not required and they transition to ECA-0. 1. It is plausible for a License Applicant to make the same mistake.

Basis for meeting the K The K/A is matched because it requires the applicant to interpret the indications on the Subcooling Margin Monitor (SMM) to determine the correct procedural flowpath related to a Loss of Offsite Power.

Basis for Hi Cog This is a higher cognitive level question because it requires more than one mental step.

First, it requires the applicant to analyze the conditions given in the stem and determine that, based on subcooling, Safety Injection is required.

Next, the applicant must recall from memory the procedural requirements from ECA-0. I and ECA-0.2 to determine that those procedure have priority over FR-H. I for implementation.

Basis for SRO only This question meets the following criteria for an SRO only question as described in the Clarification Guidance for SRO-only Questions Rev I lated 03/11/2010 for screening questions linked to 10CFR55.43(b)(5) (Assessment and selection of procedures):

1) The question can NOT be answered solely by knowing systems knowledge.

AJI of the knowledge required to answer this question is related to procedure content.

2) The question can NOT be answered by knowing immediate operator actions.

There are NO immediate actions associated with the procedures in question.

3) The question can NOT be answered solely by knowing entry conditions for AOP or direct entry conditions for EOPs.

This question is not related to entry conditions since none of the procedures are direct-entry procedures. The knowledge is related to procedure transition requirements related to the procedures.

4) The question can NOT be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of the procedure.

This is detailed knowledge of procedure content not sequence of events within the procedure or overall mitigative strategy.

5) This question requires the applicant to perform a diagnosis which will determine to which contingency procedure a transition should be made.

Therefore, it is SRO knowledge.

Thursday, May 29, 2014 Page 262 of 298

SRO FOR REVIEW ONLY DO NOT DISTRIBUTE ILT-30 MNS SRO NRC Examination QUESTION 88 Job Level Cognitive Level Comprehension QuestionType BANK Question Source c

88 2011 MNS SRO Audit Examin Q87 (Bank 4497)

Development References Student References Provided

References:

ECA-0. I (LOSS OF ALL AC POWER RECOVERY WITHOUT SI REQUIRED)

ECA-0.2 (LOSS OF ALL AC POWER RECOVERY WITH SI REQUIRED)

Learning Objectives:

EPECAOO6 APEO56 AA2.55 Loss of Offsite Power Ability to determine and interpret the following as they apply to the Loss of Offsite Power: (CFR: 43.5 / 45.13)

Subcooled margin monitors 401-9 Comments: RemarkslStatus Thursday, May 29, 2014 Page 263 of 298

Q88

References:

MNS LOSSOFALLACPOWER PAGENO EP/1/N5000IECA-00 2 of 309 Rev.34 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED C. Qorons CSF Status trees should be monitored for information ony. EPs referenced by them should not be implemented.

() Check Reactor Trip: Trip reactor.

. All rod bottom lights LIT a Reactor trip and bypass breakers -

OPEN

. hR amps GOING DOWN.

Check Turbine Trip: Perform the following:

. All throttle valves CLOSE [1 a. Trip the turbine

b. IF turbine will not trip, THEN perform the following:

1) Place turbine in manuaL 2> CLOSE governor valves in fast action.

33 governor valves will not close, THEN CLOSE the following valves:

a All MSlVs a All MSIV Bypass Valves.

Q88

References:

MNS LOSS OF ALL AC POWER PAGE NO.

EP/1/N5000/ECA-0.0 42 of 309 Rev. 34 UNIT I ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 41, Select recovery procedure as follows:

a. Check standby makeup pump ON. .- a. Perform the following:

1) Notify station management that further coohng of the NC pump seals will be established by natural gjg cooldown in subsequent EPs.

2) QjStep41.c.

b Pnor to startinq NC pump(s) in subsequent EPs, ensure status evaluation is performed since seal cooling was temporarily Lost at start of event.

a. Check NCsubcooling based on core c. Perform the following:

exit T/Cs GREATER THAN 0F.

1) Align additional RN valves PER Enclosure 29 (RN S/I \falve Alignment).

2) GO TO EP[1!Ai5000!ECA-0.2 (Loss Of All AC Power Recovery With S/I Required).

d. Check Pzr level GREATER THAN d Perform the following.

11% (29% ACC).

1) Align additional RN valves PER Enclosure 29 (RN Sl Valve Alicinment).

2) GO TO EP/1/AI5000IECA-0.2 (Loss Of All AC Power Recovery With 511 Required).

e. Check the following valves CLOSED: e. any NV pump on, THEN perform the following:

lNl-9A (NC Cold Leg j From NV)

1) Align additional RN valves,E a INI-IOB (NC Cold Leg tg,j From NV). Enclosure 29 (RN S/I Valve Alignment).

2) GO TO EP/1/A/5000/ECA-0.2 (Loss Of All AC Power Recovery With S/I Required).

Q88

References:

LOSS OF ALL AC POWER RECOVERY WITH PAGE NO REQUIRED 2 of 33 EP/1!N5000IECA-02 Rev13 UNTTI EXPECTED RESPONSE RESPONSE NOT OETA1NED C. Operator t CSF Status trees should be monitored for information only. EPs referenced by them should not be implemented until directed by this procedure.

2. Reset SlI. Reset SIL PER EPII!AI5000IG-1 (Generic Enclosures). Enclosure 23 (Local Reset of 511 Signal).

3. Check FWST level GREATER THAN IQ Enclosure I (ECCS Alignment 95 INCHES. Below FWST Lo Level).

0 00 m

-o CD

zi ) ci, CD 0

rn:

C) g C) CD 9

Cfl!:: grri 0

0 0

ri c_)

F z

cm m

0 (I,

-l rn c)p

Q88 Parent Question:

FOR REVIEW ONLY DO NOT DISTRIBUTE -

2011A MNS SRO Audit Examination QUESTION 87 PEO55 2L2O Lost of Of5it and Ouite Power (Station Blickont)

EPEO5S GENERIC Abiliw to inteTpret and enecute procechir tpc (CFR: 41.10 43.5 45.12)

Given the following conditions:

A Blackout has occurred on Unit I

• ECA-O.O, (Loss of All AC Power) has been implemented

• NC Subcooling is (-)i F

• A RED Path exists on the Heat Sink CSF Status Tree Subsequently, the following events occur

• Power is restore to I ETA from SATA aligned to Unit 2

• The crew is preparing to transition to the appropriate recovery procedure Which ONE (1) of the following describes the recovery strategy?

A. Transition to FR-Hi upon exit from ECA-O.O. Perform ECA-Oi, Loss of All AC Power Recovery Without SI Required, when FR-H, i is complete.

B. Transition to FR-Hi upon exit from ECA-O.O. Perform ECA-O.2, Loss of All AC Power Recovery with SI Required, when FR-Hi is complete.

C. Transition to ECA-O. 1, Loss of All AC Power Recovery Without SI Required.

and enter FR-H. I when directed by ECA-O.l.

0. Transition to ECA-O.2, Loss of All AC Power Recovery with SI Required, and enter FR-H.I when directed by ECA-O.2.

Thtn,day, kfai 26, 2011 Page 186 of 21f

Q88 Parent Question:

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2011A MNS SRO Audit Examination QUESTION 87 General Discussion Answer A Discussion

[A and B are hico ctbecaure CSFST are o onitored but not addreued undl allowed in appropriate recoveryprocedure I Answer B Discussion

[A and B are incorrect became CSFST are o1v monitored but nor adc1reted until allowed in appropriate recovery procedure I Answer C Discussion jC Sr incorrect became rubcciolinr ii 0 deareer F. requirim SI Answer D Discussion Basis for meeting the KA Basis for Hi Cog Basis for SRO only Job Level Cognitive Level CuestionType Question Source 3RD Conitrehenrion BANK 2007 1GS NRC QIl Development References I5fent References Provided EPE055 11.20 Lot: of Offtite and Ontiae Power cStatienBiaekou EPEG5S GENERIC Ability to interpret and execute procedure trept. (CFR: 41.10 43.5 45.12) 401-9 Comments: emathsfStatus Thurrday, May 26 2011 Pae1S7of2I

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ILT-30 MNS SRO NRC Examination QUESTION 89 89 APEO65 AA2.05 Loss of Instrument Air Ability to determine and interpret the following as they apply to the Loss of Instrument Air: (CFR: 43.5 / 45.13)

When to commence plant shutdown if instrument air pressure is decreasing Given the following plant conditions:

• Both units are at 100% RTP

• Both units have implemented AP-22 (LOSS OF VI) due to decreasing VI Header pressures In accordance with AP-22,:

1) what is the MINIMUM VI Header pressure below which the crew is directed to align RN per Enclosure 7 (RN ALIGNMENT DURING LOSS OF VI EVENT)?

2) which of the following conditions will FIRST require the crew to initiate a manual reactor trip?

A. S/G levels begin DECREASING in an uncontrolled manner B. Pressurizer level begins INCREASING in an uncontrolled manner A. 1. 6OPSIG 2.A B. 1. 85PSIG 2.A C. 1. 6OPSIG

2. B D. 1. 85PSIG

2. B Thursday, May 29, 2014 Page 264 of 298

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ILT-30 MNS SRO NRC Examination QUESTION 89 89 General Discussion AP-22 directs the operators to align RN to the SNSWP if VI Header pressure decreases to less than 60 PSIG.

AP-22 contains a continuous action step that direct the Operators to initiate a reactor trip and implement E-0 if S/G levels are decreasing in an uncontrolled manner.

Answer A Discussion CORRECT: See explanation above.

Answer B Discussion INCORRECT: See explanation above.

PLAUSIBLE:

The first part is plausible because this is the pressure below which the Air Dryers will be automatically bypassed.

The second part is correct.

Answer C Discussion

[rNC0RRECT: See explanation above.

PLAUSIBLE:

The first part is correct.

The second part is plausible because an uncontrolled increase in Pressurizer level is an expected condition during a loss of VI and AP-22 provides specific direction to control Pressurizer level using Enclosure 10. Additionally, an automatic reactor trip will occur if Pressurizer level increases to greater than 92%. Therefore applicant could conclude that both instances require a manual reactor trip.

Answer D Discussion INCORRECT: See explanation above.

PLAUSIBLE:

The first part is plausible because this is the pressure below which the Air Dryers will be automatically bypassed.

The second part is plausible because an uncontrolled increase in Pressurizer level is an expected condition during a loss of VI and AP-22 provides specific direction to control Pressurizer level using Enclosure 10. Additionally, an automatic reactor trip will occur if Pressurizer level increases to greater than 92%. Therefore applicant could conclude that both instances require a manual reactor trip.

Basis for meeting the K The KIA is matched because it requires the applicant to have knowledge of what conditions require a manual reactor trip during a loss of Instrument Air.

Basis for Hi Cog Basis for SRO only This question meets the following criteria for an SRO only question as described in the Clarification Guidance for SRO-only Questions Rev I dated 03/11/2010 for screening questions linked to IOCFR55.43(b)(5) (Assessment and selection of procedures):

1) The question can NOT be answered solely by knowing systems knowledge.

The knowledge required to answer this question is beyond systems level knowledge it is procedure content knowledge.

2) The question can NOT be answered by knowing immediate operator actions.

There are NO immediate actions associated with AP-22.

3) The question can NOT be answered solely by knowing entry conditions for AOP or direct entry conditions for EOPs.

The required knowledge is not related to AP-22 entry conditions rather knowledge of the content of AP-22.

4) The question can NOT be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of the procedure.

This is detailed knowledge of procedure step content, not sequence of events within the procedure, or overall mitigative strategy.

5) The question requires the applicant to have knowledge of diagnostic steps and decision points within the procedure which require the applicant to initiate actions based on a specific set of conditions (i.e. initiating a reactor trip and transitioning to another procedure AND initiating actions to align N2 backup to the PZR PORV5). Therefore, it is SRO level knowledge.

Thursday, May 29, 2014 Page 265 of 298

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ILT-30 MNS SRO NRC Examination QUESTION 89 Job Level Cognitive Level fQuestionTYPe Question Source SRO Memory NEW Development References Student References Provided

References:

AP-22 (LOSS OF VI)

Learning Objectives:

OP-MC-AP-22 Objective 8 APEO65 AA2.05 Loss of Instrument Air Ability to determine and interpret the following as they apply to the Loss of Instrument Air: (CFR: 43.5 / 45.13)

When to commence plant shutdown if instrument air pressure is decreasing 401-9 Comments: Remarks/Status Thursday, May 29, 2014 Page 266 of 298

Q89

References:

MNS LOSS OF VI PAGE NO.

AP/11A15500122 2 of 145 Rev. 32 UNIT 1 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED B. 2!ii!

• Abnormally low VI pressure

• VI pressure GOING DOWN

• VI COMP PNL TROUBLE alarm

• VINS LO PRESS alarm

• VlNS LOJ,,QPRESSalarm

• Erratic plant instrumentation andlor control

• Loss of KR flow to VI compressors.

C. Operator Actions 1 IF AT ANY TIME VI pressure is less than 60 PSIG, THEN align RN PER Enclosure 7 (RN Alignment During Loss of VI Event).

2. Announce occurrence on page.

3. Ensure at least 2 KR pumps running.

4 Ensure at least 2 KC pumps running.

Q89

References:

MNS LOSSOFVI PAGENO.

APIIIAJ55O0122 8 of 145 Rev. 32 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

12. (Continued)

m. Control NC temperature as follows:

. THROTTLE ND flow.

NOTE

• KC to ND Hx flow should be close to flow prior to loss of VI, since it is normally controlled by motor operated valves.

KG to ND Hx flow indications fail low during a loss of VI. Alternate indications are available at the following locations, if needed:

• 1A: 1 KCFT5670 (aux 733 ÷2, west of column MM-S4)

• 1B: 1KCFT-5680 (auxi.g, 733 +4, west side of columnJJ55).

a NC temperature is greater than 2O0F, THEN maintain KC flow to ND Hx greater than 2000 5PM.

a THROTTLE KG Flow to ND Hx as required.

NOTE CF Control Valves will fail closed on low VI pressure, which may result in AMSAC actuation arid Lo L. Ss level

13. Check SIG levels AT PROGRAMMED

- IF SIC levels are going down in an LEVEL, uncontrolled manner, THEN perform the following:

a Trip reactor.

b. Continue with this procedui e as time allows.

c. GO TO EP!1!A15000!E-0 (Reactor Trip or Safety lnjection).I

Q89

References:

MNS LOSS OF VI PAGE NO.

AP111A15500122 10 of 145 Rev. 32 UNIT I ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

20. Check CA pumps OFF. - IF CA flow goes up in an uncontrolled manner, THEN implement CA flow control criteria on Foldout page.

NOTE

• Pzr heaters will energize if Pzr level is 5% greater than control setpoint.

• Pzr spray is not available on loss of VI.

21. Operate Pzr heaters as required to:

a Maintain Pzr liquid space temperature at Pzr vapor space temperature.

. Minimize opening of Pzr PORVs.

NP1

• 1 NV-238 (UI Charging Control) and 1 NV241 (UI Seal Water Flow Control) fail open on loss of VI.

• Normal and excess letdown isolate on loss of VI. Enclosure 10 (Pressurizer Level Control) contains actions to limit Pzr fill rate.

22,

23. Check VI pressure LESS THAN 60

- Perform the following:

PSIG.

a. I TAYiVI pressure goes below 60 PSIG, THEN perform steps 24 through 32.

b. Step 32.

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ILT-30 MNS SRO NRC Examination QUESTION 90 APEO61 2.1.19 Area Radiation Monitoring (ARM) System Alarms APEO61 GENERIC Ability to use plant computers to evaluate system or component status. (CFR: 41.10 / 45.12)

Given the following plant conditions:

• The plant is experiencing and unexplained increase in Radiation Monitor indications

• RP is currently attempting to determine the source of the increased radiation levels

• The OSM is preparing to declare an Emergency based on observation of the following OAC EMF Graphic indications:

TIME Units EMF 08:30 08:50 09:10 1EMF-1 6.OOE+02 2.OOE+03 1.OOE+04 2EMF-4 1.50E+00 2.50E+00 3.00E+01 mRlhr IEMF-5 3.OOE+02 2.50E+03 3.80E+04 2EMF-9 2.OOE+03 2.OOE+04 3.OOE+04 1EMF-17 t5OE+02 6.00E+02 5.50E+03 EMF-36L 2.OOE+02 4.50E+04 I .OOE+06 CPM EMF-36H 1.80E+02 3.80E+03 4.80E+03 The Emergency Classification in accordance with RP-000 (EMERGENCY CLASSIFICATION) based on OAC radiation monitor indications at 08:30 is (1)

After the Initial Notification, the OSM re-evaluates the OAC radiation monitor indications at 09:10 and the Emergency Classification is (2)

Which ONE (1) of the following completes the statements above?

REFERENCE PROVIDED A. 1. Unusual Event

2. Alert B. 1. Alert

2. Site Area Emergency C. 1. Unusual Event

2. Site Area Emergency D. 1. Alert

2. Alert Thursday, May 29, 2014 Page 267 of 298

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ILT-30 MNS SRO NRC Examination QUESTION 90 90 General Discussion Based on the EMF OAC Graphics provided and RP-000 (EMERGENCY CLASSIFICATION) the applicant should determine that, at 08:30 the Emergency Classification is an Unusual Event. This is based on Enclosure 4.10 (Radiation Monitor Readings for Enclosure 4.3 EALs) of RP 000. Analyzing the 08:30 EMF OAC Graphic would reveal that EMF-l indicates >500 mR/hr which results in an Unusual Event.

Upon re-evaluation at 09:10, the applicant should determine that the Emergency Classification is now a Site Area Emergency. This is based on EMF-36H being >3.4E+03 for greater than 15 minutes. To make this determination the applicant must look at the EMF OAC graphic from 08:50 at which time EMF-36H indication is already >3.4E+03 (actually 3.8E+03) and then at 09:10 EMF-36H is still >3.4E+03 (actually 4.8E+03). In accordance with RP-000 4.3.S.1-1, this results in a Site Area Emergency.

Answer A Discussion INCORRECT: See explanation above.

PLAUSIBLE:

The first part is correct.

The second part is plausible because EMF-1 indication alone would result in classification of an Alert (>500 mR/hr).

Answer B Discussion INCORRECT: See explanation above.

PLAUSIBLE:

The first part is plausible if the applicant misreads the radiation monitor readings or misapplies RP-000 to the indications provided.

The second part is correct.

Answer C Discussion

[CORRECT: See explanation above.

Answer D Discussion INCORRECT: See explanation above.

PLAUSIBLE:

The first part is plausible if the applicant misreads the radiation monitor readings or misapplies RP-000 to the indications provided.

The second part is plausible because EMF-1 indication alone would result in classification of an Alert (>500 mRlhr).

Basis for meeting the K The K/A is matched because it requires the applicant to use plant computer indications to evaluate the plants Emergency status.

Basis for Hi Cog This is a higher cognitive level question because it requires more than one mental step.

First, it requires the applicant to evaluate the radiation monitor indications provided to determine the plant Emergency Action Level.

Next, it requires the applicant to evaluate a change in radiation monitor indications to determine if the Emergency Action Level status has changed and, if so, what the new Emergency Action Level is.

Basis for SRO only This question is SRO level because it requires the applicant to evaluate plant conditions and determine the Emergency Action Level based on application of RP-000 (Emergency Classification).

Job Level Cognitive Level QuestionType Question Source SRO Comprehension NEW Development References Student References Provided

References:

RP-000 (CLASSIFICATION OF EMERGENCY)

RP-000 (EMERGENCY CLASSIFICATION)

Learning Objectives:

OP-MC-EP-EMP Objective 2 APEO6 I 2.1.19 - Area Radiation Monitoring (ARM) System Alarms Thursday, May 29, 2014 Page 268 of 298

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ILT-30 MNS SRO NRC Examination QUESTION 90 90 APE06 1 GENERIC Ability to use plant computers to evaluate system or component status. (CFR: 41.10 /45.12) 4O1 9 Comments Remarks!Status Thursday, May 29, 2014 Page 269 of 298

Q90

References:

Enclosure 4.3 RP/01A15700!000 Abnormal Rad Levels/Radiological Effluent Page 2 of 5 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY 4.5.tJ-2 A valid indi:attc a Ca radiation iuctitor 4.3.A.l -2 A i alid inji:atic a or Note 2: If doe as.e--scent leant Note 2: If dose assessment team Lo. 2ti5F-ct4.pmfw 1

-3 radittcnunucw EMF 3rL cakulasons ctrltit be calculations cannot be 60 u-rues Ci 15 iii likely ilantinils fir of: 2 USE*16 cpni thr rcii.t bird ni 15 iiiiiil - -, completed in 15 ntinutes, 60 niri3e s.iuh in,hites that the 15 minutes Oi is ill likely inca valid monitor ratlitg then valid monitor reitnie man have exiteded the untiatuts c,.ntni,e for IS mmuftei should he used toe reading should be used cci,dttin and indtcates tw nerd to asess sslndt uthcates hit the release enlerzencs classrlinitlotx for emergency tie ielea-m- with piori-duit mac tiave exceeded the inthatinp classification H? 011 1009 211. II? 0112 1000/029, or condition and in:licates the need L3,s.I-l A s-flirt iltdii5tOli 590 0 i 2003 011. to iises the release with nit ta.lnti.ii ntcnttc.r 4.3.GJ-1 A valid indication precedure 19? 0 d 1109.010, F tF iON cf on radiation monitor LOU 1 0 ialid indicaton on radiation monitor 1-19011 1009/119 or 410?cpni 5

_ EMF-3611 of L5sU -31 (when alicited to WC or SH 0 B 2005/101 si,tamnl fri j34 E + 04 cpnl W(B of 1) [N --sS cpmfce I it_nilitci sustained lbr 260 IUtiLitCS 01 s-ill likels- (OiitiUlie tot 43.A.1-3 Gaseous effluent being released >15 minutes.

60 micIte. wlch nadicate th,t the exceeds 200 times the level of 4.0.s.1-2 Di- ssst inei,t traIls release nnv rn: e exceeded ti_c uitiatuig SLC 16.1 1-6 for> 15 minutes as ralnn90t:ons indicate L3.GJ2 Dose assessment et_dition and mOnte tile wed toasgss determined by Radiation .1- coit-r-quences team calculations ti_c trlcaic v.1111 proecd.re Protection (RP) procedure. greater than 100 indicate dose lOP OR 1139 Ott, H? OR 10001(29,01 mRem TEDE or 500 consequences 590cR 2035 OOi 4.3.A,1-4 Liquid effluent being released mRem CDE Adult greater than 1000 exceeds 200 times the level 0 f Thyroid at the site mRem TEDE or LOU 1-4 Gre.c4 rfiuot being released exceeds SLC 16.11-I for>l5 minutes as botandar. 5000 mRem CDE two times SLC 1611 Ofcr determined by Radiation Adult Thoid at the 60 minute- as determined by Radiation Protection (RI) procedure 4,3SJO Analysis of field site boundary.

Psatectnn (P.9) pnocedcte survey results or field (Continued) siuvey samples 4.3Gj-3 Analysis of field LOU 1-5 Liqrid efth...it know m,lea-t-d exceeds mdrcates dose survey results or nsotrntesSL( lOll-I fm consequences greater field survey samples 00 mnniutei. deteit,inted by Radiation than 100 mRem indicates dose

stectioit (tIP) procedire. TEDE or 500 mRern consequences IC riiiliuir.l) CDE Adult Thyroid greater than 1000 at the site boundary. mReni TEDE or END 5000 mRem CDE Adult Thyroid at the site boundary.

END

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ILT-30 MNS SRO NRC Examination QUESTION 91 APEO68 2.2.44 Control Room Evacuation APEO68 GENERIC Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions. (CFR: 41.5 / 43.5 / 45.12)

Given the following plant conditions:

• Units I and 2 are in Mode 1 with B Train equipment in service

• A fire has occurred in the Control Room

• The fire has been contained but flames are still present

• AP-45 (PLANT FIRE) and AP-24 (LOSS OF PLANT CONTROL DUE TO FIRE OR SABOTAGE) have been implemented on both units

• The control room crew has just completed all required AP-24 manual actions prior to control room evacuation Subsequently,

• The Unit I OATC observes two main turbine throttle valves indicating 12% open Based on the conditions above, the assured Safe Shutdown Train is (1)

In accordance with AP-24, the Unit I OATC will be required to (2)

Which ONE (1) of the following completes the statements above?

A. 1. SSS

2. initiate a Main Steam Isolation B. 1. TrainAorB

2. initiate a Main Steam Isolation C. 1. SSS

2. take manual turbine control and close throttle valves D. 1. TrainAorB

2. take manual turbine control and close throttle valves Thursday, May 29, 2014 Page 270 of 298

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ILT-30 MNS SRO NRC Examination QUESTION 91 9J General Discussion Per AP-45, PLANT FIRE, the assured safe shutdown train for a fire in the control room is the SSS (Train C).

Per AP-24, all throttle valves CLOSED is the indication used to determine if the main turbine is tripped. IF Unit 1 turbine will not trip, Then a Main Steam Isolation is initiated.

Answer A Discussion CORRECT: See explanation above.

Answer B Discussion INCORRECT: See explanation above.

PLAUSIBLE:

First part is plausible since Train A, Train B, Train A or B, and the SSS are all possible Safe Shutdown Train choices depending on the location of the fire.

Second part is correct and therefore plausible.

Answer C Discussion INCORRECT: See explanation above.

PLAUSIBLE:

First part is correct and therefore plausible.

Second part is plausible since these actions would be the correct actions after failure of the main turbine to trip in E-0 and FRS-l.

.nswer D Discussion INCORRECT: See explanation above.

PLAUSIBLE:

First part is plausible since Train A, Train B, Train A or B, and the SSS are all possible Safe Shutdown Train choices depending on the location of the fire.

Second part is plausible since these actions would be the correct actions after failure of the main turbine to trip in E-0 and FRS- I.

Basis for meeting the K The KJA is matched since the applicant is required to interpret plant conditions using control room indications to verify the status of the unit I main turbine and determine the correct procedural actions to place the unit in the desired configuration prior to control room evacuation.

Basis for Hi Cog Basis for SRO only This question meets the following criteria for an SRO only question as described in the Clarification Guidance for SRO-only Questions Rev I dated 03/11/2010 for screening questions linked to IOCFR55.43(b)(5) (Assessment and selection of procedures):

I) The question can NOT be answered solely by knowing systems knowledge.

The knowledge required to answer this question is beyond systems level knowledge it is procedure content knowledge.

2) The question can NOT be answered by knowing immediate operator actions.

3) The question can NOT be answered solely by knowing entry conditions for AOP or direct entry conditions for EOPs.

4) The question can NOT be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of the procedure.

This is detailed knowledge of procedure step content, not sequence of events within the procedure, or overall mitigative strategy.

5) The question requires the applicant to have knowledge of diagnostic steps and decision points within the procedure which require the applicant to initiate actions based on a specific set of conditions.

Job Level Cognitive Level QuestionType Question Source SRO Memory NEW Thursday, May 29, 2014 Page 271 of 298

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ILT-30 MNS SRO NRC Examination QUESTION 91 91 Development References Student References Provided -

References:

AP-45 (PLANT FIRE)

AP-24 (LOSS OF PLANT CONTROL DUE TO FIRE OR SABOTAGE)

Learning Objectives:

OP-MC-AP-45 Objective 4 APEO68 2.2.44 Control Room Evacuation APEO68 GENERIC Ability to interpret control room indications to verifj the status and operation of a system, and understand how operator actions and directives affect plant and system conditions. (CFR: 41.5 / 43.5 / 45.12)

[4O19 Comments: RemarkslStatus Early Submitall Comments from CE:

1. SRO-only: The proposed question does not test one of the seven 10CFR55.43(b) topics. The question information sheet indicated that procedure selection is being tested: however.

neither of the fill-in-the-blank statements deal with procedure selection.

I tried developing a question to hit the K/A and hit procedure selection.. .but was only able to get this far (you may be able to complete the effort see below)

Given the following plant conditions:

- The crew has entered AP-24, Loss of Plant Control Due to Fire or Sabotage.

- The SSF has NOT been activated due to a fire.

- The SRO is evaluating Step 3:

WOOTF completes both statements?

If the only method of safely shutting down requires using a combination of Train A and Train B equipment, THEN the answer to Step 3is .(NOv5YES)

Need another fill-in-the-blank to target PROCEDURE SELECTION; however, I dont think the K/A lends itself to this.

This is a tough K/A to hit at the SRO level. Suggest replacing K/A; contact Chief Examiner.

Obtained new KA, randomly selected by CE. Wrote new question based on new KA. HCF 04/21/14.

Thursday, May 29, 2014 Page 272 of 298

Q91

References:

MNS PLANTFtRE PAGENO.

AP101A15500145 8 of 203 Rev. 16 UNfl ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

4. (Confinuecl)

Fire Area Assured Enclosure(s) Shutdown Train S c osure n 750 - - a S r a a (AS /50 - Ca Ic SSS Roon °F e 0 cad Rourn Fre Fare rea.9 ntad2Ac.os) c sr 8 AS /5O t 2 Ca e o ea (.B 50 U 2 Co e SS Ro (EF Zo 2 S rca -

g corn Fre I F re Area 20 n 2 A t a s)

AS iO C no A e En os r (EFA Z e O3-_2. 30. B 7O Comnon Area SSS O5 F re t a d 2 F re Area 2 Ac-- rs)

A 67 - r 8 e R o S c os r 20 or B (F Zo 30, ( 76 t - t/G Se Train F e rea 22 Room t A . a s)

AS 7 -- 2 h8 e P is E cbs re 2 or SF. Zo e 3- (AS 767 2 M.G 5e ra I Fire Area 23 Poom 2 A o s r m S ba e22 SF2 e 6 ti RomFe F e rr - a 2 Ac- ns AS 767 C rnmo Ar-a S as re 23 (EFh Zc e 2 33. 6. S 6 Common Area SS Frent ad2 F- re hrea 5 Act ons)

CHARTCONTiNUESON NEXT PAGE

Q91

References:

MNS LOSS OF PLANT CONTROL DUE TO FIRE OR SABOTAGE PAGE NO.

AP111A15500124 4 of 87 Rev. 32 UNIT I 7.

r ACTION/EXPECTED RESPONSE Trip all Unit I NC pumps.

RESPONSE NOT OBTAINED Perform the following:

a. Open all Unit 1 NC pump 6900V supply breakers.

b. IF all Unit 1 NC pumps are off, THEN GO TO Step 8.

c. operator already dispatched to trip Unit 1 NC pumps in AP/ 01A15500145 (Plant Fire) THEN GO j,Q Step S.

d. E Unit 1 NC pump(s) will not trip fliN dispatch RO or other operator in Control Room to trip Unit 1 NC pump(s) from Unit I 6900V room PER Enclosure 3 (Tripping Unit I NC Pumps From Unit I 6900V Room).

8. Check Unit 1 turbine trip as follows: Perform the foDowing:

. All throttle valves CLOSED,

- a. Trip Unit 1 turbine.

b IF Unit 1 turbine will not trip THEN initiate Unit 1 Main Steam Isolation.

9. Trip both Unit I CF pumps. Dispatch operator to trip both Unit I CF pumps.

10. Dispatch personnel as follows:

a. Check SSF activation PERFORMED

- a. Observe Note prior to Step I G.e and BY APIOIA1S&10145 (Plant Fire). j,Q Step 10.0.

_b. IQStep1I.

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ILT-30 MNS SRO NRC Examination QUESTION 92 92 WE16 EA2.2 High Containment Radiation Ability to determine and interpret the following as they apply to the (High Containment Radiation)

(CFR: 43.5/45.13)

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

Regarding the use of FR-Z.3 (RESPONSE TO HIGH CONTAINMENT RADIATION),:

1) at what MINIMUM reading on 1EMF 51A (CONTAINMENT HIGH RANGE) is the YELLOW path for Containment High Radiation valid?

2) what actions does the CRS direct to mitigate the consequences of the event?

A. 1. l5RIhr

2. Start the Containment Auxiliary Charcoal Filter Unit B. 1. 35RIhr

2. Start the Containment Auxiliary Charcoal Filter Unit C. 1. l5RJhr

2. Ensure the VE system is in service and purge Containment to the Annulus D. 1. 35RIhr

2. Ensure the VE system is in service and purge Containment to the Annulus Thursday, May 29, 2014 Page 273 of 298

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ILT-30 MNS SRO NRC Examination QUESTION 92 92 General Discussion In this question, the applicant is asked what reading from the containment Area radiation monitor (1EMF-5 IA) would give a valid yellow path on SPDS for High Containment Radiation. The yellow path for containment radiation is 35 R/hr per FR- FO.

The procedural direction for mitigation is to place the containment Charcoal Filter unit in-service. While purging the containment atmosphere to the annulus would effectively lower Rad levels in containment, it is not directed for this situation.

Answer A Discussion INCORRECT: See explanation above.

PLAUSIBLE:

Part I is plausible because the value of l5mrlhour is the limit for an area monitor outside containment (Control room) which requires the declaration of an Alert because it is considered a value which would impede the operation of systems required to maintain safe operations. The SRO applicant may confuse this value (15) with that required to implement corrective actions for high containment rad levels.

Part 2 is correct.

Answer B Discussion CORRECT: See explanation above.

Answer C Discussion INCORRECT: See explanation above.

PLAUSIBLE:

Part 1 is plausible because the value of l5mrlhour is the limit for an area monitor outside containment (Control room) which requires the declaration of an Alert because it is considered a value which would impede the operation of systems required to maintain safe operations. The SRO applicant may confuse this value (15) with that required to implement corrective actions for high containment rad levels.

Part 2 is plausible because this strategy would be effective in lowering containment radiation levels by performing a controlled release of the containment atmosphere to the Annulus Ventilation (VE) filter system.

Answer 0 Discussion INCORRECT: See explanation above.

PLAUSIBLE:

Part I is correct.

Part 2 is plausible because this strategy would be effective in lowering containment radiation levels by performing a controlled release of the containment atmosphere to the Annulus Ventilation (VE) filter system.

Basis for meeting the K The KA is matched because it requires the applicant to have knowledge of the procedure for High Containment Radiation which ensures the applicants ability to adhere to the requirements of the procedure.

The question is indirectly tied to operation within the limitations in the facilitys license and amendments in that the actions within the procedure for High Containment Radiation are part of the actions contained in the facilitys transient and accident technical guidelines required by NUREG 0737. The transient and accident analysis guideline procedures required by NTJREG-0737 are a requirement of Technical Specification 5.4 (Procedures) and as such the performance of those procedures it a requirement of the facilitys license.

Basis for Hi Cog Basis for SRO only This question meets the following criteria for an SRO only question as described in the Clarification Guidance for SRO-only Questions Rev idated 03/11/2010 for screening questions linked to IOCFR55.43(b)(5) (Assessment and selection of procedures):

1) This question can NOT be answered by knowing systems knowledge alone. This is strict procedure knowledge. This is not covered during systems training or discussed in a systems lesson plan.

2) This question can NOT be answered by knowing immediate operator actions.

There are NO immediate actions associated with FR-Z.3.

3) This question can NOT be answered by knowing the entry conditions for AOPs. The steps to be taken by the crew are not based on the entry conditions provided.

Thursday, May 29, 2014 Page 274 of 298

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ILT-30 MNS SRO NRC Examination QUESTION 92 92

4) This question can NOT be answered by knowing the purpose, overall sequence of events, or overall mitigative strategy of the AOPs. The question is based on knowledge of specific procedure content.

5) The question requires the applicant to have in-depth knowledge of specific steps within FRP Z.3. Specifically, it requires the applicant to recall that FR Z.3 requires the crew to place the containment aux filter in service vs. another set of actions which would provide mitigation but not directed for the indications provided.

Job Level Cognitive Level QuestionType Question Source SRO Memory BANK 2011 MNS NRC Examin Q82 (Bank 4432)

Development References Student References Provided

References:

Lesson Plan OP-MC-EP-FRZ FR-Z.3 (Response to High Containment Radiation)

Learning Objectives:

,OP-MC-EP-FRZ Objectives 2 & 3 WE16 EA2.2 High Containment Radiation Ability to determine and interpret the following as they apply to the (High Containment Radiation)

(CFR: 43.5/45.13)

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

401-9 Comments: Remarks/Status Thursday, May 29, 2014 Page 275 of 298

Q92

References:

MNS CRITLGAL SAFETY FUNCTION STATUS TREES PAGE NO, EP11tN5OOWFO Containment Page 1 of 1 Rev5 UNITI To FRZ .1 GOTO FR-Z .1 (iOTO FR-Z 2 GO TO FR-Z.1 rh GOFR-ZTO.2 (GO FR.Z4 CSF SAT

Q92

References:

MNS RESPONSE TO HIGH CONTAINMENT RADIATION LEVEL PAGE NO EP!11A15000!FR-L3 2 of 3 Rev.3 UNIT 1 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED C. orAions Check Containment ventilation isolation as follows:

a. Check the following isolation valves - a. CLOSE valve(s)

CLOSED:

. IVQ-IA(UI ContAir Release Inside Isol)J

. IVQ-6A(U1 Cont Air Addition Inside lsol)

. IVQ-2B (UI Cont Air Release Outside lsol)

. 1VQ-5B (U 1 Cont Air Addition Outside lsol).

2.

a. a Start fans as follows:

I) Select

2) Return switch to AUTO°.

b. b Ensure the following damper mode select switches in AUTO.

. IAVS-D-7 Mode Select e IAVS-D-8 Mode Select

. IAVS-D-2 Mode Select

• IAVS-D-3 Mode Select.

Q92

References:

MNS RESPONSE TO HIGH CONTAINMENT RADIATION LEVEL PAGE NO.

EPI1IA/50001FR-Z.3 3 of 3 Rev.3 UNITI EPECTED REi:.E RESPONSE NOT OBTAINED

3. Check if Containment Aux Carbon Filter Fan can be placed in service as follows:

a Check containment sump level LESS

- a. .Q j Step 4.

THAN OR EQUAL TO 0.5 FT.

b. Start IA Containment Aux Carbon Filter Fan.

4. Notify station management of containment radiation level to obtain recommended action.

5. RNIQ procedure and step in effect.

END

Q92 Parent Question:

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2011 MNS SRO NRC Examination QUESTION 82 I___

SYSO72 2.4.31 Area Radiation Monitoring (ARM) System SYSO72 GENERIC Knowledge of anmmciator alarms, intheations, or response procedures. (CFR: 41.10/45.3)

Regarding the use of FR-Z.3 (Response To High Containment Radiation):

1) At what MINIMUM reading on 1EMF 51A (Containment High Range) is the YELLOW path for Containment High Radiation valid?

2) What actbns does the CRS direct to mitigate the consequences of the event?

A. 1.35R/hr

2. Start the Containment Auxiliary Charcoal Filter Unit.

B 1. 15R/hr

2. Start the Containment Auxiliary Charcoal Filter Unit.

C. 1.35R/hr

2. Ensure the VE system is in service and purge containment to the annulus D. 1. l5Rflir

2. Ensure the VE system is in service and purge containment to the annulus Tnesda, August 23, 2011 Page 245 of 302

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2011 MNS SRO NRC Examination QUESTION 82 General Discussion

[In this question, the applicant is asked what reading from the containment Area radiation monitor (lElvW-51A) would give a valid yellow path on SPDS for Hub Containment Radiation. The yellow path for containment radiation is 35 RibrperFR/FO. The procedural direction for mitigation is to place the containment Charcoal Filter unit in-service While purging the containment auncuphere to the annulus would

[effectively lower Red levels in containment, it is not directed for this situation.

Answer A Discussion CORRECT: See explanation above.

Answer B Discussion NCORRECT: See explanation above.

1jUSlBLEc bat (1) is plausible because the value of l5mr?hour is the limit for an area monitor outside containment (Control roon which requires the Leclaratiom of an Alert because it is considered a value which would inwede the operation of systems required to maintain safe operations. The SRO applicant may confuse this value (15) with that required to implement corrective actions for high containment rad levels.

bat (2) is correct and therefore plausibin Answer C Discussion NCORRECT: See explanation above.

LAUSlBLE:

brt (1) is correct and therefore plausible.

bit (2) is plausible because this strategy would be eflbctive in lowering containment radiation levels byperfonning a controlled release of the ontainment atusosphere to the Annulus Ventilation (VE) ifiter systeni Answer D Discussion NCORRECT: See elanation above.

1AUSBLE:

Part (1) is plausible because the value of l5mr/bour is the limit for an area monitor outside containment (Coutsol roonl which requires the declaration of an Alert because it is considered a value which would impede the operation of systems required to maintain safe operations. The SRO applicant may confuse this value (15) with that required to implement corrective actions for high containment red levels.

bit (2) is plausible because this strategy would be effective in lowering containment radiation levels bypexfonning a controlled release ofthe

ontainment abnosphere to the Annulus Ventilation (VE) filter system Basis for meeting the KA

[lie K/A is matched because the implensentation ofFRP Z.3 (Response to High Containment Ra is a response procedure for a abnonnal aidication associated with the containment ARM (IEIvff.5lA). This is also a direct alarm response for the Yellow path for SPDS Cntainment which would come in alarm at 35 Rdir. The Mann response on the QAC directs implementation of the FRZ.3. The question equires the applicant to denronstiate both what reading would be required to implement and detailed knowledge ofwhat is directed by this rrccednre.

Basis for Hi Cog Basis for SRO only This question meets the following criteria for an SRO only question as described in the ClarifIcation Guidance for SRO-only Questions Rev lthted 03/11/2010 for screening questions linked to 10CFRS5.43i)(5) (Assessment and selection ofprocedures):

1) This question can NOT be answered by knowing systems knowledge alone. This is strict procedure knowledge. This is not covered during

.ystems training or discussed in a systems lesson plan.

2) This question can NOT be answered by knowing immediate operator actions.

3) This question can NOT be answered by knowing the entry conditions for AOPs, The steps to be taken by the crew are not based on the entsy onditions provided.

I) This question can NOT be answered by knowing the puapose, overall sequence of events, or overall mitigative strategy of the AOPs. The luestiosi is based on knowledge of specific procedure content.

5) The question requires the applicant to have in-depth knowledge of specific steps within FRP Z3. Specificall it requires the applicant to acall that FRZ.3 requires the crew to place the containment aim filter in service vs. another set of actions which would provide mitigation but sot directed for the indications Fosded.

Tuesday, August 23, 2011 Page 246 of 302

Q92 Parent Question:

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2011 MNS SRO NRC Examination QUESTION 82 Job Level Cognitive Level QuestionType Question Source SRO Memory BANK Bank 591 Development References Student References Provided esson Plan OP-MC-EP-FRZ Objedives 2 & 3 (Rev 20)

.ruon Plan OP-MC-EP-FRZPg 53 of 89

/FR-Z.3 (Response to Higr Containment Rad Level)

SYSO72 14.31 Area Radiation Monitoring (ARM) System SYSO72 GENERIC Knowledge ofannunciator aIanns indications orresponseprocedures. (CFR 41.10/45.3) 401-9 Comments: Remarks/Status I 4(fl..9 Comment SAT.

RESOLUflON N/A HCF5/16/ll Tuesday, August 23, 2011 Page 247 of 302

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ILT-30 MNS SRO NRC Examination QUESTION 93 93 WEO8 EA2.2 Pressurized Thermal Shock Ability to determine and interpret the following as they apply to the (Pressurized Thermal Shock)

(CFR: 435/45.13)

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

Given the following on Unit 1:

• A LOCA has occurred

• E-i (LOSS OF REACTOR OR SECONDARY COOLANT) has been implemented The following conditions exist:

• NC System pressure is 950 PSIG

• Containment pressure is 7 PSIG

• NC System Cold Leg temperature has decreased from 547°F to 248°F in the last hour Procedure Legend:

FR-Ri (RESPONSE TO IMMINENT PRESSURIZED THERMAL SHOCK CONDITION)

FR-P.2 (RESPONSE TO ANTICIPATED PRESSURIZED THERMAL SHOCK CONDITION)

Based on the conditions above, the NEXT procedure transition will be to (1)

After transition, the Safety Injection Termination Criteria will be (2) restrictive than the termination criteria in E-1.

Which ONE (1) of the following completes the statements above?

A. I. FR-P.i

2. LESS B. i. FR-P.2

2. LESS C. i. FR-P.i

2. MORE D. 1. FR-P.2

2. MORE Thursday, May 29, 2014 Page 276 of 298

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ILT-30 MNS SRO NRC Examination QUESTION 93 93 General Discussion Based on the conditions given, transition will be to FR-P.l since there has been a cooldown of greater than 100°F in the last hour and all cold leg temperatures are NOT greater than 250°F.

The Safety Injection Termination criteria is less restrictive in FR-P. 1 than in ES- 1.1 because of the need to minimize/limit the NC system cooldown.

Answer A Discussion CORRECT: See explanation above.

Answer B Discussion TNCORRECT: See explanation above.

PLAUSIBLE:

Part I is plausible if the applicant does not correct recall the graph for the break points related to loss of integrity severity. If so, the applicant could conclude that FR-P.2 would apply instead of FR-P. 1.

Part 2 is correct.

Answer C Discussion INCORRECT: See explanation above.

PLAUSIBLE:

Part I is correct.

Part 2 is plausible because it would be logical for the applicant to conclude that SI termination criteria is always more restrictive in EPs outside of ES-l.l. If the applicant does not understand why SI is being terminated in FR-P.1, they might conclude that ES-l.l is less restrictive.

Answer D Discussion INCORRECT: See explanation above.

PLAUSIBLE:

Part 1 is plausible if the applicant does not correct recall the graph for the break points related to loss of integrity severity. If so, the applicant could conclude that FR-P.2 would apply instead of FR-Pa.

Part 2 is plausible because it would be logical for the applicant to conclude that SI termination criteria is always more restrictive in Eps outside of ES-I .1. If the applicant does not understand why SI is being terminated in FR-P. 1, they might conclude that ES-l.l is less restrictive.

Basis for meeting the K For this question, the applicant must demonstrate a knowledge of the EPs related to Pressurized Thermal Shock to demonstrate that they have the ability to adhere to those procedures. Additionally, while operation within the EPs is not DIRECTLY related to operation within the facilitys license, adherence to the plants Emergency Procedures ensures that actions taken to mitigate the consequences of a design-basis accident are consistent with the mitigation strategy set forth in the plants accident analysis which is a requirement in the facilities license.

Basis for Hi Cog This is a higher cognitive level question because it requires more than one mental step.

First, the applicant has to recall from memory the chart related to the Integrity Critical Safety Function and the Integrity Critical Safety Function Status Tree Flowchart.

Next, the applicant has to compare the information given in the stem to the Integrity Critical Safety Function criteria to determine which transition is appropriate.

Finally, the applicant has to recall from memory the Safety Injection Termination criteria from both ES- 1.1 and FR-P. 1, compare the two, and determine which of the two is more restrictive.

Basis for SRO only This question meets the following criteria for an SRO only question as described in the Clarification Guidance for SRO-only Questions Rev 1 dated 03/11/2010 for screening questions linked to IOCFR55.43(b)(5) (Assessment and selection of procedures):

I) The question can NOT be answered solely by knowing systems knowledge.

None of the information asked for in this question is related to systems knowledge.

2) The question can NOT be answered by knowing immediate operator actions.

This is NOT related to immediate actions. It is related to a procedure transition from one EP to another.

Thursday, May 29, 2014 Page 277 of 298

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ILT-30 MNS SRO NRC Examination QUESTION 93

3) The question can NOT be answered solely by knowing entry conditions for AOP or direct entry conditions for EOPs.

This question cannot be answered solely by knowing direct entry conditions for EOPs. The question requires the applicant to have knowledge of Safety Injection Termination Criteria from two different EOPs, and identify which criteria is less restrictive.

4) The question can NOT be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of the procedure.

This is detailed knowledge of procedure content related to criteria for terminating Safety Injection and identifying which is less restrictive.

5) The question requires detailed knowledge of procedure content. Therefore, it is SRO knowledge.

Job Level Cognitive Level Questionrype Question Source SRO Comprehension BANK 2010 MNS Audit Exam Q84 (Bank 3397)

Development References Student References Provided

References:

F-0 (CRITICAL SAFETY FUNCTION STATUS TREES)

FR-P.l (RESPONSE TO IMMINENT PRESSURIZED THERMAL SHOCK CONDITION)

Learning Objectives:

EPFRPOO3 WEO8 EA2.2 Pressurized Thermal Shock Ability to determine and interpret the following as they apply to the (Pressurized Thermal Shock)

(CFR: 43.5 /45.13)

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

401-9 Comments: RemarkslStatus Thursday, May 29, 2014 Page 278 of 298

o U,

.4-f3 uJ

- .4-40 U,

9 gc. Li H

[,1 Ui ti 0

H 0

U)

C) Li Li Li Li Li Li Li Li Li Li Li Li Li Li Li 0 Li Li Li Li Li Li Li Li Li Li Li Li Li Li CL) U, C%I Li Ui W LL C%l Li Ci) Ll Li tl C) C.J i 4\j Cj l I I Li IJ1Ll tIJX C.IfjliUi C)

C) a

Q93

References:

MNS CRITFCAL SAFElY FUNCTION STATUS TREES PAGE NO.

EPII!N5000IF-O Reactor Coolant Integrity Page 1 of I Rev5 0010 FR-P.l (1010 FR-P.l 0010 FR-P .2 (1010 FR-PJ 0010 FR-P 2 csF SX CSF SAT

Q93

References:

RESPONSE TO IMMINENT PRESSURIZED THERMAL PAGE NO SHOCK CONDITION EPI1!Al&000IFR-R1 Rev.13 irrri r E:cuExPEc6r S5E

7. Check it Sit can be temiinated: Perform the following:

. NCsubcoolingbased oncoreexitTlCs a. ffiNcsubcoolingbasedoncoreexit GREATER TI-IAN 5DF. TiCs greaterthari OF all NC pumps off, Iftj,4 start an NC pump

• Check RVLIS indication: ER EPI1fAI5000IG-1 (Generic Enclosures), Enclosure 6 (NC Pump

(( all NC pumps off, THEN check Startup).

REACTOR VESSEL LR LEVEL -

GREATER THAN 60%. _b. IQStep26.

OR IF atleastoneNDpumpcn.THEN check REACTOR VESSEL DIP -

GREATER THAN REQUIRED DELTA P FROM TAELE EELO:

,JE Et ESSE_

R 3 4,: . L -c -tr .

r CFF a:;

3 35 6 2 23

3E

Q93

References:

MNS LOSS OF REACTOR ORSEGONDARY COOLANT PAGE NO.

EPJIIN5000IE-1 5of23 urri ACTOUEXPECD RE5PGNS RESPD5E NOT OBANEO

6. tCantinued)

c. At least onePzr PORVisolationvalve- c. OPEN one Pzr PORVisolationvalve OPEN. unless itwas closed to isolatean open Pzr PORV.

d. IF AT ANY TIME an Pzr PORV opens dieto hicihpressrire.THEN after pressuregoes below22lSPSIG.

erisbrePzr PORV CLOSES oris isOiMed.

7. Check SI termination criteria:

. NO sLIbc-oollng based on core exitTlCs a. QIQ Step7.f.

- GREATER TH.SN DF.

b. Secondarheatsinl:: b. JQStep7.f, lJJRIe,eIiriatlestcneintactSiG GRE/-.TERTHAIJ 11% 32% ACCi OR

• Total feed fio to intact SLG..

GREATER THAN 450 GPT.

e NC pressure- STAELE OR GOING LIP. c. Step 7.f.

d. P:r leel- GREATERTHAN 11% d. Performthefollowing:

25% ACC.

_1) NCpressuregoingupJjpPzr spray availablejIjj4 try to stabilize NC pressureusing normal Pzr spray,

2) QIQStep7.f.

e. GO TO EPIiLJ50CDES-i.1 (Safety InjectionTermination.j

f. MonitorSIt&rrninatioricriteriaER Enclosure2 (S.lTerrrunation Criterial hilein this procedure.

ci. JF ATANYTIME while in this procedure SI terriiiriatiori criteria are met. THEN RETURN TO Step 7.

Q93 Parent Question:

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2010 MNS SRO Audit Examination QUESTION 84 I EPEOO EA2.14 Small Bteak L0C Ability to deteimine or interpret the following as they apply to a small break LOCk (CFR 43.5 45.13)

Actions to be taken if PTS limits are violated Given the foHowing conditions on Unit 1:

ALOCA has occurred

• The crew is performing E-i, Loss of Reactor or Secondary Coolant

• The following conditions exist:

o NC System pressure is 950 PSIG o Containment Pressure is 7 PSIG o NC System Cold Leg temperature has decreased from 547F to 248T in the last hour Which ONE (1) of the following describes the status of the Integrity CSF Status Tree, and whether SI Termination criteria is more or less restrictive than the criteria in ES-i .1 (SI Termination)?

A. Yellow; Enter FR-P.2, Response to Anticipated Pressurized Thermal Shock Condition; 51 Termination Criteria is j. restrictive than ES-i.i B. Orange; Enter FR-Ri, Response to Imminent Pressurized Thermal Shock Condition; SI Termination Criteria is less restrictive than ES-I .1 C. Orange; Enter FR-P,i, Response to Imminent Pressurized Thermal Shock Condition; SI Termination Criteria is more restrictive than ES-i .1 D. Yellow; Enter FR-P.2, Response to Anticipated Pressurized Thermal Shock Condition; SI Termination Criteria is more restrictive than ES-i. 1 Friday, October 15,2010 Page 181 of 220

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2010 MNS SRO Audit Examination QUESTION 84 General Discussion Answer A Discussion jiOPRECT: See explanation above.

PLAUSIBLE:

Answer B Discussion ICPT: See explanation above.

Answer C Discussion llCORRECT: See explanation above.

PLAUSIBLE:

Answer D Discussion INCORRECT: See explanation above PLAUSIBLE:

Basis for meeting the KA Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source SRO Comprehension BANK kINS 2005 Audit Exam (Q75)

Development References Student References Provided F-O.4 EPEOQP EA2.14 -Small Break LOCA Abiit to detennine or interpret the following as they appIvto a small break LOCA: (CFR 435 45.13)

Actions to be taken if PTS limits are violated 401-9 Comments: jRemarkslStatus Friday, October 15, 2010 Page 182 of 220

FOR REVIEW ONLY DO NOT DISTRIBUTE ILT-30 MNS SRO NRC Examination QUESTION 94 94 GEN2.1 2.1.3 GENERIC Conduct of Operations Conduct of Operations Knowledge of shift or short-term relief turnover practices. (CFR: 41.10 / 45.13)

Given the following plant conditions:

• Both units are at 100% RTP

• The Control Room Supervisor (CRS) and the relief SRO need to be absent from the Control Room for a short period of time

• The Shift Technical Advisor holds an active SRO license and is available to provide relief In accordance with Tech Spec 5.1 (RESPONSIBILITY), the STA may assume the duties of the CRS provided:

1) the CRS or relief SRO is available to return to the control room within minutes, AND

2) the periods during which the STA assumes SRO duties do not exceed minutes in duration.

Which ONE (1) of the following completes the statements above?

A. 1.10

2. 15 B. 1.10

2. 60 C. 1.15

2. 15 D. 1.15

2. 60 Thursday, May 29, 2014 Page 279 of 298

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT-30 MNS SRO NRC Examination QUESTION 94 94 General Discussion Technical Specifications allows the Shift Technical Advisor to assume the control room command fl.inction and perform the duties of the control room SRO in Modes 1, 2, 3, and 4 during periods when the CRSRO and the relief SRO are required to be absent from the control room.

However, the following requirements must be met:

1) The STA must hold an SRO license for the unit.

2) The CRSRO or relief SRO must be available to return to the control room within 10 minutes.

3) The periods during which the STA may perform the control room SRO duties may not exceed 15 minutes in duration or a total of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> for the entire shift.

Answer A Discussion

[CORRECT: See explanation above.

Answer B Discussion INCORRECT: See explanation above.

PLAUSIBLE:

The first part is correct.

The second part is plausible because 60 minutes is the TOTAL amount of time specified in Tech Spec 5.1 that the STA is allowed to relieve the CRS during one shift.

Answer C Discussion INCORRECT: See explanation above.

PLAUSIBLE:

The first part is plausible because this is the time period that the STA is allowed to relieve per Tech Spec 5.1.

The second part of the question is plausible because this the time requirement specified in Tech Spec 5.1 that the CRS must be able to return to the Control Room. And, it does not exceed the time period that the STA is allowed to relieve.

Answer D Discussion INCORRECT: See explanation above.

PLAUSIBLE:

The first part is plausible because this is the time period that the STA is allowed to relieve per Tech Spec 5.1.

The second part is plausible because 60 minutes is the TOTAL amount of time specified in Tech Spec 5.1 that the STA is allowed to relieve the CRS during one shift.

Basis for meeting the K KA is matched because the candidate must understand the control room manning requirements for the individual fulfilling the control room command function.

Basis for Hi Cog Basis for SRO only This is an SRO Only question linked to IOCFR55.43(b)(2), Tech Specs. This questions can NOT be answered by knowing less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Tech Spec or TRM action statements. It can NOT be answered by knowing the LCO/TRM information listed above-the-line (since this is an Administrative Control). It can NOT be answered by knowing Tech Spec Safety Limits or their basis. The candidate must apply requirements from Section 5.0, Administrative Controls of Technical specifications. Requirements in Section 5.0 are NOT expected knowledge for ROs.

Job Level Cognitive Level QuestionType Question Source SRO Memory BANK 2009 MNS NRC Exam NRC Q86 (Bank 3101)

Development References Student References Provided

References:

Tech Spec 5.1 (RESPONSIBILITIES)

Learning Objectives:

OP-MC-ADM-OMP Objective 12 Thursday, May 29, 2014 Page 280 of 298

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ILT-30 MNS SRO NRC Examination QUESTION 94 94 GEN2.1 2.1.3 GENERIC Conduct of Operations Conduct of Operations Knowledge of shift or short-term relief turnover practices. (CFR: 41.10 / 45.13) 401-9 Comments: RemarkslStatus Thursday, May 29, 2014 Page 281 of 298

Q94

References:

piity 5.1 5.0 ADMINISTRATIVE CONTROLS 5.1 Responsibility 5.1.1 The Station Manager shall be responsible for overall unit operation and shall delegate in writing the succession to this responsibility during his absence.

5.1.2 The Control Room Senior Reactor Operator CRSROi shall be responsible for the control room corrimand function. During any absence of the CRSRO from the control room while the unit is in MODE 1 2. 3.

or 4. an individual lother than the Shift Technical Adisor (STA1] with an active Senior Reactor Operator (SRO) license shall be designated to assume the control room command function. During any absence ofthe CRSRO from the control room while the unit is in MODE 5 or 6. an individual with an actie SRO licei se or Reactor Operator license shail be designatedto assume the control room command function.

On occasion when there is a need for both the CRSRO andthe relief SROto be absent fromthe control room in MODEl. 2 3. cr4. the STA shall be allowed to assume the control room command function and serve as the SRO in the control room providedthat

a. the ORSRO or the relief SRO is available to return to the control room within 10 minutes.

b. the assumption of SRO duties by the STA is limited to periods not in excess of 15 minutes duration and a total time not to exceed during any shift. and

c. the STA has a SRO license on the unit

j McGi.nre Units and 2

Q94 Parent Question (2009 MNS NRC Q86 Bank 3101): -

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross- Level RO SRO reference: x Tier# 3 Finat Group#

KIA# (321.5 Importance Rating 3.9 Conduct of operations Ability to Iocste and use procedures related to shift stalling, such as minimum crew complement, overtime limitations, etc.

Proposed Question: SRO 86 1 Pt Unit 1 is operating at 100% RTR Under which ONE (1) of the following conditions may an active licensed STA assume the duties of the Control Room Supervisor?

The CRS or relief SRO is available to return to the control room within (1) AND the periods during which the STA assumes SRO duties do not exceed (2) in duration, A. (1) 15 minutes (2)15 minutes B. (1) 15 minutes (2>10 minutes C. (1) 10 minutes (2)15 minutes D. (1) 10 minutes (2) 10 minutes Proposed Answer: C Page 271 of 320 Rev Final

Q94 Parent Question (2009 MNS NRC Q86 Bank 3101): -

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Explanation (Optional):

Technical Specifications allows the Shift Technical Advisor to assume the control room command function and perform the duties of the control room SRO in Modes 1, 2, 3, and 4 during periods when the CRSRO and the relief SRO are required to be absent from the control room. However, the following requirements must be met:

• The STA must hold an SRO license for the unit

• The CRSRO or relief SRO must be available to return to the control room within 10 minutes.

• The periods during which the STA may perform the control room SRO duties may not exceed 15 minutes in duration or a total of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> for the entire shift.

A. Incorrect: See explanation above. Plausible if the candidate confuses the time for the CRSRO or relief SRO to return to the control room with the allowable duration of the relief by the STA.

B. Incorrect: See explanation above. Plausible if the candidate confuses the time for the CRSRO or relief SRO to return to the control room with the allowable duration of the relief by the STA.

C. Correct.

D. Incorrect: See explanation above. Plausible if the candidate confuses the time for the CRSRO or relief SRO to return to the control room with the allowable duration of the relief by the STk Technical Reference(s) Technical Specification (Attach if not previously 5.1.2, amendment 213 and provided) 194 (Including version or revision Proposed references to be provided to applicants during None examination:

Learning Objective: OPMCADM-OMP, Obj 3 (As available)

Page 272 of 320 Rev Final

Q94 Parent Question (2009 MNS NRC Q86 Bank 3101): -

ES-401 Sample Written Examination Form ES-401 -5 Question Worksheet Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Memory or Fundamental Knowledge X Level:

Comprehension or Analysis 10 CFR Part 55 5541 Content:

5543 43_5 Comments:

Conduct of operations Ability to locate and use procedures related to shift staffing, such as minimum crew complement, overtime limitations, etc KA is matched because the candidate must understand the control room manning requirements for the individual fulfilling the control room command function.

This is an SRO Only question linked to 10CFR5543(b)(2), Tech Specs. This questions can NOT be answered by knowing less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Tech Spec or TRM action statements It can NOT be answered by knowing the LCOITRM information listed above-the-Iine (since this is an Administrative Control). It can NOT be answered by knowing Tech Spec Safety Limits or their basis The candidate must apply requirements from Section 50, Administrative Controls of Technical specifications. Requirements in Section 5.0 are NOT expected knowledge for ROs Page 273 of 320 Rev Final

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ILT-30 MNS SRO NRC Examination QUESTION 95 95 GEN2.1 2.1.45 GENERIC Conduct of Operations Conduct of Operations Ability to identify and interpret diverse indications to validate the response of another indication. (CFR: 41.7 / 43.5/45.4)

Given the following conditions on Unit 1:

• A unit startup is in progress

• The unit is at 15% RTP

• A Steam Generator Tube Leak occurs Indications:

• S/G NR Level going up in an uncontrolled manner

• S/G or Steam Line EMFs above normal

• Comparison of S/G Secondary Chemistry samples

1. S/G CF Flow comparison

2. Comparison of RP frisk of S/G cation columns

3. Comparison of RP frisk of main steam lines In accordance with AP-lO (NC SYSTEM LEAKAGE WITHIN THE CAPACITY OF BOTH NV PUMPS) Case I (STEAM GENERATOR TUBE LEAKAGE), ALL of the redundant indications that will be used to identify the leaking S/G are the bulleted indications and Which ONE (1) of the following completes the statement above?

A. 1 ONLY B. 3 ONLY C. 1AND2ONLY D. 2AND3ONLY Thursday, May 29, 2014 Page 282 of 298

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ILT-30 MNS SRO NRC Examination QUESTION 95 95 General Discussion In accordance with AP- 10, the following indications are used to identify the S/G with the tube leak:

1. S/G NR Level going up in an uncontrolled manner

2. S!G or Steam Line EMFs above normal

3. Comparison of S/G Secondary Chemistry samples

4. S/G CF Flow comparison

5. Comparison of RP frisk of S/G cation columns Answer A Discussion INCORRECT: See explanation above.

PLAUSIBLE:

This answer is plausible because this is correct. If the applicant does not recall that an RP frisk of the cation columns is also listed in the AP, they would conclude that this is the correct response.

Answer B Discussion INCORRECT: See explanation above.

PLAUSIBLE:

This answer is plausible if the applicant confuses the cation column frisk with a frisk of the main steam lines and does not recall that a comparison of S/G CF Flows is also listed in the AP.

Answer C Discussion CORRECT: See explanation above.

Answer D Discussion INCORRECT: See explanation above.

PLAUSIBLE:

This answer is plausible because a comparison of S/G cation column frisks is listed in the AP. If the applicant recalls that there are five indications listed and does not recall that S/G CF Flow comparison is listed, they could logically conclude that the main steam line frisk was the remaining indication.

Basis for meeting the K The K.A is matched because it required the applicant to have knowledge of the diverse indications that are used in AP-lO to identify the S/G with the tube leak.

Basis for Hi Cog Basis for SRO only This question meets the following criteria for an SRO only question as described in the Clarification Guidance for SRO-only Questions Rev 1 dated 03/11/2010 for screening questions linked to 10CFR55.43(b)(5) (Assessment and selection of procedures):

1) The question can NOT be answered solely by knowing systems knowledge.

This is not a systems knowledge based question.

2) The question can NOT be answered by knowing immediate operator actions.

There are no immediate actions associated with AP-lO.

3) The question can NOT be answered solely by knowing entry conditions for AOP or direct entry conditions for EOPs.

This question is not related to entry conditions for AP- 10.

4) The question can NOT be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of the procedure.

This is detailed knowledge of steps within the body of the procedure.

5) The question requires detailed knowledge of procedure content. Therefore, it is SRO knowledge.

Job Level Cognitive Level QuestionType Question Source SRO Memory BANK 2013 MNS NRC Q95 Thursday, May 29, 2014 Page 283 of 298

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ILT-30 MNS SRO NRC Examination QUESTION 95 95 Development References Student References Provided

References:

AP-lO (NC System Leak Within The Capacity of Both NV Pumps)

Learning Objectives:

OP-MC-AP-1O Objective 7 GEN2.l 2.1.45 GENERIC Conduct of Operations Conduct of Operations Ability to identify and interpret diverse indications to validate the response of another indication. (CFR: 41.7 / 43.5 / 45.4) 401-9 Comments: RemarkslStatus Thursday, May 29, 2014 Page 284 of 298

Q95

References:

NC SYSTEM LEAKAGE WITHIN THE CAPACITY OF PAGE NO BOTH NV PUMPS 4 of 131 APIIIN5500IID Rev. 23 viii i Steam Generator Tube Leakage ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

2. iiA[M4uIiME Pzr level qoes down in an uncontrolled manner OR cannot be maintained qreater than 4%, IHi perform Step 1.

NOTE ri subsequent steps affected S/G is considered the SIC v:ith primary to secondar leakaqe requinriq unit shutdown

3. Identify affected SIG as follows:

• Any S.C N.R level - GOING UP IN AN UNCONTROLLED r.tANNER OR

• Check any of the followinq EMFs ABOVE NORMAL:

a 1EMF-24 (S1G A Starniia Hi Radi a 1EMF-25 SIC B .SL.awJiaa. Hi Rad

• 1EMF-2G (S1G C .SJ.awJja Hi RacI)

• 1EMF-27 (SIC D taujliaa Hi Rad)

• 1EMF 71 (S/C ALeakaqe Hi Ract) a 1EMF 72 (SIC B Leakaqe Hi Rad a IEMF 13 (S/C C Leakaqe Hi RaclI a 1EMF 74 (S/C D Leakacle Hi Rad).

OR a Oheck CF Flow LOVI.ER IN ANY S/G COMPARED TO ALL.

OR a Secondar Chemistry or RP has determined affected SIC by samplinq or evaluation of available EMF data.

OR a Notify RP to frisk all Unit 1 SIC cation columns (CT Lab) to determine if activity level is siqnificantlv hiqher for any SIC.

4 Announce occurrence on paqe.

Q95 Parent Question:

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2013A MNS SRO NRC Examination QUESTION 95 I GENIi 2.1.45 GENERIC Conduct of Operations Conduct of Operalions Abdityto idenndy and intespret diverse indications to validate the response of another indication. (CFR: 41.7143.5/45.4)

Given the following conditions on Unit 1:

A unit startup is in progress The unit is at 15% RTP

• A Steam Generator Tube Leak occurs Indications:

• SIG NR Level going up in an uncontrolled manner

• S/G or Steam Line EMPs above normal

• Comparison of S/G Secondary Chemistry samples 1 Comparison of RP frisk of main steam lines

2. S1G CF Flow comparison

3. Comparison of RP frisk of SIG cation columns In accordance with AP-lO (NC SYSTEM LEAKAGE WITHIN THE CAPACITY OF BOTH NV PUMPS) Case I (STEAM GENERATOR TUBE LEAKAGE), ALL of the redundant indications that will be used to identify the leaking S/G are the bulleted indications and Which ONE (1) of the following completes the statement above?

A. 1 ONLY B. 3 ONLY C. IAND2 D. 2AND3 Thursday, Deceiuber 05, 2013 Page 285 of 302

Q95 Parent Question:

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2013A MNS SRO NRC Examination QUESTION 95 General Discussion In accordance with AP-lO, the following indications are used to identify the Sf0 with the tube leak:

1. SIONR Level oina up in an uncontrolled manner

2. Sf0 or Steam Line EMFs above normal

3. Comparison of 5/0 Secondary Chemistiy samples

4. Sf0 CF Flow comparison

5. Comparison of RI frisk of SG cation columns Answer A Discussion INCORRECI: See explanation above.

PLAUSIBLE:

This answer is plausible if the applicant confuses a comparison ofthe frisk of main steam lines with a frisk of SG cation cohunns and concludes that Sf0 CF Flows axe not useful for identifying a leaking Sf0.

Answer B Discussion INCORRECT: See explanation above.

PLAUSIBLE:

This answer is plausible if the applicant and concludes that 5/0 CF Flows are not useful for identifying a leaking SG and does not recall that hand-held radiatioumonitors can be an effective means of identifying a leaking 5/0, Answer C Discussion INCORRECT: See explanation above.

PLAUSIBLE:

This answer is plausible if the applicant confuses the RI frisk of main steam lines with RI frisk of SO Cation columns.

Answer D Discussion CORRECT: See explanation above.

Basis for meeting the K The KA is matched because it required the applicant to have knowledge of the diverse indications that are used in AP-lO to identify the 5/0 with Ithetubeleak.

Basis for Hi Cog Basis for SRO only This question meets the following criteria for an SRO only question as described in the Clarification Guidance for SRO-only Questions Rev 1 dated 03/11/2010 fur screening questions linked to IOCFR55.43tb)(5) (Assessment and selection of procedures):

1) The question can NOT be answered solely by knowing systems knowledge.

This is not a systems knowledge based question.

2) The question can NOT be answered by knowing immediate operator actions.

There are no immediate actions associated with AP-lO.

3) The question can NOT be answered solely by knowing entry conditions for AOP or direct entry conditions for EOPs.

This question is not related to entry conditions fur AP-IG.

4) The question can NOT be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of the procedure.

This is detailed knowledge of steps within the body of the procedure.

5) The question requires detailed knowledge ofprocedure content. Therefore, it is SRO knowledge.

Job Level Cognitive Level QuestionType Question Source SRO Memory NEW Thursday, December 0, 2013 Page 286 of 302

Q95 Parent Question:

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2013A T S SRO NRC Examination Development References AP-lO TC Systemi Learning Objectives:

OP-MC-AP-10 Objec GENii 2145 GE4ERIC Conduct of Operations Conduct of Operations Ability to identify and intespret diverse indications to validate the response of another indication. (CFR 413/43.5/454) 401-9 Comments: RemarkslStatus This question submitted for pre-401-9 review.

Based on feedback from Chief Examiner, revised stem of guestion to be less confining. Otherwise kept question as mitten, HCF 09/05/2013.

401-9 Comment: SAT G2i45 Question appears to match the KJA This is a different cay to match this KJ& I also tend to believe this is SRO only mowledge. Need to make sure the 110 question on this K/A does not help the 5110 applicants to answer this question. (May need to change the 110K/A ifthis cannot be done)

Question could be made a little less condasing by rewording stem to state:

Indications:

• S/O NRlevel sising in anuncontolled manner S10 or Steam line EMP is above normal Comparison of 5/0 Secondasy Chemisfty samples I Comparison of RP frisk ofmain steam liner 2S/G CF flow comparison 3.Compaxison of RP frisk of 516 Cation Columns A.1 only B.3 only C.Iand2 D.2 and 3 Made changes as requested. SAT 10/2S/2013 No changes to question at this time. SUvIIOI31)2013 Q95 approved as SAT by Chief Examiner. HCF 11/19/13 Thursday, December 05,2013 Page 287 of 302

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ILT-30 MNS SRO NRC Examination QUESTION 96 GEN2.2 2.2.25 GENERIC Equipment Control Equipment Control Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits. (CFR: 41.5 / 41.7 / 43.2)

Given the following conditions on Unit 1:

• A power escalation to 100% RTP is in progress

• The OATC observed that Control Bank D rod H8 is at 185 steps

• The remaining rods in Control Bank D are at 200 steps

• l&E has determined the cause of the misalignment is a blown lift coil fuse In accordance with Tech Spec 3.1.4 (ROD GROUP ALIGNMENT LIMITS) Basis,:

1) Control Rod H8 OPERABLE.

2) maximum rod misalignment is an initial assumption in the safety analysis that directly affects Which ONE (1) of the following completes the statements above?

A. 1.lS

2. Core Power Distribution ONLY B. 1.IS

2. Core Power Distribution AND Shutdown Margin (SDM)

C. 1. IS NOT

2. Core Power Distribution ONLY D. 1. IS NOT

2. Core Power Distribution AND Shutdown Margin (SDM)

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ILT-30 MNS SRO NRC Examination QUESTION 96 96 General Discussion Since the cause of the misaligned rod is a blown lift coil fuse, the applicant should conclude that the misaligned control rod is trippable and therefore, in accordance with TS 3.1.4 Basis, remains OPERABLE.

Until recently, it was a long-held misconception at MNS that control rod operability was tied to BOTH trippability and proper alignment (because of the way the LCO is worded). And, even though this point has been clarified by Reactor Engineering, it remains a concept that is frequently misunderstood by Licensed Operators.

In accordance with TS 3.1.4 Basis, maximum rod misalignment is an initial assumption in the safety analysis that directly effects core power distribution and SDM.

Answer A Discussion CORRECT: See explanation above.

Answer B Discussion INCORRECT: See explanation above.

PLAUSIBLE:

The first part is correct.

The second part is plausible because maintaining Axial Flux Distribution within limits is an assumption in the safety analysis for limiting core peaking factors. Since core power distribution (peaking factors) is one of the bases for rod misalignment limits, it is plausible for an applicant to conclude that AFD is the other safety analysis assumption.

Answer C Discussion INCORRECT: See explanation above.

PLAUSIBLE:

The first part is plausible because the wording of TS 3.1.4 (All shutdown and control rods shall be OPERABLE, with all individual indicated rod positions within 12 steps of their group step counter demand position), has led to a long-standing misconception at MNS that if a control rod is misaligned by more than 12 steps, it is considered INOPERABLE.

The second part is correct.

Answer D Discussion INCORRECT: See explanation above.

PLAUSIBLE:

The first part is plausible because the wording of TS 3.1.4 (All shutdown and control rods shall be OPERABLE, with all individual indicated rod positions within 12 steps of their group step counter demand position). has led to a long-standing misconception at MNS that if a control rod is misaligned by more than 12 steps, it is considered INOPERABLE.

The second part is plausible because maintaining Axial Flux Distribution within limits is an assumption in the safety analysis for limiting core peaking factors. Since core power distribution (peaking factors) is one of the bases for rod misalignment limits, it is plausible for an applicant to conclude that AFD is the other safety analysis assumption.

Basis for meeting the K The KA is matched because the applicant must have knowledge of Tech Spec 3.1.4 Basis to determine the correct response.

Basis for Hi Cog This is a higher cognitive level question because it requires more than one mental step.

First, the applicant must analyze the conditions given in the stem to determine that the control rod is misaligned due to an electrical problem with the CRDM and NOT because it is mechanically bound. This will allow the applicant to determine that the control rod is still trippable and therefore remains OPERABLE.

Next, the applicant must recall from memory, the assumptions set forth in the TS 3.1.4 Basis regarding the reasons for maintaining control rods within alignment limits.

Basis for SRO only This question meets the following criteria for an SRO only question as described in the Clarification Guidance for SRO-only Questions Rev 1 dated 03/11/2010 for screening questions linked to 10CFR55.43(b)(2) (Tech Specs):

1) This question can NOT be answered by knowing less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Tech Specs

2) This question can NOT be answered by knowing information listed above-the-line.

3) This question can NOT be answered by knowing the TS Safety Limits or their bases.

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ILT-30 MNS SRO NRC Examination QUESTION 96

4) This question requires the applicant to have knowledge of TS 3.1.4 (ROD GROUP ALIGNMENT LIMITS) Basis and is therefore SRO-Only knowledge.

Job Level Cognitive Level QuestionType Question Source SRO Comprehension NEW Development References Student References Provided

References:

Tech Spec 3.1.4 (ROD GROUP ALIGNMENT LIMITS)

Tech Spec 3.1.4 Basis Learning Objectives:

OP-MC-IC-IRE Objective 14 GEN2.2 2.2.25 GENERIC Equipment Control Equipment Control Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits. (CFR: 41.5 / 41.7 / 43.2) 401-9 Comments: arks/Status Thursday, May 29, 2014 Page 287 of 298

Q96

References:

Rod Group Allgnment Limits 3 IA 31 REACTIVITY CONTROL SYSTEMS 31A Rod Group Alignment Limits LCO 31 A All shutdown and control rods shall be OPERABLE, with all individual indicated rod positions within 12 steps of their group step counter demand position]

APPLICABILITY: MODES I and 2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more rod(s) All Verify SDM is within the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> jppoJ limit specified in the COLR.

OR A12 Initiate boration to restore 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> SDM to within limit AND A2 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> (continued)

D MoGuLe Units 1 and 2 3 1 4-1 Amendment No 184:166

Q96

References:

Rod GroupAhgnment Limits B 1t4 83.1 REATlVffY CONTROL SYSTEMS B 3.1.4 Rod GroupAlignment Limits BASES BACKGROUND The OPERABILITY (e.g., pLty) ofthe shutdown and control rods is an initial assumption in all safety analyses that assume rod insertion upon reactor trip. Maxiraim rod misalignment is an initial assumion iii the safety analysis that dired affects core power distributions and assumptions of available SDM.I The applicable criteria for these reactivity arid power cfistribiiion design requirements are 10 CFR 50. Appendix A, GDC 10, Reactor Design.

GDC 26, Reactivity Control System Redundancy and Protection (Ref 1). arid 10 CFR 50 46, Acceptance Criteria for Emergen Core Cooling Systems for Light V.ater Nuclear Paver Plants (Ref. 2).

Mechanical or electrical failures may cause a control rodto become inoperable onto become misaligned from its group Control rod inoperability or misalignment may cause increased power peaking, clue to the asymmetric reactivit distribution and a reduction in the total available rod worth far reactor shutdown. Therefore, control rod alignment and OPERABILITY are related to core operation in design paver peaking limits and the core design requirement of a minimum SDM Limits on control rod alignment have been established, and all rod positions are monitored and controlled during paver operation to ensure that the power distribution arid reactivity limits defined by the design power peaking and 5DM lirns are presered.

Rod cluster control assemblies (ROAsi. or rods are moved by their control rod drive rrletharHsms (CRDMs Each CRDM moves its RCCA one step (approxirrtely 5:8 irichi at atimne. but at varying rates (steps per minute) depending on the sigui oLJtpLit from the Rod Control System.

The RCCAs are divided arriong control banks and shLitdown banks. Each bank may be further subdivided into two groups to provide for precise reactivity control. A group consists oftwo or more RAs tI at are electrically paralleledto step simultaneously. A bank of RCCAs consists of two groups that are moved in a staggered fashion. but always within one step of each other. The unit has four control banks and fre shutdown banks.

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ILT-30 MNS SRO NRC Examination QUESTION 97 97 GEN2.3 2.3.6 GENERIC Radiation Control Radiation Control Ability to approve release permits. (CFR: 41.13 / 43.4 / 45.10)

Unit I is at 100% RTP.

An Operator brings a gaseous radiological release permit to the SRO for approval.

Given the following information on the permit:

• Release ID = WGDT B

• Most restrictive release rate = 3.24E÷02 OEM

• Recommended release rate = 4.OOE+01 OEM

• EME-50(L) (WASTE GAS) in service = yes

• EME background = 1.58 E+01

• Trip I setpoint 5.34E+02

• Trip 2 setpoint = 4.99E+03 Which ONE (1) of the following actions is correct for approval of this release permit?

A. The release may NOT be approved because the release rate is not correct.

B. The release may NOT be approved because the EME-50(L) trip setpoints are not correct.

0. The release may be approved as presented ONLY if I EME-36(L) Unit Vent Gas is also operable.

D. The release may be approved as presented if a source check of EME-50(L) is performed successfully.

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ILT-30 MNS SRO NRC Examination QUESTION 97 97 General Discussion Per WE-RGR, pg 13, The SRO will review the GWR to ensure the Recommended Release Rate is less than the most restrictive release rate. He also reviews the expected range of the EMF and the Trip 1 and 2 setpoints. He also checks the operability of the controlling EMF (normally 0EMF-50, but could be 1EMF-36). Operability of BOTH EMF-50 and 1EMF-36 is preferred but NOT required. The SRO also ensures the controlling EMF has been source checked.

Answer A Discussion INCORRECT. See Explanation Above.

PLAUSIBLE:

This answer is plausible if the applicant incorrectly interprets or misreads the Most Restrictive Release Rate and the Recommended Release Rate OR incorrectly recalls that the Most Restrictive Release Rate must be less than the Recommended Release Rate.

Answer B Discussion INCORRECT. See explanation above.

PLAUSIBLE:

This answer is plausible if the applicant misreads the EMF trip setpoints relative to each other or relate to the EMF background reading.

Answer C Discussion INCORRECT. See explanation above.

PLAUSIBLE:

This answer is plausible because Operators are trained that it is desirable to have BOTH EMF-36(L) and EMF-50(L) operable for a waste gas release.

Answer 0 Discussion CORRECT. See explanation above.

Basis for meeting the K KA is matched because the item evaluates requirements for issuing a radioactive liquid waste release permit.

Basis for Hi Cog This is a higher cognitive level question because it requires the applicant to evaluate multiple pieces of information to determine the correct response.

Basis for SRO only SRO level because the SRO is responsible for authorizing the release based on given conditions.

Job Level Cognitive Level QuestionType Question Source SRO Comprehension BANK 2010 MNS Audit Exam Q83 (Bank 2983)

Development References Student References Provided Technical Reference(s)OP-MC-WE-RGR, Rev 9 OP/OIB/6200/019, End 4.1 (Rev 29)

Learning Objectives:

OP-MC-WE-RLR Objective 3 GEN2.3 2.3.6 GENERIC Radiation Control Radiation Control Ability to approve release permits. (CFR: 41.13 /43.4 / 45.10) 401-9 Comments: Remarks!Status Thursday, May 29, 2014 Page 289 of 298

Q97

References:

From Lesson Plan OP-MC-WE-RGR (RadoIogicaI Gaseous Releases):

Objective #2,3,4 Opetator Fundamental Focus; Monitoiing and Teamwork Reinforce the fundamental attribute for the CRS to track degraded and inoperable technical specification equipment Relate this fundamental behavior to the tact that it is preferable to make a release with either IEMF-36 or OEMF-50 inoperable.

Controlling EMF operability (IEMF-36 or OEMF-50), and any necessaly inoperable actions is a responsibility of the entire crew and communicating problems to the CRS so that the appropriate Tech Spec actions can be taken support the Operator Fundamental Teamwork The Discharge Document is then deilvered to the Control Room where the Control Room SRO ensures all paperwork is complete prior to authorizing the release This authorization serves as an acknowledgment by the Control Room SRO that a release is about to take place. He should review the following prior to authorization.

• GWR document agrees with release procedure (Le. WGDT ReLease procedure used forWG release).

• Recommended Release Rate y calculated release rates.

The Recommended Release Rate will be the most restrictive release rate based on sample activity or the maximum observed system release rate (40 CFM for WGDT releases whichever is less.

The SRO reviewing the release paperwork should ensure the Recommended Release Rate is less than or equal to the Most Restrictive Release Rate.

• Controlling EMF operability (IEMF-36 or 0EMF-50 and any necessary inoperable actions. It is preferable to make a release with either IEMF-36 or OEMF-50 inoperable.

NOTE: WGDT must be resampled if a trip 2 on either EMF auto-terminates the release.

• Expected range of EMF. Trip 1. and Trip 2 setpaints.

• Checks Waste Gas Discharge Flow Monitor operability.

• Reviews special Instructions.

jihe Control Room Operator ensures the GWR# in i$.lpg.

FOR TRAINING PURPOSES ONL V REV, 12 Page 7 of 15*

Q97

References:

RETOAS v3 .5 1 <]JCNN S Rev. 4 0> CANBERRA SDUS RELE.SE REfIT PET GWR NurLber: 2014026 Release ID: A Waste Gas Decay Tsn

== ALLOiUB1E RE LEASE RATE Total both dose release rate (cErn) N/A Szrj arid Gatm air dose release rate (cfri) N/A Tood, Ground, Inhalation dose release rate Ccfrn) N/A Most restrictive aelease rate (cErn) Unrestricted Recciended release rate (oErn Unrestricted

== SETPQINT DATA E50L Monitor Operable° Yes ENT5OL Entered Background (c) 5.OOE+Ol E350L Ezpected CPM 5.OOE+01 Trip 1 Setpoint (c 2.502E+02 Trip 2 Setpoint (c) 3.503E+02 E36L Monitor Operable? NA EE36L Entered Backcrourid (.cij NA E,E36L Ezpected CPM NA Entered Vent Ilowrate (cErn) NA Xe-lSS Equivalence (uCi/cc) 0.OOE+00 Source Check Performed :__________________ Unit 1 EEE0L Trip 142 Set/Verified by:_________________

Per Formed by:_____________________________________________________________

IZ by:

= SPEZIAL INSTRJJZTIQS CR 510 Authorization:____________________________ flate/Tine: I Release Initiation Release Cce>letion RUn______ Release Start ROW______ Release Stop /

Date Tine Date Tine ROW_______ Vol. rel ft3 Ocaspletion of Release Acknowledged CR 510 Date/Time:

RP Review__________________________ Data/Time:

Ensure aU signoffs are legible, Print name where indicated on next page.

Date/Tine: 05/05/2014 14:06 Page 2

Q97 Parent Question:

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2010 MNS SRO Audit Examination QUESTION 83 I____

APEO6D AAI.02 Accidental G eot-Wate Release Ability to operate and or otonitor the following as they apply to the Accidental Gaseous Radwaste: (CFR 41.7 45.5 45.6)

Vetadlallon system Unit 1 is at 100% RTP.

A Radwaste Operator brings a gaseous radiological release permit to the SRO for approval Given the following information on the permit:

• ReleaselD=WGDTB

• Most restrictive release rate = 3.24E+02 CFM

• Recommended release rate = 400E+01 CFM

• EMF-50(L) (WASTE GAS) in service = yes

• EMF background = 158 EtOl

• Trip 1 setpoint = 534E+02

• Trip 2 setpoint = 4.99E+03 Which ONE (1) of the following actions is correct for approval of th release permit?

The release may NOT be approved because the release rate is not correct, R The release may NOT be approved because the EMF-50(L) thp setpoints are not correct.

C. The release may be approved as presented ONLY ii IEMF-36(L) Unit Vent Gas is also operable.

D The release may be approved as presented if a source check of EMF-50(L) is pertormed successthl[

Friday, October 15 2010 Page 179 of 220

Q97 Parent Question:

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO Audit Examination QUESTION 83 24331 General Discussion Per WE-RGR. pa 13. The 5RO will view the GWR to e ure the Recommended Release Rare is less than the most restrictive release rate He also reviews the expected range of the EMF and the Trip I and 2 setpoints. He also checks the operability of the controlling EMF (notmally

[0EMr-50 but could be 1ThlF-36) Operability of BOTH EMF-30 and 1Th-36 is preferred buSNOT requirecL The SRO also ensures the jconuoiliug Eh has been source checked.

Answer A Discussion ncorrect: recommended release rate is less than the move restrictive release rate IPlmathie: if candidate doss not understand this requirement Answer B Discussion frncotrect: nothinz svrona with Ehff-50L trip sepoints IlPiansible: -backouud:aip I :thp2 Answer C Discussion Incorrect. Petniit can be approved but EMF-36 does not need to be operable.

Answer 0 Discussion Correct: The SRO ensures the EMP Source check is sianed oE Basis for meeting the KA KA is matched because the item evaluates requirements the issuing a radioactive liquid waste release pemlit 5310 level because the 5310 is frespansible for authorizing the release based on aiven conditions Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source 5310 Comprehension BANK bINS Development References Student References Provided fechulcal Referen s)OP-MC-WE-RGII Rev P lOp 03 6200 019. End 4I (Rev 29) eaauian Objective:OP-MC-WE-EiRobj 3 APEO6D AAI.02 Accidental Gaseouc-Waste Release Ability to operate and or monitor the following as they apply to the Accidental Gaseous Radwaste: (CPR 41.7 45.5 456)

Ventilation system 01-9 Comments: RemarkslStatus Friday, October 1S 201(1 Page 180 of 220

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ILT-30 MNS SRO NRC Examination QUESTION 98 GEN2.3 2.3.15 GENERIC Radiation Control Radiation Control Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc. (CFR: 41.12 / 43.4 / 45.9)

Given the following plant conditions:

• Unit I is preparing to make a radioactive liquid waste release (LWR)

• OEMF-49 (LIQUID WASTE DISCHARGE) Minimum Flow Device has failed Based on the conditions above, in accordance with SLC 16.11.2 (RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION),:

1) Radiation Monitor OE1F-49 is

2) the Liquid Waste Release Which ONE (1) of the following completes the statements above?

REFERENCE PROVIDED A. 1. FUNCTIONAL

2. may proceed with no additional actions B. 1. FUNCTIONAL

2. may proceed provided additional actions are performed C. 1. NON-FUNCTIONAL

2. may proceed provided additional actions are performed D. 1. NON-FUNCTIONAL

2. may not proceed until the minimum flow device is repaired Thursday, May 29, 2014 Page 290 of 298

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ILT-30 MNS SRO NRC Examination QUESTION 98 [ 98 General Discussion In accordance with SLC 16.11.2 (RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION), the minimum flow device must be functional for OEMF-49 to be considered functional. In accordance with SLC 16.11.2, with EMF-49 non-functional, the Liquid Waste Release may be initiated provided additional actions are taken.

Answer A Discussion INCORRECT: See explanation above.

PLAUSIBLE:

The first part is plausible because the monitor itself is functioning correctly. If the applicant misreads the SLC or misses the note which states that both the monitor and minimum flow device must be functional for the EMF to be considered functional, they could conclude that the minimum flow device is not required for the EMF to be functional.

Second part is plausible if the applicant misreads or misinterprets the SLC.

Answer B Discussion INCORRECT: See explanation above.

PLAUSIBLE:

The first part is plausible because the monitor itself is functioning correctly. If the applicant misreads the SLC or misses the note which states that both the monitor and minimum flow device must be functional for the EMF to be considered functional, they could conclude that the minimum flow device is not required for the EMF to be functional.

Second part is correct.

Answer C Discussion CORRECT: See explanation above.

Answer D Discussion INCORRECT: See explanation above.

PLAUSIBLE:

The first part is correct.

Second part is plausible because this is an option provided by the SLC should other compensatory actions not work or cannot be performed.

Basis for meeting the K The KJA is matched because it requires the applicant to have know of a radiation monitoring device as it relates to Selected Licensing Commitments and a Liquid Waste Release.

Basis for Hi Cog The is a higher cognitive level question because it requires the applicant to evaluate the conditions given and apply the requirements of the Selected Licensing Commitment to determine the correct response.

Basis for SRO only This question is SRO level knowledge because it requires the applicant to analyze given conditions and apply Technical Specifications beyond the application of one hour or less specs.

Job Level Cognitive Level QuestionType Question Source SRO Comprehension NEW Development References Student References Provided

References:

SLC 16.1 1.2 (Radioactive Liquid Effluent Lesson Plan OP-MC-WE-EMF Monitoring Instrumentation)

Learning Objectives:

OP-MC-WE-EMF Objective 10 GEN2.3 2.3.15 GENERIC Radiation Control Radiation Control Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc. (CFR: 41.12 / 43.4 / 45.9)

Thursday, May 29, 2014 Page 291 of 298

FOR REVIEW ONLY DO NOT DISTRIBUTE ILT-30 MNS SRO NRC Examination QUESTION 98 401-9 Comments: RemarkslStatus I Thursday, May 29, 2014 Page 292 of 298

Q98

References:

Raftoa te Li iud E ffluet lonitot ing [nst umentati in 16112 1611 RADIOLOGICAL EFFLUENT CONTROLS 1611.2 Radioactive Uquid Effluent Monitoring Instrumentation COMMITMENT The radioactive liquid effluent monitoring instrumentation channels shown in Table 1611.21 shall be FUNCTIONAL with their Alarm/Trip Setpoints set to ensure that the limits of SLC 1611.1 are not exceeded.

AND The Alarm/Trip Setpoints of these channels shall be determined and adjusted in accordance with the methodology and parameters in the OFFSITE DOSE CALCULATION MANUAL (ODCM).

APPLICABILITY As shown in Table 1611.2-1.

REMEDIALACTIONS Separate Condition entry is allowed for each Function.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more radioactive A,1 Suspend the release of Immediately liquid effluent monitoring radioactive liquid effluents I channels Alarm/Trip setpoint less monitored by the affected channel.

conservative than required. OR A2 Declare the channel non. Immediately functional.

OR A.3 Adjust setpoint to within Immediately limit.

B. One or more radioactive B.1 Enter the Remedial Action Immediately liquid effluent monitoring specified in Table 16.11.2-instrument channels 1 for the channel(s).

non-functional.

(continued)

LioGuire Units 1 and 2 Revision 124

Q98

References:

Radoao%ve LkiUd. EffLUent OflOflfl:g Lnstrum:ertation 16112 4IREMEDIALAC1IONS (continued CONDITION REOLIE RED ACTION COMPLETION liME C One channel non CII Analyze two independent functional samples perTR 161111.

AND 012 Perform independent verification of the discharge line valving AND C 1 3 1 Perform independent verification of manual portion of the computer input for the release rate calculations performed by cornputer OR Ci 32Perforrn independent verification of entire release rate calculations for calculations performed manually.

AND C1.4 Restore channel to FUNCTIONAL status.

OR 02 Suspend the release of radioactive effluents via this pathway.

iJccLi U MoGuire UnLts I and 2 I. 6 ii 2-I RevLsion 134

Q98

References:

Radaotive Liq.uid Effi.uent Monitoring. nstrurnentation 1.6.112 fREMEDIAL ACTIONS (continued CONDITION REQUIRED ACTION COMPLETION TIME

0. One or more channels D.i Obtain grab samples from Once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> non-functional. the effluent pathway. during releases.

AND D.2 Perform an analysis of To meet LLD grab samples for requirements per radioactivity. Table 1611.1-1.

AND D.3 Restore the channel to 30 days FUNCTIONAL status E. One or more flow rate El -----NOTE measurement channels Pump performance curves non-functional. generated in place may be used to estimate flow.

Estimate the flow rate of Once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> the release. during releases AND E.2 Restore the channel to 30 days FUNCIIONAL status F. RC minimum flow Fl Verify that the number of Once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> interlock non-functional. pumps providing dilution is during releases greater than or equal to the number of pumps required, AND F.2 Restore the channel to 30 days FUNCI1ONAL status.

MoGuire Units 1 and 2 Revision 134

Q98

References:

RadDacfr a Liqu d Eff1uetMnibring Instil mantation 1611 2 REMEDIAL ACTIONS (continued CONDITION REQUIRED ACTION COMPLETION TIME G Required Action and G. 1 Explain why the non- ln the next associated Completion functionality was not scheduled Annual Time of Condition C, D, corrected within the Radioactive Effluent E or F not met. specified Completion Time Release Report in the Annual Radioactive Effluent Release Report.

MoGuire Units 1 and 2 Revision I 34

Q98

References:

P adioactiie Liqud Efflue I onLo ing tnstrurnnta:on 1611 2 TESTING REQUIREMENTS Refer to Table 16.11.2-1 to determine which TRs apply for each Radioactive Liquid Effluent Monitoring channel.

TEST FREQUENCY TR 16.11.2.1 Perform CHANNEL CHECK. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> TR 1611.2.2 ---NOTE The CHANNEL CHECK shall consist of verifying indication of flow.

Perform CHANNEL CHECK. Every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> during periods of release TR 16,11.2.3 Perform SOURCE CHECK. Prior to each release TR 16.11.2.4 Perform SOURCE CHECK. 31 days TR 16.11.2.5 -----_____

1. For Instrument 1, the COT shall also demonstrate that automatic isolation of the pathway occurs it the instrument indicates measured levels above the AlarmIThp Setpoint.

2. For Instruments 1 and 2, the COT shall also demonstrate that control room alarm annunciation occurs if the instrument indicates measured levels above the AlarmlTrip Setpoint; circuit failure and, a downscate failure.

Perform CHANNEL OPERATIONAL TEST. 92 days TR 16.11.2.6 Perform a CHANNEL CALIBRATION. 18 months F leG uire Jnits 1 a.ia 2 1611 2 Reision 134

Q98

References:

Radoactje Liquid Effluent Monitorinq ustrumentation 1611.2

÷TESTING REQUIREMENTS_(continued)

TEST FREQUENCY TR 16112.7 The initial CHANNEL CALIBRA11ON shall be performed using standards certified by the National Institute of Standards and Technology (NIST) or using standards obtained from suppliers that participate in measurement assurance activities with NIST. These standards shall permit calibrating the system over its intended range of energy and measurement range. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration shall be used.

Perform a CHANNEL CALIBRATION. 24 months Mot3uire Units 1 and 2 16.11.2-6 Revision 134

Q98

References:

Radioactive Liquid Efti.uen.i Monitoring instrumentation 16112 TABLE 1611.2-i RADKJACTIVE UQUID EFFLUENT MONITORING INSTRUMENTATION INSTRUMENT MINIMUM REMEDIAL APPLICABlL( TESTING CHANNELS ACTION REQUIREMENTS FUNCTIONAL Radioaciivity Monitors ProvidingAlarm And AutomaticTermination of Release

a. WasteLiquidEffluentLine(EMF-49) 1 perstatian A,CG During liquid TR 16.11.2.1 effluentreleases TR 16.11.2.3 TR 16.11.2.6 TR 16.11.2.7

b. EMF-49MinimumFlowDevice 1 per station CG During liquid TR 16.11.2.5 (2) effiuentrelease.s TR 16.11.2.7

c. ContainmetaVentiIationUnitConcsnsate I A, DG At all times TR 16.11.2.1 Line{EMF-44) TR 16.11.2.4 TR 16.11.2.5 TR 16.11.2.7

d. EMF-44MinimumFlowDeviee I DG Atailtimes TR 16.11.2.5 (2) TR 16.11.2.7

2. Radioaivity Monitors ProvidingAlarm But Not AutomaticTermination of Release

a. Conventional Waste Water Treatment I A. D. G At all times TR 16.11.2.1 Line or Turbine Building Sump to RC TR 16.11.2.4 (EMF- 31)

TR 16.11.2.6 TR 16.11.2.7 1 DG Atall times TR 16.11.2.5

b. EMF-31 Minimum Flow Device (2) TR 16.11.2.7

3. GontinuousComposite Samplers

a. ContainmentVentilationUnitConcIsnsate 1 DG Mall times TR 16.11.2.2 Line TR 16.11.2.5 TR 16.11.2.6

b. ConventionalWasteWaterTreatmentLine 1 per station DG At all times TR 16.11.2.2 TR 16.11.2.5 TR 16.11.2.6

c. TurbineBuildingSumptoRC 1 D,G Atall times TR 16.11.2.2 TR_16.11.2.6 (Continued)

MeG uire Units 1 and 2

Q98

References:

Radioactive Liquki Effluent Monitoring Instrurnentaiton 16.1t2

4. Flow Rate Measurement Devices

a. WasteLiquidEffluentLine I perstation E.G During liquid TR 16.11.2.2 effluent releases TR 16,112.5 TR 16.11.2.6

b. co fffi6i VifiltfflSiiU5ft!ni TR 16.1122 Line TR 16.11.2.5 TR 16.11.2.6

c. Conventional Waste WaterTreatment Line I per station E, G At all times TR 16.11.22 TR 16112.6 TR 16.112,6

d. TurbineBuildingsumptoRC 1 EG Atalltimes TR 16.11.2.2 TR 16.11.2.6

6. RCMinimumFlowlntaiock{1) I perstation F,G Atall times TR 16.11.2.5 NOTE

1. Miriirnumflou dilution is ssuredbj an interlocluhid,terrniniatesasteIiquid release fthenLmber ofRC pumps ruririingfalls belc thenumber of purnpsreqvireclfor clIL.tion ThereqLriredrlurnbr ofRD purnpsfor dilution is determinad pa station procedures

2. Radioacti.it Ucnitc.r,EMFishallrict bedeclaredfundioral unless boththe EMF aridthe assodted EMFs Minimum F Ic El ice are rendered fun di ona I BASES The radioactive liquid effluent instrumentation is provided to monitor and control. as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents. The minimum flow devices for EMFs listed in Table 16.1 1.2-1 are required to provide assurance of representative sampling during actual or potential releases of liquid effluents. An interlock between the EM Fs minimum flow device arid its associated flow rate rneasLlrenient device disables the remove alarm during nan-release timeframes for the purpose of the control room black board annunciator criteria that disable expected alarms An EM F flow rate measurement device measures total flaw of the effluent while the. EMF minimum flow device measures the sample flow rate throLlgh the EMF. The AlarmsTrip Setpoints of these instruments shall be calculated and adjusted in accordance with the methodology and parameters in the 00CM to ensure that the Alarm/Trip will occur prior to exceeding the limits stated in SLC 16,11.1. The FIJNCTIONALITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60. 63. and 64 of Appendix A to 10 CFR Part 50. The Turbine Building Sump to RC Discharge Flow Measurement and Sampler Devices are for monitoring only and do riot alarm or have any controls that require a COT.

REFEREN CES

1. McGuire Nuclear Station Offsite Dose Calculation Manual (00CM)

2. 10 CFR Part 50. Appendix A

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ILT-30 MNS SRO NRC Examination QUESTION 99 99 GEN2.4 2.4.2 1 GENERIC Emergency Procedures! Plan Emergency Procedures / Plan Knowledge of the parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc. (CFR: 41.7 / 43.5 /45.12)

Given the following conditions on Unit 1:

• A Loss of Offsite Power occurred, resulting in a stuck open PZR safety valve

• Reactor tripped automatically and SI actuated automatically

• lB DIG failed to start

• Actions of E-0 (REACTOR TRIP OR SAFETY INJECTION) are complete at Time = 0645 Conditions at Time 0700: Conditions at Time = 0720:

• 1 B D/G will not start

• 1 B DIG will not start

• NC pressure=l 650 PSIG

• NC pressure=1610 PSIG

• RVLIS Lower Range level = 35%

• RVLIS Lower Range level = 33%

• PZR safety valve is stuck open

• PZR safety valve is stuck open

• lNl-9A (NC Cold Leg Inj from NV)

• I NI-9A is closed is closed

• CETs=715°F

• CETs = 705°F

• Subcooling Monitor= -35°F

• Subcooling Monitor= -35°F Based on the conditions at 0720, which ONE (1) of the following is the classification for this event? (For the purposes of this question, do NOT consider Emergency Coordinator judgment as a basis of classification)

REFERENCE PROVIDED A. Notification of Unusual Event (NOUE)

B. Alert C. Site Area Emergency D. General Emergency Thursday, May 29, 2014 Page 293 of 298

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ILT-30 MNS SRO NRC Examination QUESTION 99 99 General Discussion Given the plant conditions the applicant should determine a loss of subcooling (based on NCS pressure and CET temperature), no NCPs running (due to Loss of Offsite Power) and a Red Path on the Core Cooling safety function (>700 °F and <39% reactor vessel level with NCPs off).

Per RP/0/A15700/000, the correct classification is General Emergency based on 4.1 .G.2. (Loss of Any Two Barriers AND Potential Loss of the Third)

Containment Barrier: 1 points (Due to Core Cooling RED path indicated for greater than 15 minutes)

NCS Barrier: 5 points (Due to Loss of Subcooling) LOSS Fuel Clad Barrier: 5 points (Due to Core Cooling RED path) LOSS TOTAL POINTS = 11 (General Emergency)

Answer A Discussion INCORRECT. See explanation above.

PLAUSIBLE:

This answer is plausible if the applicant ONLY included the points for a RED path on Core Cooling for greater than 15 minutes. This would result in an Unusual Event declaration.

Answer B Discussion INCORRECT: See explanation above.

PLAUSIBLE:

This answer is plausible if the applicant ONLY includes the points for a LOSS of NCS Barrier (5 points) and Potential Loss of Containment Barrier (I point) as the total (6 points) would result in the declaration of an ALERT.

Answer C Discussion INCORRECT: See explanation above.

PLAUSIBLE:

This answer is plausible if the applicant excludes the POTENTIAL LOSS of the Containment Barrier as a result of the existence of a Core Cooling RED path for> 15 minutes but correctly includes the LOSS of NCS Barrier due to leak rate that results in a loss of subcooling (5 points) AND includes the LOSS of Fuel Clad Barrier as a result of the RED path on Core Cooling (5 points) as the total (10 points) would result in the declaration of a Site Area Emergency.

Answer D Discussion CORRECT: See explanation above.

Basis for meeting the K The KJA is matched because it requires the applicant to analyze the conditions given using the parameters and logic associated with Critical Safety Functions and based on that analysis, determine the correct emergency classification.

Basis for Hi Cog This is a higher cognitive level question because it requires more than one mental step.

First, it requires the applicant to recall from memory the setpoints for decision points in the Critical Safety Function Status Trees.

Next, it requires the applicant to perform a detailed analysis of the conditions given to determine the status of the plant.

Finally, it requires the applicant to analyze the Fission Product Barrier Matrix in RP-000 to determine the appropriate Emergency Classification.

Basis for SRO only This question is linked to IOCFR55.43 (b)(7) Emergency Classification. Per the guidance in IOCFR55.43 and per the MNS objective referenced for this question, assessing plant conditions and determining the proper classification of emergency is considered SRO level.

Job Level Cognitive Level QuestionType Question Source SRO Comprehension BANK 2012 MNS NRC Q77 (Bank 4428)

Thursday, May 29, 2014 Page 294 of 298

FOR REVIEW ONLY DO NOT DISTRIBUTE 99D ILT-30 MNS SRO NRC Examination QUESTION 99 Development References Student References Provided References; RP-000 (CLASSIFICATION OF EMERGENCY)

RP-000 (CLASSIFICATION OF EMERGENCY)

Learning Objectives:

OP-MC-EP-EAL Objective 6 OP-MC-EP-FRC Objective 2 GEN2.4 2.4.21 GENERIC Emergency Procedures / Plan Emergency Procedures / Plan Knowledge of the parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc. (CFR: 41.7 / 43.5 / 45.12) 401-9 Comments: RemarkslStatus Thursday, May 29, 2014 Page 295 of 298

Q99

References:

MNS CRITICAL SAFETY FUNCTION STATUS TREES PAGE NO.

EP/1IAJ5OOOIFO 4 of 11 Core Cooling Page 1 of 2 Rev. 5 UNIT I GO TO FR-C,l GO TO FR-Cl GO TO FR-C .2 GO TO FR-C .2 GO TO FR-C .3 JGOTO FR-C .2 REACFOR VESSFL wr GREflk ThAN REQUIRED FOR PURE COM1IINATION (SEE TAHLE RENT PAGE) YES CSF SAT

Q99

References:

Enclosure 4.1 RP/O!A!5700/000 Fistion Product Barrier Matrix Page 3 of S 4LC CONTAIX\IENT BARRIER 4JN NCS BARRIER 4.LF FUEL CLAD BARRIER POTENTIAL LOSS - LOSS - POTENTIAL LOSS - LOSS - POTENTIALLOSS- LOSS-(1 Point) (3 Points) (4 Points) (5 Points) (4 Points) (5 Points)

1. Ceitleal Safety Function Status I. ChOral SatetS Ilinrilon Status

• Containment-RED. Not applicable. NC.S Integrity- Not applicable. Core Cooling-

• Core Cooling-RED.

RED. ORANGE.

Cure Cooling RED Path is Heat Sink-RED. Heat Sink-RED.

indicated for IS minutes.

2. Coutainmeut Conditions 2. NCR Leak Rate 2. Primary Coolant Activity Level Containment Rapid unexplained Unisolable leak

• GItLAIER THAN Not applicable Coolant Activity Pressure> 15 decrease in exceeding the asadable n.akeup GREATER THAN PSIG. containment capacity of one capacity 300 pCi/cc Dose pressure following charging pimap in n.dicat:l Lvi a lus Equivalent Iodine 112 concentration initial increase, the nomial ofDCS vas.u (DEl) 1-131 charging mode Cnntainnsent with letdown Contaimnent pressure or stung isolated pressure greater than level response not 3 psig with either a consistent with failure of both trains LOCA conditions of NS OR failure of both trains of VX-CARF.

CONTINUED CONTINUED CONTINUED

Q99

References:

Enclosure 4.1 RPMAJs700/000 Fission Product Barrier Maids Page 2 of 5 NOTE: If a barrier is affected, it has a single point value based on a potential loss or a loss. Not Applicable is included in the matrix as a place holder only, and has no point value assigned.

Barrier Points (1-5) Potential Loss (X) Loss X) Total Points Classification Containment 1 3 1 3 Unusual Event 1

NCS 4 5 46 Alert Fuel Clad 4 7 10 Site Area 5

Emergency Total Points alEieigency I. Compare plant conditions against the Fission Product Barrier Matrix on pastes 3 through 5 of 5.

2. Determine the potential loss or loss status for each barrier (Containment, NCS and Fuel Clad) based on the SAL symptom description,

3. For each banier. write tire highest single point value applicable for the barrier in the Points colunm and mark the appropriate potential loss OR loss column,

4. Add the points in the Points column and record the stun as Total Points.

5. Detennine the classification level based on the number ofTotal Points.

6. in the table on page 1 of this enclosure, underone of the fourclassification columns, select the event (e.g. 4.l.A.1 forLoss of Nuclear Coolant System) that best fits the loss of barrier descnption.

7. Using that EAI. mnnber (e.g. 4.1 .A.l) select the preprinted notification form OR a blank form and complete the required information for Emergency CoordinatorfEOF Director approval and transmittal.

Q99 Parent Question:

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2012 MNS SRO NRC Examination QUESTION 77 I SYSOG6 A2.02 Emergency Core Cooing Syttem (ECCS)

Ability to (a) predict the impacts of the following m lfonctiom oroperaimos on the ECCS; and (b) based on those predictions, ore procedures to correct, control, or mitigate the consequences of those maifitoctions or operations: (CFR: 4L5 145.5)

Loss of flow path Given the following conditions on Unit 1:

• A Loss of Ofisite Power occurred, resulting in a stuck open PZR safety valve a Reactor tripped automatically and SI actuated automatically a lB DIG failed to start

• Actions of E-0 (Reactor Trip or Safety Injection) are complete at Time = 0645 Conditions at Time = 0700: Conditions at lime = 0720:

a 1BD/Gwillnotstart

• lBDlGwillnotstart a NC pressure=1 650 PSIG a NC pressure=1610 PSIG a RVLIS Lower Range level = 35% a RVLIS Lower Range level = 33%

a PZR safety valve is stuck open

• PZR safety valve is stuck open a 1 Nl-9A (NC Cold Leg inj from NV) a lNl-9Aisclosed is closed

• CET5 = 71 5°F a CETs=705°F

• Subcooling Monitor= -35°F a Subcooling Monitor= -35°F Based on the conditions at 0720, which ONE (1) of the following is the classification for this event? (For the purposes of this question, do NQI conskfer Emergency Cooclinatorjudgement as a basis of classification)

REFERENCE PROVIDED A. Notification of Unusual Event (NOUE)

B. Alert C Site Area Emergency D General Emergency Wednesday, August 29 2012 Page 229 of 300

Q99 Parent Question:

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2012 MNS SRO NRC Examination QUESTION 77 I General Disoussion Given the plant conditions the applicant should determine a loss of subcooling (based on NCS pressure and CET temperature), no NCPs nrnrring (due to Loss of Offsite Power) anda Red Path on the Core Cooling safety ftmclion (>700 F and 39% reactorvessel level with NCP5 ofl)

Per RP!0L4157001000, the correctclassi&ationis GeneralEmergencybased on4,l.G2, (Lass of Any Two Ban-iersAND Potential Loss of the Third)

Containment Barrier: 1 points (Due to Core Cooling RED path indicated for greater than 15 minutes)

NCS Barrier: 5 points (Due to Loss of Snbcooling) LOSS-Fuel Clad Barrier: 5 points (Due to Core Cooling RED path) LOSS -

TOTAL POINTS 11 (General Emergency)

Answer A Discussion NCORRECT See exulanation above.

PLAUSIBLE:

This answer is plausible if the applicant ONLY included the points for a RED path on Core Cooling for greater than 15 minutes, This would result in an Unusual Event declaration.

Answer B Discus5ion INCORRECT: See explanation above.

PLAUSIBLE:

This answer is plausible if the applicant ONLY includes the paints foraLOSS of NCS Barrier (5 paints) and Potential Loss of Containment Barrier (1 point) as the total (6 points) would result in the declaration of an ALERT Answer C Discussion INCORRECT: See explanation above.

PLAUSIBLE:

This answer is plausible if the applicant excludes the POTENTIAL LOSS of the Containment Battier as a result of the existence of a Core Cooling RED path for> 15 minutes but correctiv includes the LOSS of NCS Ranier due to leak rate that results in a lass of subcoohag (S points) AND includes the LOSS of Fuel Clad Barrier as a result of the RED path on Core Cooling (5 points) as the total (10 points) would result in the declaration of a Site Area Exiergency.

Answer 0 Discussion CORRECT: See explanation above.

Basis for meeting the KA KA is matched as follows:

The question examines the applicants knowledge of the effects of multiple malfonctians on the high pressure injection valves (Loss of flowpath) in a small break LOCA event The applicant must evaluate the overall effect on the plant and determine the emergency reugoase procedure to mitigate the consequences of the event Basis for Hi Cog This Question represents a higher caguitive level of Application because it involves a multi-part mental process of assembling difibment combinations of given information to select a correct classification.

Basis for SRO only This question is linked to IOCFR55.43 (bX7) Emergency Classification Per the guidance in 10CFRSS.43 and per the bINS objective referenced for this question, assessing plant conditions and determining the proper classification of emergency is considered SRO level.

Job Level Cognitive Leve QuestionType Question Source SRO Comprehension MODIBD 2011 MNS NRC Q77 MODIFIED (Bank 4423)

Weduerday, August 29, Z012 Page 230 of 300

Q99 Parent Question:

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2012 MNS SRO NRC Examination QUESTION 77 I Development References Student References Provided Referencer: RPOAf5700000. ClassfficatioaofEmersency RP-000 Cc astiflction of Emergency)

Leaini Objective5:

OP-MC-EP-E41 Objective 6 OP-MC-EP-FRC Objective 2 SYSOO6 A2.02 Emergency Core Cooling Syctem (ECCS)

AbiJity to (a) predict the impacts of the following ma!ftmctioris or operatksos on the ECCS; and (b) based on those predictions, use procedires so correct, courroL or mitigate the consequences of those malfonctions or operations: (CFR: 413 f 453)

Loss of flow path 401-9 Comments: RemarksIStatus Weduetday. Augittt 29, 2012 Page 231 of 300

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ILT-30 MNS SRO NRC Examination QUESTION 100 GEN2.4 2.4.40 GENERIC Emergency Procedures I Plan Emergency Procedures / Plan Knowledge of SRO responsibilities in emergencyplan implementation. (CFR: 41.10 /43.5 /45.11)

Given the following plant conditions:

• An Unusual Event was declared on Unit 2

• Initial Notification to the State, Counties and the NRC has been completed

• The Emergency Coordinator has just made the decision to upgrade the classification to an Alert After the initial notification of the upgrade to the Alert is made, follow up notifications to the State and Counties are required to be made every (1) until the emergency is terminated. (Assume no time extension has been agreed upon between the site and each agency.)

In accordance with the Emergency Plan procedures, the LOWEST Emergency Classification level that requires offsite Protective Action Recommendations be made isa (2)

Which ONE of the following completes the statements above?

A. 1. hour

2. Site Area Emergency B. 1. 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />

2. General Emergency C. 1. hour

2. General Emergency D. 1. 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />

2. Site Area Emergency Thursday, May 29, 2014 Page 296 of 298

FOR REVIEW ONLY DO NOT DISTRIBUTE ILT-30 MNS SRO NRC Examination QUESTION 100 ioo General Discussion In the scenario given, the applicant is presented with a situation where an NOUE was declared and all required initial notifications to the State, Counties, and the NRC has been completed. Subsequently. an escalation to an Alert occurs and the applicant is asked to evaluate the current notification requirements both to the NRC and the affect on the requirement for follow up notifications. Per our procedure RP-29. the follow-up notification requirement will change from 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> (NOUE) to a new requirement of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> for the new Alert classification.

In accordance with the Emergency Plan procedures. Protective Action Recommendations are REQUIRED for a General Emergency.

Answer A Discussion INCORRECT: See explanation above.

PLAUSIBLE:

Part I is correct.

Part 2 is plausible because onsite Protective Actions are required for a Site Area Emergency.

Answer B Discussion INCORRECT: See explanation above.

PLAUSIBLE:

Part I is plausible because the follow notification requirement for a NOUE is 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> which was in effect prior to the upgrade in classification.

Part 2 is correct.

Answer C Discussion CORRECT: See explanation above.

Answer D Discussion INCORRECT: See explanation above.

PLAUSIBLE:

Part I is plausible because the follow notification requirement for a NOUE is 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> which was in effect prior to the upgrade in classification.

Part 2 is plausible because onsite Protective Actions are required for a Site Area Emergency.

Basis for meeting the K The K.A is matched because the applicant must have knowledge of the SRO (OSM) responsibilities for implementing the Emergency Plan (i.e.

notification requirements to offsite agencies after an escalation in emergency classification) and requirements for Protective Action Recommendations.

Basis for Hi Cog This is a higher cognitive level question because it requires more than one mental step. The applicant must first evaluate all of the information provided and then apply multiple rules to a change in a given situation. This requires the applicant relate understanding the rules pertaining to offsite notification and apply them to a dynamic situation.

Basis for SRO only This question is not tied to I OCFR5O.43 (b) but can be classified as an SRO Plant Specific Example. This question requires additional knowledge required for the higher license level and is unique to the SRO/OSM position. At MNS it is the responsibility of the SRO to complete the notifications to offsite agencies and NRC notification to the NRC in the event that an emergency is declared. Per Lesson plan OP- MC-EP EMP (Emergency Plan) the objectives, #12 (Complete the ENF) and #14 (Complete the NRC event notification worksheet) are SRO ONLY objectives. (LPSO). Both the understanding of the requirements and the actual completion of the required paperwork along with the transmittal are SRO ONLY tasks at MNS.

Job Level Cognitive Level QuestionType Question Source SRO Comprehension NEW Development References Student References Provided

References:

RP-29 (NOTIFICATIONS TO OFF SITE AGENCIES FROM THE CONTROL ROOM)

RP- 10 (NRC IMMEDIATE NOTIFICATION REQUIREMENTS)

Learning Objectives:

Thursday, May 29, 2014 Page 297 of 298

FOR REVIEW ONLY DO NOT DISTRIBUTE C

ILT-30 MNS SRO NRC Examination QUESTION 100 ioo OP-MC-EP-EMP Objectives 12 & 14 GEN2.4 2.4.40 GENERIC Emergency Procedures / Plan Emergency Procedures / Plan Knowledge of SRO responsibilities in emergency plan implementation. (CFR: 41 10 I 43.5 / 45.11) r4019 Comments: RemarkslStatus Thursday, May 29, 2014 Page 298 of 298

Q100

References:

Enclosure 4.2 RP!O/B/5700/029 Completion and Transmission of a Page 1 of8 Follow-up Message NOTE: New initial messages for higher classification upgrades are addressed in Enclosure 4.1. {PIP M-O1-3711}

1. l lake followup notifications according to the following table:

Follwup Notifications

1. Followup notifications in the Stcire(s and Cou:uies must be made according to the following schedule:

ALERT. SAE. or GE every hour until the emergency is iennmatcd.

OR If there is any significant change to the situation mal:e notification a soon a possible>.

OR A agi cccl upon with an Enierencv Manacmnt official from each indivi:htal a2encv.

Documentation shall he maintained for an areed upon *Jiedule change. The interval foi ALERT. SAE. and GE shall not be reater than 2 horns to any aency.

2. ifa followup is due and an upgmde to a higher classificaiion is declared. there is no need to complete the fruiowup ENF In this case, the offsitc aencies must he notified that the pending followup is being superseded by an upgrade to a highe: classification and information u ill be provided.

3. Follow-up messages in the General Eniergencv classification that involve an upgrade in PARs must be conununicateci to the offsite agencies as soon as possible and within 15 minutes.

2. Complete an Emergency Notification Form by one of the following:

C Obtain a preprinted ENF.

OR C Obtain a blank ENF.

Q100

References:

Enclosure 4.4 RP0IB/5700i029 Offsite Protective Action Recommendations Page 1 of S NOTE: 1. Protective Action Recommendations (PARs) for the public apply during a General Emergency. and include sheltering, evacuation and consideration of K! use. PARs are based on plant conditions independent of projected dose. and can also be based on projected dose. Protective Action Guides PAGs) are levels of radiation dose at which prompt protective actions should be initiated and are based on EPA-400-R-92-00 1.

Manual of protective Action Guides and Protective Actions for Nuclear Incidents. The projected dose PARs specified in this enclosure are based on the PAGs listed below.

The PAG for KI is taken from Potassium Iodide as a Thyroid Blocking Agent in Radiation Emergencies. FDA Guidance. November 2001 and Guidance for Industry.

KI in Radiation Emergencies. Questions and Answers. FDA. December 2002. {23}

PROTECTIVE ACTION GUIDES (PAGs)

Projected Dose Total Effective Dose Committed Dose Recommendation Equivalent (TEDE Equivalent (CDE)

ThyToid

< 1 rein < 5 rem No Protective Action is required based on projected dose.

1 rein > 5 rem Evacuate affected zones and shelter the remainder of the 10-mile EPZ not evacuated.

NA > 5 rem Consider the use of KI (potassium iodide) in accordance with State Plans and Policy.

2. IF desired, you may refer to the flow chart of page 2 of this enclosure, {PIP M 5137. C.A.3}