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{{#Wiki_filter:Name:                                                                          1305 NRC RO Exam Form: 0 Version: 0
: 1. 007 EK1.02 001 Given the following plant conditions:
              -  Unit 1 is operating at 100% at EOL (15,000 MWD/MTU)
              -  Subsequently a reactor trip occurs on Unit 1.
              -  ES-O.1, Reactor Trip Response, has been implemented.
              -  Shutdown Bank B rod G3 remains at 228 steps.
              -  Control Bank D rod M8 sticks at 10 steps while inserting.
              -  Tavg dropped to 539°F before stabilizing.
Which ONE of the following completes the statement below?
Conditions indicate Emergency Boration is              to satisfy Shutdown Margin.
A. NOT required B    required due to the RCS temperature only C. required due to the stuck control rods only D. required due to both the RCS temperature and the stuck control rods Feedback DISTRACTOR ANAL YSIS:
A. lncorrect Plausible because if the RCS temperature had been greater than 540°F then emergency boratiuon would not have been required.
B. Correct, In accordance with ES-0. 1, Step 6b (See below) emergency boration is required due to the RCS coo/down being less than 540°F.
C. Incorrect, Plausible because if the RCS temperature had been greater than 540°F and the position of M8 rod had been greater than 12 steps then emergency boration would have been required due to the stuck rods.
: 0. Incorrect, Plausible because with the RCS temperature not greater than 540°F and the M8 rod had been greater more than 12 steps withdrawn, then emergency boration would have been required for both the excessive coo/down and 2 rods not being inserted Wednesday, June 05, 2013 8:16:13AM
 
1305 NRC RO Exam Rod position indicators              Instrument Rack BTransferSwch less than or equal to 12 steps.      to ALTERNATE. [M-7, lower switch]
IF any of the following conditions exists:
* two or more RPIs indicate greater than 12 steps OR
* two or more control rod positions CANNOT be determined, THEN EMERGENCY BORATE USING EA-68-4, Emergency Boration.
: b. MONITOR RCS temperature:              b. EMERGENCY BORATE as necessary to maintain shutdown
* T-avg greater than 540cF              margin USING EA-68-4. Emergency if any RCP running                    Boration.
OR
* T-cold greater than 540°F if all RCPs stopped.
2 Wednesday, June 05, 2013 8:16:13 AM
 
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1305 NRC RD Exam Notes Question Number:            1 Tier:      1    Group        1 K/A:    007 Reactor Trip EK1 .02 Knowledge of the operational implications of the following concepts as they apply to the reactor trip:
Shutdown Margin Importance Rating:        3.4 / 3.8 1OCFRPart55:            41.8/41.10 1OCFR55.43.b:            Not applicable K/A Match:      This question matches the K/A by having the candidate recall the operational effect on shutdown margin of excessive cooldown of the RCS, a stuck control rod, and the failure of an additional rod to fully insert to 0 steps following a reactor trip on shutdown margin.
Technical
 
==Reference:==
ES-0.1, Reactor Trip Response, Revision 36 Proposed references            None to be provided:
Learning Objective:            0PL271 ES-0. 1
: 6. Given the procedure and a set of initial plant conditions, determine actions required to mitigate the event in progress.
Question Source:
New Modified Bank            X Bank Question History:              WBN bank question 007 Eki .02 201 used on the WBN 10/2011 exam modified to make applicable to SQN and to make a different answer correct.
Comments:
3 Wednesday, June 05, 2013 8:16:14 AM
 
1305 NRC PC Exam
: 2. 008 AK2.01 002 Given the following plant conditions:
          -  A safety injection has occurred.
          -    RCS pressure is 1720 psig and still dropping.
          -    Pressurizer level is rising.
          -    All reactor coolant pumps are in operation.
Which one of the following identifies the leak location?
A Pressurizer safety valve B. Reactor Vessel Head vent line C. Lower pressurizer level tap D. Pressurizer heater well Feedback DIS TRACTOR ANAL YSIS:
A. Correct only a vapor space leak would result in the pressurizer level rising while the pressure continued to drop. Other leaks would result in the level dropping until the pressure stabilized or started to recover.
B. Incorrect, with the leak in this location, the pressure would not be continuing to drop if the pressurizer level were rising. Plausible because the pressurizer level could be rising with a leak on the vessel head vent if SI flow was greater than the leak flow.
C. lncorrect with the leak in the lower PZR level tap, the pressure would not be continuing to drop if the pressurizer level were rising. Plausible because the leak is in the pressurizer.
: 0. Incorrect, with the leak a PZR heater well, the pressure would not be continuing to drop if the pressurizer level were rising. Plausible because the leak is in the pressurizer.
4 Wednesday, June 05, 2013 8:16:14 AM
 
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1305 NRC RO Exam Notes Question Number:            2 Tier:      1    Group      1 K/A:    008 AK2.01 Pressurizer (PZR) Vapor Space Accident (Relief Valve Stuck Open)
Knowledge of the interrelations between the Pressurizer Vapor Space Accident and the following:
Valves Importance Rating:        2.7* / 2.7 1OCFRPart55:            41.7/45.7 1OCFR55.43.b:            Not applicable K/A Match:      Question requires the knowledge of the interrelations of a pressurizer safety valve leak and conditions in the pressurizer during a vapor space accident Technical
 
==Reference:==
WOG E-1 Background document, Rev 2 WOG E-0 Background document, Rev 2 EGT200.713, TAA LOCAs
                                                      -
Proposed references            None to be provided:
Learning Objective:            EGT200.71 3, TAA LOCAs
                                                      -
                                    #11 Describe the dynamic behavior of the reactor, from a thermodynamic and hydrolic point of view, following a LOCA for each of the following catagories:
: c. Vapor Space Break Cognitive Level:
Higher                  X Lower Question Source:
New Modified Bank Bank                  X Question History:              Watts Bar bank question question 008 AK2.01 002 used on the WBN 08/2010 exam Comments:
5 Wednesday, June 05, 2013 8:16:14 AM
 
1305 NRC RO Exam
: 3. 009 EKI.02 003 Given the following plant conditions:
          -  In response to a small break LOCA, the crew is performing ES-i .2, Post LOCA Cooldown and Depressurization.
          -  The next step is to depressurize the RCS to refill the pressurizer.
          -  Core Exit Temperature is 546°F and lowering.
            -  RCS Tavg is 531 °F and lowering.
            -  RCS wide range pressure is 1520 psig.
            -  RCPs have been removed from service.
Which ONE of the following identifies the current RCS subcooling margin and the operational impact if subcooling is lost during the depressurization?
A. 53°F; The RCS cooldown will stop.
B. 68°F; The RCS cooldown will stop.
C 53°F; Cause rapid increase in Pressurizer level.
D. 68°F; Cause rapid increase in Pressurizer level.
Wednesday, June 05, 2013 8:16:14 AM 6
 
1305 NRC RO Exam Feedback DIS TRACTOR ANAL YSIS:
A. Incorrect, RCS subcooling is 53°F is correct, but the loss would not prevent the initiated RCS coo/down from the previous step. Plausible because 53°F subcooling is correct and because natural circulation could be affected by steam voids and the presence of head voiding would affect natural circulation cooling but not stop it. In addition, with the sudden pressure drop (RCS pressure 1000 psig) that caused subcooling to go to zero, ECCS flow is going to increase to the point that it will cool the RCS rapidly.
B. Incorrect, RCS subcooling is not 68°F (it is 53°F) Plausible because 68°F could be calculated if the Tavg were used instead of Core Exit Thermocouple temperature.
599°F 531°F = 68°F and because natural circulation could be affected by steam
                    -
voids and the presence of head voiding would affect natural circulation cooling but not stop it. In addition, with the sudden pressure drop (RCS pressure 1000 psig) that caused subcooling to go to zero, ECCS flow is going to increase to the point that it will cool the RCS rapidly.
C. Correct, with core exit temperature 546°F and saturation for 1520 psig (1535 psia) being 599°F, the RCS subcooling is 53°F. The loss of subcooling could result in upper head voiding causing a rapid increase in pressurizer level as stated in the caution preceding the procedure step.
D. Incorrect, RCS subcooling is not 68°F (it is 53°F) and the loss would not prevent the initia ted RCS cooldown from the previous step. Plausible because 68°F could be calculated if the Tavg were used instead of Core Exit Thermocouple temperature. 599°F 531°F = 68°F and because the rapid increase in pressurizer
                                    -
level is correct.
Wednesday, June 05, 2013 8:16:14 AM 7
 
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1305 NRC RO Exam Notes Question Number:          3 Tier:    1    Group      1 K/A:    009 Small Break LOCA EK1 .02 Knowledge of the operational implications of the following concepts as they apply to the small break LOCA:
Use of steam tables Importance Rating:        3.5 / 4.2 1OCFRPart55:            41.8/41.10/45.3 1OCFR55.43.b:            Not applicable K/A Match:      K/A is matched because the questions requires knowledge of how to use steam tables to determine subcooling and the implication of how the the loss of subcooling could reult in or explain conditions of the unit.
Technical
 
==Reference:==
Steam Tables, ALSTOM Power Services ES-i .2, Post LOCA Cooldown and Depressurization, Rev 17 Proposed references            Steam Tables to be provided:
Learning Objective:            0PL271 ES-i .2
                                    #4 Summarize the mitigating strategy for ES-i .2.
                                    #5 Describe the basis of all limits, notes and cautions.
Question Source:
New Modified Bank Bank                  X Question History:              SON Bank Question 009 EK1 .02 002 used on the SON 1/2009 RETAKE exam Comments:
Qu Wednesday, June 05, 2013 8:16:14 AM 8
 
1305 NRC RO Exam
: 4. 011 EG2.1.31 004 Given the following plant conditions:
        -      Unit 1 is initially at 100% power when an event occurred.
          -    The crew has just restored power to FCV-63-1 at step 6 of ES-i .3, Transfer to RHR Containment Sump.
          -    RWST level is 20% and lowering.
          -    CNTMT pressure is 1 .5 psig and lowering.
          -    CNTMT sump level is 41% and rising.
          -    i-HS-63-72A, CNTMT Sump SuctTo RHR Pump 1A, Red light LIT
          -    1 -HS-63-73A, CNTMT Sump Suct To RHR Pump 1 B, Red light LIT
          -    1-HS-72-39A, CNTMT Spray Hdr 1A Isol, Green light LIT
          -    i-HS-72-2A, CNTMT Spray Hdr lB Isol, Green light LIT Which of the following correctly complete the statement below?
Based on the above indications the RHR CNTMT Sump suction valves              (1) in the expected positions and the CNTMT Spray Header isolations          (2) in the expected positions.
A (1) are (2) are B. (1)are (2) are NOT C. (1) are NOT (2) are D. (1) are NOT (2) are NOT Wednesday, June 05, 2013 8:16:14 AM                                                    9
 
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1305 NRC RO Exam Feedback DISTRACTOR ANAL YSIS:
A. Correct, With RWST level less than 27% the RHR suction auto switchover should have occurred and the 72 and 73 valves should be open and is normal. The CNTMT Spray Hdr discharge valves will be open as well with CNTMT pressure above the phase B setpoint and CSPs running. At this point only one CSP would be running, however its discharge valve is left open. The discharge valves would have been closed if CNTMT pressure had dropped to less than 2 psig and the CSPs secured. This action would have occurred by the time the RWST level had went from 27% to 20%.
B. Incorrect, The first part is correct. The second part is incorrect as both CS discharge Hdr valves would be open. Plausible as the second step in ES-1.3 secures one spray pump and it is logical to shut the header isolation valve for that pump. In fact the RNO for this step secures both CSPs and closes the Discharge Hdr Isolation valves.
C. Incorrect, The RHR Sump suction valves will be open on the auto switchover on RWST level <27%. This is plausible as other ECCS suction valves do not auto switch over (SIPs). The second part is correct.
D. Inorrect, The RHR Sump suction valves will be open on the auto switchover on RWST level <27%. This is plausible as other ECCS suction valves do not auto switch over (SIPs). The second part is correct. The second part is incorrect as both CS discharge Hdr valves would be open. Plausible as the second step in ES- 1.3 secures one spray pump and it is logical to shut the header isolation valve for that pump. In fact the RNO for this step secures both CSPs and closes the Discharge Hdr Isolation valves.
Wednesday, June 05, 2013 8:16:14AM 10
 
1305 NRC AC Exam Notes Question Number:          4 Tier:    1      Group      1 K/A:    011 Large Break LOCA EG2.1 .31 Ability to locate control room switches, controls, and indications and to determine that they are correctly reflecting the desired plant lineup.
Importance Rating:        4.6 / 4.3 10 CFR Part 55:        41.10 1OCFR55.43.b:            Not applicable K/A Match:      Applicant is required assess indications and determine if the correct lineup exist for current plant conditins during a LB LOCA.
Technical
 
==Reference:==
ES-1.3, Transfer to RHR Containment Sump, Rev 19, 1-AR-M6-E, E-3 R23 Proposed references            None to be provided:
Learning Objective:          OPT200.ECCS
                                    #1 Describe the purpose and/or functions of the ECCS amd subsystems, and major components:
: c. valves automatically operated upon SUI actuation 0PL271 ES-i .3 B.5.6 a&b Given a set of initial plant conditions use ES-i .3 to correctly:
: a. Identify required actions
: b. Respond to Contigencies Question Source:
New                  X Modified Bank Bank Question History:            New for NRC ILT 1305 Exam Comments:
Wednesday, June 05, 2013 8:16:14 AM                                                          11
 
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1305 NRC RO Exam 5.015 AG2.I.28005 Given the following plant conditions:
            -    Unit 2 is operating at 100% power.
            -    FS-62-1 1 REAC COOL PMPS SEAL LEAKOFF HIGH FLOW alarm is LIT.
            -    LS-62-45A REAC COOL PMP 4 STANDPIPE LVL HIGH-LOW alarm is LIT
            -    No. 1 seal leakoff flow recorder for RCP #4 indicates 7 gpm.
Which ONE of the following completes the statements below:
(1)  The #2 seal on RCP #4 becomes a                      in response to these conditions.
(2)    In accordance with AOP-R.04, Reactor Coolant Pump Malfunctions, the minimum No. 1 Seal Leakoff flow that would require a reactor trip and shutdown of
            #4RCPis {
A (1) film riding seal (2) 8 gpm
: 0. (1) film riding seal (2) 9 gpm C. (1)    rubbing face seal (2)  8 gpm D. (1)    rubbing face seal (2)    9gpm Wednesday, June 05, 2013 8:16:14AM 12
 
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1305 NRC RO Exam Feedback DISTRACTOR ANAL YSIS:
A. Correct, According to plant design, the #2 seal in the RCP seal package is a pressure balanced and spring loaded rubbing face type seal. Upon a loss of the #1 seal, full RCS pressure is applied to the #2 seal. When applied full pressure, #2 seal designed to convert from a rubbing-face type seal to a film-riding seal. Also in accordance with AOP-R.04, if an RCP experiences high seal leak-off flow which is greater than 6 gpm but less than 8 gpm, the operators are directed to ensure that a minimum of 9 gpm seal injection flow is maintained to the affected seal package until the RCP can be shutdown.
B. Incorrect, Plausible since the first part is correct, the #2 seal package is normally a face rubbing type seal during normal operation and converts to a film rubbing type seal during times of a failure of #1 seal. Also plausible if the candidate gets confused on the minimum seal injection flow and thinks that the normal minimum of 6 gpm is adequate for these conditions.
C. Incorrect, Plausible if the candidates gets confused on the design of the RCP seals and determines that the #2 seal converts from a film riding seal (which is the design of #1 seal) to a rubbing face type seal when full RCS pressure is applied. Also the second part is correct, if high leak off flow is seen on an RCP seal package, AOP-R.04 directs the operators to maintain a minimum of 9 gpm of seal injection flow to any RCP with a seal leak-off flow of greater than 6 gpm but less than 8 gpm.
D. Incorrect, Plausible if the candidates gets confused on the design of the RCP seals and determines that the #2 seal converts from a film riding seal (which is the design of #1 seal) to a rubbing face type seal when full RCS pressure is applied. Also plausible if the candidate gets confused on the minimum seal injection flow and thinks that the normal minimum of 6 gpm is adequate for these conditions.
Wednesday, June 05, 2013 8:16:14 AM                                                                13
 
1305 NRC RO Exam Notes Question Number:          5 Tier:    1      Group      1 K/A:    015 Reactor Coolant Pump (RCP) Malfunctions AG 2.1 .28 Knowledge of the purpose and function of major system components and controls.
Importance Rating:        4.1 /4.1 10 CFR Part 55:          41.7 1OCFR55.43.b:            Not applicable K/A Match:      This question matches the K/A by having the candidate identify RCP seal design that mitigates an RCP seal malfunction and the operator actions necessary to minimize the effect of a failed RCP seal.
Technical
 
==Reference:==
AOP-R.04, Reactor Coolant Pump Malfunctions OPT200.RCP STG rev 8 Proposed references            None to be provided:
Learning Objective:            OPT200. RCP obj:
: 1. Describe the purpose and/or functions of the Reactor Coolant Pumps (RCPs) subsystems, and the major system components listed below:
: g. Shaft seals
: 3. Given plant conditions, Determine if any on the following RCP alarms would be present and Describe actions required by the ARP:
: h. FS-62-1 1, Reac Cool PMP5 Shaft Seal Leakoff High Flow.
: j. RCP 1 (2) (3) (4) Standpipe Level HI/Lo Question Source:
New                  X Modified Bank Bank Question History:              New question written for 1305 NRC exam Comments:
Wednesday, June 05, 2013 8:16:14AM 14
 
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1305 NRC RO Exam
: 6. 022 AK3.05 006 Given the following plant conditions:
            -    Due to indications of gas binding of the 1A CCP, the operators have entered AOP-M.09, Loss of Charging and secured the pump.
Which ONE of the following identifies the reason that plant power should remain stable?
A If a power reduction is required, a manual trip must be initiated due to the inability to borate the RCS.
B. If a power reduction is required the rate of change would be limited to being lowered at no greater than 2% per minute in accordance with AOP-C.03, Rapid Shutdown or Load Reduction using control rods.
C. Reactor power should remain stable until a CCP can be started for RCP seal injection to prevent possible damage to RCP seals caused by changes in temperature.
D. Reactor power should remain stable because the crew will not be able to borate to compensate for the initial Xenon effects.
Feedback DIS TRACTOR ANAL YSIS:
A. Correct, In accordance with AQP-M.09, Note prior to step 21 If any condition in the following step is met, the reactor will be tripped. Manual reactor trip is used to shut down the reactor due to inability to borate. Any condition which would cause a plant transient should be avoided so as to prevent the need for a reactor trip.
B. Incorrec1, Plausible as this is a caution in AOP-C.03, Rapid Shutdown or Load Rejection, for loss of normal and emergency boration. Shutdown is still allowed using the RWST if the CCPs where still available. It is not until step 6 this procedure that you cannot continue the shutdown until you can get some kind of boration going to the RCS.
C. Incorrect, Plausible as this is a concern and a caution just prior to step 6 in AOP-M.09. In this case however in this case the RCP seals should be able to stay below their high temperature limits with CCS still running. The high RCP temperature steps in M.09 are trip criteria and not reasons to keep power stable.
D. Incorrect, Plausible as there will be xenon effects that the crew would be challenged to control with just rods and having no boration. However, the reason is not being able to keep up with power defect reactivity effects as turbine load is reduced.
Wednesday, June 05, 2013 8:16:14 AM 15
 
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1305 NRC RO Exam Question Number:          6 Tier:    1    Group      1 K/A:    022 Loss of Reactor Coolant Makeup AK3.05 Knowledge of the reasons for the following responses as they apply to Loss of Reactor Coolant Makeup:
Need to avoid plant transients Importance Rating:        3.2 / 3.4 10 CFR Part 55:          41.5, 41.10 1OCFR55.43.b:            Not applicable K/A Match:      This question matches the KA as it requires the examinee to address a loss of reactor coolant makeup and determine the reason to avoid a plant downpower.
Technical
 
==Reference:==
AOP-M.09, Loss Of Charging R5 AOPC.03, Rapid Shutdown or Load Reduction R27 Proposed references            None to be provided:
Learning Objective:            OPL271AOP-M.09 #s 4 & 8 Question Source:
New                    X Modified Bank Bank Question History:            New question written for 1305 ILT exam Comments:
Wednesday, June 05, 2013 8:16:14AM 16
 
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1305 NRC RO Exam
: 7. 025 AK2.05 007 Given the following plant conditions:
            -    Unit 1 was operating at 100% power when a LOCA occurred.
            -    Both trains of ECCS pump suction have been transferred to the containment sump in accordance with ES-i .3, Transfer to RHR Containment Sump.
            -    The 1 B RHR pump beings to cavitate.
            -  Containmnt pressure is 4.1 psig and slowly trending down.
Which ONE of the following completes the statements below?
The conditions above require the operating crew to ensure    jJ    are stopped and placed to Pull to Lock.
After stopping required pumps in (1) above, if 1A RHR pump starts cavitating, t?L are required to be stopped.
A. (1) only the 1 B RHR pump and one Cntmt Spray pump (2) all running ECCS and Containment Spray pumps B. (1) only the 1 B RHR pump and one Cntmt Spray pump (2) only all running ECCS pumps C (1) 1 B RHR, one CCP, one SI pump and one Cntmt Spray pump (2) all running ECCS and Containment Spray pumps D. (1) lB RHR, one CCP, one SI pump and one Cntmt Spray pump (2) only all running ECCS pumps Wednesday, June 05, 2013 8:16:14 AM 17
 
1305 NRC RO Exam Feedback DISTRACTOR ANAL YSIS:
A. Incorrect: Incorrect as in addition to the RHR and CSP, the SI and CCP on that train are secured too per ES-1.3. This is plausible as there are places within ES-1.3 where the suction to the SI and CPPs are from the RWST with the RHR pumps from the sump and unaffected by the loss of a RHR pump. In addition, shutting down the same train CPP and SIP logically would be dependent on the pumps themselves cavitating verses shutting down the RHR pump. The second part is correct.
B. Incorrect: Incorrect as in addition to the RHR and CSP, the SI and CCP on that train are secured too per ES- 1.3. This is plausible as there are places within ES-1.3 where the suction to the SI and CPPs are from the RWST with the RHR pumps from the sump and unaffected by the loss of a RHR pump. In addition, shutting down the same train CPP and SIP logically would be dependent on the pumps themselves cavitating verses shutting down the RHR pump. The second part is plausible as the CSP is directed to be secured only if it is cavitating in ES- 1.3, step 23 RNO second IF/THEN step. Once the transition is made to ECA-1.3 the CSP is directed to be stopped. A conclusion could be made that only the ECCS pumps should be secured.
C. Correct: Per ES-1.3 step 23, if a RHR pump starts cavitating the affected train of ECCS pumps and the CSP on that train are placed in PTL. Per the RNO of that step, once the second RHR pump cavitates and is stopped, a transition is made to ECA-1.3, CNTMT Sump Blockage and all ECCS pumps and CSPs are secured.
These actions could also be applied using step 23 again.
D. Incorrect: The first part is correct. The second part is plausible as the CSP is directed to be secured only if it is cavitating in ES- 1.3, step 23 RNO second IF/THEN step. Once the transition is made to ECA-1.3 the CSP is directed to be stopped. A conclusion could be made that only the ECCS pumps should be secured.
Wednesday, June 05, 2013 8:16:14 AM                                                            18
 
1305 NRC PC Exam Notes Question Number:            7 Tier:      1    Group        1 K/A:      APE 025 Loss of Residual Heat Removal System (RHRS)
AK2.05 Knowledge of the interrelations between the Loss of Residual Heat Removal System and the following, Reactor building sump Importance Rating:          2.6 / 2.6 10 CFR Part 55:          41.7 / 45.7 1OCFR55.43.b:            Not applicable K/A Match:    This question matches the K/A by having the candidate determine how the level of water in the Containment building sump affects the ability of RHR pumps to draw water for supply to the ECCS pumps during recirculation.
Technical
 
==Reference:==
ES-i .3, Transfer RHR to CNTMT Sump P19 ECA-1 .3, CNTMT Sump Blockage R2 Proposed references            None to be provided:
Learning Objective:            0PL271 ES-i .3,
                                      #4 Summarize the mitigating strategy for ES-i .3 0PL271 ECA-1 .3,
                                      #4 Summarize the mitigating strategy for ECA-1 .3 Question Source:
New Modified Bank Bank                    X Question History:              SON bank question 025 AK2.05 007 used on the SQN 1/2008 exam with wording changes in the stem.
Comments:
Wednesday, June 05, 20138:16:14 AM                                                        19
 
1305 NRC RO Exam 8.026 AAI.06 008 Given the following:
        -    Unit 1 is in Mode 3 with RCS cooldown in progress.
        -    Unit 2 is operating at 100% power.
        -    Spent Fuel Pool Cooling is being supplied from Unit 1 Train A CCS.
          -  A leak upstream of the 1A CCS heat exchangers required all Unit 1 Train A CCS to be shutdown.
Which ONE of the following completes the statement below?
The cooling supply to the Spent Fuel Cooling System will be realigned to be supplied from fl.
After CCS is restored to the in-service Spent Fuel Pool Heat Exchanger, if the CCS flow rate is 3400 gpm through the heat exchanger, the flow rate is required to be { in accordance with 0-SO-78-1, Spent Fuel Pit Coolant System.
LII A. Unit 1 CCS Train B            lowered B. Unit 1 CCS Train B            raised C Unit 2 CCS Train A                lowered D. Unit 2 CCS Train A            raised Wednesday, June 05, 2013 8:16:14 AM 20
 
_
_____
_____
1305 NRC_RO_Exam Feedback DISTRACTOR ANAL YSIS:
A. Incorrect, Plausible because a leak in any A header of a coo/mg system would be replaced by the B header. This is incorrect as the supply is shifted to the unaffected units A CCS header. The second part is correct.
B. Incorrect, Plausible because a leak in any A header of a cooling system would be replaced by the B header. This is incorrect as the supply is shifted to the unaffected units A CCS header. The second part is incorrect as f/ow is too high and will need to be lowered. Plausible as flowrate could change with the shift to the new header and 3400 is close to the max flow of 3300.
C. Correct The supply will be realigned to the Unit 2 Train A header and 3400 gpm is above the maximum allowed CCS flow rate of 3300 and the SO directs the throttling of 0-FCV 1 1(handswitch on O-M-27B) to adjust and control the flow rate.
D. Incorrect, The first part is correct. The second part is plausible as flowrate could change with the shift to the new header and 3400 is close to the max flow of 3300.
Wednesday, June 05, 2013 8:16:14 AM 21
 
_
_____
1305 NRC RO Exam Notes Question Number:            8 Tier:    1    Group      1 K/A:    026 Loss of Component Cooling Water (CCW)
AA1 .06 Ability to operate and / or monitor the following as they apply to the Loss of Component Cooling Water:
Control of flow rates to components cooled by the CCWS Importance Rating:        2.9 / 2.9 10 CFR Part 55:          41 .7 I 45.5 / 45.6 1OCFR55.43.b:            Not applicable K/A Match:      K/A is matched because the question requires applicant to operate the CCS system and control flow rates of the CCS to the SEP HX.
Technical
 
==Reference:==
SO-78-1, Spent Euel Cooling System, Revision 0059 EA-70-1, CCS Operation R4 Proposed references            None to be provided:
Learning Objective:            OPT200.CCS
                                      #10 Given specific plant conditions, ANALYSE the effect taht a loss of CCS will have on the following:
: a. All loads cooled by the CCS Question Source:
New                      X Modified Bank Bank Question History:              New question for the SQN 05/2013 exam Comments:
Wednesday, June 05, 2013 8:16:14 AM 22
 
1305 NRC RO Exam
: 9. 029 EAI.09 009 Given the following:
          -    Unit 2 is operating at 100% power when a turbine trip occurs.
          -    Control rods begin inserting with 2-HS-85-51 10, ROD CONTROL MODE SELECTOR, in AUTO.
          -    Sl-412, ROD SPEED, indicates 72 steps/mm.
          -    The reactor fails to trip and cannot be tripped from the MCR reactor trip handswitches.
          -    The crew enters the EOP network and has transitioned to FR-S.1, Nuclear Power Generation / ATWS.
Which ONE of the following completes the statement below in accordance with FR-S.1?
FR-Si A. allows ROD CONTROL to remain in AUTO until the rod insertion rate first drops to less than 64 steps/mm before requiring manual rod insertion By allows ROD CONTROL to remain in AUTO until the rod insertion rate first drops to less than 48 steps/rn in before requiring manual rod insertion C. requires ROD CONTROL to be immediately placed in MAN and insertion continued resulting in the ROD SPEED indicating 64 steps/mm D. requires ROD CONTROL to be immediately placed in MAN and insertion continued resulting in the ROD SPEED indicating 48 steps/mm Wednesday, June 05, 2013 8:16:15 AM                                                      23
 
1305 NRC RO Exam Feedback DIS TRACTOR ANAL YSIS:
A. Incorrect, allowing Rod control to remain in AUTO is correct but 64 steps/mm is not the minimum speed allowing AUTO operation. Plausible because 64 steps/mm is the speed of the shutdown rod movement.
B. Correct, FR-S. 1 Step 1 RNO states iF reactor trip breakers will NOT open, THEN MAINTAIN auto or manua rod insertion at max achievable rate UNTIL rods are at bottom. When Rod Control is in manual the speed is set at 48 steps/mm.
As long as rod are moving at a sped above the manual speed, ROD control can be left in AUTO.
C. lncorrect, plausible as FR-S. 1 does allow going to manual contro4 but does not require it. The second part is plausible because 64 steps/mm is the speed of shutdown rod movement.
D. Incorrect, plausible as FR-S. 1 does allow going to manual contro4 but does not require it. The second part is correct.
Wednesday, June 05, 2013 8:16:15 AM                                                              24
 
1305 NRC RO Exam Notes Question Number:            9 Tier:      1    Group        1 K/A:    029 Anticipated Transient Without Scram (ATWS)
EA1 .09 Ability to operate and monitor the following as they apply to a ATWS:
Manual rod control Importance Rating:        4.0 / 3.6 10 CFR Part 55:          41.7/45.5 / 45.6 1OCFR55.43.b:            Not applicable K/A Match:      Question requires the operation and monitor of control rod insertion during an ATWS event Technical
 
==Reference:==
FR-S.1, Nuclear Power Generation / ATWS, Revision 23 Proposed references            None to be provided:
Learning Objective:            0PL271 FR-S.1
: 3. Given a set of initial plant conditions, determine initial Operator response to stabilize the plant, including applicable Immediate Actions of FR-S.1.
Question Source:
New                      X Modified Bank Bank Question History:              New question for the SQN 05/2013 exam Comments:
Wednesday, June 05, 2013 8:16:15 AM                                                                  25
 
1305 NRC RO Exam
: 10. 040 AA2.02 010 Given the following:
Unit 1 is operating at 67% power steady state conditions with Rod Control
          -
in Manual.
          -  A transient occurs resulting in the following:
Rx Power      Turb Power      Tavci      RCS Press          MWe 0700      67%            67%        567&deg;F      2235 psig      785 MWe 0701      68%            66.5%      565&deg;F      2231 psig      780 MWe 0702      69%            66.5%      563&deg;F      2227 psig      770 MWe 0703      70%            65.5%      561&deg;F      2223 psig      765 MWe 0704      71%            65.5%      560&deg;F      2219 psig      760 MWe Which ONE of the following completes the statement below?
The earliest time the conditions require a reactor trip to be initiated in accordance with AOP-S.05, Steam or Feedwater Leak, is A. 0701 B. 0702 C    0703 D. 0704 Feedback DISTRACTOR ANAL YSIS:
A. Incorrect, Plausible because the conditions indicate an uncontrolled rise in reactor power. This is incorrect as the trip criteria for this is the main turbine off line.
B. Incorrect, Plausible as the difference between turbine and reactor power is close to 3%, which is close to 35 Mwe (30 Mwe) and is another set of trip criteria.
C. Correct, a reactor power rise of 3% is the first time trip criteria is met.
: 0. Incorrect, Plausible because the MWe has changed by greater than 35 MWe and there are conditions in AQP-S.05 where MWe changing by greater than 35 MWe requires a reactor trip is correct. Incorrect as this is not where trip criteria is met at the earliest.
Wednesday, June 05, 2013 8:16:15 AM                                                                    26
 
1305 NRC RO Exam Notes Question Number:            10 Tier:      1    Group        2 K/A:      040 Steam Line Rupture AA2.02 Ability to determine and interpret the following as they apply to the Steam Line Rupture:
Conditions requiring a reactor trip Importance Rating:        4.6 / 4.7 10 CFR Part 55:          43.5 /45.13 1OCFR55.43.b:            n/a K/A Match:      Question match the KA because the applicant is required to interpret data to determine the conditions that require the reactor trip.
Technical
 
==Reference:==
AOP-S.05, Steam or Feedwater Leak, Revision 12 AOP-C-.02, Uncontrolled RCS Boron Concentration Changes, Revision 0008 Proposed references          None to be provided:
Learning Objective:          0PL271 AOP-S.05
: 9. List any condition(s) that require a Reactor trip, Turbine Trip or Safety Injection in AOP-S.05.
Question Source:
New                  X Modified Bank Bank Question History:              New question written for SQN 05/2013 NRC exam.
Comments:
Wednesday, June 05, 2013 8:16:15 AM                                                          27
 
1305 NRC RO Exam 11.054  AK1.01 011 Given the following plant conditions:
          -    Unit2isatl00%RTP.
          -    SIG #4 main feedwater line develops a leak inside containment at the containment penetration wall.
          -    The S/G #4 level is able to be maintained with increased feedwater flow.
          -    Containment pressure is beginning to rise.
          -    Condenser hotwell level is 20 and lowering slowly.
          -    The operating crew enters AOP-S.05, Steam or Feedwater Leak.
          -    While performing AOP-S.05, the AOP directs the operating crew to trip the reactor, SI and close the MSIVs.
Which ONE of the following completes the statements below?
The AOP-S.05 criteria used to trip the reactor, SI and close the MSIVs is due to After the MSIVs are closed, SG #4 pressure will      (2)  drop uncontrolled.
A. imminent loss of hotwell level                will B. imminent loss of hotwell level                will NOT C containment pressure                          will D. containment pressure                          will NOT Wednesday, June 05, 2013 8:16:15 AM                                                      28
 
1305 NRC RO Exam Feedback DISTRACTOR ANAL YSIS:
A. Incorrect, Plausible as imminent loss of hotwell is trip criteria, hotwell level is lowering, but at 20 it has 8 more before the alarm is received. The second part is correct.
B. Incorrect, Plausible as imminent loss of hotwell is trip criteria, hotwell level is lowering, but at 20 it has 8 more before the alarm is received. The second part is plausible as some systems have check valves inside CTMT and would isolate RCS components from the leak. The leak would stop one the feed forward was isolated to that system.
C. Correct, The reactor will be tripped based on CTMT pressure approaching 1.5 psig.
The SG will blow down to containment due to the check valve being outside of CTMT.
D. Incorrect, The first part is correct. The second part is plausible as some systems have check valves inside CTMT and would isolate RCS components from the leak.
The leak would stop once the feed forward was isolated to that system.
Wednesday, June 05, 20138:16:15 AM                                                              29
 
1305 NRC RO Exam Notes Question Number:          11 Tier:      1    Group      1 K/A:      054 Loss of Main Feedwater AK1 .01 Knowledge of the operational implications of the following concepts as they apply to Loss of Main Feedwater (MEW):
MEW line break depressurizes the SIG (similar to a steam line break)
Importance Rating:        4.1 / 4.3 1OCFRPart55:            41.8/41.10/45.3 1 OCFR55.43.b:          Not applicable K/A Match:      The question matches the KA as it requires the examinee to understand the implications of a MEW leak and the status of the affected SG.
Technical
 
==Reference:==
AOP-S.05, Steam or Feedwater Leak, Revision 12 Proposed references          None to be provided:
Learning Objective:          0PL271 AOPS.05
                                    #8 Given a set of initial plant conditions, determine the most likely location of a Steam Line Rupture.
                                    #9 List any condition(s) that require a reactor trip, turbine trip or SI in AOP-S.05 Question Source:
New Modified Bank        X Bank Question History:            Significantly modified SQN bank question AOP-S.05-B.1 which was used on the ILT 1002 Audit Exam.
Comments:
Wednesday, June 05, 20138:16:15 AM                                                            30
 
1305 NRC RO Exam
: 12. 056 G2.4.4 012 Given the following plant conditions:
            -  Unit 1 is operating at 100% rated thermal power when a reactor trip occurs
            -  The following annunicators are observed in alarm:
M26-A, A-5, Diesel GEN lA-A Running  >  40 RPM.
M26-B, A-5, Diesel GEN lB-B Running >40 RPM.
M26-C, A-5, Diesel GEN 2A-A Running  >  40 RPM.
M26-D, A-5, Diesel GEN 2B-B Running  >  40 RPM.
Mi-B, B-2, 6900V Unit BD lB Failure Or Undervoltage.
M26-A, C-7,6900V SD BD iA-A Failure or Bus Feeder UV alarms and clears.
Which ONE of the following completes the statement below?
The crew would respond by performing            (1)      in parallel with applicable EOPs, and the Emergency Diesel status is            (2)
Av (1) AOP-P.Oi, Loss of Offsite Power (2) normal B. (1) AOP-P.01, Loss of Offsite Power (2) abnormal C. (1) AOP-P.05, Loss of Unit 1 Shutdown Boards (2) normal D. (1) AOP-P.05, Loss of Unit 1 Shutdown Boards (2) abnormal Wednesday, June 05, 2013 8:16:15 AM                                                          31
 
1305 NRC RO Exam Feedback DISTRACTOR ANAL YSIS:
A. Correct, If a shutdown board loses its normal power supply and is re-energized by its respective diesel it is considered a partial loss of offsite power and meets the entry criteria for AOP-P.O1. Per EPM-4, the crew would decide to perform AOP-P.O1 in parallel with E-O (typically the AQP would be assigned to one crew member as reader doer). All four diesels start on a UV condition on one shutdown board.
B. Incorrect, The first part is correct. The second part is plausible as it would be logical for a Unit 1 SOB UV condition to start only Unit 1 DGs and the conditions in the stem would be bnormal.
C. Incorrect. The first part is plausible as a loss of the shutdown board would be entry criteria for AOP-P.05. In facI, there is a directed transition from AOP-P.O1 to AOPO5 if the board remains de-energized. The second part is correct.
: 0. lncorrect The first part is plausible as a loss of the shutdown board would be entry criteria for AOP-P.05. In fact, there is a directed transition from AOP-P.O1 to AOPO5 if the board remains de-energized. The second part is plausible as it would be logical for a Unit 1 SOB UV condition to start only Unit 1 DGs and the conditions in the stem would be bnormaI.
Wednesday, June 05, 2013 8:16:15 AM                                                              32
 
1305 NRC RO Exam Notes Question Number:          12 Tier:    1      Group      1 K/A:      056 Loss of Offsite Power G2.4.4 Ability to recognize abnormal indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures.
Importance Rating:        4.5/4.7 10 CFR Part 55:          (CFR: 41.10/ 43.2 / 45.6) 1OCFR55.43.b:            Not applicable K/A Match:      The examinee is given indications of a shutdown board being de-energized and subsequently re-energized by its associated DG with the UV alarm clearing. They then have to choose the correct procedure to use to mitigate the event.
Technical
 
==Reference:==
AOP-P.01 R30 AOP-P.05 R20 0-AR-M26-A A-5 R30 0-AR-M26-A 0-7 R30 1-AR-Mi-B, B-2 R23 EPM-4 R22 Proposed references          None to be provided:
Learning Objective:          OPL271AOP-P.01, # 2 Question Source:
New                  X Modified Bank Bank Question History:            New for SQN ILT 1305 exam Comments:                    Low Cognitive Wednesday, June 05, 2013 8:16:15 AM                                                    33
 
1305 NRC RO Exam
: 13. 062 AA2.01 013 Given the following plant conditions:
            -    Both Units are at 100% power.
            -    ERCW is in normal alignment.
            -    ERCW header 1A & 2A are indicating LOW flow.
The following MCR alarms are LIT:
            -    M-15A Window B-6, MECH EQUIP SUMP LVL HI.
              -  M-27A Window A-i, UNIT 1 HEADER A PRESSURE LOW.
            -    M-27A Window B-3, UNIT 2 HEADER A PRESSURE LOW.
            -    NO OTHER alarms are lit associated with the ERCW system.
Which ONE of the following ERCW conditions accounts for the above indications?
A. Supply header 1AI2A has ruptured in the Yard Area.
B. A discharge header has ruptured in the Yard Area.
C. A rupture has occurred upstream of the 2A strainer.
D. A rupture has occurred in the CCW Intake Pumping Station.
Feedback DIS TRACTOR ANAL YSIS:
A. lncorrecl, Plausible since these would also be indications of a supply header rupture, however there would be High system flow associated with this failure not Low flow.
B. Incorrect, Plausible since a pipe rupture would cause system pressure to do down, but a rupture is this location would be accompanied with high system flow, not low flow.
C. Correct, The diagnostic section (Section 2.1) of AOP-M.O1, Loss of ERCW, uses the annunciators and indications listed in the stem to indicate that a supply header has ruptured upstream of a train A supply strather. Since both Unit 1 and Unit 2 supply headers are cross connected upstream of the strainers a leak or rupture on one strainer will affect the other train.
D. lncorrect, Plausible since a pipe rupture would cause system pressure to do down, but a rupture is this location would be accompanied with high system flow, not low flow. The main ERCW headers go right through the CCW pumping station.
Wednesday, June 05, 20138:16:15 AM                                                                34
 
1305 NRC RO Exam Notes Question Number:            13 Tier:      1    Group      1 K/A:      062 Loss of Nuclear Service Water AA2.01 Ability to determine and intepret the following as they apply to Loss of Nuclear Service Water:
Location of a leak in the SWS Importance Rating:        2.9 I 3.5 10 CFR Part 55:          41.7 1OCFR55.43.b:          Not applicable K/A Match:      Questions matches the KIA by requiring the candidate to interpret the indications presented and determine the location the a leak in the SWS (ERCW).
Technical
 
==Reference:==
AOP-M.01, Loss of ERCW, Revision 23 1 ,2-47W845-5 rev 55 1-AR-Mi 5-A R33 0-AR-M27-A R20 Proposed references          None to be provided:
Learning Objective:          0PL271 .AOP-M.01
                                    #3 Given a set of initial plant conditions, determine initial operator response to stabilize the plant.
Question Source:
New                    X Modified Bank Bank Question History:            New question for the SQN 05/2013 NRC exam.
Comments:
Wednesday, June 05, 2013 8:16:15 AM                                                                35
 
1305 NRC RO Exam
: 14. 065 AK3.08 014 Given the following plant conditions:
            -  Unit 1 is in Mode 3 following a manual reactor trip required due to a control air line break in the non-essential control air header.
            -  The operating crew performed the applicable emergency instructions and has stabilized the plant.
            -  The crew has implemented AOP-M.02, Loss of Control Air, to address the loss of air.
Which ONE of the following identifies a local action that would be required in Auxiliary Building during the performance of AOP-M.02 and why?
A. Close 1 -FCV-32-1 10, Non Essential Air to the RX Bldg Ui lsol.to isolate non-essential loads.
B. Establish local control of SG PORVs to control RCS temperature.
C. Establish local control of AFW LCVs to prevent SG overfill.
D Adjust RCP seal injection flow to minimize PZR level rise.
Wednesday, June 05, 2013 8:16:15 AM                                                    36
 
1305 NRC RO Exam Feedback The air system at SON is comprised of non-essential control air headers, two essential control air headers (also called aux air headers) and the seivice air header. During normal operations the service air compressors supply all the station air needs. During a air leak situation the essential control air headers isolate from the non-essential headers and have their own compressors (low capacity) that supply air for essential loads. For a loss of non-essential control air one of the biggest problems (after the unit tripping) is controlling PZR level, as the charging system (including seal injection flow control) valves fails open, letdown isolates and the PZR will continue to fill. Most of the other critical control valves for maintaining the primary systems are supplied from essential air.
DIS TRACTOR ANAL YSIS:
A. lncorrec1, Plausible because the AOP directs local action to ensure the valves are open in the same section of the procedure that the crew is directed to throttle seal injection. The valve is ensured to be open verses closed which runs counter intuitive with an air leak in progress.
B. lncorreci, Plausible because controlling the SG PORV locally would be required if the loss had been a loss of Essential Control Air (Auxiliary Air) instead of a loss of control air in the turbine building. See AQP-M.02 Section 2.1
                                                              ,
C. Incorrect, Plausible because controlling the AFW LCVs locally would be required if the loss had been a loss of Essential Control Air (Auxiliary Air) instead of a loss of control air in the turbine building. See AOP-M.02 Section 2.1
                                                              ,
D. Correct, The AOP directs local seal injection flow control to minimize PZR level rise.
Wednesday, June 05, 2013 8:16:15 AM                                                                37
 
____
1305 NRC RO Exam Notes Question Number:            14 Tier:      1    Group        1 K/A:      065 Loss of Instrument Air AK3.08 Knowledge of the reasons for the following responses as they apply to the Loss of Instrument Air:
Actions contained in EOP for loss of instrument air.
Importance Rating:        3.7 / 3.9 10 CFR Part 55:          41.5, 41.10 1OCFR55.43.b:            Not applicable K/A Match:      This question matches the KJA by having the candidate recall a local operator action associated with loss of air and the reason for the action.
Technical
 
==Reference:==
AOP-M.02, Loss of Control Air, Revision 0021 Proposed references            None to be provided:
Learning Objective:            0PL271 .AOP-M.02
                                      #3 Given a set of initial plant conditions, determine initial operator response to stabilize the plant.
                                      #7 Given the procedure and a set of initial set of conditions, determine actions requried to mitigate the event in progress.
Cognitive Level:
Higher                  X Lower Question Source:
New Modified Bank Bank                  X Question History:              Significantly modified from WBN 06/2011 NRC exam, Q079 G 2.1.30. SQN ILT 1305 NRC Exam Comments:
Wednesday, June 05, 2013 8:16:15 AM                                                                38
 
1305 NRC RO Exam
: 15. 077 AAI.03 05 Given the following:
          -    Unit 1 is at 100% power.
          -  The Transmission Operator has notified the plant that system grid voltage is high and forecasted to go higher.
          -    The Transmission Operator requests the plant to take in the maximum value of MVARs to help stabilize the grid.
Which ONE of the following transmission lines out of service affects the maximum allowed MVAR incominci value on Unit 1, and how is the adjustment made in accordance with 0-GO-5, Normal Power Operation?
TRANSMISSION LINE                        METHOD OF ADJUSTMENT A.            A 161 KV line                              Exciter Voltage Auto Adjuster B.          A 161 KV line                              Exciter Voltage Base Adjuster C          A 500 KV line                              Exciter Voltage Auto Adjuster D.          A 500 KV line                              Exciter Voltage Base Adjuster Feedback DIS TRACTOR ANAL YSIS:
A. Incorrect, 161 KV line availability determines Unit-2 VAR limit. Plausible because off-site power Operability is determined by 161 yard voltage and exciter auto voltage adjuster is the normal method adjuster is the method of varying VARS.
B. Incorrect, 161 KV line availability determines Unit-2 VAR limit. Plausible because off-site power Operability is determined by 161-yard voltage and the exciter voltage base adjuster is the method of varying VARS if the voltage regulator is not in automatic.
C. Correct, GOl-6 specifies the Unit 1 incoming VAR limits based on 500 KV line availability. 0-GO-5 specifies the Exciter Auto Voltage Adjuster as the means for VAR adjustment unless the voltage regulator is in Manual.
D. Incorrect, The exciter auto voltage adjuster is the normal method of varying VARS.
Plausible because the incoming VAR limits for Unit- lare based on 500 KV line availability.
Wednesday, June 05, 20138:16:15 AM                                                                39
 
1305 NRC RO Exam Notes Question Number:            15 Tier:      1    Group        1 K/A:      077 Generator Voltage and Electric Grid Disturbances AA1 .03 Ability to operate and/or monitor the following as they apply to Generator Voltage and Electric Grid Disturbances:
Voltage Regulator controls Importance Rating:        3.8 I 3.7 10 CFR Part 55:          41.5, 41.10 1OCFR55.43.b:            Not applicable K/A Match:      Questions requires abiltiy to monitor conditions to determine the limit for incoming reactive load and then identify the device used to adjust amps.
Technical
 
==Reference:==
0-GO-5, Normal Power Operation, Revision 78 GOl-6, Apparatus Operations, Revision 148, Section E Proposed references          None to be provided:
Learning Objective:          OPT200GEN
: 7. EXPLAIN the Main Generator design features and/or interlocks that provide the following:
: a. Generator Voltage Regulation
: c. Generator capability, including power factor, VARs and hydrogen pressure Question Source:
New Modified Bank Bank                    X Question History:            SON bank question 077 AA1.03 014 used on the SQN 09/2010 exam.
Comments:
Wednesday, June 05, 2013 8:16:15 AM                                                            40
 
1305 NRC RO Exam
: 16. W/EO4EA2.2 016 Given the following plant conditions:
          -    Following a reactor trip, abnormal radiation was noted in the Aux. Building due to a loss of RCS inventory outside containment.
Which ONE of the following identifies a required action and the subsequent check used to determine whether or not the leak is isolated in accordance with ECA-1 .2, LOCA Outside Containment?
A. Isolate SI pump Cold Leg Injection; Pressurizer level rising B. Isolate SI pump Cold Leg Injection; RCS pressure rising C. Isolate RHR Cold Leg Injection; Pressurizer level rising D Isolate RHR Cold Leg Injection; RCS pressure rising Wednesday, June 05, 2013 8:16:15 AM                                                        41
 
1305 NRC RO Exam Feedback DIS TRACTOR ANAL YSIS:
A. Incorrect; Plausible because portions of SI discharge piping have a design pressure of 1750 psig which is 500 psig below normal system pressure. This is not part of the isolation strategy for ECA-1.2. Also, Pressurizer level rising is plausible since the student could reason that it may be rising if the leak was isolated. The procedure directs the use of RCS pressure increasing as the method used to indicate the leak has been isolated.
B. Incorrect; Plausible because portions of SI discharge piping have a design pressure of 1750 psig which is 500 psig below normal system pressure. Also the
                                    -
procedure directs the use of RCS pressure increasing as the method to indicate the leak has been isolated.
C. Incorrect; Plausible because isolation of RHR Cold Leg injection is a strategy contained in the procedure to isolate a LOCA outside CNMT. Pressurizer level rising is a plausible indication since the student could reason that it may be rising if the leak was isolated. The procedure directs the use of RCS pressure increasing as the method used to indicate the leak has been isolated.
D. Correct; Isolation of RHR Cold Leg injection is a strategy contained in the procedure to isolate a LOCA outside CNMT. The procedure directs the use of RCS pressure increasing as the method used to indicate the leak has been isolated.
Wednesday, June 05, 2013 8:16:15 AM                                                                  42
 
_____
1305 NRC RO Exam Notes Question Number:            16 Tier:    1    Group        1 K/A:      W/E04 LOCA Outside Containment (CTMT)
EA2.2 Ability to determine and interpret the following as they apply to the (LOCA Outside CTMT):
Adherence to appropriate procedures and operation within the limitations in the facilitys license and amendments.
Importance Rating:          3.6 / 4.2 10 CFR Part 55:          41.7/45.5/45.6 1OCFR55.43.b:            Not applicable K/A Match:      This question matches the KJA by having the candidate determine the correct actions given a LOCA outside CNMT and the expected plant response for the actions taken.
Technical
 
==Reference:==
ECA-1 .2, LOCA Outside Containment, Revision 10 DWG 47W810-1 Proposed references            None to be provided:
Learning Objective:            OPL271 ECA-1 .2
: 4. Summarize the mitigating strategy for ECA-1 .2.
: 6. Given the procedure and a set of initial plant conditions, determine actions required to mitigate the event in progress.
Question Source:
New Modified Bank Bank                    X Question History:              SQN bank question used on the SON 2010 exam Comments:
Wednesday, June 05, 2013 8:16:15 AM                                                              43
 
1305 NRC RO Exam
: 17. W/E05 EK2.2 017 Given the following:
          -    The crew is implementing FR-H.1, Loss of Secondary Heat Sink.
          -    CST level is 25%.
          -    No Steam Generator is Intact.
Which ONE of the following identifies the preference for restoring a SG as a heat sink and the order in which the feed water sources are attempted in accordance with FR-H.1, Loss of Secondary Heat Sink?
A. Feed a ruptured SG before feeding a faulted SG; MEW, Condensate, MDAFW using ERCW B Feed a ruptured SG before feeding a faulted SG; MDAEW, TDAFW, MEW, Condensate C. Feed a faulted SG before feeding a ruptured SG; MEW, Condensate, MDAEW using ERCW D. Feed a faulted SG before feeding a ruptured SG; MDAFW, TDAFW, MEW, Condensate Feedback DISTRACTOR ANAL YSIS:
A. Incorrect, The priority of using a ruptured Steam generator before using a faulted steam generator is correct but the order of water sources is not correct. Plausible as if is most likely that the AFW systems have failed to require entry into H. 1. Then MFW, Condensate and MDAFW with Essential Raw Water cooling, which is the emergency backup source for MDAFW B. Correct, Step 7 of FR-H. 1 lists the priority of SG to feed and following steps list order of preference for source of feedwater. The preference is to use a ruptured steam generator before using a faulted steam generator.
C. Incorrect Plausible if it is determined that a faulted steam generator would have priority over a ruptured SG due to spread of contamination when a ruptured SG is steamed for heat removal. The second part is plausible as if is most likely that the AFW systems have failed to require entry into H. 1. Then MFW, Condensate and MDAFW with Essential Raw Water cooling, which is the emergency backup source for MDAFW.
: 0. Incorrect, Plausible if it is determined that a faulted steam generator would have priority over a ruptured SG due to spread of contamination when a ruptured SG is steamed for heat removal. The second part is correct.
Wednesday, June 05, 2013 8:16:15 AM                                                                44
 
1305 NRC RO Exam Notes Question Number:            17 Tier:    1    Group        1 K/A:      W/E05 Loss of Secondary Heat Sink EK2.2 Knowledge of the interrelationships between the (loss of Secondary Heat Sink) and the following: Facilitys heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility.
Importance Rating:          3.9 I 4.2 10 CFR Part 55:            41.10 1OCFR55.43.b:              Not applicable K/A Match:      This question matches the K/A due to having the candidate identify the sources of water for secondary heat removal following a loss of MFW event and has the candidate recall the order of priority for sources of makeup water and order of SGs to be fed.
Technical
 
==Reference:==
FR-H.1 Loss of Secondary Heat Sink, Revision 18 Proposed references            None to be provided:
Learning Objective:            OPL271FR-H.1
: 4. Summarize the mitigating strategy for FR-H.1.
: 6. Given the procedure and a set of initial plant conditions, determine actions required to mitigate the event in progress.
Question Source:
New Modified Bank Bank                    X Question History:              SQN bank question W/E05 EK2.2 018 used on the SQN 02/2010 exam.
Comments:
Wednesday, June 05, 20138:16:15 AM                                                                45
 
1305 NRC RO Exam
: 18. W/EI1 EK3.4018 Given the following plant conditions:
          -  At 0900 the Unit 1 Reactor Trips.
          -  At 0920 a small break LOCA occurs.
          -  At 0950 the crew transitioned to ECA-1 .1, Loss of RHR Sump Recirculation, due to the failure of both RHR pumps.
          -  Crew has established one train of ECCS flow per ECA-1 .1.
          -  SI flow cannot be terminated due to lack of subcooling.
          -  At 1030 the crew is performing ECA-1 .1 Step 20, Monitor if ECCS flow should be terminated:
          -  RVLIS indications are adequate.
          -  The RNO states Establish minimum ECCS flow:
Which one of the following correctly identifies the minimum ECCS flow rate per Curve 9 of ECA 1 .1 that meets the intent of ECA-1 .1, Step 20 RNO, AND the reason for performing this action?
REFERENCE PROVIDED A 325 gpm ECCS flow. To establish the minimum ECCS flow needed and delay RWST depletion.
B. 325 gpm ECCS flow. To ensure adequate RVLIS indications and establish conditions to start an RCP.
C. 400 gpm ECCS flow. To ensure adequate RVLIS indications and establish conditions to start an RCP.
D. 400 gpm ECCS flow. To establish the minimum ECCS flow needed and delay RWST depletion.
Wednesday, June 05, 2013 8:16:15 AM                                                      46
 
1305 NRC RD Exam Feedback DIS TRACTOR ANAL YSIS:
A. Correct, From 0900 1030 (90 Mm) Using ECA-1. 1, curve 9, the minimum amount
                                  -
of SI flow needed to match decay heat is approximately 325 gpm. The value of 325 gpm is in the acceptable region using the graph from time of trip AND meets the requirement of Minimum Flow to delay RWST depletion. The Basis states the operator is then instructed to establish the minimum ECCS flow needed to match decay heat in order to further decrease ECCS pump Flow and delay RWST depletion.
B. lncorrec1, Plausible since the value of 325 gpm would meet the requirements of the step to match the SI flow needed to match decay heat. Also plausible since reducing SI flow could be assumed to match the flow needed to maintain RVLIS and it is mentioned in the basis that when reducing flow the charging pumps could be realigned from CPIT flow to normal charging for seal cooling, however the actual reason is match decay heat to slow the decrease in RWST level.
C. Incorrect, Plausible since this flow rate would be in the acceptable range on ECA-1. 1 curve 9, however this does not meet the intent of the step which is to reduce flow as low as possible to slow the rate of RWST depletion. Also plausible since reducing SI flow could be assumed to match the flow needed to maintain RVLIS and it is mentioned in the basis that when reducing flow the charging pumps could be realigned from CPIT flow to normal charging for seal cooling, however the actual reason is match decay heat to slow the decrease in RWST level.
D. Incorrect, Plausible since this flow rate would be in the acceptable range on ECA- 1.1 curve 9, however this does not meet the intent of the step which is to reduce flow as low as possible to slow the rate of RWST depletion. Also plausible since the reason is correct per the basis document.
Wednesday, June 05, 2013 8:16:15 AM                                                            47
 
1305 NRC RO Exam Notes Question Number:          18 Tier:    1      Group      1 K/A:      W/E1 1 Loss of Emergency Coolant Recirculation EK3.4 Knowledge of the reasons for the following responses as they apply to the (Loss of Emergency Coolant Recirculation)
RO or SRO function within the control room team as appropriate to the assigned position, in such a way that procedures are adhered to and the limitations in the facilities license and amendments are not violated.
Importance Rating:        3.6 / 3.8 1OCFRParI55:            41.5/41.10 1OCFR55.43.b:            Not applicable K/A Match:      This question matches the K/A by having the candidate recall the reason for the actions that the crew will take to comply with the requirements of ECA-1 .1 while attempting to reduce the depletion rate of the RWST while unable to transfer to CNMT sump recirculation.
Technical
 
==Reference:==
ECA-1.1, rev 12 EPM-3-ECA-1 .1, Basis Document for ECA-1 .1, Loss of ECCS Sump Recirculation rev 5 Proposed references            ECA-1 .1 Loss of ECCS Sump Recirculation, Curve 9, to be provided:                Minimum ECCS Flow for Decay Heat vs. Time After Trip Learning Objective:            0PL271 .ECA-1 .1
: 4. Summarize the mitigating strategy for ECA-1 .1.
: 6. Given the procedure and a set of initial plant conditions, determine actions required to mitigate the event in progress.
Question Source:
New                    X Modified Bank Bank Question History:              New question written for 1305 ILT exam Comments:
Wednesday, June 05, 2013 8:16:15 AM                                                          48
 
1305 NRC RO Exam
: 19. 001 AA2.05 019 Given the following plant conditions:
Attime=TO
            -  Tavg Tref deviation is 0&deg;F.
                        -
              -  Pressurizer level is 45% and stable.
            -  Reactor Power is approximately 75% and stable.
            -  Control Bank D step counters are at 166 steps.
Attime=T+2 mm
              -  Tavg is 2&deg;F > Tref and rising.
              -  Pressurizer level 46% and slowly rising.
              -  Pressurizer spray valves have throttled open.
              -  Reactor Power is approximately 76% and slowly rising.
              -  Control Bank D step counters are at 178 steps and rising at 8 steps per minute Which ONE of the following identifies (1) the correct procedure to enter and (2) the FIRST action that must be performed?
A. (1)    AOP-C.O1, Rod Control System Malfunctions.
(2)  Trip the reactor and enter E-0, Reactor Trip or Safety Injection.
Bw (1)  AOP-C.01, Rod Control System Malfunctions.
(2)  Place the rod control mode selector switch to MANUAL.
C. (1)    AOP-C.02, Uncontrolled RCS Boron Concentration Changes.
(2)  Trip the reactor and enter E-0, Reactor Trip or Safety Injection.
D. (1)    AOP-C.02, Uncontrolled RCS Boron Concentration Changes.
(2)  Place the rod control mode selector switch in MANUAL.
Wednesday, June 05, 2013 8:16:16 AM                                                        49
 
1305 NRC RO Exam Feedback DIS TRACTOR ANALYSIS:
A. Incorrect, The first part is correct. The second part is incorrect as itis the second action required if placing the rods in manual does not stop rod motion. Plausible as it is the step in the RNO column of the correct answer and it may be determined as a correct answer over placing the rod select in manual because the rods are moving in the wrong direction.
B. Correct, The conditions given in the stem meet the entry criteria for AOP-C.O1 and it is the correct procedure to mitigate the event in progress. The first action to take in AOP-C.O1 is to place the rods in manual.
C. lncorrec1, The entry into AOP-C.02 should be considered as an inadvertant dilution could be the cause of the temperature change and temperature, rod and NIS changes are symptoms of an inadvertant boration or dilution. This is incorrect though as clear entry conditions are met for AOP-C.O1. Plausible as it is an action in AOP-C.02.
D. Incorrect, The entry into AOP-C.02 should be considered as an inadvertant dilution could be the cause of the temperature change and temperature, rod and NIS changes are symptoms of an inadvertant boration or dilution. This is incorrect though as clear entry conditions are met for AOP-C.O1. The second part is plausible as it is correct and it is also an action in AOP-C.02 to move the rod control switch to manual and move rods.
Wednesday, June 05, 20138:16:16 AM                                                                  50
 
1305 NRC RO Exam Notes Question Number:          19 Tier:    1      Group      2 K/A:    001 Continuous Rod Withdrawal AA2.05 Ability to determine and interpret the following as they apply to the Continuous Rod Withdrawal:
Uncontrolled rod withdrawal from available indications.
Importance Rating:        4.4 / 4.6 10 CFR Part 55:        43.5/45.13 1OCFR55.43.b:            Not applicable K/A Match:      The question meets the KA in that it requires the applicant to diagnose an uncontrolled rod withdrawal from available indications.
Technical
 
==Reference:==
AOP-C.01 R22 AOP-C.02 R8 Proposed references            None to be provided:
Learning Objective:            0PL271 .AOP-C.01
: 5. Summarize the mitigating strategy for AOP-C.01.
: 7. Given the procedure and a set of initial plant conditions, determine actions required to mitigate the event in progress.
Question Source:
New Modified Bank Bank                  X Question History:              SQN Bank, SQN ILT 1305 NRC Exam Comments:
Wednesday, June 05, 2013 8:16:16 AM                                                          51
 
1305 NRC RO Exam
: 20. 036 AK2.02 020 Given the following plant conditions:
          -    Refueling is in progress on Unit 1 when a report is made to the control room that an irradiated fuel assembly has been dropped.
          -    The following alarms are received on 0-XA-55-12A:
                -  1 -RA-90-1 1 2A CNMT BLDG UP COMPT AIR MON HIGH RAD
                -  1-RA-90-59A AX BLDG AREA RAD MON HIGH RAD
                -  1-RA-90-131A CNTMT PURGE AIR EXH MON HIGH RAD Which ONE of the following identifies the required actions to be taken per AOP-M.04, Refueling Malfunctions?
A. Evacuate non-essential personnel from Containment, and if RM-90-400, Shield Bldg and/or 0-RM-90-1 01, Aux Bldg Vent rad monitors are trending up, manually initiate Containment Ventilation Isolation.
B Evacuate all personnel from Containment and Ensure Containment Ventilation Isolation has Actuated.
C. Evacuate the Fuel Handling Area and maintain SEP and Rx cavity levels as directed by Fuel Handling SRO.
D. Evacuate the immediate area and Verify that Auxiliary Building Isolation has Actuated.
Wednesday, June 05, 2013 8:16:16 AM                                                      52
 
1305 NRC RO Exam Feedback DISTRACTOR ANAL YSIS:
A. Incorrect, Plausible since Evacuate non-essential personnel is the direction given for Reactor Cavity Seal Failure, in that case personnel are needed to stay in containment and perform actions. On fuel assembly drop/damage Containment is directed to be completely evacuated. Second part also is incorrect. If rad levels are rising operators are directed to obtain gas release rate data.
B. Correct, For the given Rad Monitor alarms, the Containment is directed to be evacuated. Also 1-RA-90-131A will cause automatic Containment Ventilation isolation. Operators are directed to verify isolation.
C. Incorrect, Plausible since Evacuate Fuel Handling area is direction in AOP-M.04 but for dropped fuel assembly in the Spent Fuel pit area not inside CNMT. Also second part of distractor is direction for reactor cavity seal failure.
: 0. Incorrect, Plausible since Evacuate the immediate area is direction in AOP-M.04 but for dropped or damaged new fuel assembly. Also second part of distractor is wrong since Aux building Isolation would have to be manually initia ted with the given rad monitor alarms.
Wednesday, June 05, 20138:16:16 AM                                                                53
 
1305 NRC RO Exam Notes Question Number:          20 Tier:    1    Group      2 K/A:      036 Fuel Handling Incidents AK2.02 Knowledge of the interrelations between the Fuel Handling Incidents and the following:
Radiation monitoring equipment (portable and installed)
Importance Rating:        3.4 / 3.9 10 CFR Part 55:          41.7 1OCFR55.43.b:            Not applicable K/A Match:      This question matches the K/A by asking if the candidate knowledge of the interrelations for a Fuel Handling Incident inside containment and the expected automatic actions of high radiation.
Technical
 
==Reference:==
AOP-M.04, Refueling Malfunctions, section 2.2, Rev 7 Proposed references            None to be provided:
Learning Objective:            OPL271AOP-M.04
: 5. Summarize the mitigating strategy for AOP-M.04.
: 7. Given the procedure and a set of initial plant conditions, determine actions required to mitigate the event in progress.
Question Source:
New Modified Bank Bank                  X Question History:              SQN Bank Comments:
Wednesday, June 05, 2013 8:16:16 AM                                                          54
 
1305 NRC RO Exam
: 21. 037AG2.2.44 021 Given the following:
* Unit 2 was at 100% RTP when a reactor trip occurred due to a sheared shaft on
                #1 RCP.
* While the crew was in ES-0.1, Reactor Trip Response with the plant stable, chemistry reports that #2 SG has developed tube leak.
* PZR Level is dropping very slowly with the 2A COP at 114 gpm.
* AOP-R.01, Steam Generator Tube Leak was entered and the crew is evaluating step 1.
The crew should first        (1)      and the crew will use    (2)    during the follow on step to depressurize the RCS in AOP-R.01.
A. (1) isolate letdown (2) auxiliary spray B (1) isolate letdown (2) normal spray C. (1) start the 2B CCP (2) auxiliary spray D. (1)start the 2B CCP (2) normal spray Wednesday, June 05, 2013 8:16:16 AM                                                        55
 
1305 NRC RO Exam Feedback DIS TRACTOR ANAL YSIS:
A. Incorrect: Plausible as isolating letdown is correct. The second part is plausible as normal spray is no longer available due to the loss of some RCPs (#2 RCP is critical) and auxiliary spray is the first choice for a backup if letdown was available.
B. Correct: isolating letdown is the next step if PZR level is still dropping after maximizing one CCP. Normal spray is first choice for RCS depressurization and is available as long as #2 RCP is available.
C. Incorrect: The first part is plausible as it is a valid step in recovering PZR level after LD has been isolated. The second part is plausible as normal spray is no longer available due to the loss of some RCPs (#2 RCP is critical) and auxiliary spray is the first choice for a backup if letdown was available.
D. Incorrect: The first part is plausible as it is a valid step in recovering PZR level after LD has been isolated. The second part is correct.
Wednesday, June 05, 2013 8:16:16 AM                                                                    56
 
1305 NRC RO Exam Notes Question Number:          21 Tier:      1    Group      2 K/A:    037 Steam Generator Tube Leaks G 2.2.44 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.
Importance Rating:        4.2/4.4 10 CFR Part 55:        (CFR: 41.5 / 43.5 / 45.12) 1OCFR55.43.b:            Not applicable K/A Match:      The examinee needs to determine correct actions from given stem indications of system status of PZR level slowly dropping and the running COP at maximized flow. This indication will require action to isolate letdown. The examinee(s) will then need to understand how their actions affected the systems in play with the event during follow on implementation of the AOP.
Technical
 
==Reference:==
AOP-R.O1, SGTL R31 E-0, R35 Proposed references            None to be provided:
Learning Objective:            OPL271AOP-R.O1
: 5. Summarize the mitigating strategy for AOP-R.01.
: 7. Given the procedure and a set of initial plant conditions, determine actions required to mitigate the event in progress.
Cognitive Level:
Higher Lower                  X Question Source:
New                    X Modified Bank Bank Question History:              New question for ILT 1305 NRC Exam Comments:
Wednesday, June 05, 2013 8:16:16 AM                                                            57
 
1305 NRC RO Exam
: 22. 051 AA1.04 022 Given the following:
          -    Unit 1 is operating at 100%.
          -    Main Generator output is 1200 MWe.
          -    Main Turbine is operating in IMP OUT.
          -    Air inleakage causes condenser pressure to change from 0.8 to 1.8 psia.
Which one of the following identifies:
(1)      the effect on the Main Generator output (MWe)
AND (2)      the required rod motion to maintain reactor power at 100%
A. (1)    rise (2)  none B. (1)    rise (2)  withdraw C. (1)    lower (2)  withdraw Dv (1)    lower (2)  none Feedback DIS TRACTOR ANAL YSIS:
A. Incorrect, the first part is incorrect. With condenser vacuum at a higher value, turbine efficency is decreased and generator ouput will lower. This is plausible if examThee confuses vacuum rising with increase generator output. The second part is correct. Since reactor power is a function of steam demand and steam demand has not changed, no rod motion is required.
B. Incorrect, the first part is incorrect (see item A). The second part is incorrect. This is plausible if examinee determThe that since generator output is decreased, reactor power will also decrease since normally a decrease Th generator output would indicate that the turbine load has decreased and reactor power would have decreased, therefore rods will need to be withdrawn to mathtain 100% RTP.
C. Incorrect, the first part is correct. With condenser vacuum at a higher value, turbine efficency is decreased and generator ouput will lower. The second part is Thcorrect (see item B).
D. Correcl, the first part is correct. With condenser vacuum at a higher value, turbine efficency is decreased and generator ouput will lower, the second part is correct.
Since reactor power is a function of steam demand and steam demand has not changed, no rod motion is required (Turbine is in IMPOUT).
Wednesday, June 05, 2013 8:16:16 AM                                                                  58
 
1305 NRC P0 Exam Notes Question Number:          22 Tier:    1      Group      2 K/A:    051 Loss of Condenser Vacuum AA1 .04 Ability to operate and / or monitor the following as they apply to the Loss of Condenser Vacuum: Rod Position Importance Rating:        2.5* / 2.5*
10 CFR Part 55:        41.7/45.5/45.6 1OCFR55.43.b:            Not applicable K/A Match:      Question match the KA because the applicant is required to determine rod position requirements resulting from the loss of condenser vacuum.
Technical
 
==Reference:==
Generic Fundamentals PWR Topic 193004 for turbine efficiency Generic Fundamentals PWR Topic 192008 for relationship of steam flow and reactor power AOP-S.02, Loss of Condenser Vacuum Proposed references            None to be provided:
Learning Objective:            OPL271AOP-S.02
: 3. Given a set of initial plant conditions, determine initial operator response to stabilize the plant.
Question Source:
New                    X Modified Bank Bank Question History:              New question for the SQN 05/2013 exam Comments:
Wednesday, June 05, 2013 8:16:16 AM                                                                59
 
1305 NRC RO Exam
: 23. 059AK3.01 023 ODCM LCO 1 .1.1 states that Liquid Radwaste Effluent Line radiation monitor 0-RM-90-1 22 shall be operable with its alarm setpoint set at a particular value.
Which one of the following completes the statement below?
The alarm setpoint of 0-RM-90-122 is based on                  (1) and upon alarming, the radiation monitor            (2)      terminate the release.
A (1) the limits of 10 CFR 20, Standards for Protection Against Radiation (2) will B. (1) the limits of 10 CFR 20, Standards for Protection Against Radiation (2) will NOT C. (1) the limits of 10 CFR 100, Reactor Site Criteria (2) will D. (1) the limits of 10 CFR 100, Reactor Site Criteria (2) will NOT Feedback DISTRACTOR ANAL YSIS:
A. Correct: the alarm set point is based on not exceeding ten times the limits of 10 CFR 20 and the alarming RM will terminate the release.
B. Incorrect: The first part is correct. The second part is plausible as there are rad monitors at SON that do not termThate the release when they alarm (for example, 0-RM-90-212, Station Discharge RM does not terminate the release).
C. Incorrect: The first part is plausible as 1OCFR 100 establishes and defines the site boundary, exclusion areas, etc and sets radiation limits at the site boundaries.
 
The second part is correct.
a    Incorrect: The first part is plausible as 1OCFR 100 establishes and defines the site boundary, exclusion areas, etc and sets radiation limits at the site boundaries.
 
The second part is plausible as there are rad monitors at SON that do not termThate the release when they alarm (for example, 0-RM-90-212, Station Discharge RM does not terminate the release).
Wednesday, June 05, 2013 8:16:16 AM                                                              60
 
1305 NRC RO Exam Notes Question Number:          23 Tier:    1    Group      2 K/A:    059 Accidental Liquid Pad Waste Release AK3.03 Knowledge of the reasons for the following responses as they apply to the Accidental Liquid RadWaste Release:
Termination of a release of radioactive liquid Importance Rating:        3.5/3.9 10 CFR Part 55:          (CFR41.5,41.1O/45.6145.13) 1OCFR55.43.b:            Not applicable K/A Match:      The question meets the KA as it test the examinee on the reason the rad monitor wouldterminate the release.
Technical
 
==Reference:==
ODCM, R57 0-AR-M12-B, C-i, rev 29 Proposed references            None to be provided:
Learning Objective:            OPT200.LRW #s 4,7,10 & 13 Question Source:
New Modified Bank          X Bank Question History:              Modiified question LRW-B.6 002 Comments:                      Low Cognitive Wednesday, June 05, 2013 8:16:16 AM                                                        61
 
1305 NRC RO Exam
: 24. 074 EA2.07 024 Given the following plant conditions:
            -  A Reactor Trip and Safety Injection has occurred on Unit 1.
            -  While performing the actions of E-0, Reactor Trip or Safety Injection, the following plant conditions are noted:
            -  Containment pressure is 6.5 psig and stable.
            -  RCPs have been stopped.
            -    RVLIS Lower Range is indicating 40%.
            -  Core Exit Thermocouples are indicating 710&deg;F.
            -  PZR level is off scale low.
            -  PZR pressure is 400 psig.
            -  RCS Wide Range Hot Leg Temperatures are indicating 680&deg;F.
Which ONE of the following identifies the accident that has occurred and the required procedure to be entered?
A. A PZR steam space break has occurred and a transition to FR-C.1, Response to Inadequate Core Cooling is required.
B. A PZR steam space break has occurred and a transition to FR-C.2, Response to Degraded Core Cooling is required.
C An RCS hot or cold leg break has occurred and a transition to FR-C.1, Response to Inadequate Core Cooling is required.
D. An RCS hot or cold leg break has occurred and a transition to FR-C.2, Response to Degraded Core Cooling is required.
Wednesday, June 05, 2013 8:16:16 AM                                                      62
 
1305 NRC RO Exam Feedback DISTRACTOR ANAL YSIS:
A. Incorrect; Plausible since a PZR Steam Space accident is a type of Loss of Coolant Accident, however a PZR steam space accident can be identified by indicated PZR level along with low RCS Pressure and Low RVLIS inventory. The candidate may get the RCS indications confused. Also the second part is correct the stated plant conditions do hdicate Inadequate Core Coo/hg.
B. Incorrect, Plausible since a PZR Steam Space accident is a type of Loss of Coolant Accident; however a PZR steam space accident can be identified by indicated PZR level along with low RCS Pressure and Low RVLIS inventory. The candidate may get the RCS indications confused. Also plausible if the candidate gets confused on the Core Exit T/C readings (not 1200&deg;F), or misses the fact that the RCPs have been secured. This would indicate that FR-C.2 Degraded Core Coo/mg would exist.
C. Correct, Due to the loss of inventory and pressure (RCS pressure 400 psig, and RVLIS at 40%) the plant has experienced a Loss of Coolant Accident. As indicated by PZR level (off scale low), the break would be in either the hot leg or cold leg.
Also the accident has progressed to a condition of Inadequate Core Cooling as indicated by RVLIS <42% and CETS >700 &deg;F. lnaccordance with 1-FR-0, Unit 1 Status Trees, the conditions of No RCPS with Core Exit T/Cs >700&deg;F and RVL1S Lower Range of <42% this would be an Inadequate Core Cooling condition.
D. Incorrect, Plausible since due to the loss of inventory and pressure (RCS pressure 400 psig, and RVLIS at 40%) the plant has experienced a Loss of Coolant Accident. As hdica ted by PZR level (off scale low), the break would be in either the hot leg or cold leg and is correct. The second part is plausible since if the candidate gets confused on the Core Exit T/C readings (not 1200&deg;F), or misses the fact that the RCPs have been secured. This would indicate that FR-C.2 Degraded Core Cooling would exist.
Wednesday, June 05, 2013 8:16:16 AM 63
 
1305 NRC RO Exam Notes Question Number:          24 Tier:    1    Group      2 K/A:    074 Inadequate Core Cooling EA2.07 Ability to determine or interpret the following as they apply to a Inadequate Core Cooling:
The difference between a LOCA and inadequate core cooling, from trends and indicators.
Importance Rating:        4.1 / 4.7 10 CFR Part 55:          43.5 / 45.13 1OCFR55.43.b:            Not applicable K/A Match:      This question matches the K/A by having the candidate determine the type of LOCA that has occurred and having them identify that conditions of Inadequate Core Cooling have developed.
Technical
 
==Reference:==
1-FR-0, Unit 1 Status Trees, Core Cooling Proposed references            None to be provided:
Learning Objective:            0PL271 FR-0, r3, FR-0, Status Trees Obj. 2 Given a set of initial conditions, determine if FR-0 entry is required.
Obj. 6 Given a set of initial plant conditions use FR-0 to correctly identify the:
: b. FR procedure applicable to the current state of each Critical Safety Function.
Question Source:
New                    X Modified Bank Bank Question History:              New question written for 1305 exam.
Comments:
Wednesday, June 05, 2013 8:16:16 AM                                                                64
 
1305 NRC RO Exam
: 25. WIEO3EK1.1 025 Given the following plant conditions:
* A small break LOCA has occurred.
* The crew is implementing ES-i .2, Post LOCA Cooldown and Depressurization.
* One Centrifugal Charging Pump (CCP) is running.
* Both Safety Injection Pumps (SIPs) are running.
* The crew has determined that one SIP can be stopped.
Which ONE of the following:
(1) Explains what will happen to the subcooling value when the SIP is stopped And (2) the next pump to be stopped if subcooling is adequate for further ECCS pump reduction?
A. (1) Lowers due to reduced ECCS injection flow and stablizes at a lower value when break flow and ECCS flow equal.
(2) The running CCP B. (1) Remains the same due to reduced ECCS injection flow causing RCS temperature and pressure to rise.
(2) The running CPP C (1) Lowers due to reduced ECCS injection flow and stablizes at a lower value when break flow and ECCS flow equal.
(2) The running SIP D. (1) Remains the same due to reduced ECCS injection flow causing RCS temperature and pressure to rise.
(2) The running SIP Wednesday, June 05, 2013 8:16:16AM                                                      65
 
1305 NRC RO Exam Feedback DISTRA CTOR ANAL YSIS:
A. Incorrect, The first part is correct. The second part is plausible because the SI reduction sequence begins with stopping a charging pump, goes on to an SI pump and it is logical to assume that if further pump reduction is required to go back to a charging pump.
B. Incorrect, Plausible some events (SG FauIt, LBLOCA) will have RCS temperature rise in securing an SIP due to reduce cooling from reduced ECCS flow. On a SG Fauli the dynamics of the SG blowing dry and not being a heat sink causes RCS temperature and pressure to rise reducing ECCS flow with the result of subcooling lowering. The second part is plausible because the SI reduction sequence begins with stopping a charging pump, goes on to an SI pump and it is logical to assume that if further pump reduction is required to go back to a charging pump.
C. Correct, Subcooling will lower in this situation as RCS pressure lowers and stabilize at a lower value when break flow and ECCS flow equal. The next pump to be secured is the second SIP as the procedure will try to keep a CPP in service to establish h normal charging.
D. Incorrect, Plausible some events (SG Fault, LBLOCA) will have RCS temperature rise in securing an SIP due to reduce cooling from reduced ECCS flow. On a SQ FaulI the dynamics of the SG blowing thy and not being a heat sink causes RCS temperature and pressure to rise reducing ECCS flow with the result of subcooling lowering. The second part is correct.
Wednesday, June 05, 2013 8:16:16 AM                                                                66
 
1305 NRC RO Exam Notes Question Number:          25 Tier:    1    Group      2 K/A:      W/E03 LOCA Cooldown and Depressurization EK1 .1 Knowledge of the operational implications of the following concepts as they apply to the (LOCA Cooldown and Depressurization)
Components, capacity, and function of emergency systems.
Importance Rating:        3.4 / 4.0 1OCFR Part 55:          41.8/41.10 1OCFR55.43.b:            Not applicable K/A Match:      This question matches the KJA by having the candidate demonstrate the knowledge of the components used in the SI reduction sequence in post LOCA cooldown and depressurization and the operational implications of securing ECCS pumps (subcooling reduction) and upon securing an ECCS pump which pump will be next in the reduction sequence.
Technical
 
==Reference:==
ES-i .2, Post LOCA Cooldown and Depressurization, rev 18.
EPM-3-ES-i .2, Basis document Proposed references            None to be provided:
Learning Objective:            0PL271 ES-i .2
                                      #4 Summarize the mitigating strategy for ES-i .2
                                      #13 Analyze and explain the process taht leads to a new RCS equalibrium pressure following the shutdown of an ECCS pump during the ES-i .2 reduction sequence.
Question Source:
New Modified Bank          X Bank Question History:              Moidified from SQN bank ES-i .2-B.2 Comments:                      High Cognitive Wednesday, June 05, 2013 8:16:16 AM                                                          67
 
1305 NRC RO Exam
: 26. WIEI4EAI.I 026 Which one of the following identifies the interlocks that must be met before valve FCV-72-23 (Train A Containment Spray Suction from Containment Sump) can be opened?
A. Both FCV-74-3 (RHR Suction from RWST) closed and FCV-72-40 (RHR Discharge to RHR Spray) must be closed.
B. Both FCV-72-40 (RHR Discharge to RHR Spray) and FCV-72-34 (Containment Spray Pump Recirc) must be closed.
C Both FCV-72-22 (Containment Spray Suction from RWST) and FCV-74-3 (RHR Suction from RWST) must be closed.
D. Both FCV-72-34 (Containment Spray Pump Recirc) and FCV-72-22 (Containment Spray Suction from RWST) must be closed.
Feedback DISTRACTOR ANALYSIS:
A. Incorrect, FCV-74-3 is interlocked but FCV-72-40 is not interlocked with opening FCV-72-23. Plausible because FCV-72-3 is correct and FCV-72-40 is interlocked with other valves associated with sump swapover.
B. Incorrect, FCV-72-40 and FCV-72-34 are not interlocked with opening FCV-72-23. Plausible because both FCV-72-40 and FCV-72-34 are containment spray valves. Other pump recirc valves do have interlocks in transferring to the containment sump and. FCV-72-40 is interlocked with other valves associated with sump swapover.
C. CORRECT Both FCV-72-22 (Containment Spray Suction from RWST) and FCV-74-3 (RHR Suction from RWST) must be closed as shown on print 1-47W61 1-72-1.
: 0. lncorrect, FCV-72-34 is not interlocked with opening FCV-72-23. Plausible because the FCV-72-34 is a containment spray valve and other pump recirc valves do have interlocks in transferring to the containment sump.
Wednesday, June 05, 2013 8:16:16 AM                                                          68
 
1305 NRC RO Exam Notes Question Number:          26 Tier:    1    Group      2 K/A:      W/E14 High Containment Pressure EA1 .1 Ability to operate and/or monitor the following as they apply to the (High Containment Pressure):
Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Importance Rating:        3.7 / 3.7 10 CFR Part 55:          41.7 1OCFR55.43b:            Not applicable K/A Match:      This question matches the K/A by having the candidate demonstrate understanding of CNTMT spray system interlocks during a CNTMT high pressure situation.
Technical
 
==Reference:==
ES-1.3, rev 19 Proposed references            None to be provided:
Learning Objective:            OPT200CS #1 & 7 Question Source:
New Modified Bank Bank                  X Question History:              SQN Bank, 0109 ILT NRC Exam Comments:
Wednesday, June 05, 2013 8:16:16 AM                                                          69
 
1305 NRC RO Exam
: 27. W/EI5EK3.I 027 Given the following plant conditions:
            -  A large break LOCA has occurred on Unit 1.
            -  Accumulators have discharged and are isolated.
            -  ES-i .3, Transfer to Containment Sump, has been completed.
            -  Containment sump level is now at 84% and slowly rising.
            -  FR-Z.2, Containment Flooding, is in progress.
Which of the following describes; (1) where the FR-Z.2 required sample is taken, and (2) the reason for sampling the containment sump?
Ab    RHR system                  To determine the level of activity, to allow the TSC to determine if excess sump water can be transferred to tanks outside of containment.
B. Containment sump                To determine the level of activity, to allow the TSC to determine if excess sump water can be transferred to tanks outside of containment.
C.      RHR system                  To ensure shutdown margin is being maintained, since non-borated water has entered the containment sump.
D. Containment sump                To ensure shutdown margin is being maintained, since non-borated water has entered the containment sum p.
Wednesday, June 05, 2013 8:16:16 AM                                                        70
 
1305 NRC RO Exam Feedback DISTRACTOR ANAL YSIS:
A. Correct, Since sump swapover has occurred, FR-Z.2 directs obtaining a sample from the RHR system to aid in determining if activity levels will allow transferring water to locations outside containment, to alleviate containment flooding.
B. Incorrect, Plausible, since containment sump would be the correct sample point if sump swapover had NOT been completed. Further plausibility is added because the reason given is correct.
C. Incorrect, Plausible, since sampling the sump is an action directed by FR-Z2.
Also, per the above excerpt from the WOG Background Document, if the crew is in FR-Z2, then non-bora ted water has entered containment, and it is plausible that shutdown margin would be a concern.
: 0. Incorrect, Plausible, since containment sump would be the correct sample point if sump swapover had NOT been completed. If the crew is in FR-Z2, then non-bora ted water has entered containment, and it is plausible that shutdown margin would be a concern.
Wednesday, June 05, 2013 8:16:16 AM                                                              71
 
1305 NRC RO Exam Notes Question Number:            27 Tier:      1    Group      2 K/A:    W/E15 Containment Flooding EK3.1 Knowledge of the reasons for the following responses as they apply to the (Containment Flooding):
Facility operating characteristics during transient conditions, including coolant chemistry and the effects of temperature, pressure, and reactivity changes and operating limitations and reasons for these operating characteristics.
Importance Rating:        2.7 / 2.9 10 CFR Part 55:        41.5/41.10,45.6,45.13 1OCFR55.43.b:            Not applicable K/A Match:      Question meets the KA by examining the applicant on where the chemistry sample is drawn during transient conditions and asking the reason for sampling.
Technical
 
==Reference:==
FRZ.2, WOG Background Document, 2.
DESCRIPTION, Page 2:
FR-Z.2, step 2. rev 7 Proposed references            None to be provided:
Learning Objective:            0PL271 FRZ.2, rev 2 Obj. 4 Summarize the mitigating strategy for FR-Z.2 Question Source:
New Modified Bank Bank                    X Question History:              Sqn bank question used on the SQN 09/2010 exam.
Comments:
Wednesday, June 05, 2013 8:16:16 AM                                                        72
 
1305 NRC RO Exam
: 28. 003 K6.02 028 Given the following plant conditions:
            -  Unit 1 is at 100% rated thermal power.
            -  1 -FCV-62-93, Charging Flow Control Valve is in Manual.
            -  1 -FCV-62-89, Charging Seal Water Flow Control Valve, is operating at 60%
open.
            -  Due to a positioner failure, 1 -FCV-62-89 throttles close, and sticks at the 30% open position.
What effect will this malfunction have on charging pump discharge pressure and RCP seal injection flow?
Charging Pump                        RCP Seal Discharge Press                  Injection Flow A.            Lowers                          Rises B.            Rises                          Lowers Rises                          Rises D.            Lowers                          Lowers Wednesday, June 05, 2013 8:16:16 AM                                                        73
 
1305 NRC RD Exam Feedback DISTRACTOR ANAL YSIS:
A. Incorrect, Plausible if the candidate thinks that FCV-62-89 is located in the RCP seal supply line such that opening the valve would cause more flow and closing the valve would decrease flow, thus causing CCP discharge pressure to lower due to increased flow. Also plausible since the second part is correct.
B. Incorrect, Plausible since the first part is correct, the increased backpressure on the system would cause CCP discharge pressure to increase. Also the second part is plausible if the candidate thinks that FC V-62-89 is located in the RCP seal supply line such that opening the valve would increase flow and closing the valve would decrease flow.
C. Correct, the 1-FCV-62-89 valve is inline with the charging flow control valve and seal injection is supplied by a connection between the two valves. FCV-62-89 provides sufficient backpressure that RCP seals are supplied by charging. By increasing the position of FCV-62-89, less backpressure is provided, RCP seal injection will lower as charging flow increases and CCP discharge pressure would lower. Thus if FCV-62-89 closes down, it would provide more backpressure to the CCP, raising its discharge pressure and forcing more flow to the RCP seals.
D. Incorrect, Plausible if the candidate thinks that FCV-62-89 is located in the RCP seal supply line such that opening the valve would cause more flow and closing the valve would decrease flow, thus causing CCP discharge pressure to lower due to increased flow. Also the second part is plausible if the candidate thinks that FCV-62-89 is located in the RCP seal supply line such that opening the valve would increase flow and closing the valve would decrease flow.
Wednesday, June 05, 2013 8:16:16 AM                                                                74
 
1305 NRC RO Exam Notes Question Number:            28 Tier:      2    Group:      1 K/A:    003 Reactor Coolant Pump System (RCPS)
K6.02 Knowledge of the effect of a loss or malfunction on the following will have on the RCPS:
RCP seals and seal water supply Importance Rating:        2.7 / 3.1 1OCFRPart55:            41.7/45.5 1OCFR55.43.b:          Not applicable K/A Match:      K/A is matched because the question requires knowledge of the how an RCP seal supply malfunction will affect charging pump operation and seal injection flow to the RCPs Technical
 
==Reference:==
1 -47W809-1, R79 Proposed references            None to be provided:
Learning Objective:                OPT200.RCP
                                          #9 Given specific plant conditions, Analyze the effect that a loss or malfunction of the following will have on the RCP
: d. Seal injection supply Cognitive Level:
Higher                  X Lower Question Source:
New Modified Bank Bank                    X Question History:              SON Bank from DC Cook 2007 NRC exam Comments:
Wednesday, June 05, 2013 8:16:17 AM                                                                75
 
1305 NRC RO Exam
: 29. 004 A2.06 029 Given the following plant conditions:
            -    Unit 1 is operating at 100% power after restart following a refueling outage.
            -    Rod Control in MANUAL.
            -    VCT level is currently at 32%.
            -    An AUO places an un-borated mixed bed demineralizer in service.
Which ONE of the following completes the statements below?
Assuming NO operator action is taken, the VCT level over time will In accordance with AOP-C.02, Uncontrolled RCS Boron Concentration Changes, the first corrective action the RO will take that will stop the event in progress is to Lfl A. remain constant                place 1-HS-62-79A, Mixed Bed Hi Temp Bypass, to V.C. TK position B. remain constant                initiate normal boration C      rise                          place 1-HS-62-79A, Mixed Bed Hi Temp Bypass, to V.C. TK position D.      rise                          initiate normal boration Wednesday, June 05, 20138:16:17 AM                                                              76
 
1305 NRC RO Exam Feedback DISTRACTOR ANAL YSIS:
A. Incorrect, Plausible because, unlike an inadvertent dilution due to makeup, no inventory is directly added to the VCT due to the unborated mixed bed and placing 1-HS-62-79A to the VCT position will divert letdown around the demin that is the source of the problem.
B. Incorrect, Plausible because, unlike an inadvertent dilution due to makeup, no inventory is directly added to the VCT due to the unborated mixed bed and AOP-C.02, Uncontrolled RCS Boron Concentration Changes, does refer operators to borate the RCS, however it will not stop the event in progress.
C. Correct, the dilution event in progress does not add any inventory but it will increase Tavg which will cause pressurizer level to increase above setpoint which will lower charging and place more coolant into the VCT. Also, placing 1-HS-62-79A to the VCT position will divert letdown around the demin that is the source of the problem.
: 0. Incorrect, Plausible because the dilution event in progress does not add any inventory but it will increase Tavg which will cause pressurizer level to increase above setpoint, which will lower charging and place more coolant into the VCT.
Also plausible because AOP-C.02, Uncontrolled RCS Boron Concentration Changes, does refer operators to borate the RCS, however it will not stop the event in progress.
Notes Question Number:            29 Tier:      2    Group:      1 K/A:      004 Chemical and Volume Control System A2.06 Ability to (a) predict the impacts of the following malfunctions or operations on the CVCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:
Inadvertent boration/dilution.
Importance Rating:        4.2 / 4.3 10 CFR Part 55:          41.5/43.5/45.3/45.5 1OCFR55.43.b:            Not applicable K/A Match:      K/A is matched because the question requires knowledge of the operation of CVCS, i.e. how charging will respond as Tavg and pressurizer level change and that effect on VCT level and knowledge of how to stop an unborated mixed bed from diluting the RCS through the CVCS.
Wednesday, June 05, 2013 8:16:17 AM                                                              77
 
1305 NRC PC Exam Technical
 
==Reference:==
AOP-C.02, Uncontrolled RCS Boron Concentration Changes, Revision 0008 1 -SO-62-9 R44 Proposed references          None to be provided:
Learning Objective:          0PT271 AOP-C.02
: 3. Given a set of initial plant conditions, determine initial Operator response to stabilize the plant.
: 7. Given the procedure and a set of of initial plant conditions, determine actions required to mitigate the event in progress Cognitive Level:
Higher                X Lower Question Source:
New Modified Bank Bank                  X Question History:            WBN bank question 004 A2.06 130 used on the WBN 10/2011 NRC exam with minor changes to make applicable for use on SQN 05/2013 NRC exam Comments:
Wednesday, June 05, 2013 8:16:17 AM                                                        78
 
1305 NRC RO Exam
: 30. 004 A4.12 030 Given the following plant conditions:
              -  Unit 1 is at 100% rated thermal power.
              -  A routine dilution has just occurred.
              -  The integrater has counted out, however the OATC has NOT returned the CVCS Makeup Selector Switch to the AUTO position.
Which ONE of the following identifes the expected positions of the following valves?
Note:
1-FCV 140D, Boric Acid Valve to the Blender 1-FCV-62-128, Inletto the Top of the VCT 1 -FCV-62-1 40D            1 -FCV-62-1 28 A.          Closed                    Open B          Closed                  Closed C.          Open                    Closed D.          Open                    Open Feedback DIS TRACTOR ANAL YSIS:
A. Incorrect The 140D will be closed until the Makeup Selector Switch is returned to auto and is correct. The 128 valve will be closed at the end of the dilution, it is plausible to the examinee that it will be open as the 1400 valve will not return to normal position until after the makeup selector switch is returned to auto.
B. Correct: The 1400 valve is normally open and closes once the Makeup Selector Switch is placed in dilute or alt dilute. The 128 valve opens for a normal dilute and feeds to the top of the VC7 it will close as soon as the integrated completes the required makeup amount.
C. Incorrect The 1400 valve being open is plausible as it is normally open and will re-open once the Makeup Control Switch is returned to auto. The 128 as closed is correct.
: 0. lncorrecl The 1400 valve being open is plausible as it is normally open and will re-open once the Makeup Control Switch is returned to auto. The 128 valve will be closed at the end of the dilution, it is plausible to the examinee that it will be open as the 1400 valve will not return to normal position until after the makeup selector switch is returned to auto.
Wednesday, June 05, 2013 8:16:17 AM                                                                  79
 
1305 NRC AC Exam Notes Question Number:          30 Tier:    2    Group      1 K/A:      004 Chemical and Volume Control System A4.12 Ability to manually operate and/or monitor in the control room: Boration/dilution batch control Importance Rating:        3.8 / 3.3 10 CFR Part 55:          CFR: 41.7 / 45.5 to 45.8 1OCFR55.43.b:            Not applicable K/A Match:      Meets the K/A due to requireing the examinee to monitor the correct component positions during a dilution.
Technical
 
==Reference:==
0-SO-62-7, R66 1 -47W61 1-62-2 R5 Proposed references            None to be provided:
Learning Objective:            OPT200.CVCS
                                      #6 Explain the CVCS design features and/or interlocks that provide the following:
: j. RCS boron concentration control and modes of operation of the reactor makeup control system (blender controls)
Question Source:
New                    X Modified Bank Bank Question History:              New for SQN ILT 1305 NRC Exam Comments:                      Low Cognitive- memory Wednesday, June 05, 2013 8:16:17 AM                                                          80
 
1305 NRC RO Exam
: 31. 005 AI.03 031 Given the following plant conditions:
            -  Unit 1 is in Mode 5, midloop operation.
            -  RHR Train A in service at a flow rate of 2100 gpm.
            -  The RCS temperature is stable at 126&deg;F.
            -  The operator throttles open 1 -FCV-74-32, RHR HXS BYPASS.
Which ONE of the following completes the statement below?
As 1 -FCV-74-32 is throttled open, the RCS temperature will jJ and the RHR flow rate indicated on 1-M-6 will {
w A.      lower                      rise B.      lower                      lower C      rise                      rise D.      rise                    lower Wednesday, June 05, 20138:16:17 AM                                            81
 
1305 NRC RO Exam Feedback DIS TRACTOR ANAL YSIS:
A. Incorrect, Plausible because the typical thought process is when bypassing a heat exchanger less heat will be picked up allowing the temperature to drop and because of where the flow is measured the indica ted flow rate increasing is correct.
B. Incorrect, Plausible because the typical thought process is when bypassing a heat exchanger less heat will be picked up allowing the temperature to drop and because there is a flow element on the flow through the heat exchanger that would sense a lower flow but it is not the flow element that provides the indication on the control board.
C. Correct, with the valve being opened further (until it is stopped by a restricting device placed on the valve during midloop operations), more flow to bypass the RHR HX. Less flow through the heat exchanger results in less cooling allowing RCS temperature to increase. Since total flow is measured downstream of where the HX Bypass connects to the HX discharge line (less overall system resistance),
the flow indication will rise.
: 0. Plausible because the RCS temperature increasing is correct and because there is a flow element on the flow through the heat exchanger that would sense a lower flow but it is not the flow element that provides the indication on the control board.
Wednesday, June 05, 2013 8:16:17 AM                                                                82
 
1305 NRC RO Exam Notes Question Number:          31 Tier:    2    Group      1 K/A:      005 Residual Heat Removal System (RHRS)
Al .03 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the RHRS controls including: Closed cooling water flow rate and temperature Importance Rating:        2.5 I 2.6 10 CFR Part 55:          (CFR: 41.5/ 45.5) 1OCFR55.43.b:            Not applicable K/A Match:      Question matches KA by adjusting the bypass valve and having the exam inee determine its effects on flow within the RHR system and temperature of the RCS Technical
 
==Reference:==
1-47W810-1 R22 l-47W811-l R74 Proposed references            None to be provided:
Learning Objective:            OPT200.RHR
                                      #8 Explain the RHR system design features and/or interlocks that provide the following:
: a. RHR heat exchanger bypass flow control Question Source:
New Modified Bank Bank                  X Question History:              Original question from WBN bank Comments:                      High Cognitive Wednesday, June 05, 2013 8:16:17 AM                                                        83
 
1305 NRC RO Exam
: 32. 006 K5.05 032 Given the following plant conditions:
              -  Unit 2 is at 100% power
              -  An inadvertent Safety Injection Actuation has occurred.
Which ONE of the following identifies the adverse affect of allowing Safety Injection to continue without performing Safety Injection Termination?
A. ECCS Pumps will be running for extended time periods at minimum flow.
B. Loss of Instrument Air to Containment will not allow the use of the normal Pressurizer Spray Valves to control Pressurizer Pressure.
C Centrifugal Charging Pumps running in Injection Mode will collapse the Pressurizer bubble and pressurize the RCS to the PORV setpoint.
D. Reactor Coolant Pumps will be running without adequate pump seal cooling.
Feedback DIS TRACTOR ANAL YSIS:
A. Incorrect, Plausible as the SI and RHR pumps would be running on minimum flows and the examinee may give expected concern for this condition. However, in ES-i. 1 space this is not the reason for getting ECCS off B. Incorrect, Plausible because it could be thought that using PZR Spray Valves could prevent an over pressure condition, however, the PZR would continue to fill and pressurize the RCS until inventory was controlled. PZR Spray is mentioned in ES-i. 1 in restoring instrument air and in restoring RCPs.
C. Correct, The high head CCPs will continue to increase RCS inventory (shutoff head 2500 psig) resulting in high pressures up to the PORV setpoint if steps to reduce flow and restore letdown as part of SI termination are not performed.
: 0. Incorrect, Plausible because the RCPs would be running with seal injection but not normal seal return flow. Seal return would be through the seal return relief valve to the PRT and the examinee could conclude that seal cooling is not adequate for the RCPs without a complete understanding of CVCS/CCS conditions at this time. It is also a step in ES-i. 1 to address restoring normal seal return flow to the VCT.
Wednesday, June 05, 20138:16:17 AM                                                                84
 
1305 NRC RO Exam Notes Question Number:          32 Tier:      2    Group      1 K/A:      006 Emergency Core Cooling System (ECCS)
K5.05 Knowledge of the operational implications of the following concepts as they apply to ECCS: Effects of pressure on a solid system Importance Rating:        3.4 / 3.8 10 CFR Part 55:          (CFR: 41.5 / 45.7) 1OCFR55.43.b:            Not applicable K/A Match:      The question matches the KA in that it test the examinee on the effects of the CCPs continuing the run on RCS pressure control. The PZR will go solid and make RCS pressure control very difficult and in fact create a LOCA in the RCS.
Technical
 
==Reference:==
EPM-3-ES-1.1 R5, Basis Document for ES-1.1 ES-l.1 R12, SI Termination Proposed references            None to be provided:
Learning Objective:            0PL271 ES-i .1
                                      #4 Summarize the mitagating strategy for ES-i .1.
                                      #5 Describe the basis for all limits, notes, cautions and steps of ES-i .1.
Question Source:
New Modified Bank Bank                  X Question History:              CPNPP March 2010 NRC Written Exam Comments:                      High Cognitive Wednesday, June 05, 2013 8:16:17 AM                                                            85
 
1305 NRC RO Exam
: 33. 007 A3.0I 033 Which ONE of the following identifies the pressure that relief valve 63-637, RHR Pump Discharge, will start relieving and the tank where the flow through the valve will be routed?
Pressure                        Tank A. 550 psig                      RCDT B. 550 psig                        PRT C. 600 psig                      RCDT Dv    600 psig                        PRT Feedback DIS TRACTOR ANAL YSIS:
A. Incorrect, Plausible because 550 psig is the maximum RHR pump discharge pressure allowed to be maintained in accordance with the System Operating Instruction when RHR system is in service and the RCDT is an RCS tank inside containment (like the PRT) which does receive flow and leakoifs from RCS related components.
B. Incorrect, Plausible because 550 psig is the maximum RHR pump discharge pressure allowed to be maintained in accordance with the System Operating Instruction when RHR system is in service and the PRT being the tank that receives flow passing through the valve is correct.
C. Incorreci, Plausible because 600 psig is the pressure that the valve starts relieving and the RCDT is an RCS tank inside containment (like the PRT) which does receive flow and leakoffs from RCS related components.
: 0. CorrecI, the RHR discharge relief valve, 63-63, starts relieving at 600 psig and is routed to the PAT.
wednesday, June 05, 2013 8:16:17 AM                                                                86
 
1305 NRC RO Exam Notes Question Number:            33 Tier:    2    Group        1 K/A:    007 Pressurizer Relief Tank/Quench Tank System (PRTS)
A3.01 Ability to monitor automatic operation of the PRTS, including:
Components which discharge to the PRT Importance Rating:        2.7 /2.9 10 CFR Part 55:          41.7 I 45.5 to 45.8 1OCFR55.43.b:            Not applicable K/A Match:      K/A is matched because the question requires the ability to monitor the automatic operation of the RHR discharge pressure and PRT level to know if the relief valve is relieving (which did occur at SQN allowing approximately 10,000 gallons to be passed to the PRT prior to termination)
Technical
 
==Reference:==
1-47W811-1 R74 0-47W813-1 R55 1 -SO-74-1, Residual Heat Removal System, Revision 0086 Proposed references            None to be provided:
Learning Objective:            OPT200.RHR
: 18. STATE the RHR design pressure and flow capacities.
OPT200.PRT
: 19. LIST the components that discharge to the PRT Cognitive Level:
Higher Lower                  X Question Source:
New Modified Bank Bank                    X Question History:              WBN bank question 007A3.01 33 used on the WBN 06/2011 NRC exam.
Comments:
Wednesday, June 05, 2013 8:16:17 AM                                                          87
 
1305 NRC RO Exam
-  34. 007 G2.1.20 034 Given the following plant conditions:
              -  Unit 2 is at 90% power.
              -  Due to a leaking PORV, 2-FCV-68-332 PORV block valve was closed.
              -  PRT level is 80%
              -  PRT pressure is 7 psig
              -  PRT Temperature is 145&deg;F Which ONE of the following describes the action to be taken in 2-SO-68-5, Pressurizer Relief Tank, to return the PRT to normal?
A. Start a Waste Gas Compressor, open 2-PCV-68-301, PRT VENT TO WDS VENT HDR, and reduce PRT pressure to < 4 psig.
B. Align the B RCDT pump, open FCV-68-305, N2 SUPPLY TO PRT, open 2-LCV-68-310, PRT DRAIN TO RCDT, and lower PRT level to 60%.
C. Open 2-FCV-68-303, PRIMARY WATER TO PRT, and return PAT level to 88%.
Dv Open 2-FCV-68-303, PRIMARY WATER TO PRT, and reduce PRT temperature to < 120&deg;F.
Feedback DIS TRACTOR ANAL YSIS:
A. lncorrect Plausible as pressure is high and this is the correct action to take to reduce PRT pressure in 2-SO-68-5.
B. Incorrect, Plausible as level is higher than normal (although below alarm set point of 88%) this is the correct action to take to reduce PRT level in 2-SO-68-5.
C. Incorrect, Plausible as one of the stopping criteria in 2-S0-68-5 for lowering PRT temperature is to feed primary water to lower temperature to < 120 &deg;F or until level is 88%. When level reaches 88%, the PRT is drained and this process is repeated.
D. Correct, The temperature in the PRT exceeds the alarm set point and this will be addressed first per the AR. Per 2-S0-68-2 the correct action to take for a high temperature condition is to Open 2-FC V-68-303 to feed primary water to the PRT.
Wednesday, June 05, 2013 8:16:17 AM                                                              88
 
1305 NRC RO Exam Notes Question Number:          34 Tier:    2    Group      1 K/A:      007 Pressurizer Relief Tank/Quench Tank System (PRTS)
G2.1 .20 Ability to interpret and execute procedure steps.
Importance Rating:        4.6/4.6 10 CFR Part 55:          (CFR: 41.10 /43.5 /45.12) 1OCFR55.43.b:            Not applicable K/A Match:      This question matches the K/A by testing the candidates knowledge of the procedures associated with PZR PRT level, pressure and temperature.
Technical
 
==Reference:==
2-SO-68-5 Ri 7 2-AR-M5-A, C-i P25 Proposed references            None to be provided:
Learning Objective:            OPT200. PZR-PRT, 8f, Given specific plant conditions, analyze the effect that a loss or malfunction of the PZR pressure control system will have on tne PRT.
Question Source:
New Modified Bank Bank                  X Question History:              SQN Bank, SQN ILT 1211 Audit Exam Comments:                      High Cognitive Wednesday, June 05, 2013 8:16:17AM                                                                    89
 
1305 NRC RO Exam
: 35. 008 A3.01 035 Given the following plant conditions:
              -  Unit 1 is operating at 100% power.
              -  1-RA-90-123A CCS LIQ EFF MON HIGH RAD Alarm is LIT.
              -  The red HIGH light is LIT on CCS LIQUID EFFLUENT RADMON 0-RM-90-1 23A.
              -  CCS surge tank level was increasing but is now stable.
              -  RC PUMPS THRM BARRIER RETURN HEADER FLOW LOW Alarm is LIT.
Which ONE of the following completes the statement below concerning the automatic actions that are designed to occur?
The thermal barrier heat exchanger containment isolation inlet and outlet valves to close and the thermal barrier booster pumps A. (1) the affected RCP ONLY (2) trip B. (1)    the affected RCP ONLY (2)  continue to run with minif low valves open.
C (1)    all four (4) RCPs (2)  trip.
D. (1)    all four (4) RCPs (2)  continue to run with minif low valves open.
Wednesday, June 05, 2013 8:16:17 AM                                                        90
 
1305 NRC RO Exam Feedback DIS TRACTOR ANAL YSIS:
A. Incorrect: The first part is incorrect. RCP thermal barriers do not have pump specific containment isolation valves (CIVs). The four valves which auto close isolates CCS to all RCPs. This is plausible since some components can be isolated individually. The second part is correct. The Thermal Barrier Booster pumps will trip with a differential flow across the thermal barrier since this is an indication of a thermal barrier leak.
B. Incorrect: The first part is incorrect. RCP thermal barriers do not have pump specific ClVs. The four valves which auto close isolates CCS to all RCPs. This is plausible since some components can be isolated individually. The second part is incorrect. The Thermal Barrier Booster pumps do not have miniflow valves to ensure pump cooling. This is plausible since other pumps do have this configuration.
C. Correct: The first part is correct. The four valves which auto close isolates CCS to all RCPs. The second part is correct. The Thermal Barrier Booster pumps will trip with a differential flow across the thermal barrier since this is an indication of a thermal barrier leak.
D. Incorrect: The first part is correct. The four valves which auto close isolates CCS to all RCPs. The second part is incorrect. The Thermal Barrier Booster pumps do not have miniflow valves to ensure pump cooling. This is plausible since other pumps do have this configuration.
Wednesday, June 05, 20138:16:17 AM                                                                91
 
1305 NRC RO Exam Notes Question Number:          35 Tier:    2      Group      1 K/A:    008 Component Cooling Water System (COWS)
A3.01 Ability to monitor automatic operation of the COWS, including:
Setpoints on instrument signal levels for normal operations, warnings, and trips that are applicable to the COWS.
Importance Rating:        3.2* / 3.0 10 CFR Part 55:          41.7 1OCFR55.43.b:            Not applicable K/A Match:      This question matches the K/A by having the candidate monitor the automatic operation of COWS components when an alarm setpoint is reached which causes components of the COWS to operate.
Technical
 
==Reference:==
0-AR-M12-A (B-i) rev 52 0-AR-M27-B-A (B-i) rev 12 47W610-90-2 rev 78 47W 61 0-70-3 rev 22 Proposed references            None to be provided:
Learning Objective:            OPT200.OOS Rev 9 Obj 11 .d Given specific plant conditions, ANALYZE the effect that a loss of malfunction of the following will have on the CCS: Heat Exchanger leaks.
Question Source:
New Modified Bank Bank                  X Question History:              SQN Bank Comments:
Wednesday, June 05, 2013 8:16:17 AM                                                            92
 
1305 NRC RO Exam
: 36. 010 K6.01 036 Given the following plant conditions:
              -    Unit 1 is operating at 100% RTP.
              -    Pressure Control Channel Selector Switch, 1 -XS-68-340D is selected to PT-68-340 & 334 position.
              -    1-PT-68-334, Pressurizer Pressure Transmitter, fails LOW.
              -    The operating crew enters AOP-l.04, Pressurizer Instrument and Control Malfunctions.
Which ONE of the following completes the statements below?
The failure    JJL      result in the pressurizer back-up heaters being energized.
When AOP-l.04 performance is completed L of the PORVs will be able to automatically open if pressurizer pressure begins rising.
Lil A.      will                    both B.      will                  only one C will NOT                      both D. will NOT                    only one Wednesday, June 05, 2013 8:16:17 AM                                                            93
 
1305 NRC RO Exam Feedback DIS TRACTOR ANAL YSIS:
A. Incorrect, Plausible because if the failure had been on 1-PT-68-340 the pressuizer backup heaters would have been auatomatically energized and both of the PORVS being able to automatically opening on highe pressure is correct.
B. Incorrect, Plausible because if the failure had been on 1-PT-68-340 the pressuizer backup heaters would have been auatomatically energized and because if the failure had been on 1-PT-68-323 or 1-PT-68-322, then one of the PORVs would have been made incapable of automatically opening if the pressure rose to the setpoint..
C. Correct, with the selector switch in the PT-68-340 & 334 position, the failure will not affect the pressurizer backup heaters but will result in a loss of one of the 2 required channels to allow PORV 334 to open in automatic. During performance of AOP-I.04 the selector switch XS-68-340D will be repositioned to PT-68-340 & 322 which restores the lost input for PORV 334 operation.
D. Incorrect, Plausible because the operation of the pressuizer backup heaters not being affected is correct but they would have been if the failure had been on 1-PT-68-340. Also plausible because if the failure had been on 1-PT-68-323 or 1-PT-68-322, then one of the PORVs would have been made incapable of automatically opening if the pressure rose to the setpoint.
Wednesday, June 05, 2013 8:16:17 AM                                                                  94
 
1305 NRC RO Exam Notes Question Number:            36 Tier:      2    Group        1 K/A:      010 Pressurizer Pressure Control System (PZR PCS)
K6.01 Knowledge of the effect of a loss or malfunction of the following will have on the PZR PCS:
Pressure detection systems Importance Rating:        2.7 / 3.1 1OCFRPart55:            41.7/45.7 1OCFR55.43.b:            Not applicable K/A Match:      Qustion requires knoowledge of how a malfunction of a pressure detection system transmitter will affect compontents in the Pressure Pressure Control System and how the required procedure prosponse will mitigate the consequence to the malfucntion.
Technical
 
==Reference:==
AOP-l.04, Pressurizer Instrument and Control Malfunctoins, Revision 12 Proposed references            None to be provided:
Learning Objective:            OPL271AOP-l.04
                                      #5 Summarize AOP-I.04 mitigating strategy for each operator section.
                                      #10 Identify automatic actions associated with dropping/rising RCS pressure.
Question Source:
New                    X Modified Bank Bank Question History:              New question for the SQN 05/2013 exam Comments:
Wednesday, June 05, 2013 8:16:17 AM                                                          95
 
1305 NRC RO Exam
: 37. 010 K6.03 037 Given the following plant conditions:
            -  Unit 2 was operating at 100% power.
            -  The Loop 1 pressurizer spray valve controller failed causing the spray valve to fully open.
Assuming No Operator Actions are taken, which ONE of the following identifies the response of the pressurizer pressure control system?
A. Master controller output would INCREASE.
PZR pressure would be maintained above the Reactor Trip setpoint.
B. Master controller output would INCREASE.
PZR pressure would decrease to the Reactor Trip setpoint.
C. Master controller output would DECREASE.
PZR pressure would be maintained above the Reactor Trip setpoint.
D Master controller output would DECREASE.
PZR pressure would decrease to the Reactor Trip setpoint.
Wednesday, June 05, 2013 8:16:17 AM                                                      96
 
1305 NRC RO Exam Feedback DIS TRACTOR ANAL YSIS:
A. Incorrect The output of the master controller increases as pressure goes high, not as pressure drops below setpoint. to turn on heaters not increasing, but with the spray valve fully open, the heaters would not be able to terminate the pressure drop, and a reactor trip on low pressurizer pressure would occur. Plausible if the applicant knowing that the heaters should be turned on but confuses the direction of the change in the output of the master controller or believes the heaters coming on would prevent the pressure from continuing to drop to the reactor trip setpoint.
B. Incorrect, The output of the master controller increases as pressure goes high, not as pressure drops below setpoint. to turn on heaters not increasing, but with the spray valve fully open, the heaters would not be able to termThate the pressure drop, and a reactor trip on low pressurizer pressure would occur. Plausible if the applicant knowmg that the heaters should be turned on but confuses the direction of the change in the output of the master controller and knows that the heaters commg on would not prevent the pressure from contmumg to drop to the reactor trip setpoint.
C. Incorrec1, The output of the master controller does decrease as the pressure drops below setpoint to turn on heaters, but with the spray valve fully open, the heaters would not be able to terminate the pressure drop, and a reactor trip on low pressurizer pressure would occur. Plausible if the applicant knowing that the heaters should be turned on and which direction the output of the master controller would change, but believes the heaters coming on would prevent the pressure from contThuThg to drop to the reactor trip setpoint.
D. Correct, The pressurizer pressure will be droppmg due to the spray valve being open. As the lower pressure is compared to the setpoTht pressure, the output of the master controller will start droppmg to turn on heaters. With the spray valve fully open, the heaters would not be able to termThate the pressure drop, and a reactor trip on low pressurizer pressure would occur.
Wednesday, June 05, 20138:16:17 AM                                                              97
 
1305 NRC RO Exam Notes Question Number:          37 Tier:    2    Group      1 K/A:    010 Pressurizer Pressure Control System K6.03 Knowledge of the effect of a loss or malfunction of the following will have on the PZR PCS: PZR sprays and heaters Importance Rating:        3.2 / 3.6 1OCFR Part 55:          41.7/45.7 1OCFR55.43.b:          Not applicable K/A Match:      Question requires the applicant to understand how a pressurizer spray valve failing open will affect the pressurizer control system and whether the heaters in the control system are designed to prevent a reactor trip due to the valve failure.
Technical
 
==Reference:==
1 -,2-47W61 1-68-3 AOP-L04 R12 Proposed references          None to be provided:
Learning Objective:          OPT200. PZR-PRT
: 9. Given specific plant conditions, ANALYZE the effect that a loss or malfunction of the following will have on the Pressurizer Level or Pressure Control System:
: a. PZR sprays and heaters.
Question Source:
New Modified Bank Bank                    X Question History:            SQN bank question 010 K6.03 037 used on the SQN 1/2008 exam Comments:
Wednesday, June 05, 20138:16:17 AM                                                                98
 
1305 NRC RO Exam
: 38. 012 K2.01 038 Which ONE of the following identifies the plant electrical boards that supply power to the listed components on Unit 1?
SSPS Train B Reactor                            Reactor Trip Bypass Breaker A TriD Breaker 48v UV coil                          (BYA Control Power Circuit A 120v AC Vital Instrument                            125V DC Vital Battery Board I Power Boards II and IV B. 120v AC Vital Instrument                            125V DC Vital Battery Board II Power Boards II and IV C. 120v AC Vital Instrument                            125V DC Vital Battery Board I Power Board II ONLY D. 120v AC Vital Instrument                          125V DC Vital Battery Board II Power Board II ONLY Feedback DIS TRACTOR ANAL YSIS:
A. Correct 120v AC Vital Instrument Power Boards II and IV supply the 4.9v Reactor Trip Undervoltage relay through an auctioneered circuit and the 125V DC Battery Board II is the control power to BYA.
B. lncorrect, Plausible because the 120v AC Vital Instrument Power Boards II and IV supplying the 48v reactor Trip Undervoltage relay through an auctioneered circuit is correct and the 125V DC Battery Board Ills the control circuit power supply Train B reactor trip breakers and BYA receives trip signal from Train B circuits.
C. Incorrect, Plausthie because the 120v AC Vital Instrument Power Boards Ills the only power supply to other components in SSPS Train B (e.g. Slave relays) and the 125V DC Battery Board I is the control power supply to BYA.
D. lncorrect Pta usthie because the 120v AC Vital Instrument Power Boards II is the only power supply to other components in SSPS Train B (e.g. Slave relays) and the 125V DC Battery Board Ills the control circuit power supply Train B reactor trip breakers and BYA receives trip signals from Train B SSPS Reactor Trip circuits.
Wednesday, June 05, 2013 8:16:17 AM                                                              99
 
1305 NRC RO Exam Notes Question Number:          38 Tier:      2    Group        1 K/A:    012 Reactor Protection System K2.01 Knowledge of bus power supplies to the following:
RPS channels, components, and interconnections.
Importance Rating:        3.3 / 3.7 10 CFR Part 55:          41.7 1OCFR55.43.b:            Not applicable K/A Match:      K/A is matched because the question requires the knowledge of the bus power supplies to Reactor Protection System components.
Technical
 
==Reference:==
1 ,2-45W699-1 RiO Proposed references            None to be provided:
Learning Objective:            OPT200.RPS
: 5. LIST the bus power supplies to the following Reactor Protection System components:
: a. RPS channels
: b. Reactor Trip Breaker Control Power Cognitive Level:
Higher Lower                  X Question Source:
New Modified Bank Bank                    X Question History:              WBN bank question used on the WBN 10/2011 NRC exam.
Comments:
Wednesday, June 05, 2013 8:16:17 AM                                                          100
 
1305 NRC RO Exam
: 39. 013 K4.07 039 Given the following plant conditions:
              -  Unit 1 is operating at 100% power.
              -  The operating crew is responding to a loss of 1 20V AC Vital Instrument Power Board 1-I.
              -  PZR pressure transmitter 1-PT-68-334 (Channel II) fails LOW.
Which ONE of the following identifies how SSPS and ECCS will respond?
A. Both trains of SSPS SI master relays will actuate AND both trains of ECCS equipment auto start.
B Both trains of SSPS SI master relays will actuate BUT only B train ECCS equipment auto starts.
C. Only the B train SSPS SI master relays will actuate BUT both trains of ECCS equipment auto start.
D. Only the B train SSPS SI master relays will actuate AND only B train ECCS equipment auto starts.
Feedback DIS TRACTOR ANAL YSIS:
A. Incorrect, Master Relays on both trains will have power. Train A from Channel Ill via an auctioneering circuit, however, with the 1-I AC vital Instrument Power Board deenergized (Channel 1), the slave relays that control the Train A equipment will not have a power supply. Plausible if the candidate mistakes the source of the power supply or thinks that the circuit that auctioneers power in the logic cabinet provides power to the slave relays.
B. Correct, Master Relays on both trains will have power. Train A from Channel III via an auctioneering circuit, however, with the 1-I AC vital Instrument Power Board deenergized, the slave relays that control the Train A equipment will not have power.
C. Incorrect, Master Relays on both trains will have power. Train A from Channel III via the auctioneering circuit, however, Channel 1 is the only power supply for the slave relays that control the Train A equipment. Plausible if the candidate mistakes the source of the power supply or thinks that the circuit that auctioneers power in the logic cabinet provides power to the slave relays instead of the master relays.
D. Incorrect, Master Relays on both trains will have power. Train A from Channel III via an auctioneering circuit, however, Channel 1 is the only power supply for the slave relays that control the Train A equipment. Plausible if the candidate mistakes the function of the circuit that auctioneers power in the logic cabinet.
Wednesday, June 05, 2013 8:16:18 AM                                                              101
 
1305 NRC RO Exam Notes Question Number:          39 Tier:      2    Group        1 K/A:      013 Engineered Safety Features Actuation System K4.07 Knowledge of ESFAS design feature(s) and/or interlock(s) which provide for the following: Power supply loss Importance Rating:        3.7/4.1 10 CFR Part 55:          41.7 1OCFR55.43.b:            Not applicable K/A Match:      Thsi question matches the KA by testing the candidates knowledge of power supplies to portions of the ESFAS and how a loss of power will effect the systems ability to fully, or partially, actuate trains of ECCS components.
Technical
 
==Reference:==
47W61 1-63-1, R4 AOP-P.03, Loss of Unit 1 Vital Instrument Power Board, R25 Proposed references            None to be provided:
Learning Objective:            OPT200.RPS B.4 & 5 Question Source:                    SQN Bank RPS-B.9.A 002 New Modified Bank Bank                  X Question History:              SQN ILT 1002 exam, SQN ILT 1305 NRC Exam Comments:                      High Cognitive Wednesday, June 05, 20138:16:18 AM                                                              102
 
1305 NRC RO Exam
: 40. 022 K4.03 040 Which ONE of the following Containment Cooling System fans will trip and isolate as a DIRECT result of a Containment Isolation Phase-A Signal?
A. Lower Compartment Coolers B. Upper Compartment Coolers C Incore Instrument Room Coolers D. Control Rod Drive Motor Coolers Feedback All 4 of the choices are containment coolers that are tripped by one of the containment isolation signals.
DIS TRACTOR ANAL YSIS:
A.      Incorrect, These coolers are tripped for a phase B signal, the examinee could mistake these as being tripped by a Phase A isolation signal. Plausible because the Lower Compartment Cooler fans do get a signal to trip from a containment isolation signaI (Phase B, not Phase A)
B.      Incorrect, These coolers are tripped for a phase B signaI the examinee could mistake these as being tripped by a Phase A isolation signal. Plausible because the Upper Compartment Cooler fans do get a signal to trip from a containment isolation signal, (Phase B, not Phase A)
C. CQRRECT The Incore Instrument Coolers are tripped by a Phase A isolation signal.
: 0.      Incorrect, These coolers are tripped for a phase B signa) the examinee could mistake these as being tripped by a Phase A isolation signal. Plausible because the CRDM Cooler fans do get a signal to trip from a containment isolation signal, (Phase B, not Phase A)
Wednesday, June 05, 2013 8:16:18 AM                                                103
 
1305 NRC RO Exam Notes Question Number:            40 Tier:      2    Group        1 K/A:      022 Containment Cooling System K4.03 Knowledge of CCS design feature(s) and/or interlock(s) which provide for the following:
Automatic containment isolation Importance Rating:        3.6 / 4.0 10 CFR Part 55:          41 .7 1OCFR55.43.b:            Not applicable K/A Match:      Question requires the knowldege of the desing feature that results in the tripping of the containment coolers on a Phase A containment isolation.
Technical
 
==Reference:==
1 ,2-47W61 1-30-2 R2 1 ,2-47W61 1-30-3 R6 1 ,2-47W61 1-30-4 Ri 8 Proposed references            None to be provided:
Learning Objective:            OPT200.CNTMCLG&PURGE
: 8. EXPLAIN the Containment Cooling and Purge Systems design features and/or interlocks that provide the following:
: a. Automatic containment isolation Question Source:
New Modified Bank Bank                    X Question History:              SQN Bank question 022 K4.03 used on the SQN 1/2009 exam with correct answer location changed and minor wording change for use on the SQN 05/2013 exam Comments:
Wednesday, June 05, 2013 8:16:18 AM                                                          104
 
1305 NRC RO Exam
: 41. 025 A4.02 041 Given the following plant conditions:
              -  Unit 1 at 100% RTP when a LOCA occurs.
              -  A Safety Injection occurs due to containment pressure rising.
              -  The containment pressure has continued to rise and is now 3.2 psig.
Which ONE of the following completes the statements below?
The Containment Air Return Fans would automatically start 10 minutes after the.jJ signal.
If the 1 A-A Air Return Fan tripped on excessive current when the start was attempted, the            indicating lights on the MCR handswitch would be LIT?
Ui A.      Safety Injection            GREEN and WHITE only B.      Safety Injection            GREEN, WHITE and RED C.      Phase B Isolation            GREEN and WHITE, only D      Phase B Isolation            GREEN, WHITE, and RED Wednesday, June 05, 20138:16:18 AM                                                      105
 
1305 NRC RO Exam Feedback Dwg 1,2-45N779-5 shows schematic for the lA-A Air Return Fan and shows the auto start 10 mm after a phase B w/o BO. Using a combination of 779-5 and 779-1 (detail Al) shows that component overcurrent protection is from an amptector. When the amptector actuates it trips open the breaker and toggles the OTS contacts shown on 779-1. The breaker 52b contact at conection 8 closes to turn on the green lite, the OTS contact at conection 9 closes to turn on the red lite. The OTS closes at connection 11 to pick up the 30X relay which closes a contact in the white lite circuit.
All three lites will be LIT.
DISTRACTOR ANAL YSIS:
A. Incorrect, The Containment Air Return Fans receive a start signal 10 minutes after the Phase B isolation signal (2.8 psig) is initiated (not 10 minutes after the SI initiation) and all 3 lights on the handswitch will be lit (not just the white.) Plausible because the applicant could misapply the timer start time to the Hi containment pressure SI signal instead of the HI-HI (Phase B signal) and following the trip of many motors the green and white lights are the only lights LIT B. Incorrect, The Containment Air Return Fans receive a start signal 10 minutes after the Phase B isolation signal (2.8 psig) is initiated (not 10 minutes after the SI initiation) but all 3 lights on the handswitch will be lit. Plausible because the applicant could misapply the timer start time to the Hi containment pressure SI signal instead of the HI-HI (Phase B signal) and all 3 indicating lights being lit is correct.
C. Incorrect The Containment Air Return Fans do receive a start signal 10 minutes after the Phase B isolation signal (2.8 psig) is initiated, however all 3 lights on the handswitch will be lit, (not just the white.) Plausible because the start signal initiation is correct and and following the trip of many motors the green and white lights are the only lights LIT D. CorrecI, The Containment Air Return Fans are to maintain air flow through the ice condenser by pulling air from upper containment and discharge into lower containment during the accident. The fans receive a start signal 10 minutes after the Phase B isolation signal (2.8 psig) is initiated. Overload protection for the fan motor is provided by an amptector device and its oepration results in all 3 lights on the handswitch being lit.
Wednesday, June 05, 2013 8:16:18 AM                                                                    106
 
1305 NRC RO Exam Notes Question Number:            41 Tier:      2    Group        1 K/A:      025 Ice Condenser System A4.02 Ability to manually operate and/or monitor in the control room:
Containment vent fans Importance Rating:        2.7 / 2.5 10 CFR Part 55:          41.7 / 45.5 to 45.8 1OCFR55.43.b:            Not applicable K/A Match:      Question matches the K/A by requiring the applicant to be able to identify when the status of the Containment Air Return fans will change during a LOCA and be able identify a tripped condition of one of the fans after the automatic start is attempted.
Technical
 
==Reference:==
1 ,2-45N7791 R5 1 ,2-45N779-5 Ri 9 1 ,2-47W61 1-30-3 R6 Proposed references            None to be provided:
Learning Objective:            OPT200.ICE
: 6. EXPLAIN, (or SKETCH as applicable), the physical connections and/or cause-effect relationships between the Ice Condenser System and the following systems:
: d. Containment ventilation Question Source:
New Modified Bank            X Bank Question History:              SQN bank question 025 A4.02 040 used on the SQN 1/2009 retake exam with the format changed along with 2 distractors and the stem modified.
Comments:
Wednesday, June 05, 2013 8:16:18 AM                                                        107
 
1305 NRC RO Exam
: 42. 026 A1.02 042 Given the following plant conditions:
              -  Unit 1 was operating at 100% power when a design basis LOCA occurred inside containment.
              -  The operating crew is performing E-1, Loss of Reactor or Secondary Coolant.
Which ONE of the following completes the statement below?
A design basis of the containment spray system is to ensure the containment design temperature of          (1)  is not exceeded.
After an automatic initiation of the Containment Spray System, E-1 first allows the containment spray pumps to be stopped and placed in A-AUTO after the containment pressure drops to less than below                (2)
A. 125&deg;F                  2.8 psig B. 125&deg;F                  2.0 psig C. 250&deg;F                  2.8 psig D 250&deg;F                  2.0 psig Feedback DIS TRACTOR ANAL YSIS:
A. Incorrect, The first part is plausible as 125 &deg;F is the is the CNTMT lower compartment TS temperature limit. The second part is plausible as it would be logical to conclude that getting below the phase B setpoint of 2.8 psig would be the time to secure CNTMT spray.
B. lncorrect The first part is plausible as 125 &deg;F is the is the CNTMT lower compartment TS temperature limit. The second part is correct.
C. Incorrect, The first part is correct The second part is plausible as it would be logical to conclude that getting below the phase B setpoint of 2.8 psig would be the time to secure CNTMT spray.
D. Correcl, The design basis temperature for CNTMT is 250 &deg;F and per E- 1, CNTMT spray pumps are secured when CNTMT pressure is <2.0 psig.
Wednesday, June 05, 2013 8:16:18 AM                                                            108
 
1305 NRC RO Exam Notes Question Number:            42 Tier:      2    Group        1 K/A:    026 Containment Spray System (CSS)
Al .02 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the CSS controls including:
Containment temperature Importance Rating:        3.6* / 3*9 10 CFR Part 55:          41.5 / 45.5 1OCFR55.43.b:            Not applicable K/A Match:      This question matches the KA by examining the applicants knowlegde of CNTMT temperature limits and criteria to secure CNTMT spray.
Technical
 
==Reference:==
FSAR E-l, Loss of Reactor or Secondary Coolant P25 Proposed references            None to be provided:
Learning Objective:            OPL271E-l
: 4. Summarize the mitigating strategy or E-l.
OPT200.CNTMTSTRUCTU RE Sd. Explain the CNTMT structure design features and/or operational interlocks that provide:
Tempurature and pressure control during normal and during DBA conditions.
Question Source:
New Modified Bank Bank                  X Question History:              SQN Bank Comments:
Wednesday, June 05, 2013 8:16:18 AM                                                          109
 
1305 NRC RO Exam
: 43. 026 K1.01 043 Which ONE of the following identifies an electrical interlock associated with opening 1-FCV-72-40, RHR Spray Header A Isolation?
1-FCV-72-40 cannot be opened unless A. 1-FCV-74-33, RHR Crosstie, is fully open B. 1-FCV-63-1, RWSTto RHR Suction, is fully open C. 1-FCV-74-3, RHR Pump lA-A Suction, is fully closed D 1-FCV-63-72, Containment Sump to RHR Pump lA-A Isolation, is fully open Feedback DIS TRACTOR ANAL YSIS:
A. Incorrect, Plausible because 1-FCV-74-33 is closed when prior to opening the RHR spray valve 1-FCV-72-40.
B. Incorrect, Plausible because 1-FCV-63-1 is normally open to supply all ECCS pump suction when the RWST is being used prior to swapover to the containment sump.
C. Incorrect, Plausible because 1-FCV-74-3 is the suction valve to the RHR pump and normally the suction valve would have to be open to supply suction to the pump.
: 0. Correct as shown on i-47W61 1-72-1, there is an electrical interlock that requires 1-FCV-63-72 to be fully open to allow 1-FCV-72-40 to be opened.
Wednesday, June 05, 2013 8:16:18 AM                                                            110
 
1305 NRC RO Exam Notes Question Number:          43 Tier:      2    Group      1 K/A:      026 Containment Spray:
K1 .01 Knowledge of the physical connections and/or cause effect relationships between the CSS and the following systems:
ECCS Importance Rating:        4.2 / 4.2 10 CFR Part 55:          41.2 to 41.9/ 45.7 to 45.8 1OCFR55.43.b:            Not applicable K/A Match:      K/a is matched because the question requires applicant to recall interlock (cause and effect relationship) between a component in a containment spray subsystem and a component in the ECCS system.
Technical
 
==Reference:==
1-47W611-72-1 R13 Proposed references            None to be provided:
Learning Objective:            6. EXPLAIN, (or SKETCH as applicable), the physical connections and/or cause-effect relationships between the Containment Spray system and the following systems:
: a. RHR system
: b. Containment Spray heat exchangers cooling water
: c. Containment spray system fill and makeup water
: d. Safety Injection
: e. CVCS Question Source:
New Modified Bank Bank                    X Question History:              WBN bank question used on the WBN 2006 exam.
Comments:
Wednesday, June 05, 2013 8:16:18 AM                                                          111
 
1305 NRC RO Exam
: 44. 039 K3.06 044 Given the following:
            -    Unit 1 reactor power is stable at 90%.
            -    Turbine Impulse pressure transmitter 1-PT-i -72 fails LOW.
Which ONE of the following identifes the status of the white lights on 1-M-4 for (1) 1-XI-1-103D, STEAM DUMPS ACTUATED D FSVS ENERGIZED and (2) 1-XI-1-1O3AIB, STM DUMPS ARMED?
1-Xl-1-103D                    1-Xl-1-103A!B A.        DARK                            DARK B        DARK                            LIT C.        LIT                            DARK D.        LIT                            LIT Feedback DISTRACTOR ANAL YSIS:
A. lncorrect, Plausible because 1-XI-103D being dark is correct and 1-XI-1-103A/B would be dark if the instrument had failed high.
B. CorrecI, 1-PT- 1-72 is the pressure transmitter that is used to sense a loss of load signal to arm the steam dumps system. When the transmitter fails low the steam dumps will arm resulting in 1 -Xl- 1-1 03A/B being energized but 1 -Xl- 1 03D will remain dark because there would not be a larger difference in temperature between Tavg and Tref created by the transmitter failure.
C. Incorrect, Plausible because 1-XI-103D would be lit and 1-Xl-1-103A/B would be dark if the failed transmitter had been the other Impulse Pressure Transmitter 1-PT-1-73.
D. lncorrect Plausible because both 1-XI-103D and 1-XI-1-103A/B would be lit during an actual rapid load reduction and the applicant could determine both are supplied form the same transmitter.
Notes Wednesday, June 05, 2013 8:16:18 AM                                                                112
 
1305 NRC RO Exam Question Number:          44 Tier:    2    Group        1 K/A:      039 Main and Reheat Steam System (MRSS)
K3.06 Knowledge of the effect that a loss or malfunction of the MRSS will have on the following:
SDS Importance Rating:        2.8*/3.1 1OCFRPart55:            41.7/45.6 1OCFR55.43b:            Not applicable K/A Match:      K/A is matched because the question requires the knowledge of how the steam dumps system will be affected by a malfunction of a pressure transmitter on the Main and Reheat Steam System.
Technical
 
==Reference:==
1-47W611-1-2 R13 l,2-45N601-1 R3 1,2-45N601-1 R23 1-47W610-i-3 R5 0-50-1 -2, Steam Dump System, R12 OPT200SDCS, Steam Dump Lesson Plan Proposed references            None to be provided:
Learning Objective:            OPT200SDCS
                                      #8 Given specific plant conditions, analyze the effect that a loss or malfunction of the following will have on the SDCS:
PT-i -72/73 Cognitive Level:
Higher                  X Lower Question Source:
New Modified Bank Bank                    X Question History:              WBN Bank question 039K3.06 045 used on the 06/2011 WBN NRC exam Comments:
Wednesday, June 05, 2013 8:16:18 AM                                                            113
 
1305 NRC RO Exam
: 45. 039 K5.08 045 Given the following:
            -  Unit 2 startup in progress.
            -  The Main Generator has just been synchronized and loaded to 40 MWe.
            -  The Steam Pressure (2-PT-i -33) input to the steam dump controller fails HIGH.
Assuming NO action by the crew, which ONE of the following describes...
(1) the effect on core reactivity, and (2) how many of the steam dump valves would be responding to the failure?
A. Negative reactivity addition            3 valves only B. Negative reactivity addition            All 12 valves C. Positive reactivity addition            3 valves only D Positive reactivity addition            All 12 valves Wednesday, June 05, 2013 8:16:18 AM                                                    114
 
1305 NRC RO Exam Feedback DIS TRACTOR ANAL YSIS:
A. Incorrect, the reactivity addition is not negative; It is positive due to the cooldown.
Only 3 valves responding is not correct but only 3 valves would respond under different conditions with the Steam Dumps in Tavg Mode and the Lo-Lo Tavg interlocked bypassed (by procedure the SDMSS switch would be in steam pressure mode at this power level).
B. Incorrect, the reactivity addition is not negative; It is positive due to the cooldown but all 12 dumps valves responding is correct.
C. Incorrect, the positive reactivity addition is correct but only 3 valves responding is not correct. Only 3 valves can respond under different conditions with the Steam dumps in Tavg Mode and the Lo-Lo Tavg interlocked bypassed.
D. Correct, If the pressure input falls high, the controller will try to reduce pressure by opening dump valves. This would cause a positive reactivity addition to the reactor core. All 12 of the steam dump valves would be opening as a result of the failure.
while the MSIVs would close eventually close to stop the cooldown, with the pressure transmitter failed high all 12 dump valves would be open.
Notes Question Number:            45 Tier:      2    Group        1 K/A:      039 Main and Reheat Steam System (MRSS)
K5.08 Knowledge of the operational implications of the following concepts as the apply to the MRSS:
Effect of steam removal on reactivity Importance Rating:        3.6 / 3.6 10 CFR Part 55:          41.5/45.7 1OCFR55.43.b:            Not applicable K/A Match:      Questions requires the applicant to determine the effect a failure on the steam dump system will have on core reactivity and how the failure will impact the steam dump system operation.
Technical
 
==Reference:==
1-47W611-1-2 R13 OPT200SDCS, Steam Dump Lesson Plan Proposed references            None to be provided:
Wednesday, June 05, 2013 8:16:18 AM                                                                    115
 
1305 NRC RO Exam Learning Objective:          OPT200.SDCS
: 5. EXPLAIN, (or SKETCH as applicable), the physical connections and/or cause-effect relationships between the Steam Dump Control System and the following systems:
: i. Reactor Coolant System
: 7. EXPLAIN the Steam Dump Control System design features and/or interlocks that provide the following:
: c. RCS Cooldown
: e. Steam pressure control
: f. Reactor temperature control
: 13. Given specific plant conditions, ANALYZE the effect that a loss or malfunction of the Steam Dump Control System will have on the following:
: b. Reactor Coolant system
: d. Reactor Power Question Source:
New Modified Bank Bank                X Question History:            SQN bank question 039 K5.08 042 used on the 02/2010 audit exam Comments:
Wednesday, June 05, 2013 8:16:18 AM                                                          116
 
1305 NRC RO Exam
: 46. 059 A3.04 046 Given the following plant conditions:
              -  Unit 1 is at 72% power.
              -  PT-i -33, Steam Header Pressure to DOS, indicates 920 psig.
              -  PT-i -33A, Steam Header Pressure to DOS, indicates 900 psig.
              -  PT-i -33B, Steam Header Pressure to DOS, indicates 890 psig.
              -  PT-i -33 fails Low.
Which ONE of the following describes the effect this failure will have on the Unit 1 Main Feed Pump Turbine Master speed controller?
A. Output Increases.
B Output Decreases.
: 0. Output remains the same.
D. Transfers to Manual Feedback DIS TRACTOR ANAL YSIS:
A. Incorrect. Plausible if candiate confuses the selection criteria from median to average control input. For example, the examinee could conclude that with the channel failing low the average now is less and output would need to increase to maintain feedilow.
B. Correc1, Based on the design of the DCS with all three channels inputting to the OCS control, the circuit is designed to be a Median Select. Thus the Median signal would be 900 psig (the middle signal). With one channel failing, the logic goes to the average of the two remaining signals. The average of the two remaining signals would be 895 psig. Thus the controller would remain in automatic and the output would be going down (from 900 psig to 895 psig) and decrease the speed of MFP turbines.
C. Incorrect, Plausible since the control circuit could bypass the failed channel and go the single element control (another type of control in DCS for SG level control) and remain in Auto with no change in controller output.
: 0. lncorrecl, Plausible since the examinee could think that the failure caused the transfer feed pump speed control to manua this used to be the case with the previous control scheme.
Wednesday, June 05, 2013 8:16:18 AM                                                                117
 
1305 NRC RO Exam Notes Question Number:          46 Tier:    2    Group      1 K/A:      059 Main Feedwater A3.04 Ability to monitor automatic operation of the MEW, including:
Turbine driven feed pump Importance Rating:        2.5 / 2.6 10 CFR Part 55:          41.7 1OCFR55.43.b:            Not applicable K/A Match:      This question matches the KJA by testing the candidates knowledge of how the inputs from steam header pressure are integrated automatically into MEP turbine speed control and how those inputs are processed when one signal falls outside of the tolerance band and then affects control of MEP turbine speed.
Technical
 
==Reference:==
OPT200.DCS/1 -SO-98 r6 Proposed references            None to be provided:
Learning Objective:            OPT200.DCS
                                      #9 State the DCS response to:
Failing of one of two channels Question Source:
New Modified Bank Bank                  X Question History:              SQN Bank, SQN ILT 1305 NRC Exam Comments:                      High Cognitive Wednesday, June 05, 2013 8:16:18 AM                                                      118
 
1305 NRC RO Exam
: 47. 061 K2.01 047 Which ONE of the following is the Alternate power supply for the Unit 2 TDAFW pump Trip and Throttle valve, 2-FCV-1-51A-S?
A. 125v DC Vital Board I B 125v DC Vital Board II C. 125v DC Vital Board Ill D. 125v DC Vital Board IV Feedback DISTRACTOR ANAL YSIS:
A. Incorrect, 125v DC Vital Board I is the normal power supply for the Unit 2 TD-AFW pump Trip and Throttle valve.
B. Correct, the alternate power supply for the Unit 2 TD-AFW pump Trip and Throttle valve, 2-FCV- 1-5 lA-S is from 125v DC Vital Board II.
C. Incorrect, 125v DC Vital Board III is the normal power supply for the Unit 1 TD-AFW pump Trip and Throttle valve.
: 0. Incorrect, 125v DC Vital Board IV is the alternate power supply for the Unit 1 TD-AFW pump Trip and Throttle valve.
Wednesday, June 05, 2013 8:16:18 AM                                                            119
 
1305 NRC RO Exam Notes Question Number:            47 Tier:      2    Group        1 K/A:      061 Auxiliary / Emergency Feedwater (AFW) System K2.01 Knowledge of bus power supplies to the following:
AFW system MOVs Importance Rating:        3.2* / 33 1OCFRPart55:            41.7 1OCFR55.43.b:            Not applicable K/A Match:      Questions requires knowledge of the power supply to an MOV on the steam supply to Unit 2 TDAFW pump.
Technical
 
==Reference:==
AOP-P.02, Loss of 125V DC Vital Battery Board, Revision 13 Proposed references            None to be provided:
Learning Objective:            OPT200.DC
                                      #4 Explain the physical connections and/or cause and effect relationaships between DC and teh following systems:
AC electrical( DC Loads)
Question Source:
New                    X Modified Bank Bank Question History:              New question for the SQN 05/2013 NRC exam.
Comments:
Wednesday, June 05, 2013 8:16:18 AM                                                        120
 
1305 NRC RO Exam
: 48. 062 A2.09 048 Given the following plant conditions:
          -  Unit 1 is at 3% power.
          -  The monthly surveillance for lA-A DIG is in progress and the D/G is paralleled to the board.
          -  An inadvertent trip of 1 B 6.9kV Unit Board Normal Supply breaker results in operation of lA-A 6.9kV Shutdown Board DG Supply breaker 50 overcurrent relay.
Which ONE of the following completes the statements below?
The operation of the overcurrent relay will cause lA-A 6.9kV Shutdown Board DG supply breaker to trip open and      iL The conditions above      L    the reactor to be tripped.
A. (1) lock-out (2) require B. (1) lock-out (2) do NOT require Cv (1) subsequently recloses in automatic (2) require D. (1) subsequently recloses in automatic (2) do NOT require Wednesday, June 05, 2013 8:16:18 AM                                                      121
 
1305 NRC RO Exam Feedback DIS TRACTOR ANAL YSIS:
A. Incorrect, plausible as some overcurrent devices will lockout the board. The second part is correct.
B. Incorrect, plausible as some overcurrent devices will lockout the board. The second part is plausible as the unit is in Mode 2 which has the <P7 trips disabled, one of which is the loss of flow trip due to loss of one RCP.
C. Correct, with the diesel tied on to the SDB in parallel with offsite power the instantaneous 600 amp trip is in play. With a COW and RCP pump running the 600 amp 51 trip will drop the diesel supply breaker causing a load shed and stripping the bus. The diesel will then reclose (1912 breaker) and supply the SDB.
Procedurally in AOP-R.04, POP Malfunctions the crew is required to trip[ the reactor in step 2.
D. Incorrect, The first part is correct. The second part is plausible as the unit is in Mode 2 which has the <P7 trips disabled, one of which is the loss of flow trip due to loss of one RCP.
Wednesday, June 05, 2013 8:16:18AM                                                                122
 
1305 NRC P0 Exam Notes Question Number:          48 Tier:    2    Group      1 K/A:    062 AC Electrical Distribution System A2.09 Ability to (a) predict the impacts on 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:
Consequences of exceeding current limitations Importance Rating:        2.7 / 3.0*
10 CFR Part 55:          41.5 1OCFR55.43.b:            Not applicable K/A Match:      The question meets the KA as it requries to determine that an overcurrent condition exist, actions occurr and actions must be taken to mitigate the event.
Technical
 
==Reference:==
AOP-R.04 R27 AR-Mi-B, B-i R23 45N724-1, R2i 45N765-i, R24 45N765-2, R25 Proposed references            None to be provided:
Learning Objective:            OPL271AOP-R.04
                                      #9. List any conditions(s) that requires a Reactor or Reactor Coolant Pump trip in AOP-R.04.
OPT200.AC6.9KV Explain the operational implication of the following concept as it applies to the 6.9 KV Distribution system:
: f. exceeding current limitations Question Source:
New Modified Bank Bank                  X Question History:            WBN Bank Comments:
Wednesday, June 05, 2013 8:16:18 AM                                                            123
 
1305 NRC RO Exam
: 49. 063 A1.01 049 Which ONE of the following identifies the required capacity for the 125v Vital DC batteries to satisfy coping time requirements?
A. Must be able to mitigate a Station Blackout event for 2 hours without any required operator action relating to the 125v Vital DC system.
B. Must be able to mitigate a Station Blackout event for 4 hours without any required operator action relating to the 125v Vital DC system.
C Must be able to mitigate a Station Blackout event for 4 hours and to meet the requirement, loads must be stripped from the batteries within 45 minutes into the Station Blackout.
D. Must be able to mitigate a Station Blackout event for 2 hours and to meet the requirement, loads must be stripped from the batteries within 45 minutes into the Station Blackout.
Feedback DIS TRACTOR ANAL YSIS:
A. Incorrect, FSAR 8.3.2.1.1 identifies that the safety related DC power system is required to mitigate a station blackout event for 4 hours (not 2 hours) and that the stripping of loads is necessitated before 45 minutes into the event.
B. Incorrect, Being required for 4 hours is correct in accordance with FSAR 8.3.2.1.1, however, operator actions are required to strip of loads prior to 45 minutes into the event.
C. Correct, FSAR 8.3.2.1.1 identifies that the safety related DC power system is required to mitigate a station blackout event for 4 hours and that the stripping of loads is necessitated before 45 minutes into the event. The capacity of the 125v Vital DC batteries is that with the batteries in a fully charged condition each batteiy has the capacity to supply the connected loads for 45 minutes and to supply a reduced load for an additional 195 minutes during a loss of all AC power.
D. Incorrect, FSAR 8.3.2.1.1 identifies that the safety related DC power system is required to mitigate a station blackout event for 4 hours (not 2 hours). The stripping of loads being necessary before 45 minutes into the event is correct.
Wednesday, June 05, 20138:16:19AM                                                                    124
 
1305 NRC RO Exam Notes Question Number:          49 Tier:    2    Group      1 K/A:    063 D.C. Electrical Distribution Al .01 Ability to predict and/or monitor changes in parameters associated with operating the DC electrical system controls including:
Battery capacity as it is affected by discharge rate Importance Rating:        2.5 / 3.3 1OCFR Part 55:          41.5/45.5 1OCFR55.43.b:            Not applicable K/A Match:      Question requires applicant to predict the manual action required to reduce the discharge rate on the battery to ensure the Vital DC electrical distribution system is able to maintain voltage throughout its required Coping Time Technical
 
==Reference:==
FSAR 8.3.2.1.1 Proposed references            None to be provided:
Learning Objective:            OPT200.DC
: 7. EXPLAIN the operational implication of the following concept as it applies to the DC Systems:
: c. Discharge rate effect on battery capacity Question Source:
New Modified Bank Bank                    X Question History:              SQN bank question used on SQN1/2009 Audit exam Comments:
Wednesday, June 05, 2013 8:16:19 AM                                                              125
 
1305 NRC AC Exam
: 50. 063 K3.02 050 Given the following plant conditions:
            -    Unit 1 was operating at 100 % power when a safety injection occurred.
            -    Eighteen (18) seconds after Safety Injection, a loss of 125v Vital DC Power Channel II occurs.
Which ONE of the following identifies the current status of RHR pump 1 B-B?
A. RHR pump 1 B-B is NOT running but can be started from the MCR handswitch.
B. RHR pump 1 B-B is NOT running and can NOT be started from the MCR handswitch.
C. RHA pump 1 B-B is running and can be stopped from the MCR handswitch.
D RHR pump 1 B-B is running but can NOT be stopped from the MCR handswitch.
Feedback DIS TRACTOR ANAL YSIS:
A. Incorrect, Plausible if the time delays associated with the pump starting with a blackout present are used. The delay times would exceed the 18 seconds (DG Start and RHR blackout time delay relay) and there are 3 other Vital DC boards available to supply the control power. One of which does supply the breaker but it is a manual transfer, not an automatic transfer.
B. IncorrecI, Plausible since RHR pump 18-B cannot be started or stopped from the control room handswitch after the loss of 125v DC Vital channel!! and because if a blackout signal had been concurrent with the SI condition, then the DG start time and pump start delay time would have exceeded the time prior to the loss of the control power.
C. Incorrect, Plausible because the pump being running is correct and there are 3 other channels of 125v DC available that could have been determined to be the control power supply for the pumps breaker.
D. Correct, RHR pump 18-B would have started immediately when the Safety injection was initiated but after the 125v DC Channel!! power was losI the pump could not be stopped from its handswitch in the math control room.
Wednesday, June 05, 2013 8:16:19 AM                                                                126
 
1305 NRC RO Exam Notes Question Number:            50 Tier:      2    Group        1 K/A:    063 D.C. Electrical Distribution K3.02 Knowledge of the effect that a loss or malfunction of the DC electrical system will have on the following:
Components using DC control power Importance Rating:        3.5 / 3.7 1OCFRPart55:            41.7/45.6 1OCFR55.43b:            Not applicable K/A Match:      K/A is matched because the question requires the applicant to know a major breaker supplied with control power from 125v DC Vital Channel II and how a loss of the power supply to the control power affects the ability to start and stop the component.
Technical
 
==Reference:==
AOP-P.02, Loss Of 125V DC Vital Battery Board R2 1,2-45N765-13 R18 1 -45N724-2 R22 Proposed references            None to be provided:
Learning Objective:            OPL271AOP-P.02
                                      #11, Given a specific de-energized DC board, describe the affect on plant equipment and operations.
Cognitive Level:
Higher                  X Lower Question Source:
New Modified Bank Bank                    X Question History:              WBN bank question 063 K3.02 050 used on the 10/2011 exam 10-2009.
Comments:
Wednesday, June 05, 2013 8:16:19 AM                                                          127
 
1305 NRC RO Exam
: 51. 064 K3.03 051 The following plant conditions exist:
Unit 1 is at 100% rated thermal power Diesel Generator (DG) lA-A is being tested for its monthly surveillance test which requires loading to 4400 kw.
Current load is 2000 kw.
Subsequently, 1 B-B 6.9KV SDBD experiences a complete loss of voltage.
Which of the following describes the control the operator has over continued loading of the lA-A DG?
A. lA-A DG    j    be loaded by the operator to 4400kw, VAR loading cannot be changed by the operator.
B lA-A DG can be loaded by the operator to 4400kw, VAR loading          jj  be changed by the operator.
C. lA-A DG cannot be loaded by the operator to 4400kw, VAR loading cannot be changed by the operator.
D. lA-A DG cannot be loaded by the operator to 4400kw, VAR loading can be changed by the operator.
Feedback DIS TRACTOR ANAL YSIS:
A. Incorrect, plausible as the first part is correct. The second part is plausible as the LOR 86 relay does on certain conditions remove the load and voltage control from the MCR.
B. Correct, the DG while under test and parallel with offsite power, relay 62X prevents the 86 LOR from going to trip from the CES signal so the operator maintains control of both load and voltage and the DG stays in droop mode.
C. Incorrect, plausible as the LOR 86 relay does on certain conditions remove the load and voltage control from the MCR.
D. Incorrect, plausible as the LOR 86 relay does on certain conditions remove the voltage control from the MCR.
Wednesday, June 05, 2013 8:16:19 AM                                                                  128
 
1305 NRC RO Exam Notes Question Number:          51 Tier:      2    Group      1 K/A:    064 Emergency Diesel Generators (ED/G)
K3.03 Knowledge of the effect that a loss or malfunction of the ED/G system will have on the following:
ED/G (manual loading)
Importance Rating:        3.6 / 3*9*
10 CFR Part 55:        41.7 1OCFR55.43.b:          n/a K/A Match:      The question matches the K/A by testing the candidates knowledge of operation of the ED/G while being manually loaded and requiring understanding of the DG control circuits under a malfunction under manual load conditions.
Technical
 
==Reference:==
DWG. 45N767-4 R21(LOR 86 relay Ckt., PM)
Proposed references          None to be provided:
Learning Objective:          OPT200.DG Obj 6.c Explain the DG design features and/or operational interlocks that provide the following: Parallel operation, local and remote.
Cognitive Level:
Higher                X Lower Question Source:
New                  X Modified Bank Bank Question History:              New for ILT 1305 NRC Exam Comments:
Wednesday, June 05, 20138:16:19 AM                                                              129
 
___________
__
1305 NRC RO Exam
: 52. 073 K1.01 052 Given the following plant alarms:
            -  0-RA-90-125A, MAIN CNTRL RM INTAKE MON HIGH RAD
            -  0-RA-90-126A, MAIN CNTRL RM INTAKE MON HIGH RAD Which ONE of the following describes the Main Control Room ventilation alignment?
The MCR is maintained at a A. positive pressure by the Main Control Room Air Handling Units B. negative pressure by the Main Control Room Air Handling Units C positive pressure by the Control Building Emergency Air Pressurization Fans D. negative pressure by the Control Building Emergency Air Pressurization Fans Feedback DIS TRACTOR ANAL YSIS:
A. lncorrect, RM 125 and 126 initiate a CR1. A CR1 isolates the ductwork from the Control Building Air Pressurization Fans and start the Control Building Emergency Air Pressurization Fans. A positive pressure is maintained to minimize outside contaminants from entering the MCR. Plausible since this is the normal ventilation flowpath/pressure condition in MCR.
B. Incorrect, RM-90-125 and 126 initiate a CR! A CR! isolates the ductwork from the Control Building Air Pressurization Fans and start the Control Building Emergency Air Pressurization Fans. A positive pressure is maintained to minimize outside contaminants from entering the MCR. Plausible since MCR is a slight negative pressure during normal operation and normal ventilation flowpath.
C. Correct, RM 125 and 126 initiate a CR!. A CR! isolates the ductwork from the Control Building Air Pressurization Fans and start the Control Building Emergency Air Pressurization Fans. A positive pressure is maintained to minimize outside contaminants from entering the MCR.
D. lncorreci, RM 125 and 126 initiate a CR1. A CR! isolates the ductwork from the Control Building Air Pressurization Fans and start the Control Buildhg Emergency Air Pressurization Fans. A positive pressure is maintained to minimize outside contaminants from entering the MCR. Plausible since MCR is a slight negative pressure during normal operation.
Wednesday, June 05, 2013 8:16:19AM 130
 
_____
1305 NRC RO Exam Notes Question Number:          52 Tier:      2    Group      1 K/A:      073 Process Radiation Monitoring (PRM) System Ki .01 Knowledge of the physical connections and/or cause-effect relationships between the PRM system and the following systems:
Those systems served by PRMs Importance Rating:        3.6 / 3.9 10 CFR Part 55:        41.2 to 41.9 / 45.7 to 45.8 1OCFR55.43.b:          Not applicable K/A Match:      This question matches the K/A by having the candidate determine the cause-effect of Pad Monitors RM-90-125 & 126 and the Control Room Ventilation system.
Technical
 
==Reference:==
47W61 1-31-1 rev 28 0-XA-55-12B rev 29 Proposed references            None to be provided:
Learning Objective:          OPT200.RM Ob] 4.m Explain the physical connections and/or cause-effect relationships between the Radiation Monitoring System and the following systems:
Control Building Vent OPT200.CBVENT Obj 4.e Explain the physical connections and/or cause-effect relationships between the CBVENT and the following systems:
Control Room Emergency Ventilation (CREV) alignment/flow-path following initiation of Control Room Isolation (CR1) signal.
Question Source:
New Modified Bank Bank                    X Question History:            SQN bank question 073 Ki .01 052 used on the SON 09/2010 NRC exam.
Comments:
Wednesday, June 05, 20138:16:19 AM                                                            131
 
1305 NRC RO Exam
: 53. 076 A4.02 053 Which ONE of the following identifies how 1 -FCV-67-66A, ERCW HDR 1 A SUP TO DG1A-A HX Al &A2 responds to a normal start and stop of Diesel generator lA-A?
The ERCW valve will automatically open after the start signal when the DG speed rises to iL When the diesel generator is stopped, the ERCW valve will be closed t?L.
LU START                                      STOP Av        40 rpm                    using the handswitch on 0-M-26 B.          40 rpm            automatically when DG stops following the idle speed run C.        200 rpm                    using the handswitch on 0-M-26 D.        200 rpm            automatically when DG stops following the idle speed run Feedback DISTRACTOR ANAL YSIS:
A. Correct, Speed switch 1 is picked up when the DG speed rises to 40rpm and one of its functions is to open the ERCW normal supply valve to provide cooling to the EDG. Following the diesel generator being stopped the valve remains open until it is closed by operator action.
B. Incorrect, Plausible because the valve is opened automatically by a speed switch that actuates when the speed rises to 40 rpm and there is an interconnection between the valve and the idle run in the SO instruction. However, it is to verify the valve is opened during idle run checks following a DG start, not to verify the valve closed following the idle run during shutdown.
C. Incorrec1, Plausible because the valve is opened automatically by a speed switch but it is not the 200 rpm speed switch and the valve being required to be closed manually is correct.
: 0. Incorrect, Plausible because the valve is opened automatically by a speed switch but it is not the 200 rpm speed switch and there is an interconnection between the valve and the idle run in the SO instruction. Howe ver it is to verify the valve is opened during idle run checks following a DG stari, not to verify the valve closed following the idle run during shutdown.
Wednesday, June 05, 2013 8:16:19 AM                                                                132
 
1305 NRC RO Exam Notes Question Number:            53 Tier:      2    Group        1 K/A:    076 Service Water System (SWS)
A4.02 Ability to manually operate and/or monitor in the control room:
SWS valves Importance Rating:        2.6 / 2.6 10 CFR Part 55:        41.7/45.5 to 45.8 1OCFR55.43.b:            Not applicable K/A Match:      the question requries the ability to monitor the automatic oepration of an ERCW (Service Water System) valve that supply water to an EDG and a condition that will require manual operation of the valve.
Technical
 
==Reference:==
0-SO-82-1, Diesel Generator lA-A, rev 0040 45N767-3 rev 24 Proposed references            None to be provided:
Learning Objective:            OPT200.DG Obj 4.g Explain the physical connections and/or cause-effect relationships between the DGs and the following systems or subsystems:
ERCW Question Source:
New                      X Modified Bank Bank Question History:              new question for the SQN Comments:
Wednesday, June 05, 2013 8:16:19 AM                                                            133
 
1305 NRC RO Exam
: 54. 078 G2.4.35 054 Given the following plant conditions:
          -    Unit 1 was at 100% power when a LOCA occurred.
          -    AUOs are performing EA-32-1, Establishing Control Air to Containment, Appendix A, Re-establishing Air to Containment on Unit 1.
          -    Steps for opening FCV-32-80, Rx Bldg Train A Essential Air, are in progress.
The reason for holding 1-HS-32-80A in OPEN while depressing the test button for 30 seconds is to ensure A. Phase B closure signal for FCV-32-80 is reset B. auxiliary air compressor shutdown relays are bypassed while compressor loads C header pressure downstream of FCV-32-80 is greater than 50 psig D. Train A essential air pressure on 0-PI-32-1 04A remains greater than 77 psig Feedback DIS TRACTOR ANAL YSIS:
A. Incorrect. The Phase B closure signal is verified reset when the Phase B is reset in step 2 of EA 1. This is plausible if examinee determines that holding the test button will bypass the Phase B signal.
B. Incorrect. The aux air compressor shutdown relays are not bypassed by this action. This is plausible if examinee determines that since these actions are contained in an Emergency Abnormal (EA) procedure, special actions are needed to facilitate restoration of air to containment.
C. Correct. The circuit for controlling the position of FCV-32-80, Rx Bldg Train A Essential Air(as well as 80 and 102) require that pressure downstream of the valve be greater than 50 psig to allow the valve to remain open.
: 0. Incorrect. The essential air pressure is an input for maintaining the containment air isolation valves in the open position however this pressure is upstream of FCV-32-80 and is not part of the valve position curcuit. This is plausible since this value is used in EA 1 for ensuring Aux Air Compressors are running and air dryers are properly configured.
Wednesday, June 05, 2013 8:16:19 AM                                                                134
 
1305 NRC RO Exam Notes Question Number:          54 Tier:      2    Group      1 K/A:      078 Instrument Air System G2.4.35 Knowledge of local auxiliary operator tasks during an emergency and the resultant operational effects.
Importance Rating:        3.8 / 4.0 1OCFR Part 55:          41.10/43.5/45.13 1OCFR55.43.b:            Not applicable K/A Match:      Question matches the K/A by since the examinee is required to recall the local actions required re-establish air to containment following an Phase B.
Technical
 
==Reference:==
1 ,2-47W61 1-32-2 Rev 9 EA-32-1, Establishing Control Air to Containment Proposed references          None to be provided:
Learning Objective:          OPT200.CSA Obj 7.d Explain the Control and Service Air System design features and/or interlocks that provide the following:
Automatic isolation of sections of the air system.
Question Source:
New                    X Modified Bank Bank Question History:            SQN ILT 1305 Comments:
Wednesday, June 05, 20138:16:19 AM                                                            135
 
1305 NRC RO Exam
: 55. 103 K1.O1 055 Given the following:
          -    In accordance with 0-SO-30-5, Lower Compartment Cooling Units, which of the following identifies:
(1) the preferred cooler that should always be in service for Unit 1 and (2) if this cooler trips, what effect this will have?
A. (1)A-A (2) PZR Enclosure heats up B.    (1)B-B (2) PZR Eclosure heats up.
C.    (1)A-A (2) Auto start of the Standby Cooler D.    (1)B-B (2) Auto start of the Standby Cooler Wednesday, June 05, 20138:16:19 AM                                              136
 
1305 NRC RO Exam Feedback DIS TRACTOR ANAL YSIS:
A. Correct, In accordance with O-SO-30-5 the preferred CNMT cooler is the A-A cooler due to its location to the exhaust plenums that supply air flow to the PZR enclosure. Also as indicated in O-SO-30-5, if the A-A cooler is not in service then all other coolers must be started to provide adequate air flow to the PZR enclosure or the equipment in the enclosure will start to heat up and instrument readings may become erratic.
B. Incorrect, Plausible since the B-B cooler is located near the A-A cooler and since it is on plenun that supplies air to the PZR enclosure it would be a logical choice. Also since the B-B cooler is physically closer to the PZR enclosure the candidate would conclude that if the B-B cooler tripped the PZR enclosure would heat up. However the A-A cooler is the preferred cooler.
C. Incorrect, Plausible since the A-A cooler is the preferred cooler. Also plausible since there is an Auto Start associated with the CNCT Coolers and a trip of the running cooler would cause low air flow which would start the cooler selected for Standby, however this feature is overriden by the switch position in the control room such that no CNMT cooler auto start would be permitted.
: 0. Incorrect, Plausible since the B-B cooler is located near the A-A cooler and since it is on plenun that supplies air to the PZR enclosure it would be a logical choice. Also plausible since there is an Auto Start associated with the CNCT Coolers and a trip of the running cooler would cause low air flow which would start the cooler selected for Standby, however this feature is overriden by the switch position in the control room such that no CNMT cooler auto start would be permitted.
Wednesday, June 05, 20138:16:19 AM                                                                137
 
1305 NRC RO Exam Notes Question Number:          55 Tier:    2      Group    1 K/A:    103 Containment System Ki .01 Knowledge of the physical connections and/or cause-effect relationships between the Containment System and the following systems:
CCS Importance Rating:        3.6 / 3.9 1OCFRPart55:            41.2to41.9 1OCFR55.43.b:          Not applicable K/A Match:      This question matches the K/A by testing the candidates knowledge of the cause-effect relationship between the Containment Cooling System unit differences and the Containment System.
Technical
 
==Reference:==
0-S0-30-5, Lower Compartment Cooling Units, P35 1 (2)-47W845-3 ERCW Proposed references          None to be provided:
Learning Objective:          OPT200.CNTMTCLG & PURGE Obj 9.d Given specific plant conditions, analyze the effect that a loss or malfuction of the Containment Cooling and Purge Systems will have on the following:
Containment parameters (pressure, temperature and humidity)
Question Source:
New                  X Modified Bank Bank Question History:            New question written for the 1305 NRC exam Comments:
Wednesday, June 05, 20138:16:19 AM                                                              138
 
1305 NRC RO Exam
: 56. 002 K5.07 056 Given the following plant conditions:
              -  A Unit 1 reactor start up following a refueling outage is in progress lAW 0-GO-2, Unit Startup from Hot Standby to Reactor Critical.
              -  RCS boron concentration was 1350 ppm for the ECP calculation.
              -  Just prior to beginning rod withdrawal for the startup chemistry reports the following boron concentrations:
              -    RCS boron concentration: 1380 ppm
              -    Pressurizer boron concentration: 1325 ppm Which one of the following completes the statements below?
Critical rod height will be      (1)
PZR boron concentration          (2)    required to be raised before the startup can continue.
A (1) higher (2) is B. (1) higher (2) is NOT C. (1) lower (2) is D. (1) lower (2) is NOT Wednesday, June 05, 20138:16:19 AM                                                            139
 
1305 NRC RO Exam Feedback DISTRACTOR ANAL YSIS:
A. Correct: A higher RCS boron concentration will require the control rod withdrawal to be greater to overcome the added negative reactivity from it. 0-GO-2 requires that the RCS and PZR be equalized if they are not within 50 ppm of each other.
B. Incorrect, Plausible as the first part is correct. The second part is plausible as the examinee will have to remember the 50 ppm requirement within the startup procedure or will determine that the delta between the RCS and PZR is within limits.
C. Incorrect, Plausible if the examinee does not understand that a lower boron concentration will require less negative reactivity for the control rods to overcome and the rod height at criticality will be lower than the ECP. The second part is correct.
: 0. Incorrect, Plausible if the examinee does not understand that a lower boron concentration will require less negative reactivity for the control rods to overcome and the rod height at criticality will be lower than the ECP. The second part is plausible as the examinee will have to remember the 50 ppm requirement within the startup procedure or will determine that the delta between the RCS and PZR is within limits.
Wednesday, June 05, 2013 8:16:19 AM                                                                140
 
1305 NRC RO Exam Notes Question Number:          56 Tier:      2    Group      2 K/A:    002 Reactor Coolant System (RCS)
K5.07 Knowledge of the operational implications of the following concepts as they apply to the RCS: Reactivity effects of RCS boron, pressure and temperature Importance Rating:        3.6 / 3.9 10 CFR Part 55:        CFR: 41.5/45.7 1OCFR55.43.b:          Not applicable K/A Match:      The question matches the KA by requiring the examinee to understand reactivity effects on a reactor startup with changing boron concentrations in the RCS and PZR.
Technical
 
==Reference:==
0-GO-2, Unit Startup from Hot Standby to Reactor Critical R35 Proposed references          None to be provided:
Learning Objective:          OPT200. RCS 9.b Given plant conditions, identify and apply the following RCS limits and precautions related to the following:
GO-2, Unit Startup from Hot Standby to Reactor Critical Question Source:
New                  X Modified Bank Bank Question History:            New for SQN ILT 1305 NRC Exam Comments:                    Low Cognitive Wednesday, June 05, 20138:16:19 AM                                                            141
 
1305 NRC RO Exam
: 57. 015 A1.02 057 Given the following plant conditions:
              -  A reactor start up without Physics Testing is in progress lAW 0-GO-2, Unit Startup from Hot Standby to Reactor Critical.
            -  The crew has blocked P-6.
Which one of the following is the lowest of the listed values that exceeds the maximum SUR limit allowed per 0-GO-2?
A. 0.4 dpm B. 0.6 dpm C. 0.8 dpm D 1.1 dpm Feedback DISTRACTOR ANAL YSIS:
A. lncorrect, Plausible as 0.3 is the expected targeted SUR as delineated by 0-GO-2 in the section for raising power from criticality to block P-6.
B. incorrect, Plausible as 0.5 is the expected targeted SUR as delineated by 0-GO-2 in the section for raising power greater than P-6 up to the POAH.
C. Incorrect, plausible as 0.8 is higher than both 0.3 dpm and .5 dpm.
D. Correct, 1.0 dpm is the maximum limit of SUR listed in 0-GO-2.
Wednesday, June 05, 20138:16:19 AM                                                            142
 
1305 NRC RO Exam Notes Question Number:          57 Tier:      2    Group      2 KIA:      015 Nuclear Instrumentation System Al .02 Ability to predict and/or monitor changes in parameters to prevent exceeding design limits) associated with operating the NIS controls including: SUR Importance Rating:        3.5 / 3.6 10 CFR Part 55:          (CFR: 41.5.45.5) 1OCFR55.43.b:            Not applicable K/A Match:    The question places the examinee in an area of a reactor startup where the procedure has specific target SURs that should be expected and monitored. It also requires the examinee to know the maximu, SUP allowed.
Technical
 
==Reference:==
0-GO-2, Unit Startup from Hot Standby to Reactor Critical R35 Proposed references          None to be provided:
Learning Objective:          OPL271GO-2 Obj 10 State the startup rate limit.
Question Source:
New                    X Modified Bank Bank Question History:            New qusetion for SQN ILT 1305 exam Comments:
Wednesday, June 05, 20138:16:19 AM                                                      143
 
1305 NRC RO Exam
: 58. 016 A3.01 058 Given the following plant conditions:
          -  Unit 1 is operating steady state at 70% reactor power.
          -  Rod control is in MANUAL.
Compare the effects of either one of the following RCS Loop 1 RTD5 failing HIGH.
: 1. ThotRTD#1
: 2. Tcold RTD #1 Assuming NO operator action, which ONE of the following identifies the RCS RTD failure...
(1) having the larger effect on the pressurizer level control system and (2) how the pressurizer level would be affected?
Largest effect                      Level would...
A. Thot failure                rise and be controlled at a level higher than the 70% power steady state level.
B. Thot failure                rise but be restored to the 70% power steady state level by the control system.
C Tcold failure                  rise and be controlled at a level higher than the 70% power steady state level.
D. Tcold failure                  rise but be restored to the 70% power steady state level by the control system.
Wednesday, June 05, 20138:16:19 AM                                                          144
 
1305 NRC RO Exam Feedback DISTRACTOR ANAL YSIS:
A. Incorrect Plausible because the Thot failure would cause a change in the loop Tavg if the failed RTD was not rejected in the Tavg calculation and the change would then cause the pressurizer level control program setpoint to increase to a higher setpoint which would result in a charging flow increase to increase the pressurizer level.
B. Incorrec1, Plausible because the Thot failure would cause a change in the loop Tavg if the failed RTD was not rejected in the Tavg calculation and because it is rejected the level would return to the steady state 70% power setpoint after the failure.
C. Correct, The Tcold failure will cause the Auctioneered High RCS Tavg to signal to rise. This signal programs the pressurizer level control program setpoint resulting in an increase in the setpoint. The charging flow will increase to bring the pressurizer level up to the new setpoint.
D. Incorrect, Plausible because the Tcold failure having the largest effect is correct and because the level would return to the steady state 70% power setpoint if the failed RTD had been rejected in the Tavg calculation as is the Thot RTD failure.
Wednesday, June 05, 2013 8:16:19AM                                                                145
 
1305 NRC RO Exam Notes Question Number:          58 Tier:    2      Group      2 K/A:    016 Non-Nuclear Instrumentation System (NNIS)
A3.01 Ability to monitor automatic operation of the NNIS, including:
Automatic selection of NNIS inputs to control systems Importance Rating:        2.9* / 2.9*
10 CFR Part 55:        CFR: 41.7/45.5 1OCFR55.43.b:            Not applicable K/A Match:      This question matches the K/A by testing the candidates knowledge of how the inputs from loop temperature RTDs are selected automatically under a failure condition.
Technical
 
==Reference:==
1-AR-M5-A D-6 R36 47W610-68-1 thru 10 Proposed references            None to be provided:
Learning Objective:            OPT200.EAGLE 21 Obj 4 Explain the physical connections and/or cause-effect relationships between the Eagle 21 system and the following:
RCS transmitters for indication, control and protection.
Question Source:
New Modified Bank Bank                  X Question History:              WBN 06/2011 NRC exam, SQN ILT 1305 NRC Exam Comments:                      High Cognitive Wednesday, June 05, 2013 8:16:19 AM                                                            146
 
1305 NRC RO Exam
: 59. 027 K2.0I 059 Given the following conditions:
          -    Unit 1 is in Mode 3 when a Safety Injection occurs.
          -    Following the Safety Injection, both the lA-A 6.9KV Shutdown Board and the 2B-B 6.9 KV Shutdown Boards trip on differential.
Which ONE of the following describes the status of the EGTS system?
A. Neither EGTS fans are running.
B Only EGTS train B fan is running.
C. Only EGTS train A fan is running.
D. Both trains EGTS fans are running.
Feedback DISTRACTOR ANAL YSIS:
A. Incorrect, the EGTS fans do receive shutdown power and for neither to be running is plausible if the source of shutdown power to the 2 fans is thought to be from boards being fed from the 2 boards that are de-energized.
B. Correct, EGTS Train B is supplied from Unit 1 Train B Shutdown Power which is energized, but EGTS Train A is supplied from the Unit 1 Train A Shutdown power which is not energized.
C. Incorrect, Plausible if the EGTS fan power supplies are confused with the ABGTS fan power supplies. The ABGTS fans are supplied from Unit 2 and only the Train A ABGTS fan would be running with the 2 boards de-energized as stated in the question.
D. lncorrect the EGTS fan do receive shutdown power and for both to be running is plausible if the source of shutdown power to the 2 fans is thought to be from boards being fed from the 2 boards that remain energized.
Wednesday, June 05, 2013 8:16:20 AM                                                              147
 
1305 NRC RO Exam Notes Question Number:            59 Tier:      2    Group      2 K/A:    027 Containment Iodine Removal System (CIRS)
K2.01 Knowledge of bus power supplies to the following:
Fans Importance Rating:        3.1*/3.4*
10 CFR Part 55:        41.7 1OCFR55.43.b:          Not applicable K/A Match:      Question requires knowledge of the power supplies to the Emergency Gas Treatment System fans in order to determine which fans will be in service.
Technical
 
==Reference:==
AOP-P.05 rev 20 (App A, B and G for fan A, App L, M and R for fan B)
Proposed references            None to be provided:
Learning Objective:            OPT200.EGTS
: 5. LIST the bus power supplies to the following EGTS components:
: a. EGTS Fans Question Source:
New Modified Bank Bank                    X Question History:              WBN bank question 027K2.01 060 used on the WBN 05/2009 exam.
Comments:
Wednesday, June 05, 2013 8:16:20 AM                                                        148
 
1305 NRC RO Exam
: 60. 033 A2.01 060 Given the following plant conditions:
          -  Unit 1 is in Mode 6.
          -  The current Spent Fuel Pool boron concentration is 2120 ppm.
          -  Refueling Water Storage Tank (RWST) is being used as the TRM borated water source and is at the lowest boron concentration allowed as identified in SO-78-1 ,Spent Fuel Pit Coolant System.
          -  A leak on the Spent Fuel Pool cooling system results in the need for makeup from the RWST.
Which ONE of the following completes the statement below relative to the Spent Fuel Pool?
Using the RWST to makeup to the Spent Fuel Pool will result in alan in the Spent Fuel Pool boron concentration.
To meet LCD 3/4.7.13, SPENT FUEL MINIMUM BORON CONCENTRATION, requires a minimum boron concentration of { ppm.
A. decrease              1950 B. decrease              2000 C. increase              1950 D increase              2000 Wednesday, June 05, 2013 8:16:20 AM                                                  149
 
1305 NRC RO Exam Feedback DISTRACTOR ANAL YSIS:
A. Incorrect, The first part is incorrect. Tech Spec requires RWST to be greater than 2500 ppm and the SFP to be greater than 2000 ppm. Plausible if you use the Tech Spec value for SFP boron (2000 ppm) and the minimum assumed boron concentration for RWST used in UFSAR Chapter 15 Accident Analysis (1950 ppm).
The second part is incorrect. Tech Spec requires SFP to be >2000 ppm. This is plausible the minimum assumed boron concentration for RWST used in UFSAR Chapter 15 Accident Analysis (1950 ppm), which could also apply to the SFP since the UFSAR states that SPF Keff will remain 0.95 with SFP boron at 700 ppm under the most severe postulated fule mis-location accident.
B. Incorrect, The first part is incorrect. Tech Spec requires RWST to be greater than 2500 ppm and the SFP to be greater than 2000 ppm. Plausible if you use the Tech Spec value for SFP boron (>2000 ppm) and the minimum assumed boron concentration for RWST used in UFSAR Chapter 15 Accident Analysis (1950 ppm).
The second part is correct. Tech Spec requires SFP to be >2000 ppm.
C. Incorrect, The first part is correct. In accordance with SQ-78-1, the RWST will be
            > 2500 ppm, thus the addition of RWST water will raise the SFP boron concentration. The second part is incorrect. Tech Spec requires SFP to be >2000 ppm. This is plausible the minimum assumed boron concentration for RWST used in UFSAR Chapter 15 Accident Analysis (1950 ppm), which could also apply to the SFP since the UFSAR states that SPF Keff will remain <0.95 with SFP boron at 700 ppm under the most severe postulated fule mis-location accident.
D. Correct The first part is correct. In accordance with SO-78-1, the RWST will be>
2500 ppm, thus the addition of RWST water will raise the SFP boron concentration.
The second part is correct. Tech Spec requires SFP to be >2000 ppm.
Wednesday, June 05, 2013 8:16:20 AM                                                            150
 
1305 NRC RO Exam Notes Question Number:          60 Tier:      2    Group      2 K/A:    033 Spent Fuel Pool Cooling System (SFPCS)
A2.01 Ability to (a) predict the impacts of the following malfunctions or operations on the Spent Fuel Pool Cooling System ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:
Inadequate SDM.
Importance Rating:        3.0 / 3.5 10 CFR Part 55:          41.5 / 43.5/45.3/45.13 1OCFR55.43.b:            Not applicable K/A Match:
Technical
 
==Reference:==
Tech Spec 3.7.13 0-SO-78-1 rev 59 UFSAR rev 22, Chapter 4 Nuclear Design pg 4.3-29 UFSAR rev 22, Chapter 15 Accident Analysis pg 15.2-37 Proposed references            None to be provided:
Learning Objective:            OPT200.SFPC Obj 8.e Explaing the Spent Fuel Pit Cooling system design features and/or interlocks that provide the following:
Adequate Shutdown Margin (boron concentration)
Question Source:
New                    X Modified Bank Bank Question History:              New for NRC ILT 1305 exam Comments:                      Low Cog Wednesday, June 05, 2013 8:16:20 AM                                                          151
 
1305 NRC RO Exam
: 61. 034 K6.02 061 Given the following plant conditions:
          -  Unit 1 is at 100% RTP.
          -    Fuel Assembly shuffles are being made in the Spent Fuel Pit.
          -  0-RM-90-102, Spent Fuel Pit Radiation Monitor, has been declared INOPERABLE and removed from service due to an instrument malfunction.
Which ONE of the following completes the statements below?
Technical Specifications would          continued movement of fuel assemblies in the Spent Fuel Pit.
If 0-RM-90-1 03, Spent Fuel Pit Radiation Monitor, subsequently detects a high Radiation condition, of Auxiliary Building Isolation equipment will actuate with no operator actions.
LU A      allow                    gjjy one train B.      allow                        both trains C. NOT allow                    pjjjy one train D. NOT allow                      both trains Wednesday, June 05, 2013 8:16:20 AM                                                      152
 
1305 NRC RO Exam Feedback DIS TRACTOR ANAL YSIS:
A. Correct, Tech Specs only require one SFP Radiation monitor and one ABGTS system to be operable for fuel movement in the Spent Fuel Pit The SFP Radiation Monitors are trained in respect to isolating the Auxiliary Building with 102 being Train A and 103 being Train B.
B. incorrect, Tech Specs only require one SFP Radiation monitor and one ABGTS system to be operable for fuel movement in the Spent Fuel Pit Both trains would not be isolated if radiation was detected by the other SFP Radiation Monitor. The monitors are trained in respect to isolating the Auxiliary Building with 102 being Train A and 103 being Train B. Plausible because fuel shuffle continuing is correct and RM 101 (on the Aux. Building Stack) would cause isolation of both trains.
C. lncorrecl Fuel shuffles would not have to be stopped, the movement could continue because only one Rad Mon is required in the Spent fuel Pit. Plausible because other conditions would cause fuel movement to be stopped and only one train of the isolation in the Aux Building is correct.
D. IncorrecI, Fuel shuffles would not have to be stopped, the movement could continue because only one Rad Mon is required in the Spent fuel Pit and both trains of isolation of the Aux Building would not be isolated if radiation was detected by the other SFP Radiation Monitor. The monitors are trained in respect to isolating the Auxiliary Building with 102 being Train A and 103 being Train B. Plausible because other conditions would cause fuel movement to be stopped and RM-09-101(on the Aux. Building Stack) would cause isolation of both trains.
Wednesday, June 05, 2013 8:16:20 AM                                                                  153
 
1305 NRC PC Exam Notes Question Number:          61 Tier:    2    Group      2 K/A:    034 Fuel Handling Equipment System (FHES)
K6.02 Knowledge of the effect of a loss or malfunction on the following will have on the Fuel Handling System:
Radiation monitoring systems Importance Rating:        2.6 / 3.3 1OCFRPart55:            41.7/45.7 1OCFR55.43.b:            Not applicable K/A Match:      Question requires the knowledge of how a loss of Radiaition monitors affects the use of fuel handling equipment.
Technical
 
==Reference:==
1 ,2-47W61 1-30-5 R9 1 ,2-47W61 1-30-6 Ri 1 Tech Spec Table 3.3-6 Tech Spec 3.9.12 Proposed references            None to be provided:
Learning Objective:            OPT200.ABVENT
: 4. EXPLAIN the physical connections and/or cause-effect relationships between the ABVENT system and the following systems:
: a. Radiation Monitoring 0-RM-90-1 01 B, Aux Bldg Exhaust Stack Monitor 0-RM-90-.102 & 103, Fuel Pool Monitors Question Source:
New Modified Bank Bank                    X Question History:              SQN bank question 034 K6.02 061 used on the SQN 1/2009 retake exam formatted and the correct answer relocated for use on the SQN 05/2013 exam.
Comments:
Wednesday, June 05, 2013 8:16:20 AM                                                          154
 
1305 NRC RO Exam
: 62. 071 K3.05 062 Given the following:
          -  Unit 1 is operating at 100% power.
          -  0-HS-77-245 is postioned to align 0-FCV-77-245, waste gas vent flow control valve to Unit 1.
          -  Waste Gas Decay Tank J relief valve develops a flancie leak and the tank contains high activity gas.
Which ONE of the following identifies how the radiation monitors listed below will respond to the gas release?
Note:
0-RE 118, Waste Gas Rad Monitor 0-RE 101, Auxiliaty Building Ventilation Monitor 1-RE-90-400, Unit 1 Shield Building Vent Monitor A Only 0-RE-90-1 01 will detect the release.
B. Only 1 -RE-90-400 will detect the release.
C. Both 0-RE-90-1 18 and 1 -RE-90-400 will detect the release.
D. Both 0-RE-90-118 and 0-RE-90-101 will detect the release.
Wednesday, June 05, 2013 8:16:20 AM                                                      155
 
1305 NRC AC Exam Feedback DISTRACTOR ANAL YSIS:
A. Correct, leakage must enter the WGDT release line to pass by waste gas radiation monitor 0-RE-90-1 18. Flange leakage would not enter this line but rather the general Aux Building Spaces where it would eventually pass by Aux building stack radiation monitor 0-RE-90-101. Therefore for flange leakage, 0-RE-90-1 18 nor 1-RE-90-400 (the normal release point) would detect the release but 0-RE-90-101 would detect the release.
B. Incorrect; leakage must enter the WGDT release line to pass by waste gas radiation monitor 0-RE-90-1 18 and 0-.RE-90-400. Flange leakage would not enter this line but rather the general Aux Building Spaces where it would eventually pass by Aux building stack radiation monitor 0-RE 101. If the leakage had been through the seat of the valve and the normal release point was aligned 0-RE-90-400 would have detected the release. Therefore for flange leakage, 0-RE-90-1 18 would not detect the release but 0-RE-90-101 would detect the release. Plausible if applicant does not understand the relationship between the ventilation system and rad monitors and thinks the gas would exit via the Shield Building Stack which is where a normal gas decay tanks release is routed.
C. Incorrect; leakage must enter the WGDT release line to pass by waste gas radiation monitor 0-RE 118 and its normal release point monitored by 1 -RE-90-400. Flange leakage would not enter this line but rather the general Aux Building Spaces where it would eventually pass by Aux building stack radiation monitor 0-RE-90-101. Therefore for flange leakage, 0-RE-90-1 18 nor 1-RE-90-400 would detect the release but 0-RE-90-101 would. Plausible if applicant does not understand the relationship between the ventilation system and rad monitors.
: 0. Incorrect; leakage must enter the WGDT release line to pass by waste gas radiation monitor 0-RE-90-1 18. Flange leakage would not enter this line but rather the general Aux Building Spaces where it would eventually pass by Aux building stack radiation monitor 0-RE-90-101. Therefore for flange leakage, 0-RE-90-1 18 would not detect the release but 0-RE-90-101 would detect the release. Plausible if applicant does not understand the relationship between the ventilation system and rad monitors.
Wednesday, June 05, 2013 8:16:20 AM                                                              156
 
1305 NRC RO Exam Notes Question Number:          62 Tier:      2    Group        2 K/A:    071 Waste Gas Disposal System (WGDS)
K3.05 Knowledge of the effect that a loss or malfunction of the Waste Gas Disposal System will have on the following:
ARM and PRM systems Importance Rating:        3.2 / 3.2 1OCFRPart55:            41.7/45.6 1OCFR55.43.b:            Not applicable K/A Match:      K/A is matched because the question requires knowledge of how a leaking flange on the will affect the radiation monitoring systems at the station.
Technical
 
==Reference:==
1 ,2-47W830-4 R47 1 -47W866-i R41 1 ,2-47W866-2 Ri 3 i,2-47W866-10 R19 Proposed references            None to be provided:
Learning Objective:            OPT200.GRW
: 4. EXPLAIN (or SKETCH as applicable) the physical connections and/or cause-effect relationships between the Gaseous Radwaste System and the following systems:
: c. Radiation Monitors, RM-90-1 18
: g. Ventilation Question Source:
New Modified Bank Bank                    X Question History:              Originally modified from SQN question 060 AA1 .02 022 that was used on the SQN January 2008 exam Comments:
Wednesday, June 05, 2013 8:16:20 AM                                                          157
 
1305 NRC RO Exam
: 63. 072 A4.01 063 Given the following plant conditions:
              -  Both units are operating at 100% power.
              -  1-RA-90-.1C AUX BLDG AREA RAD MON INSTR MALFUNC (M12A, B-7) is LIT Which ONE of the following completes the statements below?
1-RA-90-1C INSTR MALFUNC alarm                _JJL        result in an automatic action.
{)_      is a condition that would cause the alarm.
A. (1) will (2) Operate/Calibrate switch set to calibrate B. (1) will (2) Sample flow less than setpoint C (1) will NOT (2) Operate/Calibrate switch set to calibrate D. (1) will NOT (2) Sample flow less than setpoint Feedback DISTRACTOR ANAL YSIS:
A. Incorrect, Plausible because other ARMs have automatic actions (0-RM 102 and -103 Spent Fuel Pit Rad Monitors). The Operate/Calibrate switch set to calibrate position is listed as a cause of the alarm and is correct.
B. Incorrect, Plausible because other ARMs have automatic actions (0-RM-90-102 and -103 Spent Fuel Pit Rad Monitors). The second part is plausible as a sample flow less than setpoint will cause an instrument malfunction alarm on a process radiation monitor.
C. Correct, 1-RA iC AUX BLDG AREA RAD MON does not have any auto actions. The Operate/Calibrate switch set to calibrate position is listed as a cause of the alarm and is correct.
: 0. Incorrect, the first part is correct. The second part is plausible as a sample flow less than setpoint will cause an instrument malfunction alarm on a process radiation monitor.
Notes Question Number:            63 Wednesday, June 05, 2013 8:16:20 AM                                                              158
 
1305 NRC RO Exam Tier:    2    Group      2 K/A:    072 Area Radiation Monitoring System A4.01 Ability to manually operate and/or monitor in the control room:
Alarm and interlock setpoint checks and adjustments Importance Rating:        3.0* /33 10 CFR Part 55:          41.7 / 45.5 to 45.8 1OCFR55.43.b:            Not applicable K/A Match:      K/A is matched because the question requires the ability to monitor the conditions of the area radiation monitor given a MCR alarm and the knowledge of the effect of setting the Operate/Calibrate switch to calibrate.
Technical
 
==Reference:==
0-AR-M12-A rev 52, window B-7 Proposed references            None to be provided:
Learning Objective:            OPT200.RM
: 2. STATE the location of the following listed Radiation Monitoring System components, and using in plant locations, LOCATE or IDENTIFY the associated indications and controls:
: c. Area RM5 (ARM)
: 3. Given plant conditions, DETERMINE if any of the following Radiation Monitoring System alarms would be present and DESCRIBE actions required by the ARP:
: f. [0-XA-55-12AJD, B/A7j AUX BLDG AREA RAD MON INSTR MALFUNC
: 6. EXPLAIN the Radiation Monitoring System design features and/or operational interlocks that provide the following:
: c. System failure/malfunction indication Question Source:
New                    X Modified Bank Bank Question History:              New for ILT 1305 exam Comments:
Wednesday, June 05, 2013 8:16:20 AM                                                            159
 
1305 NRC RO Exam
: 64. 079 K4.0I 064 Given the following plant conditions:
          -    Unit 2 is in Mode 6 with fuel movement in progress.
          -    Maintenance personnel report that a Service Air connection providing air to a pneumatic grinder has broken and cannot be isolated.
          -    Air Header Pressure indications are as follows:
PI-32-104A, AUX CONT AIR HDR A PRESS            92 psig Pl-32-105A, AUX CONT      AIR  HDR  B PRESS    92 psig Pl-32-200, CONT AIR HDR PRESS                    94 psig and slowly rising PI-33-199, SERV AIR HDR PRESS                    90 psig and slowly lowering Which ONE of the following completes the statements below?
(1)    At the current pressure, PCV-33-4, SERVICE AIR ISOL            (1)
(2)    PCV-33-4 valve        L      after the Service Header is repressurized.
A (1)    remains open but will be closed if pressure continues to drop (2)  must be reset B. (1)    is closed (2)  must be reset C. (1)    remains open but will be closed if pressure continues to drop (2) will reopen automatically D. (1)    is closed (2)  will reopen automatically Wednesday, June 05, 2013 8:16:20 AM                                                            160
 
1305 NRC RO Exam Feedback DISTRACTOR ANAL YSIS:
A. Correct. PCV-33-4 closes at 88 psig and lowering. AR-M15B window D-7, ZS-33-4 SERVICE AIR ISOL CLOSED provides indication of PCV-33-4 and is not in alarm. The second part is correct. AOP-M.02 and O-SO 1 provide instructions for depressing the REST pushbutton on P5-33-4 after Service Header is pressurized.
B. Incorrect. The valve is open as pressure in the service air header is> 88 psig.
AR-M15B window D-7, ZS-33-4 SERVICE AIR ISOL CLOSED provides indication of PCV-33-4 and is not in alarm. The second part is incorrect. AOP-M.02 and O-S0-33-1 provide instructions for depressing the RESET pushbutton on PS-33-4 after Service Header is pressurized. This is plausible as most pressure regulating valves do not require a reset.
C. Incorrect. The first part is correct. The second part is incorrect. AOP-M.02 and O-S0-33-1 provide instructions for depressing the RESET pushbutton on PS-33-4 after Service Header is pressurized. This is plausible as most pressure regulating valves do not require a reset.
D. Incorrect. The valve is open as pressure in the service air header is> 88 psig.
AR-M15B window D-7, ZS-33-4 SERVICE AIR ISOL CLOSED provides indication of PCV-33-4 and is not in alarm. The second part is correct.
Wednesday, June 05, 2013 8:16:20 AM                                                            161
 
1305 NRC RO Exam Notes Question Number:            64 Tier:    2      Group      2 K/A:    079 Station Air System K4.0i Knowledge of SAS design feature(s) and/or interlock(s) which provide for the following: Cross-connect with lAS.
Importance Rating:        2.9 / 3.2 10 CFR Part 55:        41.7 1OCFR55.43.b:            Not applicable K/A Match:      Question matches the K/A by since the examinee is required to recall the actions necessary to re-open the valve and remember at what pressure the valve wil automtically isolate.
Technical
 
==Reference:==
AR-Mi 5-B rev 35, window D-7 AOP-M.02 rev 21 Proposed references            None to be provided:
Learning Objective:            OPT200.CSA Obj 7.d Explain the Control and Service Air System design features and/or interlocks that provide for the following:
Automatic isolation of sections of the air system.
Question Source:
New                    X Modified Bank Bank Question History:              New for SQN ILT 1305 Comments:
Wednesday, June 05, 2013 8:16:20 AM                                                              162
 
1305 NRC RO Exam
: 65. 086 K1.03 065 Given the following;
          -    Unit 1 is in Mode 4.
          -  AOP-N.03, External Flooding, has been implemented.
Which ONE of the following completes the statement below relative to the Flood Mode Spool pieces that connect the HPFP system to the Unit 1 Aux Feedwater System?
The spool pieces are installed during        JJI_ actions and they connect to the AFW piping on the discharge of the            fL AFW pump(s).
L1 A.      Stage I                      Motor Driven B      Stage II                      Motor Driven C.      Stage I                      Turbine Driven D.      Stage II                    Turbine Driven Feedback DIS TRACTOR ANAL YSIS:
A. Incorrect, Plausible because there are actions that occur during Stage I of the AOP and the second part is correct.
B. Correct, The spool pieces are installed during Stage 2 activities (See note for AOP-N.03 below) and they connect to the Motor Driven AFW piping on the discharge side of the AFW pumps.
C. Incorrect, Plausible because there are actions that occur during Stage I of the AOP and the connection to the TDAFW pump could achieve the same result.
D. Incorrect, The first part is correct. The second part is correct as the connection to the TDAFW pump could achieve the same result.
Wednesday, June 05, 2013 8:16:20 AM                                                                163
 
1305 NRC RO Exam Notes Question Number:            65 Tier:      2    Group      2 K/A:      086 Fire Protection System (FPS)
Ki .03 Knowledge of the physical connections and/or cause effect relationships between the Fire Protection System and the following systems:
AFW system Importance Rating:        3*4* /35*
10 CFR Part 55:          41.2 to 41.9 / 45.7 to 45.8 1OCFR55.43.b:            Not applicable K/A Match:      Question requires knowledge of the physical condition between the Fire Protection System and the Auxiliary Feedwater system and the condition that would result in the staged spool pieces being installed to connect the system together.
Technical
 
==Reference:==
AOP-N.03, External Flooding, Revision 43 Proposed references            None to be provided:
Learning Objective:            OPL271AOP-N.03 Ob] 5 Summarize AOP-N.03s mitigating strategies for Flooding.
Question Source:
New                    X Modified Bank Bank Question History:              New question for the SQN 05/2013 exam Comments:
Wednesday, June 05, 2013 8:16:20 AM                                                          164
 
1305 NRC RO Exam
: 66. G2.1.15 066 Which ONE of the following identifies the maximum time Standing Orders for administrative issues and whos approval is required in accordance with ODM-Y, Standing Orders and Shift Orders?
Standinci Orders                        Approval A.      30 days                      Operations Manager B.      30 days                      Operations Superintendent C.      1 year                        Operations Manager Dv    1 year                        Operations Superintendent Feedback DIS TRACTOR ANAL YSIS:
A. Incorrect, Plausible because 30 days is a time period for shift orders and the Operations Manager is logical choice for all administrative/policy procedures.
B. Incorrect, Plausible because 30 days is a time period for shift orders and the Operations Superintendent is the approver.
C. Incorrect, Plausible because 1 year being the maximum normal time for a Standing Orders for administrative issues is correct and the Operations Manager is logical choice for all administrative/policy procedures.
D. Correct, Operations Directive Manual (0DM) -Y states that a Standing Order for administrative issues should not remain in effect beyond 1 year and the Operations Superintendent is the approver.
Wednesday, June 05, 2013 8:16:20 AM                                                            165
 
1305 NRC RO Exam Notes Question Number:          66 Tier:      3    Group      n/a K/A:      G 2.1.15 Knowledge of administrative requirements for temporary management directives, such as standing orders, night orders, Operations memos, etc.
Importance Rating:        2.7 / 3.4 10 CFR Part 55:        41.10 / 45.12 1OCFR55.43.b:          Not applicable K/A Match:      KA is matched because the question requires knowledge of administrative requirements (maximum time restrictions) for standing orders and night orders, Technical
 
==Reference:==
ODM-Y, Standing Orders and Shift Orders, Revision 0 OPDP-1, Conduct of Operations, R24 Proposed references          None to be provided:
Learning Objective:          OPL271OPSMGMTL Obj 10 Describe the difference between Shift Orders and Standing Orders. Identify the requirements for both relative to assuming a shift operating position.
Cognitive Level:
Higher Lower                X Question Source:
New Modified Bank Bank                  X Question History:
Comments:
Wednesday, June 05, 20138:16:20 AM                                                          166
 
1305 NRC RO Exam
: 67. G2.1.43 067 Given the following plant condition:
          -  Unit 1 has entered Section 5.4, Power Coastdown at End of Life, of GO-5, Normal Power Operation.
Which ONE of the following identifies...
(1) how the unit will be operated during performance of GO-5, Section 5.4 and (2) why reactor power changes should be limited to 1% per hour?
A. (1) The crew will reduce turbine load as needed to maintain Tavg on program.
(2) To prevent MTC from exceeding Tech Spec limits.
B (1) The crew will reduce turbine load as needed to maintain Tavg on program.
(2) To avoid xenon peaking which could force a plant shutdown.
C. (1) The crew allow Tavg to drop while maintaining reactor power as stable as possible with AFD being maintained less than the positive limit.
(2) To prevent MTC from exceeding Tech Spec limits.
D. (1) The crew allow Tavg to drop while maintaining reactor power as stable as possible with AFD being maintained less than the positive limit.
(2) To avoid xenon peaking which could force a plant shutdown.
Wednesday, June 05, 2013 8:16:20 AM                                                167
 
1305 NRC RO Exam Feedback DISTRACTOR ANAL YSIS:
A. Incorrect, Plausible because turbine load being reduced daily to maintain Tavg on program is correct and because the MTC is at its most negative value at the end of core life do to very low boron concentration and has magnitude limits established in the TSandtheNDR.
B. Correct; 00-5, Section 5.4 identifies that turbine load will be reduced daily to maintain Ta vg on program and reactor power changes should be limited to
            <1 %/min to avoid xenon peaking which could force a plant shutdown (implied in section 5.4.8 and verified with RE as the basis for the slow power reduction with rods out.
C. Incorrect, Plausible because the direction to maintain AFD less than the positive limit is included in the Notes in Section 5.4. As power is reduced it shifts up in the core and is controlled by control rods not power reduction limits. Some Westinghouse will use this method of maintaining power constant and letting Tavg drift down. The second part is plausible because the MTC is at its most negative value at the end of core life do to very low boron concentration and has magnitude limits established in the TS and the NDR.
D. Incorrect, Plausible because the direction to maintain AFD less than the positive limit is included in the Notes in Section 5.4. As power is reduced it shifts up in the core and is controlled by control rods not power reduction limits. Some Westinghouse will use this method of maintaining power constant and letting Tavg drift down. The second part is correct.
Wednesday, June 05, 2013 8:16:20 AM                                                                  168
 
1305 NRC RO Exam Notes Question Number:          67 Tier:    3    Group K/A:      G 2.1.43 Ability to use procedures to determine the effects on reactivity of plant changes, such as reactor coolant system temperature, secondary plant, fuel depletion, etc.
Importance Rating:        4.1/4.3 10 CFR Part 55:          41.10/43.6 / 45.6 1OCFR55.43.b:            Not applicable K/A Match:      K/A is matched because the question requires the abiity to recall the directions and constraints of GO-4 (as related to RCS temperature and secondary plant changes) when the fuel is depleted and the plant is in a coastdown condition.
Technical
 
==Reference:==
GO-5, Noraml Power Operations, R82 TS LCO 3.1.1.3 Unit 1 NDRCycIe19 Unit 1 COLR Cycle 19 Proposed references            None to be provided:
Learning Objective:            OPL271GO-5 Obj 1 State the reason for each precaution and limitation provided in 0-GO-5.
Question Source:
New Modified Bank Bank                  X Question History:              WBN 10-2011 Audit Exam Comments:
Wednesday, June 05, 2013 8:16:20 AM                                                          169
 
1305 NRC RO Exam
: 68. G 2.2.22 068 Given the following plant conditions:
          -    Unit 1 is in MODE 1 with RCS pressure at 2235 psia.
          -  The following RCS leakages were determined per 1-Sl-OPS-068-137.0, Reactor Coolant System Water Inventory.
Total RCS leakage            12.41 gpm PRT leakage                    5.32 gpm CLA #1 leakage                0.88 gpm RCDT leakage                  3.18 gpm S/G #2 leakage                0.07 gpm SIG #1 ,#3,#4 leakage          0.00 gpm
[Note: Assume leakages other than those given as zero (0) gpm].
Which ONE of the following completes the statement below?
LCO 3.4.6.2, Reactor Coolant System Operational Leakage, action is required to be entered because the            limit has been exceeded?
: 1. IDENTIFIED LEAKAGE
: 2. UNIDENTIFIED LEAKAGE
: 3. PRIMARY-TO-SECONDARY LEAKAGE A. 1 only By 2 only C. 2and3only D. 1 and 3 only Wednesday, June 05, 2013 8:16:20 AM                                                  170
 
1305 NRC RO Exam Feedback DISTRACTOR ANAL YSIS:
Identified Leakage = PAT + CLA + RCDT + S/G Leakages
                            =5.32+ 0.88+3.18&#xf7;0.07 =9.45gpm Unidentified leakage = RCS leakage Identified leakage = 12.41- 9.45
                                              -                              = 2.96 gpm Primary-to Secondary Leakage = 0.07 gpm = 100.8 gpd A. Incorrect, The IDENTIFIED Leakage at 9.45 gpm is less than the LCO limit of 10 gpm. Plausible because the Identified Leakage is elevated to be close to the maximum allowed leakage.
B. Correct, The UNIDENTIFIED Leakage at 2.96 gpm is greater than the LCO limit of 1 gpm.
C. Incorrect, The first part is correct and the PRIMARY-TO-SECONDARY Leakage at 100.8 gpm is less than the LCO limit of 150 gpd. Plausible because the PRIMARY-TO-SECONDARY Leakage is greater than the leakage which requires a plant shutdown in accordance with the AOP for steam generator tube leakage.
D. Incorrect, The IDENTIFIED Leakage at 9.45 gpm is less than the LCO limit of 10 gpm. Plausible because the Identified Leakage is elevated to be close to the maximum allowed leakage. The second part is plausible because the PRIMARY-TO-SECONDARY Leakage is greater than the leakage which requires a plant shutdown in accordance with the AOP for steam generator tube leakage.
Wednesday, June 05, 20138:16:20 AM                                                        171
 
____
1305 NRC RO Exam Notes Question Number:            68 Tier:      3    Group      n/a K/A:    G 2.2.22 Knowledge of limiting conditions for operations and safety limits.
Importance Rating:        4.0 / 4.7 10 CFR Part 55:          41.5/43.2 / 45.2 1OCFR55.43.b:            Not applicable K/A Match:      Question requires knowledge of the LCO for Reactor Coolant System leakage.
Technical
 
==Reference:==
Tech Spec 3.4.6.2 Proposed references            None to be provided:
Learning Objective:            OPT200.RCS
: 11. Using the Technical Specifications, Technical Requirements Manual, and the ODCM,
: c. Given a set of plant conditions/parameters, DETERMINE entry level conditions for Reactor Coolant System Tech Spec LCO actions, Technical Requirements and/or ODCM Controls.
Question Source:
New Modified Bank Bank                    X Question History:              SQN bank question G 2.2.22 069 used on an Audit exam in 1/2008.
Comments:                      High Cognitive Wednesday, June 05, 2013 8:16:21 AM                                                        172
 
1305 NRC RO Exam
: 69. G2.2.42 069 Given the following plant conditions:
While performing a cooldown on Unit 1 from Mode 3 to Mode 4 the following parameters were logged:
Time            RCS Press    RCS Temp  PZR LIQ Space Temp 0200            2200 psig      553&deg;F          650&deg;F 0230            1550 psig      527&deg;F          606&deg;F 0300            1135 psig      505&deg;F          560&deg;F 0330              765 psig      447&deg;F          494&deg;F 0400              400 psig      402&deg;F          440&deg;F Which ONE of the following describes the Tech Spec I TRM implications of these conditions?
A RCS cooldown rate limits were exceeded; Tech Spec action is required within a maximum of 30 minutes.
B. RCS cooldown rate limits were exceeded; Tech Spec action is required within a maximum of 60 minutes.
C. PZR cooldown rate limits were exceeded; TRM action is required within a maximum of 30 minutes.
D. PZR cooldown rate limits were exceeded; TRM action is required within a maximum of 60 minutes.
Wednesday, June 05, 20138:16:21 AM                                                    173
 
1305 NRC RO Exam Feedback DISTRACTOR ANAL YSIS:
A. Correct, between 0300 and 0400 RCS coo/down rate was greater than 100&deg;F in one hour exceeds the limit, action is required within 30 minutes.
B. Incorrect, the RCS cooldown rate was exceeded however the Tech Spec action time is incorrect. This is plausible since 60 minutes is a common action time limit.
C. Incorrect, between 0300 and 0400, PZR cooldown rate limit was greater than 1 00&deg;F, however the TRM limit is 200&deg;F in any one hour. This is plausible if candidate uses the PZR heatup rate limit of 100&deg;F per hour rate. The required action time limit is correct.
: 0. Incorrect, the PZR coo/down rate limit was not exceeded (see item C). The required action time limit is also incorrect. This is plausible since 60 minutes is a common action time limit.
Wednesday, June 05, 2013 8:16:21 AM                                                              174
 
1305 NRC RD Exam Notes Question Number:          69 Tier:      3    Group    n/a K/A:    G 2.2.42 Ability to recognize system parameters that are entry-level conditions for Technical Specifications.
Importance Rating:        3.9 / 4.6 10 CFR Part 55:        41.7/41.10 / 43.2 / 43.3 / 45.3 1OCFR55.43.b:          Not applicable K/A Match:      Applicant must know entry level condition in order to know if the RCS or PZR cooldown limits have been exceeded.
Technical
 
==Reference:==
Tech Spec 3.4.9.1 PTLR rev 4, July 2003 TRM 3.4.9.2 Proposed references          None to be provided:
Learning Objective:          OPT200.RCS Obj 11 .a and c Using the Technical Specification, Technical Requirements Manual, and the ODCM:
: a. List from memory, Reactor Coolant System Tech Spec LCOs and/or Technical Requirement having action times <one hour.
: c. Given a set of plant conditions/parameters, determine entry level conditions for RCS Tech Spec LCD actions, TRM and or ODCM controls.
Question Source:
New Modified Bank Bank                  X Question History:            SON bank question Comments:                    High Cog Wednesday, June 05, 20138:16:21 AM                                                            175
 
1305 NRC P0 Exam
: 70. G 2.2.43 070 A main control room annunciator, 1-M5-B, A-7, HIGH PRESS IN AUX BLDG, has alarmed repeatedly over the past several hours. It is determined to be a nuisance alarm. All compensatory measures have been taken.
Aroval for disablement shall be obtained from the                    (1)      and the disablement shall be logged in the                  (2)
A. (1) Shift Manager ONLY (2) Disabled Annunciator Book ONLY B. (1) Shift Manager ONLY (2) Disabled Annunciator Book and the narrative log C. (1) Unit Supervisor or the Shift Manager (2) Disabled Annunciator Book ONLY D (1) Unit Supervisor or the Shift Manager (2) Disabled Annunciator Book and the narrative log Feedback DIS TRACTOR ANAL YSIS:
A. Incorrect, Plausible as the SM is one of the two positions that can approve the disablement. The second part is plausible as the Disabled Annunciator Book is one of the two places were the disablement is logged.
B. lncorrecI, Plausible as the SM is one of the two positions that can approve the disablement. The second part is correct.
C. lncorrect, The first part is correct. The second part is plausible as the Disabled Annunciator Book is one of the two places were the disablement is logged.
D. Correct, Either the SM or the US can sign the disabled alarm form OPDP 1 approving the alarm disablement. OPDP-4 requires that the disablement be logged in both the Disabled Annunciator Book and the narrative log.
Wednesday, June 05, 2013 8:16:21 AM                                                            176
 
____
1305 NRC RO Exam Notes Question Number:          70 Tier:    3    Group K/A:    2.2.43 Knowledge of the process used to track inoperable alarms.
Importance Rating:        3.0 / 3.3 10 CFR Part 55:          (CFR: 41.10/43.5 /45.13) 1OCFR55.43.b:            Not applicable K/A Match:      Question matches the KA by testing the examinees knowledge of the processs used to track inoperable alarms.
Technical
 
==Reference:==
OPDP-4, Annuciator Disablement, rev 5 Proposed references            None to be provided:
Learning Objective:            0PL271 OPDP-4 Obj 2 and 3
: 2. Describe the General Requirements for Annunciator Disablement
: 3. Describe the procedure to disable and alarm.
Question Source:
New                    X Modified Bank Bank Question History:              New for SQN ILT NRC 1305 Exam Comments:                      Low Cognitive Wednesday, June 05, 2013 8:16:21 AM                                                        177
 
1305 NRC RO Exam
: 71. G2.3.11 071 Given the folowing plant conditions:
          -  A Reactor Trip and safety injection have occurred on Unit 1 due to a SGTR.
          -  1-RA-90-1 19A CNDS VAC PMP LO RNG AIR EXH MON HIGH PAD alarm is LIT.
          -  1-PA-90-120A1121A STM GEN BLDN LIQ SAMP MON HI RAD alarm is LIT.
          -  Steam Generator parameters are as follows:
SG1        SG2      SG3      SG4 NP Level              27%        28%      32%      21%
(stable)  (lowering) (rising)  (rising)
AFW Flow            70 gpm      0 gpm    0 gpm    200 gpm Which ONE of the following is an action required to be taken to minimize the radiation release?
A. Raise #2 SG Atmospheric relief valve setpoint.
B Raise #3 SG Atmospheric relief valve setpoint.
C. Isolate the Steam Supply from the #1 SG to the TD AFW Pump turbine.
D. Isolate the Steam Supply from the #4 SG to the TD AFW Pump turbine.
Wednesday, June 05, 20138:16:21 AM                                                  178
 
1305 NRC RO Exam Feedback DISTRACTOR ANAL YSIS:
A. lncorrect, Steam Generator #3 is the ruptured steam generator (not steam generator #2) but if the #2 SQ had been ruptured raising the setpoint of the #2 SQ Atmospheric relief valve would be a correct action. Plausible if the wrong steam generator is diagnosed as ruptured because the action would be a correct action.
B. Correct, Steam Generator #3 is the ruptured steam generator as identified by the level rising with no AFW flow and an action to control the release of radiation is to raise the setpoint of the #3 SQ Atmospheric relief valve.
C. Incorrect, Steam Generator #3 is the ruptured steam generator (not steam generator #1) but if the #1 SQ had been ruptured isolating the steam supply from the SG#1 to the TD AFW pump turbine would be a correct action. Plausible if the wrong steam generator is diagnosed as ruptured because the action would be a correct action.
D. Incorrect, Steam Generator #3 is the ruptured steam generator (not steam generator #4) but if the #4 SQ had been ruptured ensuring the TD AFW pump turbine was not being supplied from the #4 SQ would be a correct action. Plausible if the wrong steam generator is diagnosed as ruptured because the action would be a correct action.
Wednesday, June 05, 2013 8:16:21 AM                                                              179
 
1305 NRC RO Exam Notes Question Number:            71 Tier:    3      Group      n/a K/A:      G 2.3.11 Ability to control radiation releases.
Importance Rating:          3.8 / 4.3 10 CFR Part 55:          41.11 /43.4/45.10 1OCFR55.43.b:            Not applicable K/A Match:      Question requires the ability to determine action needed usisng conditions provided during a SGTR to control a radaition release Technical
 
==Reference:==
E-3, Steam Generator Tube Rupture, Revision 20 Proposed references            None to be provided:
Learning Objective:            0PL271.E-3
: 6. Given the procedure and a set of initial plant conditions, determine actions required to mitigate the event in progress.
Question Source:
New Modified Bank Bank                    X Question History:              SON bank question G 2.3.11 070 used during the 1/2009 Retake exam with minor wording changes in the stem.
Comments:
Wednesday, June 05, 2013 8:16:21 AM                                                              180
 
1305 NRC RO Exam
: 72. G 2.3.5 072 Given the following plant conditions:
          -    A source check is to be performed on CCS radiation monitor 1-RE-90-123.
Which ONE of the following completes the statement below?
The source check is verified by observing the                jJ        and the isolation function of the Surge Tanks vent                    be manually blocked during the source check in accordance with 1 -SO-90-1, Liquid Process Radiation Monitors.
A. (1)    analog rate meter trending upscale (2)  can B. (1)    analog rate meter trending upscale (2)  can NOT C. (1)    bargraph display responds upscale (2)  can D (1)    bargraph display responds upscale (2)  can NOT Feedback DIS TRACTOR ANAL YSIS:
A. Incorrect, Plausible since observing the rate meter trending upscale is correct for different radiation monitors and because some radiation monitrors outputs can be blocked during the performance of a source check but not this one.
B. IncorrecI, Plausible since observing the rate meter trending upscale is correct for different radiation monitors and because the isolation function can not be blocked on this meter during the source check is correct.
C. Incorrect, Plausible since a succeesful source check being determined by observing the bargraph trending upscale is correct and some radiation monitors outputs can be blocked during the performance of a source check but not this one.
D. Correct, a successful source check is determined by observing the bargraph trending upscale. Also the isolation function is not available on this type monitor thus the valves could automatically isolate during the source check requiring them to be checked open after the source check is performed.
Wednesday, June 05, 2013 8:16:21 AM                                                                181
 
1305 NRC RO Exam Notes Question Number:            72 Tier:    3      Group      n/a K/A:      G 2.3.5 Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.
Importance Rating:          2.9/2.9 1OCFRPart55:            41.11 1OCFR55.43.b:            Not applicable K/A Match:      This question matches the K/A by requiring the applicant to identify how to perform a source check on fixed radiation monitoring equipment and to determine how functions performed by the monitor are affected by the source check.
Technical
 
==Reference:==
1-SO-90-1, Liquid Process Radiation Monitors, Rev 13 Proposed references            None to be provided:
Learning Objective:            OPT200RM Obj 7.1 Explain the operational implication of the following concept as it applies to the Radiation Monitoring System: Check source operation without blocking process function.
Question Source:
New Modified Bank Bank                    X Question History:              Original question used on Feb 2010 NRC exam Comments:
Wednesday, June 05, 2013 8:16:21 AM                                                          182
 
1305 NRC RO Exam
: 73. G 2.4.8 073 Given the following plant conditions:
          -    Unit 2 at 100% power.
          -    Panel 2-M-1 Annunciator Window 125V DC VITAL CHGR Ill FAILJVITAL BAT Ill DISCHARGE alarms.
          -    The crew enters AOP-P.02, Loss Of 125V DC Vital Battery Board.
          -    A reactor trip occurs on high PZR pressure.
Which ONE of the following identifies the allowed usage of AOP-P.02 after the Emergency Operating Procedure network is entered following the reactor trip?
Continued performance of AOP-P.02 is                    (1) because                    (2)
A. (1)    allowed after the crew enters ES-0.1, Reactor Trip Response (2)    ES-0.1 is NOT an accident mitigation EOP.
B (1)      allowed after the crew enters ES-0.1, Reactor Trip Response (2)    restoring power could have an impact on meeting the goals of the EOP.
C. (1)      NOT allowed until the crew exits ES-0.1, Reactor Trip Response (2)  the procedure reader must remain dedicated to the EOP in effect until the EOPs are exited.
D. (1)      NOT allowed until the crew exits ES-0.1, Reactor Trip Response (2)  actions taken in AOP-P.02 could degrade the performance of the EOP.
Wednesday, June 05, 2013 8:16:21 AM                                                183
 
1305 NRC RO Exam Feedback DIS TRACTOR ANAL YSIS:
A. Incorrect, As identified in EPM-4, selected AOPs, such as loss of vital power can be implemented concurrently with the EOPs. While ES-O. 1 is not an accident mitigation procedure, that is not reason for parallel implementation, it is because the loss of power AOPs can have a significant impact on the ability of the EOP to achieve it goals. Plausible because the parallel implementation is correct but the reason for parallel implementation is not correct.
B. Correct, As identified in EPM-4, selected AOPs, such as loss of vital power can be implemented concurrently with the EOPs because the loss of power AOPs can have a significant impact on the ability of the EOP to achieve it goals.
C. Incorrect, EPM-4 provides that EOPs have priority over AOPs, and normally a dedicated procedure reader is utilized in the EOP network however, selected AOPs are allowed to be performed concurrently with EOPs. Plausible if candidate fails to recognize that AOP-P.02 is allowed to be used while performing ES-O. 1.
: 0. Incorrect, EPM-4 provides that EOPs have priority over AOPs, however, selected AOPs are allowed to be performed concurrently with EOPs. Plausible if candidate fails to recognize that AOP-P.02 is allowed to be used while performing ES-O. 1 and knows that the AOP actions can not be taken if the action would degrade the EOP performance.
Wednesday, June 05, 2013 8:16:21 AM                                                            184
 
_____
1305 NRC RO Exam Notes Question Number:          73 Tier:    3      Group      na K/A:      G 2.4.8 Knowledge of how abnormal operating procedures are used in conjunction with EOP5.
Importance Rating:          3.8 / 4.5 10 CFR Part 55:          41.10 1OCFR55.43.b:            Not applicable K/A Match:      This question matches the K/A by testing the candidates knowledge of how AOP-P.02 is to be used in conjuction with the EOPs during a loss of vitla DC bus.
Technical
 
==Reference:==
EPM-4 rev 22 Proposed references            None to be provided:
Learning Objective:            0PL271 EPM-4 Obj 1 Determine/identify the correct procedural application(s) based on the operating procedures network for abnormal and emergency evolutions.
Question Source:
New Modified Bank Bank                  X Question History:              Bank question used on 1/2009 NRC exam Comments:
Wednesday, June 05, 2013 8:16:21 AM                                                            185
 
1305 NRC RO Exam
: 74. G 2.4.3 074 Given the following picture of various SG#1 pressure indications:
4          H Which ONE of the following identifies only Post Accident Monitoring (PAM) indications?
A. 1-Pl-1-2A and 1-Pl-1-5 B 1-Pl-1-2A and 1-PR-1-2 C. 1-PI-1-1C and 1-PR-1-2 D. 1-Pl-1-5 and 1-PI-1-1C Wednesday, June 05, 2013 8:16:21 AM                                                      186
 
1305 NRC RO Exam Feedback DISTRACTOR ANAL YSIS:
A. Correct, Inaccordance with the guidance in EPM-4, Users Guide, sect 3.6.2, the Category 1, Post Accident Monitoring (PAM) instrumentation is identified by a black background with white lettering and a Cl/n the corner of the label.
B. Incorrect, Plausible since this is an indicator on the main control board for SG#l pressure, however it is not desgina ted as a PAM instrument.
C. lncorrect, Plausible since this indication is on the remote shutdown paneI however it is not designated as a PAM instrument.
D. Incorrect, Plausible since this is a PAM instrument on the main control board, however it is a catergory 2 instrument which is designated with a white label with black lettering.
Wednesday, June 05, 2013 8:16:22 AM                                                            187
 
1305 NRC RO Exam Notes Question Number:          74 Tier:      3    Group      n/a K/A:      G 2.4.3 Ability to identify post-accident instrumentation Importance Rating:        3.7 / 3.9 10 CFR Part 55:          41.6 1OCFR55.43.b:            n/a K/A Match:      This question matches the K/A by having the candidates identify which of the pictured indications of SG#1 pressure is the one that is Post Accident Monitor instrumentation used for Steam Line Break indications.
Technical
 
==Reference:==
EPM-4, rev 22 sect 3.6.2 Post accident monitoring instrumentation Proposed references            None to be provided:
Learning Objective:            0PL271 EPM-4 obj 6 Identify Post-accident instrumentation and determin if its use is required.
Cognitive Level:
Higher                  X Lower Question Source:
New Modified Bank Bank                  X Question History:              New question developed for 1009 NRC exam Comments:                      Print on color copier and supply to exam Wednesday, June 05, 2013 8:16:22 AM                                                              188
 
1305 NRC RO Exam
: 75. G 2.4.9 075 Given the following plant conditions:
              -  Unit 1 reactor is shutdown.
              -  RCS in solid water operation.
              -  All RCS temperatures are approximately 125&deg;F.
              -  RCS pressure is 330 psig.
              -  RHR Train A is in service.
Subsequently:
              -  RCS pressure increases to 600 psig.
Which ONE of the following describes:
: 1)    the effect the pressure increase would have on the PORVs, AND
: 2)    the action(s) which is/are directed to be taken in accordance with AOP-R.03, RHR System Malfunction if RCS pressure is NOT promptly reduced below 600 psig?
REFERENCE PROVIDED A. Only ONE PORV OPENS              Stop the Charging pump.
B. Only ONE PORV OPENS              Stop A Train RHR Pump and ensure RHR suction valves are closed.
C. BOTH PORVS OPEN                  Stop the Charging pump.
D. BOTH PORVS OPEN                      Stop A Train RHR Pump and ensure RHR suction valves are closed.
Wednesday, June 05, 2013 8:16:22 AM                                                189
 
1305 NRC RO Exam Feedback DISTRACTOR ANAL YSIS:
A. Incorrect, The first part is incorrect. Both PORVs would lift at this pressure. This is plausible because LTOPS does have a staggered lift setpoint design feature (approximately 35 psi setpoint difference between the PORVs as shown in PTLR graph associated with LTOPS setpoints), thus the candidates could read the graph and determine that only one PORV would acuate at the given pressure. The second part is incorrect. AOP-R.03 ensures one charging pump is left running.
This is plausible since AOP-R.03 has the operator reduce charging flow which would be met by stopping the charging pump.
B. Incorrect, The first part is incorrect. Both PORVs would lift at this pressure. This is plausible because LTOPS does have a staggered lift setpoint design feature (approximately 35 psi setpoint difference between the PORVs as shown in PTLR graph associated with LTOPS setpoints), thus the candidates could read the graph and determine that only one PORV would acuate at the given pressure. The second part is correct. AOP-R.03 CAUTION prior to step 1 and step 2 RNO directs that RHR pump be stopped and step 2 RNO directs that RHR isolated from the RCS.
C. IncorrecI, The first part is correct. Both PORVs would lift at this pressure as noted on PTLR graph associated wtth LTOPS setpoint). The second part is incorrect.
AOP-R.03 ensures one charging pump is left running. This is plausible since AOP-R.03 has the operator reduce charging flow which would be met by stopping the charging pump.
D. Correct, The first part is correct. Both PORVs would lift at this pressure as noted on PTLR graph associated wtih LTOPS setpoint). The second part is correct.
AOP-R.03 CAUTION prior to step 1 and step 2 RNO directs that RHR pump be stopped and step 2 RNO directs that RHR isolated from the RCS.
Wednesday, June 05, 2013 8:16:22 AM                                                                190
 
1305 NRC RO Exam Notes Question Number:          75 Tier:    3      Group      na K/A:    G 2.4.9 Knowledge of low power/shutdown implications in accident (e.g.,
loss of coolant accident or loss of residual heat removal) mitigation stategies.
Importance Rating:        3.8 / 4.2 1OCFRPart55:            41.10 1OCFR55.43.b:          Not applicable K/A Match:      This question matches the K/A by having the candidate determine the mitigation strategy for reducing the RCS pressure and minimizing the potential for a shutdown LOCA by isolating the low pressure (RHR) system from the high pressure system during an overpressure event while in Mode 5 Technical
 
==Reference:==
AOP-R.03, RHR System Malfunctions, rev 30 Pressure-Temperature Limits Report, rev 4 pg 10 & 11 Proposed references          Pressure-Temperature Limits Report, rev 4 pg 11 to be provided:
Learning Objective:          OPL271AOP-R.03 Obj 12 List any condition(s) that require a RHR pump trip in AOP-R.03 Question Source:
New Modified Bank Bank                  X Question History:            SQN bank question Comments:
You have completed the test!
Wednesday, June 05, 20138:16:22 AM                                                          191}}

Revision as of 13:11, 4 November 2019

301 Draft RO Written Exam
ML13297A318
Person / Time
Site: Sequoyah  Tennessee Valley Authority icon.png
Issue date: 10/18/2013
From:
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To:
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References
Download: ML13297A318 (191)
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Text

Name: 1305 NRC RO Exam Form: 0 Version: 0

1. 007 EK1.02 001 Given the following plant conditions:

- Unit 1 is operating at 100% at EOL (15,000 MWD/MTU)

- Subsequently a reactor trip occurs on Unit 1.

- ES-O.1, Reactor Trip Response, has been implemented.

- Shutdown Bank B rod G3 remains at 228 steps.

- Control Bank D rod M8 sticks at 10 steps while inserting.

- Tavg dropped to 539°F before stabilizing.

Which ONE of the following completes the statement below?

Conditions indicate Emergency Boration is to satisfy Shutdown Margin.

A. NOT required B required due to the RCS temperature only C. required due to the stuck control rods only D. required due to both the RCS temperature and the stuck control rods Feedback DISTRACTOR ANAL YSIS:

A. lncorrect Plausible because if the RCS temperature had been greater than 540°F then emergency boratiuon would not have been required.

B. Correct, In accordance with ES-0. 1, Step 6b (See below) emergency boration is required due to the RCS coo/down being less than 540°F.

C. Incorrect, Plausible because if the RCS temperature had been greater than 540°F and the position of M8 rod had been greater than 12 steps then emergency boration would have been required due to the stuck rods.

0. Incorrect, Plausible because with the RCS temperature not greater than 540°F and the M8 rod had been greater more than 12 steps withdrawn, then emergency boration would have been required for both the excessive coo/down and 2 rods not being inserted Wednesday, June 05, 2013 8:16:13AM

1305 NRC RO Exam Rod position indicators Instrument Rack BTransferSwch less than or equal to 12 steps. to ALTERNATE. [M-7, lower switch]

IF any of the following conditions exists:

  • two or more RPIs indicate greater than 12 steps OR
  • two or more control rod positions CANNOT be determined, THEN EMERGENCY BORATE USING EA-68-4, Emergency Boration.
b. MONITOR RCS temperature: b. EMERGENCY BORATE as necessary to maintain shutdown
  • T-avg greater than 540cF margin USING EA-68-4. Emergency if any RCP running Boration.

OR

  • T-cold greater than 540°F if all RCPs stopped.

2 Wednesday, June 05, 2013 8:16:13 AM

____

___

_____

1305 NRC RD Exam Notes Question Number: 1 Tier: 1 Group 1 K/A: 007 Reactor Trip EK1 .02 Knowledge of the operational implications of the following concepts as they apply to the reactor trip:

Shutdown Margin Importance Rating: 3.4 / 3.8 1OCFRPart55: 41.8/41.10 1OCFR55.43.b: Not applicable K/A Match: This question matches the K/A by having the candidate recall the operational effect on shutdown margin of excessive cooldown of the RCS, a stuck control rod, and the failure of an additional rod to fully insert to 0 steps following a reactor trip on shutdown margin.

Technical

Reference:

ES-0.1, Reactor Trip Response, Revision 36 Proposed references None to be provided:

Learning Objective: 0PL271 ES-0. 1

6. Given the procedure and a set of initial plant conditions, determine actions required to mitigate the event in progress.

Question Source:

New Modified Bank X Bank Question History: WBN bank question 007 Eki .02 201 used on the WBN 10/2011 exam modified to make applicable to SQN and to make a different answer correct.

Comments:

3 Wednesday, June 05, 2013 8:16:14 AM

1305 NRC PC Exam

2. 008 AK2.01 002 Given the following plant conditions:

- A safety injection has occurred.

- RCS pressure is 1720 psig and still dropping.

- Pressurizer level is rising.

- All reactor coolant pumps are in operation.

Which one of the following identifies the leak location?

A Pressurizer safety valve B. Reactor Vessel Head vent line C. Lower pressurizer level tap D. Pressurizer heater well Feedback DIS TRACTOR ANAL YSIS:

A. Correct only a vapor space leak would result in the pressurizer level rising while the pressure continued to drop. Other leaks would result in the level dropping until the pressure stabilized or started to recover.

B. Incorrect, with the leak in this location, the pressure would not be continuing to drop if the pressurizer level were rising. Plausible because the pressurizer level could be rising with a leak on the vessel head vent if SI flow was greater than the leak flow.

C. lncorrect with the leak in the lower PZR level tap, the pressure would not be continuing to drop if the pressurizer level were rising. Plausible because the leak is in the pressurizer.

0. Incorrect, with the leak a PZR heater well, the pressure would not be continuing to drop if the pressurizer level were rising. Plausible because the leak is in the pressurizer.

4 Wednesday, June 05, 2013 8:16:14 AM

_____

_____

_____

_____

1305 NRC RO Exam Notes Question Number: 2 Tier: 1 Group 1 K/A: 008 AK2.01 Pressurizer (PZR) Vapor Space Accident (Relief Valve Stuck Open)

Knowledge of the interrelations between the Pressurizer Vapor Space Accident and the following:

Valves Importance Rating: 2.7* / 2.7 1OCFRPart55: 41.7/45.7 1OCFR55.43.b: Not applicable K/A Match: Question requires the knowledge of the interrelations of a pressurizer safety valve leak and conditions in the pressurizer during a vapor space accident Technical

Reference:

WOG E-1 Background document, Rev 2 WOG E-0 Background document, Rev 2 EGT200.713, TAA LOCAs

-

Proposed references None to be provided:

Learning Objective: EGT200.71 3, TAA LOCAs

-

  1. 11 Describe the dynamic behavior of the reactor, from a thermodynamic and hydrolic point of view, following a LOCA for each of the following catagories:
c. Vapor Space Break Cognitive Level:

Higher X Lower Question Source:

New Modified Bank Bank X Question History: Watts Bar bank question question 008 AK2.01 002 used on the WBN 08/2010 exam Comments:

5 Wednesday, June 05, 2013 8:16:14 AM

1305 NRC RO Exam

3. 009 EKI.02 003 Given the following plant conditions:

- In response to a small break LOCA, the crew is performing ES-i .2, Post LOCA Cooldown and Depressurization.

- The next step is to depressurize the RCS to refill the pressurizer.

- Core Exit Temperature is 546°F and lowering.

- RCS Tavg is 531 °F and lowering.

- RCS wide range pressure is 1520 psig.

- RCPs have been removed from service.

Which ONE of the following identifies the current RCS subcooling margin and the operational impact if subcooling is lost during the depressurization?

A. 53°F; The RCS cooldown will stop.

B. 68°F; The RCS cooldown will stop.

C 53°F; Cause rapid increase in Pressurizer level.

D. 68°F; Cause rapid increase in Pressurizer level.

Wednesday, June 05, 2013 8:16:14 AM 6

1305 NRC RO Exam Feedback DIS TRACTOR ANAL YSIS:

A. Incorrect, RCS subcooling is 53°F is correct, but the loss would not prevent the initiated RCS coo/down from the previous step. Plausible because 53°F subcooling is correct and because natural circulation could be affected by steam voids and the presence of head voiding would affect natural circulation cooling but not stop it. In addition, with the sudden pressure drop (RCS pressure 1000 psig) that caused subcooling to go to zero, ECCS flow is going to increase to the point that it will cool the RCS rapidly.

B. Incorrect, RCS subcooling is not 68°F (it is 53°F) Plausible because 68°F could be calculated if the Tavg were used instead of Core Exit Thermocouple temperature.

599°F 531°F = 68°F and because natural circulation could be affected by steam

-

voids and the presence of head voiding would affect natural circulation cooling but not stop it. In addition, with the sudden pressure drop (RCS pressure 1000 psig) that caused subcooling to go to zero, ECCS flow is going to increase to the point that it will cool the RCS rapidly.

C. Correct, with core exit temperature 546°F and saturation for 1520 psig (1535 psia) being 599°F, the RCS subcooling is 53°F. The loss of subcooling could result in upper head voiding causing a rapid increase in pressurizer level as stated in the caution preceding the procedure step.

D. Incorrect, RCS subcooling is not 68°F (it is 53°F) and the loss would not prevent the initia ted RCS cooldown from the previous step. Plausible because 68°F could be calculated if the Tavg were used instead of Core Exit Thermocouple temperature. 599°F 531°F = 68°F and because the rapid increase in pressurizer

-

level is correct.

Wednesday, June 05, 2013 8:16:14 AM 7

_

_____

_____

1305 NRC RO Exam Notes Question Number: 3 Tier: 1 Group 1 K/A: 009 Small Break LOCA EK1 .02 Knowledge of the operational implications of the following concepts as they apply to the small break LOCA:

Use of steam tables Importance Rating: 3.5 / 4.2 1OCFRPart55: 41.8/41.10/45.3 1OCFR55.43.b: Not applicable K/A Match: K/A is matched because the questions requires knowledge of how to use steam tables to determine subcooling and the implication of how the the loss of subcooling could reult in or explain conditions of the unit.

Technical

Reference:

Steam Tables, ALSTOM Power Services ES-i .2, Post LOCA Cooldown and Depressurization, Rev 17 Proposed references Steam Tables to be provided:

Learning Objective: 0PL271 ES-i .2

  1. 4 Summarize the mitigating strategy for ES-i .2.
  1. 5 Describe the basis of all limits, notes and cautions.

Question Source:

New Modified Bank Bank X Question History: SON Bank Question 009 EK1 .02 002 used on the SON 1/2009 RETAKE exam Comments:

Qu Wednesday, June 05, 2013 8:16:14 AM 8

1305 NRC RO Exam

4. 011 EG2.1.31 004 Given the following plant conditions:

- Unit 1 is initially at 100% power when an event occurred.

- The crew has just restored power to FCV-63-1 at step 6 of ES-i .3, Transfer to RHR Containment Sump.

- RWST level is 20% and lowering.

- CNTMT pressure is 1 .5 psig and lowering.

- CNTMT sump level is 41% and rising.

- i-HS-63-72A, CNTMT Sump SuctTo RHR Pump 1A, Red light LIT

- 1 -HS-63-73A, CNTMT Sump Suct To RHR Pump 1 B, Red light LIT

- 1-HS-72-39A, CNTMT Spray Hdr 1A Isol, Green light LIT

- i-HS-72-2A, CNTMT Spray Hdr lB Isol, Green light LIT Which of the following correctly complete the statement below?

Based on the above indications the RHR CNTMT Sump suction valves (1) in the expected positions and the CNTMT Spray Header isolations (2) in the expected positions.

A (1) are (2) are B. (1)are (2) are NOT C. (1) are NOT (2) are D. (1) are NOT (2) are NOT Wednesday, June 05, 2013 8:16:14 AM 9

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1305 NRC RO Exam Feedback DISTRACTOR ANAL YSIS:

A. Correct, With RWST level less than 27% the RHR suction auto switchover should have occurred and the 72 and 73 valves should be open and is normal. The CNTMT Spray Hdr discharge valves will be open as well with CNTMT pressure above the phase B setpoint and CSPs running. At this point only one CSP would be running, however its discharge valve is left open. The discharge valves would have been closed if CNTMT pressure had dropped to less than 2 psig and the CSPs secured. This action would have occurred by the time the RWST level had went from 27% to 20%.

B. Incorrect, The first part is correct. The second part is incorrect as both CS discharge Hdr valves would be open. Plausible as the second step in ES-1.3 secures one spray pump and it is logical to shut the header isolation valve for that pump. In fact the RNO for this step secures both CSPs and closes the Discharge Hdr Isolation valves.

C. Incorrect, The RHR Sump suction valves will be open on the auto switchover on RWST level <27%. This is plausible as other ECCS suction valves do not auto switch over (SIPs). The second part is correct.

D. Inorrect, The RHR Sump suction valves will be open on the auto switchover on RWST level <27%. This is plausible as other ECCS suction valves do not auto switch over (SIPs). The second part is correct. The second part is incorrect as both CS discharge Hdr valves would be open. Plausible as the second step in ES- 1.3 secures one spray pump and it is logical to shut the header isolation valve for that pump. In fact the RNO for this step secures both CSPs and closes the Discharge Hdr Isolation valves.

Wednesday, June 05, 2013 8:16:14AM 10

1305 NRC AC Exam Notes Question Number: 4 Tier: 1 Group 1 K/A: 011 Large Break LOCA EG2.1 .31 Ability to locate control room switches, controls, and indications and to determine that they are correctly reflecting the desired plant lineup.

Importance Rating: 4.6 / 4.3 10 CFR Part 55: 41.10 1OCFR55.43.b: Not applicable K/A Match: Applicant is required assess indications and determine if the correct lineup exist for current plant conditins during a LB LOCA.

Technical

Reference:

ES-1.3, Transfer to RHR Containment Sump, Rev 19, 1-AR-M6-E, E-3 R23 Proposed references None to be provided:

Learning Objective: OPT200.ECCS

  1. 1 Describe the purpose and/or functions of the ECCS amd subsystems, and major components:
c. valves automatically operated upon SUI actuation 0PL271 ES-i .3 B.5.6 a&b Given a set of initial plant conditions use ES-i .3 to correctly:
a. Identify required actions
b. Respond to Contigencies Question Source:

New X Modified Bank Bank Question History: New for NRC ILT 1305 Exam Comments:

Wednesday, June 05, 2013 8:16:14 AM 11

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1305 NRC RO Exam 5.015 AG2.I.28005 Given the following plant conditions:

- Unit 2 is operating at 100% power.

- FS-62-1 1 REAC COOL PMPS SEAL LEAKOFF HIGH FLOW alarm is LIT.

- LS-62-45A REAC COOL PMP 4 STANDPIPE LVL HIGH-LOW alarm is LIT

- No. 1 seal leakoff flow recorder for RCP #4 indicates 7 gpm.

Which ONE of the following completes the statements below:

(1) The #2 seal on RCP #4 becomes a in response to these conditions.

(2) In accordance with AOP-R.04, Reactor Coolant Pump Malfunctions, the minimum No. 1 Seal Leakoff flow that would require a reactor trip and shutdown of

  1. 4RCPis {

A (1) film riding seal (2) 8 gpm

0. (1) film riding seal (2) 9 gpm C. (1) rubbing face seal (2) 8 gpm D. (1) rubbing face seal (2) 9gpm Wednesday, June 05, 2013 8:16:14AM 12

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1305 NRC RO Exam Feedback DISTRACTOR ANAL YSIS:

A. Correct, According to plant design, the #2 seal in the RCP seal package is a pressure balanced and spring loaded rubbing face type seal. Upon a loss of the #1 seal, full RCS pressure is applied to the #2 seal. When applied full pressure, #2 seal designed to convert from a rubbing-face type seal to a film-riding seal. Also in accordance with AOP-R.04, if an RCP experiences high seal leak-off flow which is greater than 6 gpm but less than 8 gpm, the operators are directed to ensure that a minimum of 9 gpm seal injection flow is maintained to the affected seal package until the RCP can be shutdown.

B. Incorrect, Plausible since the first part is correct, the #2 seal package is normally a face rubbing type seal during normal operation and converts to a film rubbing type seal during times of a failure of #1 seal. Also plausible if the candidate gets confused on the minimum seal injection flow and thinks that the normal minimum of 6 gpm is adequate for these conditions.

C. Incorrect, Plausible if the candidates gets confused on the design of the RCP seals and determines that the #2 seal converts from a film riding seal (which is the design of #1 seal) to a rubbing face type seal when full RCS pressure is applied. Also the second part is correct, if high leak off flow is seen on an RCP seal package, AOP-R.04 directs the operators to maintain a minimum of 9 gpm of seal injection flow to any RCP with a seal leak-off flow of greater than 6 gpm but less than 8 gpm.

D. Incorrect, Plausible if the candidates gets confused on the design of the RCP seals and determines that the #2 seal converts from a film riding seal (which is the design of #1 seal) to a rubbing face type seal when full RCS pressure is applied. Also plausible if the candidate gets confused on the minimum seal injection flow and thinks that the normal minimum of 6 gpm is adequate for these conditions.

Wednesday, June 05, 2013 8:16:14 AM 13

1305 NRC RO Exam Notes Question Number: 5 Tier: 1 Group 1 K/A: 015 Reactor Coolant Pump (RCP) Malfunctions AG 2.1 .28 Knowledge of the purpose and function of major system components and controls.

Importance Rating: 4.1 /4.1 10 CFR Part 55: 41.7 1OCFR55.43.b: Not applicable K/A Match: This question matches the K/A by having the candidate identify RCP seal design that mitigates an RCP seal malfunction and the operator actions necessary to minimize the effect of a failed RCP seal.

Technical

Reference:

AOP-R.04, Reactor Coolant Pump Malfunctions OPT200.RCP STG rev 8 Proposed references None to be provided:

Learning Objective: OPT200. RCP obj:

1. Describe the purpose and/or functions of the Reactor Coolant Pumps (RCPs) subsystems, and the major system components listed below:
g. Shaft seals
3. Given plant conditions, Determine if any on the following RCP alarms would be present and Describe actions required by the ARP:
h. FS-62-1 1, Reac Cool PMP5 Shaft Seal Leakoff High Flow.
j. RCP 1 (2) (3) (4) Standpipe Level HI/Lo Question Source:

New X Modified Bank Bank Question History: New question written for 1305 NRC exam Comments:

Wednesday, June 05, 2013 8:16:14AM 14

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1305 NRC RO Exam

6. 022 AK3.05 006 Given the following plant conditions:

- Due to indications of gas binding of the 1A CCP, the operators have entered AOP-M.09, Loss of Charging and secured the pump.

Which ONE of the following identifies the reason that plant power should remain stable?

A If a power reduction is required, a manual trip must be initiated due to the inability to borate the RCS.

B. If a power reduction is required the rate of change would be limited to being lowered at no greater than 2% per minute in accordance with AOP-C.03, Rapid Shutdown or Load Reduction using control rods.

C. Reactor power should remain stable until a CCP can be started for RCP seal injection to prevent possible damage to RCP seals caused by changes in temperature.

D. Reactor power should remain stable because the crew will not be able to borate to compensate for the initial Xenon effects.

Feedback DIS TRACTOR ANAL YSIS:

A. Correct, In accordance with AQP-M.09, Note prior to step 21 If any condition in the following step is met, the reactor will be tripped. Manual reactor trip is used to shut down the reactor due to inability to borate. Any condition which would cause a plant transient should be avoided so as to prevent the need for a reactor trip.

B. Incorrec1, Plausible as this is a caution in AOP-C.03, Rapid Shutdown or Load Rejection, for loss of normal and emergency boration. Shutdown is still allowed using the RWST if the CCPs where still available. It is not until step 6 this procedure that you cannot continue the shutdown until you can get some kind of boration going to the RCS.

C. Incorrect, Plausible as this is a concern and a caution just prior to step 6 in AOP-M.09. In this case however in this case the RCP seals should be able to stay below their high temperature limits with CCS still running. The high RCP temperature steps in M.09 are trip criteria and not reasons to keep power stable.

D. Incorrect, Plausible as there will be xenon effects that the crew would be challenged to control with just rods and having no boration. However, the reason is not being able to keep up with power defect reactivity effects as turbine load is reduced.

Wednesday, June 05, 2013 8:16:14 AM 15

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1305 NRC RO Exam Question Number: 6 Tier: 1 Group 1 K/A: 022 Loss of Reactor Coolant Makeup AK3.05 Knowledge of the reasons for the following responses as they apply to Loss of Reactor Coolant Makeup:

Need to avoid plant transients Importance Rating: 3.2 / 3.4 10 CFR Part 55: 41.5, 41.10 1OCFR55.43.b: Not applicable K/A Match: This question matches the KA as it requires the examinee to address a loss of reactor coolant makeup and determine the reason to avoid a plant downpower.

Technical

Reference:

AOP-M.09, Loss Of Charging R5 AOPC.03, Rapid Shutdown or Load Reduction R27 Proposed references None to be provided:

Learning Objective: OPL271AOP-M.09 #s 4 & 8 Question Source:

New X Modified Bank Bank Question History: New question written for 1305 ILT exam Comments:

Wednesday, June 05, 2013 8:16:14AM 16

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1305 NRC RO Exam

7. 025 AK2.05 007 Given the following plant conditions:

- Unit 1 was operating at 100% power when a LOCA occurred.

- Both trains of ECCS pump suction have been transferred to the containment sump in accordance with ES-i .3, Transfer to RHR Containment Sump.

- The 1 B RHR pump beings to cavitate.

- Containmnt pressure is 4.1 psig and slowly trending down.

Which ONE of the following completes the statements below?

The conditions above require the operating crew to ensure jJ are stopped and placed to Pull to Lock.

After stopping required pumps in (1) above, if 1A RHR pump starts cavitating, t?L are required to be stopped.

A. (1) only the 1 B RHR pump and one Cntmt Spray pump (2) all running ECCS and Containment Spray pumps B. (1) only the 1 B RHR pump and one Cntmt Spray pump (2) only all running ECCS pumps C (1) 1 B RHR, one CCP, one SI pump and one Cntmt Spray pump (2) all running ECCS and Containment Spray pumps D. (1) lB RHR, one CCP, one SI pump and one Cntmt Spray pump (2) only all running ECCS pumps Wednesday, June 05, 2013 8:16:14 AM 17

1305 NRC RO Exam Feedback DISTRACTOR ANAL YSIS:

A. Incorrect: Incorrect as in addition to the RHR and CSP, the SI and CCP on that train are secured too per ES-1.3. This is plausible as there are places within ES-1.3 where the suction to the SI and CPPs are from the RWST with the RHR pumps from the sump and unaffected by the loss of a RHR pump. In addition, shutting down the same train CPP and SIP logically would be dependent on the pumps themselves cavitating verses shutting down the RHR pump. The second part is correct.

B. Incorrect: Incorrect as in addition to the RHR and CSP, the SI and CCP on that train are secured too per ES- 1.3. This is plausible as there are places within ES-1.3 where the suction to the SI and CPPs are from the RWST with the RHR pumps from the sump and unaffected by the loss of a RHR pump. In addition, shutting down the same train CPP and SIP logically would be dependent on the pumps themselves cavitating verses shutting down the RHR pump. The second part is plausible as the CSP is directed to be secured only if it is cavitating in ES- 1.3, step 23 RNO second IF/THEN step. Once the transition is made to ECA-1.3 the CSP is directed to be stopped. A conclusion could be made that only the ECCS pumps should be secured.

C. Correct: Per ES-1.3 step 23, if a RHR pump starts cavitating the affected train of ECCS pumps and the CSP on that train are placed in PTL. Per the RNO of that step, once the second RHR pump cavitates and is stopped, a transition is made to ECA-1.3, CNTMT Sump Blockage and all ECCS pumps and CSPs are secured.

These actions could also be applied using step 23 again.

D. Incorrect: The first part is correct. The second part is plausible as the CSP is directed to be secured only if it is cavitating in ES- 1.3, step 23 RNO second IF/THEN step. Once the transition is made to ECA-1.3 the CSP is directed to be stopped. A conclusion could be made that only the ECCS pumps should be secured.

Wednesday, June 05, 2013 8:16:14 AM 18

1305 NRC PC Exam Notes Question Number: 7 Tier: 1 Group 1 K/A: APE 025 Loss of Residual Heat Removal System (RHRS)

AK2.05 Knowledge of the interrelations between the Loss of Residual Heat Removal System and the following, Reactor building sump Importance Rating: 2.6 / 2.6 10 CFR Part 55: 41.7 / 45.7 1OCFR55.43.b: Not applicable K/A Match: This question matches the K/A by having the candidate determine how the level of water in the Containment building sump affects the ability of RHR pumps to draw water for supply to the ECCS pumps during recirculation.

Technical

Reference:

ES-i .3, Transfer RHR to CNTMT Sump P19 ECA-1 .3, CNTMT Sump Blockage R2 Proposed references None to be provided:

Learning Objective: 0PL271 ES-i .3,

  1. 4 Summarize the mitigating strategy for ES-i .3 0PL271 ECA-1 .3,
  1. 4 Summarize the mitigating strategy for ECA-1 .3 Question Source:

New Modified Bank Bank X Question History: SON bank question 025 AK2.05 007 used on the SQN 1/2008 exam with wording changes in the stem.

Comments:

Wednesday, June 05, 20138:16:14 AM 19

1305 NRC RO Exam 8.026 AAI.06 008 Given the following:

- Unit 1 is in Mode 3 with RCS cooldown in progress.

- Unit 2 is operating at 100% power.

- Spent Fuel Pool Cooling is being supplied from Unit 1 Train A CCS.

- A leak upstream of the 1A CCS heat exchangers required all Unit 1 Train A CCS to be shutdown.

Which ONE of the following completes the statement below?

The cooling supply to the Spent Fuel Cooling System will be realigned to be supplied from fl.

After CCS is restored to the in-service Spent Fuel Pool Heat Exchanger, if the CCS flow rate is 3400 gpm through the heat exchanger, the flow rate is required to be { in accordance with 0-SO-78-1, Spent Fuel Pit Coolant System.

LII A. Unit 1 CCS Train B lowered B. Unit 1 CCS Train B raised C Unit 2 CCS Train A lowered D. Unit 2 CCS Train A raised Wednesday, June 05, 2013 8:16:14 AM 20

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1305 NRC_RO_Exam Feedback DISTRACTOR ANAL YSIS:

A. Incorrect, Plausible because a leak in any A header of a coo/mg system would be replaced by the B header. This is incorrect as the supply is shifted to the unaffected units A CCS header. The second part is correct.

B. Incorrect, Plausible because a leak in any A header of a cooling system would be replaced by the B header. This is incorrect as the supply is shifted to the unaffected units A CCS header. The second part is incorrect as f/ow is too high and will need to be lowered. Plausible as flowrate could change with the shift to the new header and 3400 is close to the max flow of 3300.

C. Correct The supply will be realigned to the Unit 2 Train A header and 3400 gpm is above the maximum allowed CCS flow rate of 3300 and the SO directs the throttling of 0-FCV 1 1(handswitch on O-M-27B) to adjust and control the flow rate.

D. Incorrect, The first part is correct. The second part is plausible as flowrate could change with the shift to the new header and 3400 is close to the max flow of 3300.

Wednesday, June 05, 2013 8:16:14 AM 21

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1305 NRC RO Exam Notes Question Number: 8 Tier: 1 Group 1 K/A: 026 Loss of Component Cooling Water (CCW)

AA1 .06 Ability to operate and / or monitor the following as they apply to the Loss of Component Cooling Water:

Control of flow rates to components cooled by the CCWS Importance Rating: 2.9 / 2.9 10 CFR Part 55: 41 .7 I 45.5 / 45.6 1OCFR55.43.b: Not applicable K/A Match: K/A is matched because the question requires applicant to operate the CCS system and control flow rates of the CCS to the SEP HX.

Technical

Reference:

SO-78-1, Spent Euel Cooling System, Revision 0059 EA-70-1, CCS Operation R4 Proposed references None to be provided:

Learning Objective: OPT200.CCS

  1. 10 Given specific plant conditions, ANALYSE the effect taht a loss of CCS will have on the following:
a. All loads cooled by the CCS Question Source:

New X Modified Bank Bank Question History: New question for the SQN 05/2013 exam Comments:

Wednesday, June 05, 2013 8:16:14 AM 22

1305 NRC RO Exam

9. 029 EAI.09 009 Given the following:

- Unit 2 is operating at 100% power when a turbine trip occurs.

- Control rods begin inserting with 2-HS-85-51 10, ROD CONTROL MODE SELECTOR, in AUTO.

- Sl-412, ROD SPEED, indicates 72 steps/mm.

- The reactor fails to trip and cannot be tripped from the MCR reactor trip handswitches.

- The crew enters the EOP network and has transitioned to FR-S.1, Nuclear Power Generation / ATWS.

Which ONE of the following completes the statement below in accordance with FR-S.1?

FR-Si A. allows ROD CONTROL to remain in AUTO until the rod insertion rate first drops to less than 64 steps/mm before requiring manual rod insertion By allows ROD CONTROL to remain in AUTO until the rod insertion rate first drops to less than 48 steps/rn in before requiring manual rod insertion C. requires ROD CONTROL to be immediately placed in MAN and insertion continued resulting in the ROD SPEED indicating 64 steps/mm D. requires ROD CONTROL to be immediately placed in MAN and insertion continued resulting in the ROD SPEED indicating 48 steps/mm Wednesday, June 05, 2013 8:16:15 AM 23

1305 NRC RO Exam Feedback DIS TRACTOR ANAL YSIS:

A. Incorrect, allowing Rod control to remain in AUTO is correct but 64 steps/mm is not the minimum speed allowing AUTO operation. Plausible because 64 steps/mm is the speed of the shutdown rod movement.

B. Correct, FR-S. 1 Step 1 RNO states iF reactor trip breakers will NOT open, THEN MAINTAIN auto or manua rod insertion at max achievable rate UNTIL rods are at bottom. When Rod Control is in manual the speed is set at 48 steps/mm.

As long as rod are moving at a sped above the manual speed, ROD control can be left in AUTO.

C. lncorrect, plausible as FR-S. 1 does allow going to manual contro4 but does not require it. The second part is plausible because 64 steps/mm is the speed of shutdown rod movement.

D. Incorrect, plausible as FR-S. 1 does allow going to manual contro4 but does not require it. The second part is correct.

Wednesday, June 05, 2013 8:16:15 AM 24

1305 NRC RO Exam Notes Question Number: 9 Tier: 1 Group 1 K/A: 029 Anticipated Transient Without Scram (ATWS)

EA1 .09 Ability to operate and monitor the following as they apply to a ATWS:

Manual rod control Importance Rating: 4.0 / 3.6 10 CFR Part 55: 41.7/45.5 / 45.6 1OCFR55.43.b: Not applicable K/A Match: Question requires the operation and monitor of control rod insertion during an ATWS event Technical

Reference:

FR-S.1, Nuclear Power Generation / ATWS, Revision 23 Proposed references None to be provided:

Learning Objective: 0PL271 FR-S.1

3. Given a set of initial plant conditions, determine initial Operator response to stabilize the plant, including applicable Immediate Actions of FR-S.1.

Question Source:

New X Modified Bank Bank Question History: New question for the SQN 05/2013 exam Comments:

Wednesday, June 05, 2013 8:16:15 AM 25

1305 NRC RO Exam

10. 040 AA2.02 010 Given the following:

Unit 1 is operating at 67% power steady state conditions with Rod Control

-

in Manual.

- A transient occurs resulting in the following:

Rx Power Turb Power Tavci RCS Press MWe 0700 67% 67% 567°F 2235 psig 785 MWe 0701 68% 66.5% 565°F 2231 psig 780 MWe 0702 69% 66.5% 563°F 2227 psig 770 MWe 0703 70% 65.5% 561°F 2223 psig 765 MWe 0704 71% 65.5% 560°F 2219 psig 760 MWe Which ONE of the following completes the statement below?

The earliest time the conditions require a reactor trip to be initiated in accordance with AOP-S.05, Steam or Feedwater Leak, is A. 0701 B. 0702 C 0703 D. 0704 Feedback DISTRACTOR ANAL YSIS:

A. Incorrect, Plausible because the conditions indicate an uncontrolled rise in reactor power. This is incorrect as the trip criteria for this is the main turbine off line.

B. Incorrect, Plausible as the difference between turbine and reactor power is close to 3%, which is close to 35 Mwe (30 Mwe) and is another set of trip criteria.

C. Correct, a reactor power rise of 3% is the first time trip criteria is met.

0. Incorrect, Plausible because the MWe has changed by greater than 35 MWe and there are conditions in AQP-S.05 where MWe changing by greater than 35 MWe requires a reactor trip is correct. Incorrect as this is not where trip criteria is met at the earliest.

Wednesday, June 05, 2013 8:16:15 AM 26

1305 NRC RO Exam Notes Question Number: 10 Tier: 1 Group 2 K/A: 040 Steam Line Rupture AA2.02 Ability to determine and interpret the following as they apply to the Steam Line Rupture:

Conditions requiring a reactor trip Importance Rating: 4.6 / 4.7 10 CFR Part 55: 43.5 /45.13 1OCFR55.43.b: n/a K/A Match: Question match the KA because the applicant is required to interpret data to determine the conditions that require the reactor trip.

Technical

Reference:

AOP-S.05, Steam or Feedwater Leak, Revision 12 AOP-C-.02, Uncontrolled RCS Boron Concentration Changes, Revision 0008 Proposed references None to be provided:

Learning Objective: 0PL271 AOP-S.05

9. List any condition(s) that require a Reactor trip, Turbine Trip or Safety Injection in AOP-S.05.

Question Source:

New X Modified Bank Bank Question History: New question written for SQN 05/2013 NRC exam.

Comments:

Wednesday, June 05, 2013 8:16:15 AM 27

1305 NRC RO Exam 11.054 AK1.01 011 Given the following plant conditions:

- Unit2isatl00%RTP.

- SIG #4 main feedwater line develops a leak inside containment at the containment penetration wall.

- The S/G #4 level is able to be maintained with increased feedwater flow.

- Containment pressure is beginning to rise.

- Condenser hotwell level is 20 and lowering slowly.

- The operating crew enters AOP-S.05, Steam or Feedwater Leak.

- While performing AOP-S.05, the AOP directs the operating crew to trip the reactor, SI and close the MSIVs.

Which ONE of the following completes the statements below?

The AOP-S.05 criteria used to trip the reactor, SI and close the MSIVs is due to After the MSIVs are closed, SG #4 pressure will (2) drop uncontrolled.

A. imminent loss of hotwell level will B. imminent loss of hotwell level will NOT C containment pressure will D. containment pressure will NOT Wednesday, June 05, 2013 8:16:15 AM 28

1305 NRC RO Exam Feedback DISTRACTOR ANAL YSIS:

A. Incorrect, Plausible as imminent loss of hotwell is trip criteria, hotwell level is lowering, but at 20 it has 8 more before the alarm is received. The second part is correct.

B. Incorrect, Plausible as imminent loss of hotwell is trip criteria, hotwell level is lowering, but at 20 it has 8 more before the alarm is received. The second part is plausible as some systems have check valves inside CTMT and would isolate RCS components from the leak. The leak would stop one the feed forward was isolated to that system.

C. Correct, The reactor will be tripped based on CTMT pressure approaching 1.5 psig.

The SG will blow down to containment due to the check valve being outside of CTMT.

D. Incorrect, The first part is correct. The second part is plausible as some systems have check valves inside CTMT and would isolate RCS components from the leak.

The leak would stop once the feed forward was isolated to that system.

Wednesday, June 05, 20138:16:15 AM 29

1305 NRC RO Exam Notes Question Number: 11 Tier: 1 Group 1 K/A: 054 Loss of Main Feedwater AK1 .01 Knowledge of the operational implications of the following concepts as they apply to Loss of Main Feedwater (MEW):

MEW line break depressurizes the SIG (similar to a steam line break)

Importance Rating: 4.1 / 4.3 1OCFRPart55: 41.8/41.10/45.3 1 OCFR55.43.b: Not applicable K/A Match: The question matches the KA as it requires the examinee to understand the implications of a MEW leak and the status of the affected SG.

Technical

Reference:

AOP-S.05, Steam or Feedwater Leak, Revision 12 Proposed references None to be provided:

Learning Objective: 0PL271 AOPS.05

  1. 8 Given a set of initial plant conditions, determine the most likely location of a Steam Line Rupture.
  1. 9 List any condition(s) that require a reactor trip, turbine trip or SI in AOP-S.05 Question Source:

New Modified Bank X Bank Question History: Significantly modified SQN bank question AOP-S.05-B.1 which was used on the ILT 1002 Audit Exam.

Comments:

Wednesday, June 05, 20138:16:15 AM 30

1305 NRC RO Exam

12. 056 G2.4.4 012 Given the following plant conditions:

- Unit 1 is operating at 100% rated thermal power when a reactor trip occurs

- The following annunicators are observed in alarm:

M26-A, A-5, Diesel GEN lA-A Running > 40 RPM.

M26-B, A-5, Diesel GEN lB-B Running >40 RPM.

M26-C, A-5, Diesel GEN 2A-A Running > 40 RPM.

M26-D, A-5, Diesel GEN 2B-B Running > 40 RPM.

Mi-B, B-2, 6900V Unit BD lB Failure Or Undervoltage.

M26-A, C-7,6900V SD BD iA-A Failure or Bus Feeder UV alarms and clears.

Which ONE of the following completes the statement below?

The crew would respond by performing (1) in parallel with applicable EOPs, and the Emergency Diesel status is (2)

Av (1) AOP-P.Oi, Loss of Offsite Power (2) normal B. (1) AOP-P.01, Loss of Offsite Power (2) abnormal C. (1) AOP-P.05, Loss of Unit 1 Shutdown Boards (2) normal D. (1) AOP-P.05, Loss of Unit 1 Shutdown Boards (2) abnormal Wednesday, June 05, 2013 8:16:15 AM 31

1305 NRC RO Exam Feedback DISTRACTOR ANAL YSIS:

A. Correct, If a shutdown board loses its normal power supply and is re-energized by its respective diesel it is considered a partial loss of offsite power and meets the entry criteria for AOP-P.O1. Per EPM-4, the crew would decide to perform AOP-P.O1 in parallel with E-O (typically the AQP would be assigned to one crew member as reader doer). All four diesels start on a UV condition on one shutdown board.

B. Incorrect, The first part is correct. The second part is plausible as it would be logical for a Unit 1 SOB UV condition to start only Unit 1 DGs and the conditions in the stem would be bnormal.

C. Incorrect. The first part is plausible as a loss of the shutdown board would be entry criteria for AOP-P.05. In facI, there is a directed transition from AOP-P.O1 to AOPO5 if the board remains de-energized. The second part is correct.

0. lncorrect The first part is plausible as a loss of the shutdown board would be entry criteria for AOP-P.05. In fact, there is a directed transition from AOP-P.O1 to AOPO5 if the board remains de-energized. The second part is plausible as it would be logical for a Unit 1 SOB UV condition to start only Unit 1 DGs and the conditions in the stem would be bnormaI.

Wednesday, June 05, 2013 8:16:15 AM 32

1305 NRC RO Exam Notes Question Number: 12 Tier: 1 Group 1 K/A: 056 Loss of Offsite Power G2.4.4 Ability to recognize abnormal indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures.

Importance Rating: 4.5/4.7 10 CFR Part 55: (CFR: 41.10/ 43.2 / 45.6) 1OCFR55.43.b: Not applicable K/A Match: The examinee is given indications of a shutdown board being de-energized and subsequently re-energized by its associated DG with the UV alarm clearing. They then have to choose the correct procedure to use to mitigate the event.

Technical

Reference:

AOP-P.01 R30 AOP-P.05 R20 0-AR-M26-A A-5 R30 0-AR-M26-A 0-7 R30 1-AR-Mi-B, B-2 R23 EPM-4 R22 Proposed references None to be provided:

Learning Objective: OPL271AOP-P.01, # 2 Question Source:

New X Modified Bank Bank Question History: New for SQN ILT 1305 exam Comments: Low Cognitive Wednesday, June 05, 2013 8:16:15 AM 33

1305 NRC RO Exam

13. 062 AA2.01 013 Given the following plant conditions:

- Both Units are at 100% power.

- ERCW is in normal alignment.

- ERCW header 1A & 2A are indicating LOW flow.

The following MCR alarms are LIT:

- M-15A Window B-6, MECH EQUIP SUMP LVL HI.

- M-27A Window A-i, UNIT 1 HEADER A PRESSURE LOW.

- M-27A Window B-3, UNIT 2 HEADER A PRESSURE LOW.

- NO OTHER alarms are lit associated with the ERCW system.

Which ONE of the following ERCW conditions accounts for the above indications?

A. Supply header 1AI2A has ruptured in the Yard Area.

B. A discharge header has ruptured in the Yard Area.

C. A rupture has occurred upstream of the 2A strainer.

D. A rupture has occurred in the CCW Intake Pumping Station.

Feedback DIS TRACTOR ANAL YSIS:

A. lncorrecl, Plausible since these would also be indications of a supply header rupture, however there would be High system flow associated with this failure not Low flow.

B. Incorrect, Plausible since a pipe rupture would cause system pressure to do down, but a rupture is this location would be accompanied with high system flow, not low flow.

C. Correct, The diagnostic section (Section 2.1) of AOP-M.O1, Loss of ERCW, uses the annunciators and indications listed in the stem to indicate that a supply header has ruptured upstream of a train A supply strather. Since both Unit 1 and Unit 2 supply headers are cross connected upstream of the strainers a leak or rupture on one strainer will affect the other train.

D. lncorrect, Plausible since a pipe rupture would cause system pressure to do down, but a rupture is this location would be accompanied with high system flow, not low flow. The main ERCW headers go right through the CCW pumping station.

Wednesday, June 05, 20138:16:15 AM 34

1305 NRC RO Exam Notes Question Number: 13 Tier: 1 Group 1 K/A: 062 Loss of Nuclear Service Water AA2.01 Ability to determine and intepret the following as they apply to Loss of Nuclear Service Water:

Location of a leak in the SWS Importance Rating: 2.9 I 3.5 10 CFR Part 55: 41.7 1OCFR55.43.b: Not applicable K/A Match: Questions matches the KIA by requiring the candidate to interpret the indications presented and determine the location the a leak in the SWS (ERCW).

Technical

Reference:

AOP-M.01, Loss of ERCW, Revision 23 1 ,2-47W845-5 rev 55 1-AR-Mi 5-A R33 0-AR-M27-A R20 Proposed references None to be provided:

Learning Objective: 0PL271 .AOP-M.01

  1. 3 Given a set of initial plant conditions, determine initial operator response to stabilize the plant.

Question Source:

New X Modified Bank Bank Question History: New question for the SQN 05/2013 NRC exam.

Comments:

Wednesday, June 05, 2013 8:16:15 AM 35

1305 NRC RO Exam

14. 065 AK3.08 014 Given the following plant conditions:

- Unit 1 is in Mode 3 following a manual reactor trip required due to a control air line break in the non-essential control air header.

- The operating crew performed the applicable emergency instructions and has stabilized the plant.

- The crew has implemented AOP-M.02, Loss of Control Air, to address the loss of air.

Which ONE of the following identifies a local action that would be required in Auxiliary Building during the performance of AOP-M.02 and why?

A. Close 1 -FCV-32-1 10, Non Essential Air to the RX Bldg Ui lsol.to isolate non-essential loads.

B. Establish local control of SG PORVs to control RCS temperature.

C. Establish local control of AFW LCVs to prevent SG overfill.

D Adjust RCP seal injection flow to minimize PZR level rise.

Wednesday, June 05, 2013 8:16:15 AM 36

1305 NRC RO Exam Feedback The air system at SON is comprised of non-essential control air headers, two essential control air headers (also called aux air headers) and the seivice air header. During normal operations the service air compressors supply all the station air needs. During a air leak situation the essential control air headers isolate from the non-essential headers and have their own compressors (low capacity) that supply air for essential loads. For a loss of non-essential control air one of the biggest problems (after the unit tripping) is controlling PZR level, as the charging system (including seal injection flow control) valves fails open, letdown isolates and the PZR will continue to fill. Most of the other critical control valves for maintaining the primary systems are supplied from essential air.

DIS TRACTOR ANAL YSIS:

A. lncorrec1, Plausible because the AOP directs local action to ensure the valves are open in the same section of the procedure that the crew is directed to throttle seal injection. The valve is ensured to be open verses closed which runs counter intuitive with an air leak in progress.

B. lncorreci, Plausible because controlling the SG PORV locally would be required if the loss had been a loss of Essential Control Air (Auxiliary Air) instead of a loss of control air in the turbine building. See AQP-M.02 Section 2.1

,

C. Incorrect, Plausible because controlling the AFW LCVs locally would be required if the loss had been a loss of Essential Control Air (Auxiliary Air) instead of a loss of control air in the turbine building. See AOP-M.02 Section 2.1

,

D. Correct, The AOP directs local seal injection flow control to minimize PZR level rise.

Wednesday, June 05, 2013 8:16:15 AM 37

____

1305 NRC RO Exam Notes Question Number: 14 Tier: 1 Group 1 K/A: 065 Loss of Instrument Air AK3.08 Knowledge of the reasons for the following responses as they apply to the Loss of Instrument Air:

Actions contained in EOP for loss of instrument air.

Importance Rating: 3.7 / 3.9 10 CFR Part 55: 41.5, 41.10 1OCFR55.43.b: Not applicable K/A Match: This question matches the KJA by having the candidate recall a local operator action associated with loss of air and the reason for the action.

Technical

Reference:

AOP-M.02, Loss of Control Air, Revision 0021 Proposed references None to be provided:

Learning Objective: 0PL271 .AOP-M.02

  1. 3 Given a set of initial plant conditions, determine initial operator response to stabilize the plant.
  1. 7 Given the procedure and a set of initial set of conditions, determine actions requried to mitigate the event in progress.

Cognitive Level:

Higher X Lower Question Source:

New Modified Bank Bank X Question History: Significantly modified from WBN 06/2011 NRC exam, Q079 G 2.1.30. SQN ILT 1305 NRC Exam Comments:

Wednesday, June 05, 2013 8:16:15 AM 38

1305 NRC RO Exam

15. 077 AAI.03 05 Given the following:

- Unit 1 is at 100% power.

- The Transmission Operator has notified the plant that system grid voltage is high and forecasted to go higher.

- The Transmission Operator requests the plant to take in the maximum value of MVARs to help stabilize the grid.

Which ONE of the following transmission lines out of service affects the maximum allowed MVAR incominci value on Unit 1, and how is the adjustment made in accordance with 0-GO-5, Normal Power Operation?

TRANSMISSION LINE METHOD OF ADJUSTMENT A. A 161 KV line Exciter Voltage Auto Adjuster B. A 161 KV line Exciter Voltage Base Adjuster C A 500 KV line Exciter Voltage Auto Adjuster D. A 500 KV line Exciter Voltage Base Adjuster Feedback DIS TRACTOR ANAL YSIS:

A. Incorrect, 161 KV line availability determines Unit-2 VAR limit. Plausible because off-site power Operability is determined by 161 yard voltage and exciter auto voltage adjuster is the normal method adjuster is the method of varying VARS.

B. Incorrect, 161 KV line availability determines Unit-2 VAR limit. Plausible because off-site power Operability is determined by 161-yard voltage and the exciter voltage base adjuster is the method of varying VARS if the voltage regulator is not in automatic.

C. Correct, GOl-6 specifies the Unit 1 incoming VAR limits based on 500 KV line availability. 0-GO-5 specifies the Exciter Auto Voltage Adjuster as the means for VAR adjustment unless the voltage regulator is in Manual.

D. Incorrect, The exciter auto voltage adjuster is the normal method of varying VARS.

Plausible because the incoming VAR limits for Unit- lare based on 500 KV line availability.

Wednesday, June 05, 20138:16:15 AM 39

1305 NRC RO Exam Notes Question Number: 15 Tier: 1 Group 1 K/A: 077 Generator Voltage and Electric Grid Disturbances AA1 .03 Ability to operate and/or monitor the following as they apply to Generator Voltage and Electric Grid Disturbances:

Voltage Regulator controls Importance Rating: 3.8 I 3.7 10 CFR Part 55: 41.5, 41.10 1OCFR55.43.b: Not applicable K/A Match: Questions requires abiltiy to monitor conditions to determine the limit for incoming reactive load and then identify the device used to adjust amps.

Technical

Reference:

0-GO-5, Normal Power Operation, Revision 78 GOl-6, Apparatus Operations, Revision 148, Section E Proposed references None to be provided:

Learning Objective: OPT200GEN

7. EXPLAIN the Main Generator design features and/or interlocks that provide the following:
a. Generator Voltage Regulation
c. Generator capability, including power factor, VARs and hydrogen pressure Question Source:

New Modified Bank Bank X Question History: SON bank question 077 AA1.03 014 used on the SQN 09/2010 exam.

Comments:

Wednesday, June 05, 2013 8:16:15 AM 40

1305 NRC RO Exam

16. W/EO4EA2.2 016 Given the following plant conditions:

- Following a reactor trip, abnormal radiation was noted in the Aux. Building due to a loss of RCS inventory outside containment.

Which ONE of the following identifies a required action and the subsequent check used to determine whether or not the leak is isolated in accordance with ECA-1 .2, LOCA Outside Containment?

A. Isolate SI pump Cold Leg Injection; Pressurizer level rising B. Isolate SI pump Cold Leg Injection; RCS pressure rising C. Isolate RHR Cold Leg Injection; Pressurizer level rising D Isolate RHR Cold Leg Injection; RCS pressure rising Wednesday, June 05, 2013 8:16:15 AM 41

1305 NRC RO Exam Feedback DIS TRACTOR ANAL YSIS:

A. Incorrect; Plausible because portions of SI discharge piping have a design pressure of 1750 psig which is 500 psig below normal system pressure. This is not part of the isolation strategy for ECA-1.2. Also, Pressurizer level rising is plausible since the student could reason that it may be rising if the leak was isolated. The procedure directs the use of RCS pressure increasing as the method used to indicate the leak has been isolated.

B. Incorrect; Plausible because portions of SI discharge piping have a design pressure of 1750 psig which is 500 psig below normal system pressure. Also the

-

procedure directs the use of RCS pressure increasing as the method to indicate the leak has been isolated.

C. Incorrect; Plausible because isolation of RHR Cold Leg injection is a strategy contained in the procedure to isolate a LOCA outside CNMT. Pressurizer level rising is a plausible indication since the student could reason that it may be rising if the leak was isolated. The procedure directs the use of RCS pressure increasing as the method used to indicate the leak has been isolated.

D. Correct; Isolation of RHR Cold Leg injection is a strategy contained in the procedure to isolate a LOCA outside CNMT. The procedure directs the use of RCS pressure increasing as the method used to indicate the leak has been isolated.

Wednesday, June 05, 2013 8:16:15 AM 42

_____

1305 NRC RO Exam Notes Question Number: 16 Tier: 1 Group 1 K/A: W/E04 LOCA Outside Containment (CTMT)

EA2.2 Ability to determine and interpret the following as they apply to the (LOCA Outside CTMT):

Adherence to appropriate procedures and operation within the limitations in the facilitys license and amendments.

Importance Rating: 3.6 / 4.2 10 CFR Part 55: 41.7/45.5/45.6 1OCFR55.43.b: Not applicable K/A Match: This question matches the KJA by having the candidate determine the correct actions given a LOCA outside CNMT and the expected plant response for the actions taken.

Technical

Reference:

ECA-1 .2, LOCA Outside Containment, Revision 10 DWG 47W810-1 Proposed references None to be provided:

Learning Objective: OPL271 ECA-1 .2

4. Summarize the mitigating strategy for ECA-1 .2.
6. Given the procedure and a set of initial plant conditions, determine actions required to mitigate the event in progress.

Question Source:

New Modified Bank Bank X Question History: SQN bank question used on the SON 2010 exam Comments:

Wednesday, June 05, 2013 8:16:15 AM 43

1305 NRC RO Exam

17. W/E05 EK2.2 017 Given the following:

- The crew is implementing FR-H.1, Loss of Secondary Heat Sink.

- CST level is 25%.

- No Steam Generator is Intact.

Which ONE of the following identifies the preference for restoring a SG as a heat sink and the order in which the feed water sources are attempted in accordance with FR-H.1, Loss of Secondary Heat Sink?

A. Feed a ruptured SG before feeding a faulted SG; MEW, Condensate, MDAFW using ERCW B Feed a ruptured SG before feeding a faulted SG; MDAEW, TDAFW, MEW, Condensate C. Feed a faulted SG before feeding a ruptured SG; MEW, Condensate, MDAEW using ERCW D. Feed a faulted SG before feeding a ruptured SG; MDAFW, TDAFW, MEW, Condensate Feedback DISTRACTOR ANAL YSIS:

A. Incorrect, The priority of using a ruptured Steam generator before using a faulted steam generator is correct but the order of water sources is not correct. Plausible as if is most likely that the AFW systems have failed to require entry into H. 1. Then MFW, Condensate and MDAFW with Essential Raw Water cooling, which is the emergency backup source for MDAFW B. Correct, Step 7 of FR-H. 1 lists the priority of SG to feed and following steps list order of preference for source of feedwater. The preference is to use a ruptured steam generator before using a faulted steam generator.

C. Incorrect Plausible if it is determined that a faulted steam generator would have priority over a ruptured SG due to spread of contamination when a ruptured SG is steamed for heat removal. The second part is plausible as if is most likely that the AFW systems have failed to require entry into H. 1. Then MFW, Condensate and MDAFW with Essential Raw Water cooling, which is the emergency backup source for MDAFW.

0. Incorrect, Plausible if it is determined that a faulted steam generator would have priority over a ruptured SG due to spread of contamination when a ruptured SG is steamed for heat removal. The second part is correct.

Wednesday, June 05, 2013 8:16:15 AM 44

1305 NRC RO Exam Notes Question Number: 17 Tier: 1 Group 1 K/A: W/E05 Loss of Secondary Heat Sink EK2.2 Knowledge of the interrelationships between the (loss of Secondary Heat Sink) and the following: Facilitys heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility.

Importance Rating: 3.9 I 4.2 10 CFR Part 55: 41.10 1OCFR55.43.b: Not applicable K/A Match: This question matches the K/A due to having the candidate identify the sources of water for secondary heat removal following a loss of MFW event and has the candidate recall the order of priority for sources of makeup water and order of SGs to be fed.

Technical

Reference:

FR-H.1 Loss of Secondary Heat Sink, Revision 18 Proposed references None to be provided:

Learning Objective: OPL271FR-H.1

4. Summarize the mitigating strategy for FR-H.1.
6. Given the procedure and a set of initial plant conditions, determine actions required to mitigate the event in progress.

Question Source:

New Modified Bank Bank X Question History: SQN bank question W/E05 EK2.2 018 used on the SQN 02/2010 exam.

Comments:

Wednesday, June 05, 20138:16:15 AM 45

1305 NRC RO Exam

18. W/EI1 EK3.4018 Given the following plant conditions:

- At 0900 the Unit 1 Reactor Trips.

- At 0920 a small break LOCA occurs.

- At 0950 the crew transitioned to ECA-1 .1, Loss of RHR Sump Recirculation, due to the failure of both RHR pumps.

- Crew has established one train of ECCS flow per ECA-1 .1.

- SI flow cannot be terminated due to lack of subcooling.

- At 1030 the crew is performing ECA-1 .1 Step 20, Monitor if ECCS flow should be terminated:

- RVLIS indications are adequate.

- The RNO states Establish minimum ECCS flow:

Which one of the following correctly identifies the minimum ECCS flow rate per Curve 9 of ECA 1 .1 that meets the intent of ECA-1 .1, Step 20 RNO, AND the reason for performing this action?

REFERENCE PROVIDED A 325 gpm ECCS flow. To establish the minimum ECCS flow needed and delay RWST depletion.

B. 325 gpm ECCS flow. To ensure adequate RVLIS indications and establish conditions to start an RCP.

C. 400 gpm ECCS flow. To ensure adequate RVLIS indications and establish conditions to start an RCP.

D. 400 gpm ECCS flow. To establish the minimum ECCS flow needed and delay RWST depletion.

Wednesday, June 05, 2013 8:16:15 AM 46

1305 NRC RD Exam Feedback DIS TRACTOR ANAL YSIS:

A. Correct, From 0900 1030 (90 Mm) Using ECA-1. 1, curve 9, the minimum amount

-

of SI flow needed to match decay heat is approximately 325 gpm. The value of 325 gpm is in the acceptable region using the graph from time of trip AND meets the requirement of Minimum Flow to delay RWST depletion. The Basis states the operator is then instructed to establish the minimum ECCS flow needed to match decay heat in order to further decrease ECCS pump Flow and delay RWST depletion.

B. lncorrec1, Plausible since the value of 325 gpm would meet the requirements of the step to match the SI flow needed to match decay heat. Also plausible since reducing SI flow could be assumed to match the flow needed to maintain RVLIS and it is mentioned in the basis that when reducing flow the charging pumps could be realigned from CPIT flow to normal charging for seal cooling, however the actual reason is match decay heat to slow the decrease in RWST level.

C. Incorrect, Plausible since this flow rate would be in the acceptable range on ECA-1. 1 curve 9, however this does not meet the intent of the step which is to reduce flow as low as possible to slow the rate of RWST depletion. Also plausible since reducing SI flow could be assumed to match the flow needed to maintain RVLIS and it is mentioned in the basis that when reducing flow the charging pumps could be realigned from CPIT flow to normal charging for seal cooling, however the actual reason is match decay heat to slow the decrease in RWST level.

D. Incorrect, Plausible since this flow rate would be in the acceptable range on ECA- 1.1 curve 9, however this does not meet the intent of the step which is to reduce flow as low as possible to slow the rate of RWST depletion. Also plausible since the reason is correct per the basis document.

Wednesday, June 05, 2013 8:16:15 AM 47

1305 NRC RO Exam Notes Question Number: 18 Tier: 1 Group 1 K/A: W/E1 1 Loss of Emergency Coolant Recirculation EK3.4 Knowledge of the reasons for the following responses as they apply to the (Loss of Emergency Coolant Recirculation)

RO or SRO function within the control room team as appropriate to the assigned position, in such a way that procedures are adhered to and the limitations in the facilities license and amendments are not violated.

Importance Rating: 3.6 / 3.8 1OCFRParI55: 41.5/41.10 1OCFR55.43.b: Not applicable K/A Match: This question matches the K/A by having the candidate recall the reason for the actions that the crew will take to comply with the requirements of ECA-1 .1 while attempting to reduce the depletion rate of the RWST while unable to transfer to CNMT sump recirculation.

Technical

Reference:

ECA-1.1, rev 12 EPM-3-ECA-1 .1, Basis Document for ECA-1 .1, Loss of ECCS Sump Recirculation rev 5 Proposed references ECA-1 .1 Loss of ECCS Sump Recirculation, Curve 9, to be provided: Minimum ECCS Flow for Decay Heat vs. Time After Trip Learning Objective: 0PL271 .ECA-1 .1

4. Summarize the mitigating strategy for ECA-1 .1.
6. Given the procedure and a set of initial plant conditions, determine actions required to mitigate the event in progress.

Question Source:

New X Modified Bank Bank Question History: New question written for 1305 ILT exam Comments:

Wednesday, June 05, 2013 8:16:15 AM 48

1305 NRC RO Exam

19. 001 AA2.05 019 Given the following plant conditions:

Attime=TO

- Tavg Tref deviation is 0°F.

-

- Pressurizer level is 45% and stable.

- Reactor Power is approximately 75% and stable.

- Control Bank D step counters are at 166 steps.

Attime=T+2 mm

- Tavg is 2°F > Tref and rising.

- Pressurizer level 46% and slowly rising.

- Pressurizer spray valves have throttled open.

- Reactor Power is approximately 76% and slowly rising.

- Control Bank D step counters are at 178 steps and rising at 8 steps per minute Which ONE of the following identifies (1) the correct procedure to enter and (2) the FIRST action that must be performed?

A. (1) AOP-C.O1, Rod Control System Malfunctions.

(2) Trip the reactor and enter E-0, Reactor Trip or Safety Injection.

Bw (1) AOP-C.01, Rod Control System Malfunctions.

(2) Place the rod control mode selector switch to MANUAL.

C. (1) AOP-C.02, Uncontrolled RCS Boron Concentration Changes.

(2) Trip the reactor and enter E-0, Reactor Trip or Safety Injection.

D. (1) AOP-C.02, Uncontrolled RCS Boron Concentration Changes.

(2) Place the rod control mode selector switch in MANUAL.

Wednesday, June 05, 2013 8:16:16 AM 49

1305 NRC RO Exam Feedback DIS TRACTOR ANALYSIS:

A. Incorrect, The first part is correct. The second part is incorrect as itis the second action required if placing the rods in manual does not stop rod motion. Plausible as it is the step in the RNO column of the correct answer and it may be determined as a correct answer over placing the rod select in manual because the rods are moving in the wrong direction.

B. Correct, The conditions given in the stem meet the entry criteria for AOP-C.O1 and it is the correct procedure to mitigate the event in progress. The first action to take in AOP-C.O1 is to place the rods in manual.

C. lncorrec1, The entry into AOP-C.02 should be considered as an inadvertant dilution could be the cause of the temperature change and temperature, rod and NIS changes are symptoms of an inadvertant boration or dilution. This is incorrect though as clear entry conditions are met for AOP-C.O1. Plausible as it is an action in AOP-C.02.

D. Incorrect, The entry into AOP-C.02 should be considered as an inadvertant dilution could be the cause of the temperature change and temperature, rod and NIS changes are symptoms of an inadvertant boration or dilution. This is incorrect though as clear entry conditions are met for AOP-C.O1. The second part is plausible as it is correct and it is also an action in AOP-C.02 to move the rod control switch to manual and move rods.

Wednesday, June 05, 20138:16:16 AM 50

1305 NRC RO Exam Notes Question Number: 19 Tier: 1 Group 2 K/A: 001 Continuous Rod Withdrawal AA2.05 Ability to determine and interpret the following as they apply to the Continuous Rod Withdrawal:

Uncontrolled rod withdrawal from available indications.

Importance Rating: 4.4 / 4.6 10 CFR Part 55: 43.5/45.13 1OCFR55.43.b: Not applicable K/A Match: The question meets the KA in that it requires the applicant to diagnose an uncontrolled rod withdrawal from available indications.

Technical

Reference:

AOP-C.01 R22 AOP-C.02 R8 Proposed references None to be provided:

Learning Objective: 0PL271 .AOP-C.01

5. Summarize the mitigating strategy for AOP-C.01.
7. Given the procedure and a set of initial plant conditions, determine actions required to mitigate the event in progress.

Question Source:

New Modified Bank Bank X Question History: SQN Bank, SQN ILT 1305 NRC Exam Comments:

Wednesday, June 05, 2013 8:16:16 AM 51

1305 NRC RO Exam

20. 036 AK2.02 020 Given the following plant conditions:

- Refueling is in progress on Unit 1 when a report is made to the control room that an irradiated fuel assembly has been dropped.

- The following alarms are received on 0-XA-55-12A:

- 1 -RA-90-1 1 2A CNMT BLDG UP COMPT AIR MON HIGH RAD

- 1-RA-90-59A AX BLDG AREA RAD MON HIGH RAD

- 1-RA-90-131A CNTMT PURGE AIR EXH MON HIGH RAD Which ONE of the following identifies the required actions to be taken per AOP-M.04, Refueling Malfunctions?

A. Evacuate non-essential personnel from Containment, and if RM-90-400, Shield Bldg and/or 0-RM-90-1 01, Aux Bldg Vent rad monitors are trending up, manually initiate Containment Ventilation Isolation.

B Evacuate all personnel from Containment and Ensure Containment Ventilation Isolation has Actuated.

C. Evacuate the Fuel Handling Area and maintain SEP and Rx cavity levels as directed by Fuel Handling SRO.

D. Evacuate the immediate area and Verify that Auxiliary Building Isolation has Actuated.

Wednesday, June 05, 2013 8:16:16 AM 52

1305 NRC RO Exam Feedback DISTRACTOR ANAL YSIS:

A. Incorrect, Plausible since Evacuate non-essential personnel is the direction given for Reactor Cavity Seal Failure, in that case personnel are needed to stay in containment and perform actions. On fuel assembly drop/damage Containment is directed to be completely evacuated. Second part also is incorrect. If rad levels are rising operators are directed to obtain gas release rate data.

B. Correct, For the given Rad Monitor alarms, the Containment is directed to be evacuated. Also 1-RA-90-131A will cause automatic Containment Ventilation isolation. Operators are directed to verify isolation.

C. Incorrect, Plausible since Evacuate Fuel Handling area is direction in AOP-M.04 but for dropped fuel assembly in the Spent Fuel pit area not inside CNMT. Also second part of distractor is direction for reactor cavity seal failure.

0. Incorrect, Plausible since Evacuate the immediate area is direction in AOP-M.04 but for dropped or damaged new fuel assembly. Also second part of distractor is wrong since Aux building Isolation would have to be manually initia ted with the given rad monitor alarms.

Wednesday, June 05, 20138:16:16 AM 53

1305 NRC RO Exam Notes Question Number: 20 Tier: 1 Group 2 K/A: 036 Fuel Handling Incidents AK2.02 Knowledge of the interrelations between the Fuel Handling Incidents and the following:

Radiation monitoring equipment (portable and installed)

Importance Rating: 3.4 / 3.9 10 CFR Part 55: 41.7 1OCFR55.43.b: Not applicable K/A Match: This question matches the K/A by asking if the candidate knowledge of the interrelations for a Fuel Handling Incident inside containment and the expected automatic actions of high radiation.

Technical

Reference:

AOP-M.04, Refueling Malfunctions, section 2.2, Rev 7 Proposed references None to be provided:

Learning Objective: OPL271AOP-M.04

5. Summarize the mitigating strategy for AOP-M.04.
7. Given the procedure and a set of initial plant conditions, determine actions required to mitigate the event in progress.

Question Source:

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

Wednesday, June 05, 2013 8:16:16 AM 54

1305 NRC RO Exam

21. 037AG2.2.44 021 Given the following:
  • Unit 2 was at 100% RTP when a reactor trip occurred due to a sheared shaft on
  1. 1 RCP.
  • While the crew was in ES-0.1, Reactor Trip Response with the plant stable, chemistry reports that #2 SG has developed tube leak.
  • PZR Level is dropping very slowly with the 2A COP at 114 gpm.

The crew should first (1) and the crew will use (2) during the follow on step to depressurize the RCS in AOP-R.01.

A. (1) isolate letdown (2) auxiliary spray B (1) isolate letdown (2) normal spray C. (1) start the 2B CCP (2) auxiliary spray D. (1)start the 2B CCP (2) normal spray Wednesday, June 05, 2013 8:16:16 AM 55

1305 NRC RO Exam Feedback DIS TRACTOR ANAL YSIS:

A. Incorrect: Plausible as isolating letdown is correct. The second part is plausible as normal spray is no longer available due to the loss of some RCPs (#2 RCP is critical) and auxiliary spray is the first choice for a backup if letdown was available.

B. Correct: isolating letdown is the next step if PZR level is still dropping after maximizing one CCP. Normal spray is first choice for RCS depressurization and is available as long as #2 RCP is available.

C. Incorrect: The first part is plausible as it is a valid step in recovering PZR level after LD has been isolated. The second part is plausible as normal spray is no longer available due to the loss of some RCPs (#2 RCP is critical) and auxiliary spray is the first choice for a backup if letdown was available.

D. Incorrect: The first part is plausible as it is a valid step in recovering PZR level after LD has been isolated. The second part is correct.

Wednesday, June 05, 2013 8:16:16 AM 56

1305 NRC RO Exam Notes Question Number: 21 Tier: 1 Group 2 K/A: 037 Steam Generator Tube Leaks G 2.2.44 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.

Importance Rating: 4.2/4.4 10 CFR Part 55: (CFR: 41.5 / 43.5 / 45.12) 1OCFR55.43.b: Not applicable K/A Match: The examinee needs to determine correct actions from given stem indications of system status of PZR level slowly dropping and the running COP at maximized flow. This indication will require action to isolate letdown. The examinee(s) will then need to understand how their actions affected the systems in play with the event during follow on implementation of the AOP.

Technical

Reference:

AOP-R.O1, SGTL R31 E-0, R35 Proposed references None to be provided:

Learning Objective: OPL271AOP-R.O1

5. Summarize the mitigating strategy for AOP-R.01.
7. Given the procedure and a set of initial plant conditions, determine actions required to mitigate the event in progress.

Cognitive Level:

Higher Lower X Question Source:

New X Modified Bank Bank Question History: New question for ILT 1305 NRC Exam Comments:

Wednesday, June 05, 2013 8:16:16 AM 57

1305 NRC RO Exam

22. 051 AA1.04 022 Given the following:

- Unit 1 is operating at 100%.

- Main Generator output is 1200 MWe.

- Main Turbine is operating in IMP OUT.

- Air inleakage causes condenser pressure to change from 0.8 to 1.8 psia.

Which one of the following identifies:

(1) the effect on the Main Generator output (MWe)

AND (2) the required rod motion to maintain reactor power at 100%

A. (1) rise (2) none B. (1) rise (2) withdraw C. (1) lower (2) withdraw Dv (1) lower (2) none Feedback DIS TRACTOR ANAL YSIS:

A. Incorrect, the first part is incorrect. With condenser vacuum at a higher value, turbine efficency is decreased and generator ouput will lower. This is plausible if examThee confuses vacuum rising with increase generator output. The second part is correct. Since reactor power is a function of steam demand and steam demand has not changed, no rod motion is required.

B. Incorrect, the first part is incorrect (see item A). The second part is incorrect. This is plausible if examinee determThe that since generator output is decreased, reactor power will also decrease since normally a decrease Th generator output would indicate that the turbine load has decreased and reactor power would have decreased, therefore rods will need to be withdrawn to mathtain 100% RTP.

C. Incorrect, the first part is correct. With condenser vacuum at a higher value, turbine efficency is decreased and generator ouput will lower. The second part is Thcorrect (see item B).

D. Correcl, the first part is correct. With condenser vacuum at a higher value, turbine efficency is decreased and generator ouput will lower, the second part is correct.

Since reactor power is a function of steam demand and steam demand has not changed, no rod motion is required (Turbine is in IMPOUT).

Wednesday, June 05, 2013 8:16:16 AM 58

1305 NRC P0 Exam Notes Question Number: 22 Tier: 1 Group 2 K/A: 051 Loss of Condenser Vacuum AA1 .04 Ability to operate and / or monitor the following as they apply to the Loss of Condenser Vacuum: Rod Position Importance Rating: 2.5* / 2.5*

10 CFR Part 55: 41.7/45.5/45.6 1OCFR55.43.b: Not applicable K/A Match: Question match the KA because the applicant is required to determine rod position requirements resulting from the loss of condenser vacuum.

Technical

Reference:

Generic Fundamentals PWR Topic 193004 for turbine efficiency Generic Fundamentals PWR Topic 192008 for relationship of steam flow and reactor power AOP-S.02, Loss of Condenser Vacuum Proposed references None to be provided:

Learning Objective: OPL271AOP-S.02

3. Given a set of initial plant conditions, determine initial operator response to stabilize the plant.

Question Source:

New X Modified Bank Bank Question History: New question for the SQN 05/2013 exam Comments:

Wednesday, June 05, 2013 8:16:16 AM 59

1305 NRC RO Exam

23. 059AK3.01 023 ODCM LCO 1 .1.1 states that Liquid Radwaste Effluent Line radiation monitor 0-RM-90-1 22 shall be operable with its alarm setpoint set at a particular value.

Which one of the following completes the statement below?

The alarm setpoint of 0-RM-90-122 is based on (1) and upon alarming, the radiation monitor (2) terminate the release.

A (1) the limits of 10 CFR 20, Standards for Protection Against Radiation (2) will B. (1) the limits of 10 CFR 20, Standards for Protection Against Radiation (2) will NOT C. (1) the limits of 10 CFR 100, Reactor Site Criteria (2) will D. (1) the limits of 10 CFR 100, Reactor Site Criteria (2) will NOT Feedback DISTRACTOR ANAL YSIS:

A. Correct: the alarm set point is based on not exceeding ten times the limits of 10 CFR 20 and the alarming RM will terminate the release.

B. Incorrect: The first part is correct. The second part is plausible as there are rad monitors at SON that do not termThate the release when they alarm (for example, 0-RM-90-212, Station Discharge RM does not terminate the release).

C. Incorrect: The first part is plausible as 1OCFR 100 establishes and defines the site boundary, exclusion areas, etc and sets radiation limits at the site boundaries.

The second part is correct.

a Incorrect: The first part is plausible as 1OCFR 100 establishes and defines the site boundary, exclusion areas, etc and sets radiation limits at the site boundaries.

The second part is plausible as there are rad monitors at SON that do not termThate the release when they alarm (for example, 0-RM-90-212, Station Discharge RM does not terminate the release).

Wednesday, June 05, 2013 8:16:16 AM 60

1305 NRC RO Exam Notes Question Number: 23 Tier: 1 Group 2 K/A: 059 Accidental Liquid Pad Waste Release AK3.03 Knowledge of the reasons for the following responses as they apply to the Accidental Liquid RadWaste Release:

Termination of a release of radioactive liquid Importance Rating: 3.5/3.9 10 CFR Part 55: (CFR41.5,41.1O/45.6145.13) 1OCFR55.43.b: Not applicable K/A Match: The question meets the KA as it test the examinee on the reason the rad monitor wouldterminate the release.

Technical

Reference:

ODCM, R57 0-AR-M12-B, C-i, rev 29 Proposed references None to be provided:

Learning Objective: OPT200.LRW #s 4,7,10 & 13 Question Source:

New Modified Bank X Bank Question History: Modiified question LRW-B.6 002 Comments: Low Cognitive Wednesday, June 05, 2013 8:16:16 AM 61

1305 NRC RO Exam

24. 074 EA2.07 024 Given the following plant conditions:

- A Reactor Trip and Safety Injection has occurred on Unit 1.

- While performing the actions of E-0, Reactor Trip or Safety Injection, the following plant conditions are noted:

- Containment pressure is 6.5 psig and stable.

- RCPs have been stopped.

- RVLIS Lower Range is indicating 40%.

- Core Exit Thermocouples are indicating 710°F.

- PZR level is off scale low.

- PZR pressure is 400 psig.

- RCS Wide Range Hot Leg Temperatures are indicating 680°F.

Which ONE of the following identifies the accident that has occurred and the required procedure to be entered?

A. A PZR steam space break has occurred and a transition to FR-C.1, Response to Inadequate Core Cooling is required.

B. A PZR steam space break has occurred and a transition to FR-C.2, Response to Degraded Core Cooling is required.

C An RCS hot or cold leg break has occurred and a transition to FR-C.1, Response to Inadequate Core Cooling is required.

D. An RCS hot or cold leg break has occurred and a transition to FR-C.2, Response to Degraded Core Cooling is required.

Wednesday, June 05, 2013 8:16:16 AM 62

1305 NRC RO Exam Feedback DISTRACTOR ANAL YSIS:

A. Incorrect; Plausible since a PZR Steam Space accident is a type of Loss of Coolant Accident, however a PZR steam space accident can be identified by indicated PZR level along with low RCS Pressure and Low RVLIS inventory. The candidate may get the RCS indications confused. Also the second part is correct the stated plant conditions do hdicate Inadequate Core Coo/hg.

B. Incorrect, Plausible since a PZR Steam Space accident is a type of Loss of Coolant Accident; however a PZR steam space accident can be identified by indicated PZR level along with low RCS Pressure and Low RVLIS inventory. The candidate may get the RCS indications confused. Also plausible if the candidate gets confused on the Core Exit T/C readings (not 1200°F), or misses the fact that the RCPs have been secured. This would indicate that FR-C.2 Degraded Core Coo/mg would exist.

C. Correct, Due to the loss of inventory and pressure (RCS pressure 400 psig, and RVLIS at 40%) the plant has experienced a Loss of Coolant Accident. As indicated by PZR level (off scale low), the break would be in either the hot leg or cold leg.

Also the accident has progressed to a condition of Inadequate Core Cooling as indicated by RVLIS <42% and CETS >700 °F. lnaccordance with 1-FR-0, Unit 1 Status Trees, the conditions of No RCPS with Core Exit T/Cs >700°F and RVL1S Lower Range of <42% this would be an Inadequate Core Cooling condition.

D. Incorrect, Plausible since due to the loss of inventory and pressure (RCS pressure 400 psig, and RVLIS at 40%) the plant has experienced a Loss of Coolant Accident. As hdica ted by PZR level (off scale low), the break would be in either the hot leg or cold leg and is correct. The second part is plausible since if the candidate gets confused on the Core Exit T/C readings (not 1200°F), or misses the fact that the RCPs have been secured. This would indicate that FR-C.2 Degraded Core Cooling would exist.

Wednesday, June 05, 2013 8:16:16 AM 63

1305 NRC RO Exam Notes Question Number: 24 Tier: 1 Group 2 K/A: 074 Inadequate Core Cooling EA2.07 Ability to determine or interpret the following as they apply to a Inadequate Core Cooling:

The difference between a LOCA and inadequate core cooling, from trends and indicators.

Importance Rating: 4.1 / 4.7 10 CFR Part 55: 43.5 / 45.13 1OCFR55.43.b: Not applicable K/A Match: This question matches the K/A by having the candidate determine the type of LOCA that has occurred and having them identify that conditions of Inadequate Core Cooling have developed.

Technical

Reference:

1-FR-0, Unit 1 Status Trees, Core Cooling Proposed references None to be provided:

Learning Objective: 0PL271 FR-0, r3, FR-0, Status Trees Obj. 2 Given a set of initial conditions, determine if FR-0 entry is required.

Obj. 6 Given a set of initial plant conditions use FR-0 to correctly identify the:

b. FR procedure applicable to the current state of each Critical Safety Function.

Question Source:

New X Modified Bank Bank Question History: New question written for 1305 exam.

Comments:

Wednesday, June 05, 2013 8:16:16 AM 64

1305 NRC RO Exam

25. WIEO3EK1.1 025 Given the following plant conditions:
  • A small break LOCA has occurred.
  • The crew is implementing ES-i .2, Post LOCA Cooldown and Depressurization.
  • One Centrifugal Charging Pump (CCP) is running.
  • Both Safety Injection Pumps (SIPs) are running.
  • The crew has determined that one SIP can be stopped.

Which ONE of the following:

(1) Explains what will happen to the subcooling value when the SIP is stopped And (2) the next pump to be stopped if subcooling is adequate for further ECCS pump reduction?

A. (1) Lowers due to reduced ECCS injection flow and stablizes at a lower value when break flow and ECCS flow equal.

(2) The running CCP B. (1) Remains the same due to reduced ECCS injection flow causing RCS temperature and pressure to rise.

(2) The running CPP C (1) Lowers due to reduced ECCS injection flow and stablizes at a lower value when break flow and ECCS flow equal.

(2) The running SIP D. (1) Remains the same due to reduced ECCS injection flow causing RCS temperature and pressure to rise.

(2) The running SIP Wednesday, June 05, 2013 8:16:16AM 65

1305 NRC RO Exam Feedback DISTRA CTOR ANAL YSIS:

A. Incorrect, The first part is correct. The second part is plausible because the SI reduction sequence begins with stopping a charging pump, goes on to an SI pump and it is logical to assume that if further pump reduction is required to go back to a charging pump.

B. Incorrect, Plausible some events (SG FauIt, LBLOCA) will have RCS temperature rise in securing an SIP due to reduce cooling from reduced ECCS flow. On a SG Fauli the dynamics of the SG blowing dry and not being a heat sink causes RCS temperature and pressure to rise reducing ECCS flow with the result of subcooling lowering. The second part is plausible because the SI reduction sequence begins with stopping a charging pump, goes on to an SI pump and it is logical to assume that if further pump reduction is required to go back to a charging pump.

C. Correct, Subcooling will lower in this situation as RCS pressure lowers and stabilize at a lower value when break flow and ECCS flow equal. The next pump to be secured is the second SIP as the procedure will try to keep a CPP in service to establish h normal charging.

D. Incorrect, Plausible some events (SG Fault, LBLOCA) will have RCS temperature rise in securing an SIP due to reduce cooling from reduced ECCS flow. On a SQ FaulI the dynamics of the SG blowing thy and not being a heat sink causes RCS temperature and pressure to rise reducing ECCS flow with the result of subcooling lowering. The second part is correct.

Wednesday, June 05, 2013 8:16:16 AM 66

1305 NRC RO Exam Notes Question Number: 25 Tier: 1 Group 2 K/A: W/E03 LOCA Cooldown and Depressurization EK1 .1 Knowledge of the operational implications of the following concepts as they apply to the (LOCA Cooldown and Depressurization)

Components, capacity, and function of emergency systems.

Importance Rating: 3.4 / 4.0 1OCFR Part 55: 41.8/41.10 1OCFR55.43.b: Not applicable K/A Match: This question matches the KJA by having the candidate demonstrate the knowledge of the components used in the SI reduction sequence in post LOCA cooldown and depressurization and the operational implications of securing ECCS pumps (subcooling reduction) and upon securing an ECCS pump which pump will be next in the reduction sequence.

Technical

Reference:

ES-i .2, Post LOCA Cooldown and Depressurization, rev 18.

EPM-3-ES-i .2, Basis document Proposed references None to be provided:

Learning Objective: 0PL271 ES-i .2

  1. 4 Summarize the mitigating strategy for ES-i .2
  1. 13 Analyze and explain the process taht leads to a new RCS equalibrium pressure following the shutdown of an ECCS pump during the ES-i .2 reduction sequence.

Question Source:

New Modified Bank X Bank Question History: Moidified from SQN bank ES-i .2-B.2 Comments: High Cognitive Wednesday, June 05, 2013 8:16:16 AM 67

1305 NRC RO Exam

26. WIEI4EAI.I 026 Which one of the following identifies the interlocks that must be met before valve FCV-72-23 (Train A Containment Spray Suction from Containment Sump) can be opened?

A. Both FCV-74-3 (RHR Suction from RWST) closed and FCV-72-40 (RHR Discharge to RHR Spray) must be closed.

B. Both FCV-72-40 (RHR Discharge to RHR Spray) and FCV-72-34 (Containment Spray Pump Recirc) must be closed.

C Both FCV-72-22 (Containment Spray Suction from RWST) and FCV-74-3 (RHR Suction from RWST) must be closed.

D. Both FCV-72-34 (Containment Spray Pump Recirc) and FCV-72-22 (Containment Spray Suction from RWST) must be closed.

Feedback DISTRACTOR ANALYSIS:

A. Incorrect, FCV-74-3 is interlocked but FCV-72-40 is not interlocked with opening FCV-72-23. Plausible because FCV-72-3 is correct and FCV-72-40 is interlocked with other valves associated with sump swapover.

B. Incorrect, FCV-72-40 and FCV-72-34 are not interlocked with opening FCV-72-23. Plausible because both FCV-72-40 and FCV-72-34 are containment spray valves. Other pump recirc valves do have interlocks in transferring to the containment sump and. FCV-72-40 is interlocked with other valves associated with sump swapover.

C. CORRECT Both FCV-72-22 (Containment Spray Suction from RWST) and FCV-74-3 (RHR Suction from RWST) must be closed as shown on print 1-47W61 1-72-1.

0. lncorrect, FCV-72-34 is not interlocked with opening FCV-72-23. Plausible because the FCV-72-34 is a containment spray valve and other pump recirc valves do have interlocks in transferring to the containment sump.

Wednesday, June 05, 2013 8:16:16 AM 68

1305 NRC RO Exam Notes Question Number: 26 Tier: 1 Group 2 K/A: W/E14 High Containment Pressure EA1 .1 Ability to operate and/or monitor the following as they apply to the (High Containment Pressure):

Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Importance Rating: 3.7 / 3.7 10 CFR Part 55: 41.7 1OCFR55.43b: Not applicable K/A Match: This question matches the K/A by having the candidate demonstrate understanding of CNTMT spray system interlocks during a CNTMT high pressure situation.

Technical

Reference:

ES-1.3, rev 19 Proposed references None to be provided:

Learning Objective: OPT200CS #1 & 7 Question Source:

New Modified Bank Bank X Question History: SQN Bank, 0109 ILT NRC Exam Comments:

Wednesday, June 05, 2013 8:16:16 AM 69

1305 NRC RO Exam

27. W/EI5EK3.I 027 Given the following plant conditions:

- A large break LOCA has occurred on Unit 1.

- Accumulators have discharged and are isolated.

- ES-i .3, Transfer to Containment Sump, has been completed.

- Containment sump level is now at 84% and slowly rising.

- FR-Z.2, Containment Flooding, is in progress.

Which of the following describes; (1) where the FR-Z.2 required sample is taken, and (2) the reason for sampling the containment sump?

Ab RHR system To determine the level of activity, to allow the TSC to determine if excess sump water can be transferred to tanks outside of containment.

B. Containment sump To determine the level of activity, to allow the TSC to determine if excess sump water can be transferred to tanks outside of containment.

C. RHR system To ensure shutdown margin is being maintained, since non-borated water has entered the containment sump.

D. Containment sump To ensure shutdown margin is being maintained, since non-borated water has entered the containment sum p.

Wednesday, June 05, 2013 8:16:16 AM 70

1305 NRC RO Exam Feedback DISTRACTOR ANAL YSIS:

A. Correct, Since sump swapover has occurred, FR-Z.2 directs obtaining a sample from the RHR system to aid in determining if activity levels will allow transferring water to locations outside containment, to alleviate containment flooding.

B. Incorrect, Plausible, since containment sump would be the correct sample point if sump swapover had NOT been completed. Further plausibility is added because the reason given is correct.

C. Incorrect, Plausible, since sampling the sump is an action directed by FR-Z2.

Also, per the above excerpt from the WOG Background Document, if the crew is in FR-Z2, then non-bora ted water has entered containment, and it is plausible that shutdown margin would be a concern.

0. Incorrect, Plausible, since containment sump would be the correct sample point if sump swapover had NOT been completed. If the crew is in FR-Z2, then non-bora ted water has entered containment, and it is plausible that shutdown margin would be a concern.

Wednesday, June 05, 2013 8:16:16 AM 71

1305 NRC RO Exam Notes Question Number: 27 Tier: 1 Group 2 K/A: W/E15 Containment Flooding EK3.1 Knowledge of the reasons for the following responses as they apply to the (Containment Flooding):

Facility operating characteristics during transient conditions, including coolant chemistry and the effects of temperature, pressure, and reactivity changes and operating limitations and reasons for these operating characteristics.

Importance Rating: 2.7 / 2.9 10 CFR Part 55: 41.5/41.10,45.6,45.13 1OCFR55.43.b: Not applicable K/A Match: Question meets the KA by examining the applicant on where the chemistry sample is drawn during transient conditions and asking the reason for sampling.

Technical

Reference:

FRZ.2, WOG Background Document, 2.

DESCRIPTION, Page 2:

FR-Z.2, step 2. rev 7 Proposed references None to be provided:

Learning Objective: 0PL271 FRZ.2, rev 2 Obj. 4 Summarize the mitigating strategy for FR-Z.2 Question Source:

New Modified Bank Bank X Question History: Sqn bank question used on the SQN 09/2010 exam.

Comments:

Wednesday, June 05, 2013 8:16:16 AM 72

1305 NRC RO Exam

28. 003 K6.02 028 Given the following plant conditions:

- Unit 1 is at 100% rated thermal power.

- 1 -FCV-62-93, Charging Flow Control Valve is in Manual.

- 1 -FCV-62-89, Charging Seal Water Flow Control Valve, is operating at 60%

open.

- Due to a positioner failure, 1 -FCV-62-89 throttles close, and sticks at the 30% open position.

What effect will this malfunction have on charging pump discharge pressure and RCP seal injection flow?

Charging Pump RCP Seal Discharge Press Injection Flow A. Lowers Rises B. Rises Lowers Rises Rises D. Lowers Lowers Wednesday, June 05, 2013 8:16:16 AM 73

1305 NRC RD Exam Feedback DISTRACTOR ANAL YSIS:

A. Incorrect, Plausible if the candidate thinks that FCV-62-89 is located in the RCP seal supply line such that opening the valve would cause more flow and closing the valve would decrease flow, thus causing CCP discharge pressure to lower due to increased flow. Also plausible since the second part is correct.

B. Incorrect, Plausible since the first part is correct, the increased backpressure on the system would cause CCP discharge pressure to increase. Also the second part is plausible if the candidate thinks that FC V-62-89 is located in the RCP seal supply line such that opening the valve would increase flow and closing the valve would decrease flow.

C. Correct, the 1-FCV-62-89 valve is inline with the charging flow control valve and seal injection is supplied by a connection between the two valves. FCV-62-89 provides sufficient backpressure that RCP seals are supplied by charging. By increasing the position of FCV-62-89, less backpressure is provided, RCP seal injection will lower as charging flow increases and CCP discharge pressure would lower. Thus if FCV-62-89 closes down, it would provide more backpressure to the CCP, raising its discharge pressure and forcing more flow to the RCP seals.

D. Incorrect, Plausible if the candidate thinks that FCV-62-89 is located in the RCP seal supply line such that opening the valve would cause more flow and closing the valve would decrease flow, thus causing CCP discharge pressure to lower due to increased flow. Also the second part is plausible if the candidate thinks that FCV-62-89 is located in the RCP seal supply line such that opening the valve would increase flow and closing the valve would decrease flow.

Wednesday, June 05, 2013 8:16:16 AM 74

1305 NRC RO Exam Notes Question Number: 28 Tier: 2 Group: 1 K/A: 003 Reactor Coolant Pump System (RCPS)

K6.02 Knowledge of the effect of a loss or malfunction on the following will have on the RCPS:

RCP seals and seal water supply Importance Rating: 2.7 / 3.1 1OCFRPart55: 41.7/45.5 1OCFR55.43.b: Not applicable K/A Match: K/A is matched because the question requires knowledge of the how an RCP seal supply malfunction will affect charging pump operation and seal injection flow to the RCPs Technical

Reference:

1 -47W809-1, R79 Proposed references None to be provided:

Learning Objective: OPT200.RCP

  1. 9 Given specific plant conditions, Analyze the effect that a loss or malfunction of the following will have on the RCP
d. Seal injection supply Cognitive Level:

Higher X Lower Question Source:

New Modified Bank Bank X Question History: SON Bank from DC Cook 2007 NRC exam Comments:

Wednesday, June 05, 2013 8:16:17 AM 75

1305 NRC RO Exam

29. 004 A2.06 029 Given the following plant conditions:

- Unit 1 is operating at 100% power after restart following a refueling outage.

- Rod Control in MANUAL.

- VCT level is currently at 32%.

- An AUO places an un-borated mixed bed demineralizer in service.

Which ONE of the following completes the statements below?

Assuming NO operator action is taken, the VCT level over time will In accordance with AOP-C.02, Uncontrolled RCS Boron Concentration Changes, the first corrective action the RO will take that will stop the event in progress is to Lfl A. remain constant place 1-HS-62-79A, Mixed Bed Hi Temp Bypass, to V.C. TK position B. remain constant initiate normal boration C rise place 1-HS-62-79A, Mixed Bed Hi Temp Bypass, to V.C. TK position D. rise initiate normal boration Wednesday, June 05, 20138:16:17 AM 76

1305 NRC RO Exam Feedback DISTRACTOR ANAL YSIS:

A. Incorrect, Plausible because, unlike an inadvertent dilution due to makeup, no inventory is directly added to the VCT due to the unborated mixed bed and placing 1-HS-62-79A to the VCT position will divert letdown around the demin that is the source of the problem.

B. Incorrect, Plausible because, unlike an inadvertent dilution due to makeup, no inventory is directly added to the VCT due to the unborated mixed bed and AOP-C.02, Uncontrolled RCS Boron Concentration Changes, does refer operators to borate the RCS, however it will not stop the event in progress.

C. Correct, the dilution event in progress does not add any inventory but it will increase Tavg which will cause pressurizer level to increase above setpoint which will lower charging and place more coolant into the VCT. Also, placing 1-HS-62-79A to the VCT position will divert letdown around the demin that is the source of the problem.

0. Incorrect, Plausible because the dilution event in progress does not add any inventory but it will increase Tavg which will cause pressurizer level to increase above setpoint, which will lower charging and place more coolant into the VCT.

Also plausible because AOP-C.02, Uncontrolled RCS Boron Concentration Changes, does refer operators to borate the RCS, however it will not stop the event in progress.

Notes Question Number: 29 Tier: 2 Group: 1 K/A: 004 Chemical and Volume Control System A2.06 Ability to (a) predict the impacts of the following malfunctions or operations on the CVCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Inadvertent boration/dilution.

Importance Rating: 4.2 / 4.3 10 CFR Part 55: 41.5/43.5/45.3/45.5 1OCFR55.43.b: Not applicable K/A Match: K/A is matched because the question requires knowledge of the operation of CVCS, i.e. how charging will respond as Tavg and pressurizer level change and that effect on VCT level and knowledge of how to stop an unborated mixed bed from diluting the RCS through the CVCS.

Wednesday, June 05, 2013 8:16:17 AM 77

1305 NRC PC Exam Technical

Reference:

AOP-C.02, Uncontrolled RCS Boron Concentration Changes, Revision 0008 1 -SO-62-9 R44 Proposed references None to be provided:

Learning Objective: 0PT271 AOP-C.02

3. Given a set of initial plant conditions, determine initial Operator response to stabilize the plant.
7. Given the procedure and a set of of initial plant conditions, determine actions required to mitigate the event in progress Cognitive Level:

Higher X Lower Question Source:

New Modified Bank Bank X Question History: WBN bank question 004 A2.06 130 used on the WBN 10/2011 NRC exam with minor changes to make applicable for use on SQN 05/2013 NRC exam Comments:

Wednesday, June 05, 2013 8:16:17 AM 78

1305 NRC RO Exam

30. 004 A4.12 030 Given the following plant conditions:

- Unit 1 is at 100% rated thermal power.

- A routine dilution has just occurred.

- The integrater has counted out, however the OATC has NOT returned the CVCS Makeup Selector Switch to the AUTO position.

Which ONE of the following identifes the expected positions of the following valves?

Note:

1-FCV 140D, Boric Acid Valve to the Blender 1-FCV-62-128, Inletto the Top of the VCT 1 -FCV-62-1 40D 1 -FCV-62-1 28 A. Closed Open B Closed Closed C. Open Closed D. Open Open Feedback DIS TRACTOR ANAL YSIS:

A. Incorrect The 140D will be closed until the Makeup Selector Switch is returned to auto and is correct. The 128 valve will be closed at the end of the dilution, it is plausible to the examinee that it will be open as the 1400 valve will not return to normal position until after the makeup selector switch is returned to auto.

B. Correct: The 1400 valve is normally open and closes once the Makeup Selector Switch is placed in dilute or alt dilute. The 128 valve opens for a normal dilute and feeds to the top of the VC7 it will close as soon as the integrated completes the required makeup amount.

C. Incorrect The 1400 valve being open is plausible as it is normally open and will re-open once the Makeup Control Switch is returned to auto. The 128 as closed is correct.

0. lncorrecl The 1400 valve being open is plausible as it is normally open and will re-open once the Makeup Control Switch is returned to auto. The 128 valve will be closed at the end of the dilution, it is plausible to the examinee that it will be open as the 1400 valve will not return to normal position until after the makeup selector switch is returned to auto.

Wednesday, June 05, 2013 8:16:17 AM 79

1305 NRC AC Exam Notes Question Number: 30 Tier: 2 Group 1 K/A: 004 Chemical and Volume Control System A4.12 Ability to manually operate and/or monitor in the control room: Boration/dilution batch control Importance Rating: 3.8 / 3.3 10 CFR Part 55: CFR: 41.7 / 45.5 to 45.8 1OCFR55.43.b: Not applicable K/A Match: Meets the K/A due to requireing the examinee to monitor the correct component positions during a dilution.

Technical

Reference:

0-SO-62-7, R66 1 -47W61 1-62-2 R5 Proposed references None to be provided:

Learning Objective: OPT200.CVCS

  1. 6 Explain the CVCS design features and/or interlocks that provide the following:
j. RCS boron concentration control and modes of operation of the reactor makeup control system (blender controls)

Question Source:

New X Modified Bank Bank Question History: New for SQN ILT 1305 NRC Exam Comments: Low Cognitive- memory Wednesday, June 05, 2013 8:16:17 AM 80

1305 NRC RO Exam

31. 005 AI.03 031 Given the following plant conditions:

- Unit 1 is in Mode 5, midloop operation.

- RHR Train A in service at a flow rate of 2100 gpm.

- The RCS temperature is stable at 126°F.

- The operator throttles open 1 -FCV-74-32, RHR HXS BYPASS.

Which ONE of the following completes the statement below?

As 1 -FCV-74-32 is throttled open, the RCS temperature will jJ and the RHR flow rate indicated on 1-M-6 will {

w A. lower rise B. lower lower C rise rise D. rise lower Wednesday, June 05, 20138:16:17 AM 81

1305 NRC RO Exam Feedback DIS TRACTOR ANAL YSIS:

A. Incorrect, Plausible because the typical thought process is when bypassing a heat exchanger less heat will be picked up allowing the temperature to drop and because of where the flow is measured the indica ted flow rate increasing is correct.

B. Incorrect, Plausible because the typical thought process is when bypassing a heat exchanger less heat will be picked up allowing the temperature to drop and because there is a flow element on the flow through the heat exchanger that would sense a lower flow but it is not the flow element that provides the indication on the control board.

C. Correct, with the valve being opened further (until it is stopped by a restricting device placed on the valve during midloop operations), more flow to bypass the RHR HX. Less flow through the heat exchanger results in less cooling allowing RCS temperature to increase. Since total flow is measured downstream of where the HX Bypass connects to the HX discharge line (less overall system resistance),

the flow indication will rise.

0. Plausible because the RCS temperature increasing is correct and because there is a flow element on the flow through the heat exchanger that would sense a lower flow but it is not the flow element that provides the indication on the control board.

Wednesday, June 05, 2013 8:16:17 AM 82

1305 NRC RO Exam Notes Question Number: 31 Tier: 2 Group 1 K/A: 005 Residual Heat Removal System (RHRS)

Al .03 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the RHRS controls including: Closed cooling water flow rate and temperature Importance Rating: 2.5 I 2.6 10 CFR Part 55: (CFR: 41.5/ 45.5) 1OCFR55.43.b: Not applicable K/A Match: Question matches KA by adjusting the bypass valve and having the exam inee determine its effects on flow within the RHR system and temperature of the RCS Technical

Reference:

1-47W810-1 R22 l-47W811-l R74 Proposed references None to be provided:

Learning Objective: OPT200.RHR

  1. 8 Explain the RHR system design features and/or interlocks that provide the following:
a. RHR heat exchanger bypass flow control Question Source:

New Modified Bank Bank X Question History: Original question from WBN bank Comments: High Cognitive Wednesday, June 05, 2013 8:16:17 AM 83

1305 NRC RO Exam

32. 006 K5.05 032 Given the following plant conditions:

- Unit 2 is at 100% power

- An inadvertent Safety Injection Actuation has occurred.

Which ONE of the following identifies the adverse affect of allowing Safety Injection to continue without performing Safety Injection Termination?

A. ECCS Pumps will be running for extended time periods at minimum flow.

B. Loss of Instrument Air to Containment will not allow the use of the normal Pressurizer Spray Valves to control Pressurizer Pressure.

C Centrifugal Charging Pumps running in Injection Mode will collapse the Pressurizer bubble and pressurize the RCS to the PORV setpoint.

D. Reactor Coolant Pumps will be running without adequate pump seal cooling.

Feedback DIS TRACTOR ANAL YSIS:

A. Incorrect, Plausible as the SI and RHR pumps would be running on minimum flows and the examinee may give expected concern for this condition. However, in ES-i. 1 space this is not the reason for getting ECCS off B. Incorrect, Plausible because it could be thought that using PZR Spray Valves could prevent an over pressure condition, however, the PZR would continue to fill and pressurize the RCS until inventory was controlled. PZR Spray is mentioned in ES-i. 1 in restoring instrument air and in restoring RCPs.

C. Correct, The high head CCPs will continue to increase RCS inventory (shutoff head 2500 psig) resulting in high pressures up to the PORV setpoint if steps to reduce flow and restore letdown as part of SI termination are not performed.

0. Incorrect, Plausible because the RCPs would be running with seal injection but not normal seal return flow. Seal return would be through the seal return relief valve to the PRT and the examinee could conclude that seal cooling is not adequate for the RCPs without a complete understanding of CVCS/CCS conditions at this time. It is also a step in ES-i. 1 to address restoring normal seal return flow to the VCT.

Wednesday, June 05, 20138:16:17 AM 84

1305 NRC RO Exam Notes Question Number: 32 Tier: 2 Group 1 K/A: 006 Emergency Core Cooling System (ECCS)

K5.05 Knowledge of the operational implications of the following concepts as they apply to ECCS: Effects of pressure on a solid system Importance Rating: 3.4 / 3.8 10 CFR Part 55: (CFR: 41.5 / 45.7) 1OCFR55.43.b: Not applicable K/A Match: The question matches the KA in that it test the examinee on the effects of the CCPs continuing the run on RCS pressure control. The PZR will go solid and make RCS pressure control very difficult and in fact create a LOCA in the RCS.

Technical

Reference:

EPM-3-ES-1.1 R5, Basis Document for ES-1.1 ES-l.1 R12, SI Termination Proposed references None to be provided:

Learning Objective: 0PL271 ES-i .1

  1. 4 Summarize the mitagating strategy for ES-i .1.
  1. 5 Describe the basis for all limits, notes, cautions and steps of ES-i .1.

Question Source:

New Modified Bank Bank X Question History: CPNPP March 2010 NRC Written Exam Comments: High Cognitive Wednesday, June 05, 2013 8:16:17 AM 85

1305 NRC RO Exam

33. 007 A3.0I 033 Which ONE of the following identifies the pressure that relief valve 63-637, RHR Pump Discharge, will start relieving and the tank where the flow through the valve will be routed?

Pressure Tank A. 550 psig RCDT B. 550 psig PRT C. 600 psig RCDT Dv 600 psig PRT Feedback DIS TRACTOR ANAL YSIS:

A. Incorrect, Plausible because 550 psig is the maximum RHR pump discharge pressure allowed to be maintained in accordance with the System Operating Instruction when RHR system is in service and the RCDT is an RCS tank inside containment (like the PRT) which does receive flow and leakoifs from RCS related components.

B. Incorrect, Plausible because 550 psig is the maximum RHR pump discharge pressure allowed to be maintained in accordance with the System Operating Instruction when RHR system is in service and the PRT being the tank that receives flow passing through the valve is correct.

C. Incorreci, Plausible because 600 psig is the pressure that the valve starts relieving and the RCDT is an RCS tank inside containment (like the PRT) which does receive flow and leakoffs from RCS related components.

0. CorrecI, the RHR discharge relief valve, 63-63, starts relieving at 600 psig and is routed to the PAT.

wednesday, June 05, 2013 8:16:17 AM 86

1305 NRC RO Exam Notes Question Number: 33 Tier: 2 Group 1 K/A: 007 Pressurizer Relief Tank/Quench Tank System (PRTS)

A3.01 Ability to monitor automatic operation of the PRTS, including:

Components which discharge to the PRT Importance Rating: 2.7 /2.9 10 CFR Part 55: 41.7 I 45.5 to 45.8 1OCFR55.43.b: Not applicable K/A Match: K/A is matched because the question requires the ability to monitor the automatic operation of the RHR discharge pressure and PRT level to know if the relief valve is relieving (which did occur at SQN allowing approximately 10,000 gallons to be passed to the PRT prior to termination)

Technical

Reference:

1-47W811-1 R74 0-47W813-1 R55 1 -SO-74-1, Residual Heat Removal System, Revision 0086 Proposed references None to be provided:

Learning Objective: OPT200.RHR

18. STATE the RHR design pressure and flow capacities.

OPT200.PRT

19. LIST the components that discharge to the PRT Cognitive Level:

Higher Lower X Question Source:

New Modified Bank Bank X Question History: WBN bank question 007A3.01 33 used on the WBN 06/2011 NRC exam.

Comments:

Wednesday, June 05, 2013 8:16:17 AM 87

1305 NRC RO Exam

- 34. 007 G2.1.20 034 Given the following plant conditions:

- Unit 2 is at 90% power.

- Due to a leaking PORV, 2-FCV-68-332 PORV block valve was closed.

- PRT level is 80%

- PRT pressure is 7 psig

- PRT Temperature is 145°F Which ONE of the following describes the action to be taken in 2-SO-68-5, Pressurizer Relief Tank, to return the PRT to normal?

A. Start a Waste Gas Compressor, open 2-PCV-68-301, PRT VENT TO WDS VENT HDR, and reduce PRT pressure to < 4 psig.

B. Align the B RCDT pump, open FCV-68-305, N2 SUPPLY TO PRT, open 2-LCV-68-310, PRT DRAIN TO RCDT, and lower PRT level to 60%.

C. Open 2-FCV-68-303, PRIMARY WATER TO PRT, and return PAT level to 88%.

Dv Open 2-FCV-68-303, PRIMARY WATER TO PRT, and reduce PRT temperature to < 120°F.

Feedback DIS TRACTOR ANAL YSIS:

A. lncorrect Plausible as pressure is high and this is the correct action to take to reduce PRT pressure in 2-SO-68-5.

B. Incorrect, Plausible as level is higher than normal (although below alarm set point of 88%) this is the correct action to take to reduce PRT level in 2-SO-68-5.

C. Incorrect, Plausible as one of the stopping criteria in 2-S0-68-5 for lowering PRT temperature is to feed primary water to lower temperature to < 120 °F or until level is 88%. When level reaches 88%, the PRT is drained and this process is repeated.

D. Correct, The temperature in the PRT exceeds the alarm set point and this will be addressed first per the AR. Per 2-S0-68-2 the correct action to take for a high temperature condition is to Open 2-FC V-68-303 to feed primary water to the PRT.

Wednesday, June 05, 2013 8:16:17 AM 88

1305 NRC RO Exam Notes Question Number: 34 Tier: 2 Group 1 K/A: 007 Pressurizer Relief Tank/Quench Tank System (PRTS)

G2.1 .20 Ability to interpret and execute procedure steps.

Importance Rating: 4.6/4.6 10 CFR Part 55: (CFR: 41.10 /43.5 /45.12) 1OCFR55.43.b: Not applicable K/A Match: This question matches the K/A by testing the candidates knowledge of the procedures associated with PZR PRT level, pressure and temperature.

Technical

Reference:

2-SO-68-5 Ri 7 2-AR-M5-A, C-i P25 Proposed references None to be provided:

Learning Objective: OPT200. PZR-PRT, 8f, Given specific plant conditions, analyze the effect that a loss or malfunction of the PZR pressure control system will have on tne PRT.

Question Source:

New Modified Bank Bank X Question History: SQN Bank, SQN ILT 1211 Audit Exam Comments: High Cognitive Wednesday, June 05, 2013 8:16:17AM 89

1305 NRC RO Exam

35. 008 A3.01 035 Given the following plant conditions:

- Unit 1 is operating at 100% power.

- 1-RA-90-123A CCS LIQ EFF MON HIGH RAD Alarm is LIT.

- The red HIGH light is LIT on CCS LIQUID EFFLUENT RADMON 0-RM-90-1 23A.

- CCS surge tank level was increasing but is now stable.

- RC PUMPS THRM BARRIER RETURN HEADER FLOW LOW Alarm is LIT.

Which ONE of the following completes the statement below concerning the automatic actions that are designed to occur?

The thermal barrier heat exchanger containment isolation inlet and outlet valves to close and the thermal barrier booster pumps A. (1) the affected RCP ONLY (2) trip B. (1) the affected RCP ONLY (2) continue to run with minif low valves open.

C (1) all four (4) RCPs (2) trip.

D. (1) all four (4) RCPs (2) continue to run with minif low valves open.

Wednesday, June 05, 2013 8:16:17 AM 90

1305 NRC RO Exam Feedback DIS TRACTOR ANAL YSIS:

A. Incorrect: The first part is incorrect. RCP thermal barriers do not have pump specific containment isolation valves (CIVs). The four valves which auto close isolates CCS to all RCPs. This is plausible since some components can be isolated individually. The second part is correct. The Thermal Barrier Booster pumps will trip with a differential flow across the thermal barrier since this is an indication of a thermal barrier leak.

B. Incorrect: The first part is incorrect. RCP thermal barriers do not have pump specific ClVs. The four valves which auto close isolates CCS to all RCPs. This is plausible since some components can be isolated individually. The second part is incorrect. The Thermal Barrier Booster pumps do not have miniflow valves to ensure pump cooling. This is plausible since other pumps do have this configuration.

C. Correct: The first part is correct. The four valves which auto close isolates CCS to all RCPs. The second part is correct. The Thermal Barrier Booster pumps will trip with a differential flow across the thermal barrier since this is an indication of a thermal barrier leak.

D. Incorrect: The first part is correct. The four valves which auto close isolates CCS to all RCPs. The second part is incorrect. The Thermal Barrier Booster pumps do not have miniflow valves to ensure pump cooling. This is plausible since other pumps do have this configuration.

Wednesday, June 05, 20138:16:17 AM 91

1305 NRC RO Exam Notes Question Number: 35 Tier: 2 Group 1 K/A: 008 Component Cooling Water System (COWS)

A3.01 Ability to monitor automatic operation of the COWS, including:

Setpoints on instrument signal levels for normal operations, warnings, and trips that are applicable to the COWS.

Importance Rating: 3.2* / 3.0 10 CFR Part 55: 41.7 1OCFR55.43.b: Not applicable K/A Match: This question matches the K/A by having the candidate monitor the automatic operation of COWS components when an alarm setpoint is reached which causes components of the COWS to operate.

Technical

Reference:

0-AR-M12-A (B-i) rev 52 0-AR-M27-B-A (B-i) rev 12 47W610-90-2 rev 78 47W 61 0-70-3 rev 22 Proposed references None to be provided:

Learning Objective: OPT200.OOS Rev 9 Obj 11 .d Given specific plant conditions, ANALYZE the effect that a loss of malfunction of the following will have on the CCS: Heat Exchanger leaks.

Question Source:

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

Wednesday, June 05, 2013 8:16:17 AM 92

1305 NRC RO Exam

36. 010 K6.01 036 Given the following plant conditions:

- Unit 1 is operating at 100% RTP.

- Pressure Control Channel Selector Switch, 1 -XS-68-340D is selected to PT-68-340 & 334 position.

- 1-PT-68-334, Pressurizer Pressure Transmitter, fails LOW.

- The operating crew enters AOP-l.04, Pressurizer Instrument and Control Malfunctions.

Which ONE of the following completes the statements below?

The failure JJL result in the pressurizer back-up heaters being energized.

When AOP-l.04 performance is completed L of the PORVs will be able to automatically open if pressurizer pressure begins rising.

Lil A. will both B. will only one C will NOT both D. will NOT only one Wednesday, June 05, 2013 8:16:17 AM 93

1305 NRC RO Exam Feedback DIS TRACTOR ANAL YSIS:

A. Incorrect, Plausible because if the failure had been on 1-PT-68-340 the pressuizer backup heaters would have been auatomatically energized and both of the PORVS being able to automatically opening on highe pressure is correct.

B. Incorrect, Plausible because if the failure had been on 1-PT-68-340 the pressuizer backup heaters would have been auatomatically energized and because if the failure had been on 1-PT-68-323 or 1-PT-68-322, then one of the PORVs would have been made incapable of automatically opening if the pressure rose to the setpoint..

C. Correct, with the selector switch in the PT-68-340 & 334 position, the failure will not affect the pressurizer backup heaters but will result in a loss of one of the 2 required channels to allow PORV 334 to open in automatic. During performance of AOP-I.04 the selector switch XS-68-340D will be repositioned to PT-68-340 & 322 which restores the lost input for PORV 334 operation.

D. Incorrect, Plausible because the operation of the pressuizer backup heaters not being affected is correct but they would have been if the failure had been on 1-PT-68-340. Also plausible because if the failure had been on 1-PT-68-323 or 1-PT-68-322, then one of the PORVs would have been made incapable of automatically opening if the pressure rose to the setpoint.

Wednesday, June 05, 2013 8:16:17 AM 94

1305 NRC RO Exam Notes Question Number: 36 Tier: 2 Group 1 K/A: 010 Pressurizer Pressure Control System (PZR PCS)

K6.01 Knowledge of the effect of a loss or malfunction of the following will have on the PZR PCS:

Pressure detection systems Importance Rating: 2.7 / 3.1 1OCFRPart55: 41.7/45.7 1OCFR55.43.b: Not applicable K/A Match: Qustion requires knoowledge of how a malfunction of a pressure detection system transmitter will affect compontents in the Pressure Pressure Control System and how the required procedure prosponse will mitigate the consequence to the malfucntion.

Technical

Reference:

AOP-l.04, Pressurizer Instrument and Control Malfunctoins, Revision 12 Proposed references None to be provided:

Learning Objective: OPL271AOP-l.04

  1. 5 Summarize AOP-I.04 mitigating strategy for each operator section.
  1. 10 Identify automatic actions associated with dropping/rising RCS pressure.

Question Source:

New X Modified Bank Bank Question History: New question for the SQN 05/2013 exam Comments:

Wednesday, June 05, 2013 8:16:17 AM 95

1305 NRC RO Exam

37. 010 K6.03 037 Given the following plant conditions:

- Unit 2 was operating at 100% power.

- The Loop 1 pressurizer spray valve controller failed causing the spray valve to fully open.

Assuming No Operator Actions are taken, which ONE of the following identifies the response of the pressurizer pressure control system?

A. Master controller output would INCREASE.

PZR pressure would be maintained above the Reactor Trip setpoint.

B. Master controller output would INCREASE.

PZR pressure would decrease to the Reactor Trip setpoint.

C. Master controller output would DECREASE.

PZR pressure would be maintained above the Reactor Trip setpoint.

D Master controller output would DECREASE.

PZR pressure would decrease to the Reactor Trip setpoint.

Wednesday, June 05, 2013 8:16:17 AM 96

1305 NRC RO Exam Feedback DIS TRACTOR ANAL YSIS:

A. Incorrect The output of the master controller increases as pressure goes high, not as pressure drops below setpoint. to turn on heaters not increasing, but with the spray valve fully open, the heaters would not be able to terminate the pressure drop, and a reactor trip on low pressurizer pressure would occur. Plausible if the applicant knowing that the heaters should be turned on but confuses the direction of the change in the output of the master controller or believes the heaters coming on would prevent the pressure from continuing to drop to the reactor trip setpoint.

B. Incorrect, The output of the master controller increases as pressure goes high, not as pressure drops below setpoint. to turn on heaters not increasing, but with the spray valve fully open, the heaters would not be able to termThate the pressure drop, and a reactor trip on low pressurizer pressure would occur. Plausible if the applicant knowmg that the heaters should be turned on but confuses the direction of the change in the output of the master controller and knows that the heaters commg on would not prevent the pressure from contmumg to drop to the reactor trip setpoint.

C. Incorrec1, The output of the master controller does decrease as the pressure drops below setpoint to turn on heaters, but with the spray valve fully open, the heaters would not be able to terminate the pressure drop, and a reactor trip on low pressurizer pressure would occur. Plausible if the applicant knowing that the heaters should be turned on and which direction the output of the master controller would change, but believes the heaters coming on would prevent the pressure from contThuThg to drop to the reactor trip setpoint.

D. Correct, The pressurizer pressure will be droppmg due to the spray valve being open. As the lower pressure is compared to the setpoTht pressure, the output of the master controller will start droppmg to turn on heaters. With the spray valve fully open, the heaters would not be able to termThate the pressure drop, and a reactor trip on low pressurizer pressure would occur.

Wednesday, June 05, 20138:16:17 AM 97

1305 NRC RO Exam Notes Question Number: 37 Tier: 2 Group 1 K/A: 010 Pressurizer Pressure Control System K6.03 Knowledge of the effect of a loss or malfunction of the following will have on the PZR PCS: PZR sprays and heaters Importance Rating: 3.2 / 3.6 1OCFR Part 55: 41.7/45.7 1OCFR55.43.b: Not applicable K/A Match: Question requires the applicant to understand how a pressurizer spray valve failing open will affect the pressurizer control system and whether the heaters in the control system are designed to prevent a reactor trip due to the valve failure.

Technical

Reference:

1 -,2-47W61 1-68-3 AOP-L04 R12 Proposed references None to be provided:

Learning Objective: OPT200. PZR-PRT

9. Given specific plant conditions, ANALYZE the effect that a loss or malfunction of the following will have on the Pressurizer Level or Pressure Control System:
a. PZR sprays and heaters.

Question Source:

New Modified Bank Bank X Question History: SQN bank question 010 K6.03 037 used on the SQN 1/2008 exam Comments:

Wednesday, June 05, 20138:16:17 AM 98

1305 NRC RO Exam

38. 012 K2.01 038 Which ONE of the following identifies the plant electrical boards that supply power to the listed components on Unit 1?

SSPS Train B Reactor Reactor Trip Bypass Breaker A TriD Breaker 48v UV coil (BYA Control Power Circuit A 120v AC Vital Instrument 125V DC Vital Battery Board I Power Boards II and IV B. 120v AC Vital Instrument 125V DC Vital Battery Board II Power Boards II and IV C. 120v AC Vital Instrument 125V DC Vital Battery Board I Power Board II ONLY D. 120v AC Vital Instrument 125V DC Vital Battery Board II Power Board II ONLY Feedback DIS TRACTOR ANAL YSIS:

A. Correct 120v AC Vital Instrument Power Boards II and IV supply the 4.9v Reactor Trip Undervoltage relay through an auctioneered circuit and the 125V DC Battery Board II is the control power to BYA.

B. lncorrect, Plausible because the 120v AC Vital Instrument Power Boards II and IV supplying the 48v reactor Trip Undervoltage relay through an auctioneered circuit is correct and the 125V DC Battery Board Ills the control circuit power supply Train B reactor trip breakers and BYA receives trip signal from Train B circuits.

C. Incorrect, Plausthie because the 120v AC Vital Instrument Power Boards Ills the only power supply to other components in SSPS Train B (e.g. Slave relays) and the 125V DC Battery Board I is the control power supply to BYA.

D. lncorrect Pta usthie because the 120v AC Vital Instrument Power Boards II is the only power supply to other components in SSPS Train B (e.g. Slave relays) and the 125V DC Battery Board Ills the control circuit power supply Train B reactor trip breakers and BYA receives trip signals from Train B SSPS Reactor Trip circuits.

Wednesday, June 05, 2013 8:16:17 AM 99

1305 NRC RO Exam Notes Question Number: 38 Tier: 2 Group 1 K/A: 012 Reactor Protection System K2.01 Knowledge of bus power supplies to the following:

RPS channels, components, and interconnections.

Importance Rating: 3.3 / 3.7 10 CFR Part 55: 41.7 1OCFR55.43.b: Not applicable K/A Match: K/A is matched because the question requires the knowledge of the bus power supplies to Reactor Protection System components.

Technical

Reference:

1 ,2-45W699-1 RiO Proposed references None to be provided:

Learning Objective: OPT200.RPS

5. LIST the bus power supplies to the following Reactor Protection System components:
a. RPS channels
b. Reactor Trip Breaker Control Power Cognitive Level:

Higher Lower X Question Source:

New Modified Bank Bank X Question History: WBN bank question used on the WBN 10/2011 NRC exam.

Comments:

Wednesday, June 05, 2013 8:16:17 AM 100

1305 NRC RO Exam

39. 013 K4.07 039 Given the following plant conditions:

- Unit 1 is operating at 100% power.

- The operating crew is responding to a loss of 1 20V AC Vital Instrument Power Board 1-I.

- PZR pressure transmitter 1-PT-68-334 (Channel II) fails LOW.

Which ONE of the following identifies how SSPS and ECCS will respond?

A. Both trains of SSPS SI master relays will actuate AND both trains of ECCS equipment auto start.

B Both trains of SSPS SI master relays will actuate BUT only B train ECCS equipment auto starts.

C. Only the B train SSPS SI master relays will actuate BUT both trains of ECCS equipment auto start.

D. Only the B train SSPS SI master relays will actuate AND only B train ECCS equipment auto starts.

Feedback DIS TRACTOR ANAL YSIS:

A. Incorrect, Master Relays on both trains will have power. Train A from Channel Ill via an auctioneering circuit, however, with the 1-I AC vital Instrument Power Board deenergized (Channel 1), the slave relays that control the Train A equipment will not have a power supply. Plausible if the candidate mistakes the source of the power supply or thinks that the circuit that auctioneers power in the logic cabinet provides power to the slave relays.

B. Correct, Master Relays on both trains will have power. Train A from Channel III via an auctioneering circuit, however, with the 1-I AC vital Instrument Power Board deenergized, the slave relays that control the Train A equipment will not have power.

C. Incorrect, Master Relays on both trains will have power. Train A from Channel III via the auctioneering circuit, however, Channel 1 is the only power supply for the slave relays that control the Train A equipment. Plausible if the candidate mistakes the source of the power supply or thinks that the circuit that auctioneers power in the logic cabinet provides power to the slave relays instead of the master relays.

D. Incorrect, Master Relays on both trains will have power. Train A from Channel III via an auctioneering circuit, however, Channel 1 is the only power supply for the slave relays that control the Train A equipment. Plausible if the candidate mistakes the function of the circuit that auctioneers power in the logic cabinet.

Wednesday, June 05, 2013 8:16:18 AM 101

1305 NRC RO Exam Notes Question Number: 39 Tier: 2 Group 1 K/A: 013 Engineered Safety Features Actuation System K4.07 Knowledge of ESFAS design feature(s) and/or interlock(s) which provide for the following: Power supply loss Importance Rating: 3.7/4.1 10 CFR Part 55: 41.7 1OCFR55.43.b: Not applicable K/A Match: Thsi question matches the KA by testing the candidates knowledge of power supplies to portions of the ESFAS and how a loss of power will effect the systems ability to fully, or partially, actuate trains of ECCS components.

Technical

Reference:

47W61 1-63-1, R4 AOP-P.03, Loss of Unit 1 Vital Instrument Power Board, R25 Proposed references None to be provided:

Learning Objective: OPT200.RPS B.4 & 5 Question Source: SQN Bank RPS-B.9.A 002 New Modified Bank Bank X Question History: SQN ILT 1002 exam, SQN ILT 1305 NRC Exam Comments: High Cognitive Wednesday, June 05, 20138:16:18 AM 102

1305 NRC RO Exam

40. 022 K4.03 040 Which ONE of the following Containment Cooling System fans will trip and isolate as a DIRECT result of a Containment Isolation Phase-A Signal?

A. Lower Compartment Coolers B. Upper Compartment Coolers C Incore Instrument Room Coolers D. Control Rod Drive Motor Coolers Feedback All 4 of the choices are containment coolers that are tripped by one of the containment isolation signals.

DIS TRACTOR ANAL YSIS:

A. Incorrect, These coolers are tripped for a phase B signal, the examinee could mistake these as being tripped by a Phase A isolation signal. Plausible because the Lower Compartment Cooler fans do get a signal to trip from a containment isolation signaI (Phase B, not Phase A)

B. Incorrect, These coolers are tripped for a phase B signaI the examinee could mistake these as being tripped by a Phase A isolation signal. Plausible because the Upper Compartment Cooler fans do get a signal to trip from a containment isolation signal, (Phase B, not Phase A)

C. CQRRECT The Incore Instrument Coolers are tripped by a Phase A isolation signal.

0. Incorrect, These coolers are tripped for a phase B signa) the examinee could mistake these as being tripped by a Phase A isolation signal. Plausible because the CRDM Cooler fans do get a signal to trip from a containment isolation signal, (Phase B, not Phase A)

Wednesday, June 05, 2013 8:16:18 AM 103

1305 NRC RO Exam Notes Question Number: 40 Tier: 2 Group 1 K/A: 022 Containment Cooling System K4.03 Knowledge of CCS design feature(s) and/or interlock(s) which provide for the following:

Automatic containment isolation Importance Rating: 3.6 / 4.0 10 CFR Part 55: 41 .7 1OCFR55.43.b: Not applicable K/A Match: Question requires the knowldege of the desing feature that results in the tripping of the containment coolers on a Phase A containment isolation.

Technical

Reference:

1 ,2-47W61 1-30-2 R2 1 ,2-47W61 1-30-3 R6 1 ,2-47W61 1-30-4 Ri 8 Proposed references None to be provided:

Learning Objective: OPT200.CNTMCLG&PURGE

8. EXPLAIN the Containment Cooling and Purge Systems design features and/or interlocks that provide the following:
a. Automatic containment isolation Question Source:

New Modified Bank Bank X Question History: SQN Bank question 022 K4.03 used on the SQN 1/2009 exam with correct answer location changed and minor wording change for use on the SQN 05/2013 exam Comments:

Wednesday, June 05, 2013 8:16:18 AM 104

1305 NRC RO Exam

41. 025 A4.02 041 Given the following plant conditions:

- Unit 1 at 100% RTP when a LOCA occurs.

- A Safety Injection occurs due to containment pressure rising.

- The containment pressure has continued to rise and is now 3.2 psig.

Which ONE of the following completes the statements below?

The Containment Air Return Fans would automatically start 10 minutes after the.jJ signal.

If the 1 A-A Air Return Fan tripped on excessive current when the start was attempted, the indicating lights on the MCR handswitch would be LIT?

Ui A. Safety Injection GREEN and WHITE only B. Safety Injection GREEN, WHITE and RED C. Phase B Isolation GREEN and WHITE, only D Phase B Isolation GREEN, WHITE, and RED Wednesday, June 05, 20138:16:18 AM 105

1305 NRC RO Exam Feedback Dwg 1,2-45N779-5 shows schematic for the lA-A Air Return Fan and shows the auto start 10 mm after a phase B w/o BO. Using a combination of 779-5 and 779-1 (detail Al) shows that component overcurrent protection is from an amptector. When the amptector actuates it trips open the breaker and toggles the OTS contacts shown on 779-1. The breaker 52b contact at conection 8 closes to turn on the green lite, the OTS contact at conection 9 closes to turn on the red lite. The OTS closes at connection 11 to pick up the 30X relay which closes a contact in the white lite circuit.

All three lites will be LIT.

DISTRACTOR ANAL YSIS:

A. Incorrect, The Containment Air Return Fans receive a start signal 10 minutes after the Phase B isolation signal (2.8 psig) is initiated (not 10 minutes after the SI initiation) and all 3 lights on the handswitch will be lit (not just the white.) Plausible because the applicant could misapply the timer start time to the Hi containment pressure SI signal instead of the HI-HI (Phase B signal) and following the trip of many motors the green and white lights are the only lights LIT B. Incorrect, The Containment Air Return Fans receive a start signal 10 minutes after the Phase B isolation signal (2.8 psig) is initiated (not 10 minutes after the SI initiation) but all 3 lights on the handswitch will be lit. Plausible because the applicant could misapply the timer start time to the Hi containment pressure SI signal instead of the HI-HI (Phase B signal) and all 3 indicating lights being lit is correct.

C. Incorrect The Containment Air Return Fans do receive a start signal 10 minutes after the Phase B isolation signal (2.8 psig) is initiated, however all 3 lights on the handswitch will be lit, (not just the white.) Plausible because the start signal initiation is correct and and following the trip of many motors the green and white lights are the only lights LIT D. CorrecI, The Containment Air Return Fans are to maintain air flow through the ice condenser by pulling air from upper containment and discharge into lower containment during the accident. The fans receive a start signal 10 minutes after the Phase B isolation signal (2.8 psig) is initiated. Overload protection for the fan motor is provided by an amptector device and its oepration results in all 3 lights on the handswitch being lit.

Wednesday, June 05, 2013 8:16:18 AM 106

1305 NRC RO Exam Notes Question Number: 41 Tier: 2 Group 1 K/A: 025 Ice Condenser System A4.02 Ability to manually operate and/or monitor in the control room:

Containment vent fans Importance Rating: 2.7 / 2.5 10 CFR Part 55: 41.7 / 45.5 to 45.8 1OCFR55.43.b: Not applicable K/A Match: Question matches the K/A by requiring the applicant to be able to identify when the status of the Containment Air Return fans will change during a LOCA and be able identify a tripped condition of one of the fans after the automatic start is attempted.

Technical

Reference:

1 ,2-45N7791 R5 1 ,2-45N779-5 Ri 9 1 ,2-47W61 1-30-3 R6 Proposed references None to be provided:

Learning Objective: OPT200.ICE

6. EXPLAIN, (or SKETCH as applicable), the physical connections and/or cause-effect relationships between the Ice Condenser System and the following systems:
d. Containment ventilation Question Source:

New Modified Bank X Bank Question History: SQN bank question 025 A4.02 040 used on the SQN 1/2009 retake exam with the format changed along with 2 distractors and the stem modified.

Comments:

Wednesday, June 05, 2013 8:16:18 AM 107

1305 NRC RO Exam

42. 026 A1.02 042 Given the following plant conditions:

- Unit 1 was operating at 100% power when a design basis LOCA occurred inside containment.

- The operating crew is performing E-1, Loss of Reactor or Secondary Coolant.

Which ONE of the following completes the statement below?

A design basis of the containment spray system is to ensure the containment design temperature of (1) is not exceeded.

After an automatic initiation of the Containment Spray System, E-1 first allows the containment spray pumps to be stopped and placed in A-AUTO after the containment pressure drops to less than below (2)

A. 125°F 2.8 psig B. 125°F 2.0 psig C. 250°F 2.8 psig D 250°F 2.0 psig Feedback DIS TRACTOR ANAL YSIS:

A. Incorrect, The first part is plausible as 125 °F is the is the CNTMT lower compartment TS temperature limit. The second part is plausible as it would be logical to conclude that getting below the phase B setpoint of 2.8 psig would be the time to secure CNTMT spray.

B. lncorrect The first part is plausible as 125 °F is the is the CNTMT lower compartment TS temperature limit. The second part is correct.

C. Incorrect, The first part is correct The second part is plausible as it would be logical to conclude that getting below the phase B setpoint of 2.8 psig would be the time to secure CNTMT spray.

D. Correcl, The design basis temperature for CNTMT is 250 °F and per E- 1, CNTMT spray pumps are secured when CNTMT pressure is <2.0 psig.

Wednesday, June 05, 2013 8:16:18 AM 108

1305 NRC RO Exam Notes Question Number: 42 Tier: 2 Group 1 K/A: 026 Containment Spray System (CSS)

Al .02 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the CSS controls including:

Containment temperature Importance Rating: 3.6* / 3*9 10 CFR Part 55: 41.5 / 45.5 1OCFR55.43.b: Not applicable K/A Match: This question matches the KA by examining the applicants knowlegde of CNTMT temperature limits and criteria to secure CNTMT spray.

Technical

Reference:

FSAR E-l, Loss of Reactor or Secondary Coolant P25 Proposed references None to be provided:

Learning Objective: OPL271E-l

4. Summarize the mitigating strategy or E-l.

OPT200.CNTMTSTRUCTU RE Sd. Explain the CNTMT structure design features and/or operational interlocks that provide:

Tempurature and pressure control during normal and during DBA conditions.

Question Source:

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

Wednesday, June 05, 2013 8:16:18 AM 109

1305 NRC RO Exam

43. 026 K1.01 043 Which ONE of the following identifies an electrical interlock associated with opening 1-FCV-72-40, RHR Spray Header A Isolation?

1-FCV-72-40 cannot be opened unless A. 1-FCV-74-33, RHR Crosstie, is fully open B. 1-FCV-63-1, RWSTto RHR Suction, is fully open C. 1-FCV-74-3, RHR Pump lA-A Suction, is fully closed D 1-FCV-63-72, Containment Sump to RHR Pump lA-A Isolation, is fully open Feedback DIS TRACTOR ANAL YSIS:

A. Incorrect, Plausible because 1-FCV-74-33 is closed when prior to opening the RHR spray valve 1-FCV-72-40.

B. Incorrect, Plausible because 1-FCV-63-1 is normally open to supply all ECCS pump suction when the RWST is being used prior to swapover to the containment sump.

C. Incorrect, Plausible because 1-FCV-74-3 is the suction valve to the RHR pump and normally the suction valve would have to be open to supply suction to the pump.

0. Correct as shown on i-47W61 1-72-1, there is an electrical interlock that requires 1-FCV-63-72 to be fully open to allow 1-FCV-72-40 to be opened.

Wednesday, June 05, 2013 8:16:18 AM 110

1305 NRC RO Exam Notes Question Number: 43 Tier: 2 Group 1 K/A: 026 Containment Spray:

K1 .01 Knowledge of the physical connections and/or cause effect relationships between the CSS and the following systems:

ECCS Importance Rating: 4.2 / 4.2 10 CFR Part 55: 41.2 to 41.9/ 45.7 to 45.8 1OCFR55.43.b: Not applicable K/A Match: K/a is matched because the question requires applicant to recall interlock (cause and effect relationship) between a component in a containment spray subsystem and a component in the ECCS system.

Technical

Reference:

1-47W611-72-1 R13 Proposed references None to be provided:

Learning Objective: 6. EXPLAIN, (or SKETCH as applicable), the physical connections and/or cause-effect relationships between the Containment Spray system and the following systems:

a. RHR system
b. Containment Spray heat exchangers cooling water
c. Containment spray system fill and makeup water
d. Safety Injection
e. CVCS Question Source:

New Modified Bank Bank X Question History: WBN bank question used on the WBN 2006 exam.

Comments:

Wednesday, June 05, 2013 8:16:18 AM 111

1305 NRC RO Exam

44. 039 K3.06 044 Given the following:

- Unit 1 reactor power is stable at 90%.

- Turbine Impulse pressure transmitter 1-PT-i -72 fails LOW.

Which ONE of the following identifes the status of the white lights on 1-M-4 for (1) 1-XI-1-103D, STEAM DUMPS ACTUATED D FSVS ENERGIZED and (2) 1-XI-1-1O3AIB, STM DUMPS ARMED?

1-Xl-1-103D 1-Xl-1-103A!B A. DARK DARK B DARK LIT C. LIT DARK D. LIT LIT Feedback DISTRACTOR ANAL YSIS:

A. lncorrect, Plausible because 1-XI-103D being dark is correct and 1-XI-1-103A/B would be dark if the instrument had failed high.

B. CorrecI, 1-PT- 1-72 is the pressure transmitter that is used to sense a loss of load signal to arm the steam dumps system. When the transmitter fails low the steam dumps will arm resulting in 1 -Xl- 1-1 03A/B being energized but 1 -Xl- 1 03D will remain dark because there would not be a larger difference in temperature between Tavg and Tref created by the transmitter failure.

C. Incorrect, Plausible because 1-XI-103D would be lit and 1-Xl-1-103A/B would be dark if the failed transmitter had been the other Impulse Pressure Transmitter 1-PT-1-73.

D. lncorrect Plausible because both 1-XI-103D and 1-XI-1-103A/B would be lit during an actual rapid load reduction and the applicant could determine both are supplied form the same transmitter.

Notes Wednesday, June 05, 2013 8:16:18 AM 112

1305 NRC RO Exam Question Number: 44 Tier: 2 Group 1 K/A: 039 Main and Reheat Steam System (MRSS)

K3.06 Knowledge of the effect that a loss or malfunction of the MRSS will have on the following:

SDS Importance Rating: 2.8*/3.1 1OCFRPart55: 41.7/45.6 1OCFR55.43b: Not applicable K/A Match: K/A is matched because the question requires the knowledge of how the steam dumps system will be affected by a malfunction of a pressure transmitter on the Main and Reheat Steam System.

Technical

Reference:

1-47W611-1-2 R13 l,2-45N601-1 R3 1,2-45N601-1 R23 1-47W610-i-3 R5 0-50-1 -2, Steam Dump System, R12 OPT200SDCS, Steam Dump Lesson Plan Proposed references None to be provided:

Learning Objective: OPT200SDCS

  1. 8 Given specific plant conditions, analyze the effect that a loss or malfunction of the following will have on the SDCS:

PT-i -72/73 Cognitive Level:

Higher X Lower Question Source:

New Modified Bank Bank X Question History: WBN Bank question 039K3.06 045 used on the 06/2011 WBN NRC exam Comments:

Wednesday, June 05, 2013 8:16:18 AM 113

1305 NRC RO Exam

45. 039 K5.08 045 Given the following:

- Unit 2 startup in progress.

- The Main Generator has just been synchronized and loaded to 40 MWe.

- The Steam Pressure (2-PT-i -33) input to the steam dump controller fails HIGH.

Assuming NO action by the crew, which ONE of the following describes...

(1) the effect on core reactivity, and (2) how many of the steam dump valves would be responding to the failure?

A. Negative reactivity addition 3 valves only B. Negative reactivity addition All 12 valves C. Positive reactivity addition 3 valves only D Positive reactivity addition All 12 valves Wednesday, June 05, 2013 8:16:18 AM 114

1305 NRC RO Exam Feedback DIS TRACTOR ANAL YSIS:

A. Incorrect, the reactivity addition is not negative; It is positive due to the cooldown.

Only 3 valves responding is not correct but only 3 valves would respond under different conditions with the Steam Dumps in Tavg Mode and the Lo-Lo Tavg interlocked bypassed (by procedure the SDMSS switch would be in steam pressure mode at this power level).

B. Incorrect, the reactivity addition is not negative; It is positive due to the cooldown but all 12 dumps valves responding is correct.

C. Incorrect, the positive reactivity addition is correct but only 3 valves responding is not correct. Only 3 valves can respond under different conditions with the Steam dumps in Tavg Mode and the Lo-Lo Tavg interlocked bypassed.

D. Correct, If the pressure input falls high, the controller will try to reduce pressure by opening dump valves. This would cause a positive reactivity addition to the reactor core. All 12 of the steam dump valves would be opening as a result of the failure.

while the MSIVs would close eventually close to stop the cooldown, with the pressure transmitter failed high all 12 dump valves would be open.

Notes Question Number: 45 Tier: 2 Group 1 K/A: 039 Main and Reheat Steam System (MRSS)

K5.08 Knowledge of the operational implications of the following concepts as the apply to the MRSS:

Effect of steam removal on reactivity Importance Rating: 3.6 / 3.6 10 CFR Part 55: 41.5/45.7 1OCFR55.43.b: Not applicable K/A Match: Questions requires the applicant to determine the effect a failure on the steam dump system will have on core reactivity and how the failure will impact the steam dump system operation.

Technical

Reference:

1-47W611-1-2 R13 OPT200SDCS, Steam Dump Lesson Plan Proposed references None to be provided:

Wednesday, June 05, 2013 8:16:18 AM 115

1305 NRC RO Exam Learning Objective: OPT200.SDCS

5. EXPLAIN, (or SKETCH as applicable), the physical connections and/or cause-effect relationships between the Steam Dump Control System and the following systems:
i. Reactor Coolant System
7. EXPLAIN the Steam Dump Control System design features and/or interlocks that provide the following:
c. RCS Cooldown
e. Steam pressure control
f. Reactor temperature control
13. Given specific plant conditions, ANALYZE the effect that a loss or malfunction of the Steam Dump Control System will have on the following:
b. Reactor Coolant system
d. Reactor Power Question Source:

New Modified Bank Bank X Question History: SQN bank question 039 K5.08 042 used on the 02/2010 audit exam Comments:

Wednesday, June 05, 2013 8:16:18 AM 116

1305 NRC RO Exam

46. 059 A3.04 046 Given the following plant conditions:

- Unit 1 is at 72% power.

- PT-i -33, Steam Header Pressure to DOS, indicates 920 psig.

- PT-i -33A, Steam Header Pressure to DOS, indicates 900 psig.

- PT-i -33B, Steam Header Pressure to DOS, indicates 890 psig.

- PT-i -33 fails Low.

Which ONE of the following describes the effect this failure will have on the Unit 1 Main Feed Pump Turbine Master speed controller?

A. Output Increases.

B Output Decreases.

0. Output remains the same.

D. Transfers to Manual Feedback DIS TRACTOR ANAL YSIS:

A. Incorrect. Plausible if candiate confuses the selection criteria from median to average control input. For example, the examinee could conclude that with the channel failing low the average now is less and output would need to increase to maintain feedilow.

B. Correc1, Based on the design of the DCS with all three channels inputting to the OCS control, the circuit is designed to be a Median Select. Thus the Median signal would be 900 psig (the middle signal). With one channel failing, the logic goes to the average of the two remaining signals. The average of the two remaining signals would be 895 psig. Thus the controller would remain in automatic and the output would be going down (from 900 psig to 895 psig) and decrease the speed of MFP turbines.

C. Incorrect, Plausible since the control circuit could bypass the failed channel and go the single element control (another type of control in DCS for SG level control) and remain in Auto with no change in controller output.

0. lncorrecl, Plausible since the examinee could think that the failure caused the transfer feed pump speed control to manua this used to be the case with the previous control scheme.

Wednesday, June 05, 2013 8:16:18 AM 117

1305 NRC RO Exam Notes Question Number: 46 Tier: 2 Group 1 K/A: 059 Main Feedwater A3.04 Ability to monitor automatic operation of the MEW, including:

Turbine driven feed pump Importance Rating: 2.5 / 2.6 10 CFR Part 55: 41.7 1OCFR55.43.b: Not applicable K/A Match: This question matches the KJA by testing the candidates knowledge of how the inputs from steam header pressure are integrated automatically into MEP turbine speed control and how those inputs are processed when one signal falls outside of the tolerance band and then affects control of MEP turbine speed.

Technical

Reference:

OPT200.DCS/1 -SO-98 r6 Proposed references None to be provided:

Learning Objective: OPT200.DCS

  1. 9 State the DCS response to:

Failing of one of two channels Question Source:

New Modified Bank Bank X Question History: SQN Bank, SQN ILT 1305 NRC Exam Comments: High Cognitive Wednesday, June 05, 2013 8:16:18 AM 118

1305 NRC RO Exam

47. 061 K2.01 047 Which ONE of the following is the Alternate power supply for the Unit 2 TDAFW pump Trip and Throttle valve, 2-FCV-1-51A-S?

A. 125v DC Vital Board I B 125v DC Vital Board II C. 125v DC Vital Board Ill D. 125v DC Vital Board IV Feedback DISTRACTOR ANAL YSIS:

A. Incorrect, 125v DC Vital Board I is the normal power supply for the Unit 2 TD-AFW pump Trip and Throttle valve.

B. Correct, the alternate power supply for the Unit 2 TD-AFW pump Trip and Throttle valve, 2-FCV- 1-5 lA-S is from 125v DC Vital Board II.

C. Incorrect, 125v DC Vital Board III is the normal power supply for the Unit 1 TD-AFW pump Trip and Throttle valve.

0. Incorrect, 125v DC Vital Board IV is the alternate power supply for the Unit 1 TD-AFW pump Trip and Throttle valve.

Wednesday, June 05, 2013 8:16:18 AM 119

1305 NRC RO Exam Notes Question Number: 47 Tier: 2 Group 1 K/A: 061 Auxiliary / Emergency Feedwater (AFW) System K2.01 Knowledge of bus power supplies to the following:

AFW system MOVs Importance Rating: 3.2* / 33 1OCFRPart55: 41.7 1OCFR55.43.b: Not applicable K/A Match: Questions requires knowledge of the power supply to an MOV on the steam supply to Unit 2 TDAFW pump.

Technical

Reference:

AOP-P.02, Loss of 125V DC Vital Battery Board, Revision 13 Proposed references None to be provided:

Learning Objective: OPT200.DC

  1. 4 Explain the physical connections and/or cause and effect relationaships between DC and teh following systems:

AC electrical( DC Loads)

Question Source:

New X Modified Bank Bank Question History: New question for the SQN 05/2013 NRC exam.

Comments:

Wednesday, June 05, 2013 8:16:18 AM 120

1305 NRC RO Exam

48. 062 A2.09 048 Given the following plant conditions:

- Unit 1 is at 3% power.

- The monthly surveillance for lA-A DIG is in progress and the D/G is paralleled to the board.

- An inadvertent trip of 1 B 6.9kV Unit Board Normal Supply breaker results in operation of lA-A 6.9kV Shutdown Board DG Supply breaker 50 overcurrent relay.

Which ONE of the following completes the statements below?

The operation of the overcurrent relay will cause lA-A 6.9kV Shutdown Board DG supply breaker to trip open and iL The conditions above L the reactor to be tripped.

A. (1) lock-out (2) require B. (1) lock-out (2) do NOT require Cv (1) subsequently recloses in automatic (2) require D. (1) subsequently recloses in automatic (2) do NOT require Wednesday, June 05, 2013 8:16:18 AM 121

1305 NRC RO Exam Feedback DIS TRACTOR ANAL YSIS:

A. Incorrect, plausible as some overcurrent devices will lockout the board. The second part is correct.

B. Incorrect, plausible as some overcurrent devices will lockout the board. The second part is plausible as the unit is in Mode 2 which has the <P7 trips disabled, one of which is the loss of flow trip due to loss of one RCP.

C. Correct, with the diesel tied on to the SDB in parallel with offsite power the instantaneous 600 amp trip is in play. With a COW and RCP pump running the 600 amp 51 trip will drop the diesel supply breaker causing a load shed and stripping the bus. The diesel will then reclose (1912 breaker) and supply the SDB.

Procedurally in AOP-R.04, POP Malfunctions the crew is required to trip[ the reactor in step 2.

D. Incorrect, The first part is correct. The second part is plausible as the unit is in Mode 2 which has the <P7 trips disabled, one of which is the loss of flow trip due to loss of one RCP.

Wednesday, June 05, 2013 8:16:18AM 122

1305 NRC P0 Exam Notes Question Number: 48 Tier: 2 Group 1 K/A: 062 AC Electrical Distribution System A2.09 Ability to (a) predict the impacts on 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:

Consequences of exceeding current limitations Importance Rating: 2.7 / 3.0*

10 CFR Part 55: 41.5 1OCFR55.43.b: Not applicable K/A Match: The question meets the KA as it requries to determine that an overcurrent condition exist, actions occurr and actions must be taken to mitigate the event.

Technical

Reference:

AOP-R.04 R27 AR-Mi-B, B-i R23 45N724-1, R2i 45N765-i, R24 45N765-2, R25 Proposed references None to be provided:

Learning Objective: OPL271AOP-R.04

  1. 9. List any conditions(s) that requires a Reactor or Reactor Coolant Pump trip in AOP-R.04.

OPT200.AC6.9KV Explain the operational implication of the following concept as it applies to the 6.9 KV Distribution system:

f. exceeding current limitations Question Source:

New Modified Bank Bank X Question History: WBN Bank Comments:

Wednesday, June 05, 2013 8:16:18 AM 123

1305 NRC RO Exam

49. 063 A1.01 049 Which ONE of the following identifies the required capacity for the 125v Vital DC batteries to satisfy coping time requirements?

A. Must be able to mitigate a Station Blackout event for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> without any required operator action relating to the 125v Vital DC system.

B. Must be able to mitigate a Station Blackout event for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> without any required operator action relating to the 125v Vital DC system.

C Must be able to mitigate a Station Blackout event for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and to meet the requirement, loads must be stripped from the batteries within 45 minutes into the Station Blackout.

D. Must be able to mitigate a Station Blackout event for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and to meet the requirement, loads must be stripped from the batteries within 45 minutes into the Station Blackout.

Feedback DIS TRACTOR ANAL YSIS:

A. Incorrect, FSAR 8.3.2.1.1 identifies that the safety related DC power system is required to mitigate a station blackout event for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> (not 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) and that the stripping of loads is necessitated before 45 minutes into the event.

B. Incorrect, Being required for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is correct in accordance with FSAR 8.3.2.1.1, however, operator actions are required to strip of loads prior to 45 minutes into the event.

C. Correct, FSAR 8.3.2.1.1 identifies that the safety related DC power system is required to mitigate a station blackout event for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and that the stripping of loads is necessitated before 45 minutes into the event. The capacity of the 125v Vital DC batteries is that with the batteries in a fully charged condition each batteiy has the capacity to supply the connected loads for 45 minutes and to supply a reduced load for an additional 195 minutes during a loss of all AC power.

D. Incorrect, FSAR 8.3.2.1.1 identifies that the safety related DC power system is required to mitigate a station blackout event for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> (not 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />). The stripping of loads being necessary before 45 minutes into the event is correct.

Wednesday, June 05, 20138:16:19AM 124

1305 NRC RO Exam Notes Question Number: 49 Tier: 2 Group 1 K/A: 063 D.C. Electrical Distribution Al .01 Ability to predict and/or monitor changes in parameters associated with operating the DC electrical system controls including:

Battery capacity as it is affected by discharge rate Importance Rating: 2.5 / 3.3 1OCFR Part 55: 41.5/45.5 1OCFR55.43.b: Not applicable K/A Match: Question requires applicant to predict the manual action required to reduce the discharge rate on the battery to ensure the Vital DC electrical distribution system is able to maintain voltage throughout its required Coping Time Technical

Reference:

FSAR 8.3.2.1.1 Proposed references None to be provided:

Learning Objective: OPT200.DC

7. EXPLAIN the operational implication of the following concept as it applies to the DC Systems:
c. Discharge rate effect on battery capacity Question Source:

New Modified Bank Bank X Question History: SQN bank question used on SQN1/2009 Audit exam Comments:

Wednesday, June 05, 2013 8:16:19 AM 125

1305 NRC AC Exam

50. 063 K3.02 050 Given the following plant conditions:

- Unit 1 was operating at 100 % power when a safety injection occurred.

- Eighteen (18) seconds after Safety Injection, a loss of 125v Vital DC Power Channel II occurs.

Which ONE of the following identifies the current status of RHR pump 1 B-B?

A. RHR pump 1 B-B is NOT running but can be started from the MCR handswitch.

B. RHR pump 1 B-B is NOT running and can NOT be started from the MCR handswitch.

C. RHA pump 1 B-B is running and can be stopped from the MCR handswitch.

D RHR pump 1 B-B is running but can NOT be stopped from the MCR handswitch.

Feedback DIS TRACTOR ANAL YSIS:

A. Incorrect, Plausible if the time delays associated with the pump starting with a blackout present are used. The delay times would exceed the 18 seconds (DG Start and RHR blackout time delay relay) and there are 3 other Vital DC boards available to supply the control power. One of which does supply the breaker but it is a manual transfer, not an automatic transfer.

B. IncorrecI, Plausible since RHR pump 18-B cannot be started or stopped from the control room handswitch after the loss of 125v DC Vital channel!! and because if a blackout signal had been concurrent with the SI condition, then the DG start time and pump start delay time would have exceeded the time prior to the loss of the control power.

C. Incorrect, Plausible because the pump being running is correct and there are 3 other channels of 125v DC available that could have been determined to be the control power supply for the pumps breaker.

D. Correct, RHR pump 18-B would have started immediately when the Safety injection was initiated but after the 125v DC Channel!! power was losI the pump could not be stopped from its handswitch in the math control room.

Wednesday, June 05, 2013 8:16:19 AM 126

1305 NRC RO Exam Notes Question Number: 50 Tier: 2 Group 1 K/A: 063 D.C. Electrical Distribution K3.02 Knowledge of the effect that a loss or malfunction of the DC electrical system will have on the following:

Components using DC control power Importance Rating: 3.5 / 3.7 1OCFRPart55: 41.7/45.6 1OCFR55.43b: Not applicable K/A Match: K/A is matched because the question requires the applicant to know a major breaker supplied with control power from 125v DC Vital Channel II and how a loss of the power supply to the control power affects the ability to start and stop the component.

Technical

Reference:

AOP-P.02, Loss Of 125V DC Vital Battery Board R2 1,2-45N765-13 R18 1 -45N724-2 R22 Proposed references None to be provided:

Learning Objective: OPL271AOP-P.02

  1. 11, Given a specific de-energized DC board, describe the affect on plant equipment and operations.

Cognitive Level:

Higher X Lower Question Source:

New Modified Bank Bank X Question History: WBN bank question 063 K3.02 050 used on the 10/2011 exam 10-2009.

Comments:

Wednesday, June 05, 2013 8:16:19 AM 127

1305 NRC RO Exam

51. 064 K3.03 051 The following plant conditions exist:

Unit 1 is at 100% rated thermal power Diesel Generator (DG) lA-A is being tested for its monthly surveillance test which requires loading to 4400 kw.

Current load is 2000 kw.

Subsequently, 1 B-B 6.9KV SDBD experiences a complete loss of voltage.

Which of the following describes the control the operator has over continued loading of the lA-A DG?

A. lA-A DG j be loaded by the operator to 4400kw, VAR loading cannot be changed by the operator.

B lA-A DG can be loaded by the operator to 4400kw, VAR loading jj be changed by the operator.

C. lA-A DG cannot be loaded by the operator to 4400kw, VAR loading cannot be changed by the operator.

D. lA-A DG cannot be loaded by the operator to 4400kw, VAR loading can be changed by the operator.

Feedback DIS TRACTOR ANAL YSIS:

A. Incorrect, plausible as the first part is correct. The second part is plausible as the LOR 86 relay does on certain conditions remove the load and voltage control from the MCR.

B. Correct, the DG while under test and parallel with offsite power, relay 62X prevents the 86 LOR from going to trip from the CES signal so the operator maintains control of both load and voltage and the DG stays in droop mode.

C. Incorrect, plausible as the LOR 86 relay does on certain conditions remove the load and voltage control from the MCR.

D. Incorrect, plausible as the LOR 86 relay does on certain conditions remove the voltage control from the MCR.

Wednesday, June 05, 2013 8:16:19 AM 128

1305 NRC RO Exam Notes Question Number: 51 Tier: 2 Group 1 K/A: 064 Emergency Diesel Generators (ED/G)

K3.03 Knowledge of the effect that a loss or malfunction of the ED/G system will have on the following:

ED/G (manual loading)

Importance Rating: 3.6 / 3*9*

10 CFR Part 55: 41.7 1OCFR55.43.b: n/a K/A Match: The question matches the K/A by testing the candidates knowledge of operation of the ED/G while being manually loaded and requiring understanding of the DG control circuits under a malfunction under manual load conditions.

Technical

Reference:

DWG. 45N767-4 R21(LOR 86 relay Ckt., PM)

Proposed references None to be provided:

Learning Objective: OPT200.DG Obj 6.c Explain the DG design features and/or operational interlocks that provide the following: Parallel operation, local and remote.

Cognitive Level:

Higher X Lower Question Source:

New X Modified Bank Bank Question History: New for ILT 1305 NRC Exam Comments:

Wednesday, June 05, 20138:16:19 AM 129

___________

__

1305 NRC RO Exam

52. 073 K1.01 052 Given the following plant alarms:

- 0-RA-90-125A, MAIN CNTRL RM INTAKE MON HIGH RAD

- 0-RA-90-126A, MAIN CNTRL RM INTAKE MON HIGH RAD Which ONE of the following describes the Main Control Room ventilation alignment?

The MCR is maintained at a A. positive pressure by the Main Control Room Air Handling Units B. negative pressure by the Main Control Room Air Handling Units C positive pressure by the Control Building Emergency Air Pressurization Fans D. negative pressure by the Control Building Emergency Air Pressurization Fans Feedback DIS TRACTOR ANAL YSIS:

A. lncorrect, RM 125 and 126 initiate a CR1. A CR1 isolates the ductwork from the Control Building Air Pressurization Fans and start the Control Building Emergency Air Pressurization Fans. A positive pressure is maintained to minimize outside contaminants from entering the MCR. Plausible since this is the normal ventilation flowpath/pressure condition in MCR.

B. Incorrect, RM-90-125 and 126 initiate a CR! A CR! isolates the ductwork from the Control Building Air Pressurization Fans and start the Control Building Emergency Air Pressurization Fans. A positive pressure is maintained to minimize outside contaminants from entering the MCR. Plausible since MCR is a slight negative pressure during normal operation and normal ventilation flowpath.

C. Correct, RM 125 and 126 initiate a CR!. A CR! isolates the ductwork from the Control Building Air Pressurization Fans and start the Control Building Emergency Air Pressurization Fans. A positive pressure is maintained to minimize outside contaminants from entering the MCR.

D. lncorreci, RM 125 and 126 initiate a CR1. A CR! isolates the ductwork from the Control Building Air Pressurization Fans and start the Control Buildhg Emergency Air Pressurization Fans. A positive pressure is maintained to minimize outside contaminants from entering the MCR. Plausible since MCR is a slight negative pressure during normal operation.

Wednesday, June 05, 2013 8:16:19AM 130

_____

1305 NRC RO Exam Notes Question Number: 52 Tier: 2 Group 1 K/A: 073 Process Radiation Monitoring (PRM) System Ki .01 Knowledge of the physical connections and/or cause-effect relationships between the PRM system and the following systems:

Those systems served by PRMs Importance Rating: 3.6 / 3.9 10 CFR Part 55: 41.2 to 41.9 / 45.7 to 45.8 1OCFR55.43.b: Not applicable K/A Match: This question matches the K/A by having the candidate determine the cause-effect of Pad Monitors RM-90-125 & 126 and the Control Room Ventilation system.

Technical

Reference:

47W61 1-31-1 rev 28 0-XA-55-12B rev 29 Proposed references None to be provided:

Learning Objective: OPT200.RM Ob] 4.m Explain the physical connections and/or cause-effect relationships between the Radiation Monitoring System and the following systems:

Control Building Vent OPT200.CBVENT Obj 4.e Explain the physical connections and/or cause-effect relationships between the CBVENT and the following systems:

Control Room Emergency Ventilation (CREV) alignment/flow-path following initiation of Control Room Isolation (CR1) signal.

Question Source:

New Modified Bank Bank X Question History: SQN bank question 073 Ki .01 052 used on the SON 09/2010 NRC exam.

Comments:

Wednesday, June 05, 20138:16:19 AM 131

1305 NRC RO Exam

53. 076 A4.02 053 Which ONE of the following identifies how 1 -FCV-67-66A, ERCW HDR 1 A SUP TO DG1A-A HX Al &A2 responds to a normal start and stop of Diesel generator lA-A?

The ERCW valve will automatically open after the start signal when the DG speed rises to iL When the diesel generator is stopped, the ERCW valve will be closed t?L.

LU START STOP Av 40 rpm using the handswitch on 0-M-26 B. 40 rpm automatically when DG stops following the idle speed run C. 200 rpm using the handswitch on 0-M-26 D. 200 rpm automatically when DG stops following the idle speed run Feedback DISTRACTOR ANAL YSIS:

A. Correct, Speed switch 1 is picked up when the DG speed rises to 40rpm and one of its functions is to open the ERCW normal supply valve to provide cooling to the EDG. Following the diesel generator being stopped the valve remains open until it is closed by operator action.

B. Incorrect, Plausible because the valve is opened automatically by a speed switch that actuates when the speed rises to 40 rpm and there is an interconnection between the valve and the idle run in the SO instruction. However, it is to verify the valve is opened during idle run checks following a DG start, not to verify the valve closed following the idle run during shutdown.

C. Incorrec1, Plausible because the valve is opened automatically by a speed switch but it is not the 200 rpm speed switch and the valve being required to be closed manually is correct.

0. Incorrect, Plausible because the valve is opened automatically by a speed switch but it is not the 200 rpm speed switch and there is an interconnection between the valve and the idle run in the SO instruction. Howe ver it is to verify the valve is opened during idle run checks following a DG stari, not to verify the valve closed following the idle run during shutdown.

Wednesday, June 05, 2013 8:16:19 AM 132

1305 NRC RO Exam Notes Question Number: 53 Tier: 2 Group 1 K/A: 076 Service Water System (SWS)

A4.02 Ability to manually operate and/or monitor in the control room:

SWS valves Importance Rating: 2.6 / 2.6 10 CFR Part 55: 41.7/45.5 to 45.8 1OCFR55.43.b: Not applicable K/A Match: the question requries the ability to monitor the automatic oepration of an ERCW (Service Water System) valve that supply water to an EDG and a condition that will require manual operation of the valve.

Technical

Reference:

0-SO-82-1, Diesel Generator lA-A, rev 0040 45N767-3 rev 24 Proposed references None to be provided:

Learning Objective: OPT200.DG Obj 4.g Explain the physical connections and/or cause-effect relationships between the DGs and the following systems or subsystems:

ERCW Question Source:

New X Modified Bank Bank Question History: new question for the SQN Comments:

Wednesday, June 05, 2013 8:16:19 AM 133

1305 NRC RO Exam

54. 078 G2.4.35 054 Given the following plant conditions:

- Unit 1 was at 100% power when a LOCA occurred.

- AUOs are performing EA-32-1, Establishing Control Air to Containment, Appendix A, Re-establishing Air to Containment on Unit 1.

- Steps for opening FCV-32-80, Rx Bldg Train A Essential Air, are in progress.

The reason for holding 1-HS-32-80A in OPEN while depressing the test button for 30 seconds is to ensure A. Phase B closure signal for FCV-32-80 is reset B. auxiliary air compressor shutdown relays are bypassed while compressor loads C header pressure downstream of FCV-32-80 is greater than 50 psig D. Train A essential air pressure on 0-PI-32-1 04A remains greater than 77 psig Feedback DIS TRACTOR ANAL YSIS:

A. Incorrect. The Phase B closure signal is verified reset when the Phase B is reset in step 2 of EA 1. This is plausible if examinee determines that holding the test button will bypass the Phase B signal.

B. Incorrect. The aux air compressor shutdown relays are not bypassed by this action. This is plausible if examinee determines that since these actions are contained in an Emergency Abnormal (EA) procedure, special actions are needed to facilitate restoration of air to containment.

C. Correct. The circuit for controlling the position of FCV-32-80, Rx Bldg Train A Essential Air(as well as 80 and 102) require that pressure downstream of the valve be greater than 50 psig to allow the valve to remain open.

0. Incorrect. The essential air pressure is an input for maintaining the containment air isolation valves in the open position however this pressure is upstream of FCV-32-80 and is not part of the valve position curcuit. This is plausible since this value is used in EA 1 for ensuring Aux Air Compressors are running and air dryers are properly configured.

Wednesday, June 05, 2013 8:16:19 AM 134

1305 NRC RO Exam Notes Question Number: 54 Tier: 2 Group 1 K/A: 078 Instrument Air System G2.4.35 Knowledge of local auxiliary operator tasks during an emergency and the resultant operational effects.

Importance Rating: 3.8 / 4.0 1OCFR Part 55: 41.10/43.5/45.13 1OCFR55.43.b: Not applicable K/A Match: Question matches the K/A by since the examinee is required to recall the local actions required re-establish air to containment following an Phase B.

Technical

Reference:

1 ,2-47W61 1-32-2 Rev 9 EA-32-1, Establishing Control Air to Containment Proposed references None to be provided:

Learning Objective: OPT200.CSA Obj 7.d Explain the Control and Service Air System design features and/or interlocks that provide the following:

Automatic isolation of sections of the air system.

Question Source:

New X Modified Bank Bank Question History: SQN ILT 1305 Comments:

Wednesday, June 05, 20138:16:19 AM 135

1305 NRC RO Exam

55. 103 K1.O1 055 Given the following:

- In accordance with 0-SO-30-5, Lower Compartment Cooling Units, which of the following identifies:

(1) the preferred cooler that should always be in service for Unit 1 and (2) if this cooler trips, what effect this will have?

A. (1)A-A (2) PZR Enclosure heats up B. (1)B-B (2) PZR Eclosure heats up.

C. (1)A-A (2) Auto start of the Standby Cooler D. (1)B-B (2) Auto start of the Standby Cooler Wednesday, June 05, 20138:16:19 AM 136

1305 NRC RO Exam Feedback DIS TRACTOR ANAL YSIS:

A. Correct, In accordance with O-SO-30-5 the preferred CNMT cooler is the A-A cooler due to its location to the exhaust plenums that supply air flow to the PZR enclosure. Also as indicated in O-SO-30-5, if the A-A cooler is not in service then all other coolers must be started to provide adequate air flow to the PZR enclosure or the equipment in the enclosure will start to heat up and instrument readings may become erratic.

B. Incorrect, Plausible since the B-B cooler is located near the A-A cooler and since it is on plenun that supplies air to the PZR enclosure it would be a logical choice. Also since the B-B cooler is physically closer to the PZR enclosure the candidate would conclude that if the B-B cooler tripped the PZR enclosure would heat up. However the A-A cooler is the preferred cooler.

C. Incorrect, Plausible since the A-A cooler is the preferred cooler. Also plausible since there is an Auto Start associated with the CNCT Coolers and a trip of the running cooler would cause low air flow which would start the cooler selected for Standby, however this feature is overriden by the switch position in the control room such that no CNMT cooler auto start would be permitted.

0. Incorrect, Plausible since the B-B cooler is located near the A-A cooler and since it is on plenun that supplies air to the PZR enclosure it would be a logical choice. Also plausible since there is an Auto Start associated with the CNCT Coolers and a trip of the running cooler would cause low air flow which would start the cooler selected for Standby, however this feature is overriden by the switch position in the control room such that no CNMT cooler auto start would be permitted.

Wednesday, June 05, 20138:16:19 AM 137

1305 NRC RO Exam Notes Question Number: 55 Tier: 2 Group 1 K/A: 103 Containment System Ki .01 Knowledge of the physical connections and/or cause-effect relationships between the Containment System and the following systems:

CCS Importance Rating: 3.6 / 3.9 1OCFRPart55: 41.2to41.9 1OCFR55.43.b: Not applicable K/A Match: This question matches the K/A by testing the candidates knowledge of the cause-effect relationship between the Containment Cooling System unit differences and the Containment System.

Technical

Reference:

0-S0-30-5, Lower Compartment Cooling Units, P35 1 (2)-47W845-3 ERCW Proposed references None to be provided:

Learning Objective: OPT200.CNTMTCLG & PURGE Obj 9.d Given specific plant conditions, analyze the effect that a loss or malfuction of the Containment Cooling and Purge Systems will have on the following:

Containment parameters (pressure, temperature and humidity)

Question Source:

New X Modified Bank Bank Question History: New question written for the 1305 NRC exam Comments:

Wednesday, June 05, 20138:16:19 AM 138

1305 NRC RO Exam

56. 002 K5.07 056 Given the following plant conditions:

- A Unit 1 reactor start up following a refueling outage is in progress lAW 0-GO-2, Unit Startup from Hot Standby to Reactor Critical.

- RCS boron concentration was 1350 ppm for the ECP calculation.

- Just prior to beginning rod withdrawal for the startup chemistry reports the following boron concentrations:

- RCS boron concentration: 1380 ppm

- Pressurizer boron concentration: 1325 ppm Which one of the following completes the statements below?

Critical rod height will be (1)

PZR boron concentration (2) required to be raised before the startup can continue.

A (1) higher (2) is B. (1) higher (2) is NOT C. (1) lower (2) is D. (1) lower (2) is NOT Wednesday, June 05, 20138:16:19 AM 139

1305 NRC RO Exam Feedback DISTRACTOR ANAL YSIS:

A. Correct: A higher RCS boron concentration will require the control rod withdrawal to be greater to overcome the added negative reactivity from it. 0-GO-2 requires that the RCS and PZR be equalized if they are not within 50 ppm of each other.

B. Incorrect, Plausible as the first part is correct. The second part is plausible as the examinee will have to remember the 50 ppm requirement within the startup procedure or will determine that the delta between the RCS and PZR is within limits.

C. Incorrect, Plausible if the examinee does not understand that a lower boron concentration will require less negative reactivity for the control rods to overcome and the rod height at criticality will be lower than the ECP. The second part is correct.

0. Incorrect, Plausible if the examinee does not understand that a lower boron concentration will require less negative reactivity for the control rods to overcome and the rod height at criticality will be lower than the ECP. The second part is plausible as the examinee will have to remember the 50 ppm requirement within the startup procedure or will determine that the delta between the RCS and PZR is within limits.

Wednesday, June 05, 2013 8:16:19 AM 140

1305 NRC RO Exam Notes Question Number: 56 Tier: 2 Group 2 K/A: 002 Reactor Coolant System (RCS)

K5.07 Knowledge of the operational implications of the following concepts as they apply to the RCS: Reactivity effects of RCS boron, pressure and temperature Importance Rating: 3.6 / 3.9 10 CFR Part 55: CFR: 41.5/45.7 1OCFR55.43.b: Not applicable K/A Match: The question matches the KA by requiring the examinee to understand reactivity effects on a reactor startup with changing boron concentrations in the RCS and PZR.

Technical

Reference:

0-GO-2, Unit Startup from Hot Standby to Reactor Critical R35 Proposed references None to be provided:

Learning Objective: OPT200. RCS 9.b Given plant conditions, identify and apply the following RCS limits and precautions related to the following:

GO-2, Unit Startup from Hot Standby to Reactor Critical Question Source:

New X Modified Bank Bank Question History: New for SQN ILT 1305 NRC Exam Comments: Low Cognitive Wednesday, June 05, 20138:16:19 AM 141

1305 NRC RO Exam

57. 015 A1.02 057 Given the following plant conditions:

- A reactor start up without Physics Testing is in progress lAW 0-GO-2, Unit Startup from Hot Standby to Reactor Critical.

- The crew has blocked P-6.

Which one of the following is the lowest of the listed values that exceeds the maximum SUR limit allowed per 0-GO-2?

A. 0.4 dpm B. 0.6 dpm C. 0.8 dpm D 1.1 dpm Feedback DISTRACTOR ANAL YSIS:

A. lncorrect, Plausible as 0.3 is the expected targeted SUR as delineated by 0-GO-2 in the section for raising power from criticality to block P-6.

B. incorrect, Plausible as 0.5 is the expected targeted SUR as delineated by 0-GO-2 in the section for raising power greater than P-6 up to the POAH.

C. Incorrect, plausible as 0.8 is higher than both 0.3 dpm and .5 dpm.

D. Correct, 1.0 dpm is the maximum limit of SUR listed in 0-GO-2.

Wednesday, June 05, 20138:16:19 AM 142

1305 NRC RO Exam Notes Question Number: 57 Tier: 2 Group 2 KIA: 015 Nuclear Instrumentation System Al .02 Ability to predict and/or monitor changes in parameters to prevent exceeding design limits) associated with operating the NIS controls including: SUR Importance Rating: 3.5 / 3.6 10 CFR Part 55: (CFR: 41.5.45.5) 1OCFR55.43.b: Not applicable K/A Match: The question places the examinee in an area of a reactor startup where the procedure has specific target SURs that should be expected and monitored. It also requires the examinee to know the maximu, SUP allowed.

Technical

Reference:

0-GO-2, Unit Startup from Hot Standby to Reactor Critical R35 Proposed references None to be provided:

Learning Objective: OPL271GO-2 Obj 10 State the startup rate limit.

Question Source:

New X Modified Bank Bank Question History: New qusetion for SQN ILT 1305 exam Comments:

Wednesday, June 05, 20138:16:19 AM 143

1305 NRC RO Exam

58. 016 A3.01 058 Given the following plant conditions:

- Unit 1 is operating steady state at 70% reactor power.

- Rod control is in MANUAL.

Compare the effects of either one of the following RCS Loop 1 RTD5 failing HIGH.

1. ThotRTD#1
2. Tcold RTD #1 Assuming NO operator action, which ONE of the following identifies the RCS RTD failure...

(1) having the larger effect on the pressurizer level control system and (2) how the pressurizer level would be affected?

Largest effect Level would...

A. Thot failure rise and be controlled at a level higher than the 70% power steady state level.

B. Thot failure rise but be restored to the 70% power steady state level by the control system.

C Tcold failure rise and be controlled at a level higher than the 70% power steady state level.

D. Tcold failure rise but be restored to the 70% power steady state level by the control system.

Wednesday, June 05, 20138:16:19 AM 144

1305 NRC RO Exam Feedback DISTRACTOR ANAL YSIS:

A. Incorrect Plausible because the Thot failure would cause a change in the loop Tavg if the failed RTD was not rejected in the Tavg calculation and the change would then cause the pressurizer level control program setpoint to increase to a higher setpoint which would result in a charging flow increase to increase the pressurizer level.

B. Incorrec1, Plausible because the Thot failure would cause a change in the loop Tavg if the failed RTD was not rejected in the Tavg calculation and because it is rejected the level would return to the steady state 70% power setpoint after the failure.

C. Correct, The Tcold failure will cause the Auctioneered High RCS Tavg to signal to rise. This signal programs the pressurizer level control program setpoint resulting in an increase in the setpoint. The charging flow will increase to bring the pressurizer level up to the new setpoint.

D. Incorrect, Plausible because the Tcold failure having the largest effect is correct and because the level would return to the steady state 70% power setpoint if the failed RTD had been rejected in the Tavg calculation as is the Thot RTD failure.

Wednesday, June 05, 2013 8:16:19AM 145

1305 NRC RO Exam Notes Question Number: 58 Tier: 2 Group 2 K/A: 016 Non-Nuclear Instrumentation System (NNIS)

A3.01 Ability to monitor automatic operation of the NNIS, including:

Automatic selection of NNIS inputs to control systems Importance Rating: 2.9* / 2.9*

10 CFR Part 55: CFR: 41.7/45.5 1OCFR55.43.b: Not applicable K/A Match: This question matches the K/A by testing the candidates knowledge of how the inputs from loop temperature RTDs are selected automatically under a failure condition.

Technical

Reference:

1-AR-M5-A D-6 R36 47W610-68-1 thru 10 Proposed references None to be provided:

Learning Objective: OPT200.EAGLE 21 Obj 4 Explain the physical connections and/or cause-effect relationships between the Eagle 21 system and the following:

RCS transmitters for indication, control and protection.

Question Source:

New Modified Bank Bank X Question History: WBN 06/2011 NRC exam, SQN ILT 1305 NRC Exam Comments: High Cognitive Wednesday, June 05, 2013 8:16:19 AM 146

1305 NRC RO Exam

59. 027 K2.0I 059 Given the following conditions:

- Unit 1 is in Mode 3 when a Safety Injection occurs.

- Following the Safety Injection, both the lA-A 6.9KV Shutdown Board and the 2B-B 6.9 KV Shutdown Boards trip on differential.

Which ONE of the following describes the status of the EGTS system?

A. Neither EGTS fans are running.

B Only EGTS train B fan is running.

C. Only EGTS train A fan is running.

D. Both trains EGTS fans are running.

Feedback DISTRACTOR ANAL YSIS:

A. Incorrect, the EGTS fans do receive shutdown power and for neither to be running is plausible if the source of shutdown power to the 2 fans is thought to be from boards being fed from the 2 boards that are de-energized.

B. Correct, EGTS Train B is supplied from Unit 1 Train B Shutdown Power which is energized, but EGTS Train A is supplied from the Unit 1 Train A Shutdown power which is not energized.

C. Incorrect, Plausible if the EGTS fan power supplies are confused with the ABGTS fan power supplies. The ABGTS fans are supplied from Unit 2 and only the Train A ABGTS fan would be running with the 2 boards de-energized as stated in the question.

D. lncorrect the EGTS fan do receive shutdown power and for both to be running is plausible if the source of shutdown power to the 2 fans is thought to be from boards being fed from the 2 boards that remain energized.

Wednesday, June 05, 2013 8:16:20 AM 147

1305 NRC RO Exam Notes Question Number: 59 Tier: 2 Group 2 K/A: 027 Containment Iodine Removal System (CIRS)

K2.01 Knowledge of bus power supplies to the following:

Fans Importance Rating: 3.1*/3.4*

10 CFR Part 55: 41.7 1OCFR55.43.b: Not applicable K/A Match: Question requires knowledge of the power supplies to the Emergency Gas Treatment System fans in order to determine which fans will be in service.

Technical

Reference:

AOP-P.05 rev 20 (App A, B and G for fan A, App L, M and R for fan B)

Proposed references None to be provided:

Learning Objective: OPT200.EGTS

5. LIST the bus power supplies to the following EGTS components:
a. EGTS Fans Question Source:

New Modified Bank Bank X Question History: WBN bank question 027K2.01 060 used on the WBN 05/2009 exam.

Comments:

Wednesday, June 05, 2013 8:16:20 AM 148

1305 NRC RO Exam

60. 033 A2.01 060 Given the following plant conditions:

- Unit 1 is in Mode 6.

- The current Spent Fuel Pool boron concentration is 2120 ppm.

- Refueling Water Storage Tank (RWST) is being used as the TRM borated water source and is at the lowest boron concentration allowed as identified in SO-78-1 ,Spent Fuel Pit Coolant System.

- A leak on the Spent Fuel Pool cooling system results in the need for makeup from the RWST.

Which ONE of the following completes the statement below relative to the Spent Fuel Pool?

Using the RWST to makeup to the Spent Fuel Pool will result in alan in the Spent Fuel Pool boron concentration.

To meet LCD 3/4.7.13, SPENT FUEL MINIMUM BORON CONCENTRATION, requires a minimum boron concentration of { ppm.

A. decrease 1950 B. decrease 2000 C. increase 1950 D increase 2000 Wednesday, June 05, 2013 8:16:20 AM 149

1305 NRC RO Exam Feedback DISTRACTOR ANAL YSIS:

A. Incorrect, The first part is incorrect. Tech Spec requires RWST to be greater than 2500 ppm and the SFP to be greater than 2000 ppm. Plausible if you use the Tech Spec value for SFP boron (2000 ppm) and the minimum assumed boron concentration for RWST used in UFSAR Chapter 15 Accident Analysis (1950 ppm).

The second part is incorrect. Tech Spec requires SFP to be >2000 ppm. This is plausible the minimum assumed boron concentration for RWST used in UFSAR Chapter 15 Accident Analysis (1950 ppm), which could also apply to the SFP since the UFSAR states that SPF Keff will remain 0.95 with SFP boron at 700 ppm under the most severe postulated fule mis-location accident.

B. Incorrect, The first part is incorrect. Tech Spec requires RWST to be greater than 2500 ppm and the SFP to be greater than 2000 ppm. Plausible if you use the Tech Spec value for SFP boron (>2000 ppm) and the minimum assumed boron concentration for RWST used in UFSAR Chapter 15 Accident Analysis (1950 ppm).

The second part is correct. Tech Spec requires SFP to be >2000 ppm.

C. Incorrect, The first part is correct. In accordance with SQ-78-1, the RWST will be

> 2500 ppm, thus the addition of RWST water will raise the SFP boron concentration. The second part is incorrect. Tech Spec requires SFP to be >2000 ppm. This is plausible the minimum assumed boron concentration for RWST used in UFSAR Chapter 15 Accident Analysis (1950 ppm), which could also apply to the SFP since the UFSAR states that SPF Keff will remain <0.95 with SFP boron at 700 ppm under the most severe postulated fule mis-location accident.

D. Correct The first part is correct. In accordance with SO-78-1, the RWST will be>

2500 ppm, thus the addition of RWST water will raise the SFP boron concentration.

The second part is correct. Tech Spec requires SFP to be >2000 ppm.

Wednesday, June 05, 2013 8:16:20 AM 150

1305 NRC RO Exam Notes Question Number: 60 Tier: 2 Group 2 K/A: 033 Spent Fuel Pool Cooling System (SFPCS)

A2.01 Ability to (a) predict the impacts of the following malfunctions or operations on the Spent Fuel Pool Cooling System ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Inadequate SDM.

Importance Rating: 3.0 / 3.5 10 CFR Part 55: 41.5 / 43.5/45.3/45.13 1OCFR55.43.b: Not applicable K/A Match:

Technical

Reference:

Tech Spec 3.7.13 0-SO-78-1 rev 59 UFSAR rev 22, Chapter 4 Nuclear Design pg 4.3-29 UFSAR rev 22, Chapter 15 Accident Analysis pg 15.2-37 Proposed references None to be provided:

Learning Objective: OPT200.SFPC Obj 8.e Explaing the Spent Fuel Pit Cooling system design features and/or interlocks that provide the following:

Adequate Shutdown Margin (boron concentration)

Question Source:

New X Modified Bank Bank Question History: New for NRC ILT 1305 exam Comments: Low Cog Wednesday, June 05, 2013 8:16:20 AM 151

1305 NRC RO Exam

61. 034 K6.02 061 Given the following plant conditions:

- Unit 1 is at 100% RTP.

- Fuel Assembly shuffles are being made in the Spent Fuel Pit.

- 0-RM-90-102, Spent Fuel Pit Radiation Monitor, has been declared INOPERABLE and removed from service due to an instrument malfunction.

Which ONE of the following completes the statements below?

Technical Specifications would continued movement of fuel assemblies in the Spent Fuel Pit.

If 0-RM-90-1 03, Spent Fuel Pit Radiation Monitor, subsequently detects a high Radiation condition, of Auxiliary Building Isolation equipment will actuate with no operator actions.

LU A allow gjjy one train B. allow both trains C. NOT allow pjjjy one train D. NOT allow both trains Wednesday, June 05, 2013 8:16:20 AM 152

1305 NRC RO Exam Feedback DIS TRACTOR ANAL YSIS:

A. Correct, Tech Specs only require one SFP Radiation monitor and one ABGTS system to be operable for fuel movement in the Spent Fuel Pit The SFP Radiation Monitors are trained in respect to isolating the Auxiliary Building with 102 being Train A and 103 being Train B.

B. incorrect, Tech Specs only require one SFP Radiation monitor and one ABGTS system to be operable for fuel movement in the Spent Fuel Pit Both trains would not be isolated if radiation was detected by the other SFP Radiation Monitor. The monitors are trained in respect to isolating the Auxiliary Building with 102 being Train A and 103 being Train B. Plausible because fuel shuffle continuing is correct and RM 101 (on the Aux. Building Stack) would cause isolation of both trains.

C. lncorrecl Fuel shuffles would not have to be stopped, the movement could continue because only one Rad Mon is required in the Spent fuel Pit. Plausible because other conditions would cause fuel movement to be stopped and only one train of the isolation in the Aux Building is correct.

D. IncorrecI, Fuel shuffles would not have to be stopped, the movement could continue because only one Rad Mon is required in the Spent fuel Pit and both trains of isolation of the Aux Building would not be isolated if radiation was detected by the other SFP Radiation Monitor. The monitors are trained in respect to isolating the Auxiliary Building with 102 being Train A and 103 being Train B. Plausible because other conditions would cause fuel movement to be stopped and RM-09-101(on the Aux. Building Stack) would cause isolation of both trains.

Wednesday, June 05, 2013 8:16:20 AM 153

1305 NRC PC Exam Notes Question Number: 61 Tier: 2 Group 2 K/A: 034 Fuel Handling Equipment System (FHES)

K6.02 Knowledge of the effect of a loss or malfunction on the following will have on the Fuel Handling System:

Radiation monitoring systems Importance Rating: 2.6 / 3.3 1OCFRPart55: 41.7/45.7 1OCFR55.43.b: Not applicable K/A Match: Question requires the knowledge of how a loss of Radiaition monitors affects the use of fuel handling equipment.

Technical

Reference:

1 ,2-47W61 1-30-5 R9 1 ,2-47W61 1-30-6 Ri 1 Tech Spec Table 3.3-6 Tech Spec 3.9.12 Proposed references None to be provided:

Learning Objective: OPT200.ABVENT

4. EXPLAIN the physical connections and/or cause-effect relationships between the ABVENT system and the following systems:
a. Radiation Monitoring 0-RM-90-1 01 B, Aux Bldg Exhaust Stack Monitor 0-RM-90-.102 & 103, Fuel Pool Monitors Question Source:

New Modified Bank Bank X Question History: SQN bank question 034 K6.02 061 used on the SQN 1/2009 retake exam formatted and the correct answer relocated for use on the SQN 05/2013 exam.

Comments:

Wednesday, June 05, 2013 8:16:20 AM 154

1305 NRC RO Exam

62. 071 K3.05 062 Given the following:

- Unit 1 is operating at 100% power.

- 0-HS-77-245 is postioned to align 0-FCV-77-245, waste gas vent flow control valve to Unit 1.

- Waste Gas Decay Tank J relief valve develops a flancie leak and the tank contains high activity gas.

Which ONE of the following identifies how the radiation monitors listed below will respond to the gas release?

Note:

0-RE 118, Waste Gas Rad Monitor 0-RE 101, Auxiliaty Building Ventilation Monitor 1-RE-90-400, Unit 1 Shield Building Vent Monitor A Only 0-RE-90-1 01 will detect the release.

B. Only 1 -RE-90-400 will detect the release.

C. Both 0-RE-90-1 18 and 1 -RE-90-400 will detect the release.

D. Both 0-RE-90-118 and 0-RE-90-101 will detect the release.

Wednesday, June 05, 2013 8:16:20 AM 155

1305 NRC AC Exam Feedback DISTRACTOR ANAL YSIS:

A. Correct, leakage must enter the WGDT release line to pass by waste gas radiation monitor 0-RE-90-1 18. Flange leakage would not enter this line but rather the general Aux Building Spaces where it would eventually pass by Aux building stack radiation monitor 0-RE-90-101. Therefore for flange leakage, 0-RE-90-1 18 nor 1-RE-90-400 (the normal release point) would detect the release but 0-RE-90-101 would detect the release.

B. Incorrect; leakage must enter the WGDT release line to pass by waste gas radiation monitor 0-RE-90-1 18 and 0-.RE-90-400. Flange leakage would not enter this line but rather the general Aux Building Spaces where it would eventually pass by Aux building stack radiation monitor 0-RE 101. If the leakage had been through the seat of the valve and the normal release point was aligned 0-RE-90-400 would have detected the release. Therefore for flange leakage, 0-RE-90-1 18 would not detect the release but 0-RE-90-101 would detect the release. Plausible if applicant does not understand the relationship between the ventilation system and rad monitors and thinks the gas would exit via the Shield Building Stack which is where a normal gas decay tanks release is routed.

C. Incorrect; leakage must enter the WGDT release line to pass by waste gas radiation monitor 0-RE 118 and its normal release point monitored by 1 -RE-90-400. Flange leakage would not enter this line but rather the general Aux Building Spaces where it would eventually pass by Aux building stack radiation monitor 0-RE-90-101. Therefore for flange leakage, 0-RE-90-1 18 nor 1-RE-90-400 would detect the release but 0-RE-90-101 would. Plausible if applicant does not understand the relationship between the ventilation system and rad monitors.

0. Incorrect; leakage must enter the WGDT release line to pass by waste gas radiation monitor 0-RE-90-1 18. Flange leakage would not enter this line but rather the general Aux Building Spaces where it would eventually pass by Aux building stack radiation monitor 0-RE-90-101. Therefore for flange leakage, 0-RE-90-1 18 would not detect the release but 0-RE-90-101 would detect the release. Plausible if applicant does not understand the relationship between the ventilation system and rad monitors.

Wednesday, June 05, 2013 8:16:20 AM 156

1305 NRC RO Exam Notes Question Number: 62 Tier: 2 Group 2 K/A: 071 Waste Gas Disposal System (WGDS)

K3.05 Knowledge of the effect that a loss or malfunction of the Waste Gas Disposal System will have on the following:

ARM and PRM systems Importance Rating: 3.2 / 3.2 1OCFRPart55: 41.7/45.6 1OCFR55.43.b: Not applicable K/A Match: K/A is matched because the question requires knowledge of how a leaking flange on the will affect the radiation monitoring systems at the station.

Technical

Reference:

1 ,2-47W830-4 R47 1 -47W866-i R41 1 ,2-47W866-2 Ri 3 i,2-47W866-10 R19 Proposed references None to be provided:

Learning Objective: OPT200.GRW

4. EXPLAIN (or SKETCH as applicable) the physical connections and/or cause-effect relationships between the Gaseous Radwaste System and the following systems:
c. Radiation Monitors, RM-90-1 18
g. Ventilation Question Source:

New Modified Bank Bank X Question History: Originally modified from SQN question 060 AA1 .02 022 that was used on the SQN January 2008 exam Comments:

Wednesday, June 05, 2013 8:16:20 AM 157

1305 NRC RO Exam

63. 072 A4.01 063 Given the following plant conditions:

- Both units are operating at 100% power.

- 1-RA-90-.1C AUX BLDG AREA RAD MON INSTR MALFUNC (M12A, B-7) is LIT Which ONE of the following completes the statements below?

1-RA-90-1C INSTR MALFUNC alarm _JJL result in an automatic action.

{)_ is a condition that would cause the alarm.

A. (1) will (2) Operate/Calibrate switch set to calibrate B. (1) will (2) Sample flow less than setpoint C (1) will NOT (2) Operate/Calibrate switch set to calibrate D. (1) will NOT (2) Sample flow less than setpoint Feedback DISTRACTOR ANAL YSIS:

A. Incorrect, Plausible because other ARMs have automatic actions (0-RM 102 and -103 Spent Fuel Pit Rad Monitors). The Operate/Calibrate switch set to calibrate position is listed as a cause of the alarm and is correct.

B. Incorrect, Plausible because other ARMs have automatic actions (0-RM-90-102 and -103 Spent Fuel Pit Rad Monitors). The second part is plausible as a sample flow less than setpoint will cause an instrument malfunction alarm on a process radiation monitor.

C. Correct, 1-RA iC AUX BLDG AREA RAD MON does not have any auto actions. The Operate/Calibrate switch set to calibrate position is listed as a cause of the alarm and is correct.

0. Incorrect, the first part is correct. The second part is plausible as a sample flow less than setpoint will cause an instrument malfunction alarm on a process radiation monitor.

Notes Question Number: 63 Wednesday, June 05, 2013 8:16:20 AM 158

1305 NRC RO Exam Tier: 2 Group 2 K/A: 072 Area Radiation Monitoring System A4.01 Ability to manually operate and/or monitor in the control room:

Alarm and interlock setpoint checks and adjustments Importance Rating: 3.0* /33 10 CFR Part 55: 41.7 / 45.5 to 45.8 1OCFR55.43.b: Not applicable K/A Match: K/A is matched because the question requires the ability to monitor the conditions of the area radiation monitor given a MCR alarm and the knowledge of the effect of setting the Operate/Calibrate switch to calibrate.

Technical

Reference:

0-AR-M12-A rev 52, window B-7 Proposed references None to be provided:

Learning Objective: OPT200.RM

2. STATE the location of the following listed Radiation Monitoring System components, and using in plant locations, LOCATE or IDENTIFY the associated indications and controls:
c. Area RM5 (ARM)
3. Given plant conditions, DETERMINE if any of the following Radiation Monitoring System alarms would be present and DESCRIBE actions required by the ARP:
f. [0-XA-55-12AJD, B/A7j AUX BLDG AREA RAD MON INSTR MALFUNC
6. EXPLAIN the Radiation Monitoring System design features and/or operational interlocks that provide the following:
c. System failure/malfunction indication Question Source:

New X Modified Bank Bank Question History: New for ILT 1305 exam Comments:

Wednesday, June 05, 2013 8:16:20 AM 159

1305 NRC RO Exam

64. 079 K4.0I 064 Given the following plant conditions:

- Unit 2 is in Mode 6 with fuel movement in progress.

- Maintenance personnel report that a Service Air connection providing air to a pneumatic grinder has broken and cannot be isolated.

- Air Header Pressure indications are as follows:

PI-32-104A, AUX CONT AIR HDR A PRESS 92 psig Pl-32-105A, AUX CONT AIR HDR B PRESS 92 psig Pl-32-200, CONT AIR HDR PRESS 94 psig and slowly rising PI-33-199, SERV AIR HDR PRESS 90 psig and slowly lowering Which ONE of the following completes the statements below?

(1) At the current pressure, PCV-33-4, SERVICE AIR ISOL (1)

(2) PCV-33-4 valve L after the Service Header is repressurized.

A (1) remains open but will be closed if pressure continues to drop (2) must be reset B. (1) is closed (2) must be reset C. (1) remains open but will be closed if pressure continues to drop (2) will reopen automatically D. (1) is closed (2) will reopen automatically Wednesday, June 05, 2013 8:16:20 AM 160

1305 NRC RO Exam Feedback DISTRACTOR ANAL YSIS:

A. Correct. PCV-33-4 closes at 88 psig and lowering. AR-M15B window D-7, ZS-33-4 SERVICE AIR ISOL CLOSED provides indication of PCV-33-4 and is not in alarm. The second part is correct. AOP-M.02 and O-SO 1 provide instructions for depressing the REST pushbutton on P5-33-4 after Service Header is pressurized.

B. Incorrect. The valve is open as pressure in the service air header is> 88 psig.

AR-M15B window D-7, ZS-33-4 SERVICE AIR ISOL CLOSED provides indication of PCV-33-4 and is not in alarm. The second part is incorrect. AOP-M.02 and O-S0-33-1 provide instructions for depressing the RESET pushbutton on PS-33-4 after Service Header is pressurized. This is plausible as most pressure regulating valves do not require a reset.

C. Incorrect. The first part is correct. The second part is incorrect. AOP-M.02 and O-S0-33-1 provide instructions for depressing the RESET pushbutton on PS-33-4 after Service Header is pressurized. This is plausible as most pressure regulating valves do not require a reset.

D. Incorrect. The valve is open as pressure in the service air header is> 88 psig.

AR-M15B window D-7, ZS-33-4 SERVICE AIR ISOL CLOSED provides indication of PCV-33-4 and is not in alarm. The second part is correct.

Wednesday, June 05, 2013 8:16:20 AM 161

1305 NRC RO Exam Notes Question Number: 64 Tier: 2 Group 2 K/A: 079 Station Air System K4.0i Knowledge of SAS design feature(s) and/or interlock(s) which provide for the following: Cross-connect with lAS.

Importance Rating: 2.9 / 3.2 10 CFR Part 55: 41.7 1OCFR55.43.b: Not applicable K/A Match: Question matches the K/A by since the examinee is required to recall the actions necessary to re-open the valve and remember at what pressure the valve wil automtically isolate.

Technical

Reference:

AR-Mi 5-B rev 35, window D-7 AOP-M.02 rev 21 Proposed references None to be provided:

Learning Objective: OPT200.CSA Obj 7.d Explain the Control and Service Air System design features and/or interlocks that provide for the following:

Automatic isolation of sections of the air system.

Question Source:

New X Modified Bank Bank Question History: New for SQN ILT 1305 Comments:

Wednesday, June 05, 2013 8:16:20 AM 162

1305 NRC RO Exam

65. 086 K1.03 065 Given the following;

- Unit 1 is in Mode 4.

- AOP-N.03, External Flooding, has been implemented.

Which ONE of the following completes the statement below relative to the Flood Mode Spool pieces that connect the HPFP system to the Unit 1 Aux Feedwater System?

The spool pieces are installed during JJI_ actions and they connect to the AFW piping on the discharge of the fL AFW pump(s).

L1 A. Stage I Motor Driven B Stage II Motor Driven C. Stage I Turbine Driven D. Stage II Turbine Driven Feedback DIS TRACTOR ANAL YSIS:

A. Incorrect, Plausible because there are actions that occur during Stage I of the AOP and the second part is correct.

B. Correct, The spool pieces are installed during Stage 2 activities (See note for AOP-N.03 below) and they connect to the Motor Driven AFW piping on the discharge side of the AFW pumps.

C. Incorrect, Plausible because there are actions that occur during Stage I of the AOP and the connection to the TDAFW pump could achieve the same result.

D. Incorrect, The first part is correct. The second part is correct as the connection to the TDAFW pump could achieve the same result.

Wednesday, June 05, 2013 8:16:20 AM 163

1305 NRC RO Exam Notes Question Number: 65 Tier: 2 Group 2 K/A: 086 Fire Protection System (FPS)

Ki .03 Knowledge of the physical connections and/or cause effect relationships between the Fire Protection System and the following systems:

AFW system Importance Rating: 3*4* /35*

10 CFR Part 55: 41.2 to 41.9 / 45.7 to 45.8 1OCFR55.43.b: Not applicable K/A Match: Question requires knowledge of the physical condition between the Fire Protection System and the Auxiliary Feedwater system and the condition that would result in the staged spool pieces being installed to connect the system together.

Technical

Reference:

AOP-N.03, External Flooding, Revision 43 Proposed references None to be provided:

Learning Objective: OPL271AOP-N.03 Ob] 5 Summarize AOP-N.03s mitigating strategies for Flooding.

Question Source:

New X Modified Bank Bank Question History: New question for the SQN 05/2013 exam Comments:

Wednesday, June 05, 2013 8:16:20 AM 164

1305 NRC RO Exam

66. G2.1.15 066 Which ONE of the following identifies the maximum time Standing Orders for administrative issues and whos approval is required in accordance with ODM-Y, Standing Orders and Shift Orders?

Standinci Orders Approval A. 30 days Operations Manager B. 30 days Operations Superintendent C. 1 year Operations Manager Dv 1 year Operations Superintendent Feedback DIS TRACTOR ANAL YSIS:

A. Incorrect, Plausible because 30 days is a time period for shift orders and the Operations Manager is logical choice for all administrative/policy procedures.

B. Incorrect, Plausible because 30 days is a time period for shift orders and the Operations Superintendent is the approver.

C. Incorrect, Plausible because 1 year being the maximum normal time for a Standing Orders for administrative issues is correct and the Operations Manager is logical choice for all administrative/policy procedures.

D. Correct, Operations Directive Manual (0DM) -Y states that a Standing Order for administrative issues should not remain in effect beyond 1 year and the Operations Superintendent is the approver.

Wednesday, June 05, 2013 8:16:20 AM 165

1305 NRC RO Exam Notes Question Number: 66 Tier: 3 Group n/a K/A: G 2.1.15 Knowledge of administrative requirements for temporary management directives, such as standing orders, night orders, Operations memos, etc.

Importance Rating: 2.7 / 3.4 10 CFR Part 55: 41.10 / 45.12 1OCFR55.43.b: Not applicable K/A Match: KA is matched because the question requires knowledge of administrative requirements (maximum time restrictions) for standing orders and night orders, Technical

Reference:

ODM-Y, Standing Orders and Shift Orders, Revision 0 OPDP-1, Conduct of Operations, R24 Proposed references None to be provided:

Learning Objective: OPL271OPSMGMTL Obj 10 Describe the difference between Shift Orders and Standing Orders. Identify the requirements for both relative to assuming a shift operating position.

Cognitive Level:

Higher Lower X Question Source:

New Modified Bank Bank X Question History:

Comments:

Wednesday, June 05, 20138:16:20 AM 166

1305 NRC RO Exam

67. G2.1.43 067 Given the following plant condition:

- Unit 1 has entered Section 5.4, Power Coastdown at End of Life, of GO-5, Normal Power Operation.

Which ONE of the following identifies...

(1) how the unit will be operated during performance of GO-5, Section 5.4 and (2) why reactor power changes should be limited to 1% per hour?

A. (1) The crew will reduce turbine load as needed to maintain Tavg on program.

(2) To prevent MTC from exceeding Tech Spec limits.

B (1) The crew will reduce turbine load as needed to maintain Tavg on program.

(2) To avoid xenon peaking which could force a plant shutdown.

C. (1) The crew allow Tavg to drop while maintaining reactor power as stable as possible with AFD being maintained less than the positive limit.

(2) To prevent MTC from exceeding Tech Spec limits.

D. (1) The crew allow Tavg to drop while maintaining reactor power as stable as possible with AFD being maintained less than the positive limit.

(2) To avoid xenon peaking which could force a plant shutdown.

Wednesday, June 05, 2013 8:16:20 AM 167

1305 NRC RO Exam Feedback DISTRACTOR ANAL YSIS:

A. Incorrect, Plausible because turbine load being reduced daily to maintain Tavg on program is correct and because the MTC is at its most negative value at the end of core life do to very low boron concentration and has magnitude limits established in the TSandtheNDR.

B. Correct; 00-5, Section 5.4 identifies that turbine load will be reduced daily to maintain Ta vg on program and reactor power changes should be limited to

<1 %/min to avoid xenon peaking which could force a plant shutdown (implied in section 5.4.8 and verified with RE as the basis for the slow power reduction with rods out.

C. Incorrect, Plausible because the direction to maintain AFD less than the positive limit is included in the Notes in Section 5.4. As power is reduced it shifts up in the core and is controlled by control rods not power reduction limits. Some Westinghouse will use this method of maintaining power constant and letting Tavg drift down. The second part is plausible because the MTC is at its most negative value at the end of core life do to very low boron concentration and has magnitude limits established in the TS and the NDR.

D. Incorrect, Plausible because the direction to maintain AFD less than the positive limit is included in the Notes in Section 5.4. As power is reduced it shifts up in the core and is controlled by control rods not power reduction limits. Some Westinghouse will use this method of maintaining power constant and letting Tavg drift down. The second part is correct.

Wednesday, June 05, 2013 8:16:20 AM 168

1305 NRC RO Exam Notes Question Number: 67 Tier: 3 Group K/A: G 2.1.43 Ability to use procedures to determine the effects on reactivity of plant changes, such as reactor coolant system temperature, secondary plant, fuel depletion, etc.

Importance Rating: 4.1/4.3 10 CFR Part 55: 41.10/43.6 / 45.6 1OCFR55.43.b: Not applicable K/A Match: K/A is matched because the question requires the abiity to recall the directions and constraints of GO-4 (as related to RCS temperature and secondary plant changes) when the fuel is depleted and the plant is in a coastdown condition.

Technical

Reference:

GO-5, Noraml Power Operations, R82 TS LCO 3.1.1.3 Unit 1 NDRCycIe19 Unit 1 COLR Cycle 19 Proposed references None to be provided:

Learning Objective: OPL271GO-5 Obj 1 State the reason for each precaution and limitation provided in 0-GO-5.

Question Source:

New Modified Bank Bank X Question History: WBN 10-2011 Audit Exam Comments:

Wednesday, June 05, 2013 8:16:20 AM 169

1305 NRC RO Exam

68. G 2.2.22 068 Given the following plant conditions:

- Unit 1 is in MODE 1 with RCS pressure at 2235 psia.

- The following RCS leakages were determined per 1-Sl-OPS-068-137.0, Reactor Coolant System Water Inventory.

Total RCS leakage 12.41 gpm PRT leakage 5.32 gpm CLA #1 leakage 0.88 gpm RCDT leakage 3.18 gpm S/G #2 leakage 0.07 gpm SIG #1 ,#3,#4 leakage 0.00 gpm

[Note: Assume leakages other than those given as zero (0) gpm].

Which ONE of the following completes the statement below?

LCO 3.4.6.2, Reactor Coolant System Operational Leakage, action is required to be entered because the limit has been exceeded?

1. IDENTIFIED LEAKAGE
2. UNIDENTIFIED LEAKAGE
3. PRIMARY-TO-SECONDARY LEAKAGE A. 1 only By 2 only C. 2and3only D. 1 and 3 only Wednesday, June 05, 2013 8:16:20 AM 170

1305 NRC RO Exam Feedback DISTRACTOR ANAL YSIS:

Identified Leakage = PAT + CLA + RCDT + S/G Leakages

=5.32+ 0.88+3.18÷0.07 =9.45gpm Unidentified leakage = RCS leakage Identified leakage = 12.41- 9.45

- = 2.96 gpm Primary-to Secondary Leakage = 0.07 gpm = 100.8 gpd A. Incorrect, The IDENTIFIED Leakage at 9.45 gpm is less than the LCO limit of 10 gpm. Plausible because the Identified Leakage is elevated to be close to the maximum allowed leakage.

B. Correct, The UNIDENTIFIED Leakage at 2.96 gpm is greater than the LCO limit of 1 gpm.

C. Incorrect, The first part is correct and the PRIMARY-TO-SECONDARY Leakage at 100.8 gpm is less than the LCO limit of 150 gpd. Plausible because the PRIMARY-TO-SECONDARY Leakage is greater than the leakage which requires a plant shutdown in accordance with the AOP for steam generator tube leakage.

D. Incorrect, The IDENTIFIED Leakage at 9.45 gpm is less than the LCO limit of 10 gpm. Plausible because the Identified Leakage is elevated to be close to the maximum allowed leakage. The second part is plausible because the PRIMARY-TO-SECONDARY Leakage is greater than the leakage which requires a plant shutdown in accordance with the AOP for steam generator tube leakage.

Wednesday, June 05, 20138:16:20 AM 171

____

1305 NRC RO Exam Notes Question Number: 68 Tier: 3 Group n/a K/A: G 2.2.22 Knowledge of limiting conditions for operations and safety limits.

Importance Rating: 4.0 / 4.7 10 CFR Part 55: 41.5/43.2 / 45.2 1OCFR55.43.b: Not applicable K/A Match: Question requires knowledge of the LCO for Reactor Coolant System leakage.

Technical

Reference:

Tech Spec 3.4.6.2 Proposed references None to be provided:

Learning Objective: OPT200.RCS

11. Using the Technical Specifications, Technical Requirements Manual, and the ODCM,
c. Given a set of plant conditions/parameters, DETERMINE entry level conditions for Reactor Coolant System Tech Spec LCO actions, Technical Requirements and/or ODCM Controls.

Question Source:

New Modified Bank Bank X Question History: SQN bank question G 2.2.22 069 used on an Audit exam in 1/2008.

Comments: High Cognitive Wednesday, June 05, 2013 8:16:21 AM 172

1305 NRC RO Exam

69. G2.2.42 069 Given the following plant conditions:

While performing a cooldown on Unit 1 from Mode 3 to Mode 4 the following parameters were logged:

Time RCS Press RCS Temp PZR LIQ Space Temp 0200 2200 psig 553°F 650°F 0230 1550 psig 527°F 606°F 0300 1135 psig 505°F 560°F 0330 765 psig 447°F 494°F 0400 400 psig 402°F 440°F Which ONE of the following describes the Tech Spec I TRM implications of these conditions?

A RCS cooldown rate limits were exceeded; Tech Spec action is required within a maximum of 30 minutes.

B. RCS cooldown rate limits were exceeded; Tech Spec action is required within a maximum of 60 minutes.

C. PZR cooldown rate limits were exceeded; TRM action is required within a maximum of 30 minutes.

D. PZR cooldown rate limits were exceeded; TRM action is required within a maximum of 60 minutes.

Wednesday, June 05, 20138:16:21 AM 173

1305 NRC RO Exam Feedback DISTRACTOR ANAL YSIS:

A. Correct, between 0300 and 0400 RCS coo/down rate was greater than 100°F in one hour exceeds the limit, action is required within 30 minutes.

B. Incorrect, the RCS cooldown rate was exceeded however the Tech Spec action time is incorrect. This is plausible since 60 minutes is a common action time limit.

C. Incorrect, between 0300 and 0400, PZR cooldown rate limit was greater than 1 00°F, however the TRM limit is 200°F in any one hour. This is plausible if candidate uses the PZR heatup rate limit of 100°F per hour rate. The required action time limit is correct.

0. Incorrect, the PZR coo/down rate limit was not exceeded (see item C). The required action time limit is also incorrect. This is plausible since 60 minutes is a common action time limit.

Wednesday, June 05, 2013 8:16:21 AM 174

1305 NRC RD Exam Notes Question Number: 69 Tier: 3 Group n/a K/A: G 2.2.42 Ability to recognize system parameters that are entry-level conditions for Technical Specifications.

Importance Rating: 3.9 / 4.6 10 CFR Part 55: 41.7/41.10 / 43.2 / 43.3 / 45.3 1OCFR55.43.b: Not applicable K/A Match: Applicant must know entry level condition in order to know if the RCS or PZR cooldown limits have been exceeded.

Technical

Reference:

Tech Spec 3.4.9.1 PTLR rev 4, July 2003 TRM 3.4.9.2 Proposed references None to be provided:

Learning Objective: OPT200.RCS Obj 11 .a and c Using the Technical Specification, Technical Requirements Manual, and the ODCM:

a. List from memory, Reactor Coolant System Tech Spec LCOs and/or Technical Requirement having action times <one hour.
c. Given a set of plant conditions/parameters, determine entry level conditions for RCS Tech Spec LCD actions, TRM and or ODCM controls.

Question Source:

New Modified Bank Bank X Question History: SON bank question Comments: High Cog Wednesday, June 05, 20138:16:21 AM 175

1305 NRC P0 Exam

70. G 2.2.43 070 A main control room annunciator, 1-M5-B, A-7, HIGH PRESS IN AUX BLDG, has alarmed repeatedly over the past several hours. It is determined to be a nuisance alarm. All compensatory measures have been taken.

Aroval for disablement shall be obtained from the (1) and the disablement shall be logged in the (2)

A. (1) Shift Manager ONLY (2) Disabled Annunciator Book ONLY B. (1) Shift Manager ONLY (2) Disabled Annunciator Book and the narrative log C. (1) Unit Supervisor or the Shift Manager (2) Disabled Annunciator Book ONLY D (1) Unit Supervisor or the Shift Manager (2) Disabled Annunciator Book and the narrative log Feedback DIS TRACTOR ANAL YSIS:

A. Incorrect, Plausible as the SM is one of the two positions that can approve the disablement. The second part is plausible as the Disabled Annunciator Book is one of the two places were the disablement is logged.

B. lncorrecI, Plausible as the SM is one of the two positions that can approve the disablement. The second part is correct.

C. lncorrect, The first part is correct. The second part is plausible as the Disabled Annunciator Book is one of the two places were the disablement is logged.

D. Correct, Either the SM or the US can sign the disabled alarm form OPDP 1 approving the alarm disablement. OPDP-4 requires that the disablement be logged in both the Disabled Annunciator Book and the narrative log.

Wednesday, June 05, 2013 8:16:21 AM 176

____

1305 NRC RO Exam Notes Question Number: 70 Tier: 3 Group K/A: 2.2.43 Knowledge of the process used to track inoperable alarms.

Importance Rating: 3.0 / 3.3 10 CFR Part 55: (CFR: 41.10/43.5 /45.13) 1OCFR55.43.b: Not applicable K/A Match: Question matches the KA by testing the examinees knowledge of the processs used to track inoperable alarms.

Technical

Reference:

OPDP-4, Annuciator Disablement, rev 5 Proposed references None to be provided:

Learning Objective: 0PL271 OPDP-4 Obj 2 and 3

2. Describe the General Requirements for Annunciator Disablement
3. Describe the procedure to disable and alarm.

Question Source:

New X Modified Bank Bank Question History: New for SQN ILT NRC 1305 Exam Comments: Low Cognitive Wednesday, June 05, 2013 8:16:21 AM 177

1305 NRC RO Exam

71. G2.3.11 071 Given the folowing plant conditions:

- A Reactor Trip and safety injection have occurred on Unit 1 due to a SGTR.

- 1-RA-90-1 19A CNDS VAC PMP LO RNG AIR EXH MON HIGH PAD alarm is LIT.

- 1-PA-90-120A1121A STM GEN BLDN LIQ SAMP MON HI RAD alarm is LIT.

- Steam Generator parameters are as follows:

SG1 SG2 SG3 SG4 NP Level 27% 28% 32% 21%

(stable) (lowering) (rising) (rising)

AFW Flow 70 gpm 0 gpm 0 gpm 200 gpm Which ONE of the following is an action required to be taken to minimize the radiation release?

A. Raise #2 SG Atmospheric relief valve setpoint.

B Raise #3 SG Atmospheric relief valve setpoint.

C. Isolate the Steam Supply from the #1 SG to the TD AFW Pump turbine.

D. Isolate the Steam Supply from the #4 SG to the TD AFW Pump turbine.

Wednesday, June 05, 20138:16:21 AM 178

1305 NRC RO Exam Feedback DISTRACTOR ANAL YSIS:

A. lncorrect, Steam Generator #3 is the ruptured steam generator (not steam generator #2) but if the #2 SQ had been ruptured raising the setpoint of the #2 SQ Atmospheric relief valve would be a correct action. Plausible if the wrong steam generator is diagnosed as ruptured because the action would be a correct action.

B. Correct, Steam Generator #3 is the ruptured steam generator as identified by the level rising with no AFW flow and an action to control the release of radiation is to raise the setpoint of the #3 SQ Atmospheric relief valve.

C. Incorrect, Steam Generator #3 is the ruptured steam generator (not steam generator #1) but if the #1 SQ had been ruptured isolating the steam supply from the SG#1 to the TD AFW pump turbine would be a correct action. Plausible if the wrong steam generator is diagnosed as ruptured because the action would be a correct action.

D. Incorrect, Steam Generator #3 is the ruptured steam generator (not steam generator #4) but if the #4 SQ had been ruptured ensuring the TD AFW pump turbine was not being supplied from the #4 SQ would be a correct action. Plausible if the wrong steam generator is diagnosed as ruptured because the action would be a correct action.

Wednesday, June 05, 2013 8:16:21 AM 179

1305 NRC RO Exam Notes Question Number: 71 Tier: 3 Group n/a K/A: G 2.3.11 Ability to control radiation releases.

Importance Rating: 3.8 / 4.3 10 CFR Part 55: 41.11 /43.4/45.10 1OCFR55.43.b: Not applicable K/A Match: Question requires the ability to determine action needed usisng conditions provided during a SGTR to control a radaition release Technical

Reference:

E-3, Steam Generator Tube Rupture, Revision 20 Proposed references None to be provided:

Learning Objective: 0PL271.E-3

6. Given the procedure and a set of initial plant conditions, determine actions required to mitigate the event in progress.

Question Source:

New Modified Bank Bank X Question History: SON bank question G 2.3.11 070 used during the 1/2009 Retake exam with minor wording changes in the stem.

Comments:

Wednesday, June 05, 2013 8:16:21 AM 180

1305 NRC RO Exam

72. G 2.3.5 072 Given the following plant conditions:

- A source check is to be performed on CCS radiation monitor 1-RE-90-123.

Which ONE of the following completes the statement below?

The source check is verified by observing the jJ and the isolation function of the Surge Tanks vent be manually blocked during the source check in accordance with 1 -SO-90-1, Liquid Process Radiation Monitors.

A. (1) analog rate meter trending upscale (2) can B. (1) analog rate meter trending upscale (2) can NOT C. (1) bargraph display responds upscale (2) can D (1) bargraph display responds upscale (2) can NOT Feedback DIS TRACTOR ANAL YSIS:

A. Incorrect, Plausible since observing the rate meter trending upscale is correct for different radiation monitors and because some radiation monitrors outputs can be blocked during the performance of a source check but not this one.

B. IncorrecI, Plausible since observing the rate meter trending upscale is correct for different radiation monitors and because the isolation function can not be blocked on this meter during the source check is correct.

C. Incorrect, Plausible since a succeesful source check being determined by observing the bargraph trending upscale is correct and some radiation monitors outputs can be blocked during the performance of a source check but not this one.

D. Correct, a successful source check is determined by observing the bargraph trending upscale. Also the isolation function is not available on this type monitor thus the valves could automatically isolate during the source check requiring them to be checked open after the source check is performed.

Wednesday, June 05, 2013 8:16:21 AM 181

1305 NRC RO Exam Notes Question Number: 72 Tier: 3 Group n/a K/A: G 2.3.5 Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.

Importance Rating: 2.9/2.9 1OCFRPart55: 41.11 1OCFR55.43.b: Not applicable K/A Match: This question matches the K/A by requiring the applicant to identify how to perform a source check on fixed radiation monitoring equipment and to determine how functions performed by the monitor are affected by the source check.

Technical

Reference:

1-SO-90-1, Liquid Process Radiation Monitors, Rev 13 Proposed references None to be provided:

Learning Objective: OPT200RM Obj 7.1 Explain the operational implication of the following concept as it applies to the Radiation Monitoring System: Check source operation without blocking process function.

Question Source:

New Modified Bank Bank X Question History: Original question used on Feb 2010 NRC exam Comments:

Wednesday, June 05, 2013 8:16:21 AM 182

1305 NRC RO Exam

73. G 2.4.8 073 Given the following plant conditions:

- Unit 2 at 100% power.

- Panel 2-M-1 Annunciator Window 125V DC VITAL CHGR Ill FAILJVITAL BAT Ill DISCHARGE alarms.

- The crew enters AOP-P.02, Loss Of 125V DC Vital Battery Board.

- A reactor trip occurs on high PZR pressure.

Which ONE of the following identifies the allowed usage of AOP-P.02 after the Emergency Operating Procedure network is entered following the reactor trip?

Continued performance of AOP-P.02 is (1) because (2)

A. (1) allowed after the crew enters ES-0.1, Reactor Trip Response (2) ES-0.1 is NOT an accident mitigation EOP.

B (1) allowed after the crew enters ES-0.1, Reactor Trip Response (2) restoring power could have an impact on meeting the goals of the EOP.

C. (1) NOT allowed until the crew exits ES-0.1, Reactor Trip Response (2) the procedure reader must remain dedicated to the EOP in effect until the EOPs are exited.

D. (1) NOT allowed until the crew exits ES-0.1, Reactor Trip Response (2) actions taken in AOP-P.02 could degrade the performance of the EOP.

Wednesday, June 05, 2013 8:16:21 AM 183

1305 NRC RO Exam Feedback DIS TRACTOR ANAL YSIS:

A. Incorrect, As identified in EPM-4, selected AOPs, such as loss of vital power can be implemented concurrently with the EOPs. While ES-O. 1 is not an accident mitigation procedure, that is not reason for parallel implementation, it is because the loss of power AOPs can have a significant impact on the ability of the EOP to achieve it goals. Plausible because the parallel implementation is correct but the reason for parallel implementation is not correct.

B. Correct, As identified in EPM-4, selected AOPs, such as loss of vital power can be implemented concurrently with the EOPs because the loss of power AOPs can have a significant impact on the ability of the EOP to achieve it goals.

C. Incorrect, EPM-4 provides that EOPs have priority over AOPs, and normally a dedicated procedure reader is utilized in the EOP network however, selected AOPs are allowed to be performed concurrently with EOPs. Plausible if candidate fails to recognize that AOP-P.02 is allowed to be used while performing ES-O. 1.

0. Incorrect, EPM-4 provides that EOPs have priority over AOPs, however, selected AOPs are allowed to be performed concurrently with EOPs. Plausible if candidate fails to recognize that AOP-P.02 is allowed to be used while performing ES-O. 1 and knows that the AOP actions can not be taken if the action would degrade the EOP performance.

Wednesday, June 05, 2013 8:16:21 AM 184

_____

1305 NRC RO Exam Notes Question Number: 73 Tier: 3 Group na K/A: G 2.4.8 Knowledge of how abnormal operating procedures are used in conjunction with EOP5.

Importance Rating: 3.8 / 4.5 10 CFR Part 55: 41.10 1OCFR55.43.b: Not applicable K/A Match: This question matches the K/A by testing the candidates knowledge of how AOP-P.02 is to be used in conjuction with the EOPs during a loss of vitla DC bus.

Technical

Reference:

EPM-4 rev 22 Proposed references None to be provided:

Learning Objective: 0PL271 EPM-4 Obj 1 Determine/identify the correct procedural application(s) based on the operating procedures network for abnormal and emergency evolutions.

Question Source:

New Modified Bank Bank X Question History: Bank question used on 1/2009 NRC exam Comments:

Wednesday, June 05, 2013 8:16:21 AM 185

1305 NRC RO Exam

74. G 2.4.3 074 Given the following picture of various SG#1 pressure indications:

4 H Which ONE of the following identifies only Post Accident Monitoring (PAM) indications?

A. 1-Pl-1-2A and 1-Pl-1-5 B 1-Pl-1-2A and 1-PR-1-2 C. 1-PI-1-1C and 1-PR-1-2 D. 1-Pl-1-5 and 1-PI-1-1C Wednesday, June 05, 2013 8:16:21 AM 186

1305 NRC RO Exam Feedback DISTRACTOR ANAL YSIS:

A. Correct, Inaccordance with the guidance in EPM-4, Users Guide, sect 3.6.2, the Category 1, Post Accident Monitoring (PAM) instrumentation is identified by a black background with white lettering and a Cl/n the corner of the label.

B. Incorrect, Plausible since this is an indicator on the main control board for SG#l pressure, however it is not desgina ted as a PAM instrument.

C. lncorrect, Plausible since this indication is on the remote shutdown paneI however it is not designated as a PAM instrument.

D. Incorrect, Plausible since this is a PAM instrument on the main control board, however it is a catergory 2 instrument which is designated with a white label with black lettering.

Wednesday, June 05, 2013 8:16:22 AM 187

1305 NRC RO Exam Notes Question Number: 74 Tier: 3 Group n/a K/A: G 2.4.3 Ability to identify post-accident instrumentation Importance Rating: 3.7 / 3.9 10 CFR Part 55: 41.6 1OCFR55.43.b: n/a K/A Match: This question matches the K/A by having the candidates identify which of the pictured indications of SG#1 pressure is the one that is Post Accident Monitor instrumentation used for Steam Line Break indications.

Technical

Reference:

EPM-4, rev 22 sect 3.6.2 Post accident monitoring instrumentation Proposed references None to be provided:

Learning Objective: 0PL271 EPM-4 obj 6 Identify Post-accident instrumentation and determin if its use is required.

Cognitive Level:

Higher X Lower Question Source:

New Modified Bank Bank X Question History: New question developed for 1009 NRC exam Comments: Print on color copier and supply to exam Wednesday, June 05, 2013 8:16:22 AM 188

1305 NRC RO Exam

75. G 2.4.9 075 Given the following plant conditions:

- Unit 1 reactor is shutdown.

- RCS in solid water operation.

- All RCS temperatures are approximately 125°F.

- RCS pressure is 330 psig.

- RHR Train A is in service.

Subsequently:

- RCS pressure increases to 600 psig.

Which ONE of the following describes:

1) the effect the pressure increase would have on the PORVs, AND
2) the action(s) which is/are directed to be taken in accordance with AOP-R.03, RHR System Malfunction if RCS pressure is NOT promptly reduced below 600 psig?

REFERENCE PROVIDED A. Only ONE PORV OPENS Stop the Charging pump.

B. Only ONE PORV OPENS Stop A Train RHR Pump and ensure RHR suction valves are closed.

C. BOTH PORVS OPEN Stop the Charging pump.

D. BOTH PORVS OPEN Stop A Train RHR Pump and ensure RHR suction valves are closed.

Wednesday, June 05, 2013 8:16:22 AM 189

1305 NRC RO Exam Feedback DISTRACTOR ANAL YSIS:

A. Incorrect, The first part is incorrect. Both PORVs would lift at this pressure. This is plausible because LTOPS does have a staggered lift setpoint design feature (approximately 35 psi setpoint difference between the PORVs as shown in PTLR graph associated with LTOPS setpoints), thus the candidates could read the graph and determine that only one PORV would acuate at the given pressure. The second part is incorrect. AOP-R.03 ensures one charging pump is left running.

This is plausible since AOP-R.03 has the operator reduce charging flow which would be met by stopping the charging pump.

B. Incorrect, The first part is incorrect. Both PORVs would lift at this pressure. This is plausible because LTOPS does have a staggered lift setpoint design feature (approximately 35 psi setpoint difference between the PORVs as shown in PTLR graph associated with LTOPS setpoints), thus the candidates could read the graph and determine that only one PORV would acuate at the given pressure. The second part is correct. AOP-R.03 CAUTION prior to step 1 and step 2 RNO directs that RHR pump be stopped and step 2 RNO directs that RHR isolated from the RCS.

C. IncorrecI, The first part is correct. Both PORVs would lift at this pressure as noted on PTLR graph associated wtth LTOPS setpoint). The second part is incorrect.

AOP-R.03 ensures one charging pump is left running. This is plausible since AOP-R.03 has the operator reduce charging flow which would be met by stopping the charging pump.

D. Correct, The first part is correct. Both PORVs would lift at this pressure as noted on PTLR graph associated wtih LTOPS setpoint). The second part is correct.

AOP-R.03 CAUTION prior to step 1 and step 2 RNO directs that RHR pump be stopped and step 2 RNO directs that RHR isolated from the RCS.

Wednesday, June 05, 2013 8:16:22 AM 190

1305 NRC RO Exam Notes Question Number: 75 Tier: 3 Group na K/A: G 2.4.9 Knowledge of low power/shutdown implications in accident (e.g.,

loss of coolant accident or loss of residual heat removal) mitigation stategies.

Importance Rating: 3.8 / 4.2 1OCFRPart55: 41.10 1OCFR55.43.b: Not applicable K/A Match: This question matches the K/A by having the candidate determine the mitigation strategy for reducing the RCS pressure and minimizing the potential for a shutdown LOCA by isolating the low pressure (RHR) system from the high pressure system during an overpressure event while in Mode 5 Technical

Reference:

AOP-R.03, RHR System Malfunctions, rev 30 Pressure-Temperature Limits Report, rev 4 pg 10 & 11 Proposed references Pressure-Temperature Limits Report, rev 4 pg 11 to be provided:

Learning Objective: OPL271AOP-R.03 Obj 12 List any condition(s) that require a RHR pump trip in AOP-R.03 Question Source:

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

You have completed the test!

Wednesday, June 05, 20138:16:22 AM 191

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