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| number = ML15127A501
| number = ML15127A501
| issue date = 04/21/2015
| issue date = 04/21/2015
| title = Station-2015-04-DRAFT Written Exam
| title = 2015-04-DRAFT Written Exam
| author name = Gaddy V
| author name = Gaddy V
| author affiliation = NRC/RGN-IV/DRS/OB
| author affiliation = NRC/RGN-IV/DRS/OB
Line 15: Line 15:


=Text=
=Text=
{{#Wiki_filter:U.S. Nuclear Regulatory Commission Site-Specific RO Written Examination Applicant Information Name: Date: Facility/Unit: Cooper Nuclear Station Region: I   II    III    IV     Reactor Type: W CE BW GE   Start Time:
{{#Wiki_filter:U.S. Nuclear Regulatory Commission Site-Specific RO Written Examination Applicant Information Name:
Finish Time:
Date:                                                 Facility/Unit: Cooper Nuclear Station Region: I       II    III    IV                       Reactor Type: W       CE     BW     GE Start Time:                                           Finish Time:
Instructions Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. To pass the examination, you must achieve a final grade of at least 80.00 percent. Examination papers will be collected 6 hours after the examination begins.
Instructions Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. To pass the examination, you must achieve a final grade of at least 80.00 percent. Examination papers will be collected 6 hours after the examination begins.
Applicant Certification All work done on this examination is my own. I have neither given nor received aid.
Applicant Certification All work done on this examination is my own. I have neither given nor received aid.
Applicants Signature Results Examination Value                                                                __________ Points Applicants Score                                                                __________ Points Applicants Grade                                                                __________ Percent 1


______________________________________
ES-401                             Sample Written Examination                     Form ES-401-5 Question Worksheet Examination Outline Cross-
Applicant's Signature Results Examination Value
__________  Points Applicant's Score
__________  Points Applicant's Grade
__________ Percent 1
ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
-


==Reference:==
==Reference:==
Level                  RO Tier #                __1__
Group #                __1__
K/A #                  _295001.AK3.04___
Importance Rating      _3.4__
295001 Partial or Complete Loss of Forced Core Flow Circulation- Knowledge of the reasons for the following responses as they apply to partial or complete loss of forced core flow circulation:
AK3.04 Reactor SCRAM Question: 1 With the Reactor initially operating at 72% power the following conditions exist:
* Reactor Recirculation MG A trips on motor generator high air temperature.
* PMIS shows the plant is operating in the Stability Exclusion Region (red region) of the Power to Flow map.
* SRM period swings with a fluctuation time of <3 seconds.
Why does Procedure 2.4RR (Reactor Recirculation Abnormal) require a scram?
A. To protect the fuel and ensure fuel clad integrity.
B. RPV water level control is unpredictable due to power swings.
C. It is the fastest way to exit the area of operation while in single loop.
D. To ensure the reactor is shut down before exceeding the RPV pressure limit.
Answer:
A. To protect the fuel and ensure fuel clad integrity.
Explanation:
The reactor is designed such that thermal hydraulic oscillations are prevented or can be readily detected and suppressed without exceeding specified fuel design limits. Specific directions are provided to avoid operation in this region and to immediately exit upon entry. Entries into the Stability Exclusion Region (SER) are not part of normal operation. Although operator action can prevent the occurrence of and protect the reactor from an instability, the APRM Neutron Flux-High (Flow Biased) scram function will suppress oscillations prior to exceeding the Safety Limit MCPR. A manual scram is inserted as an Immediate Operator Action if instability is observed while operating in the SER to ensure MCPR Safety Limit is not challenged, therefore protecting the fuel and ensure fuel clad integrity.
Distracters:
2


Level  RO  Tier #    __1__  Group #  __1__  K/A #  _295001.AK3.04___
B. This option is incorrect because there is no requirement within 2.4RR to scram based upon RPV level swings. The reactor feedwater pumps have anticipatory circuitry which attempts to maintain RPV level in a suitable band so the operators will not think level control is unreliable. This option is plausible because abnormal condition procedure 2.4RXLVL directs scramming the reactor if RPV level cannot be maintained between 12 inches and 50 inches. Due to the very nature of reactor pressure oscillations during neutron flux oscillations, Reactor Feed Pump Turbine speed may be affected due to variations in Reactor Feed Pump output to the reactor. The candidate who believes that power oscillations would cause unreliable or uncontrollable RPV level oscillations per 2.4RR would select this option.
Importance Rating
C. This option is incorrect because the direction to scram is not based on the fastest way of exiting the stability exclusion area. The scram is required because the core is exhibiting thermal hydraulic instabilities. Technical Specifications states to immediately exit the region but does not specifically state to scram the reactor. Procedure 2.4RR requires the operator to insert a manual reactor scram based on observed abnormal neutron flux oscillations.
_3.4__    295001 Partial or Complete Loss of Forced Core Flow Circulation
This option is plausible because scramming is an immediate way of exiting the exclusion region.
- Knowledge of the reasons for the following responses as they apply to partial or complete loss of forced core flow circulation:
D. This option is incorrect because RPV pressure limits are not threatened relatively small power swings. The RPV high pressure scram is designed to shut down the reactor before the thermal power transferred to the reactor coolant increases and challenges the integrity of the fuel cladding and the reactor coolant pressure boundary. Reactor power protection from the APRM flow biased high neutron flux will occur before the pressure spike reaches the RPV pressure limit scram. This option is plausible because power swings will cause pressure swings.
AK3.04 Reactor SCRAM
Technical Reference(s):        2.4RR Reactor Recirculation Abnormal, Rev. 40 Proposed references to be provided to applicants during examination:        ___NONE_______
Learning Objective:    COR002-22-02, OPS Reactor Recirculation System, Rev. 32
: 5.      Briefly describe the following concepts as they apply to the Reactor Recirculation system, or to the Recirculation Flow Control system:
: i.      Power to Flow Map (including normal operation/startup/shutdown and Stability Exclusion Region)
: l.      Thermal limits Question Source:                Bank #              _______
Modified Bank #      _______
New                  __X_____
Question History:               Last NRC Exam        _________
Question Cognitive Level:      Memory or Fundamental Knowledge            X Comprehension or Analysis 10 CFR Part 55 Content:        55.41 (10) 3


Q ues ti on: 1  W it h t h e R eac t o r i niti a ll y ope r a ti n g a t 72% po w e r t h e f o ll o w i n g conditi on s ex i s t:   Reactor Recirculation MG A trips on motor generator high air temperature.
Comments:
P M I S sho w s t h e p l an t i s ope r a ti n g i n t h e S t ab ilit y E xc l us i o n R eg i o n (r e d r eg i on) o f t he P o w e r t o F l o w m ap. SRM period swings with a fluctuation time of <3 seconds. W h y does Procedure 2.4RR (Reactor Recirculation Abnormal) require a scram
LOD 2 4
? A. To protect the fuel and ensure fuel clad integrity.
B. RP V water leve l con tr o l i s unp r ed i c t ab l e du e t o po w e r s w ings.
C. I t i s t h e f as t es t w a y t o ex i t t h e area of operation while i n s i ng l e l oo p. D. To ensure the r eac t o r is shu t do w n before exceeding the RPV pressure limit.
A ns w e r: A. To protect the fuel and ensure fuel clad integrity.
Explanation:
The reactor is designed such that thermal hydraulic oscillations are prevented or can be readily detected and suppressed without exceeding specified fuel design limits. Specific directions are provided to avoid operation in this region and to immediately exit upon entry. Entries into the Stability Exclusion Region (SER) are not part of normal operation. Although operator action can prevent the occurrence of and protect the reactor from an instability, the APRM Neutron Flux
-High (Flow Biased) scram function will suppress oscillations prior to exceeding the Safety Limit MCPR. A manual scram is inserted as an Immediate Operator Action if instability is observed while operating in the SER to ensure MCPR Safety Limit is not challenged, therefore protecting the fuel and ensure fuel clad integrity.
Distracters:
2 B. This option is incorrect because there is no requirement within 2.4RR to scram based upon RPV level swings. The reactor feedwater pumps have anticipatory circuitry which attempts to maintain RPV level in a suitable band so the operators will not think level control is unreliable. This option is plausible because abnormal condition procedure 2.4RXLVL directs scramming the reactor if RPV level cannot be maintained between 12 inches and 50 inches. Due to the very nature of reactor pressure oscillations during neutron flux oscillations, Reactor Feed Pump Turbine speed may be affected due to variations in Reactor Feed Pump output to the reactor. The candidate who believes that power oscillations would cause unreliable or uncontrollable RPV level oscillations per 2.4RR would select this option.
C. This option is incorrect because the direction to scram is not based on the fastest way of exiting the stability exclusion area. The scram is required because the core is exhibiting thermal hydraulic instabilities. Technical Specifications states to immediately exit the region but does not specifically state to scram the reactor. Procedure 2.4RR requires the operator to insert a manual reactor scram based on observed abnormal neutron flux oscillations. This option is plausible because scramming is an immediate way of exiting the exclusion region. D. This option is incorrect because RPV pressure limits are not threatened relatively sma ll power swings. The RPV high pressure scram is designed to shut down the reactor before the thermal power transferred to the reactor coolant increases and challenges the integrity of the fuel cladding and the reactor coolant pressure boundary. Reactor power protection from the APRM flow biased high neutron flux will occur before the pressure spike reaches the RPV pressure limit scram. This option is plausible because power swings will cause pressure swings.
Technical Reference(s):
2.4RR Reactor Recirculation Abnormal, Rev. 40 Proposed references to be provided to applicants during examination:
___NONE_______
Learning Objective:
COR002-22-02, OPS Reactor Recirculation System, Rev. 32
: 5. Briefly describe the following concepts as they apply to the Reactor Recirculation system, or to the Recirculation Flow Control system:
: i. Power to Flow Map (including normal operation/startup/shutdown and Stability Exclusion Region)
: l. Thermal limits Question Source:
Bank #  _______ Modified Bank #
_______  New  __X_____  Question History:
Last NRC Exam
_________  Question Cognitive Level:
Memory or Fundamental Knowledge X Comprehension or Analysis    10 CFR Part 55 Content:
55.41 (10)  3 Comments: LOD 2   4  


5 6
5 6
From 2.4RR procedure change 7
From 2.4RR procedure change 7
ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
 
-
ES-401                             Sample Written Examination                   Form ES-401-5 Question Worksheet Examination Outline Cross-


==Reference:==
==Reference:==
 
Level             RO Tier #             __1__
Level   RO   Tier #     __1__   Group # __1__   K/A #   _295003 AA1.03___
Group #           __1__
Importance Rating
K/A #             _295003 AA1.03___
_4.3__   295003 Partial or Complete Loss of AC:
Importance Rating _4.3__
295003 Partial or Complete Loss of AC:
Ability to operate and/or monitor the following as they apply to partial or complete loss of A.C. power:
Ability to operate and/or monitor the following as they apply to partial or complete loss of A.C. power:
 
AA1.03 Systems necessary to assure safe plant shutdown Question: 2 With the plant operating at 100% power, a loss of offsite power occurs.
AA1.03 Systems necessary to assure safe plant shutdown Q ues ti on: 2 W it h t h e p l an t ope r a ti n g a t 100% po w e r , a l os s o f o ff s it e po w e r occu r s. B o t h d iese l gene r a t o r s f a i l t o s t a r t an d CANNO T b e s t a rt ed. The Supplemental Diesel Generator is NOT available.
* Both diesel generators fail to start and CANNOT be started.
HPC I an d RC I C recove r RP V w a t e r leve l t o +3 5 inche s (N a rr o w Range). W ha t ope r a ti ona l r estri c ti o n app li e s t o t h e con ti nue d us e o f HPC I i n respons e t o t h i s even t , unti l ons it e o r o ff s it e e l ectri ca l po w e r is r es t o r ed? HPC I must be secured...
* The Supplemental Diesel Generator is NOT available.
A. after one cycle of operation and must remain off. B. just prior to the division's battery being exhausted and must remain off.
* HPCI and RCIC recover RPV water level to +35 inches (Narrow Range).
 
What operational restriction applies to the continued use of HPCI in response to this event, until onsite or offsite electrical power is restored?
C. N O LATE R THA N 1 5 m i nu t e s fr o m t h e ti m e t ha t injection fl o w w a s reduced an d m us t r e m a i n off. D. after one cycle of operation and must remain off unless RPV level lowers to +3 inches narrow range.
HPCI must be secured...
A ns w e r: A. aft e r on e cyc l e o f ope r a ti o n an d m us t r e m a i n off. Explanation:
A. after one cycle of operation and must remain off.
B. just prior to the division's battery being exhausted and must remain off.
C. NO LATER THAN 15 minutes from the time that injection flow was reduced and must remain off.
D. after one cycle of operation and must remain off unless RPV level lowers to +3 inches narrow range.
Answer:
A. after one cycle of operation and must remain off.
Explanation:
This procedure assumes that RPV water level and pressure is initially controlled by High Pressure Coolant Injection (HPCI), as directed by the EOPs. CNS has committed to secure HPCI after one cycle of operation, even if EOPs allow HPCI use, in order to extend station battery life during station blackout (SBO). One cycle of HPCI operation is ~ 10 minutes. SBO analysis assumes Reactor Core Isolation Cooling (RCIC) is operable and maintains RPV level 8
This procedure assumes that RPV water level and pressure is initially controlled by High Pressure Coolant Injection (HPCI), as directed by the EOPs. CNS has committed to secure HPCI after one cycle of operation, even if EOPs allow HPCI use, in order to extend station battery life during station blackout (SBO). One cycle of HPCI operation is ~ 10 minutes. SBO analysis assumes Reactor Core Isolation Cooling (RCIC) is operable and maintains RPV level 8
and pressure until on
-site or off
-site electrical power can be restored. If RCIC is unable to perform this function, compensatory actions must be taken to ensure adequate core cooling. This could include starting HPCI. Since RPV level is restored to +35 inches, HPCI is not needed for adequate core cooling and must be secured after one cycle and must remain off. Should RPV level subsequently lower to
-150 inches (meaning RCIC cannot maintain RPV level as assumed in the analysis), then EOPs allow HPCI to be used as operation is outside the SBO analysis. If level lowers to +3 inches, HPCI cannot be restarted as adequate core cooling is not threatened (RPV level is still approximately 168 inches above top of active fuel).


and pressure until on-site or off-site electrical power can be restored. If RCIC is unable to perform this function, compensatory actions must be taken to ensure adequate core cooling.
This could include starting HPCI. Since RPV level is restored to +35 inches, HPCI is not needed for adequate core cooling and must be secured after one cycle and must remain off.
Should RPV level subsequently lower to -150 inches (meaning RCIC cannot maintain RPV level as assumed in the analysis), then EOPs allow HPCI to be used as operation is outside the SBO analysis. If level lowers to +3 inches, HPCI cannot be restarted as adequate core cooling is not threatened (RPV level is still approximately 168 inches above top of active fuel).
Distracters:
Distracters:
B. This option is incorrect because HPCI is manually secured after approximately 10 minutes of operation. Operation until the battery is exhausted would be inconsistent with the commitment to secure HPCI after 1 Cycle of operation. The operator who does not correctly recall the restriction in 5.3SBO to extend station battery life would select this option. This option is plausible because HPCI may be utilized without significantly draining battery power as HPCI turbine speed is generally not changed too much to control injection. Significant draining of battery power would be an issue if cycling HPCI valves or allowing the HPCI Auxiliary Oil Pump to run at low turbine speeds or following shutdown of the turbine.
B.     This option is incorrect because HPCI is manually secured after approximately 10 minutes of operation. Operation until the battery is exhausted would be inconsistent with the commitment to secure HPCI after 1 Cycle of operation. The operator who does not correctly recall the restriction in 5.3SBO to extend station battery life would select this option. This option is plausible because HPCI may be utilized without significantly draining battery power as HPCI turbine speed is generally not changed too much to control injection. Significant draining of battery power would be an issue if cycling HPCI valves or allowing the HPCI Auxiliary Oil Pump to run at low turbine speeds or following shutdown of the turbine.
C. This option is incorrect because CNS has a commitment that HPCI should be secured after approximately 10 minutes of operation vs. 15 minutes from the time HPCI flow is reduced. There is no requirement based on when flow is reduced, the only basis is one cycle. Waiting 15 minutes is inconsistent with the commitment to operate HPCI for no more than one cycle (~10 minutes) during a Station Black Out event. The operator who does not correctly recall the restriction in 5.3SBO to extend station battery life would select this option. This option is plausible because the utilization of HPCI and/or RCIC during this event may be required and 15 minutes is a common number for emergency procedure usage. Examples would be 15 minutes to classify an event, 15 minutes below TAF for significant fuel damage to occur, loss of electrical power to classify an event, etc.
C.     This option is incorrect because CNS has a commitment that HPCI should be secured after approximately 10 minutes of operation vs. 15 minutes from the time HPCI flow is reduced. There is no requirement based on when flow is reduced, the only basis is one cycle. Waiting 15 minutes is inconsistent with the commitment to operate HPCI for no more than one cycle (~10 minutes) during a Station Black Out event. The operator who does not correctly recall the restriction in 5.3SBO to extend station battery life would select this option. This option is plausible because the utilization of HPCI and/or RCIC during this event may be required and 15 minutes is a common number for emergency procedure usage. Examples would be 15 minutes to classify an event, 15 minutes below TAF for significant fuel damage to occur, loss of electrical power to classify an event, etc.
D. This option is incorrect because HPCI is secured after one cycle of operation and no procedural step allows restarting per 5.3SBO. If the operator does not remember the restriction in 5.3SBO they would select this option. This option is plausible because EOPs allow the use of HPCI to maintain adequate core cooling at a much lower RPV water level. There is no reason to be concerned with RPV water level at this point because it is at +35 inches (normal band) and RPV water level would have to drop ~200 inches to reach the point where adequate core cooling is challenged. In the event that adequate core cooling is challenged, the EOPs override the emergency procedure 5.3SBO. If level lowers to +3 inches, HPCI cannot be restarted as adequate core cooling is not threatened (RPV level is still approximately 168 inches above top of active fuel). Technical Reference(s):
D.     This option is incorrect because HPCI is secured after one cycle of operation and no procedural step allows restarting per 5.3SBO. If the operator does not remember the restriction in 5.3SBO they would select this option. This option is plausible because EOPs allow the use of HPCI to maintain adequate core cooling at a much lower RPV water level. There is no reason to be concerned with RPV water level at this point because it is at +35 inches (normal band) and RPV water level would have to drop ~200 inches to reach the point where adequate core cooling is challenged. In the event that adequate core cooling is challenged, the EOPs override the emergency procedure 5.3SBO. If level lowers to +3 inches, HPCI cannot be restarted as adequate core cooling is not threatened (RPV level is still approximately 168 inches above top of active fuel).
5.3SBO Station Blackout, Rev 33.
Technical Reference(s):         5.3SBO Station Blackout, Rev 33.
Proposed references to be provided to applicants during examination:
Proposed references to be provided to applicants during examination:           ___NONE_______
___NONE_______
9
9 Learning Objective:
INT032-01-31 CNS Abnormal Procedures (RO) Electrical


W. Given plant condition(s) and the applicable Abnormal/Emergency Procedure, determine the correct subsequent actions required to mitigate the event(s).
Learning Objective:          INT032-01-31 CNS Abnormal Procedures (RO) Electrical W. Given plant condition(s) and the applicable Abnormal/Emergency Procedure, determine the correct subsequent actions required to mitigate the event(s).
Question Source:            Bank #                _13338_
Modified Bank #        _    _
New                    _______
Question History:            Last NRC Exam          __2002___
Question Cognitive Level:    Memory or Fundamental Knowledge          X Comprehension or Analysis 10 CFR Part 55 Content:      55.41 (10)
Comments:
LOD 4 10


Question Source:
11 ES-401                             Sample Written Examination                   Form ES-401-5 Question Worksheet Examination Outline Cross-
Bank #  _13338_ Modified Bank #
_
_  New  _______  Question History:
Last NRC Exam
__2002___  Question Cognitive Level:
Memory or Fundamental Knowledge X Comprehension or Analysis    10 CFR Part 55 Content:
55.41 (10)  Comments:
LOD 4 10 11 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
-


==Reference:==
==Reference:==
Level                RO Tier #                __1__
Group #              __1__
K/A #                _295004.AA2.01___
Importance Rating    _3.2__
295004 Partial or Total Loss of DC Pwr- Ability to determine and/or interpret the following as they apply to partial or complete loss of D.C. power:
AA2.01 Cause of partial or complete loss of D.C. power Question: 3 The plant is operating at power with the following conditions:
* Breaker 1FE red indicating light is illuminated.
* Breaker 1GE red indicating light is illuminated.
* 4160V buses A, C, and E indicating lights are off.
What is causing the above conditions?
A. Panel BB1 has a blown fuse.
B. Panel BB3 has a blown fuse.
C. Panel AA1 has a blown fuse.
D. Panel AA3 has a blown fuse.
Answer:
C. Panel AA1 has a blown fuse.
Explanation Panel AA1 provides DC power to the 4160V buses A, C, and E indication. Breaker 1FE indication is supplied by Panel AA3 and breaker 1GE indication is supplied by Panel BB3. In order to determine which power supply has been lost the operator must know which power supply is providing the power to the breakers listed. Not all control room breaker indication is powered from the same DC power supply. So the candidate must determine the power supply to the breaker indication. With the breakers listed the candidate determines that Breaker 1FE has indicating lights, so DC Power Panel AA3 has power. Knowing that breaker 1GE has its indicating light illuminated, the candidate knows that DC Power Panel BB3 has power. Knowing that indicating lights for breakers associated with buses A, C, and E NOT being illuminated, the candidate knows that DC Power Panel AA1 has become de-energized.
Distracters:
12


Level  RO  Tier #    __1__  Group #  __1__  K/A #  _295004.AA2.01___
A. This option is incorrect because Panel BB1 is not the power supply to any of the breakers listed in the question. This option is plausible because Panel BB1 does provide power to other 4160V breaker indications in the control room and other non-critical division components. The candidate who does not correctly recall the power supplies listed would select this option.
Importance Rating
_3.2__    295004 Partial or Total Loss of DC Pwr
- Ability to determine and/or interpret the following as they apply to partial or complete loss of D.C. power:
 
AA2.01 Cause of partial or complete loss of D.C. power Q ues ti on: 3  Th e p l an t i s ope r a ti n g a t po w e r w it h t h e f o ll o w i n g conditi ons:  B reake r 1F E red i nd i ca ti n g li gh t i s ill u m i na t ed. B reake r 1 G E red i nd i ca ti n g li gh t i s ill u m i na t ed. 4160 V buse s A , C , an d E i nd i ca ti n g li gh t s a r e o ff. W ha t i s caus i n g t h e abov e conditi ons? A. P ane l BB 1 ha s a b l o w n fuse. B. P ane l BB 3 ha s a b l o w n fuse.
C. P ane l AA 1 ha s a b l o w n fuse. D. P ane l AA 3 ha s a b l o w n fuse. A ns w e r: C. P ane l AA 1 ha s a b l o w n fuse.
E xp l anati on Panel AA1 provides DC power to the 4160V buses A, C, and E indication. Breaker 1FE indication is supplied by Panel AA3 and breaker 1GE indication is supplied by Panel BB3. In order to determine which power supply has been lost the operator must know which power supply is providing the power to the breakers listed. Not all control room breaker indication is powered from the same DC power supply. So the candidate must determine the power supply to the breaker indication. With the breakers listed the candidate determines that Breaker 1FE has indicating lights, so DC Power Panel AA3 has power. Knowing that breaker 1GE has its indicating light illuminated, the candidate knows that DC Power Panel BB3 has power. Knowing that indicating lights for breakers associated with buses A, C, and E NOT being illuminated, the candidate knows that DC Power Panel AA1 has become de
-energized.
Distracters:
12 A. This option is incorrect because Panel BB1 is not the power supply to any of the breakers listed in the question. This option is plausible because Panel BB1 does provide power to other 4160V breaker indications in the control room and other non
-critical division components. The candidate who does not correctly recall the power supplies listed would select this option.
B. This option is incorrect because breaker 1GE has indication lights which are powered from BB3. This option is plausible because BB3 does supply other breaker indication in the control room. The candidate who does not correctly recall the power supplies listed would select this option.
B. This option is incorrect because breaker 1GE has indication lights which are powered from BB3. This option is plausible because BB3 does supply other breaker indication in the control room. The candidate who does not correctly recall the power supplies listed would select this option.
D. This option is incorrect because breaker 1FE has indication lights which are powered from AA3. This option is plausible because AA3 is in the same division power supply supplies other breaker indications in the control room. The candidate who does recognize the divisional power of the breakers but does not correctly recall the power supplies listed would select this option.
D. This option is incorrect because breaker 1FE has indication lights which are powered from AA3. This option is plausible because AA3 is in the same division power supply supplies other breaker indications in the control room. The candidate who does recognize the divisional power of the breakers but does not correctly recall the power supplies listed would select this option.
Technical Reference(s):
Technical Reference(s):       APP 2.3_9-5-2 Rev. 43 Proposed references to be provided to applicants during examination:       ___NONE_______
APP 2.3_9-5-2 Rev. 43 Proposed references to be provided to applicants during examination:
Learning Objective:          COR0020702 OPS DC ELECTRICAL DISTRIBUTION
___NONE_______
: 8. Given a specific DC Electrical Distribution system malfunction, determine the effect on any of the following:
: p. AC Electrical Distribution Question Source:              Bank #                  _      _
Modified Bank #        _______
New                    __X____
Question History:            Last NRC Exam          _      _
Question Cognitive Level:    Memory or Fundamental Knowledge              _____
Comprehension or Analysis                    __X __
10 CFR Part 55 Content:      55.41 (7)
Comments:
LOD 4 13


Learning Objective:
14 15 16 ES-401                           Sample Written Examination                       Form ES-401-5 Question Worksheet Examination Outline Cross-
COR0020702 OPS DC ELECTRICAL DISTRIBUTION
: 8. Given a specific DC Electrical Distribution system malfunction, determine the effect on any of the following:
: p. AC Electrical Distribution Question Source:
Bank #  _        _ Modified Bank #
_______  New  __X____
Question History:
Last NRC Exam
_        _
Question Cognitive Level:
Memory or Fundamental Knowledge
_____ Comprehension or Analysis  __X __  10 CFR Part 55 Content:
55.41 (7)
Comments:  LOD 4  13 14 15 16 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
-


==Reference:==
==Reference:==
 
Level                 RO Tier #                 __1__
Level   RO   Tier #     __1__   Group # __1__   K/A #   _295005.2.1.23___
Group #               __1__
Importance Rating
K/A #                 _295005.2.1.23___
_4.3__   295005 Main Turbine Generator Trip 2.1.23 Ability to perform specific system and integrated plant procedures during all modes of plant operation.
Importance Rating     _4.3__
 
295005 Main Turbine Generator Trip 2.1.23 Ability to perform specific system and integrated plant procedures during all modes of plant operation.
Q ues ti on: 4 R eac t o r po w e r i s 35% du ri n g a s t a rt up. Main Turbine bearing vibrations are as follows:
Question: 4 Reactor power is 35% during a startup.
B ea ri n g 5 v i b r a ti o n ri se s r ap i d l y t o 1 5 m il s an d s t ead i e s ou t. Vibration on b ea ri ng s 4 an d 6 are 7 mils and rising.
Main Turbine bearing vibrations are as follows:
W ha t ac ti on (s) i s/a r e r equ ir ed? A. T ri p t h e M a i n Tu r b i n e ONLY. B. S c r a m t h e R eac t o r AN D tri p t h e M a i n T u r b ine. C. Lo w e r r eac t o r po w e r un ti l bea ri n g v i b r a ti o n l o w e r s t o <1 4 m il s. D. Suspen d t h e s t a rt u p t o a ll o w raising the t u r b i n e cas i n g temperature
* Bearing 5 vibration rises rapidly to 15 mils and steadies out.
. A ns w e r: B. S c r a m t h e R eac t o r AN D tri p t h e M a i n T u r b ine. Explanation:
* Vibration on bearings 4 and 6 are 7 mils and rising.
The Main Turbine (One High Pressure and two Low Pressure), Generator and Exciter are on a single shaft and incorporate 9 bearings. Bearing 1 is on the High Pressure Turbine end of the shaft and Bearing 9 is on the Exciter end of the shaft. There are three bearings (4, 5, and 6) that have shown a rise in vibration, and one of them (5) rises above the action point for tripping the Main Turbine. These bearing are associated with the #2 Low Pressure Turbine. There is no indication given that the number 9 bearing is rising. This bearing is associated with the Exciter. With reactor power >29.5% (164.5 psig first stage pressure), Annunciator 9 2/C-4 is clear. These conditions require the operator to scram the reactor per 2.4TURB.
What action(s) is/are required?
A. Trip the Main Turbine ONLY.
B. Scram the Reactor AND trip the Main Turbine.
C. Lower reactor power until bearing vibration lowers to <14 mils.
D. Suspend the startup to allow raising the turbine casing temperature.
Answer:
B. Scram the Reactor AND trip the Main Turbine.
Explanation:
The Main Turbine (One High Pressure and two Low Pressure), Generator and Exciter are on a single shaft and incorporate 9 bearings. Bearing 1 is on the High Pressure Turbine end of the shaft and Bearing 9 is on the Exciter end of the shaft. There are three bearings (4, 5, and 6) that have shown a rise in vibration, and one of them (5) rises above the action point for tripping the Main Turbine. These bearing are associated with the #2 Low Pressure Turbine. There is no indication given that the number 9 bearing is rising. This bearing is associated with the Exciter.
With reactor power >29.5% (164.5 psig first stage pressure), Annunciator 9-5-2/C-4 is clear.
These conditions require the operator to scram the reactor per 2.4TURB.
Since there is an unexpected turbine or generator vibration rise, there is an entry condition for procedure 2.4TURB. 2.4TURB Attachment 1 High Vibration is applicable.
Since there is an unexpected turbine or generator vibration rise, there is an entry condition for procedure 2.4TURB. 2.4TURB Attachment 1 High Vibration is applicable.
17 Distracters:
17


Distracters:
A. This option is incorrect because a reactor scram is required before tripping the turbine. With the reactor power level given, the reactor is scrammed and then the turbine is tripped.
A. This option is incorrect because a reactor scram is required before tripping the turbine. With the reactor power level given, the reactor is scrammed and then the turbine is tripped.
Tripping the turbine and not scramming the reactor would force a reactor scram and operators should not force an automatic RPS trip. The candidate who potentially focuses on "during a startup" and does not realize the reactor power level is high enough that a turbine trip would cause a reactor scram would select this option.
Tripping the turbine and not scramming the reactor would force a reactor scram and operators should not force an automatic RPS trip. The candidate who potentially focuses on during a startup and does not realize the reactor power level is high enough that a turbine trip would cause a reactor scram would select this option.
 
C. This option is incorrect because lowering reactor power is only taken if bearing 9 vibration is rising above 14 mils. Because this is the action to take for bearing 9, this option is plausible.
C. This option is incorrect because lowering reactor power is only taken if bearing 9 vibration is rising above 14 mils. Because this is the action to take for bearing 9, this option is plausible. The candidate who recalls lowering power to bring bearing vibration down but does not recall it being only for bearing 9 would select this option. Additionally, it would most likely require a significant amount of rod insertion and time to reduce power to the necessary value. This would most likely result in bearing damage in the interim.
The candidate who recalls lowering power to bring bearing vibration down but does not recall it being only for bearing 9 would select this option. Additionally, it would most likely require a significant amount of rod insertion and time to reduce power to the necessary value. This would most likely result in bearing damage in the interim.
D. This option is incorrect because a reactor scram is required. This is an action to be taken if the rotor becomes long during startup which makes this answer plausible due to rotor vibration may be present under this condition. There is no indication turbine expansion is excessive.
D. This option is incorrect because a reactor scram is required. This is an action to be taken if the rotor becomes long during startup which makes this answer plausible due to rotor vibration may be present under this condition. There is no indication turbine expansion is excessive. The candidate who recalls rotor long actions and believes it is causing high vibrations would select this option.
The candidate who recalls rotor long actions and believes it is causing high vibrations would select this option.
Technical Reference(s):        2.4TURB Main Turbine Abnormal, Rev. 30 2.3_9-5-2 (Panel 9-5-2 Annunciator Response Procedure), Rev.
43 Proposed references to be provided to applicants during examination:          ___NONE_______
Learning Objective:
INT0320127, CNS Abnormal Procedures (RO) Turbine/Generator O.      Given plant condition(s), determine from memory if a manual reactor scram or an emergency shutdown from power is required due to the event(s).
P.      Given plant condition(s), determine from memory if a Main Turbine trip is required due to the event(s).
Question Source:                Bank #                  _______
Modified Bank #        _24663_ (See attached)
New                    _______
Question History:              Last NRC Exam          __ _____
Question Cognitive Level:      Memory or Fundamental Knowledge                _____
Comprehension or Analysis                      __X__
10 CFR Part 55 Content:        55.41 (10)
Comments:
LOD: 3 18


Technical Reference(s):
19 20 21 22 23 ES-401                             Sample Written Examination                     Form ES-401-5 Question Worksheet Examination Outline Cross-
2.4TURB Main Turbine Abnormal, Rev. 30 2.3_9-5-2 (Panel 9 2 Annunciator Response Procedure), Rev.
43  Proposed references to be provided to applicants during examination:
___NONE_______
 
Learning Objective:
INT0320127, CNS Abnormal Procedures (RO) Turbine/Generator O. Given plant condition(s), determine from memory if a manual reactor scram or an emergency shutdown from power is required due to the event(s).
P. Given plant condition(s), determine from memory if a Main Turbine trip is required due to the event(s).
Question Source:
Bank #  _______ Modified Bank #
_24663_ (See attached)
New  _______  Question History:
Last NRC Exam
__ _____  Question Cognitive Level:
Memory or Fundamental Knowledge
_____ Comprehension or Analysis  __X__  10 CFR Part 55 Content:
55.41 (10)
Comments: LOD: 3 18 19 20 21 22 23 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
-


==Reference:==
==Reference:==
Level                  RO Tier #                __1__
Group #                __1__
K/A #                  _295006.AK1.01___
Importance Rating      _3.7__
295006 SCRAM- Knowledge of the operational implications of the following concepts as they apply to SCRAM:
AK1.01 Decay heat generation and removal Question: 5 The plant is operating at 100% power on the 221st day of continuous operation when all outboard MSIVs go closed.
* At time T=0, the reactor automatically scrams.
* At time T=3 seconds, reactor pressure spikes to 1100 psig.
* At time T=10 minutes, reactor pressure is in its expected band.
What is the status of the SRVs at T= 10 minutes?
A. 1 SRV is cycling.
B. No SRVs are open or cycling.
C. 1 SRV is open and another SRV is cycling.
D. 2 SRVs are open and another SRV is cycling.
Answer:
A. 1 SRV is cycling.
Explanation:
For the first hour after the Reactor Scram decay generation in the reactor is exponentially decreased from approximately from 7% to 1%. Each of the 8 relief valves are designed to flow 9% (power) when at rated pressure. With the MSIVs shut, the decay heat removal is via the Nuclear Pressure Relief system to the torus. Low-Low Set (LLS) arms under these conditions and RV-71 D controls RPV pressure between 875 and 1010 psig.
Distracters:
B. This answer is incorrect because the decay heat load at 10 minutes following a reactor scram requires at least one SRV (RV-71D) to be periodically cycling to control RPV pressure between 875 psig to 1010 psig. This option is plausible because several hours to one day after a scram if containment conditions are normal then the heat loss to containment would approach that of decay heat and reactor pressure would fall with SRVs all closed. Also if the candidate does not construct an accurate mental model regarding the status of the MSIVs or 24


Level  RO  Tier #    __1__  Group #  __1__  K/A #  _295006.AK1.01___
if the candidate does not remember the amount of decay heat generated following a reactor scram they may very well believe that after 10 minutes decay heat is less than ambient heat loss and SRV actuation is no longer required.
Importance Rating
C. This option is incorrect because RV-71D is controlling reactor pressure between 875 psig to 1010 psig. If one LLS set valve is not enough to maintain Reactor Pressure then the second LLS valve opens at 1040 psig also cycle to between 1040 and 875. Since there is no indication given of an ATWS, a single relief valve is capable of maintaining reactor pressure.
_3.7__    295006 SCRAM
This selection is plausible if the candidate does not recall that a single SRV is capable of maintaining RPV pressure based on an initial decay heat rate of 7%. Additionally, the volumetric flow rate across the SRV is reduced with a reduced reactor pressure. This reduction in flow may lead the candidate to an inaccurate mental model of RPV pressure control.
- Knowledge of the operational implications of the following concepts as they apply to SCRAM:
D. This option is incorrect because RV-71D is controlling RPV pressure between 875 psig to 1010 psig. If one LLS set valve is not enough to maintain Reactor Pressure in the required band, then the second LLS valve opens at 1040 psig and then a 3rd SRV opens at 1090 psig and cycle to maintain RPV pressure. This option is plausible during an ATWS event or if the candidate does not recall that 1 SRV is capable of maintaining reactor pressure based on the decay heat rate of 7%.
AK1.01 Decay heat generation and removal Q ues ti on: 5  Th e p l an t i s ope r a ti n g a t 100% power o n t h e 221 st da y o f con ti nuou s ope r a ti o n w he n a l l ou t boa r d MS I V s g o c losed. At time T=0, th e r eac t o r au t o m a ti ca ll y sc r a m s. At time T=3 seconds, r eac t o r p ressu r e sp i ke s t o 110 0 ps i g. At time T=10 minutes, reactor pressure is in its expected band.
Technical Reference(s):       2.2.1 Nuclear Pressure Relief System, Rev.38 Proposed references to be provided to applicants during examination:         ___NONE_______
Wha t i s t h e s t a t u s o f t h e SRVs at T= 10 minutes?  A. 1 SR V i s cyc ling. B. N o SRV s a r e ope n o r cyc li ng. C. 1 SR V i s ope n an d ano t he r SR V i s cyc ling. D. 2 SRV s a r e ope n an d ano t he r SR V i s cyc ling. A ns w e r: A. 1 SR V i s cyc ling. Explanation:
Learning Objective:           COR0021602 Nuclear Pressure Relief
For the first hour after the Reactor Scram decay generation in the reactor is exponentially decreased from approximately from 7% to 1%. Each of the 8 relief valves are designed to flow 9% (power) when at rated pressure. With the MSIVs shut, the decay heat removal is via the Nuclear Pressure Relief system to the torus. Low
-Low Set (LLS) arms under these conditions and RV-71 D controls RPV pressure between 875 and 1010 psig.
 
Distracters: B. This answer is incorrect because the decay heat load at 10 minutes following a reactor scram requires at least one SRV (RV
-71D) to be periodically cycling to control RPV pressure between 875 psig to 1010 psig. This option is plausible because several hours to one day after a scram if containment conditions are normal then the heat loss to containment would approach that of decay heat and reactor pressure would fall with SRVs all closed. Also if the candidate does not construct an accurate mental model regarding the status of the MSIVs or 24 if the candidate does not remember the amount of decay heat generated following a reactor scram they may very well believe that after 10 minutes decay heat is less than ambient heat loss and SRV actuation is no longer required.
C. This option is incorrect because RV
-71D is controlling reactor pressure between 875 psig to 1010 psig. If one LLS set valve is not enough to maintain Reactor Pressure then the second LLS valve opens at 1040 psig also cycle to between 1040 and 875. Since there is no indication given of an ATWS, a single relief valve is capable of maintaining reactor pressure.
This selection is plausible if the candidate does not recall that a single SRV is capable of maintaining RPV pressure based on an initial decay heat rate of 7%. Additionally, the volumetric flow rate across the SRV is reduced with a reduced reactor pressure. This reduction in flow may lead the candidate to an inaccurate mental model of RPV pressure control. D. This option is incorrect because RV
-71D is controlling RPV pressure between 875 psig to 1010 psig. If one LLS set valve is not enough to maintain Reactor Pressure in the required band, then the second LLS valve opens at 1040 psig and then a 3 rd SRV opens at 1090 psig and cycle to maintain RPV pressure. This option is plausible during an ATWS event or if the candidate does not recall that 1 SRV is capable of maintaining reactor pressure based on the decay heat rate of 7%.
 
Technical Reference(s):
 
====2.2.1 Nuclear====
Pressure Relief System, Rev.38 Proposed references to be provided to applicants during examination:
___NONE_______
Learning Objective:
COR0021602 Nuclear Pressure Relief
: 8. Predict the consequences a malfunction of the following would have on the NPR system:
: 8. Predict the consequences a malfunction of the following would have on the NPR system:
: i. Main Steam system
: i.     Main Steam system
: 12. Given plant conditions, determine if the following should occur:
: 12. Given plant conditions, determine if the following should occur:
: c. SRV/SV opening on safety function.
: c.     SRV/SV opening on safety function.
Question Source:
Question Source:               Bank #               _18065______
Bank #   _18065______
Modified Bank #       _______
Modified Bank #
New                   _______
_______  New   _______ Question History:
Question History:             Last NRC Exam         __ ___
Last NRC Exam
Question Cognitive Level:     Memory or Fundamental Knowledge               _____
__ ___ Question Cognitive Level:
Comprehension or Analysis                     __X__
Memory or Fundamental Knowledge
10 CFR Part 55 Content:       55.41 (8)
_____ Comprehension or Analysis   __X__
Comments:
10 CFR Part 55 Content:
LOD 2 25
55.41 (8)
 
Comments: LOD 2 25 2 6 27 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
26 27 ES-401                           Sample Written Examination                     Form ES-401-5 Question Worksheet Examination Outline Cross-
-


==Reference:==
==Reference:==
 
Level                 RO Tier #               __1__
Level   RO   Tier #     __1__   Group # __1__   K/A #   _295016 AK2.01___
Group #               __1__
Importance Rating
K/A #                 _295016 AK2.01___
_4.4__   295016 Control Room Abandonment:
Importance Rating     _4.4__
295016 Control Room Abandonment:
Knowledge of the interrelations between control room abandonment and the following:
Knowledge of the interrelations between control room abandonment and the following:
(CFR: 41.7 / 45.8)
(CFR: 41.7 / 45.8)
AK2.01 Remote shutdown panel: Plant
AK2.01 Remote shutdown panel: Plant-Specific Question: 6 Following a toxic gas event requiring the control room to be abandoned, the following conditions exist:
-Specific Q ues ti on: 6 Fo ll o w i n g a t ox i c ga s even t requiri n g t h e con tr o l r oo m t o b e abandoned , t h e f o ll o w i n g conditi on s ex i s t:   Th e MS I V s a r e c losed. BO T H Lo w-Lo w S e t (LL S) va l ve s a r e cyc li n g. Th e AD S I SO L A T I O N s w it c h i n t h e A lt e r na t e S hu t do w n R oo m i s no w p lace d i n I SO L A T E.
* The MSIVs are closed.
W h i c h LL S va l v e i s ab l e t o b e con tr o l l e d fr o m t h e AS D r oo m? W h i c h LL S va l v e con tinue s t o cyc l e?   C on tr o ll e d fr o m LL S V a l v e that AS D R oo m  C on ti nue s t o C yc l e  A. RV-71D RV-71F B RV-71F RV-71F C. RV-71D RV-71D D. RV-71F RV-71D A ns w e r: D. RV-71 F  RV-71D Explanation:
* BOTH Low-Low Set (LLS) valves are cycling.
The ADS ISOLATION switch in the Alternate Shutdown Room is now placed in ISOLATE.
Which LLS valve is able to be controlled from the ASD room?
Which LLS valve continues to cycle?
Controlled from      LLS Valve that ASD Room              Continues to Cycle A.       RV-71D             RV-71F B       RV-71F             RV-71F C.       RV-71D             RV-71D D.       RV-71F             RV-71D Answer:
D.       RV-71F              RV-71D Explanation:
There are two LLS valves that automatically control reactor pressure once LLS is activated (RV-71D and RV-71F). In the ASD room, the bottom section of the ADS/REC panel contains two isolation switches. One switch is for Safety Relief Valves 71E, 71F, and 71G. The isolation switch removes valve position status indication from the Control Room and prevents automatic valve operation from Low-Low Set Logic so RV-71F will no longer function in the LLS mode.
28


There are two LLS valves that automatically control reactor pressure once LLS is activated (RV
However, RV-71F will function by valve spring pressure and when manually actuated from the ASD room. No fire exists which would cause spurious equipment operation, therefore LLS valve 71D operation is not effected by the ASD panel switch operation and continues to operate in the LLS mode.
-71D and RV
-71F). In the ASD room, the bottom section of the ADS/REC panel contains two isolation switches. One switch is for Safety Relief Valves 71E, 71F, and 71G. The isolation switch removes valve position status indication from the Control Room and prevents automatic valve operation from Low
-Low Set Logic so RV
-71F will no longer function in the LLS mode.
28 However, RV
-71F will function by valve spring pressure and when manually actuated from the ASD room. No fire exists which would cause spurious equipment operation, therefore LLS valve 71D operation is not effected by the ASD panel switch operation and continues to operate in the LLS mode.
Distracters:
Distracters:
A. This option is incorrect because Low-Low set valve 71D can continue to cycle. This selection is plausible since ASD switch manipulation is an uncommon occurrence and the candidate may not internalize that the LLS logic has been removed from 71F.
B. This option is incorrect because Low-Low set valve controlled from the ASD room is not able to cycle on Low-Low set. This selection is plausible since ASD switch manipulation is an uncommon occurrence and the candidate may not remember that LLS logic has been removed from 71F with the ASD switch.
C. This option is incorrect because Low-Low set valve 71D can continue to cycle. This selection is plausible since ASD switch manipulation is an uncommon occurrence and the candidate may have the inaccurate mental model that LLS logic is removed from both 71D and 71F under these conditions.
Technical Reference(s): GE Electrical Drawing 753E253 Sheet 2.
Proposed references to be provided to applicants during examination:      ___NONE_______
Learning Objective:            COR002-34-02 Ops Alternate Shutdown
: 2.      Describe the interrelationship between ASD and the following:
: a.      Nuclear Pressure Relief (NPR) system Question Source:              Bank #                _      _
Modified Bank #        21372 _ (See attached)
New                  __ _____
Question History:              Last NRC Exam        ___2006___
Question Cognitive Level:      Memory or Fundamental Knowledge Comprehension or Analysis                  X 10 CFR Part 55 Content:        55.41 (10)
Comments:
LOD 3 29


A. This option is incorrect because Low
From CNS 2006 NRC Exam 30
-Low set valve 71D can continue to cycle. This selection is plausible since ASD switch manipulation is an uncommon occurrence and the candidate may not internalize that the LLS logic has been removed from 71F. B. This option is incorrect because Low
-Low set valve controlled from the ASD room is not able to cycle on Low
-Low set. This selection is plausible since ASD switch manipulation is an uncommon occurrence and the candidate may not remember that LLS logic has been removed from 71F with the ASD switch.
C. This option is incorrect because Low
-Low set valve 71D can continue to cycle. This selection is plausible since ASD switch manipulation is an uncommon occurrence and the candidate may have the inaccurate mental model that LLS logic is removed from both 71D and 71F under these conditions.


Technical Reference(s): GE Electrical Drawing 753E253 Sheet 2.
31 32 33 34 791E253 Sheet 35
Proposed references to be provided to applicants during examination:
___NONE_______


Learning Objective:
ES-401                           Sample Written Examination                       Form ES-401-5 Question Worksheet Examination Outline Cross-
COR002-34-02 Ops Alternate Shutdown
: 2. Describe the interrelationship between ASD and the following:
: a. Nuclear Pressure Relief (NPR) system Question Source:
Bank #  _          _
Modified Bank #
21372 _ (See attached)
New  __ _____  Question History:
Last NRC Exam
___2006___
Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis  X  10 CFR Part 55 Content:
55.41 (10)
Comments:  LOD 3  29 From CNS 2006 NRC Exam 30 31 32 33 34 791E253 Sheet 35 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
-


==Reference:==
==Reference:==
 
Level               RO Tier #               __1__
Level   RO   Tier #     __1__   Group # __1__   K/A #   _295018 AK3.03___
Group #             __1__
Importance Rating
K/A #               _295018 AK3.03___
_3.1__   295018 Partial or Total Loss of CCW:
Importance Rating   _3.1__
295018 Partial or Total Loss of CCW:
Knowledge of the reasons for the following responses as they apply to partial or complete loss of component cooling water:
Knowledge of the reasons for the following responses as they apply to partial or complete loss of component cooling water:
(CFR: 41.5 / 45.6)
(CFR: 41.5 / 45.6)
AK3.03 Securing individual components (prevent equipment damage)
AK3.03 Securing individual components (prevent equipment damage)
Q ues ti on: 7 Th e p l an t i s ope r a ti n g a t r a t e d po w e r. A l os s o f a l l RE C pu m p s occu r s. 5.2REC , LOS S o f RE C i s en t e red. A l l a tt e m p t s t o r es t o r e RE C hav e f a il ed. Th e reac t o r i s manua ll y sc r a mm ed. W h y i s t h e runn i n g CR D pu m p requ ir e d t o b e secu red? A. P reven t r eac t o r vesse l s tr atifi cati on. B. P reven t r eac t o r w a t e r leve l overfill. C. P reven t pu m p bea ri n g o ve r heati ng. D. P reven t pump m o t o r bearing ove r heati ng. A ns w e r: C. P reven t pu m p bea ri n g ove r heati ng. Explanation:
Question: 7 The plant is operating at rated power.
The REC system supplies cooling to the CRD pump bearing and oil cooler via the non
* A loss of all REC pumps occurs.
-critical supply loop. REC's only safety function is to provide cooling to the Reactor Building Quadrant room Fan Coil Units for RHR pump operation. With the loss of REC cooling the running CRD pump must be secured to prevent pump damage due to overheating. While the quadrant fan coil units provide cooling for the RHR pumps, they are not credited for keeping the CRD pump motor cool for operation. The CRD pump injects water in the bottom head region of the reactor vessel which can be a part of the reason for stratification events which occur with low core flow and cool CRD water amassing in the bottom of the reactor vessel.
* 5.2REC, LOSS of REC is entered.
With the loss of REC the reactor recirculation pumps are secured so core flow is low. Also with continued CRD injection during a scram, CRD is supplying RPV injection around 140 gpm which can eventually lead to overfilling 36 the RPV if the scram is not reset or CRD
* All attempts to restore REC have failed.
-29 valve is not closed. With the new reactor vessel level control system and setpoint setdown overfill events are not as likely, but can occur if the operator does not keep track of RPV level and take the actions described here.
* The reactor is manually scrammed.
Why is the running CRD pump required to be secured?
A. Prevent reactor vessel stratification.
B. Prevent reactor water level overfill.
C. Prevent pump bearing overheating.
D. Prevent pump motor bearing overheating.
Answer:
C. Prevent pump bearing overheating.
Explanation:
The REC system supplies cooling to the CRD pump bearing and oil cooler via the non-critical supply loop. RECs only safety function is to provide cooling to the Reactor Building Quadrant room Fan Coil Units for RHR pump operation. With the loss of REC cooling the running CRD pump must be secured to prevent pump damage due to overheating. While the quadrant fan coil units provide cooling for the RHR pumps, they are not credited for keeping the CRD pump motor cool for operation. The CRD pump injects water in the bottom head region of the reactor vessel which can be a part of the reason for stratification events which occur with low core flow and cool CRD water amassing in the bottom of the reactor vessel. With the loss of REC the reactor recirculation pumps are secured so core flow is low. Also with continued CRD injection during a scram, CRD is supplying RPV injection around 140 gpm which can eventually lead to overfilling 36
 
the RPV if the scram is not reset or CRD-29 valve is not closed. With the new reactor vessel level control system and setpoint setdown overfill events are not as likely, but can occur if the operator does not keep track of RPV level and take the actions described here.
From COR002-19-02 OPS Reactor Equipment Cooling
From COR002-19-02 OPS Reactor Equipment Cooling
: 5. A sustained loss of REC to the CRD pumps would cause damage to the pumps since the bearing and oil coolers are supplied by the REC non
: 5. A sustained loss of REC to the CRD pumps would cause damage to the pumps since the bearing and oil coolers are supplied by the REC non-critical equipment supply loop.
-critical equipment supply loop.
Distracters:
Distracters:
A. This option is incorrect because vessel stratification is not an issue requiring removing the CRD pump from service. The CRD system injects into the bottom head region via the CRD mechanisms. If core flow is low and the CRD system is injecting, the colder water can collect in the bottom head region and cause bottom head metal temperatures to be much colder than the rest of the vessel metal temperatures. The vessel is then considered stratified. Stopping CRD flow is a means of precluding stratification, if the core flow is low because the RR pumps are tripped. However, with normal reactor scram conditions, and reactor feedwater pumps controlling RPV level, the stratification can be prevented by closing CRD
A. This option is incorrect because vessel stratification is not an issue requiring removing the CRD pump from service. The CRD system injects into the bottom head region via the CRD mechanisms. If core flow is low and the CRD system is injecting, the colder water can collect in the bottom head region and cause bottom head metal temperatures to be much colder than the rest of the vessel metal temperatures. The vessel is then considered stratified.
-29, charging water isolation valve. The CRD pump is tripped because of a lack of REC cooling and not because of stratification issues.
Stopping CRD flow is a means of precluding stratification, if the core flow is low because the RR pumps are tripped. However, with normal reactor scram conditions, and reactor feedwater pumps controlling RPV level, the stratification can be prevented by closing CRD-29, charging water isolation valve. The CRD pump is tripped because of a lack of REC cooling and not because of stratification issues.
D. This option is incorrect because REC does not provide cooling to the motor bearing. REC is lost to the quad fan coil units so the quadrant temperatures rise but there are no restrictions for CRD pump operation based on the quadrant temperatures. If the candidate does not know that CRD pumps are not restricted from operating due to increased room temperature they would select this option. This option is plausible because the quadrant temperature rises and the CRD pump motor operating temperature also rises.
D. This option is incorrect because REC does not provide cooling to the motor bearing. REC is lost to the quad fan coil units so the quadrant temperatures rise but there are no restrictions for CRD pump operation based on the quadrant temperatures. If the candidate does not know that CRD pumps are not restricted from operating due to increased room temperature they would select this option. This option is plausible because the quadrant temperature rises and the CRD pump motor operating temperature also rises.
B. This option is incorrect because vessel overfill is not an issue requiring removing the CRD pump from service. With a scram present, CRD system flow into the RPV is approximately 140 gpm. This relatively cool water expands after it is injected into the RPV. With reactor feedwater pumps controlling RPV level, the overfill can be prevented by closing CRD
B. This option is incorrect because vessel overfill is not an issue requiring removing the CRD pump from service. With a scram present, CRD system flow into the RPV is approximately 140 gpm. This relatively cool water expands after it is injected into the RPV. With reactor feedwater pumps controlling RPV level, the overfill can be prevented by closing CRD-29, charging water isolation valve. With the old reactor vessel level control system it was common to overfill the RPV due to CRD injection. The new reactor vessel level control system minimizes these events but they can still happen if the operators are slow to reset the reactor scram so this option is plausible for this reason.
-29, charging water isolation valve. With the old reactor vessel level control system it was common to overfill the RPV due to CRD injection. The new reactor vessel level control system minimizes these events but they can still happen if the operators are slow to reset the reactor scram so this option is plausible for this reason.
Technical Reference(s):         5.2REC Loss of REC, Rev. 16 Proposed references to be provided to applicants during examination:         ___NONE_______
Technical Reference(s):
Learning Objective: COR002-19-02 OPS Reactor Equipment Cooling
5.2REC Loss of REC, Rev. 16 Proposed references to be provided to applicants during examination:
: 6.       Given a specific REC malfunction, determine the effect on any of the following:
___NONE_______
: g.     CRDH system Question Source:               Bank #                 _     _
Learning Objective:
Modified Bank #         _      _
COR002-19-02 OPS Reactor Equipment Cooling
New                     ____X _
: 6. Given a specific REC malfunction, determine the effect on any of the following:
Question History:               Last NRC Exam           ___ ___
: g. CRDH system Question Source:
Question Cognitive Level:       Memory or Fundamental Knowledge             _ __
Bank #   _         _
37
Modified Bank #
 
_
Comprehension or Analysis __ X _
_  New   ____X _ Question History:
10 CFR Part 55 Content: 55.41 (5)
Last NRC Exam
Comments:
___ ___ Question Cognitive Level:
LOD 3 38
Memory or Fundamental Knowledge
 
_   __ 37 Comprehension or Analysis   __ X _ 10 CFR Part 55 Content:
39 ES-401                             Sample Written Examination                     Form ES-401-5 Question Worksheet Examination Outline Cross-
55.41 (5) Comments: LOD 3   38 39 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
-


==Reference:==
==Reference:==
 
Level                 RO Tier #               __1__
Level   RO   Tier #     __1__   Group # __1__   K/A #   _295019.AA1.02___
Group #               __1__
Importance Rating
K/A #                 _295019.AA1.02___
_3.3__   295019 Partial or Total Loss of Inst. Air
Importance Rating     _3.3__
- Ability to operate and/or monitor the following as they apply to partial or complete loss of instrument air:
295019 Partial or Total Loss of Inst. Air- Ability to operate and/or monitor the following as they apply to partial or complete loss of instrument air:
AA1.02 Instrument air system valves: Plant
AA1.02 Instrument air system valves: Plant-Specific Question: 8 The plant is operating near rated power.
-Specific Q ues ti on: 8 Th e p l an t i s ope r a ti n g nea r r a t e d po w e r. A station operator has reported that a l ea k ha s deve l oped in the Augmented Radwaste Building basement a i r sys t e m. Ins tr u m en t a i r heade r p ressu re has lowered t o 7 5 ps ig and has stabilized
* A station operator has reported that a leak has developed in the Augmented Radwaste Building basement air system.
. A reactor scram has been inserted.
* Instrument air header pressure has lowered to 75 psig and has stabilized.
W ha t ac ti o n i s requ ir ed IAW Procedure 5.2AIR (Loss of Instrument Air)
* A reactor scram has been inserted.
?  A. C l os e I A-SOV-21 , D r y w e l l Instrument Air Supp l y V a lve. B. C l os e I A-MO-80 , N o n Criti ca l I ns tr u m en t A i r I so l a ti o n V a lve.
What action is required IAW Procedure 5.2AIR (Loss of Instrument Air)?
C. O pe n SA-MO-81 , Service Air t o Instrument Air C ross ti e V a lve.
A. Close IA-SOV-21, Drywell Instrument Air Supply Valve.
D. Ensu r e SA-AO-PCV-609 , S e r v i c e A i r S ys t e m I so l a ti o n V a l v e i s open. A ns w e r: B. C l os e I A-MO-80 , N o n Criti ca l I ns tr u m en t A i r I so l a ti o n V a lve.
B. Close IA-MO-80, Non Critical Instrument Air Isolation Valve.
C. Open SA-MO-81, Service Air to Instrument Air Crosstie Valve.
D. Ensure SA-AO-PCV-609, Service Air System Isolation Valve is open.
Answer:
B. Close IA-MO-80, Non Critical Instrument Air Isolation Valve.
Explanation:
Explanation:
A leak is present in the non
A leak is present in the non-critical instrument air system. This leak is large enough to lower the instrument air header pressure and has stabilized. Since the station operator has located the source of the leak in a timely manner, closing IA-MO-80 from the control room isolates the leak and allows the instrument air header pressure to recover.
-critical instrument air system. This leak is large enough to lower the instrument air header pressure and has stabilized. Since the station operator has located the source of the leak in a timely manner, closing IA
-MO-80 from the control room isolates the leak and allows the instrument air header pressure to recover.
Distracters:
Distracters:
A. This option is incorrect because IA
A. This option is incorrect because IA-SOV-21, Drywell IA Supply Valve is used to supply back up IA to the inboard MSIV in the event Nitrogen to the DW is lost. Although this valve is normally closed while operating at power, the procedure provides guidance to ensure this valve is closed which would be applicable during plant shutdown conditions and would help preserve air required to operate other plant components. The candidate would chose this answer if they believe this valve to be normally open and that by shutting the valve it would preserve pneumatics to the drywell.
-SOV-21, Drywell IA Supply Valve is used to supply back up IA to the inboard MSIV in the event Nitrogen to the DW is lost. Although this valve is normally closed while operating at power, the procedure provides guidance to ensure this valve is closed which would be applicable during plant shutdown conditions and would help preserve air required to operate other plant components. The candidate would chose this answer if they believe this valve to be normally open and that by shutting the valve it would preserve pneumatics to the drywell.
40
40 C. This option is incorrect because SA
 
-MO-81 is required to be opened by the control room operator when Air pressure lowers to 85 psig and it has been determined that a clogged instrument air dryer is the cause of lowering air pressure. The candidate would select this if he/she remembered it is an action that is identified for IA pressure below 85 psig (but does not remember the other requirement for confirmation of a clogged dryer).
C. This option is incorrect because SA-MO-81 is required to be opened by the control room operator when Air pressure lowers to 85 psig and it has been determined that a clogged instrument air dryer is the cause of lowering air pressure. The candidate would select this if he/she remembered it is an action that is identified for IA pressure below 85 psig (but does not remember the other requirement for confirmation of a clogged dryer).
D. This option is incorrect because SA
D. This option is incorrect because SA-AO-PCV-609, Service Air System Isolation Valve automatically closes at >77 psig to isolate the Service Air Header from the Instrument Air Header. The procedure has an action to reopen this valve once it has been verified there is no leak in SA header. The candidate would chose this if he/she does not remember this valve automatically closes on low air pressure or believes the valve must be opened to recover SA pressure. Since recovery of SA is not the goal of the procedure then this action is not correct. This action, if it were able to be accomplished, would jeopardize the IA system which is why it is important for a candidate to discriminate against this option Technical Reference(s): 5.2Air Loss of Instrument Air, Rev. 19 Proposed references to be provided to applicants during examination:         ___NONE_______
-AO-PCV-609, Service Air System Isolation Valve automatically closes at >77 psig to isolate the Service Air Header from the Instrument Air Header. The procedure has an action to reopen this valve once it has been verified there is no leak in SA header. The candidate would chose this if he/she does not remember this valve automatically closes on low air pressure or believes the valve must be opened to recover SA pressure. Since recovery of SA is not the goal of the procedure then this action is not correct. This action, if it were able to be accomplished, would jeopardize the IA system which is why it is important for a candidate to discriminate against this option Technical Reference(s): 5.2Air Loss of Instrument Air, Rev. 19 Proposed references to be provided to applicants during examination: ___NONE_______
Learning Objective:
Learning Objective:
7 Given a specific Plant Air system malfunction, determine the effect on any of the following:
7     Given a specific Plant Air system malfunction, determine the effect on any of the following:
: a. Plant operation Question Source:
: a. Plant operation Question Source:               Bank #                 _______
Bank #   _______ Modified Bank #
Modified Bank #         _3979_ (See attached)
_3979_ (See attached)
New                     _______
New   _______ Question History: Last NRC Exam
Question History:             Last NRC Exam           ___ ____
___ ____
Question Cognitive Level:     Memory or Fundamental Knowledge Comprehension or Analysis                     X 10 CFR Part 55 Content:       55.41 (7)
Question Cognitive Level:
Comments:
Memory or Fundamental Knowledge Comprehension or Analysis   X 10 CFR Part 55 Content:
LOD: 2 41
55.41 (7) Comments: LOD: 2 41 42 43 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
 
-
42 43 ES-401                           Sample Written Examination                               Form ES-401-5 Question Worksheet Examination Outline Cross-


==Reference:==
==Reference:==
 
Level                   RO Tier #                   __1__
Level   RO   Tier #     __1__   Group # __1__   K/A #   _295021 AA2.01___
Group #                 __1__
Importance Rating
K/A #                   _295021 AA2.01___
_3.5__   295021 Loss of Shutdown Cooling:
Importance Rating       _3.5__
Ability to determine and/or interpret the following as they apply to loss of shutdown cooling: (CFR: 41.10 / 43.5 / 45.13)
295021 Loss of Shutdown Cooling:
AA2.01 Reactor water heatup/cooldown rate Q ues ti on: 9 Th e p l an t i s shu t do w n f o r r e f ue li n g w it h t h e fo ll o w i n g conditions:   Th e reac t o r ha s bee n shu t do w n f o r 4 8 hou r s. RP V w a t e r leve l i s be i n g m a i n t a i ne d 5 0 t o 6 0 inches. RH R H X i n l e t t e m pe r a t u r e i nd i ca t e s 1 00&#xba;F. W ha t i s M I N I MU M ti m e f o r t h e reac t o r coo l an t t o r eac h 212&#xba;F i f shu t do w n coo li n g i s l os t?   A. 2.1 hou r s  B. 2.9 hours C. 4.9 hou r s  D. 13.1 hou r s  A ns w e r: A. 2.1 hou r s Explanation:
Ability to determine and/or interpret the following as they apply to loss of shutdown cooling:
Determining the Time to boil is an interpretation of heatup rate (Tfinal -Tinitial)/Time This requires knowledge of time after shutdown, initial reactor coolant temperature, and reactor coolant inventory available. Based upon the reactor being shutdown for 48 hours, initial Reactor Coolant Temperature at 100&deg;F, and RPV level at the high level trip results in an estimated time to boil of 2.1 hours. Abnormal Procedure 2.4SDC, Shutdown Cooling Abnormal, Attachment 5 contains a family of curves based upon RPV (or cavity) water level. Using the water level at the high level trip curves and hours after shutdown (not Days after shutdown) and the given reactor coolant temperature, the answer can be determined. Interpolation on the family of curves is allowed.
(CFR: 41.10 / 43.5 / 45.13)
AA2.01 Reactor water heatup/cooldown rate Question: 9 The plant is shutdown for refueling with the following conditions:
* The reactor has been shutdown for 48 hours.
* RPV water level is being maintained 50 to 60 inches.
* RHR HX inlet temperature indicates 100&#xba;F.
What is MINIMUM time for the reactor coolant to reach 212&#xba;F if shutdown cooling is lost?
A. 2.1 hours B. 2.9 hours C. 4.9 hours D. 13.1 hours Answer:
A. 2.1 hours Explanation:
Determining the Time to boil is an interpretation of heatup rate (T final -T initial )/Time This requires knowledge of time after shutdown, initial reactor coolant temperature, and reactor coolant inventory available. Based upon the reactor being shutdown for 48 hours, initial Reactor Coolant Temperature at 100&deg;F, and RPV level at the high level trip results in an estimated time to boil of 2.1 hours. Abnormal Procedure 2.4SDC, Shutdown Cooling Abnormal, Attachment 5 contains a family of curves based upon RPV (or cavity) water level. Using the water level at the high level trip curves and hours after shutdown (not Days after shutdown) and the given reactor coolant temperature, the answer can be determined. Interpolation on the family of curves is allowed.
Distracters:
Distracters:
44 B. This option is incorrect because the time to boil under the given conditions is 2.1 hours. This option is plausible if the Water Level at the flange graph is confused with the water level at the high level trip curve. The candidate who uses the 48 hours on the Water Level at the flange graph would select this option.
44


B. This option is incorrect because the time to boil under the given conditions is 2.1 hours. This option is plausible if the Water Level at the flange graph is confused with the water level at the high level trip curve. The candidate who uses the 48 hours on the Water Level at the flange graph would select this option.
C. This option is incorrect because the time to boil under the given conditions is 2.1 hours. This option is plausible if the time since shutdown is confused with 48 days vs. hours and water level at the flange. The candidate who uses the 48 days on the Water Level at high level trip graph would select this option.
C. This option is incorrect because the time to boil under the given conditions is 2.1 hours. This option is plausible if the time since shutdown is confused with 48 days vs. hours and water level at the flange. The candidate who uses the 48 days on the Water Level at high level trip graph would select this option.
D. This option is incorrect due to time to boil under the given conditions is 2.1 hours. This option is plausible if the time since shutdown is confused with 48 hours with water level flooded to 1001. The candidate who uses the 48 hours on the Water Level to Level Flooded to 1001 would select this option.
Technical Reference(s):        2.4SDC, Shutdown Cooling Abnormal, Rev 14 Proposed references to be provided to applicants during examination:        2.4SDC Attachment 5 Learning Objective:            INT0231002001170A Give a set of plant conditions and time of the reactor shutdown, determine: Time to core boiling.
Question Source:                Bank #                _      _
Modified Bank #      9181 _ (See attached)
New                  __ ____
Question History:              Last NRC Exam        ___ ___
Question Cognitive Level:      Memory or Fundamental Knowledge Comprehension or Analysis                    X 10 CFR Part 55 Content:        55.41 (14)
Comments:
LOD 3 45


D. This option is incorrect due to time to boil under the given conditions is 2.1 hours. This option is plausible if the time since shutdown is confused with 48 hours with water level flooded to 1001'. The candidate who uses the 48 hours on the Water Level to Level Flooded to 1001' would select this option.
46 47 48 49 ES-401                             Sample Written Examination                           Form ES-401-5 Question Worksheet Examination Outline Cross-
 
Technical Reference(s):
2.4SDC, Shutdown Cooling Abnormal, Rev 14
 
Proposed references to be provided to applicants during examination:
2.4SDC Attachment 5 Learning Objective:
INT0231002001170A Give a set of plant conditions and time of the reactor shutdown, determine: Time to core boiling.
 
Question Source:
Bank #  _          _
Modified Bank #
9181  _  (See attached)
New  __ ____  Question History:
Last NRC Exam
___ ___  Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis  X  10 CFR Part 55 Content:
55.41 (14)  Comments: LOD 3 45 46 47 48 49 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
-


==Reference:==
==Reference:==
 
Level                   RO Tier #                 __1__
Level   RO   Tier #     __1__   Group # __1__   K/A #   _295023 2.2.40___
Group #                 __1__
Importance Rating
K/A #                   _295023 2.2.40___
_3.4__   295023 Refueling Accidents:
Importance Rating       _3.4__
295023 Refueling Accidents:
2.2.40 Ability to apply Technical Specifications for a system.
2.2.40 Ability to apply Technical Specifications for a system.
(CFR: 41.10 / 43.2 / 43.5 / 45.3)
(CFR: 41.10 / 43.2 / 43.5 / 45.3)
Q ues ti on: 1 0  Th e p l an t i s i n da y 1 2 o f a r e f ue li n g ou tage. R e f ue li n g ope r a ti on s a r e i n p r og ress. A con tr o l r o d i s acc i den t a ll y d r oppe d i n t o t h e RPV. W ha t i s t h e requ ir e d M I N I MU M Techn i ca l S pecifi ca ti on s w a t e r leve l abov e t h e t o p o f t h e RP V flang e that ensu r es o f f s it e dos e i s m a i n t a i ne d w it h i n t h e a ll o w ab l e li m it s?   A. 6 feet B. 12 feet C. 21 feet D. 37 feet A ns w e r: C. 21 feet Explanation:
Question: 10 The plant is in day 12 of a refueling outage.
TS 3.9.6 (REFUELING OPERATIONS)  
* Refueling operations are in progress.
- Reactor Pressure Vessel (RPV) Water Level requires assemblies with in the RPV, during movement of new fuel assemblies or handling of control rods within the RPV, when irradiate fuel assemblies are seated within the RPV. This minimum level retains iodine fission product activity to limit offsite doses in the event of a refueling accident.
* A control rod is accidentally dropped into the RPV.
What is the required MINIMUM Technical Specifications water level above the top of the RPV flange that ensures off site dose is maintained within the allowable limits?
A. 6 feet B. 12 feet C. 21 feet D. 37 feet Answer:
C. 21 feet Explanation:
TS 3.9.6 (REFUELING OPERATIONS) - Reactor Pressure Vessel (RPV) Water Level requires RPV water level to be 21 ft above the top of the RPV flange during movement of irradiated fuel assemblies with in the RPV, during movement of new fuel assemblies or handling of control rods within the RPV, when irradiate fuel assemblies are seated within the RPV. This minimum level retains iodine fission product activity to limit offsite doses in the event of a refueling accident.
Distracters:
Distracters:
A. This option is incorrect because the listed water level is too low. The required level is a minimum of 21 feet above the flange. This option is plausible because it is the minimum water level above a suspended fuel bundle on the refuel bridge and the candidate may recall this number and select this option.
A. This option is incorrect because the listed water level is too low. The required level is a minimum of 21 feet above the flange. This option is plausible because it is the minimum water level above a suspended fuel bundle on the refuel bridge and the candidate may recall this number and select this option.
50 B. This option is incorrect because the listed water level is too low. The required level is a minimum of 21 feet above the flange. This option is plausible because it is the height of a fuel bundle or a control rod and the candidate may recall this number and select this option.
50
 
B. This option is incorrect because the listed water level is too low. The required level is a minimum of 21 feet above the flange. This option is plausible because it is the height of a fuel bundle or a control rod and the candidate may recall this number and select this option.
D. This option is incorrect because the listed water level is too high. The required level is a minimum of 21 feet above the flange. This option is plausible because it is the normal spent fuel water level and the candidate may recall this number and select this option.
D. This option is incorrect because the listed water level is too high. The required level is a minimum of 21 feet above the flange. This option is plausible because it is the normal spent fuel water level and the candidate may recall this number and select this option.
Technical Reference(s):      Technical Specifications 3.9.6 Proposed references to be provided to applicants during examination:        ___NONE_______
Learning Objective:          INT007-05-10 OPS CNS Tech Specs 3.9, Refueling Operations
: 1. Given a set of plant conditions, recognize non-compliance with a Section 3.9 LCO.
Question Source:              Bank #                _      _
Modified Bank #        _        _
New                    ___X___
Question History:            Last NRC Exam          ___ ___
Question Cognitive Level:    Memory or Fundamental Knowledge                __X__
Comprehension or Analysis                      __ __
10 CFR Part 55 Content:      55.41 (10)
Comments:
LOD 2 51


Technical Reference(s):
52 53 ES-401                           Sample Written Examination                   Form ES-401-5 Question Worksheet Examination Outline Cross-
Technical Specifications 3.9.6 Proposed references to be provided to applicants during examination:
___NONE_______
Learning Objective:
INT007-05-10 OPS CNS Tech Specs 3.9, Refueling Operations
: 1. Given a set of plant conditions, recognize non
-compliance with a Section 3.9 LCO.
Question Source:
Bank #  _      _ Modified Bank #
_
_  New  ___X___  Question History:
Last NRC Exam
___ ___  Question Cognitive Level:
Memory or Fundamental Knowledge
__X__ Comprehension or Analysis  __  __
10 CFR Part 55 Content:
55.41 (10)
Comments: LOD 2  51 52 53 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
-


==Reference:==
==Reference:==
 
Level               RO Tier #               __1__
Level   RO   Tier #     __1__   Group # __1__   K/A #   _295024EK1.01___
Group #             __1__
Importance Rating
K/A #               _295024EK1.01___
_4.1__   295024 High Drywell Pressure
Importance Rating   _4.1__
- Knowledge of the operational implications of the following concepts as they apply to high drywell pressure:
295024 High Drywell Pressure- Knowledge of the operational implications of the following concepts as they apply to high drywell pressure:
EK1.01 Drywell integrity: Plant
EK1.01 Drywell integrity: Plant-Specific Question: 11 During an ATWS, a LOCA occur s wit h the following conditions present:
-Specific Q ues ti on: 1 1  Du ri n g an ATWS, a L O C A occurs with t h e f o ll ow i n g conditi on s present:   EOP 6A and 7A have been entered.
* EOP 6A and 7A have been entered.
Ave r ag e d rywe l l t e m pe r a t u r e i s 210&deg;F an d s t eady. D rywe l l p ressu r e i s 1 9 ps i g an d ri s i ng. Ave r ag e supp ress i o n poo l t e m pe r a t u r e i s 205&deg;F an d ri s i ng. To r u s wa t e r leve l i s 1 4 f ee t an d ri s ing slowly. Reac t o r p ressu r e i s 100 0 ps i g an d s teady. Why is an Emergency Depressurization required at this time IAW EOP 3A (Primary Containment Control)
* Average drywell temperature is 210&deg;F and steady.
?  To prevent-A. chugging, and possible loss of the pressure suppression function.
* Drywell pressure is 19 psig and rising.
* Average suppression pool temperature is 205&deg;F and rising.
* Torus water level is 14 feet and rising slowly.
* Reactor pressure is 1000 psig and steady.
Why is an Emergency Depressurization required at this time IAW EOP 3A (Primary Containment Control)?
To prevent A. chugging, and possible loss of the pressure suppression function.
B. raising torus pressure above the PCPL A which may result in failure of containment.
B. raising torus pressure above the PCPL A which may result in failure of containment.
C. excessive torus to drywell vacuum breaker operation and possible vacuum breaker failure.
C. excessive torus to drywell vacuum breaker operation and possible vacuum breaker failure.
D. the torus water level rise which may cause loss of containment on SRV actuation due to a water column in the system discharge piping
D. the torus water level rise which may cause loss of containment on SRV actuation due to a water column in the system discharge piping.
. A ns w e r: B. raising torus pressure above the PCPL A which may result in failure of containment.
Answer:
B. raising torus pressure above the PCPL A which may result in failure of containment.
Explanation:
Explanation:
Operation is on the wrong side of the Heat Capacity Temperature limit curve for the given reactor pressure. Emergency depressurizing too far above this point has the potential to lead to exceeding the PCPL A limit which could result in loss of containment integrity. 54 Current conditions place the plant on the unsafe side of the HCTL Curve.
Operation is on the wrong side of the Heat Capacity Temperature limit curve for the given reactor pressure. Emergency depressurizing too far above this point has the potential to lead to exceeding the PCPL A limit which could result in loss of containment integrity.
A. Heat Capacity Temperature Limit (HCTL) (GRAP07)  
54
: 1. Definition  
 
- the highest torus water temperature from which emergency RPV depressurization will not raise suppression chamber pressure above Primary Containment Pressure Limit A, while the rate of energy transfer from the RPV to the containment is greater than the capacity of the containment vent.
Current conditions place the plant on the unsafe side of the HCTL Curve.
: 2. Use - The HCTL is a function of RPV pressure and primary containment water level. It is utilized to preclude failure of the containment and equipment in the containment necessary for the safe shutdown of the plant.
A.       Heat Capacity Temperature Limit (HCTL) (GRAP07)
: 1.       Definition - the highest torus water temperature from which emergency RPV depressurization will not raise suppression chamber pressure above Primary Containment Pressure Limit A, while the rate of energy transfer from the RPV to the containment is greater than the capacity of the containment vent.
: 2.       Use - The HCTL is a function of RPV pressure and primary containment water level. It is utilized to preclude failure of the containment and equipment in the containment necessary for the safe shutdown of the plant.
Flowchart 3A requires emergency RPV depressurization when RPV pressure and torus water temperature cannot be maintained within HCTL.
Flowchart 3A requires emergency RPV depressurization when RPV pressure and torus water temperature cannot be maintained within HCTL.
 
B.       Primary Containment Pressure Limits A/B (GRAP11)
B. Primary Containment Pressure Limits A/B (GRAP11)  
: 1.       Definition - the lesser of either:
: 1. Definition  
: a.       The pressure capability of the primary containment, or
- the lesser of either:
: b.       The maximum primary containment pressure at which vent valves sized to reject all decay heat from the containment can be opened and closed.
: a. The pressure capability of the primary containment, or b. The maximum primary containment pressure at which vent valves sized to reject all decay heat from the containment can be opened and closed.
55
55  
: c.       For Primary Containment Pressure Limit A, the maximum primary containment pressure at which SRVs can be opened and will remain open.
: c. For Primary Containment Pressure Limit A, the maximum primary containment pressure at which SRVs can be opened and will remain open.
: 2.       Use - Each PCPL is a function of primary containment water level and primary containment pressure. The limits are utilized to avoid challenges to primary containment vent valve operability, SRV operability, and primary containment integrity.
: 2. Use - Each PCPL is a function of primary containment water level and primary containment pressure. The limits are utilized to avoid challenges to primary containment vent valve operability, SRV operability, and primary containment integrity.
RPV vent valve operability is not a concern in derivation of the PCPL because RPV venting can be accomplished using the motor operated main steam line drain valves.
RPV vent valve operability is not a concern in derivation of the PCPL because RPV venting can be accomplished using the motor operated main steam line drain valves. Operability of these valves are not affected by containment atmospheric pressure.
Operability of these valves are not affected by containment atmospheric pressure.
Distracters:
Distracters:
A. This option is incorrect because chugging is not an issue with emergency depressurizing. Chugging is an issue when drywell sprays are initiated when suppression chamber pressure exceeds the Suppression Chamber Spray Initiation Pressure (SCSIP) to preclude chugging-the cyclic condensation of steam at the downcomer openings of the drywell vents. When a steam bubble collapses at the exit of the downcomers, the rush of water drawn into the downcomers to fill the void induces stresses at the junction of the downcomers and the vent header. Repeated application of such stresses could cause fatigue failure of these joints, thereby creating a direct path between the drywell and suppression chamber. Steam discharged through the downcomers could then bypass the suppression pool and directly pressurize the primary containment. This option is plausible because the chugging phenomenon is a real consideration in certain conditions and this is a common misconception among operators.  
A. This option is incorrect because chugging is not an issue with emergency depressurizing.
 
Chugging is an issue when drywell sprays are initiated when suppression chamber pressure exceeds the Suppression Chamber Spray Initiation Pressure (SCSIP) to preclude chuggingthe cyclic condensation of steam at the downcomer openings of the drywell vents.
C. This option is incorrect because excessive torus to drywell vacuum breaker operation is not occurring so valve failure is not an issue. This option is plausible because all the energy added to the suppression pool causes suppression pool level and air space pressure to rise. Vacuum breaker operation may occur but not excessively. If the candidate believes adding the energy to the suppression pool at the given conditions may affect vacuum breaker integrity would select this option.
When a steam bubble collapses at the exit of the downcomers, the rush of water drawn into the downcomers to fill the void induces stresses at the junction of the downcomers and the vent header. Repeated application of such stresses could cause fatigue failure of these joints, thereby creating a direct path between the drywell and suppression chamber. Steam discharged through the downcomers could then bypass the suppression pool and directly pressurize the primary containment. This option is plausible because the chugging phenomenon is a real consideration in certain conditions and this is a common misconception among operators.
C. This option is incorrect because excessive torus to drywell vacuum breaker operation is not occurring so valve failure is not an issue. This option is plausible because all the energy added to the suppression pool causes suppression pool level and air space pressure to rise.
Vacuum breaker operation may occur but not excessively. If the candidate believes adding the energy to the suppression pool at the given conditions may affect vacuum breaker integrity would select this option.
D. This option is incorrect because the primary containment water level of 16 ft is defined as the SRV Tail Pipe Level Limit (SRVTPLL). It is the highest primary containment water level at which opening of an SRV will not result in exceeding the capability of the SRV tail pipe, tail pipe supports, T quencher, or T quencher supports. By maintaining primary containment water level below this limit, SRV system damage and containment failure may be precluded.
D. This option is incorrect because the primary containment water level of 16 ft is defined as the SRV Tail Pipe Level Limit (SRVTPLL). It is the highest primary containment water level at which opening of an SRV will not result in exceeding the capability of the SRV tail pipe, tail pipe supports, T quencher, or T quencher supports. By maintaining primary containment water level below this limit, SRV system damage and containment failure may be precluded.
This option is plausible because had the torus level been higher than the conditions specified in the stem then this would have been the correct answer. So a candidate who does not recall the SRVTPLL may very well choose this option.
This option is plausible because had the torus level been higher than the conditions specified in the stem then this would have been the correct answer. So a candidate who does not recall the SRVTPLL may very well choose this option.
Technical Reference(s):
Technical Reference(s): INT0080613, OPS EOP FLOWCHART 3A - PRIMARY CONTAINMENT CONTROL, Rev 17 AMT-TB00 Appendix B (PSTG) for Step RC/P-2, Rev 8 Proposed references to be provided to applicants during examination: _HCTL Graph Learning Objective:
INT0080613, OPS EOP FLOWCHART 3A  
- PRIMARY CONTAINMENT CONTROL, Rev 17 AMT-TB00 Appendix B (PSTG) for Step RC/P
-2, Rev 8 Proposed references to be provided to applicants during examination: _HCTL Graph Learning Objective:
: 4. State the basis for primary containment control actions as they apply to the following.
: 4. State the basis for primary containment control actions as they apply to the following.
56  
56
: a. Specific setpoints
: a. Specific setpoints
: b. Primary Containment Control Systems
: b. Primary Containment Control Systems
: c. Graphs referenced on Flowchart 3A Question Source:
: c. Graphs referenced on Flowchart 3A Question Source:           Bank #             _21460______
Bank #   _21460______
Modified Bank #   _______
Modified Bank #
New               _______
_______  New   _______ Question History:
Question History:         Last NRC Exam     ____ ____
Last NRC Exam
Question Cognitive Level: Memory or Fundamental Knowledge _____
____ ____ Question Cognitive Level:
Comprehension or Analysis       __X__
Memory or Fundamental Knowledge
10 CFR Part 55 Content:   55.41 (8)
_____ Comprehension or Analysis   __X__ 10 CFR Part 55 Content:
Comments:
55.41 (8)
LOD: 4 57
Comments: LOD: 4   57 58 59 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
 
-
58 59 ES-401                           Sample Written Examination                     Form ES-401-5 Question Worksheet Examination Outline Cross-


==Reference:==
==Reference:==
Level                RO Tier #              __1__
Group #              __1__
K/A #                _295025.EK2.07___
Importance Rating    _3.7__
295025 High Reactor Pressure- Knowledge of the interrelations between high reactor pressure and the following:
EK2.07 RCIC: Plant-Specific Question: 12 The plant is operating at power and HPCI is inoperable due to maintenance on its Auxiliary Lube Oil Pump. An inadvertent PCIS Group 1 isolation occurs. RPV pressure rises until Low-Low Set arms.
What system is capable of maintaining adequate core cooling IAW EOPs as LLS actuates?
A. SLC B. RCIC C. MC/RF D. Core Spray Answer:
B. RCIC Explanation:
The high pressure injection systems per EOP 1A, Table 3 are MC/RF, High Pressure Coolant Injection (HPCI), Reactor Core Isolation Cooling (RCIC), and the smaller Control Rod Drive hydraulic (CRD) system. The MSIVs closing cause a reactor scram and reactor pressure spike above the reactor scram setpoint of 1045 psig because the reactor energy is now fully contained within the RPV. The Low-Low Set (LLS) valves actuate and control RPV pressure between 835 psig and 1040 psig. The RFPs are steam driven and lose their motive force when the MSIVs close so they are not available for injection. The Main Condensate (MC) system utilizes booster pumps that can inject below 550 psig RPV pressure. The RCIC system is the only available injection system for the conditions given in this question.
Distracters:
A. This option is incorrect because SLC is not a listed injection system per Table 3. This selection is plausible because SLC is a viable high pressure system per Table 4 of EOP 1A.
Table 4 systems are not utilized unless Table 3 systems are unavailable or ineffective and RPV water level has dropped below the ADS timer actuation setpoint.
60


Level  RO  Tier #    __1__  Group #  __1__  K/A #  _295025.EK2.07___
C. This option is incorrect because the RFPs are not available for injection due to MSIV closure.
Importance Rating
The RFPs are the normal injection system at this reactor pressure so this is a plausible answer. The candidate must realize the Group 1 closes the MSIVs and this causes the RFP turbine motive force to go away after the residual steam decays from the steam supply piping. In order to utilize main condensate, reactor pressure will have to be reduced well below the setpoint of the Low-Low Set value. If this is not realized, then this option would be chosen.
_3.7__    295025 High Reactor Pressure
D. This option is incorrect because Core Spray is a lower pressure injection system listed on Table 3. The Core Spray shutoff head value is ~350 psig so RPV pressure would have to be lowered well below the Low-Low Set value to begin injection. This is plausible because Core Spray is a listed injection system per EOP 1A, Table 3, but will not provide flow at the present RPV pressure.
- Knowledge of the interrelations between high reactor pressure and the following:
Technical Reference(s):         EOP 1A (RPV Control), Rev. 18 Vendor Manual 1869 COR0021802 OPS Reactor Core Isolation Cooling, Rev. 25 5.9SAMG Attachment 2 Plant Condition Assessment 1 Proposed references to be provided to applicants during examination:         ___NONE_______
EK2.07 RCIC: Plant
Learning Objective: COR0021802 OPS Reactor Core Isolation Cooling
-Specific  Q ues ti on: 1 2  The plant is operating at power and HPCI is inoperable due to maintenance on its Auxiliary Lube Oil Pump. An inadvertent PCIS Group 1 isolation occurs. RPV pressure rises until Low
-Low Set arms. What system is capable of maintaining adequate core cooling IAW EOPs as LLS actuates?
A. SLC  B. RCI C  C. MC/RF D. Core Spray A ns w e r: B. RCIC  Explanation:
The high pressure injection systems per EOP 1A, Table 3 are MC/RF, High Pressure Coolant Injection (HPCI), Reactor Core Isolation Cooling (RCIC), and the smaller Control Rod Drive hydraulic (CRD) system. The MSIVs closing cause a reactor scram and reactor pressure spike above the reactor scram setpoint of 1045 psig because the reactor energy is now fully contained within the RPV. The Low
-Low Set (LLS) valves actuate and control RPV pressure between 835 psig and 1040 psig. The RFPs are steam driven and lose their motive force when the MSIVs close so they are not available for injection. The Main Condensate (MC) system utilizes booster pumps that can inject below 550 psig RPV pressure. The RCIC system is the only available injection system for the conditions given in this question.
 
Distracters:
A. This option is incorrect because SLC is not a listed injection system per Table 3. This selection is plausible because SLC is a viable high pressure system per Table 4 of EOP 1A. Table 4 systems are not utilized unless Table 3 systems are unavailable or ineffective and RPV water level has dropped below the ADS timer actuation setpoint.
60 C. This option is incorrect because the RFPs are not available for injection due to MSIV closure. The RFPs are the normal injection system at this reactor pressure so this is a plausible answer. The candidate must realize the Group 1 closes the MSIVs and this causes the RFP turbine motive force to go away after the residual steam decays from the steam supply piping. In order to utilize main condensate, reactor pressure will have to be reduced well below the setpoint of the Low
-Low Set value. If this is not realized, then this option would be chosen. D. This option is incorrect because Core Spray is a lower pressure injection system listed on Table 3. The Core Spray shutoff head value is ~350 psig so RPV pressure would have to be lowered well below the Low
-Low Set value to begin injection. This is plausible because Core Spray is a listed injection system per EOP 1A, Table 3, but will not provide flow at the present RPV pressure.
Technical Reference(s):
EOP 1A (RPV Control), Rev. 18 Vendor Manual 1869 COR0021802 OPS Reactor Core Isolation Cooling, Rev. 25 5.9SAMG Attachment 2 Plant Condition Assessment 1 Proposed references to be provided to applicants during examination:
___NONE_______
Learning Objective:
COR0021802 OPS Reactor Core Isolation Cooling
: 10. Predict the consequences of the following on the RCIC system:
: 10. Predict the consequences of the following on the RCIC system:
: a. High/Low Reactor Pressure
: a.     High/Low Reactor Pressure Question Source:                Bank #
Modified Bank #
New                  X Question History:              Last NRC Exam        ____________
Question Cognitive Level:      Memory or Fundamental Knowledge              _____
Comprehension or Analysis                    __X__
10 CFR Part 55 Content:        55.41 (7)
Comments:
LOD 3 61


Question Source:
62 63 ES-401                             Sample Written Examination                     Form ES-401-5 Question Worksheet Examination Outline Cross-
Bank #    Modified Bank #
New  X  Question History:
Last NRC Exam
____________
 
Question Cognitive Level:
Memory or Fundamental Knowledge
_____ Comprehension or Analysis  __X__  10 CFR Part 55 Content:
55.41 (7)
Comments: LOD 3  61 62 63 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
-


==Reference:==
==Reference:==
 
Level                 RO Tier #                 __1__
Level   RO   Tier #     __1__   Group # __1__   K/A #   _295026EK3.04___
Group #               __1__
Importance Rating
K/A #                 _295026EK3.04___
_3.6__   295026 Suppression Pool High Water Temp
Importance Rating     _3.6__
- Knowledge of the reasons for the following responses as they apply to suppression pool high water temperature:
295026 Suppression Pool High Water Temp- Knowledge of the reasons for the following responses as they apply to suppression pool high water temperature:
EK 3.0 4 SBLC injection Q ues ti on: 1 3  The plant is experiencing an ATWS with the following conditions:
EK3.04 SBLC injection Question: 13 The plant is experiencing an ATWS with the following conditions:
Reactor Power is 15% and steady.
* Reactor Power is 15% and steady.
Suppression Pool temperature is 96&deg; F and rising. Why is boron injection required to be initiated before average suppression pool water temperature reaches 140&deg; F IAW EOP 6A {Reactor Power (Failure to Scram)}?
* Suppression Pool temperature is 96&deg; F and rising.
A. Prevents exceeding the 25% peak
Why is boron injection required to be initiated before average suppression pool water temperature reaches 140&deg; F IAW EOP 6A {Reactor Power (Failure to Scram)}?
-to-peak periodic neutron flux oscillations.
A. Prevents exceeding the 25% peak-to-peak periodic neutron flux oscillations.
B. Prevents violating Technical Specification Limit for Suppression Pool Temperature.
B. Prevents violating Technical Specification Limit for Suppression Pool Temperature.
C. Ensures the reactor will be shutdown under all conditions before the suppression pool is heated beyond its design limits.
C. Ensures the reactor will be shutdown under all conditions before the suppression pool is heated beyond its design limits.
 
D. Ensures the reactor will be shutdown under hot-standby conditions before the suppression pool reaches the Heat Capacity Temperature Limit.
D. Ensures the reactor will be shutdown under hot
Answer:
-standby conditions before the suppression pool reaches the Heat Capacity Temperature Limit.
D. Ensures the reactor will be shutdown under hot-standby conditions before the suppression pool reaches the Heat Capacity Temperature Limit.
Answer: D. Ensures the reactor will be shutdown under hot
Explanation:
-standby conditions before the suppression pool reaches the Heat Capacity Temperature Limit.
Explanation:
 
The BIIT for 15% power is 140&deg; F. Boron injection before this temperature ensures the Reactor will be shut down before the HCTL is exceeded.
The BIIT for 15% power is 140&deg; F. Boron injection before this temperature ensures the Reactor will be shut down before the HCTL is exceeded.
 
The Boron Injection Initiation Temperature (BIIT) is the greater of:
The Boron Injection Initiation Temperature (BIIT) is the greater of
* The highest suppression pool temperature at which initiation of boron injection will permit injection of the Hot Shutdown Boron Weight of boron before suppression pool temperature exceeds the Heat Capacity Temperature Limit.
:  The highest suppression pool temperature at which initiation of boron injection will permit injection of the Hot Shutdown Boron Weight of boron before suppression pool temperature exceeds the Heat Capacity Temperature Limit.
64
64 The suppression pool temperature at which a reactor scram is required by Technical Specifications.
* The suppression pool temperature at which a reactor scram is required by Technical Specifications.
The BIIT is a function of reactor power. If boron injection is initiated before suppression pool temperature reaches the BIIT, emergency RPV depressurization may be precluded at lower reactor power levels. At higher reactor power levels, however, the suppression pool heatup rate may become so high that the Hot Shutdown Boron Weight of boron cannot be injected before suppression pool temperature reaches the Heat Capacity Temperature Limit even if boron injection is initiated early in the event. Refer to Section 16 of this appendix for a detailed discussion of the BIIT. Distracters:
The BIIT is a function of reactor power. If boron injection is initiated before suppression pool temperature reaches the BIIT, emergency RPV depressurization may be precluded at lower reactor power levels. At higher reactor power levels, however, the suppression pool heatup rate may become so high that the Hot Shutdown Boron Weight of boron cannot be injected before suppression pool temperature reaches the Heat Capacity Temperature Limit even if boron injection is initiated early in the event. Refer to Section 16 of this appendix for a detailed discussion of the BIIT.
Distracters:
A. This option is incorrect because boron is injected after large neutron oscillations are observed not before. This option is plausible because boron injection is directly tied to large neutron oscillations and boron is injected to minimize them. The candidate who does not know the boron injection is performed in response to power oscillations would select this option.
A. This option is incorrect because boron is injected after large neutron oscillations are observed not before. This option is plausible because boron injection is directly tied to large neutron oscillations and boron is injected to minimize them. The candidate who does not know the boron injection is performed in response to power oscillations would select this option.
B. This option is incorrect because the Technical Specification limit for suppression pool temperature is already violated. This option is plausible because there is a Technical Specification limit on pool temperature. The candidate who recalls the Technical Specification limit and not the BIIT limit would select this option.
B. This option is incorrect because the Technical Specification limit for suppression pool temperature is already violated. This option is plausible because there is a Technical Specification limit on pool temperature. The candidate who recalls the Technical Specification limit and not the BIIT limit would select this option.
C. This option is incorrect because the suppression pool design temperature limit is not tied to the start of boron injection bases upon pool temperature. Getting the reactor shutdown before HCTL is reached ensures the pool can absorb the energy released to the pool during emergency depressurization. This option is plausible because shutting the reactor down will lessen the energy being transferred to the suppression pool. The candidate who cannot recall the reasons for the BIIT curve would select this option.
Technical Reference(s):          AMP-TB00 (CNS PSTGs) Rev. 8 Appendix B Proposed references to be provided to applicants during examination:          NONE Learning Objective:      INT008-06-06. OPS EOP Flowchart 6A - RPV Pressure & Power (Failure-to-Scram)
: 3.        Describe the conditions that require boron injection, and when boron injection can be secured.
INT008-06-18, OPS EOP and SAG Graphs and Cautions
: 1.        Using the graphs provided in the EOP and SAG Graphs Flowchart, explain how the shape of each curve or family of curves was determined.
Question Source:                  Bank #
Modified Bank #        5334 New Question History:                Last NRC Exam          ____________
Question Cognitive Level:        Memory or Fundamental Knowledge              X Comprehension or Analysis 65


C. This option is incorrect because the suppression pool design temperature limit is not tied to the start of boron injection bases upon pool temperature. Getting the reactor shutdown before HCTL is reached ensures the pool can absorb the energy released to the pool during emergency depressurization. This option is plausible because shutting the reactor down will lessen the energy being transferred to the suppression pool. The candidate who cannot recall the reasons for the BIIT curve would select this option.
10 CFR Part 55 Content: 55.41 (5)
Comments:
LOD 3 66


Technical Reference(s):
67 68 ES-401                           Sample Written Examination                     Form ES-401-5 Question Worksheet Examination Outline Cross-
AMP-TB00 (CNS PSTGs) Rev. 8 Appendix B Proposed references to be provided to applicants during examination:
NONE  Learning Objective:
INT008-06-06. OPS EOP Flowchart 6A
- RPV Pressure & Power (Failure-to-Scram)  3. Describe the conditions that require boron injection, and when boron injection can be secured.
INT008-06-18, OPS EOP and SAG Graphs and Cautions
: 1. Using the graphs provided in the EOP and SAG Graphs Flowchart, explain how the shape of each curve or family of curves was determined.
Question Source:
Bank #    Modified Bank #
5334 New    Question History:
Last NRC Exam
____________
Question Cognitive Level:
Memory or Fundamental Knowledge X Comprehension or Analysis    65 10 CFR Part 55 Content:
55.41 (5)  Comments: LOD 3    66 67 68 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
-


==Reference:==
==Reference:==
 
Level               RO Tier #               __1__
Level   RO   Tier #     __1__   Group # __1__   K/A #   _295028EA1.01___
Group #             __1__
Importance Rating
K/A #               _295028EA1.01___
_3.8__   295028 High Drywell Temperature
Importance Rating   _3.8__
- Ability to operate and/or monitor the following as they apply to high drywell temperature:
295028 High Drywell Temperature- Ability to operate and/or monitor the following as they apply to high drywell temperature:
EA1.01 Drywell spray: Mark
EA1.01 Drywell spray: Mark-I&II Question: 14 Drywell spray has been placed in service during a LOCA due to high Drywell temperature.
-I&II Q ues ti on: 1 4  Drywell spray has been placed in service during a LOCA due to high Drywell temperature.
* Drywell temperature and pressure are lowering.
Drywell temperature and pressure are lowering.
When is Drywell spray required to be stopped IAW EOP 3A (Primary Containment Control)?
When is Drywell spray required to be stopped IAW EOP 3A (Primary Containment Control)?
Drywell spray is REQUIRED to be stopped before-A. drywell pressure lowers to zero psig.
Drywell spray is REQUIRED to be stopped before A. drywell pressure lowers to zero psig.
B. the suppression chamber to drywell vacuum breakers open.
B. the suppression chamber to drywell vacuum breakers open.
C. the reactor building to suppression chamber vacuum breakers open.
C. the reactor building to suppression chamber vacuum breakers open.
D. drywell temperature and pressure lower to the UNSAFE (Red) region of the Drywell Spray Initiation Limit (DWISL) curve.
D. drywell temperature and pressure lower to the UNSAFE (Red) region of the Drywell Spray Initiation Limit (DWISL) curve.
A ns w e r: A. drywell pressure lowers to zero psig.
Answer:
A. drywell pressure lowers to zero psig.
Explanation:
Explanation:
EOP-3A override DS
EOP-3A override DS-1 directs if drywell sprays have been started, then before drywell pressure drops to 0 psig ensure drywell sprays are stopped to ensure that primary containment pressure is not reduced below atmospheric. Drywell pressure can be reduced below suppression chamber pressure causing the suppression chamber to DW vacuum breakers to open (non-condensables return to DW). DW pressure would have to be significantly negative to cause the suppression chamber pressure to lower sufficiently to cause the reactor building to suppression chamber vacuum breakers to open due to the reactor building being maintained at a negative pressure.
-1 directs if drywell sprays have been started, then before drywell pressure drops to 0 psig ensure drywell sprays are stopped to ensure that primary containment pressure is not reduced below atmospheric. Drywell pressure can be reduced below suppression chamber pressure causing the suppression chamber to DW vacuum breakers to open (non
69
-condensables return to DW). DW pressure would have to be significantly negative to cause the suppression chamber pressure to lower sufficiently to cause the reactor building to suppression chamber vacuum breakers to open due to the reactor building being maintained at a negative pressure. 69 Distracters:
 
Distracters:
B. This answer is incorrect because Drywell spray is not required to be stopped before the suppression chamber to drywell vacuum breakers open. This choice is plausible due the common misconception that operation of the suppression chamber to drywell vacuum breakers is abnormal and not desired. Candidates that have the misconception of suppression chamber to drywell vacuum breaker operation would select this answer.
B. This answer is incorrect because Drywell spray is not required to be stopped before the suppression chamber to drywell vacuum breakers open. This choice is plausible due the common misconception that operation of the suppression chamber to drywell vacuum breakers is abnormal and not desired. Candidates that have the misconception of suppression chamber to drywell vacuum breaker operation would select this answer.
C. This answer is incorrect because Drywell spray is not required to be stopped before the reactor building to suppression chamber vacuum breakers open. This choice is plausible due the operation of the reactor building to suppression chamber vacuum breakers introducing air into the torus being abnormal and not desired. Candidates that have the misconception of the reactor building to suppression chamber vacuum breaker operation and securing drywell sprays would select this answer.
C. This answer is incorrect because Drywell spray is not required to be stopped before the reactor building to suppression chamber vacuum breakers open. This choice is plausible due the operation of the reactor building to suppression chamber vacuum breakers introducing air into the torus being abnormal and not desired. Candidates that have the misconception of the reactor building to suppression chamber vacuum breaker operation and securing drywell sprays would select this answer.
D. This answer is incorrect because Drywell spray is not required to be stopped before the drywell temperature and pressure lower to the RED region of the Drywell Spray Initiation Limit (DWISL) curve. This choice is a   common misconception because drywell spray cannot be initiated in the unsafe region of the curve but if spray is already in progress it may continue. Candidates that have the misconception of always being in the SAFE region of DWSIL would select this answer
D. This answer is incorrect because Drywell spray is not required to be stopped before the drywell temperature and pressure lower to the RED region of the Drywell Spray Initiation Limit (DWISL) curve. This choice is a common misconception because drywell spray cannot be initiated in the unsafe region of the curve but if spray is already in progress it may continue. Candidates that have the misconception of always being in the SAFE region of DWSIL would select this answer Technical Reference(s):         EOP-3A Primary Containment Control, Rev. 15 AMP-TBD00 (EOP/PSTG Technical Basis), Rev. 08 Proposed references to be provided to applicants during examination:           None Learning Objective:             INT00806130010300 OPS EOP FLOWCHART 3A - PRIMARY CONTAINMENT CONTROL Explain why torus and drywell sprays must be secured before torus/drywell pressure lowers to 0 psig.
 
Question Source:                Bank #                _______
Technical Reference(s):
Modified Bank #          ___ (See attached)
EOP-3A Primary Containment Control, Rev. 15 AMP-TBD00 (EOP/PSTG Technical Basis), Rev. 08 Proposed references to be provided to applicants during examination:
New                  ___X__
None           Learning Objective:
Question History:              Last NRC Exam        ____________
INT00806130010300 OPS EOP FLOWCHART 3A  
Question Cognitive Level:      Memory or Fundamental Knowledge                X Comprehension or Analysis 10 CFR Part 55 Content:        55.41 (7)
- PRIMARY CONTAINMENT CONTROL Explain why torus and drywell sprays must be secured before torus/drywell pressure lowers to 0 psig.
Comments:
LOD 3 70


Question Source:
71 72 73 ES-401                           Sample Written Examination                       Form ES-401-5 Question Worksheet Examination Outline Cross-
Bank #  _______ Modified Bank #
___ (See attached)
New  ___X__
Question History:
Last NRC Exam
____________
Question Cognitive Level:
Memory or Fundamental Knowledge X Comprehension or Analysis    10 CFR Part 55 Content:
55.41 (7)  Comments: LOD 3  70 71 72 73 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
-


==Reference:==
==Reference:==
Level                RO Tier #                __1__
Group #              __1__
K/A #                _295030EA2.02___
Importance Rating    _3.8__
295030 Low Suppression Pool Water Level- Ability to determine and/or interpret the following as they apply to low suppression pool water level:
EA2.02 Suppression pool temperature Question: 15 Following a LOCA, the following conditions are present:
* RPV water level above TAF and slowly rising
* Torus pressure is 4.5 psig (stable).
* Primary containment water level is 5.5 feet (stable).
* Only RHR pump C is injecting at the NPSH and Vortex limit of 6500 gpm.
What is the CURRENT Torus average water temperature based upon the above conditions?
A. 140&deg;F B. 175&deg;F C. 185&deg;F D. 2 1 0 &deg;F Answer:
C. 185&deg;F Explanation:
Requires determining Suppression Pool temperature which limits RHR flow to 6500 gpm with low SP water level. With 4.5 psig pressure and 5.5 feet of water level in the torus, there is 5.14 psig overpressure in the torus (see attached calculation). The NPSH and vortex limit for RHR is 6500 gpm. With RPV level rising and above the top of active fuel, the CRS is correct in limiting flow to the NPSH and vortex limits. When using EOP graphs, Procedure 5.8 (EMERGENCY OPERATING PROCEDURES), Step 3.11 allows interpolation of graphs if SPDS is unavailable.
SPDS is unavailable because the curves and graphs are given. Step 3.11 also states interpolation is not required even though it is preferred. The question is asking for the Current Torus average water temperature so interpolation is required in this question.
74


Level  RO  Tier #    __1__  Group #  __1__  K/A #  _295030EA2.02___
Distracters:
Importance Rating
_3.8__    295030 Low Suppression Pool Water Level
- Ability to determine and/or interpret the following as they apply to low suppression pool water level:
EA2.02 Suppression pool temperature Q ues ti on: 1 5  Fo ll o w i n g a LOC A, t h e f o ll o w i n g conditi on s a r e p resen t:  RPV water level above TAF and slowly rising To r u s p ressu r e i s 4.5 ps i g (s t ab l e). P ri m a r y con t a i n m en t w a t e r leve l i s 5.5 f ee t (s t ab l e). Only RH R pu m p C is injecting at the NPS H an d V o rt e x li m it of 6500 gpm
.
W ha t i s t h e CURRENT To r u s ave r ag e w a t e r t e m pe r a t u re based upon the above conditions
?  A. 140&deg;F B. 175&deg;F C. 185&deg;F D. 210&deg;F A ns w e r: C. 185&deg;F Explanation:
Requires determining Suppression Pool temperature which limits RHR flow to 6500 gpm with low SP water level. With 4.5 psig pressure and 5.5 feet of water level in the torus, there is 5.14 psig overpressure in the torus (see attached calculation). The NPSH and vortex limit for RHR is 6500 gpm. With RPV level rising and above the top of active fuel, the CRS is correct in limiting flow to the NPSH and vortex limits. When using EOP graphs, Procedure 5.8 (EMERGENCY OPERATING PROCEDURES), Step 3.11 allows interpolation of graphs if SPDS is unavailable. SPDS is unavailable because the curves and graphs are given. Step 3.11 also states interpolation is not required even though it is preferred. The question is asking for the Current T o r u s ave r ag e w a t e r t e m pe r a t u r e so interpolation is required in this question.
74 Distracters:
A. This option is incorrect because the listed temperature is too low. This option is plausible if torus overpressure is miscalculated for 0 psig overpressure. The candidate who incorrectly calculates torus overpressure would select this answer.
A. This option is incorrect because the listed temperature is too low. This option is plausible if torus overpressure is miscalculated for 0 psig overpressure. The candidate who incorrectly calculates torus overpressure would select this answer.
B. This option is incorrect because the listed temperature is too low. This option is plausible if torus overpressure is miscalculated and interpolated for 3.9 psig overpressure. The candidate that incorrectly calculates torus overpressure would select this answer.
B. This option is incorrect because the listed temperature is too low. This option is plausible if torus overpressure is miscalculated and interpolated for 3.9 psig overpressure. The candidate that incorrectly calculates torus overpressure would select this answer.
D. This option is incorrect because the listed temperature is too high. This option is plausible if torus overpressure is miscalculated for 10 psig overpressure. The candidate that incorrectly calculates torus overpressure would select this answer.
D. This option is incorrect because the listed temperature is too high. This option is plausible if torus overpressure is miscalculated for 10 psig overpressure. The candidate that incorrectly calculates torus overpressure would select this answer.
Technical Reference(s):
Technical Reference(s):       CNS PSTG AMP TB00, Section 16, Rev. 8 Proposed references to be provided to applicants during examination:       _EOP Graphs 4 and 5, EOP Note 3_
CNS PSTG AMP TB00, Section 16, Rev. 8 Proposed references to be provided to applicants during examination:
_EOP Graphs 4 and 5, EOP Note 3_
Learning Objective:
Learning Objective:
INT0080613 OPS EOP FLOWCHART 3A  
INT0080613 OPS EOP FLOWCHART 3A - PRIMARY CONTAINMENT CONTROL
- PRIMARY CONTAINMENT CONTROL
: 4.     State the basis for primary containment control actions as they apply to the following.
: 4. State the basis for primary containment control actions as they apply to the following.
: c.     Graphs referenced on Flowchart 3A Question Source:               Bank #               _______
: c. Graphs referenced on Flowchart 3A Question Source:
Modified Bank #       _______
Bank #   _______
New                   ____X___
Modified Bank #
Question History:             Last NRC Exam         _________
_______  New   ____X___ Question History:
Question Cognitive Level:     Memory or Fundamental Knowledge             _____
Last NRC Exam
Comprehension or Analysis                   __X__
_________
10 CFR Part 55 Content:       55.41 (10)
Question Cognitive Level:
Comments:
Memory or Fundamental Knowledge
LOD: 3 75
_____ Comprehension or Analysis   __X__
 
10 CFR Part 55 Content:
76 77 ES-401                           Sample Written Examination                     Form ES-401-5 Question Worksheet Examination Outline Cross-
55.41 (10)
Comments: LOD: 3   75 76 77 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
-


==Reference:==
==Reference:==
Level                  RO Tier #                __1__
Group #                __1__
K/A #                  _295031 2.4.2___
Importance Rating      _3.8__
295031 Reactor Low Water Level:
2.4.2 Knowledge of system set points, interlocks and automatic actions associated with EOP entry conditions.
Question: 16 RPV water level lowers to the value requiring entry into EOP-1A (RPV Control) and remains steady at that level.
No other EOP entry conditions are satisfied.
What group isolation(s) automatically occur(s)?
A. Group 2 only.
B. Group 3 only.
C. Groups 3 and 6 only.
D. Groups 2, 3, and 6.
Answer:
A. Group 2 only.
Explanation:
Instrument zero RPV level at CNS is +165 inches above the top of active fuel. Normal RPV water level at CNS is +35 inches above instrument zero. EOP-1A entry condition on low RPV level is +3 inches above instrument zero. The PCIS Group 2 isolation actuates at
+3 inches RPV level. The PCIS Group 3 and 6 isolations both actuate at -42 inches wide range which is 45 inches RPV level below the Group 2 actuation point. Providing an actual RPV water level in the stem would reduce the difficulty to a point in which the question would no longer be discriminatory therefore the water level is intentionally omitted to comply with the NUREG requirement to avoid use of questions with a difficulty of 1.
Distracters:
B. This option is incorrect because a Group 3 isolation does not occur until RPV level is at
    -42 inches. The candidate who does not recall this setpoint would select this option.
This option is plausible because Group 3 does occur on a low RPV level condition.
78


Level  RO  Tier #    __1__  Group #  __1__  K/A #  _295031 2.4.2___
C. This option is incorrect because the Group 3 & 6 isolations do not occur until the -42 inches level is reached. The candidate who does not recall this setpoint would select this option. This option is plausible because the Groups 3 & 6 isolation do occur on a low RPV level.
Importance Rating
D. This option is incorrect because the Group 3 and 6 isolations do not occur until -42 inches RPV level. Group 2 isolation does actually occur at this RPV level. The candidate who does not recall these Group isolations actuation points would select this option. This option is plausible because the group 2, 3 and 6 isolations occur on a low RPV level condition.
_3.8__    295031 Reactor Low Water Level:
Technical Reference(s):        2.1.22 Recovering From A Group Isolation, Rev. 59 Proposed references to be provided to applicants during examination:      ___NONE_______
Learning Objective:    COR0020302 OPS CONTAINMENT
: 21.           Given plant conditions, determine if the following should have occurred:
: b. Any of the PCIS group isolations Question Source:              Bank #                _19694_
Modified Bank #      _______
New                  _______
Question History:              Last NRC Exam        ___ ___
Question Cognitive Level:     Memory or Fundamental Knowledge            ____
Comprehension or Analysis                  __ X__
10 CFR Part 55 Content:        55.41 (7)
Comments:
LOD 2 EOP 1A Entry Conditions 79


====2.4.2 Knowledge====
Group 2 Isolations Group 3 Isolations 80
of system set points, interlocks and automatic actions associated with EOP entry conditions.
Q ues ti on: 1 6  RP V w a t e r leve l l o w e r s t o t h e va l u e requiring en tr y into EO P-1A (RP V Contro l) and r e m a i n s s t ead y a t t ha t leve l. N o o t he r EO P en tr y conditi on s a r e sati sfi ed. W ha t g r ou p i so l a ti o n(s) au t o m a ti ca ll y occu r(s)?  A. G r ou p 2 on l y. B. G r ou p 3 on l y. C. G roup s 3 an d 6 on l y. D. G roup s 2 , 3 , an d 6. A ns w e r: A. G r ou p 2 on l y. E xp l anati on: I ns tr u m en t ze r o RP V leve l a t CN S i s +16 5 inche s abov e t h e t o p o f ac ti v e f ue l. N o r m a l RP V w a t e r leve l a t CN S i s +3 5 inche s abov e inst r u m en t ze r o. EOP-1 A en tr y conditi o n o n l o w RP V leve l i s +3 inche s abov e i ns tr u m en t ze r o. Th e PC I S G r ou p 2 i so l a ti o n ac t ua t e s a t +3 inche s RP V leve l. Th e PC I S G r ou p 3 an d 6 i so l ati on s bo t h ac t ua t e a t -4 2 inche s w i d e rang e w h i c h i s 4 5 inche s RP V leve l be l o w t h e G r ou p 2 ac t ua ti o n po i n t. Providing an actual RPV water level in the stem would reduce the difficulty to a point in which the question would no longer be discriminatory therefore the water level is intentionally omitted to comply with the NUREG requirement to avoid use of questions with a difficulty of 1.
D i s tr ac t e r s: B. This option is incorrect because a Group 3 isolation does not occur until RPV level is at
-42 inches. The candidate who does not recall this setpoint would select this option. This option is plausible because Group 3 does occur on a low RPV level conditio n. 78 C. Th i s option i s inco rr ec t becaus e t h e G r ou p 3 & 6 i so l a ti on s d o no t occu r un ti l t h e -4 2 inches leve l i s reached. Th e cand i da t e w h o doe s no t reca l l t h i s se t po i n t w ou l d se l ec t t h i s option. Th i s option i s p l aus i b l e becaus e t h e G r oup s 3 & 6 i so l a ti o n d o occu r o n a l o w RP V leve l. D. Th i s option i s inco rr ec t becaus e t h e G r ou p 3 an d 6 i so l a ti on s d o no t occu r un ti l -4 2 inches RP V leve l. Group 2 isolation does actually occur at this RPV level. Th e cand i da t e w h o doe s no t reca l l t hes e G r ou p i so l a ti on s ac t ua ti o n po i n t s w o u l d se l ec t t h i s option. Th i s option i s p l aus i b l e becaus e t h e g r ou p 2 , 3 an d 6 i so l a ti on s occu r o n a l o w RP V leve l condition. Technical Reference(s):
2.1.22 Recovering From A Group Isolation, Rev. 59 Proposed references to be provided to applicants during examination:
___NONE_______


Learning Objective:
Group 6 Isolations 81
COR0020302 OPS CONTAINMENT
: 21. Given plant conditions, determine if the following should have occurred:
: b. Any of the PCIS group isolations


Question Source:
ES-401                             Sample Written Examination                     Form ES-401-5 Question Worksheet Examination Outline Cross-
Bank #  _19694_ Modified Bank #
_______  New  _______  Question History:
Last NRC Exam
___ ___
Question Cognitive Level:
Memory or Fundamental Knowledge
____ Comprehension or Analysis  __ X__
10 CFR Part 55 Content:
55.41 (7)
Comments:
LOD 2 EOP 1A Entry Conditions 79 Group 2 Isolations Group 3 Isolations 80 Group 6 Isolations 81 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
-


==Reference:==
==Reference:==
Level                  RO Tier #                __1__
Group #                __1__
K/A #                  _295037 EK1.05___
Importance Rating      _3.4__
295037 SCRAM Condition Present and Reactor Power above APRM Downscale or Unknown:
Knowledge of the operational implications of the following concepts as they apply to scram condition present and reactor power above APRM downscale or unknown:
(CFR: 41.8 to 41.10)
EK1.05 Cold shutdown boron weight: Plant-Specific Question: 17 The plant is operating at full power when the following occurs:
* All MSIVs close resulting in a Reactor Scram with 27 rods failing to fully insert.
* RPV water level is intentionally lowered due to level/power conditions being met.
* SLC is injecting Boron.
When is a normal reactor cooldown FIRST allowed to commence IAW EOP 6A {RPV Pressure (Failure to Scram)}?
A. When all APRMs indicate downscale.
B. When Hot Shutdown Boron Weight is injected.
C. When Cold Shutdown Boron Weight is injected.
D. When one control rod is at 48 and all other control rods are at position 02.
Answer:
C. When Cold Shutdown Boron Weight is injected.
Explanation:
When the reactor is not shutdown and lowering RPV level to lower reactor power is required, boron is being injected into the RPV. When hot shutdown boron weight (26% of the Standby Liquid Control tank level) has been injected into the RPV, the RPV level is raised via outside the shroud injection systems. RPV level is raised to bring more of the boron into fuel region to aid in keeping reactor power low. The normal RPV level band of +3 inches to +54 inches is directed at this point. Once cold shutdown boron weight is injected (60% of the Standby Liquid Control tank level) or it is determined the reactor will remain shutdown without relying on the 82


Level  RO  Tier #    __1__  Group #  __1__  K/A #  _295037 EK1.05___
boron concentration in the RPV, reactor pressure can be lowered so the plant can be placed in a cold shutdown condition.
Importance Rating
From training material INT0080606 OPS EOP Flowchart 6A - RPV Pressure & Power (Failure-to-Scram):
_3.4__    295037 SCRAM Condition Present and Reactor Power above APRM Downscale or Unknown:  Knowledge of the operational implications of the following concepts as they apply to scram condition present and reactor power above APRM downscale or unknown:
FS/Q Injection normally continues until the entire usable contents of the SLC tank have been injected. Actions in other EOP steps, however, are conditioned upon lesser amounts of boron:
(CFR: 41.8 to 41.10)
: a. When the Hot Shutdown Boron Weight has been injected, RPV water level may be restored above the low level scram setpoint. The second override in Flowchart 7A Step FS/L-11 becomes active.
EK1.05 Cold shutdown boron weight: Plant
: b. When the Cold Shutdown Boron Weight has been injected, RPV cooldown may be initiated in accordance with Step FS/P-6.
-Specific  Q ues ti on: 1 7  Th e p l an t i s ope r a ti n g a t f u l l p o w e r w he n t h e f o ll o w i n g occu r s:  All MS I Vs c lose resu lting i n a R eac t o r S c r a m w it h 2 7 r od s f a ili n g t o f u ll y inse rt. RP V w a t e r leve l i s i n t en ti ona ll y l o w e r e d du e t o leve l/po w e r conditi on s be i n g m e t. S L C i s i n j ec ti n g B o r on. When is a normal reactor cooldown FIRST allowed to commence IAW EOP 6A {RPV Pressure (Failure to Scram)}?
FS/P RPV depressurization and cooldown may not proceed until the reactor is shutdown with no boron injected or the amount of boron injected into the RPV is sufficient to keep the reactor shut down.
A. W he n a l l APR M s i nd i ca t e do wnsca l e.
B. W he n H o t S hu t do w n B o r o n W e i gh t i s i n j ec t ed. C. W he n C o l d S hu t do w n B o r o n W e i gh t i s i n j ec t ed. D. W he n one control r o d i s at 48 and all other control rods are at positi o n 0 2. A ns w e r: C. W he n C o l d S hu t do w n B o r o n W e i gh t i s i n j ec t ed. Explanation:
When the reactor is not shutdown and lowering RPV level to lower reactor power is required, boron is being injected into the RPV. When hot shutdown boron weight (26% of the Standby Liquid Control tank level) has been injected into the RPV, the RPV level is raised via outside the shroud injection systems. RPV level is raised to bring more of the boron into fuel region to aid in keeping reactor power low. The normal RPV level band of +3 inches to +54 inches is directed at this point. Once cold shutdown boron weight is injected (60% of the Standby Liquid Control tank level) or it is determined the reactor will remain shutdown without relying on the 82 boron concentration in the RPV, reactor pressure can be lowered so the plant can be placed in a cold shutdown condition.
 
From training material INT0080606 OPS EOP Flowchart 6A  
- RPV Pressure & Power (Failure
-to-Scram):
FS/Q Injection normally continues until the entire usable contents of the SLC tank have been injected. Actions in other EOP steps, however, are conditioned upon lesser amounts of boron: a. When the Hot Shutdown Boron Weight has been injected, RPV water level may be restored above the low level scram setpoint. The second override in Flowchart 7A Step FS/L-11 becomes active.  
: b. When the Cold Shutdown Boron Weight has been injected, RPV cooldown may be initiated in accordance with Step FS/P
-6. FS/P RPV depressurization and cooldown may not proceed until the reactor is shutdown with no boron injected or the amount of boron injected into the RPV is sufficient to keep the reactor shut down.
A volume of boron solution equivalent to 60% of the SLC tank (or 2258 lbs of boric acid and 2321 lbs of borax) is called the Cold Shutdown Boron Weight (CSBW). The CSBW is defined to be the amount of soluble boron which, if injected into the RPV and mixed uniformly, will maintain the reactor shutdown under all conditions.
A volume of boron solution equivalent to 60% of the SLC tank (or 2258 lbs of boric acid and 2321 lbs of borax) is called the Cold Shutdown Boron Weight (CSBW). The CSBW is defined to be the amount of soluble boron which, if injected into the RPV and mixed uniformly, will maintain the reactor shutdown under all conditions.
The CSBW is determined assuming:
The CSBW is determined assuming:
Line 700: Line 726:
: c. No Xenon in the core
: c. No Xenon in the core
: d. No voids in the core
: d. No voids in the core
: e. f. Shutdown cooling and RWCU are in service
: e. Water is at most reactive temperature (68F)
: f. Shutdown cooling and RWCU are in service
: g. RPV water level at 54 in.
: g. RPV water level at 54 in.
NOTE 1 of EOP 6A provides a method of calculating the SLC tank level resulting from injection of the CSBW. Weights of borax and boric acid are provided if boron injection must be performed by preparing the borate solution at the RWCU precoat tank.
NOTE 1 of EOP 6A provides a method of calculating the SLC tank level resulting from injection of the CSBW. Weights of borax and boric acid are provided if boron injection must be performed by preparing the borate solution at the RWCU precoat tank.
If any amount of boron less than the CSBW has been injected into the RPV, cooldown is not permitted unless it can be determined that control rod insertion alone assures the reactor will remain shutdown under all conditions. The core reactivity response from cooldown in a partially borated core is unpredictable and subsequent steps may not prescribe the correct actions for such conditions if criticality were to occur.
If any amount of boron less than the CSBW has been injected into the RPV, cooldown is not permitted unless it can be determined that control rod insertion alone assures the reactor will remain shutdown under all conditions. The core reactivity response from cooldown in a partially borated core is unpredictable and subsequent steps may not prescribe the correct actions for such conditions if criticality were to occur.
The reactor is shutdown under all conditions without boron if all control rods are inserted to or beyond position 02, the shutdown margin is met (theoretically strongest control rod withdrawn and all other control rods fully inserted), or engineering determination.
The reactor is shutdown under all conditions without boron if all control rods are inserted to or beyond position 02, the shutdown margin is met (theoretically strongest control rod withdrawn and all other control rods fully inserted), or engineering determination.
83 Distracters:
83


Distracters:
A. This option is incorrect because APRM downscales are not used to determine when the cooldown can commence. If the candidate does not understand that cold shutdown boron weight (CSBW) must be injected prior to cooling down the reactor, he/she would select this option. This option is plausible because APRM downscale indication is used to determine if EOPs are entered.
A. This option is incorrect because APRM downscales are not used to determine when the cooldown can commence. If the candidate does not understand that cold shutdown boron weight (CSBW) must be injected prior to cooling down the reactor, he/she would select this option. This option is plausible because APRM downscale indication is used to determine if EOPs are entered.
 
B. This option is incorrect because HSBW only allows water level to be restored to +3 inches to
B. This option is incorrect because HSBW only allows water level to be restored to +3 inches to +54 inches to promote boron mixing in the core. If the candidate does not understand that cold shutdown boron weight (CSBW) must be injected prior to cooling down the reactor, he/she would select this option. This option is plausible because HSBW does allow a major parameter change to be made during ATWS conditions.
    +54 inches to promote boron mixing in the core. If the candidate does not understand that cold shutdown boron weight (CSBW) must be injected prior to cooling down the reactor, he/she would select this option. This option is plausible because HSBW does allow a major parameter change to be made during ATWS conditions.
 
D. This option is incorrect because with 1 control rod not at 02 and all other control rods at 02, the reactor cannot be considered shut down under all conditions without boron. The RE would be required to evaluate this and inform the control room. If the candidate thinks he/she can make this call without RE evaluation they would select this option. This option is plausible because 1 control rod at position 48 and all other control rods at position 00 is the definition of shutdown under all conditions without boron. The reactor can be considered shutdown at this point and the cooldown could commence.
D. This option is incorrect because with 1 control rod not at 02 and all other control rods at 02, the reactor cannot be considered shut down under all conditions without boron. The RE would be required to evaluate this and inform the control room. If the candidate thinks he/she can make this call without RE evaluation they would select this option. This option is plausible because 1 control rod at position 48 and all other control rods at position 00 is the definition of shutdown under all conditions without boron. The reactor can be considered shutdown at this point and the cooldown could commence.
Technical Reference(s):
Technical Reference(s):       EOP-6A, Rev 16/EOP-7A, Rev 17-Failure to Scram EOPs Proposed references to be provided to applicants during examination:       ___NONE_______
EOP-6A, Rev 16/EOP
Learning Objective:            INT0080606 OPS EOP Flowchart 6A - RPV Pressure & Power (Failure-to-Scram)
-7A, Rev 17
: 12. Given an EOP flowchart 6A, RPV PRESSURE/POWER step, state the reason for the actions contained in the step.
-Failure to Scram EOPs Proposed references to be provided to applicants during examination:
Question Source:              Bank #                _      _
___NONE_______
Modified Bank #      _      _
New                  ___X___
Question History:              Last NRC Exam        ___ ___
Question Cognitive Level:      Memory or Fundamental Knowledge              X Comprehension or Analysis 10 CFR Part 55 Content:        55.41 (10)
Comments:
LOD 3 84


Learning Objective:
85 ES-401                           Sample Written Examination                   Form ES-401-5 Question Worksheet Examination Outline Cross-
INT0080606 OPS EOP Flowchart 6A
- RPV Pressure & Power (Failure-to-Scram)  12. Given an EOP flowchart 6A, RPV PRESSURE/POWER step, state the reason for the actions contained in the step.
Question Source:
Bank #  _          _
Modified Bank #
_
_  New  ___X___  Question History:
Last NRC Exam
___ ___  Question Cognitive Level:
Memory or Fundamental Knowledge X Comprehension or Analysis    10 CFR Part 55 Content:
55.41 (10)
Comments:  LOD 3  84 85 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
-


==Reference:==
==Reference:==
 
Level                 RO Tier #                 __1__
Level   RO   Tier #     __1__   Group # __1__   K/A #   _295038 EK2.02___
Group #               __1__
Importance Rating
K/A #                 _295038 EK2.02___
_3.6__   295038 High Off
Importance Rating     _3.6__
-site Release Rate:
295038 High Off-site Release Rate:
Knowledge of the interrelations between high off
Knowledge of the interrelations between high off-site release rate and the following:
-site release rate and the following:
(CFR: 41.7 / 45.8)
(CFR: 41.7 / 45.8)
EK2.02 Offgas system Question: 18 The plant is operating at rated power when the following annunciator is received:
EK2.02 Offgas system Q ues ti on: 1 8  Th e p l an t i s ope r a ti n g a t r a t e d po w e r w he n the following annunciator is received:
MAIN STM LINE               PANEL/WINDOW:
MAIN STM LINE HIGH RAD    PANEL/WINDOW:
HIGH RAD                    9-4-1/A-5 1 minute later the following annunciator is received:
9-4-1/A-5 1 minute later the following annunciator is received:
OFFGAS                 PANEL/WINDOW:
OFFGAS HIGH RAD    PANEL/WINDOW:
HIGH RAD                    9-4-1/C-5 What automatic action minimizes off-site release rates if the ERP, MSL and OG radiation monitors continue rising?
9-4-1/C-5 What automatic action minimizes off
-site release rates if the ERP, MSL and OG radiation monitors continue rising?
What procedure(s) is/are required to be entered due to these alarms?
What procedure(s) is/are required to be entered due to these alarms?
A. MSIV closure 5.2FUEL (Fuel Failure) ONLY.
A. MSIV closure 5.2FUEL (Fuel Failure) ONLY.
B. MSIV closure 2.4OG (Off
B. MSIV closure 2.4OG (Off-Gas Abnormal) and 5.2FUEL (Fuel Failure).
-Gas Abnormal) and 5.2FUEL (Fuel Failure).
C. Offgas isolation 5.2FUEL (Fuel Failure) ONLY.
C. Offgas isolation 5.2FUEL (Fuel Failure) ONLY.
D. Offgas isolation 2.4OG (Off
D. Offgas isolation 2.4OG (Off-Gas Abnormal) and 5.2FUEL (Fuel Failure).
-Gas Abnormal) and 5.2FUEL (Fuel Failure).
Answer:
A ns w e r: D. Offgas isolation 2.4OG (Off
D. Offgas isolation 2.4OG (Off-Gas Abnormal) and 5.2FUEL (Fuel Failure).
-Gas Abnormal) and 5.2FUEL (Fuel Failure).
86
86 Explanation:
 
The interrelationship between high off-site release rate and the Offgas system is that the Offgas system automatically isolates to minimize/terminate off
Explanation:
-site releases.
The interrelationship between high off-site release rate and the Offgas system is that the Offgas system automatically isolates to minimize/terminate off-site releases.
During fuel failure events, MSL radiation level rise along with RB area radiation levels followed by SJAE radiation. MSL Radiation Hi and OG Radiation Hi annunciators are received long before the ERP rad alarms. As radiation levels continue to rise, with no operator action, the OG system will isolate 15 minutes after reaching the Hi
During fuel failure events, MSL radiation level rise along with RB area radiation levels followed by SJAE radiation. MSL Radiation Hi and OG Radiation Hi annunciators are received long before the ERP rad alarms. As radiation levels continue to rise, with no operator action, the OG system will isolate 15 minutes after reaching the Hi-Hi radiation value thus terminating the release from the main condenser. If MSL rad Hi-Hi is received, the reactor is required to be Scrammed and if shutdown, the MSIVs and MSL drains are required to be closed. Per OFFGAS High Rad annunciator, procedures 2.4OG & 5.2Fuel are required to be entered. Per MSL Rad High annunciator, Procedure 5.2FUEL is required to be entered if raised radiation levels are not due to hydrogen injection. The Control Room Operator is expected to recognize entry conditions from memory in Abnormal and Emergency procedures.
-Hi radiation value thus terminating the release from the main condenser. If MSL rad Hi
Distracters:
-Hi is received, the reactor is required to be Scrammed and if shutdown, the MSIVs and MSL drains are required to be closed. Per OFFGAS High Rad annunciator, procedures 2.4OG & 5.2Fuel are required to be entered. Per MSL Rad High annunciator, Procedure 5.2FUEL is required to be entered if raised radiation levels are not due to hydrogen injection. The Control Room Operator is expected to recognize entry conditions from memory in Abnormal and Emergency procedures. Distracters:
A. This answer is incorrect because MSIVs do not auto close due to MSL Rad Hi or Hi Hi and 2.4OG is not the ONLY procedure required to be entered. This answer is plausible because the original design of MSL Hi Hi was to isolate the MSIVs. If the stem were changed to reflect Manual action - MSIV closure would be correct and MSL Rad Hi only requires entry into 5.2FUEL. The candidate that confuses Auto vs. Manual action or the original design of MSL Rad Hi Hi and does not remember the entry conditions to AOP 5.2FUEL and 2.4OG would select this answer.
A. This answer is incorrect because MSIVs do not auto close due to MSL Rad Hi or Hi Hi and 2.4OG is not the ONLY procedure required to be entered. This answer is plausible because the original design of MSL Hi Hi was to isolate the MSIVs. If the stem were changed to reflect Manual action  
B. This answer is incorrect because MSIVs do not auto close due to MSL Rad Hi or Hi Hi.
- MSIV closure would be correct and MSL Rad Hi only requires entry into 5.2FUEL. The candidate that confuses Auto vs. Manual action or the original design of MSL Rad Hi Hi and does not remember the entry conditions to AOP 5.2FUEL and 2.4OG would select this answer.
This answer is plausible because the original design of MSL Hi Hi was to isolate the MSIVs.
B. This answer is incorrect because MSIVs do not auto close due to MSL Rad Hi or Hi Hi. This answer is plausible because the original design of MSL Hi Hi was to isolate the MSIVs. If the stem were changed to reflect Manual action  
If the stem were changed to reflect Manual action - MSIV closure would be correct. The candidate that confuses Auto vs. Manual action or the original design of MSL Rad Hi Hi and correctly identifies the entry conditions to AOP 5.2FUEL and 2.4OG would select this answer.
- MSIV closure would be correct. The candidate that confuses Auto vs. Manual action or the original design of MSL Rad Hi Hi and correctly identifies the entry conditions to AOP 5.2FUEL and 2.4OG would select this answer.
C. This answer is incorrect because 2.4OG is not the ONLY procedure required to be entered.
C. This answer is incorrect because 2.4OG is not the ONLY procedure required to be entered. This answer is plausible because and MSL Rad Hi only requires entry into 5.2FUEL. The candidate that correctly identifies Offgas isolation and does not remember the entry conditions to AOP 5.2FUEL and 2.4OG would select this answer.
This answer is plausible because and MSL Rad Hi only requires entry into 5.2FUEL. The candidate that correctly identifies Offgas isolation and does not remember the entry conditions to AOP 5.2FUEL and 2.4OG would select this answer.
Technical Reference(s):
Technical Reference(s):         SOP 2.2.62 Off Gas System, Rev. 27 Proposed references to be provided to applicants during examination:       ___NONE_______
SOP 2.2.62 Off Gas System, Rev. 27 Proposed references to be provided to applicants during examination:
Learning Objective:             COR001-161-1 OPS Off Gas, Rev 27
___NONE_______
Learning Objective:
COR001-161-1 OPS Off Gas, Rev 27  
: 11. Given an Off Gas system component manipulation, predict and explain the change in the following parameters:
: 11. Given an Off Gas system component manipulation, predict and explain the change in the following parameters:
: b. Radioactive release rate 87 Question Source:
: b. Radioactive release rate 87
Bank #   _         _
 
Modified Bank #
Question Source:         Bank #             _     _
_
Modified Bank #   _      _
_  New   ___X___ Question History:
New               ___X___
Last NRC Exam
Question History:         Last NRC Exam     ___ ___
___ ___ Question Cognitive Level:
Question Cognitive Level: Memory or Fundamental Knowledge __ __
Memory or Fundamental Knowledge
Comprehension or Analysis       __ X__
__ __ Comprehension or Analysis   __ X__ 10 CFR Part 55 Content:
10 CFR Part 55 Content:   55.41 (7)
55.41 (7) Comments:
Comments:
LOD 2 88 89 90 91 92 93 94 Per 2.0.3 R88 Conduct Of Operations the following recognition requirements apply:
LOD 2 88
 
89 90 91 92 93 94 Per 2.0.3 R88 Conduct Of Operations the following recognition requirements apply:
Per 2.0.1.2 R44 Operations Procedure Policy the following apply:
Per 2.0.1.2 R44 Operations Procedure Policy the following apply:
95  
95
\ ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
 
-
\
ES-401                           Sample Written Examination                   Form ES-401-5 Question Worksheet Examination Outline Cross-


==Reference:==
==Reference:==
 
Level                 RO Tier #               __1__
Level   RO   Tier #     __1__   Group # __1__   K/A #   _600000AK3.04___
Group #               __1__
Importance Rating
K/A #                 _600000AK3.04___
_2.8__   600000 Plant Fire On Site Knowledge of the reasons for the following responses as they apply to plant fire on site:
Importance Rating     _2.8__
 
600000 Plant Fire On Site Knowledge of the reasons for the following responses as they apply to plant fire on site:
AK3.04 Actions contained in the abnormal procedure for plant fire on site Q ues ti on: 1 9  Th e p l an t i s ope r a ti n g a t 100% w he n t h e S h if t Manage r d ir ec t s a C on tr o l R oo m evacua ti o n du e t o a fir e i n t h e con tr o l r oo m.
AK3.04 Actions contained in the abnormal procedure for plant fire on site Question: 19 The plant is operating at 100% when the Shift Manager directs a Control Room evacuation due to a fire in the control room.
W h y a r e a l l t h e A C po w e r e d R eac t o r Feed w a t e r Pu m p l ub e o i l pu m p control s w itche s p lace d i n PU LL-T O-LOCK IAW P rocedu r e 5.4F I RE-S/D (F i r e Induce d Shu t do w n from Outside Con t ro l Room)? T o ensure-  A. a r eac t o r water overfil l event is prevented. B. r eac t o r w a t e r leve l is intentionally lowered to aid in FW preheating
Why are all the AC powered Reactor Feedwater Pump lube oil pump control switches placed in PULL-TO-LOCK IAW Procedure 5.4FIRE-S/D (Fire Induced Shutdown from Outside Control Room)?
. C. automatic start of DC lube oil pumps to maintain RFP bearing lubrication during pump operation. D. automatic start of D C lube o i l pu m p s to m a i n t a i n RFP bea ri n g l ub ri cation du ri n g pu m p coast down
To ensure A. a reactor water overfill event is prevented.
. A ns w e r: A. a r eac t o r water overfil l event is prevented
B. reactor water level is intentionally lowered to aid in FW preheating.
.
C. automatic start of DC lube oil pumps to maintain RFP bearing lubrication during pump operation.
D. automatic start of DC lube oil pumps to maintain RFP bearing lubrication during pump coast down.
Answer:
A. a reactor water overfill event is prevented.
Explanation:
Explanation:
Prior to evacuation of the Control Room, the RFP AC Lube oil pumps are placed in PULL
Prior to evacuation of the Control Room, the RFP AC Lube oil pumps are placed in PULL-TO-LOCK to prevent an overfill event from occurring. This action results in the RFPs tripping due to low bearing oil pressure with DC Lube oil pumps auto starting to provide bearing lubrication during pump coast down. Reactor water level lowers and HPCI may be used for RPV level control from the ASD Panel.
-TO-LOCK to prevent an overfill event from occurring. This action results in the RFPs tripping due to low bearing oil pressure with DC Lube oil pumps auto starting to provide bearing lubrication during pump coast down. Reactor water level lowers and HPCI may be used for RPV level control from the ASD Panel.
 
Distracters:
Distracters:
1 B. This answer is incorrect because the intent of the procedure guidance is to prevent an overfill event. RPV level is intentionally lowered during ATWS events to preheat the feedwater and lower reactor power which makes this answer plausible. The candidate who creates an incorrect mental model would recall the requirement for intentionally lowering RPV level by tripping the RFPs.
1


C. This answer is incorrect because the intent of the procedure guidance is to prevent an overfill event. The action of placing the AC lube oil pumps to Pu ll-to-Lock automatically starts the DC lube oil pumps for the RFPs but these pumps are only designed to protect the RFPs during coast down and are not of sufficient capacity to support continued operation and automatic level control. Additionally, with the MSIVs closed, there will be no motive force for the RFPs to maneuver. The candidate selects this if it is believed the RFPs can continue to be controlled by RVLC in automatic with only the DC lube oil pumps. This answer is plausible because the DC oil pumps are provided to ensure proper lubrication
B. This answer is incorrect because the intent of the procedure guidance is to prevent an overfill event. RPV level is intentionally lowered during ATWS events to preheat the feedwater and lower reactor power which makes this answer plausible. The candidate who creates an incorrect mental model would recall the requirement for intentionally lowering RPV level by tripping the RFPs.
. D. This answer is incorrect because the intent of the procedure guidance is to prevent an RPV overfill event by securing the control oil and allowing the RFPTs to trip. Stopping the AC lube oil pumps will auto start the DC lube oil pumps to support RFP bearing lubrication during coast down following RFP trip, but this is not the reason that the AC lube oil pumps are secured IAW 5.4FIRE. This selection is plausible and the candidate may select it knowing how the RFLO system operates and not the reason for performance IAW 5.4FIRE.
C. This answer is incorrect because the intent of the procedure guidance is to prevent an overfill event. The action of placing the AC lube oil pumps to Pull-to-Lock automatically starts the DC lube oil pumps for the RFPs but these pumps are only designed to protect the RFPs during coast down and are not of sufficient capacity to support continued operation and automatic level control. Additionally, with the MSIVs closed, there will be no motive force for the RFPs to maneuver. The candidate selects this if it is believed the RFPs can continue to be controlled by RVLC in automatic with only the DC lube oil pumps. This answer is plausible because the DC oil pumps are provided to ensure proper lubrication.
Technical Reference(s):
D. This answer is incorrect because the intent of the procedure guidance is to prevent an RPV overfill event by securing the control oil and allowing the RFPTs to trip. Stopping the AC lube oil pumps will auto start the DC lube oil pumps to support RFP bearing lubrication during coast down following RFP trip, but this is not the reason that the AC lube oil pumps are secured IAW 5.4FIRE. This selection is plausible and the candidate may select it knowing how the RFLO system operates and not the reason for performance IAW 5.4FIRE.
5.4 FIRE-SD FIRE INDUCED SHUTDOWN FROM OUTSIDE CONTROL ROOM, Rev. 62.
Technical Reference(s):         5.4 FIRE-SD FIRE INDUCED SHUTDOWN FROM OUTSIDE CONTROL ROOM, Rev. 62.
Proposed references to be provided to applicants during examination:
Proposed references to be provided to applicants during examination:         ___NONE_______
___NONE_______
Learning Objective:             INT032-01-34 OPS CNS Abnormal Procedures (RO) Fire H. Given plant condition(s) and the applicable Abnormal/Emergency Procedure, determine the correct subsequent actions required to mitigate the event(s).
Learning Objective:
Question Source:                Bank #                _      _
INT032-01-34 OPS CNS Abnormal Procedures (RO) Fire
Modified Bank #        _      _
New                    ___X___
Question History:              Last NRC Exam          ___ ___
Question Cognitive Level:      Memory or Fundamental Knowledge              __X__
Comprehension or Analysis                    __ __
10 CFR Part 55 Content:        55.41 (10)
Comments:
LOD 3 2


H. Given plant condition(s) and the applicable Abnormal/Emergency Procedure, determine the correct subsequent actions required to mitigate the event(s).
3 4
Question Source:
5 6
Bank #  _          _
ES-401                           Sample Written Examination                 Form ES-401-5 Question Worksheet Examination Outline Cross-
Modified Bank #
_          _
New  ___X___  Question History:
Last NRC Exam
___ ___  Question Cognitive Level:
Memory or Fundamental Knowledge
__X__ Comprehension or Analysis  __  __
10 CFR Part 55 Content:
55.41 (10)
Comments:  LOD 3  2 3
4 5
6 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
-


==Reference:==
==Reference:==
 
Level               RO Tier #               __1__
Level   RO   Tier #     __1__   Group # __1__   K/A #   _700000 AA1.01___
Group #             __1__
Importance Rating
K/A #               _700000 AA1.01___
_3.6__   700000 Generator Voltage and Electric Grid Disturbances:
Importance Rating   _3.6__
700000 Generator Voltage and Electric Grid Disturbances:
Ability to operate and/or monitor the following as they apply to generator voltage and electric grid disturbances:
Ability to operate and/or monitor the following as they apply to generator voltage and electric grid disturbances:
AA1.01 Grid frequency and voltag e  Question: 20 The plant is operating at 70% power when Doniphan Control Center (DCC) System Operator notifies the Control Room that a grid disturbance near Cooper is occurring.
AA1.01 Grid frequency and voltage Question: 20 The plant is operating at 70% power when Doniphan Control Center (DCC) System Operator notifies the Control Room that a grid disturbance near Cooper is occurring.
 
The following conditions exist:
The following conditions exist:
345 KV voltage is RISING.
* 345 KV voltage is RISING.
Main Generator exciter AMPs are RISING.
* Main Generator exciter AMPs are RISING.
NPPD System frequency is 59.8 hertz and steady.
* NPPD System frequency is 59.8 hertz and steady.
What is causing the above indications?
What is causing the above indications?
What is the required position of the GEN VOLTAGE REGULATOR switch under these conditions IAW Procedure 5.3GRID (Degraded Grid Voltage)?
What is the required position of the GEN VOLTAGE REGULATOR switch under these conditions IAW Procedure 5.3GRID (Degraded Grid Voltage)?
A. The Grid OFF B. The Grid ON C. CNS OFF D. CNS ON Answer:
C. CNS OFF 7


A. The Grid OFF  B. The Grid ON  C. CNS  OFF  D. CNS  ON  Answer: C. CNS  OFF    7 Explanation:
Explanation:
Actions in this procedure to place the voltage regulator to OFF are mandated under conditions indicative of a CNS voltage regulator oscillation causing 345 kV voltage variations. If the grid voltage is oscillating and voltage regulator is working as designed, as grid voltage lowers regulator will raise field amps in an attempt to maintain terminal voltage. If regulator is causing the oscillations, field amps rising will cause 345 kV volts to rise. So, if 345 kV voltage rises as field amps lower, regulator is working properly and there is no need to transfer to the base adjuster. If 345 kV voltage rises as our field amps rise, then the CNS voltage regulator is driving the voltage oscillations and transferring to the base adjuster should stabilize conditions.
Actions in this procedure to place the voltage regulator to OFF are mandated under conditions indicative of a CNS voltage regulator oscillation causing 345 kV voltage variations. If the grid voltage is oscillating and voltage regulator is working as designed, as grid voltage lowers regulator will raise field amps in an attempt to maintain terminal voltage. If regulator is causing the oscillations, field amps rising will cause 345 kV volts to rise. So, if 345 kV voltage rises as field amps lower, regulator is working properly and there is no need to transfer to the base adjuster. If 345 kV voltage rises as our field amps rise, then the CNS voltage regulator is driving the voltage oscillations and transferring to the base adjuster should stabilize conditions.
Distracters:
Distracters:
A. This answer is incorrect because the CNS voltage regulator is causing the problem. This answer is plausible because the voltage regulator switch position is correct. The candidate who does not realize rising voltage regulator output and corresponding grid voltage rising is caused by the voltage regulator would select this answer.
A. This answer is incorrect because the CNS voltage regulator is causing the problem. This answer is plausible because the voltage regulator switch position is correct. The candidate who does not realize rising voltage regulator output and corresponding grid voltage rising is caused by the voltage regulator would select this answer.
B. This answer is incorrect because the CNS voltage regulator is causing the problem and 5.3GRID directs placing the voltage regulator switch in OFF. If the candidate does not understand that the current condition on the grid is occurring due to the CNS main generator voltage regulator failing high in automatic they may chose this answer.
B. This answer is incorrect because the CNS voltage regulator is causing the problem and 5.3GRID directs placing the voltage regulator switch in OFF. If the candidate does not understand that the current condition on the grid is occurring due to the CNS main generator voltage regulator failing high in automatic they may chose this answer.
D. This answer is incorrect because the CNS voltage regulator is causing the problem so its control switch must be placed in OFF. The candidate that does not recall the correct procedure guidance from 5.3GRID would select this answer.
D. This answer is incorrect because the CNS voltage regulator is causing the problem so its control switch must be placed in OFF. The candidate that does not recall the correct procedure guidance from 5.3GRID would select this answer.
Technical Reference(s):
Technical Reference(s):         5.3GRID Degraded Grid Voltage, Rev. 41 Proposed references to be provided to applicants during examination:          ___NONE_______
5.3GRID Degraded Grid Voltage, Rev. 41
Learning Objective:              INT032-01-31 CNS Abnormal Procedures (RO) Electrical S. Given plant condition(s) and the applicable Abnormal/Emergency Procedure, determine the correct subsequent actions required to mitigate the event(s).
Question Source:                Bank #                _        _
Modified Bank #      _        _
New                  ___X___
Question History:                Last NRC Exam        ___ ___
Question Cognitive Level:        Memory or Fundamental Knowledge              __ __
Comprehension or Analysis                    __ X_
10 CFR Part 55 Content:          55.41 (10)
Comments:
LOD 3 8


Proposed references to be provided to applicants during examination:
9 10 11 12 13 ES-401                           Sample Written Examination                       Form ES-401-5 Question Worksheet Examination Outline Cross-
___NONE_______
Learning Objective:
INT032-01-31 CNS Abnormal Procedures (RO) Electrical S. Given plant condition(s) and the applicable Abnormal/Emergency Procedure, determine the correct subsequent actions required to mitigate the event(s).
 
Question Source:
Bank #  _          _
Modified Bank #
_
_  New  ___X___  Question History:
Last NRC Exam
___ ___  Question Cognitive Level:
Memory or Fundamental Knowledge
__ __ Comprehension or Analysis  __ X_
10 CFR Part 55 Content:
55.41 (10)  Comments: LOD 3  8 9
10 11 12 13 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
-


==Reference:==
==Reference:==
 
Level                 RO Tier #               __1__
Level   RO   Tier #     __1__   Group # __1__   K/A #   _295002 2.4.50___
Group #               __1__
Importance Rating
K/A #                 _295002 2.4.50___
_4.2__   295002 Loss of Main Condenser Vacuum:
Importance Rating     _4.2__
295002 Loss of Main Condenser Vacuum:
2.4.50 Ability to verify system alarm setpoints and operate controls identified in the alarm response manual.
2.4.50 Ability to verify system alarm setpoints and operate controls identified in the alarm response manual.
Q ues ti on: 2 1  The P l an t i s ope r a ti n g a t f u l l po w e r with the f o ll o w i n g conditi on s: A l a r m B-1/B-3 T G L O W VACUU M PR E TR I P sounds. M a i n G ene r a t o r l oa d i s 80 5 M W e.
Question: 21 The Plant is operating at full power with the following conditions:
(1) W he r e i s t h e a l a r m se t po i n t validated? (2) W ha t ac ti o n i s d ir ec t e d b y B-1/B-3 T G L O W VACUU M PR E TR I P, when vacuu m " H g? A. (1) CONDENSE R PRESSUR E TR I P GRAP H o n DE H HM I.   (2) P e rf o r m R ap i d P o w e r reducti on. B. (1) CONDENSE R PRESSUR E TR I P GRAP H o n RV L C HM I.   (2) S c r a m an d tri p t h e t u r b i ne. C. (1) CONDENSE R PRESSUR E TR I P GRAP H o n RV L C HM I.   (2) P e rf o r m R ap i d P o w e r reducti on. D. (1) CONDENSE R PRESSUR E TR I P GRAP H o n DE H HM I.   (2) S c r a m an d tri p t h e t u r b i ne. A ns w e r: D. (1) CONDENSE R PRESSUR E TR I P GRAP H o n DE H HM I.   (2) S c r a m an d tri p t h e t u r b i ne. Explanation:
* Alarm B-1/B-3 TG LOW VACUUM PRE TRIP sounds.
The main condenser low vacuum alarms (Pre
* Main Generator load is 805 MWe.
-trip and Trip) are received from the ALARMS program module of the Trip Tricon unit. The DEH HMI contains a dynamic display, CONDENSER PRESSURE TRIP GRAPH. The control room operator verifies the alarm by noting the current plant operating status. With the condition given, the CONDENSER PRESSURE TRIP GRAPH indicates the operating point very close to the 8.25 absolute pressure trip point. Alarm B
(1) Where is the alarm setpoint validated?
-1/B-3 directs tripping the main turbine if condenser vacuum cannot 14 Distracters:
(2) What action is directed by B-1/B-3 TG LOW VACUUM PRE TRIP, when vacuum CANNOT be maintained 23" Hg?
A. This answer is incorrect because rapid power reduction is not the correct action to take. This answer is plausible because the correct HMI is listed and lowering reactor power aids in condenser vacuum recovery. The candidate who knows the correct location to verify the alarm and knows reactor power reduction helps in mitigating the low vacuum would select this answer.
A. (1) CONDENSER PRESSURE TRIP GRAPH on DEH HMI.
(2) Perform Rapid Power reduction.
B. (1) CONDENSER PRESSURE TRIP GRAPH on RVLC HMI.
(2) Scram and trip the turbine.
C. (1) CONDENSER PRESSURE TRIP GRAPH on RVLC HMI.
(2) Perform Rapid Power reduction.
D. (1) CONDENSER PRESSURE TRIP GRAPH on DEH HMI.
(2) Scram and trip the turbine.
Answer:
D. (1) CONDENSER PRESSURE TRIP GRAPH on DEH HMI.
(2) Scram and trip the turbine.
Explanation:
The main condenser low vacuum alarms (Pre-trip and Trip) are received from the ALARMS program module of the Trip Tricon unit. The DEH HMI contains a dynamic display, CONDENSER PRESSURE TRIP GRAPH. The control room operator verifies the alarm by noting the current plant operating status. With the condition given, the CONDENSER PRESSURE TRIP GRAPH indicates the operating point very close to the 8.25 absolute pressure trip point. Alarm B-1/B-3 directs tripping the main turbine if condenser vacuum cannot be maintained 23" Hg.
14


B. This answer is incorrect because the Main Turbine low vacuum pre trip alarm cannot be verified on the RVLCS HMI. This answer is plausible because the RVLCS HMI can display condenser vacuum and the RFP turbine receives a trip on loss of vacuum. The candidate who believes the vacuum pre
Distracters:
-trip alarm can be verified on the RVLCS HMI would select this answer. C. This answer is incorrect because rapid power reduction is not the correct action to take and the RVLCS HMI is not the correct place to verify the alarm. The abnormal procedure 2.4VAC, Loss of Condenser Vacuum provides direction to lower reactor power but does not direct rapid power reduction so this answer is plausible. The alarm procedure, B
A. This answer is incorrect because rapid power reduction is not the correct action to take.
-1/B-3 does not direct lowering reactor power, but does provide direction to scram the reactor if vacuum is below 23" Hg. If the candidate cannot recall the scram action from B
This answer is plausible because the correct HMI is listed and lowering reactor power aids in condenser vacuum recovery. The candidate who knows the correct location to verify the alarm and knows reactor power reduction helps in mitigating the low vacuum would select this answer.
-1/B-3 would select this answer. This answer is plausible because rapidly lowering reactor power aids in condenser vacuum recovery.
B. This answer is incorrect because the Main Turbine low vacuum pre trip alarm cannot be verified on the RVLCS HMI. This answer is plausible because the RVLCS HMI can display condenser vacuum and the RFP turbine receives a trip on loss of vacuum. The candidate who believes the vacuum pre-trip alarm can be verified on the RVLCS HMI would select this answer.
 
C. This answer is incorrect because rapid power reduction is not the correct action to take and the RVLCS HMI is not the correct place to verify the alarm. The abnormal procedure 2.4VAC, Loss of Condenser Vacuum provides direction to lower reactor power but does not direct rapid power reduction so this answer is plausible. The alarm procedure, B-1/B-3 does not direct lowering reactor power, but does provide direction to scram the reactor if vacuum is below 23 Hg. If the candidate cannot recall the scram action from B-1/B-3 would select this answer. This answer is plausible because rapidly lowering reactor power aids in condenser vacuum recovery.
Technical Reference(s):
Technical Reference(s):       2.4VAC LOSS OF CONDENSER VACUUM, Rev. 25 ALARM PROCEDURE 2.3_B-1 PANEL B - ANNUNCIATOR B-1, Rev. 34 Proposed references to be provided to applicants during examination:       ___NONE_______
2.4VAC LOSS OF CONDENSER VACUUM, Rev. 25 ALARM PROCEDURE 2.3_B
Learning Objective: INT032-01-32 CNS Abnormal Procedures (RO) Off Gas/Vacuum J.     Given plant condition(s), determine from memory if a manual reactor scram or an emergency shutdown from power is required due to the event(s).
-1 PANEL B  
K.     Given plant condition(s), determine from memory if a Main Turbine trip is required due to the event(s).
- ANNUNCIATOR B
Question Source:               Bank #
-1,   Rev. 34   Proposed references to be provided to applicants during examination:
Modified Bank #
___NONE_______
New X Question History:             Last NRC Exam Question Cognitive Level:      Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content:        55.41 (10)
Learning Objective:
Comments:
INT032-01-32 CNS Abnormal Procedures (RO) Off Gas/Vacuum J. Given plant condition(s), determine from memory if a manual reactor scram or an emergency shutdown from power is required due to the event(s).
LOD 3 15
K. Given plant condition(s), determine from memory if a Main Turbine trip is required due to the event(s).
Question Source:
Bank #   Modified Bank #
New X Question History:
Last NRC Exam


Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content:
16 17
55.41 (10)
\
Comments: LOD 3  15 16 17  
ES-401                           Sample Written Examination                   Form ES-401-5 Question Worksheet Examination Outline Cross-
\ ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
-


==Reference:==
==Reference:==
 
Level                 RO Tier #               __1__
Level   RO   Tier #     __1__   Group # __1__   K/A #   _295013 AK1.03___
Group #               __1__
Importance Rating
K/A #                 _295013 AK1.03___
_3.0__   295013 High Suppression Pool Temp.:
Importance Rating     _3.0__
295013 High Suppression Pool Temp.:
Knowledge of the operational implications of the following concepts as they apply to high suppression pool temperature:
Knowledge of the operational implications of the following concepts as they apply to high suppression pool temperature:
(CFR: 41.8 to 41.10)
(CFR: 41.8 to 41.10)
 
AK1.03 Localized heating Question: 22 The plant is operating at 100% power.
AK1.03 Localized heating Question: 22 The plant is operating at 100% power.
* All MSIVs close causing a Reactor Scram.
All MSIVs close causing a Reactor Scram.
* The operator takes manual control of RPV pressure using the SRVs.
The operator takes manual control of RPV pressure using the SRVs.
 
Why are the SRVs alternated while controlling pressure IAW Procedure 2.2.1 (Nuclear Pressure Relief System)?
Why are the SRVs alternated while controlling pressure IAW Procedure 2.2.1 (Nuclear Pressure Relief System)?
To prevent-A. localized torus overheating.
To prevent A. localized torus overheating.
 
B. SRV failure due to excessive cycling.
B. SRV failure due to excessive cycling.
C. SRV failure due to excessive valve temperature.
C. SRV failure due to excessive valve temperature.
D. inaccurate average torus temperature indication.  
D. inaccurate average torus temperature indication.
Answer:
A. localized torus overheating.
Explanation:
The SRV discharge is into the suppression pool. The energy is transported thru the SRV tailpipe to a T-Quencher located approximately 6 feet below normal Suppression Pool water level. If the same valve is re-opened repeatedly, the area in the Suppression Pool does not have a chance of mixing and a localized high temperature results.
Distracters:
B. This answer is incorrect because the procedure requires a minimum of 3 seconds between re-opening the same SRV. If the SRV was re-opened without the 3 second wait, then the SRV tailpipe could sustain damage due to the water slug being drawn into the tailpipe 18


Answer: A. localized torus overheatin
because the vacuum breakers are not given enough time to clear the column of water. The candidate may pick this if he/she does not recall the wait period between valve re-opening.
: g. Explanation:
The SRV discharge is into the suppression pool. The energy is transported thru the SRV tailpipe to a T
-Quencher located approximately 6 feet below normal Suppression Pool water level. If the same valve is re
-opened repeatedly, the area in the Suppression Pool does not have a chance of mixing and a localized high temperature results.
Distracters:
B. This answer is incorrect because the procedure requires a minimum of 3 seconds between re-opening the same SRV. If the SRV was re
-opened without the 3 second wait, then the SRV tailpipe could sustain damage due to the water slug being drawn into the tailpipe 18 because the vacuum breakers are not given enough time to clear the column of water. The candidate may pick this if he/she does not recall the wait period between valve re
-opening.
C. This answer is incorrect because the valve is designed to withstand the temperature of the fluid flowing through it. The candidate may select this answer if he/she does not understand that using a single SRV would not result in valve overheating.
C. This answer is incorrect because the valve is designed to withstand the temperature of the fluid flowing through it. The candidate may select this answer if he/she does not understand that using a single SRV would not result in valve overheating.
D. This answer is incorrect because alternating the location of the heat addition aids in torus mixing and allows an accurate average torus water temperature but this is not the reason for alternating the opening of the SRVs. The candidate may pick this if he/she believes the reason for alternating the opening sequence may invalidate an accurate average temperature. This answer is plausible because the energy is added to the same body of water at various locations.
D. This answer is incorrect because alternating the location of the heat addition aids in torus mixing and allows an accurate average torus water temperature but this is not the reason for alternating the opening of the SRVs. The candidate may pick this if he/she believes the reason for alternating the opening sequence may invalidate an accurate average temperature. This answer is plausible because the energy is added to the same body of water at various locations.
Technical Reference(s):
Technical Reference(s):       SOP 2.2.1 Nuclear Pressure Relief System, Rev. 38 Proposed references to be provided to applicants during examination:       ___NONE_______
SOP 2.2.1 Nuclear Pressure Relief System, Rev. 38 Proposed references to be provided to applicants during examination:
Learning Objective:           COR002-16-02 OPS Nuclear Pressure Relief
___NONE_______
: 4.     Given a Nuclear Pressure Relief system component manipulation, predict and explain the changes in the following parameters:
Learning Objective:
: e. Suppression pool temperature Question Source:               Bank #               _       _
COR002-16-02 OPS Nuclear Pressure Relief  
Modified Bank #     _      _
: 4. Given a Nuclear Pressure Relief system component manipulation, predict and explain the changes in the following parameters:  
New                 ___X __
: e. Suppression pool temperature Question Source:
Question History:             Last NRC Exam       ___ ___
Bank #   _         _
Question Cognitive Level:     Memory or Fundamental Knowledge             __X__
Modified Bank #
Comprehension or Analysis                   __ __
_
10 CFR Part 55 Content:       55.41 (10)
_  New   ___X __ Question History: Last NRC Exam
Comments:
___ ___ Question Cognitive Level:
LOD 2 19
Memory or Fundamental Knowledge
 
__X__ Comprehension or Analysis   __ __
20 ES-401                           Sample Written Examination                     Form ES-401-5 Question Worksheet Examination Outline Cross-
10 CFR Part 55 Content:
55.41 (10)
Comments: LOD 2 19 20 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
-


==Reference:==
==Reference:==
 
Level                 RO Tier #                 __1__
Level   RO   Tier #     __1__   Group # __1__   K/A #   _295002 AK2.02___
Group #               __1__
Importance Rating
K/A #                 _295002 AK2.02___
_3.1__   295022 Loss of CRD Pumps:
Importance Rating     _3.1__
295022 Loss of CRD Pumps:
Knowledge of the interrelations between loss of CRD pumps and the following:
Knowledge of the interrelations between loss of CRD pumps and the following:
AK2.02 CRD mechanism Q ues ti on: 2 3 A p l an t s t a rt u p i s i n p r og r es s w it h pressure set a t 92 6 ps i g an d reac t o r po w e r a t 8%. CR D P u m p A tri p s an d canno t b e r es t a rt ed. A l l a tt e m p t s b y t h e c r e w t o s t a r t CR D P u m p B a r e unsuccess f u l. W ha t i s t h e consequenc e t o t h e C on tr o l R o d D ri v e Mechan i s m (CRDM) i f ope r a ti o n continue s w it h t hes e conditi ons? A. Reduce d sea l lif e.
AK2.02 CRD mechanism Question: 23 A plant startup is in progress with pressure set at 926 psig and reactor power at 8%.
* CRD Pump A trips and cannot be restarted.
* All attempts by the crew to start CRD Pump B are unsuccessful.
What is the consequence to the Control Rod Drive Mechanism (CRDM) if operation continues with these conditions?
A. Reduced seal life.
B. Unlatching of the collet fingers.
B. Unlatching of the collet fingers.
C. Cooling water orifice plugging.
C. Cooling water orifice plugging.
D. Fa il u r e o f t h e p i s t o n t ub e asse m b l y. A ns w e r: A. Reduce d sea l lif e. Explanation:
D. Failure of the piston tube assembly.
From 2.4CRD Loss of or inadequate cooling water to the CRDMs causes the inability to move rods and elevated CRDM temperatures. The CRDMs can operate without cooling water flow but seal life may be shortened by exposure to reactor operating temperatures.
Answer:
CRDM temperatures over 350&deg;F may result in a measurable delay in scram response times. A rise to 400&deg;F could result in up to a 0.150 second rise in the 90% insertion time for an otherwise normally performing CRD. The evaluation of Scram time Tau correction is performed by Procedure 10.35 when temperature exceeds 350&deg;F.
A. Reduced seal life.
Explanation:
From 2.4CRD Loss of or inadequate cooling water to the CRDMs causes the inability to move rods and elevated CRDM temperatures. The CRDMs can operate without cooling water flow but seal life may be shortened by exposure to reactor operating temperatures. CRDM temperatures over 350&deg;F may result in a measurable delay in scram response times. A rise to 400&deg;F could result in up to a 0.150 second rise in the 90% insertion time for an otherwise normally performing CRD. The evaluation of Scram time Tau correction is performed by Procedure 10.35 when temperature exceeds 350&deg;F.
Distracters:
Distracters:
21 B This answer is incorrect because unlatching of the collet fingers is not caused by loss of CRD flow. This answer is plausible because loss of CRD flow through the mechanism would allow more weight of the drive to rest on the collet fingers. The candidate that misinterprets the CRDM internal flow paths and effects would select this answer.
21
 
B This answer is incorrect because unlatching of the collet fingers is not caused by loss of CRD flow. This answer is plausible because loss of CRD flow through the mechanism would allow more weight of the drive to rest on the collet fingers. The candidate that misinterprets the CRDM internal flow paths and effects would select this answer.
C. This answer is incorrect because cooling water plugging is not an immediate issue. This answer is plausible because suddenly stopping cooling water flow could cause impurities settled in the mechanism to become disturbed. The screen at the mechanism cooling water inlet should filter out the larger debris that could cause orifice plugging which makes this answer plausible.
C. This answer is incorrect because cooling water plugging is not an immediate issue. This answer is plausible because suddenly stopping cooling water flow could cause impurities settled in the mechanism to become disturbed. The screen at the mechanism cooling water inlet should filter out the larger debris that could cause orifice plugging which makes this answer plausible.
D. This answer is incorrect because there is no concern with the piston tube assembly. Elevated drive temperatures result in elevated seal temperatures and the candidate may believe seal failure could result in elevated d/p across the piston tube assembly and damage to the piston tube. This answer is plausible because piston tube assembly damage can occur in CRDMs.
D. This answer is incorrect because there is no concern with the piston tube assembly.
Technical Reference(s):
Elevated drive temperatures result in elevated seal temperatures and the candidate may believe seal failure could result in elevated d/p across the piston tube assembly and damage to the piston tube. This answer is plausible because piston tube assembly damage can occur in CRDMs.
2.4CRD CRD Trouble, Rev 15 Proposed references to be provided to applicants during examination:
Technical Reference(s):           2.4CRD CRD Trouble, Rev 15 Proposed references to be provided to applicants during examination:           ___NONE_______
___NONE_______
Learning Objective:               COR002-04-02 Control Rod Drive Hydraulics
Learning Objective:
COR002-04-02 Control Rod Drive Hydraulics
: 12. Given a specific CRDH system malfunction, determine the effect on any of the following:
: 12. Given a specific CRDH system malfunction, determine the effect on any of the following:
: c. Control Rod Drive Mechanisms (CRDMs)
: c.       Control Rod Drive Mechanisms (CRDMs)
Question Source:                  Bank #                  19962 Modified Bank #
New Question History:                Last NRC Exam          ___ ___
Question Cognitive Level:        Memory or Fundamental Knowledge Comprehension or Analysis                    X 10 CFR Part 55 Content:          55.41 (7)
Comments:
LOD 3 22


Question Source:
23 24 25 ES-401                           Sample Written Examination                     Form ES-401-5 Question Worksheet Examination Outline Cross-
Bank #      19962  Modified Bank #
New    Question History:
Last NRC Exam
___ ___
Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis  X  10 CFR Part 55 Content:
55.41 (7)
Comments: LOD 3  22 23 24 25 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
-


==Reference:==
==Reference:==
 
Level               RO Tier #               __1__
Level   RO   Tier #     __1__   Group # __1__   K/A #   _295033 EK3.02___
Group #             __1__
Importance Rating
K/A #               _295033 EK3.02___
_3.5__   295033 High Secondary Containment Area Radiation Levels:
Importance Rating   _3.5__
295033 High Secondary Containment Area Radiation Levels:
Knowledge of the reasons for the following responses as they apply to high secondary containment area radiation levels:
Knowledge of the reasons for the following responses as they apply to high secondary containment area radiation levels:
EK3.02 Reactor SCRAM Q ues t i on: 2 4  The plant is operating at 100% power when an un
EK3.02 Reactor SCRAM Question: 24 The plant is operating at 100% power when an un-isolable RCIC steam line failure results in high radiation in the secondary containment.
-isolable RCIC steam line failure results in high radiation in the secondary containment.
What is the reason for inserting a reactor scram before secondary containment radiation levels reach the Max Safe value?
What is the reason for inserting a reactor scram before secondary containment radiation levels reach the Max Safe value? A. To preclude flooding of the Northeast Quadrant.
A. To preclude flooding of the Northeast Quadrant.
B. To prevent ANY component environmental qualification (EQ) limit from being exceeded. C. To reduce the driving head of the primary system discharging into secondary containment
B. To prevent ANY component environmental qualification (EQ) limit from being exceeded.
. D. To ensure the reactor is shut down before radiation release rates exceed the values for a General Emergency
C. To reduce the driving head of the primary system discharging into secondary containment.
. A ns w e r: C. To reduce the driving head of the primary system discharging into secondary containment
D. To ensure the reactor is shut down before radiation release rates exceed the values for a General Emergency.
. Explanation:
Answer:
C. To reduce the driving head of the primary system discharging into secondary containment.
Explanation:
With a primary system discharging into secondary containment, the EPGs give the basis for when to scram the plant and conduct an emergency depressurization. Scramming reduces the driving head of a primary system that is discharging into the secondary containment and in anticipation of performing an emergency depressurization if radiation levels continue to rise.
With a primary system discharging into secondary containment, the EPGs give the basis for when to scram the plant and conduct an emergency depressurization. Scramming reduces the driving head of a primary system that is discharging into the secondary containment and in anticipation of performing an emergency depressurization if radiation levels continue to rise.
D i s trac t e r s: A. This answer is incorrect because according to the 0
Distracters:
-BARRIER series of procedures, the NE quad (S1 Stair area on attached drawing) is not an area of flooding concern from the steam tunnel or for the RCIC area. Water from this area is normally routed to the 1B Sump which is located in the NE Quad. Due to the nature of a steam leak, the fire detection system will actuate and close RW
A. This answer is incorrect because according to the 0-BARRIER series of procedures, the NE quad (S1 Stair area on attached drawing) is not an area of flooding concern from the steam tunnel or for the RCIC area. Water from this area is normally routed to the 1B Sump which is located in the NE Quad. Due to the nature of a steam leak, the fire detection system will actuate and close RW-AO-771 which redirects any water from the 26
-AO-771 which redirects any water from the 26 steam tunnel drainage to an open discharge in the torus area for holding, therefore the NE Quad area is not jeopardized. This selection is plausible due to the Northeast Quad's location to the MSL tunnel, its potential flooding path and the potential radiological contribution to the surrounding environment. Additionally, there are secondary containment level values associated with Max Normal and Max Safe which require SCRAM and/or ED.
 
B. Th i s ans w e r i s inco rr ec t becaus e t h e sc r a m i s no t base d o n p r o t ec ti n g E Q equ i p m en t i n seconda r y con t a i n m en t. Seconda r y con t a i n m en t ha s nu m e r ou s E Q equ i p m en t t ha t a r e des igne d t o w it hs t an d DBA , an d Spec i a l Even t s s o t h e cand i da t e m a y be li e v e thes e ac ti on s a r e nee d t o p r o t ec t the E Q rating of equ i p m en t. T h i s ans w e r i s p l aus i b l e becaus e p r o t ec ti n g sa f e shu t do w n equ i p m en t i s a p ruden t ac ti o n t o take. D. Th i s ans w e r i s incorr ec t becaus e t h e sc r am is required to reduce the driving head of the leak into secondary containment. Additionally, significant fuel damage would be required to reach the values of a General Emergency. This answer is plausible because it is the reason for a scram if the leak were outside primary and secondary containment. In other words, if the candidate were on the Radioactive Release Control leg of EOP
steam tunnel drainage to an open discharge in the torus area for holding, therefore the NE Quad area is not jeopardized. This selection is plausible due to the Northeast Quads location to the MSL tunnel, its potential flooding path and the potential radiological contribution to the surrounding environment. Additionally, there are secondary containment level values associated with Max Normal and Max Safe which require SCRAM and/or ED.
-5A and a radioactive release to the environment were occurring, then a scram and ED would be required. The candidate who is aware of the reason for the scram in EOP 5A, Radioactive Release Control, and not realize the question does not pertain to radioactive release would select this option.
B. This answer is incorrect because the scram is not based on protecting EQ equipment in secondary containment. Secondary containment has numerous EQ equipment that are designed to withstand DBA, and Special Events so the candidate may believe these actions are need to protect the EQ rating of equipment. This answer is plausible because protecting safe shutdown equipment is a prudent action to take.
Technical Reference(s):
D. This answer is incorrect because the scram is required to reduce the driving head of the leak into secondary containment. Additionally, significant fuel damage would be required to reach the values of a General Emergency. This answer is plausible because it is the reason for a scram if the leak were outside primary and secondary containment. In other words, if the candidate were on the Radioactive Release Control leg of EOP-5A and a radioactive release to the environment were occurring, then a scram and ED would be required. The candidate who is aware of the reason for the scram in EOP 5A, Radioactive Release Control, and not realize the question does not pertain to radioactive release would select this option.
EOP-5A Secondary Containment Control, Rev. 15 0-BARRIER-MAPS Rev. 4 Attachment 3 B&R Drawing 2038 Sheet 1 Rev. 54 Proposed references to be provided to applicants during examination:
Technical Reference(s):       EOP-5A Secondary Containment Control, Rev. 15 0-BARRIER-MAPS Rev. 4 Attachment 3 B&R Drawing 2038 Sheet 1 Rev. 54 Proposed references to be provided to applicants during examination:       ___NONE_______
___NONE_______
Learning Objective:           INT008-06-17 EOP Flowchart 5A Secondary Containment and Radioactivity Release Control
Learning Objective:
INT008-06-17 EOP Flowchart 5A Secondary Containment and Radioactivity Release Control
: 7. Given plant conditions and EOP flowchart 5A, SECONDARY CONTAINMENT CONTROL and RADIOACTIVITY RELEASE CONTROL, state the reasons for the actions contained in the steps.
: 7. Given plant conditions and EOP flowchart 5A, SECONDARY CONTAINMENT CONTROL and RADIOACTIVITY RELEASE CONTROL, state the reasons for the actions contained in the steps.
Question Source:
Question Source:             Bank #                 _       _
Bank #   _         _
Modified Bank #         _        _
Modified Bank #
New                     __X____
_
Question History:             Last NRC Exam Question Cognitive Level:     Memory or Fundamental Knowledge               __X__
_  New   __X____ Question History:
Comprehension or Analysis                     __ __
Last NRC Exam Question Cognitive Level:
10 CFR Part 55 Content:       55.41 (10)
Memory or Fundamental Knowledge
Comments:
__X__ Comprehension or Analysis   __ __
LOD 3 27
10 CFR Part 55 Content:
 
55.41 (10)
28 Technical Reference redacted due to SUNSI considerations
Comments: LOD 3 27 28  
 
Section of B&R Drawing 2038 Sheet 1 Rev.54 30


Technical Reference redacted due to SUNSI considerations Section of B&R Drawing 2038 Sheet 1 Rev.54 30 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
ES-401                             Sample Written Examination                       Form ES-401-5 Question Worksheet Examination Outline Cross-
-


==Reference:==
==Reference:==
 
Level               RO             SRO Tier #               __1__           _____
Level   RO SRO Tier #     __1__ _____ Group # __2__ _____ K/A #   _295034 EA1.03
Group #             __2__           _____
___ Importance Rating
K/A #               _295034 EA1.03___
_4.0 _ _____ 295034 Secondary Containment Ventilation High Radiation / 9  
Importance Rating   _4.0 _         _____
- Ability to operate and / or monitor the following as they apply to secondary containment ventilation high radiation: (CFR: 41.7) EA1.03 Secondary containment ventilation Q ues ti on: 2 5  Th e following R eac tor B u il d i n g V entilation parameters are observed while operating at power
295034 Secondary Containment Ventilation High Radiation / 9 - Ability to operate and / or monitor the following as they apply to secondary containment ventilation high radiation:
:  Exhaus t R ad C hanne l A 1 1 m r/h r  Exhaus t R ad C hanne l B 6 m r/h r  Exhaus t R ad C hanne l C  5 m r/h r  Exhaus t R ad C hanne l D 1 0 m r/h r  R eac tor B u il d i n g DP -0.20" wg N O p l an t sys t e m response s occur. Which of the following is a requ ir e d Con tr o l R oo m Ope r a t o r action f o r Seconda r y Con t a i n m en t Ventil ati o n sys t e m s? A. S t a r t ON E SG T tr a in ONLY. B. S t a r t BO T H SG T tr a i ns. C. Start the Standby RB supply fan
(CFR: 41.7) EA1.03 Secondary containment ventilation Question: 25 The following Reactor Building Ventilation parameters are observed while operating at power:
. D. S t a r t the Standby RB exhaust fan
* Exhaust Rad Channel      A     11 mr/hr
. A ns w e r: B. S t a r t BO T H SG T tr a i ns.
* Exhaust Rad Channel      B     6 mr/hr
* Exhaust Rad Channel      C     5 mr/hr
* Exhaust Rad Channel      D     10 mr/hr
* Reactor Building DP             -0.20 wg NO plant system responses occur.
Which of the following is a required Control Room Operator action for Secondary Containment Ventilation systems?
A. Start ONE SGT train ONLY.
B. Start BOTH SGT trains.
C. Start the Standby RB supply fan.
D. Start the Standby RB exhaust fan.
Answer:
B. Start BOTH SGT trains.
Explanation:
Explanation:
Radiation Monitors A and C are in division 1 while B and D are in division 2. Either division will trip upon receipt of one (1) Radiation monitor upscale trip at 10 mr/hr. With a trip signal present in both divisions, a full Group 6 isolation will be initiated. The Group 6 signal isolates Reactor BLDG Ventilation which causes the initiation of both trains of Standby Gas Treatment.
Radiation Monitors A and C are in division 1 while B and D are in division 2. Either division will trip upon receipt of one (1) Radiation monitor upscale trip at 10 mr/hr. With a trip signal present in both divisions, a full Group 6 isolation will be initiated. The Group 6 signal isolates Reactor BLDG Ventilation which causes the initiation of both trains of Standby Gas Treatment.
 
Distracters:
Distracters:
A. This answer is incorrect because the action is to ensure both SGT trains is required.
A.     This answer is incorrect because the action is to ensure both SGT trains is required.
This answer is plausible because only 1 SGT is required to maintain Secondary Containment negative pressure and there is direction to shutdown 1 SGT per Procedure 31 2.2.73 within 1 hour of receiving Group 6. The candidate that recognizes a Group 6 signal is present and does not know the requirement to start BOTH SGT trains and knows 1 SGT is required to be shut down within 1 hour would select this answer.
This answer is plausible because only 1 SGT is required to maintain Secondary Containment negative pressure and there is direction to shutdown 1 SGT per Procedure 31
C. This is incorrect because starting the Standby RB Supply fan is not required with a Group 6 isolation signal present.
This option is plausible because RB DP is below the procedural minimum (2.2.47) which would require starting an additional fan. The candidate that does not recognize the Group 6 isolation and confuses RB DP as being high would select this answer.


2.2.73 within 1 hour of receiving Group 6. The candidate that recognizes a Group 6 signal is present and does not know the requirement to start BOTH SGT trains and knows 1 SGT is required to be shut down within 1 hour would select this answer.
C. This is incorrect because starting the Standby RB Supply fan is not required with a Group 6 isolation signal present. This option is plausible because RB DP is below the procedural minimum (2.2.47) which would require starting an additional fan. The candidate that does not recognize the Group 6 isolation and confuses RB DP as being high would select this answer.
D. This is incorrect because starting the Standby RB Exhaust fan is not required with a Group 6 isolation signal present. This option is plausible because RB DP is below the procedural minimum (2.2.47) which would require starting an additional fan. The candidate that does not recognize the Group 6 isolation and recognizes RB DP as being low would select this answer.
D. This is incorrect because starting the Standby RB Exhaust fan is not required with a Group 6 isolation signal present. This option is plausible because RB DP is below the procedural minimum (2.2.47) which would require starting an additional fan. The candidate that does not recognize the Group 6 isolation and recognizes RB DP as being low would select this answer.
Technical Reference(s):
Technical Reference(s):       Procedure 2.1.22 (Recovering from a Group Isolation), Rev. 59 Procedure 2.2.47, HVAC Reactor Building, Rev. 51 Procedure 4.7.5, Reactor Building Vent Exhaust Radiation Monitoring System, Rev 18 Proposed references to be provided to applicants during examination:       ___None_________
Procedure 2.1.22 (Recovering from a Group Isolation), Rev. 59 Procedure 2.2.47, HVAC Reactor Building, Rev. 51 Procedure 4.7.5, Reactor Building Vent Exhaust Radiation Monitoring System, Rev 18
Learning Objective:
 
COR001-08-01
Proposed references to be provided to applicants during examination:
: 6.     Describe the interrelationships between HVAC systems and the following:
___None_________  Learning Objective:
: e.     Process Radiation Monitoring system
COR001-08-01 6. Describe the interrelationships between HVAC systems and the following:  
: e. Process Radiation Monitoring system
: 11. Describe the HVAC design features and interlocks that provide for the following:
: 11. Describe the HVAC design features and interlocks that provide for the following:
: b. Secondary containment isolation COR001-18-01 5. Describe the interrelationship between the RM system and the following:
: b.     Secondary containment isolation COR001-18-01
: r. Reactor Building Ventilation system
: 5.     Describe the interrelationship between the RM system and the following:
: 8. Describe the Radiation Monitoring system design feature(s) and/or interlock(s) that provide for the following:
: r.     Reactor Building Ventilation system
: b. Automatic action to contain the radioactive release in the event that the predetermined release rates are exceeded.
: 8.     Describe the Radiation Monitoring system design feature(s) and/or interlock(s) that provide for the following:
Question Source:
: b.     Automatic action to contain the radioactive release in the event that the predetermined release rates are exceeded.
Bank #   __       _ Modified Bank #
Question Source:               Bank #               __     _
_6021__ (See attached)
Modified Bank #       _6021__ (See attached)
New   _______ Question History:
New                   _______
Last NRC Exam
Question History:             Last NRC Exam         ____________
____________
Question Cognitive Level:      Memory or Fundamental Knowledge              __X __
Comprehension or Analysis                    __ __
10 CFR Part 55 Content:        55.41 _(7) _
Comments:
LOD 2 32


Question Cognitive Level:
33 34 35 36 37 38 39 40 41 ES-401                             Sample Written Examination                       Form ES-401-5 Question Worksheet Examination Outline Cross-
Memory or Fundamental Knowledge
__X __ Comprehension or Analysis  __ __
10 CFR Part 55 Content:
55.41 _(7) _
Comments: LOD 2 32 33 34 35 36 37 38 39 40 41 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
-


==Reference:==
==Reference:==
 
Level                   RO             SRO Tier #                   __1__         _____
Level   RO SRO Tier #     __1__ _____ Group # __2__ _____ K/A #   _295035   EA2.02
Group #                 __2__         _____
__ Importance Rating
K/A #                   _295035 EA2.02 __
_2.8 _ _____ 295035 Secondary Containment High Differential Pressure /5  
Importance Rating       _2.8 _         _____
- Ability to determine and/or interpret the following as they apply to secondary containment high differential pressure: (CFR:
295035 Secondary Containment High Differential Pressure /5 - Ability to determine and/or interpret the following as they apply to secondary containment high differential pressure: (CFR: 41.8) EA2.02 Off-site release rate: Plant-Specific Question: 26 The plant is conducting refueling operations involving the movement of recently irradiated fuel assemblies in the secondary containment.
41.8) EA2.02 Off
* A fuel bundle is dropped on the top of the core.
-site release rate: Plant
* A Group 6 isolation occurs due to bundle damage.
-Specific Question: 26 The plant is conducting refueling operations involving the movement of recently irradiated fuel assemblies in the secondary containment.
* Calm winds are indicated on the SPDS Weather Display.
A fuel bundle is dropped on the top of the core.
* HV-MO-262 MG SET-1A INLET and HV-AO-263 MG SET-1A INLET fail to fully isolate.
A Group 6 isolation occurs due to bundle damage.
* Reactor Building Average DP indicates -0.04 wg and stable.
Calm winds are indicated on the SPDS Weather Display.
Which of the following identifies the actual Off-Site radiological release rate response?
HV-MO-262 MG SET
-1A INLET and HV
-AO-263 MG SET
-1A INLET fail to fully isolate.
Reactor Building Average DP indicates  
-0.04" wg and stable.
Which of the following identifies the actual Off
-Site radiological release rate response?
A. ERP release rate rises.
A. ERP release rate rises.
B. ERP release rate lowers then rises.
B. ERP release rate lowers then rises.
C. Reactor Building monitored release rate rises.
C. Reactor Building monitored release rate rises.
D. Reactor Building unmonitored release rate rises.
D. Reactor Building unmonitored release rate rises.
Answer: A. ERP release rate rises.
Answer:
A. ERP release rate rises.
Explanation:
Explanation:
A fuel handling accident involving handling of recently irradiated fuel inside of the secondary containment is one of the two principal accident scenarios for which credit is taken for secondary containment operabilof vacuum water gauge to maintain OPERABILITY. In this case however, the Reactor Building Average DP is below 0 inches water gauge (negative) and stable so both Standby Gas Trains are able to maintain negative building pressure under calm wind conditions and no RB unmonitored release is indicated. Since the Group 6 isolation occurred, the RB exhaust fans have terminated the release from the reactor building exhaust plenum. The ERP release rate rises because both SGTs are now operating and providing increased flow of airborne radiation due to the dropped fuel bundle damage and failure of the HV-MO-262 MG SET
A fuel handling accident involving handling of recently irradiated fuel inside of the secondary containment is one of the two principal accident scenarios for which credit is taken for secondary containment operability. Typically the Secondary Containment requires 0.25 inches of vacuum water gauge to maintain OPERABILITY. In this case however, the Reactor Building Average DP is below 0 inches water gauge (negative) and stable so both Standby Gas Trains are able to maintain negative building pressure under calm wind conditions and no RB unmonitored release is indicated. Since the Group 6 isolation occurred, the RB exhaust fans have terminated the release from the reactor building exhaust plenum. The ERP release rate rises because both SGTs are now operating and providing increased flow of airborne radiation due to the dropped fuel bundle damage and failure of the HV-MO-262 MG SET-1A INLET and 42
-1A INLET and 42 HV-AO-263 MG SET
 
-1A INLET to fully isolate. This flow that was being processed through the RB exhaust plenum is now routed to the ERP via the SGT system.
HV-AO-263 MG SET-1A INLET to fully isolate. This flow that was being processed through the RB exhaust plenum is now routed to the ERP via the SGT system.
Distracters:
Distracters:
B. This answer is incorrect because the ERP release rate will not lower but only rise. The candidate may incorrectly assume that the start of the SGT causes the ERP release rate to lower because of added dilution to the ERP KAMAN calculation is based on the increased ERP flow. However, the plant is in a refueling outage so there is no radiation exiting the ERP. Once the SGT stream containing the radioactive gases from the damaged fuel bundle reach the ERP, the release rate will rise but it will not go down (for quite some time).
B. This answer is incorrect because the ERP release rate will not lower but only rise. The candidate may incorrectly assume that the start of the SGT causes the ERP release rate to lower because of added dilution to the ERP KAMAN calculation is based on the increased ERP flow. However, the plant is in a refueling outage so there is no radiation exiting the ERP. Once the SGT stream containing the radioactive gases from the damaged fuel bundle reach the ERP, the release rate will rise but it will not go down (for quite some time).
 
C. This answer is incorrect because the RB ventilation system has isolated. This choice is plausible if the candidate thinks that the failure of the HV-MO-262 MG SET-1A INLET and HV-AO-263 MG SET-1A INLET to fully isolate contribute to a rise in Reactor Building ventilation release rate. In this case, the building is being maintained negative by the SGT system and not the RB ventilation system.
C. This answer is incorrect because the RB ventilation system has isolated. This choice is plausible if the candidate thinks that the failure of the HV-MO-262 MG SET
-1A INLET and HV-AO-263 MG SET
-1A INLET to fully isolate contribute to a rise in Reactor Building ventilation release rate. In this case, the building is being maintained negative by the SGT system and not the RB ventilation system.
D. This answer is incorrect in this instance, because the SGT system in conjunction with the partial integrity of the Secondary Containment is adequate to maintain a negative RB pressure. This answer is plausible if the candidate does not recognize that the RB pressure is sufficiently negative to prevent an unmonitored release for the current wind conditions.
D. This answer is incorrect in this instance, because the SGT system in conjunction with the partial integrity of the Secondary Containment is adequate to maintain a negative RB pressure. This answer is plausible if the candidate does not recognize that the RB pressure is sufficiently negative to prevent an unmonitored release for the current wind conditions.
 
Technical Reference(s):                   SOP 2.2.47, HVAC Reactor Building, Rev 51 (Attach if not previously provided)       SOP2.2.73, Standby Gas Treatment System, Rev 52.
Technical Reference(s):
Proposed references to be provided to applicants during examination:           _None__________
SOP 2.2.47, HVAC Reactor Building, Rev 51 (Attach if not previously provided)
SOP2.2.73, Standby Gas Treatment System, Rev 52.
Proposed references to be provided to applicants during examination:
_None__________
Learning Objective:
Learning Objective:
COR001-08-01 13. Briefly describe the following concepts as they apply to HVAC:
COR001-08-01
: a. Airborne contamination control COR002-28-02R22 1. State the purpose of the following items related to the Standby Gas Treatment (SGT)
: 13. Briefly describe the following concepts as they apply to HVAC:
System: e. High efficiency inlet filter (HEPA)
: a.       Airborne contamination control COR002-28-02R22
: f. Activated carbon iodine adsorber (charcoal filter)
: 1.     State the purpose of the following items related to the Standby Gas Treatment (SGT)
: e. High efficiency final filter
System:
: 7. Given a specific Standby Gas Treatment System malfunction, determine the effect on any of the following:
: e.       High efficiency inlet filter (HEPA)
: a. Secondary Containment differential pressure
: f.       Activated carbon iodine adsorber (charcoal filter)
: b. Off-Site release rate COR002-03-02R30 7. Describe the interrelationship between Secondary Containment and the following:
: e.       High efficiency final filter
: a. Reactor Building Ventilation
: 7.     Given a specific Standby Gas Treatment System malfunction, determine the effect on any of the following:
: c. SGT d. ERP 19. Predict the consequences of the following items on Secondary Containment:
: a.       Secondary Containment differential pressure
43  
: b.       Off-Site release rate COR002-03-02R30
: c. High airborne radiation
: 7.     Describe the interrelationship between Secondary Containment and the following:
: a.       Reactor Building Ventilation
: c.       SGT
: d.       ERP
: 19. Predict the consequences of the following items on Secondary Containment:
43
: c.     High airborne radiation
: 25. Predict the consequences of a malfunction of the following on Secondary Containment:
: 25. Predict the consequences of a malfunction of the following on Secondary Containment:
: a. Reactor Building Ventilation Question Source:
: a.     Reactor Building Ventilation Question Source:             Bank #             _______
Bank #   _______ Modified Bank #
Modified Bank #     _______
_______  New   ___X___ Question History:
New                 ___X___
Last NRC Exam
Question History:             Last NRC Exam       ____________
____________
Question Cognitive Level:    Memory or Fundamental Knowledge Comprehension or Analysis                X 10 CFR Part 55 Content:      55.41 _(9)_
Comments:
LOD 4 44


Question Cognitive Level:
45 46 47 48 ES-401                             Sample Written Examination                     Form ES-401-5 Question Worksheet Examination Outline Cross-
Memory or Fundamental Knowledge Comprehension or Analysis  X 10 CFR Part 55 Content:
55.41 _(9)_  Comments:
LOD 4  44 45 46 47 48 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
-


==Reference:==
==Reference:==
 
Level                 RO             SRO Tier #               _1___         _____
Level   RO SRO Tier #     _1___ _____ Group # _2___ _____ K/A #   _500000 2.4.21
Group #               _2___         _____
___ Importance Rating
K/A #                 _500000 2.4.21 ___
_4.0 _ _____ 500000 High CTMT Hydrogen Conc. / 5  
Importance Rating     _4.0 _         _____
- 2.4.21 Knowledge of parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactive release control, etc. (CFR: 41.7)
500000 High CTMT Hydrogen Conc. / 5 - 2.4.21 Knowledge of parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactive release control, etc. (CFR: 41.7)
Q ues ti on: 2 7  The plant has experienced a LOCA.
Question: 27 The plant has experienced a LOCA.
Which of the following identifies the MINIMUM Drywell Hydrogen concentration that requires venting and purging Primary Containment (PC)?
Which of the following identifies the MINIMUM Drywell Hydrogen concentration that requires venting and purging Primary Containment (PC)?
During this venting are ODAM Release Rates allowed to be exceeded IAW Procedure 5.8.21 {PC Venting AND Hydrogen Control (Less Than Combustible Limits)}?
During this venting are ODAM Release Rates allowed to be exceeded IAW Procedure 5.8.21 {PC Venting AND Hydrogen Control (Less Than Combustible Limits)}?
PC H2 concentration at... A. 0.34%; Exceeding ODAM limits IS NOT allowed.
PC H2 concentration at...
B. 0.34%; Exceeding ODAM limits IS allowed.  
A. 0.34%; Exceeding ODAM limits IS NOT allowed.
 
B. 0.34%; Exceeding ODAM limits IS allowed.
C. 1%; Exceeding ODAM limits IS NOT allowed.
C. 1%; Exceeding ODAM limits IS NOT allowed.
D. 1%; Exceeding ODAM limits IS allowed.
D. 1%; Exceeding ODAM limits IS allowed.
Answer: C. 1%; Exceeding ODAM limits IS NOT allowed.
Answer:
C. 1%; Exceeding ODAM limits IS NOT allowed.
Explanation:
Explanation:
Requires knowledge of containment H 2 concentration which requires venting (Rad Release Control) and the impact on offsite release. EOP 3A (PCCP) requires venting & purging PC when H 2 concentration reaches 1% only if offsite radioactivity release rate is expected to remain below the offsite release rate limits specified in ODAM. PC H 2 concentration above 1% is an entry condition per EOP
Requires knowledge of containment H 2 concentration which requires venting (Rad Release Control) and the impact on offsite release. EOP 3A (PCCP) requires venting & purging PC when H 2 concentration reaches 1% only if offsite radioactivity release rate is expected to remain below the offsite release rate limits specified in ODAM. PC H 2 concentration above 1% is an entry condition per EOP-3A and is required to be memorized by the CRO.
-3A and is required to be memorized by the CRO.
Distracters:
Distracters:
A. This answer is incorrect due to H 2 concentration being less than 1%. This choice is plausible if the H 2 Hi & Hi Hi alarm setpoints are confused (0.34% is 10% of the Hi Hi alarm setpoint). The candidate who confuses Hi & Hi Hi H 2 alarm setpoints and correctly recognizes release within ODAM limits would select this option.
A. This answer is incorrect due to H 2 concentration being less than 1%. This choice is plausible if the H 2 Hi & Hi Hi alarm setpoints are confused (0.34% is 10% of the Hi Hi alarm setpoint). The candidate who confuses Hi & Hi Hi H 2 alarm setpoints and correctly recognizes release within ODAM limits would select this option.
49 B. This answer is incorrect due to H 2 concentration being less than 1% and having release above ODAM limits. This choice is plausible if the H 2 Hi & Hi Hi alarm setpoints are confused (0.34% is 10% of the Hi Hi alarm setpoint) and if venting PC is confused with the emergency release rate which requires a General Emergency (above ODAM limit). The candidate who confuses Hi & Hi Hi H 2 alarm setpoints and confuses emergency release above ODAM limits would select this option.
49
 
B. This answer is incorrect due to H 2 concentration being less than 1% and having release above ODAM limits. This choice is plausible if the H 2 Hi & Hi Hi alarm setpoints are confused (0.34% is 10% of the Hi Hi alarm setpoint) and if venting PC is confused with the emergency release rate which requires a General Emergency (above ODAM limit). The candidate who confuses Hi & Hi Hi H 2 alarm setpoints and confuses emergency release above ODAM limits would select this option.
D. This answer is incorrect due to having releases above ODAM limits. This choice is plausible if venting PC is confused with the emergency release rate which requires a General Emergency (above ODAM limit). The candidate who confuses Hi & Hi Hi H 2 alarm setpoints and confuses emergency release above ODAM limits would select this option.
D. This answer is incorrect due to having releases above ODAM limits. This choice is plausible if venting PC is confused with the emergency release rate which requires a General Emergency (above ODAM limit). The candidate who confuses Hi & Hi Hi H 2 alarm setpoints and confuses emergency release above ODAM limits would select this option.
Technical Reference(s):
Technical Reference(s):
EOP 5.8.21 {PC Venting AND Hydrogen Control (Less Than Combustible Limits)}, Rev 18 EOP-3A (Primary Containment Control), Rev 15 5.9 H2O2 { Primary Containment Combustible Gas Control (SAG 3)}, Rev 8 Procedure 5.8.22 Proposed references to be provided to applicants during examination:
EOP 5.8.21 {PC Venting AND Hydrogen Control (Less Than Combustible Limits)}, Rev 18 EOP-3A (Primary Containment Control), Rev 15 5.9 H2O2 { Primary Containment Combustible Gas Control (SAG 3)}, Rev 8 Procedure 5.8.22 Proposed references to be provided to applicants during examination:       __None     _________
__None   _________ Learning Objective:
Learning Objective:
COR002-03-02 12. Describe the Containment design features and/or interlocks that provide for the following:
COR002-03-02
: d. Hydrogen control Question Source:
: 12. Describe the Containment design features and/or interlocks that provide for the following:
Bank #   _         _
: d. Hydrogen control Question Source:             Bank #               _       _
Modified Bank #
Modified Bank #       _______
_______  New   ___X___ Question History:
New                   ___X___
Last NRC Exam
Question History:             Last NRC Exam         _       _
_ _ Question Cognitive Level:
Question Cognitive Level:     Memory or Fundamental Knowledge             __ __
Memory or Fundamental Knowledge
Comprehension or Analysis                   __ X __
__ __ Comprehension or Analysis   __ X __ 10 CFR Part 55 Content:
10 CFR Part 55 Content:       55.41 _(7)__
55.41 _(7)__ Comments: DIF 4     50 51 52 53 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
Comments:
-
DIF     4 50
 
51 52 53 ES-401                             Sample Written Examination                     Form ES-401-5 Question Worksheet Examination Outline Cross-


==Reference:==
==Reference:==
 
Level                 RO           SRO Tier #                 __2__         _____
Level   RO SRO Tier #     __2__ _____ Group # __1__ _____ K/A #   _203000 A3.03
Group #               __1__         _____
____ Importance Rating
K/A #                 _203000 A3.03____
_3.7_ _____ 203000 RHR/LPCI: Injection Mode  
Importance Rating     _3.7_         _____
- Ability to monitor automatic operation of the RHR/LPCI: injection mode (plant specific) including: (CFR: 41.7) A3.03 Pump discharge pressure Q ues ti on: 2 8  W it h t h e p l an t ope r a ti n g a t r a t e d po w e r a LOC A occu r s. Th e f o ll o w i n g conditi on s ex i s t: R eac t o r p ressu r e i s 60 0 ps i g an d l o w e ri ng. D r y w e l l p ressu r e i s 5.2 ps i g an d ri s i n g s l o w l y. To r u s p ressu r e i s 4.0 ps i g an d ri s i n g s l o w l y. When does L PC I injection fl o w i n t o t h e RP V FI RS T occur for the listed pressures below
203000 RHR/LPCI: Injection Mode - Ability to monitor automatic operation of the RHR/LPCI: injection mode (plant specific) including: (CFR: 41.7) A3.03 Pump discharge pressure Question: 28 With the plant operating at rated power a LOCA occurs.
?  A. 55 0 ps i g  B. 43 5 ps i g  C. 230 ps i g  D. 105 ps i g  A ns w e r: C. 230 ps i g  Explanation:
The following conditions exist:
With drywell pressure greater than 1.84 psig a LPCI initiation signal is present. The RHR pumps receive a start signal and operate on minimum flow. As reactor pressure continues to lower the RHR inboard injection valves open when reactor pressure reaches 436 psig. Flow however does not occur until reactor pressure falls below the shutoff head of the RHR pumps at 230 psig. At or below this pressure indications of flow from the RHR would first occur. It should be noted that Plant Condition Assessment (PCA
* Reactor pressure is 600 psig and lowering.
-1) chart indicates that RHR Pump shutoff head is ~300 psig, thus making 230 psig the only plausible choice.
* Drywell pressure is 5.2 psig and rising slowly.
 
* Torus pressure is 4.0 psig and rising slowly.
When does LPCI injection flow into the RPV FIRST occur for the listed pressures below?
A. 550 psig B. 435 psig C. 230 psig D. 105 psig Answer:
C. 230 psig Explanation:
With drywell pressure greater than 1.84 psig a LPCI initiation signal is present. The RHR pumps receive a start signal and operate on minimum flow. As reactor pressure continues to lower the RHR inboard injection valves open when reactor pressure reaches 436 psig. Flow however does not occur until reactor pressure falls below the shutoff head of the RHR pumps at 230 psig. At or below this pressure indications of flow from the RHR would first occur. It should be noted that Plant Condition Assessment (PCA-1) chart indicates that RHR Pump shutoff head is ~300 psig, thus making 230 psig the only plausible choice.
Distracters:
Distracters:
A. This option is incorrect because, with reactor pressure at 550 psig, the RHR pumps are not injecting, as this pressure is greater than the shutoff head of the RHR pumps. As reactor 54


A. This option is incorrect because, with reactor pressure at 550 psig, the RHR pumps are not injecting, as this pressure is greater than the shutoff head of the RHR pumps. As reactor 54 pressure lowers to this value, there would be condensate and condensate booster pump flow if they are operating. A candidate who has seen these conditions during training in the in the simulator may confuse the where the source of injection came from and choose this option.
pressure lowers to this value, there would be condensate and condensate booster pump flow if they are operating. A candidate who has seen these conditions during training in the in the simulator may confuse the where the source of injection came from and choose this option.
B. This option is incorrect because, with reactor pressure at 435 psig, the RHR pumps are not injecting, as this pressure is greater than the shutoff head of the RHR pumps. The RHR Inboard injection valves however are interlocked to open at this pressure establishing a flow path from the RHR pumps to the reactor vessel. A candidate may choose this answer however believing that flow begins when the inboard injection valve opens.
B. This option is incorrect because, with reactor pressure at 435 psig, the RHR pumps are not injecting, as this pressure is greater than the shutoff head of the RHR pumps. The RHR Inboard injection valves however are interlocked to open at this pressure establishing a flow path from the RHR pumps to the reactor vessel. A candidate may choose this answer however believing that flow begins when the inboard injection valve opens.
D. This option is incorrect because with reactor pressure at 105 psig this is not the FIRST pressure at which the RHR pumps are injecting. This is the pressure where pressure main condensate pumps (only) inject. This value is plausible because the candidate may recall the condensate pump injection pressure and confuse this with the discharge pressure where RHR injects.
D. This option is incorrect because with reactor pressure at 105 psig this is not the FIRST pressure at which the RHR pumps are injecting. This is the pressure where pressure main condensate pumps (only) inject. This value is plausible because the candidate may recall the condensate pump injection pressure and confuse this with the discharge pressure where RHR injects.
Technical Reference(s):
Technical Reference(s):                 USAR Section VI, Table VI-5-4. Plant ECCS Parameters (Attach if not previously provided)     5.9SAMG Attachment 2 Plant Condition Assessment 1 R7 (including version/revision number)
USAR Section VI, Table VI 4. Plant ECCS Parameters (Attach if not previously provided) 5.9SAMG Attachment 2 Plant Condition Assessment 1 R7 (including version/revision number)
Proposed references to be provided to applicants during examination:       ___None     _______
Proposed references to be provided to applicants during examination:
Learning Objective:
___None
_______  Learning Objective:
COR002-23-02, OPS Residual Heat Removal System
COR002-23-02, OPS Residual Heat Removal System
: 4. Describe the interrelationship between the RHR system and the following:
: 4. Describe the interrelationship between the RHR system and the following:
: n. Reactor pressure Question Source:
: n.     Reactor pressure Question Source:               Bank #                 _______
Bank #   _______ Modified Bank #
Modified Bank #       _______
_______  New   ___X___ Question History:
New                   ___X___
Last NRC Exam
Question History:               Last NRC Exam         ____________
____________
Question Cognitive Level:       Memory or Fundamental Knowledge             __ __
Question Cognitive Level:
Comprehension or Analysis                   ___X__
Memory or Fundamental Knowledge
10 CFR Part 55 Content:         55.41 _(7)_
__   __ Comprehension or Analysis   ___X__ 10 CFR Part 55 Content:
Comments:
55.41 _(7)_ Comments:
LOD 3 55
LOD 3     55 56 57 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
 
-
56 57 ES-401                           Sample Written Examination                 Form ES-401-5 Question Worksheet Examination Outline Cross-


==Reference:==
==Reference:==
Level                RO            SRO Tier #              __2__        _____
Group #              __1__        _____
K/A #                _205000 A.4.07 ___
Importance Rating    _3.7_        _____
205000Shutdown Cooling - Ability to manually operate and/or monitor in the control room: (CFR: 41.7) A4.07 Reactor temperatures (moderator, vessel, flange)
Question: 29 Residual Heat Removal (RHR) loop A is aligned for shutdown cooling mode of operation with RHR pump A operating. The following conditions exist:
* The vessel head is tensioned.
* Reactor Recirculation pump B is operating.
* SDC system flow is 5000 gpm.
* Temperatures are logged as follows:
RHR-TR-131, CH 9 NBI-TR-89, CH 9 (reactor NBI-TR-89, CH6 (reactor TIME (RHR HX inlet temp)    vessel metal temp)      vessel flange temp) 0100            210&#xba;F                  222&#xba;F                    332&#xba;F 0115            188&#xba;F                  196&#xba;F                    331&#xba;F 0130            166&#xba;F                  173&#xba;F                    331&#xba;F 0145            143&#xba;F                  147&#xba;F                    330&#xba;F What RHR action is required and why?
A. Throttle CLOSED RHR-MO-27A, Inboard Injection Valve, to reduce the cooldown rate.
B. Throttle OPEN RHR-MO-66, RHR Heat Exchanger Bypass Valve, to reduce the cooldown rate.
C. Throttle CLOSED RHR-MO-27A, Inboard Injection Valve, to ensure accurate temperature indication at the RHR HX Inlet.
D. Throttle OPEN RHR-MO-66, RHR Heat Exchanger Bypass Valve, to ensure accurate temperature indication at the RHR HX Inlet.
Answer:
B. Throttle OPEN RHR-MO-66, RHR Heat Exchanger Bypass Valve, to reduce the cooldown rate.
58


Level  RO  SRO Tier #    __2__  _____ Group #  __1__  _____ K/A #  _205000  A.4.07
Explanation:
___ Importance Rating
When the RHR system is placed in operation for SDC the heat up and cooldown rates are adjusted to average heatup/cooldown rate  90&deg;F/hr averaged over any 1 hour period. RHR-MO-27A, RHR-MO-66A and RHR-MO-12A are the valves that are procedurally adjusted to manipulate cooldown rate. From the data provided, if the cooldown rate is continued at the current rate, the 90&deg;F/hr administrative limit for cooldown will be exceeded. The cooldown rate for the reactor vessel metal temperature is currently excessive and if the current rate continues the cooldown rate will exceed the limit.
_3.7_  _____  205000Shutdown Cooling
Opening RHR-MO-66A will reduce the cooldown rate. Although both RHR-MO-12 and 27 could be closed to reduce cooldown the SDC flow is already low and cannot be lowered to less than 5000 gpm and closing either of these valves would reduce SDC flow.
- Ability to manually operate and/or monitor in the control room: (CFR: 41.7) A4.07 Reactor temperatures (moderator, vessel, flange)
NBI-TR-89, CH 6 is measured at the vessel flange which is in the air space above the coolant level. The RTD is on the outer flange surface so the flange temperature rate change does not directly follow the coolant rate of change. If the coolant rate of change raised by 10&deg;F one hour, the flange rate of change would most likely not change. (NOTE: this data was gathered from actual plant parameters)
Question:  29
There is a NOTE in Procedure 2.2.69.2 that alerts the operator that vessel metal temperatures changing at rates different than reactor coolant rates may indicate core flow is too low and the temperatures may not be as accurate as needed and that coolant temperature may be higher than indicated. The options C and D A. This option is incorrect because even though the cooldown rate is excessive, the conditions given have SDC flow at 5000 gpm and so throttling closed either RHR-MO-27A or 12A would reduce flow rate and would therefore be inappropriate with flow at the low limit. The candidate who fails to completely evaluate all the provided conditions may choose this option because this is a procedural method for reducing the cooldown rate.
 
C. This option is incorrect because the cooldown rate is excessive. The temperatures are trending together correctly so the change needed is to reduce the cooldown rate.
Residual Heat Removal (RHR) loop A is aligned for shutdown cooling mode of operation with RHR pump A operating. The following conditions exist:
Surveillance Procedure 6.RCS.601 requires monitoring of various temperatures during the cooldown to ensure proper trending so it a candidate may believe that the trends are not correct and that a SDC manipulation is required for accurate indication. MO-27 is a valve that can be manipulated to adjust SDC flow. However since flow is already at 5000 gpm this action is inappropriate because it would cause flow to go below 5000 gpm.
The vessel head is tensioned.
D. This option is incorrect because the cooldown rate is excessive which requires action.
Reactor Recirculation pump B is operating.
Surveillance Procedure 6.RCS.601 requires monitoring of various temperatures during the cooldown and if the temperatures are not trending together then it requires the manipulation of MO-66 to ensure that the water temperature measured is accurate. Because this is a possible manipulation that may be required, a candidate may believe that these values are not trending correctly and choose this answer.
SDC system flow is 5000 gpm. Temperatures are logged as follows:
TIME RHR-TR-131, CH 9 (RHR HX inlet temp)
NBI-TR-89, CH 9 (reactor vessel metal temp)
NBI-TR-89, CH6 (reactor vessel flange temp) 0100 210&#xba;F 222&#xba;F 332&#xba;F 0115 188&#xba;F 196&#xba;F 331&#xba;F 0130 166&#xba;F 173&#xba;F 331&#xba;F 0145 143&#xba;F 147&#xba;F 330&#xba;F What RHR action is required and why?
 
A. Throttle CLOSED RHR
-MO-27A, Inboard Injection Valve, to reduce the cooldown rate.
B. Throttle OPEN RHR
-MO-66, RHR Heat Exchanger Bypass Valve, to reduce the cooldown rate. C. Throttle CLOSED RHR-MO-27A, Inboard Injection Valve, to ensure accurate temperature indication at the RHR HX Inlet.
 
D. Throttle OPEN RHR
-MO-66, RHR Heat Exchanger Bypass Valve, to ensure accurate temperature indication at the RHR HX Inlet.
Answer: B. Throttle OPEN RHR
-MO-66, RHR Heat Exchanger Bypass Valve, to reduce the cooldown rate. 58 Explanation:
When the RHR system is placed in operation for SDC the heat up and cooldown rates are adjusted to average heatup/cooldown rate  90F/hr averaged over any 1 hour period. R HR-MO-27A, RHR-MO-66A and RHR
-MO-12A are the valves that are procedurally adjusted to manipulate cooldown rate. From the data provided, if the cooldown rate is continued at the current rate, the 90F/hr administrative limit for cooldown will be exceeded.
The cooldown rate for the reactor vessel metal temperature is currently excessive and if the current rate continues the cooldown rate will exceed the limit.
Opening RHR
-MO-66A will reduce the cooldown rate. Although both RHR
-MO-12 and 27 could be closed to reduce cooldown the SDC flow is already low and cannot be lowered to less than 5000 gpm and closing either of these valves would reduce SDC flow.
 
NBI-TR-89, CH 6 is measured at the vessel flange which is in the air space above the coolant level. The RTD is on the outer flange surface so the flange temperature rate change does not directly follow the coolant rate of change. If the coolant rate of change raised by 10 F one hour, the flange rate of change would most likely not change. (NOTE: this data was gathered from actual plant parameters)
There is a NOTE in Procedure 2.2.69.2 that alerts the operator that vessel metal temperatures changing at rates different than reactor coolant rates may indicate core flow is too low and the temperatures may not be as accurate as needed and that coolant temperature may be higher than indicated. The options C and D A. This option is incorrect because even though the cooldown rate is excessive, the conditions given have SDC flow at 5000 gpm and so throttling closed either RHR
-MO-27A or 12A would reduce flow rate and would therefore be inappropriate with flow at the low limit. The candidate who fails to completely evaluate all the provided conditions may choose this option because this is a procedural method for reducing the cooldown rate.
C. This option is incorrect because the cooldown rate is excessive. The temperatures are trending together correctly so the change needed is to reduce the cooldown rate. Surveillance Procedure 6.RCS.601 requires monitoring of various temperatures during the cooldown to ensure proper trending so it a candidate may believe that the trends are not correct and that a SDC manipulation is required for accurate indication. MO
-27 is a valve that can be manipulated to adjust SDC flow. However since flow is already at 5000 gpm this action is inappropriate because it would cause flow to go below 5000 gpm.  
 
D. This option is incorrect because the cooldown rate is excessive which requires action. Surveillance Procedure 6.RCS.601 requires monitoring of various temperatures during the cooldown and if the temperatures are not trending together then it requires the manipulation of MO-66 to ensure that the water temperature measured is accurate. Because this is a possible manipulation that may be required, a candidate may believe that these values are not trending correctly and choose this answer.
 
Technical Reference(s):
Technical Reference(s):
SP 6.RCS.601, RCS Heatup/Cooldown Rate Monitoring, Rev 21 SOP 2.2.69.2, RHR System Shutdown Operations, Rev 89 Proposed references to be provided to applicants during examination:
SP 6.RCS.601, RCS Heatup/Cooldown Rate Monitoring, Rev 21 SOP 2.2.69.2, RHR System Shutdown Operations, Rev 89 Proposed references to be provided to applicants during examination:         __None_________
__None_________
59
59 Learning Objective:


Per COR002 02 6. Given an RHR control manipulation, predict and explain changes in the following:
Learning Objective:
Per COR002-23-02
: 6. Given an RHR control manipulation, predict and explain changes in the following:
: a. Heat exchanger temperature and flow
: a. Heat exchanger temperature and flow
: d. Reactor parameters (level, pressure, temperature)
: d. Reactor parameters (level, pressure, temperature)
Question Source:
Question Source:           Bank #                 _______
Bank #   _______ Modified Bank #
Modified Bank #       _______
_______  New   ___X___ Question History:
New                   ___X___
Last NRC Exam
Question History:         Last NRC Exam         ____________
____________  Question Cognitive Level:
Question Cognitive Level: Memory or Fundamental Knowledge             __ __
Memory or Fundamental Knowledge
Comprehension or Analysis                   __X__
__ __ Comprehension or Analysis   __X__ 10 CFR Part 55 Content:
10 CFR Part 55 Content:   55.41 _(7)_
55.41 _(7)_ Comments: LOD 3 60 PRESSMAINTTo 'B'LOOP RHR(Loop Crosstie)(HX Outlet) 2012A LT HX A TERHR SW(A Only)FI66A(HX Bypass)RHR SW TE65A(HX Inlet)
Comments:
PS PS PS PSCST13C13A(Torus)15C15AFPC 17To 'B'LOOP RHR(SDCSuction)RR'A'18From RxTo Rx 274A16A34A39A38A25A27A(TorusCooling)(TorusSpray)(LPCI)31A26A(DrywellSpray)26CV(Drywell)RHR LOOP 'A'Figure 1, Rev. 9COR002-23-02CXA05157f:\home\jyknapp\figures\cxa05157\co022302.fig\fig1.r0981A 61 62 63 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
LOD 3 60
-
 
31A    (Drywell    26A Spray)                                                                        20                                      61 PRESS                                                            To 'B' MAINT                                                          LOOP RHR (Drywell)                                                                                                          (Loop Crosstie) 25A                27A 26CV (LPCI) 38A      (Torus    39A 81A              274A Spray)                                            FI (HX Outlet) 12A From Rx                            To Rx (Torus                                                                                              HX 34A                                                                                        LT Cooling) 16A                                          TE A
RHR SW (A Only) 18 RR'A' 66A  (HX Bypass)
(Torus)                   PS  PS 17 13A FPC 15A (SDC                                                                                              PS   PS Suction)                                                                                                                                           RHR SW CST 13C To 'B' LOOP                               15C                                                                                          65A RHR (HX Inlet)
TE RHR LOOP 'A' Figure 1, Rev. 9 f:\home\jyknapp\figures\cxa05157\co022302.fig\fig1.r09 COR002-23-02                                                                              CXA05157
 
62 63 ES-401                           Sample Written Examination                       Form ES-401-5 Question Worksheet Examination Outline Cross-


==Reference:==
==Reference:==
 
Level                   RO           SRO Tier #                 __2__         _____
Level   RO SRO Tier #     __2__ _____ Group # __1__ _____ K/A #   _ 206000 2.2.39 _
Group #                 __1__         _____
__ Importance Rating
K/A #                   _ 206000 2.2.39 ___
_ 3.9 _ _____ 206000 HPCI  
Importance Rating       _ 3.9 _       _____
- 2.2.39 Knowledge of less than or equal to one hour Technical Specification action statements for systems. (CFR: 41.7)
206000 HPCI - 2.2.39 Knowledge of less than or equal to one hour Technical Specification action statements for systems. (CFR: 41.7)
Q ues ti on: 3 0  Th e p l an t i s ope r a ti n g a t 10% o f r a t e d po w e r. HPCI-MO-15, STM SUPP INBD ISOL VLV is found closed and cannot be opened with its control switch.
Question: 30 The plant is operating at 10% of rated power.
HPCI-MO-15, STM SUPP INBD ISOL VLV is found closed and cannot be opened with its control switch.
HPCI is declared Inoperable.
HPCI is declared Inoperable.
W ha t Technical Specification ACTION i s requ ir e d? A. E n t e r Techn i ca l S pecifi ca ti o n 3.0.3 i mm ed i a t e l y. B. V erif y a l l AD S SRV s a r e O pe r ab l e w it h i n 1 hou r.
What Technical Specification ACTION is required?
C. V erif y RC I C sys t e m i s O pe r ab le by administrative means w it h i n 1 hou r. D. I so l a t e HPC I b y deac ti va ti n g HPC I-MO-15 and HPCI
A. Enter Technical Specification 3.0.3 immediately.
-MO-16 w it h i n 1 hou r. A ns w e r: C. V erif y RC I C sys t e m i s O pe r ab le by administrative means w it h i n 1 hou r. Explanation:
B. Verify all ADS SRVs are Operable within 1 hour.
Per TS 3.5.1, CONDITION C; If the HPCI System is inoperable and the RCIC System is verified to be OPERABLE, the HPCI System must be restored to OPERABLE status within 14 days. In this condition, adequate core cooling is ensured by the OPERABILITY of the redundant and diverse low
C. Verify RCIC system is Operable by administrative means within 1 hour.
-pressure ECCS injection/spray subsystems in conjunction with ADS. Also, the RCIC System will automatically provide makeup water at most reactor operating pressures. Verification of RCIC OPERABILITY within 1 hour is therefore required when HPCI is inoperable. RCIC is required to be determined to be operable by administrative means within one hour.
D. Isolate HPCI by deactivating HPCI-MO-15 and HPCI-MO-16 within 1 hour.
Answer:
C. Verify RCIC system is Operable by administrative means within 1 hour.
Explanation:
Per TS 3.5.1, CONDITION C; If the HPCI System is inoperable and the RCIC System is verified to be OPERABLE, the HPCI System must be restored to OPERABLE status within 14 days. In this condition, adequate core cooling is ensured by the OPERABILITY of the redundant and diverse low-pressure ECCS injection/spray subsystems in conjunction with ADS. Also, the RCIC System will automatically provide makeup water at most reactor operating pressures.
Verification of RCIC OPERABILITY within 1 hour is therefore required when HPCI is inoperable.
RCIC is required to be determined to be operable by administrative means within one hour.
Distracters:
Distracters:
A. This option is incorrect because entry into 3.0.3 is not required because only one ECCS system is inoperable. If two were inoperable then this option would be correct. The candidate may remember that there is an immediate entry into 3.0.3 required for an inoperable ECCS. Since it is a less than 1
A. This option is incorrect because entry into 3.0.3 is not required because only one ECCS system is inoperable. If two were inoperable then this option would be correct. The candidate may remember that there is an immediate entry into 3.0.3 required for an inoperable ECCS. Since it is a less than 1-hour specification associated with HPCI a candidate may choose this option.
-hour specification associated with HPCI a candidate may choose this option.
64
64 B. This option is incorrect because verifying ADS operability is not required by this specification. Since functionally ADS provides a backup to the HPCI system (in conjunction with low pressure ECCS) and because RCIC is not an ECCS system, a candidate may believe that verifying that ADS is operable is the required action and would therefore choose this option. This option is a common misconception.
 
B. This option is incorrect because verifying ADS operability is not required by this specification. Since functionally ADS provides a backup to the HPCI system (in conjunction with low pressure ECCS) and because RCIC is not an ECCS system, a candidate may believe that verifying that ADS is operable is the required action and would therefore choose this option. This option is a common misconception.
D. This option is incorrect because this action is not required. The actions for a primary containment isolation valve inoperable are contained in 3.6.1.3. But since only one valve is inoperable and it is closed this specification does not apply. However this would be the action required if both HPCI isolation valves had failed which is why a candidate may choose this answer.
D. This option is incorrect because this action is not required. The actions for a primary containment isolation valve inoperable are contained in 3.6.1.3. But since only one valve is inoperable and it is closed this specification does not apply. However this would be the action required if both HPCI isolation valves had failed which is why a candidate may choose this answer.
 
Technical Reference(s):                 Technical Specification 3.5.1, ECCS-Operating ________
Technical Reference(s):
(Attach if not previously provided)     SOP 2.2.33, High Pressure Coolant Injection System, Rev 77 Proposed references to be provided to applicants during examination:         __None___________
Technical Specification 3.5.1, ECCS
-Operating
________ (Attach if not previously provided)
SOP 2.2.33, High Pressure Coolant Injection System, Rev 77       Proposed references to be provided to applicants during examination:
__None___________
Learning Objective:
Learning Objective:
INT007-05-06, OPS Tech Specs 3.5, Emergency Core Cooling systems (ECCS) and Reactor Core Isolation Cooling (RCIC) System
INT007-05-06, OPS Tech Specs 3.5, Emergency Core Cooling systems (ECCS) and Reactor Core Isolation Cooling (RCIC) System
: 3. Given a set of plant conditions that constitutes non
: 3.       Given a set of plant conditions that constitutes non-compliance with a Section 3.5 LCO, determine the Actions that are required.
-compliance with a Section 3.5 LCO, determine the Actions that are required.
: 4.       From memory, in MODES 1, 2, and 3, state the actions required in  one hour if HPCI System is inoperable or two or more low pressure ECCS injection/spray subsystems inoperable or HPCI System and one or more ADS valves are inoperable (LCO 3.5.1).
: 4. From memory, in MODES 1, 2, and 3, state the actions or two or more low pressure ECCS injection/spray subsystems inoperable or HPCI System and one or more ADS valves are inoperable (LCO 3.5.1).
Question Source:               Bank #                 _______
Question Source:
Modified Bank #         _______
Bank #   _______ Modified Bank #
New                     ___X___
_______  New   ___X___ Question History:
Question History:               Last NRC Exam           ____________
Last NRC Exam
Question Cognitive Level:      Memory or Fundamental Knowledge              _ X__
____________
Comprehension or Analysis                    __ __
10 CFR Part 55 Content:        55.41 _10_
Comments:
LOD 3 65


Question Cognitive Level:
66 ES-401                               Sample Written Examination                         Form ES-401-5 Question Worksheet Examination Outline Cross-
Memory or Fundamental Knowledge
_ X__ Comprehension or Analysis  __  __
10 CFR Part 55 Content:
55.41 _10_
Comments: LOD 3  65 66 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
-


==Reference:==
==Reference:==
 
Level                   RO           SRO Tier #                 __2__         _____
Level   RO SRO Tier #     __2__ _____ Group # __1__ _____ K/A #   _ 209001 K1.01_
Group #                 __1__         _____
__ Importance Rating
K/A #                   _ 209001 K1.01___
_3.1_ _____ 209001 LPCS  
Importance Rating       _3.1_         _____
- Knowledge of the physical connection and/or cause
209001 LPCS - Knowledge of the physical connection and/or cause-effect relationships between low pressure core spray and the following: (CFR: 41.2 - 41.9) K1.01 Condensate storage tank: Plant-Specific Question: 31 What is the alternate suction source for Core Spray Pump A and when is it aligned to this source?
-effect relationships between low pressure core spray and the following: (CFR: 41.2  
A. ECST, to raise torus level in Modes 1/2/3.
- 41.9) K1.01 Condensate storage tank: Plant-Specific Question: 31 What is the alternate suction source for Core Spray Pump A and when is it aligned to this source? A. ECST, to raise torus level in Modes 1/2/3.  
 
B. CST A, to raise torus level in Modes 1/2/3.
B. CST A, to raise torus level in Modes 1/2/3.
C. ECST, for core reflood capability when the torus is drained in Modes 4/5.
C. ECST, for core reflood capability when the torus is drained in Modes 4/5.
D. CST A, for core reflood capability when the torus is drained in Modes 4/5.
D. CST A, for core reflood capability when the torus is drained in Modes 4/5.
 
Answer:
Answer: D. CST A, for core reflood capability when the torus is drained in Modes 4/5.  
D. CST A, for core reflood capability when the torus is drained in Modes 4/5.
 
Explanation:
Explanation:
Per USAR Chapter VI Section 4 the CST provides an alternate suction source to CS. The suction to the CS pumps can also be lined up to Condensate Storage Tank (CST) 1A. CNS Technical Specifications allow refueling operations to be conducted with the suppression pool drained provided an operable CS or LPCI subsystem is aligned to take a suction on CST 1A, containing at least 150,000 gallons. In this condition, the reactor vessel is depressurized and the CS subsystem provides core reflooding capability.
Per USAR Chapter VI Section 4 the CST provides an alternate suction source to CS. The suction to the CS pumps can also be lined up to Condensate Storage Tank (CST) 1A.
Distracters: A. This option is incorrect because CS A is not capable of being aligned to the ECST. This is a plausible selection since other core cooling systems (HPCI and RCIC) may be aligned to the ECST, which is why a candidate may believe that CS is capable of this alignment. Additionally, the purpose of the alternate source for the CS is to provide for core reflood when in modes 4 and 5. But under certain conditions, the Core Spray system can be used to fill the torus (with pressure maintenance) so a candidate may choose this option.
CNS Technical Specifications allow refueling operations to be conducted with the suppression pool drained provided an operable CS or LPCI subsystem is aligned to take a suction on CST 1A, containing at least 150,000 gallons. In this condition, the reactor vessel is depressurized and the CS subsystem provides core reflooding capability.
Distracters:
A. This option is incorrect because CS A is not capable of being aligned to the ECST. This is a plausible selection since other core cooling systems (HPCI and RCIC) may be aligned to the ECST, which is why a candidate may believe that CS is capable of this alignment.
Additionally, the purpose of the alternate source for the CS is to provide for core reflood when in modes 4 and 5. But under certain conditions, the Core Spray system can be used to fill the torus (with pressure maintenance) so a candidate may choose this option.
B. This option is incorrect because the purpose of the alternate suction source for CS A is not to provide the capability to fill the torus it is to allow reflood in modes 4 and 5 when the torus is not available. This selection is plausible since the torus can be filled with the CS system (from pressure maintenance not the alternate suction path).
B. This option is incorrect because the purpose of the alternate suction source for CS A is not to provide the capability to fill the torus it is to allow reflood in modes 4 and 5 when the torus is not available. This selection is plausible since the torus can be filled with the CS system (from pressure maintenance not the alternate suction path).
67 C. This option is incorrect because the alternate suction source is from the CST not the ECST. Other core cooling systems (HPCI and RCIC) may be aligned to the ECST, which makes this a plausible selection if the candidate believes that CS is capable of this alignment.
67
Technical Reference(s):
 
SOP 2.2.9, Core Spray System, Rev 76
C. This option is incorrect because the alternate suction source is from the CST not the ECST.
____ (Attach if not previously provided)
Other core cooling systems (HPCI and RCIC) may be aligned to the ECST, which makes this a plausible selection if the candidate believes that CS is capable of this alignment.
COR002-06-02, Core Spray System, Rev 87
Technical Reference(s):                 SOP 2.2.9, Core Spray System, Rev 76                   ____
__ (including version/revision number)
(Attach if not previously provided)     COR002-06-02, Core Spray System, Rev 87                 __
__ USAR Chapter VI Section 4________ _____________
(including version/revision number) __ USAR Chapter VI Section 4________ _____________
Proposed references to be provided to applicants during examination:
Proposed references to be provided to applicants during examination:         __None       _______
__None
Learning Objective:
_______  Learning Objective:
COR002-06-02, Core Spray System
COR002-06-02, Core Spray System
: 3. Describe the interrelationships between the Core Spray and the following:
: 3.     Describe the interrelationships between the Core Spray and the following:
: a. Condensate Storage Tank Question Source:
: a.     Condensate Storage Tank Question Source:               Bank #                 _______
Bank #   _______ Modified Bank #
Modified Bank #       _______
_______  New   __ X __ Question History:
New                   __ X __
Last NRC Exam
Question History:               Last NRC Exam         ____________
____________
Question Cognitive Level:      Memory or Fundamental Knowledge              __X _
Comprehension or Analysis                    ____
10 CFR Part 55 Content:        55.41 8 Comments:
LOD 3 68


Question Cognitive Level:
69 70 71 ES-401                           Sample Written Examination                         Form ES-401-5 Question Worksheet Examination Outline Cross-
Memory or Fundamental Knowledge
__X _ Comprehension or Analysis  ____
10 CFR Part 55 Content:
55.41 8  Comments: LOD 3  68 69 70 71 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
-


==Reference:==
==Reference:==
 
Level                   RO             SRO Tier #                   __2__         _____
Level   RO SRO Tier #     __2__ _____ Group # __1__ _____ K/A #   _259002 K3.02
Group #                 __1__         _____
___ Importance Rating
K/A #                   _259002 K3.02 ___
_3.7 _ _____ 259002 Reactor Water Level Control
Importance Rating       _3.7 _         _____
- Knowledge of the effect that a loss or malfunction of the reactor water level control system will have on following: (CFR: 41.7 / 45.4) (CFR: 41.7 / 45.5 to 45.8)  
259002 Reactor Water Level Control - Knowledge of the effect that a loss or malfunction of the reactor water level control system will have on following: (CFR: 41.7 / 45.4) (CFR:
- K3.02 Reactor feedwater system Q ues ti on: 3 2  Th e p l an t i s ope r a ti n g a t nea r r a t e d po w e r w he n a l os s o f bo t h RV L C/R F P T COR E s w itche s causes RFPT control to transfer to MDEM.
41.7 / 45.5 to 45.8) - K3.02 Reactor feedwater system Question: 32 The plant is operating at near rated power when a loss of both RVLC/RFPT CORE switches causes RFPT control to transfer to MDEM.
How are the RFPs affected?
How are the RFPs affected?
A. Speed lowers to idle speed.
A. Speed lowers to idle speed.
B. Speed rises to upper automatic clamp.
B. Speed rises to upper automatic clamp.
C. Speed is held constant at current speed.
C. Speed is held constant at current speed.
D. Speed lowers to minimum governor speed.
D. Speed lowers to minimum governor speed.
Answer: C. Speed is held constant at current speed.
Answer:
C. Speed is held constant at current speed.
Explanation:
Explanation:
The loss of RVLC/RFPT CORE switches, which are part of the Reactor Vessel Level Control System, result in the RFPs transferring to MDEM. With the loss of the core switches, there is no control available from HMIs. Since the RFPs transfer to MDEM the RFP speed is held constant and no longer modulates to control level.
The loss of RVLC/RFPT CORE switches, which are part of the Reactor Vessel Level Control System, result in the RFPs transferring to MDEM. With the loss of the core switches, there is no control available from HMIs. Since the RFPs transfer to MDEM the RFP speed is held constant and no longer modulates to control level.
Distracters:
Distracters:
A. This option is incorrect because, with the failure of the core switches the controllers transfer to MDEM and now instead of RFP speed modulating it now locks at its current speed. A candidate could believe that with the loss of the switches that the output would be low as is the case with many analog controllers and that speed would therefore lower to idle speed (1000 RPM). This selection is plausible as idle speed is an operationally significant speed during a RFP start a candidate may choose this option.
A. This option is incorrect because, with the failure of the core switches the controllers transfer to MDEM and now instead of RFP speed modulating it now locks at its current speed. A candidate could believe that with the loss of the switches that the output would be low as is the case with many analog controllers and that speed would therefore lower to idle speed (1000 RPM). This selection is plausible as idle speed is an operationally significant speed during a RFP start a candidate may choose this option.
B. This option is incorrect because, when the RFPs transfer to MDEM, they do so at their current speed. But a candidate may believe that with the loss of the core switches RFPT speed would rise to the upper limit of 5800 RPM.
B. This option is incorrect because, when the RFPs transfer to MDEM, they do so at their current speed. But a candidate may believe that with the loss of the core switches RFPT speed would rise to the upper limit of 5800 RPM. This selection is plausible since some traditional control systems that suffer loss of input indicating a lowering or loss of speed do 72
This selection is plausible since some traditional control systems that suffer loss of input indicating a lowering or loss of speed do 72 go to their maximum values so a candidate who does not understand this system may choose this option.
 
D. This option is incorrect because, with the failure of the core switches the controllers transfer to MDEM and now instead of RFP speed modulating it now locks at its current speed. A candidate could believe that with the loss of the switches that the output would be low as is the case with many analog controllers and that speed would therefore lower to the minimum clamp on the governor (2000 RPM). This is a different speed than the idle choice listed in option "A" so the candidate who believes that the speed control functions as do many analog controllers may choose this options lending plausibility for this selection as the minimum governor speed is an operationally significant value.
Technical Reference(s):
Instrument Procedure 4.4.1, Reactor Vessel Level Control, Rev 7      COR002-32-02 Reactor Vessel Level Control, Rev 14
 
Proposed references to be provided to applicants during examination:
__None__________


go to their maximum values so a candidate who does not understand this system may choose this option.
D. This option is incorrect because, with the failure of the core switches the controllers transfer to MDEM and now instead of RFP speed modulating it now locks at its current speed. A candidate could believe that with the loss of the switches that the output would be low as is the case with many analog controllers and that speed would therefore lower to the minimum clamp on the governor (2000 RPM). This is a different speed than the idle choice listed in option A so the candidate who believes that the speed control functions as do many analog controllers may choose this options lending plausibility for this selection as the minimum governor speed is an operationally significant value.
Technical Reference(s):      Instrument Procedure 4.4.1, Reactor Vessel Level Control, Rev 7 COR002-32-02 Reactor Vessel Level Control, Rev 14 Proposed references to be provided to applicants during examination:        __None__________
Learning Objective:
Learning Objective:
COR002-32-02 Reactor Vessel Level Control
COR002-32-02 Reactor Vessel Level Control
: 9. Given a specific RVLC system malfunction, determine the effect on any of the following:
: 9. Given a specific RVLC system malfunction, determine the effect on any of the following:
: c. Feedwater System Question Source:
: c. Feedwater System Question Source:             Bank #                 _      _
Bank #   _
Modified Bank #         _______
_ Modified Bank #
New                     __X ___
_______  New   __X ___ Question History:
Question History:             Last NRC Exam           ___ ______
Last NRC Exam
Question Cognitive Level:     Memory or Fundamental Knowledge               __X__
___ ______ Question Cognitive Level:
Comprehension or Analysis                     ___
Memory or Fundamental Knowledge
10 CFR Part 55 Content:       55.41 (7)
__X__ Comprehension or Analysis   ___ 10 CFR Part 55 Content:
Comments:
55.41 (7) Comments: LOD: 3 73 74 75 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
LOD: 3 73
-
 
74 75 ES-401                           Sample Written Examination                       Form ES-401-5 Question Worksheet Examination Outline Cross-


==Reference:==
==Reference:==
 
Level                   RO             SRO Tier #                 __2__         _____
Level   RO SRO Tier #     __2__ _____ Group # __1__ _____ K/A #     211000   K2.02
Group #                 __1__         _____
_ Importance Rating
K/A #                     211000   K2.02 _
_3.1 _ _____ 211000 SLC  
Importance Rating       _3.1 _         _____
- Knowledge of electrical power supplies to the following: (CR: 41.7) K2.02 Explosive valves Question: 33
211000 SLC - Knowledge of electrical power supplies to the following: (CR: 41.7) K2.02 Explosive valves Question: 33 What is the power supply that is used to fire the A Standby Liquid Control (SLC) squib valve?
 
A. MCC K B. MCC S C. MCC M D. CPP Answer:
What is the power supply that is used to fire the "A" Standby Liquid Control (SLC) squib valve?
A. MCC K Explanation:
A. MCC K B. MCC S C. MCC M D. CPP Answer:   A. MCC K Explanation:
MCC K is the power supply to the A SLC pump and the squib valve receives its power from the pump supply breaker.
MCC K is the power supply to the A SLC pump and the squib valve receives its power from the pump supply breaker.
Distracters:
Distracters:
B. This option is incorrect as this is the power supply to the B SLC pump and the B squib valve. A candidate may confuse which power supply is associated with which pump and may therefore choose this option. This answer is plausible because this power supply does power a SLC squib valve.
B. This option is incorrect as this is the power supply to the B SLC pump and the B squib valve.
A candidate may confuse which power supply is associated with which pump and may therefore choose this option. This answer is plausible because this power supply does power a SLC squib valve.
C. This option is incorrect because this is the listed power supply for SLC heat tracing. A candidate who knows SLC loads are powered from MCC M but is not certain of the squib valve power supply may choose this answer because of its association with the SLC system.
This answer is plausible because this power supply does power components in the SLC system.
D. This option is incorrect because this is the power supply for the squib valve ready lights. A candidate who does not recall that the squib valves are fired by an auxiliary contact in the pump breaker may believe this circuit also powers the squib valves since it does provide power to a squib related component (squib ready lights). Note: that even though this power supply appears different than the other options it is highly plausible because it powers actual squib components just not to power to fire the squib.
76


C. This option is incorrect because this is the listed power supply for SLC heat tracing. A candidate who knows SLC loads are powered from MCC M but is not certain of the squib valve power supply may choose this answer because of its association with the SLC system. This answer is plausible because this power supply does power components in the SLC system.
Technical Reference(s):
D. This option is incorrect because this is the power supply for the squib valve ready lights. A candidate who does not recall that the squib valves are fired by an auxiliary contact in the pump breaker may believe this circuit also powers the squib valves since it does provide power to a squib related component (squib ready lights). Note: that even though this power supply appears different than the other options it is highly plausible because it powers actual squib components just not to power to fire the squib.
76 Technical Reference(s):
Procedure 2.2.74A (Standby Liquid Control System Component Checklist), Rev.
Procedure 2.2.74A (Standby Liquid Control System Component Checklist), Rev.
11 Procedure 7.2.25 {SLC System Explosive (Squib) Valve Trigger/Primer Chamber Assembly Replacement}, Rev. 19
11 Procedure 7.2.25 {SLC System Explosive (Squib) Valve Trigger/Primer Chamber Assembly Replacement}, Rev. 19 Proposed references to be provided to applicants during examination:      None Learning Objective:
COR0022902 R20
: 13.      State the electrical power supply to the following SLC components:
: b. Squib valves Question Source:                Bank #
Modified Bank # 32 on 2014 NRC Exam        (See attached)
New Question Cognitive Level:      Memory or Fundamental Knowledge              X Comprehension or Analysis 10 CFR Part 55 Content:        41.7 Difficulty: 2 77


Proposed references to be provided to applicants during examination:
78 79 80 81 ES-401                           Sample Written Examination                     Form ES-401-5 Question Worksheet Examination Outline Cross-
None Learning Objective:
COR0022902 R20
: 13. State the electrical power supply to the following SLC components:
: b. Squib valves Question Source:
Bank #    Modified Bank # 32 on 2014 NRC Exam (See attached)
New      Question Cognitive Level:
Memory or Fundamental Knowledge X  Comprehension or Analysis 10 CFR Part 55 Content:
41.7 Difficulty:  2 77 78 79 80 81 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
-


==Reference:==
==Reference:==
 
Level                 RO             SRO Tier #               __2__         _____
Level   RO SRO Tier #     __2__ _____ Group # __1__ _____ K/A #   _212000   K4.02   __
Group #               __1__         _____
Importance Rating
K/A #                 _212000 K4.02 __
_3.5_ _____ 212000 RPS  
Importance Rating     _3.5_         _____
- Knowledge of the reactor protection system design feature(s) and/or interlocks which provide for the following: (CFR: 41.7) K4.02 The prevention of a reactor SCRAM following a single component failure Question: 34 The reactor is operating at 100% rated power.
212000 RPS - Knowledge of the reactor protection system design feature(s) and/or interlocks which provide for the following: (CFR: 41.7) K4.02 The prevention of a reactor SCRAM following a single component failure Question: 34 The reactor is operating at 100% rated power.
 
The RPS MG Set A motor supply fused disconnect fuse blows.
The RPS MG Set A motor supply fused disconnect fuse blows.
  (1) What is the status of the RPS A Electrical Protection Assemblies (EPA)?
(1) What is the status of the RPS A Electrical Protection Assemblies (EPA)?
(2) What is the status of RPS?
(2) What is the status of RPS?
A. (1) Two (2) RPS A EPA Breakers opened.
A. (1) Two (2) RPS A EPA Breakers opened.
(2) Full reactor Scram. B. (1) Two (2) RPS A EPA Breakers opened.
(2) Full reactor Scram.
(2) 1/2 Scram ONLY on RPSPP1A.
B. (1) Two (2) RPS A EPA Breakers opened.
(2) 1/2 Scram ONLY on RPSPP1A.
C. (1) Four (4) RPS A EPA Breakers opened.
C. (1) Four (4) RPS A EPA Breakers opened.
(2) Full reactor Scram. D. (1) Four (4) RPS A EPA Breakers opened.
(2) Full reactor Scram.
(2) 1/2 Scram ONLY on RPSPP1A.
D. (1) Four (4) RPS A EPA Breakers opened.
 
(2) 1/2 Scram ONLY on RPSPP1A.
Answer: B. (1) Two (2) RPS A EPA Breakers opened.
Answer:
  (2) 1/2 Scram ONLY on RPSPP1A.
B. (1) Two (2) RPS A EPA Breakers opened.
(2) 1/2 Scram ONLY on RPSPP1A.
Explanation:
Explanation:
RPS MG Set A generator supplies RPSPP1A and only RPSPP1A becomes de
RPS MG Set A generator supplies RPSPP1A and only RPSPP1A becomes de-energized. RPS MG Set B supplies RPSPP1B. The MG sets supply power to their respective RPS power panel via a pair of EPA in series. The EPAs ensure a pure and consistent power source for the sensitive RPS instrumentation which are energized and fail safe in the de-energized state.
-energized. RPS MG Set B supplies RPSPP1B. The MG sets supply power to their respective RPS power panel via a pair of EPA in series. The EPAs ensure a pure and consistent power source for the sensitive RPS instrumentation which are energized and fail safe in the de
There is no automatic power transfer in the RPS system. Loss of power to the RPS MG Set causes two EPAs to trip on under-voltage and or under frequency which causes a half scram on the A side. Since MG set B remains energized a full scram does NOT occur.
-energized state. There is no automatic power transfer in the RPS system. Loss of power to the RPS MG Set causes two EPAs to trip on under
-voltage and or under frequency which causes a half scram on the A side. Since MG set B remains energized a full scram does NOT occur.
There are four EPA breakers for each RPS power supply. TWO series breakers for the MG Set and two series breakers for the alternate power supply. Only the two breakers associated with the degraded power supply will trip allowing power to be manually transferred without operation of any EPA Breakers.
There are four EPA breakers for each RPS power supply. TWO series breakers for the MG Set and two series breakers for the alternate power supply. Only the two breakers associated with the degraded power supply will trip allowing power to be manually transferred without operation of any EPA Breakers.
82 Distracters:
82
 
Distracters:
A. This option is incorrect because RPSPP1B remains energized. The candidate could choose this option if he/she did not correctly identify that only one side of RPS was impacted. This answer is plausible because there are some single components that cause a full scram vice a divisional partial scram and the number of EPAs that trip is correct. The correct combination of 2 EPA breakers opening (A1 & B1) will cause a full scram. The RPS Shorting Link Switches are in CLOSE during power operation thus inhibiting a full reactor scram from a single channel trip.
A. This option is incorrect because RPSPP1B remains energized. The candidate could choose this option if he/she did not correctly identify that only one side of RPS was impacted. This answer is plausible because there are some single components that cause a full scram vice a divisional partial scram and the number of EPAs that trip is correct. The correct combination of 2 EPA breakers opening (A1 & B1) will cause a full scram. The RPS Shorting Link Switches are in CLOSE during power operation thus inhibiting a full reactor scram from a single channel trip.
C. This option is incorrect because only the two breakers associated with the MG Set will trip because RPSPP1B remains energized. The candidate may choose this option if he/she did not identify that only two EPAs were impacted and did not correctly identify that only one side of RPS was impacted. This answer is plausible because there are some single components that cause a full scram vice a divisional partial scram. The RPS Shorting Link Switches are in CLOSE during power operation thus inhibiting a full reactor scram from a single channel trip.
C. This option is incorrect because only the two breakers associated with the MG Set will trip because RPSPP1B remains energized. The candidate may choose this option if he/she did not identify that only two EPAs were impacted and did not correctly identify that only one side of RPS was impacted. This answer is plausible because there are some single components that cause a full scram vice a divisional partial scram. The RPS Shorting Link Switches are in CLOSE during power operation thus inhibiting a full reactor scram from a single channel trip.
D. This option is incorrect because only the two breakers associated with the MG Set will trip. The candidate may choose this option if he/she did not identify that only two EPAs were impacted and did not correctly identify that only one side of RPS was impacted. This answer is plausible because the partial scram is correct. The correct combination of 4 EPA breakers opening (A1, 2, 3 & 4) will only cause a 1/2 scram. The RPS Shorting Link Switches are in CLOSE during power operation thus inhibiting a full reactor scram from a single channel trip.
D. This option is incorrect because only the two breakers associated with the MG Set will trip.
Technical Reference(s):
The candidate may choose this option if he/she did not identify that only two EPAs were impacted and did not correctly identify that only one side of RPS was impacted. This answer is plausible because the partial scram is correct. The correct combination of 4 EPA breakers opening (A1, 2, 3 & 4) will only cause a 1/2 scram. The RPS Shorting Link Switches are in CLOSE during power operation thus inhibiting a full reactor scram from a single channel trip.
OPS Reactor Protection System COR002-21-02, Rev 23 (Attach if not previously provided)
Technical Reference(s):                 OPS Reactor Protection System COR002-21-02, Rev 23 (Attach if not previously provided)     SOP-2.2.22, Vital Instrument Power Supply, Rev 71____
SOP-2.2.22, Vital Instrument Power Supply, Rev 71
(including version/revision number) _______________________________________________
____ (including version/revision number)
Proposed references to be provided to applicants during examination:         None Learning Objective:
_______________________________________________
Proposed references to be provided to applicants during examination:
None Learning Objective:
COR002-21-02, OPS Reactor Protection System
COR002-21-02, OPS Reactor Protection System
: 4. Describe the RPS design features and/or interlocks that provide for the following:
: 4.     Describe the RPS design features and/or interlocks that provide for the following:
: b. Scram prevention following single component failure
: b. Scram prevention following single component failure
: l. Under/over voltage and frequency protection
: l. Under/over voltage and frequency protection
: 8. Given a specific RPS malfunction, determine the effect on any of the following:
: 8. Given a specific RPS malfunction, determine the effect on any of the following:
: f. RPS logic channels Question Source:
: f. RPS logic channels Question Source:               Bank #                   _     _
Bank #   _         _
Modified Bank #         _5210__ (See attached)
Modified Bank #
New                     _______
_5210__ (See attached)
Question History:               Last NRC Exam           ____________
New   _______
Question Cognitive Level:       Memory or Fundamental Knowledge               _____
Question History:
Comprehension or Analysis                     __X__
Last NRC Exam
10 CFR Part 55 Content:         55.41 _(7)_
____________
Comments:
Question Cognitive Level:
LOD 2 83
Memory or Fundamental Knowledge
_____ Comprehension or Analysis   __X__ 10 CFR Part 55 Content:
55.41 _(7)_ Comments: LOD 2 83 Original Bank Question:
84 85 86 87 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
-


==Reference:==
Original Bank Question:
84


Level  RO  SRO Tier #    __2__  _____ Group #  __1__  _____ K/A #  _  215003  K5.01
85 86 87 ES-401                              Sample Written Examination                        Form ES-401-5 Question Worksheet Examination Outline Cross-
_ Importance Rating
_2.6 _  _____  215003 IRM
- Knowledge of the operational implications of the following concepts as they apply to intermediate range monitor (IRM) system: (CFR: 41.5) K5.01 Detector operation Question: 35  IRM A detector is installed with twice the argon fill pressure of the other IRM detectors.


==Reference:==
Level                    RO            SRO Tier #                  __2__          _____
Group #                  __1__          _____
K/A #                    _ 215003 K5.01 _
Importance Rating        _2.6 _        _____
215003 IRM - Knowledge of the operational implications of the following concepts as they apply to intermediate range monitor (IRM) system: (CFR: 41.5) K5.01 Detector operation Question: 35 IRM A detector is installed with twice the argon fill pressure of the other IRM detectors.
How does this affect the operation of the IRM A versus the others when subjected to the same neutron field?
How does this affect the operation of the IRM A versus the others when subjected to the same neutron field?
IRM A High High trip is-A. less conservative and the downscale rod block is less conservative.
IRM A High High trip is A. less conservative and the downscale rod block is less conservative.
B. less conservative and the downscale rod block is more conservative.
B. less conservative and the downscale rod block is more conservative.
C. more conservative and the downscale rod block is less conservative.
C. more conservative and the downscale rod block is less conservative.
D. more conservative and the downscale rod block is more conservative.
D. more conservative and the downscale rod block is more conservative.
 
Answer:
Answer: C. more conservative and the downscale rod block is less conservative.
C. more conservative and the downscale rod block is less conservative.
Explanation:
Explanation:
The IRM detectors operate in the ionization region and their output (in a constant neutron flux) is effected primarily by the Argon pressure. The IRMs are not sensitive to small voltage changes because of their operation in the ionization region. Fission events in the detector cause ionizations as the fission fragments ionize the detector fill gas. If there is more argon in the detector then there will be more ions created by a given fission event so for IRM A each fission event causes a greater detector output. This means that the high high trip for that IRM occurs at a lower neutron flux (conservative) and because the IRM output is higher it also means that the downscale trip does not occur until a lower (non
The IRM detectors operate in the ionization region and their output (in a constant neutron flux) is effected primarily by the Argon pressure. The IRMs are not sensitive to small voltage changes because of their operation in the ionization region. Fission events in the detector cause ionizations as the fission fragments ionize the detector fill gas. If there is more argon in the detector then there will be more ions created by a given fission event so for IRM A each fission event causes a greater detector output. This means that the high high trip for that IRM occurs at a lower neutron flux (conservative) and because the IRM output is higher it also means that the downscale trip does not occur until a lower (non-conservative) neutron flux is reached.
-conservative) neutron flux is reached.
A. This option is incorrect because the high high trip is more conservative because the output will reach the high high trip at a lower neutron flux. This selection is plausible because a candidate may believe that the additional argon in the detector would shield the ions from the anode (cathode), and would also believe that any reduction in detector output causes all associated trips to be less conservative.
A. This option is incorrect because the high high trip is more conservative because the output will reach the high high trip at a lower neutron flux. This selection is plausible because a candidate may believe that the additional argon in the detector would shield the ions from the anode (cathode), and would also believe that any reduction in detector output causes all associated trips to be less conservative.
88 B. This option is incorrect because the high high trip is more conservative because the output will reach the high high trip at a lower neutron flux. A candidate who believes that the additional argon in the detector would shield the ions from the anode (cathode) may choose this answer. This choice is plausible since it would be the correct answer if an event occurred that overall reduced the output (such as a very low detector voltage or loss of argon gas pressure).
88
 
B. This option is incorrect because the high high trip is more conservative because the output will reach the high high trip at a lower neutron flux. A candidate who believes that the additional argon in the detector would shield the ions from the anode (cathode) may choose this answer. This choice is plausible since it would be the correct answer if an event occurred that overall reduced the output (such as a very low detector voltage or loss of argon gas pressure).
D. This option is incorrect because the downscale rod block would be less conservative. With the additional argon gas the downscale rod block would occur at a lower neutron level than that of the other detectors so as neutron flux lowers the downscale for IRM A is delayed. A candidate may believe that, because there is more argon gas pressure, the raised detector output makes all the associated actions more conservative and would therefore choose this option.
D. This option is incorrect because the downscale rod block would be less conservative. With the additional argon gas the downscale rod block would occur at a lower neutron level than that of the other detectors so as neutron flux lowers the downscale for IRM A is delayed. A candidate may believe that, because there is more argon gas pressure, the raised detector output makes all the associated actions more conservative and would therefore choose this option.
Technical Reference(s):
Technical Reference(s):                 Lesson Plan COR002-12-02 (IRM), Rev 15 (Attach if not previously provided)
Lesson Plan COR002 02 (IRM), Rev 15 (Attach if not previously provided)
(including version/revision number) ______________________________________________
          (including version/revision number)
Proposed references to be provided to applicants during examination:       ___None_________
______________________________________________
Learning Objective:
Proposed references to be provided to applicants during examination:
COR001-10-02
___None_________  Learning Objective:
: 5. Describe how changes in each of the following affect detector sensitivity:
: b.      Detector gas pressure Question Source:                Bank #                  ________
Modified Bank #        _23352_ (See attached)
New                    _______
Question History:              Last NRC Exam          _CNS 2008__
Question Cognitive Level:      Memory or Fundamental Knowledge            ____
Comprehension or Analysis                  __X___
10 CFR Part 55 Content:        55.41 _(5)__
Comments:
LOD 3 89


COR001-10-02 5. Describe how changes in each of the following affect detector sensitivity:
Original Bank question:
: b. Detector gas pressure Question Source:
90
Bank #  ________ Modified Bank #
_23352_ (See attached)
New  _______
Question History:
Last NRC Exam
_CNS 2008__  Question Cognitive Level:
Memory or Fundamental Knowledge
____ Comprehension or Analysis  __X___  10 CFR Part 55 Content:
55.41 _(5)__  Comments: LOD 3  89 Original Bank question:
90 From COR002 02, Slide 31 91 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
-


==Reference:==
From COR002-12-02, Slide 31 91


Level  RO  SRO Tier #    _2___  _____ Group #  _1___  _____ K/A #  _215004  K6.01 _
ES-401                              Sample Written Examination                        Form ES-401-5 Question Worksheet Examination Outline Cross-
__ Importance Rating
_3.2 _  _____  215004 Source Range Monitor
- Knowledge of the effect that a loss or malfunction of the following will have on the source range monitor (SRM) system: (CFR: 41.7) K6.01 RPS Question: 36


The plant is performing a reactor startup requiring the SRM Shorting Link Switches to be placed in the OPEN position for testing.
==Reference:==
Level                RO              SRO Tier #                _2___            _____
Group #              _1___            _____
K/A #                _215004 K6.01 ___
Importance Rating    _3.2 _          _____
215004 Source Range Monitor - Knowledge of the effect that a loss or malfunction of the following will have on the source range monitor (SRM) system: (CFR: 41.7) K6.01 RPS Question: 36 The plant is performing a reactor startup requiring the SRM Shorting Link Switches to be placed in the OPEN position for testing.
All Shorting Link Switches have been placed in OPEN, however while placing switches for the B Channel to OPEN, the contacts remain CLOSED (switch position changes while switch contacts do not change state).
All Shorting Link Switches have been placed in OPEN, however while placing switches for the B Channel to OPEN, the contacts remain CLOSED (switch position changes while switch contacts do not change state).
What is the impact of this RPS Logic malfunction if SRM A reaches the Hi Hi setpoint?
What is the impact of this RPS Logic malfunction if SRM A reaches the Hi Hi setpoint?
A. A full Reactor Scram occurs.
A. A full Reactor Scram occurs.
B. Only a half scram from "A" RPS trip system occurs.
B. Only a half scram from "A" RPS trip system occurs.
C. Only a half scram from "B" RPS trip system occurs.
C. Only a half scram from "B" RPS trip system occurs.
D. A full Reactor Scram occurs only if a second SRM trip signal is received.
D. A full Reactor Scram occurs only if a second SRM trip signal is received.
 
Answer:
Answer: B. Only a half scram from "A" RPS trip system occurs.
B. Only a half scram from "A" RPS trip system occurs.
Explanation:
Explanation:
The shorting link switches being OPEN allow the neutron monitoring non
The shorting link switches being OPEN allow the neutron monitoring non-coincident trips (any single neutron monitoring trip causes a FULL scram) in the A3 and B3 scram logics. If the RPS B3 shorting link contacts remaining closed, the B3 channel will not allow the SRM upscale high-high trip (SRM trip is only available in the A/B3 logic) and prevents the RPS B trip from occurring. This failure will not impact the ability of the SRM channel to input a trip signal. The candidate would choose this answer based on the shorting links being open would cause the B3 RPS Manual Scram Trip logic and A3 RPS Manual Scram Trip Logic to become a two-out-of-two taken once logic. By removing the ability of the B3 logic to trip, a 1/2 scram could not occur on the B side and a full scam due to SRM inputs would not be possible. This question is a K/A match because the RPS logic failure bypasses SRM scram inputs. If the contacts were not closed, one SRM upscale trip causes both A3 and B3 logics to trip which is a full reactor scram.
-coincident trips (any single neutron monitoring trip causes a FULL scram) in the A3 and B3 scram logics. If the RPS B3 shorting link contacts remaining closed, the B3 channel will not allow the SRM upscale high
-high trip (SRM trip is only available in the A/B3 logic) and prevents the RPS B trip from occurring. This failure will not impact the ability of the SRM channel to input a trip signal. The candidate would choose this answer based on the shorting links being open would cause the B3 RPS Manual Scram Trip logic and A3 RPS Manual Scram Trip Logic to become a two
-out-of-two taken once logic. By removing the ability of the B3 logic to trip, a 1/2 scram could not occur on the B side and a full scam due to SRM inputs would not be possible. This question is a K/A match because the RPS logic failure bypasses SRM scram inputs. If the contacts were not closed, one SRM upscale trip causes both A3 and B3 logics to trip which is a full reactor scram.
 
A. This option is incorrect because the RPS failure prevents any and all SRM induced full Reactor scrams (but not half scrams). The candidate could choose this distractor if he/she 1
A. This option is incorrect because the RPS failure prevents any and all SRM induced full Reactor scrams (but not half scrams). The candidate could choose this distractor if he/she 1
did not know the relationship between RPS Manual Scram Trip logic and the non
-coincident neutron monitoring trips. Since the RPS failure is on the B3 system, it is a common misconception that this failure would only prevent SRM B and D from initiating a full scram.


C. This option is incorrect because the failure present in RPS actually prevents an RPS B trip. The candidate who does not understand that the trip condition of the B3 logic is energized would choose this option.
did not know the relationship between RPS Manual Scram Trip logic and the non-coincident neutron monitoring trips. Since the RPS failure is on the B3 system, it is a common misconception that this failure would only prevent SRM B and D from initiating a full scram.
D. This option is incorrect because the failure prevents any and all SRM induced full Reactor scrams. The candidate could choose this distractor if he/she did not understand the non
C. This option is incorrect because the failure present in RPS actually prevents an RPS B trip.
-coincident input of the SRMs through the shorting links or does not recall the Shorting Link s switch positions. This answer is plausible because a non
The candidate who does not understand that the trip condition of the B3 logic is energized would choose this option.
-coincident scram from the SRMs does occur on the upscale trip of any SRM.
D. This option is incorrect because the failure prevents any and all SRM induced full Reactor scrams. The candidate could choose this distractor if he/she did not understand the non-coincident input of the SRMs through the shorting links or does not recall the Shorting Links switch positions. This answer is plausible because a non-coincident scram from the SRMs does occur on the upscale trip of any SRM.
Technical Reference(s):              _                                                ________
(Attach if not previously provided)  791E256 Reactor Protection System Elementary Electrical (including version/revision number) _                                              __________
Proposed references to be provided to applicants during examination:        ___None      _______
Learning Objective:
Per COR002-30-02, Source Range Monitor
: 8. Predict the consequences a malfunction of the following would have on the SRM system:
: a.      RPS (including shorting switches)
Question Source:              Bank #                _            _
Modified Bank #      ___23354_____
New                  ____ ___
Question History:              Last NRC Exam        ____________
Question Cognitive Level:      Memory or Fundamental Knowledge              _____
Comprehension or Analysis                    __X__
10 CFR Part 55 Content:        55.41 _(7)_
Comments:
LOD 4 2


Technical Reference(s):
3 4
_
5 6
________ (Attach if not previously provided) 791E256 Reactor Protection System Elementary Electrical (including version/revision number)
ES-401                           Sample Written Examination                       Form ES-401-5 Question Worksheet Examination Outline Cross-
_          __________
Proposed references to be provided to applicants during examination:
___None
_______
Learning Objective:
Per COR002 02, Source Range Monitor
: 8. Predict the consequences a malfunction of the following would have on the SRM system: a. RPS (including shorting switches)
Question Source:
Bank #  _
_ Modified Bank #
___23354_____  New  ____  ___
Question History:
Last NRC Exam
____________
Question Cognitive Level:
Memory or Fundamental Knowledge
_____ Comprehension or Analysis  __X__  10 CFR Part 55 Content:
55.41 _(7)_  Comments:  LOD 4   2 3
4 5
6 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
-


==Reference:==
==Reference:==
 
Level               RO             SRO Tier #               __2__           _____
Level   RO SRO Tier #     __2__ _____ Group # __1__ _____ K/A #   __215005 A1.03
Group #             __1__           _____
___ Importance Rating
K/A #               __215005 A1.03___
_3.6__ _____ 215005 APRM / LPRM  
Importance Rating   _3.6__         _____
- Ability to predict and/or monitor changes in parameters associated with operating the average power range monitor/local power range monitor systems controls including: (CFR: 41.5) A1.03 Control rod block status Question: 37 The reactor is operating at 15% rated power when the following occurs:
215005 APRM / LPRM - Ability to predict and/or monitor changes in parameters associated with operating the average power range monitor/local power range monitor systems controls including: (CFR: 41.5) A1.03 Control rod block status Question: 37 The reactor is operating at 15% rated power when the following occurs:
IRM "B" fails upscale.
* IRM "B" fails upscale.
Then, before any operator action is taken:
Then, before any operator action is taken:
APRMs "B" and "E" both fail downscale.
* APRMs "B" and "E" both fail downscale.
What are the minimum action(s) that will allow all rod blocks and/or scrams to be cleared?
What are the minimum action(s) that will allow all rod blocks and/or scrams to be cleared?
A. Bypass IRM "B" only.
A. Bypass IRM "B" only.
Line 1,620: Line 1,594:
C. Bypass APRM "B" and IRM "B".
C. Bypass APRM "B" and IRM "B".
D. Bypass APRM "B" and APRM "E".
D. Bypass APRM "B" and APRM "E".
 
Answer:
Answer: D. Bypass APRM "B" and APRM "E".
D. Bypass APRM "B" and APRM "E".
Explanation:
Explanation:
Since the reactor is at 15% RTP, it is implied that the Reactor Mode Switch is in the RUN position. The upscale IRM "B" or IRM "H" and the downscale APRM "B" together generates an RPS trip on the "B" RPS. APRM "B" or APRM "E" failed downscale cause a rod block.
Since the reactor is at 15% RTP, it is implied that the Reactor Mode Switch is in the RUN position. The upscale IRM B or IRM H and the downscale APRM B together generates an RPS trip on the "B" RPS. APRM "B" or APRM E" failed downscale cause a rod block.
To clear the RPS trip, APRM "B" OR IRM "B" and IRM "H" must be bypassed. To clear the rod block, both APRM "B" AND APRM "E" must be bypassed.
To clear the RPS trip, APRM "B" OR IRM B and IRM H must be bypassed. To clear the rod block, both APRM B AND APRM E must be bypassed.
Bypassing both APRM "B" and APRM "E" will clear the rod block, and bypassing APRM "B" als o clears the RPS trip signal.
Bypassing both APRM B and APRM E will clear the rod block, and bypassing APRM B also clears the RPS trip signal.
7


7 Distracters:
Distracters:
A. This option is incorrect because this action will not clear the rod block or 1/2 scram. This answer is plausible because if the Mode switch were in STARTUP, this action would clear the 1/2 scram and rod block. The candidate who does not recognize the Mode switch being in RUN and understands IRM upscale causes a 1/2 scram & rod block in startup would choose this option.
A. This option is incorrect because this action will not clear the rod block or 1/2 scram. This answer is plausible because if the Mode switch were in STARTUP, this action would clear the 1/2 scram and rod block. The candidate who does not recognize the Mode switch being in RUN and understands IRM upscale causes a 1/2 scram & rod block in startup would choose this option.
B. This option is incorrect because this action will not clear either the rod block or the half scram. The candidate could choose this distractor if he/she did not understand that a rod block is generated by EITHER APRM B OR APRM E failed downscale and he/she did not know the 1/2 scram was generated by the upscale IRM "B" or IRM "H" and the downscale on APRM "B". This answer is plausible because the APRM can be bypassed and its direct trip is also bypassed clearing the 1/2 scram
B. This option is incorrect because this action will not clear either the rod block or the half scram. The candidate could choose this distractor if he/she did not understand that a rod block is generated by EITHER APRM B OR APRM E failed downscale and he/she did not know the 1/2 scram was generated by the upscale IRM B or IRM H and the downscale on APRM B. This answer is plausible because the APRM can be bypassed and its direct trip is also bypassed clearing the 1/2 scram.
. C. This option is incorrect because this action will clear the half scram but will not clear the rod block. The candidate could choose this option if he/she did not know the 1/2 scram was generated by both the upscale IRM "B" and the downscale APRM "B". This answer is plausible because the APRM and IRM can be bypassed and their direct trips are also bypassed. Technical Reference(s):
C. This option is incorrect because this action will clear the half scram but will not clear the rod block. The candidate could choose this option if he/she did not know the 1/2 scram was generated by both the upscale IRM B and the downscale APRM B. This answer is plausible because the APRM and IRM can be bypassed and their direct trips are also bypassed.
IOP 4.1.3, Average Power Range Monitoring System, Rev. 25 Proposed references to be provided to applicants during examination:
Technical Reference(s):       IOP 4.1.3, Average Power Range Monitoring System, Rev. 25 Proposed references to be provided to applicants during examination:         ___None________
___None________
Learning Objective:
Learning Objective:
Per COR002-01-02, Average Power Range Monitor
: 5. Describe the interrelationships between the Average Power Range Monitor system and the following:
: a.      Reactor Protection System (RPS)
: b.      Intermediate Range Monitoring system (IRM) f        Reactor Manual Control System (RMCS)
Question Source:              Bank #                _23450_
Modified Bank #        _______
New                    _______
Question History:              Last NRC Exam          _CNS 2006__
Question Cognitive Level:      Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content:        55.41 _(5)_
Comments:
LOD 3 8


Per COR002 02, Average Power Range Monitor
9 10 11 12 ES-401                             Sample Written Examination                   Form ES-401-5 Question Worksheet Examination Outline Cross-
: 5. Describe the interrelationships between the Average Power Range Monitor system and the following:
: a. Reactor Protection System (RPS)
: b. Intermediate Range Monitoring system (IRM) f Reactor Manual Control System (RMCS)
Question Source:
Bank #  _23450_ Modified Bank #
_______  New  _______  Question History:
Last NRC Exam
_CNS 2006__
Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X    10 CFR Part 55 Content:
55.41 _(5)_  Comments: LOD 3 8 9
10 11 12 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
-


==Reference:==
==Reference:==
 
Level               RO           SRO Tier #               __2__         _____
Level   RO SRO Tier #     __2__ _____ Group # __1__ _____ K/A #   __217000 A2.16
Group #             __1__         _____
___ Importance Rating
K/A #               __217000 A2.16___
_3.5__ _____ 217000 RCIC  
Importance Rating   _3.5__       _____
- Ability to (a) predict the impacts of the following on the reactor core isolation cooling system (RCIC) ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5) A2.16 Low condensate storage tank level
217000 RCIC - Ability to (a) predict the impacts of the following on the reactor core isolation cooling system (RCIC) ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5) A2.16 Low condensate storage tank level Question: 38 Given the following conditions:
 
* RCIC is injecting to the reactor at 400 gpm.
Question: 38 Given the following conditions:
* RCIC suction is from the ECST.
RCIC is injecting to the reactor at 400 gpm.
* Torus temperature is 125&deg;F and steady.
RCIC suction is from the ECST.
* ECST level is 25 inches and lowering.
Torus temperature is 125&deg;F and steady.
* Torus level is 12 9 and rising.
ECST level is 25 inches and lowering.
Torus level is 12' 9" and rising.
 
(1) If the conditions persist, how is RCIC impacted?
(1) If the conditions persist, how is RCIC impacted?
(2) What action is required IAW 9-4-1/F-2 (RCIC Suction Transfer)
(2) What action is required IAW 9-4-1/F-2 (RCIC Suction Transfer)?
?  (1)     (2) RCIC Suction aligns to torus-Action required A. Secure RCIC B. Makeup to the ECSTs C. . Secure RCIC D. . Makeup to the ECSTs Answer: B. Makeup to the ECSTs Explanation:
(1)                                         (2)
The alternate source of water for the RCIC pump is the Suppression Pool (torus). This source of water is used if the Emergency Condensate Storage Tank levels are low. The valve (MO
RCIC Suction aligns to torus                 Action required A.       at 23 ECST Level.                          Secure RCIC B.       at 23 ECST Level.                          Makeup to the ECSTs C.       at 13 1 Torus Level .                     Secure RCIC D.       at 13 1 Torus Level .                     Makeup to the ECSTs Answer:
-41) for RCIC pump suction from the Supprfrom the bottom of either Emergency Condensate Storage Tank. Per AP 9 1/F-2 (RCIC Suction Transfer), when the suction transfer alarm is received, action should be taken to provide makeup to the ECST per Procedure 2.2.7, Condensate Storage and Transfer System.
B.       at 23 ECST Level.                          Makeup to the ECSTs Explanation:
The alternate source of water for the RCIC pump is the Suppression Pool (torus). This source of water is used if the Emergency Condensate Storage Tank levels are low. The valve (MO-41) for RCIC pump suction from the Suppression Pool will automatically open on a low level of 23" from the bottom of either Emergency Condensate Storage Tank. Per AP 9-4-1/F-2 (RCIC Suction Transfer), when the suction transfer alarm is received, action should be taken to provide makeup to the ECST per Procedure 2.2.7, Condensate Storage and Transfer System.
Distracters:
Distracters:
13 A. This option is incorrect because there is no need to secure the RCIC system following the swap. If the torus temperature were 145&deg;F instead of 125&deg;F, RCIC would need to be secured in order to prevent overheating the lube oil. The candidate who understands the RCIC suction swap but who does not remember the maximum torus temperature that supports RCIC lube oil cooling would choose this answer.
13
 
A. This option is incorrect because there is no need to secure the RCIC system following the swap. If the torus temperature were 145&deg;F instead of 125&deg;F, RCIC would need to be secured in order to prevent overheating the lube oil. The candidate who understands the RCIC suction swap but who does not remember the maximum torus temperature that supports RCIC lube oil cooling would choose this answer.
C. This option is incorrect because the RCIC suction swap does not occur on high torus level and the torus water temperature is not high enough to prohibit RCIC operation. The HPCI system does have a suction swap on high level, as well as the low ECST, so a candidate could easily confuse which system has only one parameter that swaps the suction and what that condition causes the swap. If the torus temperature were higher as well the RCIC system would need to be secured due to high oil temperature concerns.
C. This option is incorrect because the RCIC suction swap does not occur on high torus level and the torus water temperature is not high enough to prohibit RCIC operation. The HPCI system does have a suction swap on high level, as well as the low ECST, so a candidate could easily confuse which system has only one parameter that swaps the suction and what that condition causes the swap. If the torus temperature were higher as well the RCIC system would need to be secured due to high oil temperature concerns.
 
D. This option is incorrect because the RCIC suction swap does not occur on high torus level.
D. This option is incorrect because the RCIC suction swap does not occur on high torus level. The HPCI system does have a suction swap on high level, as well as the low ECST, so a candidate could easily confuse which system has only one parameter that swaps the suction and what condition causes the swap. This answer option is a plausible misconception because candidates often confuse the interlocks associated with HPCI and RCIC which are quite similar and this option would be correct if the system asked in the stem were HPCI. Makeup to the ECSTs is correct.
The HPCI system does have a suction swap on high level, as well as the low ECST, so a candidate could easily confuse which system has only one parameter that swaps the suction and what condition causes the swap. This answer option is a plausible misconception because candidates often confuse the interlocks associated with HPCI and RCIC which are quite similar and this option would be correct if the system asked in the stem were HPCI.
Makeup to the ECSTs is correct.
Technical Reference(s):
Technical Reference(s):
    (Attach if not previously provided) SOP 2.2.67,Reactor Core Isolation Cooling Ops, Rev. 70 (including version/revision number)
(Attach if not previously provided)   SOP 2.2.67,Reactor Core Isolation Cooling Ops, Rev. 70 (including version/revision number) ARP 9-4-1/F-2, Rev. 51___________________________
ARP 9-4-1/F-2, Rev. 51___________________________
Proposed references to be provided to applicants during examination:         ___None________
Proposed references to be provided to applicants during examination:
Learning Objective:
___None________
COR002-18-02, Reactor Core Isolation Cooling
Learning Objective: COR002-18-02, Reactor Core Isolation Cooling
: 10. Describe the interrelationship between RCIC system and the following:
: 10. Describe the interrelationship between RCIC system and the following:
: h. ECSTs 8. Describe the RCIC system design features and/or interlocks that provide for the following:
: h. ECSTs
: 8.     Describe the RCIC system design features and/or interlocks that provide for the following:
: a. Alternate water supplies
: a. Alternate water supplies
: 10. Predict the consequences of the following on the RCIC system:
: 10. Predict the consequences of the following on the RCIC system:
Line 1,688: Line 1,661:
: b. RCIC suction transfer on Low ECST water level
: b. RCIC suction transfer on Low ECST water level
: 12. Given plant conditions, determine if the following RCIC actions should occur:
: 12. Given plant conditions, determine if the following RCIC actions should occur:
: c. ECST suction transfer
: c. ECST suction transfer Question Source:                Bank #                _      _
Modified Bank #        _    __
New                    ___X___
Question History:              Last NRC Exam          ____________
Question Cognitive Level:      Memory or Fundamental Knowledge              _____
Comprehension or Analysis                    __X__
10 CFR Part 55 Content:        55.41 _(5)_
14
 
Comments:
LOD 3 15


Question Source:
16 ES-401                           Sample Written Examination                       Form ES-401-5 Question Worksheet Examination Outline Cross-
Bank #  _        _
Modified Bank #
_      __  New  ___X___
Question History:
Last NRC Exam
____________
Question Cognitive Level:
Memory or Fundamental Knowledge
_____ Comprehension or Analysis  __X__  10 CFR Part 55 Content:
55.41 _(5)_ 14 Comments: LOD  3  15 16 ES-401 Sample Written Examination Form ES-401-5 Question Workshee t    Examination Outline Cross
-


==Reference:==
==Reference:==
 
Level                 RO             SRO Tier #                 _ 2__         _____
Level   RO SRO Tier #     _ 2__ _____ Group # _ 1__ _____ K/A #   _ 218000 A3.02 _
Group #               _ 1__         _____
__ Importance Rating
K/A #                 _ 218000 A3.02 ___
_ 3.6_ _____ 218000 ADS  
Importance Rating     _ 3.6_         _____
- Ability to monitor automatic operations of the automatic depressurization system including: (CFR: 41.7) A3.02 ADS valve tail pipe temperatures
218000 ADS - Ability to monitor automatic operations of the automatic depressurization system including: (CFR: 41.7) A3.02 ADS valve tail pipe temperatures Question: 39 With reactor pressure at 900 psig an automatic initiation of ADS occurs.
 
What provides positive indication that each ADS valve is open?
Q ues ti on: 3 9  With reactor pressure at 900 psig an a u t o m a ti c i niti ati o n o f AD S occu r s. W ha t p r ov i de s positi v e i nd i ca ti o n t ha t eac h AD S va l ve i s open?
A. The red light for each ADS valve is illuminated on Panel 9-3.
A. Th e r e d li gh t for each ADS valve i s ill u m i na t e d o n P ane l 9-3. B. Th e g r ee n li gh t f o r eac h AD S va l ve s i s ex tingu ishe d o n pane l 9-3. C. PC-TR-24 and 25 (SUPPR COOL TEMP RECORDER), Point 4 reading 90&deg;F.
B. The green light for each ADS valves is extinguished on panel 9-3.
D. Te m pe r a t u r e read i ng s 285&#xba;F t o 300&#xba;F f o r eac h AD S va l ve o n t e m pe ratu r e reco r de r MS-T R-166 (SAFETY AND RELIEF VALVE LEAKAGE TEMPS)
C. PC-TR-24 and 25 (SUPPR COOL TEMP RECORDER), Point 4 reading 90&deg;F.
. A ns w e r: D. Te m pe r a t u r e read i ng s 285&#xba;F t o 300&#xba;F f o r eac h AD S va l ve o n t e m pe r a t u r e reco r de r MS-T R-16 6 (SAFETY AND RELIEF VALVE LEAKAGE TEMPS)
D. Temperature readings 285&#xba;F to 300&#xba;F for each ADS valve on temperature recorder MS-TR-166 (SAFETY AND RELIEF VALVE LEAKAGE TEMPS).
. Explanation:
Answer:
When an SRV lifts, the tailpipe temperature and pressure increase to approximately 285
D. Temperature readings 285&#xba;F to 300&#xba;F for each ADS valve on temperature recorder MS-TR-166 (SAFETY AND RELIEF VALVE LEAKAGE TEMPS).
-300&#xba;F and 30 psig in a constant enthalpy process that superheats the SRV exhaust steam. The temperature recorder provides a positive indication that the ADS SRVs are open. The red indicating light associated with each SRV illuminates when the solenoid for the associated valve is energized meaning the circuitry is supposed to reposition the pilot valve allowing the SRV to open. The red light does not indicate SRV position however as the SRV may fail to open even with the solenoid energized. The green light for each SRV extinguish on an ADS signal, but do not indicate SRV position.
Explanation:
When an SRV lifts, the tailpipe temperature and pressure increase to approximately 285-300&#xba;F and 30 psig in a constant enthalpy process that superheats the SRV exhaust steam. The temperature recorder provides a positive indication that the ADS SRVs are open. The red indicating light associated with each SRV illuminates when the solenoid for the associated valve is energized meaning the circuitry is supposed to reposition the pilot valve allowing the SRV to open. The red light does not indicate SRV position however as the SRV may fail to open even with the solenoid energized. The green light for each SRV extinguish on an ADS signal, but do not indicate SRV position.
The annunciators are tied to the ADS logic being satisfied. They do not indicate SRV position.
The annunciators are tied to the ADS logic being satisfied. They do not indicate SRV position.
The SRV tailpipe temperatures actually indicate steam is flowing from the SRV, hence they indicate that the individual SRVs are open and passing steam.
The SRV tailpipe temperatures actually indicate steam is flowing from the SRV, hence they indicate that the individual SRVs are open and passing steam.
The SPDS computer program monitors Suppression Pool Temperatures at various locations around the pool and averages the temperature readings.
The SPDS computer program monitors Suppression Pool Temperatures at various locations around the pool and averages the temperature readings. The SPDS indicated temperature only means that energy is being transferred to the Suppression Pool but is not an indication all six ADS valves are open.
The SPDS indicated temperature only means that energy is being transferred to the Suppression Pool but is not an indication all six ADS valves are open.
17
17 A. This option is incorrect because the red lights only indicate that the solenoids are energized but do not provide positive indication that the valves are open. However the red lights associated with each ADS valve are energized every time ADS automatically actuates. A candidate would see the red lights illuminated whenever he/she sees an ADS actuation in the simulator. And while this does occur for every valve it is not positive indication of valve position.
 
A. This option is incorrect because the red lights only indicate that the solenoids are energized but do not provide positive indication that the valves are open. However the red lights associated with each ADS valve are energized every time ADS automatically actuates. A candidate would see the red lights illuminated whenever he/she sees an ADS actuation in the simulator. And while this does occur for every valve it is not positive indication of valve position.
B. This option is incorrect because the green lights extinguishing only indicate that the ADS logic is satisfied. It does not provide positive indication of SRV position. However, this occurs with every ADS actuation so a candidate who does not completely understand the system may associate that indication with positive indication of valve position and choose this option.
B. This option is incorrect because the green lights extinguishing only indicate that the ADS logic is satisfied. It does not provide positive indication of SRV position. However, this occurs with every ADS actuation so a candidate who does not completely understand the system may associate that indication with positive indication of valve position and choose this option.
 
C. This option is incorrect because the average suppression pool temperature rising at one point is not positive indication all the ADS valves are open, only that SRVs are open. This answer is plausible because the average temperature rises when heat is added to the suppression pool. The candidate who knows the average temperature utilizes all different area temperatures in the suppression pool and doesnt fully internalize the average temperature will rise no matter how many valves are open would select this option.
C. This option is incorrect because the average suppression pool temperature rising at one point is not positive indication all the ADS valves are open, only that SRVs are open. This answer is plausible because the average temperature rises when heat is added to the suppression pool. The candidate who knows the average temperature utilizes all different area temperatures in the suppression pool and doesn't fully internalize the average temperature will rise no matter how many valves are open would select this option.
Technical Reference(s):         Procedure 2.4SRV (Stuck Open Relief Valve), Rev. 15 Procedure 2.2.1 (Nuclear Pressure Relief System), Rev. 38 Proposed references to be provided to applicants during examination:         __None    _________
Technical Reference(s):
Learning Objective:
Procedure 2.4SRV (Stuck Open Relief Valve), Rev. 15 Procedure 2.2.1 (Nuclear Pressure Relief System), Rev. 38 Proposed references to be provided to applicants during examination:
__None    _________ Learning Objective:
Per COR002-16-02, Nuclear Pressure Relief
Per COR002-16-02, Nuclear Pressure Relief
: 6. Briefly describe the following concepts as they apply to NPR:
: 6. Briefly describe the following concepts as they apply to NPR:
: d. Tail pipe temperature monitoring Question Source:
: d.       Tail pipe temperature monitoring Question Source:               Bank #                 _______
Bank #   _______ Modified Bank #
Modified Bank #         _1860__ (See attached)
_1860__ (See attached)
New                     _______
New   _______ Question History:
Question History:               Last NRC Exam           ____________
Last NRC Exam ____________
Question Cognitive Level:       Memory or Fundamental Knowledge               _____
Question Cognitive Level:
Comprehension or Analysis                     __X__
Memory or Fundamental Knowledge
10 CFR Part 55 Content:         55.41 _(7) _
_____ Comprehension or Analysis   __X__ 10 CFR Part 55 Content:
Comments:
55.41 _(7) _ Comments: LOD 3 18 19 20 21 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
LOD 3 18
-
 
19 20 21 ES-401                           Sample Written Examination                       Form ES-401-5 Question Worksheet Examination Outline Cross-


==Reference:==
==Reference:==
 
Level                 RO             SRO Tier #                 __2__         _____
Level   RO SRO Tier #     __2__ _____ Group # __1__ _____ K/A #   _ 223002 A4.06_
Group #               __1__         _____
__ Importance Rating
K/A #                 _ 223002 A4.06___
_ 3.6 _ _____ 223002 PCIS/Nuclear Steam Supply Shutoff  
Importance Rating     _ 3.6 _       _____
- Ability to manually operate and/or monitor in the control room: (CFR: 41.7) A.4.06 Confirm initiation to completion
223002 PCIS/Nuclear Steam Supply Shutoff - Ability to manually operate and/or monitor in the control room: (CFR: 41.7) A.4.06 Confirm initiation to completion Question: 40 A manual reactor Scram is inserted due to a transient while operating at 50% power.
 
Q ues ti on: 4 0  A manual reactor Scram is inserted due to a transient while operating at 50% power.
The following conditions occur after the Mitigating Task Scram Actions are complete:
The following conditions occur after the Mitigating Task Scram Actions are complete:
RPV pressure lowered to 600 psig and is currently rising slowly.
* RPV pressure lowered to 600 psig and is currently rising slowly.
RPV water level lowered to  
* RPV water level lowered to -42 inches and is currently rising slowly.
-42 inches and is currently rising slowly.
* Drywell pressure rises to 1.2 psig and stabilizes.
Drywell pressure rises to 1.2 psig and stabilizes.
* All MSIVs are open.
All MSIVs are open.
* RHR-MO-920 {(Div 1), AOG STEAM SUPPLY VLV} is open.
RHR-MO-920 {(Div 1), AOG STEAM SUPPLY VLV} is open.
* RR-AO-741 (INBD ISOL VLV) is open.
RR-AO-741 (INBD ISOL VLV) is open.
* RCIC-MO-15 (INBD ISOL VLV) is open.
RCIC-MO-15 (INBD ISOL VLV) is open. Which valve(s) is/are required to be CLOSED to ensure PCIS initiation is complete?
Which valve(s) is/are required to be CLOSED to ensure PCIS initiation is complete?
A. MSIVs B. RR-AO-741 C. RCIC-MO-15 D. RHR-MO-920   A ns w e r: D. RHR-MO-920 E xp l anati on: Requires knowledge of Mitigating Task Scram Actions and PCIS initiation signals. The Mitigating Task Scram Actions are to depress the manual scram pushbuttons and then take the Reactor Mode Switch out of RUN and place it in REFUEL. The action of taking the Reactor MODE Switch out of RUN ensures the MSIVs do NOT close due to low equalizing header Reactor water level lowered below the level (+3 inches) that initiates a Group 2 isolation. When a Group 2 isolation occurs RHR-MO-920, AOG STEAM SUPPLY VLV is required to close and if the isolation does not complete the valve closure the operator is to close the valve.
A. MSIVs B. RR-AO-741 C. RCIC-MO-15 D. RHR-MO-920 Answer:
No 22 conditions other than those that would result in the initiation of a Group 2 isolation have occurred. Distracters:
D. RHR-MO-920 Explanation:
Requires knowledge of Mitigating Task Scram Actions and PCIS initiation signals. The Mitigating Task Scram Actions are to depress the manual scram pushbuttons and then take the Reactor Mode Switch out of RUN and place it in REFUEL. The action of taking the Reactor MODE Switch out of RUN ensures the MSIVs do NOT close due to low equalizing header pressure. The interlock is MODE Switch in RUN and equalizing header pressure 835 psig.
Reactor water level lowered below the level (+3 inches) that initiates a Group 2 isolation. When a Group 2 isolation occurs RHR-MO-920, AOG STEAM SUPPLY VLV is required to close and if the isolation does not complete the valve closure the operator is to close the valve. No 22


conditions other than those that would result in the initiation of a Group 2 isolation have occurred.
Distracters:
A. This option is incorrect because RPV level did not get low enough for the Group 1 Isolation nor did steam pressure go low enough with the mode switch in RUN to cause a Group 1 isolation. The candidate who does not recall the water level for the Group 1 isolation or who fails to analyze the effect of the mode switch position may choose this answer. If the mode switch were not in SHUTDOWN then the MSIVs would be required to be closed. This option would be correct if the mode switch were in RUN. This answer is plausible because MSIVs will close on a low RPV level.
A. This option is incorrect because RPV level did not get low enough for the Group 1 Isolation nor did steam pressure go low enough with the mode switch in RUN to cause a Group 1 isolation. The candidate who does not recall the water level for the Group 1 isolation or who fails to analyze the effect of the mode switch position may choose this answer. If the mode switch were not in SHUTDOWN then the MSIVs would be required to be closed. This option would be correct if the mode switch were in RUN. This answer is plausible because MSIVs will close on a low RPV level.
 
B. This option is incorrect because RPV level did not get low enough for the Group 7 Isolation.
B. This option is incorrect because RPV level did not get low enough for the Group 7 Isolation. Water level would have to lowe-113" for this isolation to occur. A candidate may confuse the Group 7 isolation with a Group 6 isolation which would occur at the water level provided and would therefore choose this option.
Water level would have to lower to -113 for this isolation to occur. A candidate may confuse the Group 7 isolation with a Group 6 isolation which would occur at the water level provided and would therefore choose this option.
C. This option is incorrect because reactor pressure is not low enough to cause a Group 5 isolation. The candidate who is unsure of the Group 5 isolation setpoint on reactor pressure may choose this option particularly if they confuse the isolation and initiation conditions as a RCIC initiation signal is present. This answer is plausible because the listed valve closes on a low RPV pressure signal.
C. This option is incorrect because reactor pressure is not low enough to cause a Group 5 isolation. The candidate who is unsure of the Group 5 isolation setpoint on reactor pressure may choose this option particularly if they confuse the isolation and initiation conditions as a RCIC initiation signal is present. This answer is plausible because the listed valve closes on a low RPV pressure signal.
Technical Reference(s):
Technical Reference(s):         GOP 2.1.22, Recovery From a Group Isolation, Rev. 59 Proposed references to be provided to applicants during examination:         __None___________
GOP 2.1.22, Recovery From a Group Isolation, Rev. 59 Proposed references to be provided to applicants during examination:
__None___________
Learning Objective:
Learning Objective:
COR002-03-02, Containment
COR002-03-02, Containment
: 21. Given plant conditions, determine if the following should have occurred:
: 21.     Given plant conditions, determine if the following should have occurred:
: a. Any of the PCIS group isolations Question Source:
: a. Any of the PCIS group isolations Question Source:               Bank #                 _______
Bank #   _______ Modified Bank #
Modified Bank #       _______
_______  New   ___X___ Question History: Last NRC Exam
New                   ___X___
____________
Question History:               Last NRC Exam         ____________
Question Cognitive Level:      Memory or Fundamental Knowledge              _____
Comprehension or Analysis                    __X__
10 CFR Part 55 Content:        55.41 _(7)_
Comments:
LOD 3 23


Question Cognitive Level:
24 GROUP 1 25
Memory or Fundamental Knowledge
_____ Comprehension or Analysis  __X__  10 CFR Part 55 Content:
55.41 _(7)_  Comments: LOD  3 23 24 GROUP 1 25 GROUP 2  26 27 28 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
-


==Reference:==
GROUP 2 26


Level  RO  SRO Tier #    _ 2__  _____ Group #  __1 _  _____ K/A #  _ 239002  2.2.12___ Importance Rating
27 28 ES-401                            Sample Written Examination                          Form ES-401-5 Question Worksheet Examination Outline Cross-
_3.7_  _____  239002 SRVs
- 2.2.12 Knowledge of surveillance procedures. (CFR: 41.10 / 45.13)


Q ues ti on: 4 1  While performing Surveillance 6.ADS.201, ADS Manual Valve Actuation (IST) during startup following a refuel outage, the BOP Operator reports that S R V RV-7 1 E Control Switch has been placed to OPEN. Whe n conditi on s s t ab ili ze , which of the following i nd i ca ti on s va li da t e that SRV RV-7 1 E i s full open IAW 6.ADS.201?   A. Main Generator output lowers
==Reference:==
.
Level                  RO              SRO Tier #                  _ 2__          _____
Group #                __1 _          _____
K/A #                  _ 239002 2.2.12___
Importance Rating      _3.7_          _____
239002 SRVs - 2.2.12 Knowledge of surveillance procedures. (CFR: 41.10 / 45.13)
Question: 41 While performing Surveillance 6.ADS.201, ADS Manual Valve Actuation (IST) during startup following a refuel outage, the BOP Operator reports that SRV RV-71 E Control Switch has been placed to OPEN.
When conditions stabilize, which of the following indications validate that SRV RV-71 E is full open IAW 6.ADS.201?
A. Main Generator output lowers.
B. Total indicated steam flow rises.
B. Total indicated steam flow rises.
C. D. Bypass valves throttle in the closed direction.
C. PMIS temperatures within MAX T.
A ns w e r: D. Bypass valves throttle in the closed direction.
D. Bypass valves throttle in the closed direction.
Answer:
D. Bypass valves throttle in the closed direction.
Explanation:
Explanation:
The acceptance criteria for 6.ADS.201 specify the valid parameters for verifying that an SRV   Distracters:
The acceptance criteria for 6.ADS.201 specify the valid parameters for verifying that an SRV has properly opened. The operability limit specified is a change of BPV Position 2%.
A. This option is incorrect because the Main Generator is off-line during SRV testing following a refuel outage. This is plausible if the candidate only thinks of total plant effect without realizing the Main Generator is off
Distracters:
-line. B. This option is incorrect because total steam line flow is not an approved method of verifying SRV position for surveillance purposes. This answer is plausible because the SRVs are located upstream of the main steam line flow measurement devices, therefore total indicated steam flow will actually lower. The candidate who selects this answer could have an incorrect mental model of where SRVs are located in relation to the MS Flow elements
A. This option is incorrect because the Main Generator is off-line during SRV testing following a refuel outage. This is plausible if the candidate only thinks of total plant effect without realizing the Main Generator is off-line.
. C. This option is incorrect because PMIS temperatures are not acceptance criteria for SRV opening during this surveillance. These are actually listed as the closing criteria for this surveillance. This answer is plausible because this is the opening criteria per surveillance procedure 6.SRV.302, Safety Valve and Relief Valve Position Indication Instrument and 29 ADS Pneumatic Supply Check. The candidate who recalls this surveillance criteria and not the surveillance criteria in 6.ADS.201 would select this option.
B. This option is incorrect because total steam line flow is not an approved method of verifying SRV position for surveillance purposes. This answer is plausible because the SRVs are located upstream of the main steam line flow measurement devices, therefore total indicated steam flow will actually lower. The candidate who selects this answer could have an incorrect mental model of where SRVs are located in relation to the MS Flow elements.
Technical Reference(s):
C. This option is incorrect because PMIS temperatures are not acceptance criteria for SRV opening during this surveillance. These are actually listed as the closing criteria for this surveillance. This answer is plausible because this is the opening criteria per surveillance procedure 6.SRV.302, Safety Valve and Relief Valve Position Indication Instrument and 29
6.ADS.201 ADS Manual Valve Actuation Rev.11


Proposed references to be provided to applicants during examination:
ADS Pneumatic Supply Check. The candidate who recalls this surveillance criteria and not the surveillance criteria in 6.ADS.201 would select this option.
__None___________
Technical Reference(s):        6.ADS.201 ADS Manual Valve Actuation Rev.11 Proposed references to be provided to applicants during examination:   __None___________
Learning Objective:
Learning Objective:
Per COR002 02, Nuclear Pressure Relief
Per COR002-16-02, Nuclear Pressure Relief
: 4. Given a Nuclear Pressure Relief system component manipulation, predict and explain the changes in the following parameters:  
: 4. Given a Nuclear Pressure Relief system component manipulation, predict and explain the changes in the following parameters:
: c. Reactor pressure
: c.     Reactor pressure
: f. Reactor power
: f.     Reactor power
: g. Turbine load Question Source:
: g.     Turbine load Question Source:               Bank #             _______
Bank #   _______ Modified Bank #
Modified Bank #     _______
_______  New   ___X___ Question History:
New                 ___X___
Last NRC Exam ____________
Question History:               Last NRC Exam       ____________
Question Cognitive Level:
Question Cognitive Level:       Memory or Fundamental Knowledge       _ __
Memory or Fundamental Knowledge
Comprehension or Analysis             _ X__
_   __ Comprehension or Analysis   _ X__
10 CFR Part 55 Content:         55.41 _(7)_
10 CFR Part 55 Content:
Comments:
55.41 _(7)_ Comments: LOD 3   30 FROM 6.ADS.201 31 32 VALVE NUMBER VALVE STATUS ANNUNCIATOR STATUS (ALM/CLR) LIGHT STATUS (ON/OFF) TRM LIMIT TEMPERATURE
LOD 3 30
( F) BPV POS (%) BPV** POS (%) OPERABILITY LIMITS (%) 9-3-1/A-2 9-3-1/C-1 GREEN AMBER RED MS-TR-166 PMIS PID* MS-RV-71A CLOSED     OFF   1 BPV POS  CLOSED 1 =  BPV POS CLOSED 2  BPV POS  2% OPEN      ON    CLOSED      OFF   2  MS-RV-71B CLOSED     OFF   1 O PEN      ON    CLOSED     OFF   2  MS-RV-71C CLOSED     OFF   1 OPEN      ON    CLOSED     OFF   2  MS-RV-71D CLOSED     OFF   1 O PEN      ON    CLOSED     OFF   2  MS-RV-71E CLOSED     OFF   1 OPEN      ON    CLOSED     OFF   2  MS-RV-71F CLOSED     OFF   1 O PEN      ON    CLOSED     OFF   2  MS-RV-71G CLOSED     OFF   1 O PEN      ON    CLOSED     OFF   2  MS-RV-71H CLOSED      OFF  1  OPEN     ON     CLOSED     OFF   2
 
* Use PMIS Points T142 through T149 for MS
FROM 6.ADS.201 31
-RV-71A through MS
 
-RV-71H, respectively.
32 Shaded Blocks ANNUNCIATOR STATUS           LIGHT STATUS               TEMPERATURE (ALM/CLR)             (ON/OFF)                         (&deg;F)                   BPV** OPERABILITY VALVE    VALVE              9-3-1/C-                       TRM                              BPV POS  POS      LIMITS NUMBER    STATUS 9-3-1/A-2        1     GREEN AMBER RED LIMIT MS-TR-166 PMIS PID*                 (%)    (%)        (%)
** BPV POSITION = BPV POSITION CLOSED 1 - BPV POSITION OPEN
1 CLOSED                                              OFF MS-RV-OPEN                                                ON 71A 2
. Shaded Blocks 33 34 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
CLOSED                                               OFF 1
-
CLOSED                                               OFF MS-RV-OPEN                                                ON 71B 2
CLOSED                                               OFF 1
CLOSED                                               OFF MS-RV-OPEN                                                ON 71C 2
CLOSED                                               OFF 1
CLOSED                                               OFF                                                 BPV POS MS-RV-OPEN                                                ON                                                CLOSED1 =
71D CLOSED                                               OFF                                 2              BPV POS 1             CLOSED2 CLOSED                                               OFF MS-RV-OPEN                                                ON 71E CLOSED                                               OFF                                 2          BPV POS  2%
1 CLOSED                                               OFF MS-RV-OPEN                                                ON 71F 2
CLOSED                                               OFF 1
CLOSED                                               OFF MS-RV-OPEN                                                ON 71G 2
CLOSED                                               OFF 1
CLOSED                                               OFF MS-RV-OPEN                                                 ON 71H 2
CLOSED                                               OFF
* Use PMIS Points T142 through T149 for MS-RV-71A through MS-RV-71H, respectively.
** BPV POSITION = BPV POSITION CLOSED1 - BPV POSITION OPEN.
33
 
34 ES-401                           Sample Written Examination                     Form ES-401-5 Question Worksheet Examination Outline Cross-


==Reference:==
==Reference:==
 
Level               RO               SRO Tier #               __2__           _____
Level   RO SRO Tier #     __2__ _____ Group # __1__ _____ K/A #   _ 259002 K1.07 _
Group #             __1__           _____
__ Importance Rating
K/A #               _ 259002 K1.07 ___
_ 2.6 _ _____ 259002 Reactor Water Level Control  
Importance Rating   _ 2.6 _         _____
- Knowledge of the physical connections and/or cause-effect relationships between reactor water level control system and the following: (CFR: 41.2) K1.07 Rod worth minimizer: Plant
259002 Reactor Water Level Control - Knowledge of the physical connections and/or cause-effect relationships between reactor water level control system and the following:
-Specific Q ues ti on: 4 2  In order to determine its mode of operation, what information must t h e R o d W o rt h M i n i m i ze r (R W M) receive from the Reactor Vessel Level Control System (RVLCS)
(CFR: 41.2) K1.07 Rod worth minimizer: Plant-Specific Question: 42 In order to determine its mode of operation, what information must the Rod Worth Minimizer (RWM) receive from the Reactor Vessel Level Control System (RVLCS)?
?  A. Total main steam flow and feedwater flow rates.
A. Total main steam flow and feedwater flow rates.
B. Total main steam flow and feedwater flow rate mismatch.
B. Total main steam flow and feedwater flow rate mismatch.
C. Time that total main steam flow and feedwater flow are above limits.
C. Time that total main steam flow and feedwater flow are above limits.
D. Reactor power level calculated from total main steam flow and feedwater flow rates.
D. Reactor power level calculated from total main steam flow and feedwater flow rates.
A ns w e r: A. Total main steam flow and feedwater flow rates.
Answer:
A. Total main steam flow and feedwater flow rates.
Explanation:
Explanation:
The RWM operates during plant startup or shutdown. Based upon steam flow and feedwater flow rates, the mode of operation of the RWM is determined. The algorithms come from the Reactor Vessel Level Control System and determine when the RWM starts or stops enforcing predetermined control rod movements. When operating in the Low Power Alarm Point (LPAP) mode, the RVLCS sends the total main steam flow signal to the RWM. The RWM algorithm determines if the total main steam signal has been above 35% for 60 seconds. If these conditions are met, then operation is above the LPAP. Operation in the Transition Zone (TZ)Error! Bookmark not defined.
The RWM operates during plant startup or shutdown. Based upon steam flow and feedwater flow rates, the mode of operation of the RWM is determined. The algorithms come from the Reactor Vessel Level Control System and determine when the RWM starts or stops enforcing predetermined control rod movements. When operating in the Low Power Alarm Point (LPAP) mode, the RVLCS sends the total main steam flow signal to the RWM. The RWM algorithm determines if the total main steam signal has been above 35% for 60 seconds. If these conditions are met, then operation is above the LPAP. Operation in the Transition Zone (TZ)Error! Bookmark not defined. is operation between LPSP and LPAP. Occurs when > 20%
is operation between LPSP and LPAP. Occurs when > 20% total Main Steam flow AND > 20% total feedwater flow (with each condition present for at least 60 seconds) AND and LPAP algorithms in the Reactor Vessel Level Control System, respectively. LPSP is a variable used by the RWM program and the RWM mode will be OPERATING < LPSP when either total Main Steam flow is at or below 20%
total Main Steam flow AND > 20% total feedwater flow (with each condition present for at least 60 seconds) AND 35% total Main Steam flow (for any amount of time) as sensed by the LPSP and LPAP algorithms in the Reactor Vessel Level Control System, respectively. LPSP is a variable used by the RWM program and the RWM mode will be OPERATING < LPSP when either total Main Steam flow is at or below 20% or total feedwater flow is at or below 20%
or total feedwater flow is at or below 20% (this condition being determined by the LPSP algorithm in the RVLCS program) for any period of time. Distracters:
(this condition being determined by the LPSP algorithm in the RVLCS program) for any period of time.
35 B. This option is incorrect because the RWM does not use main steam flow and feedwater flow mismatch for determining its operating mode. This answer is plausible because the RVLCS utilizes steam flow/feedwater flow mismatch algorithms for hardware error signals to be initiated and the candidate may confuse the two.
Distracters:
35
 
B. This option is incorrect because the RWM does not use main steam flow and feedwater flow mismatch for determining its operating mode. This answer is plausible because the RVLCS utilizes steam flow/feedwater flow mismatch algorithms for hardware error signals to be initiated and the candidate may confuse the two.
C. This option is incorrect because the time measurement comes from the RWM algorithm and not the RVLCS. This option is plausible because main steam and feedwater flow rates are timed to determine operating mode. The candidate who does not know which system is measuring the time that flows are at a given level would select this option.
C. This option is incorrect because the time measurement comes from the RWM algorithm and not the RVLCS. This option is plausible because main steam and feedwater flow rates are timed to determine operating mode. The candidate who does not know which system is measuring the time that flows are at a given level would select this option.
D. This option is incorrect because reactor power is not an input to the RWM. This option is plausible because the RWM will utilize main steam flow and feedwater flow rate to approximate reactor power since it is required to be in service when below 9.85% rated power. If a RFP controller or feedwater controller were to fail high, then this would create a false indicated reactor power signal, therefore algorithms must be utilized to adequately justify the current reactor power. The RWM utilizes raw feedwater flow and steam flow data, thus reactor power is not a direct input. The candidate who knows the Technical Specification requirements for BPWS and knows the reactor power level where BPWS constraints must be met would select this answer.
D. This option is incorrect because reactor power is not an input to the RWM. This option is plausible because the RWM will utilize main steam flow and feedwater flow rate to approximate reactor power since it is required to be in service when below 9.85% rated power. If a RFP controller or feedwater controller were to fail high, then this would create a false indicated reactor power signal, therefore algorithms must be utilized to adequately justify the current reactor power. The RWM utilizes raw feedwater flow and steam flow data, thus reactor power is not a direct input. The candidate who knows the Technical Specification requirements for BPWS and knows the reactor power level where BPWS constraints must be met would select this answer.
Technical Reference(s):
Technical Reference(s):             IOP 4.2, Rod Worth Minimizer, Rev. 29 Proposed references to be provided to applicants during examination:       __None___________
IOP 4.2, Rod Worth Minimizer, Rev. 29 Proposed references to be provided to applicants during examination:
__None___________
Learning Objective:
Learning Objective:
Per COR002 02, Reactor Vessel Level Control
Per COR002-32-02, Reactor Vessel Level Control
: 2. Describe the interrelationship between RVLC and the following:
: 2.     Describe the interrelationship between RVLC and the following:
: j. RWM Question Source:
: j.     RWM Question Source:               Bank #
Bank #   Modified Bank #
Modified Bank #
New X Question History:
New           X Question History:             Last NRC Exam         ____________
Last NRC Exam
Question Cognitive Level:     Memory or Fundamental Knowledge             __X__
____________
Comprehension or Analysis                   __ __
Question Cognitive Level: Memory or Fundamental Knowledge
10 CFR Part 55 Content:       55.41 _(7)_
__X__ Comprehension or Analysis   __ __ 10 CFR Part 55 Content:
Comments:
55.41 _(7)_ Comments: LOD 3 36 37 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
LOD 3 36
-
 
37 ES-401                           Sample Written Examination                       Form ES-401-5 Question Worksheet Examination Outline Cross-


==Reference:==
==Reference:==
 
Level                 RO             SRO Tier #                 _2__           _____
Level   RO SRO Tier #     _2__ _____ Group # _1 __ _____ K/A #   _261000 K3.02 _
Group #               _1 __         _____
__ Importance Rating
K/A #                 _261000 K3.02 ___
_3.6_ _____ 261000 SGTS  
Importance Rating     _3.6_         _____
- Knowledge of the effect that a loss or malfunction of the standby gas treatment system will have on the following: (CFR: 41.7) K3.02 Off
261000 SGTS - Knowledge of the effect that a loss or malfunction of the standby gas treatment system will have on the following: (CFR: 41.7) K3.02 Off-site release rate Question: 43 An accident occurred that resulted in fuel failure and a breach of the reactor coolant system boundary. The A SGT train is in service to support Primary containment venting when the following alarms:
-site release rate
* Annunciator K-1/A-2 SGT A HIGH MOISTURE What SGT system component is primarily affected and how is the offsite release rate affected?
 
Question: 43
 
An accident occurred that resulted in fuel failure and a breach of the reactor coolant system boundary. The "A" SGT train is in service to support Primary containment venting when the following alarms:
Annunciator K
-1/A-2 SGT A HIGH MOISTURE What SGT system component is primarily affected and how is the offsite release rate affected?
 
A. The Charcoal Filter and release rate rises primarily due to iodine activity.
A. The Charcoal Filter and release rate rises primarily due to iodine activity.
B. The Charcoal Filter and release rate rises primarily due to particulate activity.
B. The Charcoal Filter and release rate rises primarily due to particulate activity.
C. The High Efficiency Inlet Filter and the release rate rises primarily due to iodine activity.
C. The High Efficiency Inlet Filter and the release rate rises primarily due to iodine activity.
D. The High Efficiency Inlet Filter and the release rate rises primarily due to particulate activity.  
D. The High Efficiency Inlet Filter and the release rate rises primarily due to particulate activity.
 
Answer:
Answer: A. The charcoal filter and release rate rises primarily due to iodine activity.
A. The charcoal filter and release rate rises primarily due to iodine activity.
Explanation:
Explanation:
This matches the KA due to moisture content within the SGT impacts the amount of Iodine released. The quantity of radioactive airborne contaminants is reduced as the air passes through the SGT train. The SGT has a design flow rate of 1780 cfm. The train is comprised of multiple compartments including a moisture separator and electric heating element upstream of the HEPA filters and activated carbon iodine adsorber. The heating system is designed to reduce the relative humidity of the inlet stream from 100% to 70% when the SGT system is operating. The charcoal filters are iodide
This matches the KA due to moisture content within the SGT impacts the amount of Iodine released. The quantity of radioactive airborne contaminants is reduced as the air passes through the SGT train. The SGT has a design flow rate of 1780 cfm. The train is comprised of multiple compartments including a moisture separator and electric heating element upstream of the HEPA filters and activated carbon iodine adsorber. The heating system is designed to reduce the relative humidity of the inlet stream from 100% to 70% when the SGT system is operating. The charcoal filters are iodide-impregnated activated carbon filters capable of removing in excess of 97.5% of the methyl iodide in the air stream under entering conditions of 70% relative humidity. As relative humidity rises above 70%, the efficiency of the carbon filters decreases thus allowing more iodine components to pass. The decreased hold-up time from adsorption on the carbon filters result in elevated radiation values due to the lack of radioactive decay of the iodine. The HEPA filter is designed to remove particulate components of >0.30 microns which are impinged on the filter and is virtually unaffected by relative humidity.
-impregnated activated carbon filters capable of removing in excess of 97.5% of the methyl iodide in the air stream under entering conditions of 70% relative humidity. As relative humidity rises above 70%, the efficiency of the carbon filters decreases thus allowing more iodine components to pass. The decreased hold
At high humidity values above 70%, the charcoal filter becomes less efficient for adsorbing the iodine thus raising the Committed Dose Equivalent (CDE) for off-site release rates.
-up time from adsorption on the carbon filters result in elevated radiation values due to the lack of radioactive decay of the iodine. The HEPA filter is designed to remove particulate components of >0.30 microns which are impinged on the filter and is virtually unaffected by relative humidity.
38
At high humidity values above 70%, the charcoal filter becomes less efficient for adsorbing the iodine thus raising the Committed Dose Equivalent (CDE) for off
-site release rates.
38 Distracters:
B. This option is incorrect because the primary SGT system component that is affected by high humidity is the charcoal filter. The charcoal filter needs the SGT heater to ensure that the relative humidity of the gas stream entering the charcoal filter is sufficiently low to allow the charcoal filter to function efficiency with a relatively high adsorption rate for iodine. The roughing filter and HEPA filter are the primary filtering units for particulate material. This is a plausible selection if the candidate believes that the charcoal filter is used primarily for particulates.


C. This option is incorrect because although iodine would be a concern with the high humidity the high efficiency filter is capable of performing its function of removing particulates with high humidity. The HEPA filter has no discernable effect on iodides. This is a plausible selection if the candidate confuses the primary function of the high efficiency inlet filter and the charcoal filter.
Distracters:
D. This option is incorrect because the primary component affected is the charcoal filter which adsorbs the iodine components, vs. the HEPA filter which traps particulates. Very little particulate activity will pass through the entire SGT because after the HEPA filter, there is a carbon filter and then a second HEPA filter. This selection is plausible if a candidate believes that the HEPA filter efficiency is affected by the high humidity.
B.      This option is incorrect because the primary SGT system component that is affected by high humidity is the charcoal filter. The charcoal filter needs the SGT heater to ensure that the relative humidity of the gas stream entering the charcoal filter is sufficiently low to allow the charcoal filter to function efficiency with a relatively high adsorption rate for iodine. The roughing filter and HEPA filter are the primary filtering units for particulate material. This is a plausible selection if the candidate believes that the charcoal filter is used primarily for particulates.
Technical Reference(s):
C.     This option is incorrect because although iodine would be a concern with the high humidity the high efficiency filter is capable of performing its function of removing particulates with high humidity. The HEPA filter has no discernable effect on iodides.
USAR V, Section 3.3.4 Procedure 2.2.73, Standby Gas Treatment System, Rev. 52.
This is a plausible selection if the candidate confuses the primary function of the high efficiency inlet filter and the charcoal filter.
Proposed references to be provided to applicants during examination:
D.     This option is incorrect because the primary component affected is the charcoal filter which adsorbs the iodine components, vs. the HEPA filter which traps particulates. Very little particulate activity will pass through the entire SGT because after the HEPA filter, there is a carbon filter and then a second HEPA filter. This selection is plausible if a candidate believes that the HEPA filter efficiency is affected by the high humidity.
___None__________
Technical Reference(s):           USAR V, Section 3.3.4 Procedure 2.2.73, Standby Gas Treatment System, Rev. 52.
Proposed references to be provided to applicants during examination:             ___None__________
Learning Objective:
Learning Objective:
COR002-28-2, Standby Gas Treatment System
COR002-28-2, Standby Gas Treatment System
: 7. Given a specific Standby Gas Treatment System malfunction, determine the effect on any of the following:
: 7.     Given a specific Standby Gas Treatment System malfunction, determine the effect on any of the following:
: b. Off-site release rate Question Source:
: b.       Off-site release rate Question Source:                   Bank #
Bank #   Modified Bank #
Modified Bank #
New   X Question History: Last NRC Exam Question Cognitive Level:
New                   X Question History:                 Last NRC Exam Question Cognitive Level:         Memory or Fundamental Knowledge               _____
Memory or Fundamental Knowledge
Comprehension or Analysis                     __X__
_____ Comprehension or Analysis   __X__ 10 CFR Part 55 Content:
10 CFR Part 55 Content:           55.41 _(7)_
55.41 _(7)_
Comments:
Comments: LOD 3   39 40 From Flanders FFI Vendor manual data:
LOD 3 39
HFATS Test Los Alamos National Laboratory developed an alternate test method in the 1980's under contract to the U.S. Department of Energy (DOE). It is often referred to as the HFATS test (High Flow Alternative Test System). It was developed specifically to test filters rated at airflows higher than 100 CFM, but it can be used for lower flows. It is only limited by the size of the system fan and the aerosol generator output. This method was later standardized in the publication of a recommended practice, IEST=
-RP-CC007.1, Testing ULPA Filters, published by the Institute of Environmental Sciences and Technology. Currently, ASME AG
-1 Section FC allows for testing by this method. The filter is challenged with an acceptable polydispersed oil aerosol and the penetration through the filter is measured with a Laser Particle Counter. The Particle Counter counts and sizes individual droplets in a size range from 0.1 to 3.0 micrometers in diameter.
The ratio of the downstream counts to the upstream counts in each size range is the penetration. Although this value is not equal to the penetration measured by the Q
-107, research performed by Los Alamos National Laboratory verified it to be very similar and the method to be an acceptable alternative to the penetration measured by Mil-Std-282 Test Method. Humidity and Water Resistance
. HEPA filter media will tolerate high humidity (95% +/-5%) and some direct wetting, but excessive moisture, either from air borne droplets or condensation, can plug the filter and result in failure by over-pressure. Metal case filters are more suitable for moisture laden atmospheres. Because aluminum separators can corrode in some environments and slough particles downstream of the filter, Separatorless filters are also recommended for l=moist conditions, except in high
-temperature or caustic application.
41 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
-


==Reference:==
40 From Flanders FFI Vendor manual data:
HFATS Test Los Alamos National Laboratory developed an alternate test method in the 1980s under contract to the U.S. Department of Energy (DOE). It is often referred to as the HFATS test (High Flow Alternative Test System). It was developed specifically to test filters rated at airflows higher than 100 CFM, but it can be used for lower flows. It is only limited by the size of the system fan and the aerosol generator output. This method was later standardized in the publication of a recommended practice, IEST=-RP-CC007.1, Testing ULPA Filters, published by the Institute of Environmental Sciences and Technology. Currently, ASME AG-1 Section FC allows for testing by this method. The filter is challenged with an acceptable polydispersed oil aerosol and the penetration through the filter is measured with a Laser Particle Counter. The Particle Counter counts and sizes individual droplets in a size range from 0.1 to 3.0 micrometers in diameter. The ratio of the downstream counts to the upstream counts in each size range is the penetration. Although this value is not equal to the penetration measured by the Q-107, research performed by Los Alamos National Laboratory verified it to be very similar and the method to be an acceptable alternative to the penetration measured by Mil-Std-282 Test Method.
Humidity and Water Resistance.
HEPA filter media will tolerate high humidity (95% +/-5%) and some direct wetting, but excessive moisture, either from air borne droplets or condensation, can plug the filter and result in failure by over-pressure. Metal case filters are more suitable for moisture laden atmospheres.
Because aluminum separators can corrode in some environments and slough particles downstream of the filter, Separatorless filters are also recommended for l=moist conditions, except in high-temperature or caustic application.
41


Level  RO  SRO Tier #    __2__  _____ Group #  __1__  _____ K/A #  _262001  K2.01  _
ES-401                            Sample Written Examination                      Form ES-401-5 Question Worksheet Examination Outline Cross-
__ Importance Rating
_ 3.3 _  _____  262001 AC Electrical Distribution
- Knowledge of electrical power supplies to the following: (CFR: 41.7) K2.01 Off
-site sources of power.


Q ues ti on: 4 4  W ha t i s a po w e r sou r c e t o t h e E m e r genc y S e r v i c e S t a ti o n T r ans f o r m e r? A. D ir ec tl y fr o m t h e Aubu r n li ne. B. D ir ec tl y fr o m t h e C o r nfi e l d Substation. C. 16 1 K V subs t a ti o n v i a t h e T 6 trans f o r m e r. D. 16 1 K V subs t a ti o n v i a t h e T 7 trans f o r m e r. A ns w e r: C. 16 1 K V subs t a ti o n v i a t h e T 6 trans f o r m e r.
==Reference:==
E xp l anati on: D u ri n g normal s t a ti o n operation, the E m e r genc y Service Station Transformer (ESST) i s energized by the 69 k V transmission li n e from the 69 kV B a y o f the 16 1 k V Subs t ation through Air B r ea k Switch 5298. T h e E m e rgenc y Transformer supply can b e a ligne d t o either the Cooper 16 1 kV S ys t e m v i a Transformer T6 o r t o the OPP D 69 kV li ne. Distracters:
Level                RO              SRO Tier #              __2__          _____
Group #              __1__          _____
K/A #                _262001 K2.01 ___
Importance Rating    _ 3.3 _        _____
262001 AC Electrical Distribution - Knowledge of electrical power supplies to the following: (CFR: 41.7) K2.01 Off-site sources of power.
Question: 44 What is a power source to the Emergency Service Station Transformer?
A. Directly from the Auburn line.
B. Directly from the Cornfield Substation.
C. 161 KV substation via the T6 transformer.
D. 161 KV substation via the T7 transformer.
Answer:
C. 161 KV substation via the T6 transformer.
Explanation:
During normal station operation, the Emergency Service Station Transformer (ESST) is energized by the 69 kV transmission line from the 69 kV Bay of the 161 kV Substation through Air Break Switch 5298. The Emergency Transformer supply can be aligned to either the Cooper 161 kV System via Transformer T6 or to the OPPD 69 kV line.
Distracters:
A. This option is incorrect because the Auburn line connects with the 161 kV Substation. From the 161 kV Substation the power must go through the 69 kV Bay and the T6 transformer to connect with the ESST. The Auburn line does not connect directly with the ESST. The 161 kV switchyard has recently gone through a major design change and the candidate may not fully understand the new configuration. This answer is plausible because the Auburn line does supply the ESST, just not directly.
A. This option is incorrect because the Auburn line connects with the 161 kV Substation. From the 161 kV Substation the power must go through the 69 kV Bay and the T6 transformer to connect with the ESST. The Auburn line does not connect directly with the ESST. The 161 kV switchyard has recently gone through a major design change and the candidate may not fully understand the new configuration. This answer is plausible because the Auburn line does supply the ESST, just not directly.
B. This option is incorrect because the 69 kV Cornfield Substation has been removed during the 161 kV Substation major design change. The candidate who recalls the old arrangement would select this answer. This answer is plausible because the Cornfield Substation was previously used to directly feed the ESST.
B. This option is incorrect because the 69 kV Cornfield Substation has been removed during the 161 kV Substation major design change. The candidate who recalls the old arrangement would select this answer. This answer is plausible because the Cornfield Substation was previously used to directly feed the ESST.
D. This option is incorrect because the T7 transformer does not feed the ESST. The T7 transformer is part of the AC distribution system and is a new addition to the station so the candidate who confuses it with the T6 transformer would select this answer.
D. This option is incorrect because the T7 transformer does not feed the ESST. The T7 transformer is part of the AC distribution system and is a new addition to the station so the candidate who confuses it with the T6 transformer would select this answer.
42 Technical Reference(s):
42
SOP 2.2.17, ESST, Rev. 64
 
Proposed references to be provided to applicants during examination:
__None___________


Technical Reference(s):              SOP 2.2.17, ESST, Rev. 64 Proposed references to be provided to applicants during examination: __None___________
Learning Objective:
Learning Objective:
COR001-01-01, OPS AC Distribution
COR001-01-01, OPS AC Distribution
: 7. State the electrical power supplies to the following:
: 7. State the electrical power supplies to the following:
: a. Off-Site Sources of Power Question Source:
: a.     Off-Site Sources of Power Question Source:             Bank #                 _   _
Bank #   _     _ Modified Bank #
Modified Bank #         __     _
__       _ New   ___ X ___ Question History:
New                     ___ X ___
Last NRC Exam Question Cognitive Level:
Question History:             Last NRC Exam Question Cognitive Level:     Memory or Fundamental Knowledge       __X__
Memory or Fundamental Knowledge
Comprehension or Analysis             __ __
__X__ Comprehension or Analysis   __ __
10 CFR Part 55 Content:       55.41 _(7)_
10 CFR Part 55 Content:
Comments:
55.41 _(7)_
LOD 3 43
Comments: LOD 3   43 44 ES-401   Sample Written Examination Question Worksheet Form ES-401-5   Examination Outline Cross
 
-
44 ES-401                 Sample Written Examination Question Worksheet Form ES-401-5 Examination Outline Cross-


==Reference:==
==Reference:==
 
Level                   RO             SRO Tier #                     2 Group #                   1 K/A #                       262002 K4.01 Importance Rating         3.1 262002 Uninterruptible Power Supply (A.C. / D.C.)
Level   RO SRO       Tier #   2         Group #   1         K/A #   262002 K4.01       Importance Rating  
 
===3.1 262002===
Uninterruptible Power Supply (A.C. / D.C.)
Ability to monitor automatic operations of the UNINTERRUPTABLE POWER SUPPLY (A.C./D.C.) including:
Ability to monitor automatic operations of the UNINTERRUPTABLE POWER SUPPLY (A.C./D.C.) including:
K4.01Transfer from preferred to alternate source.
K4.01Transfer from preferred to alternate source.
Q ues ti on: 4 5  Th e p l an t i s i n a no r m a l f u l l po w e r e l ectri ca l li neup. Th e f o ll o w i n g a l a r m i s rece ived: NO BREAK INVERTER 1A VOLT FAILURE PANEL/WINDOW:
Question: 45 The plant is in a normal full power electrical lineup. The following alarm is received:
C-4/E-7 What is the current source of power to the NBPP? A. MCC-L v i a a s t e p do w n tr ans f o r m e r an d t h e inve rt e r cab i ne t s t a ti c s w itch. B. MCC-R v i a a s t e p do w n tr ans f o r m e r an d t h e inve rt e r cab i ne t s t a ti c s w it ch. C. MCC-L v i a a s t e p do w n tr ans f o r m e r and then directly to the power panel. D. MCC-R v i a a s t e p do w n tr ans f o r m e r and then directly to the power panel
NO BREAK PANEL/WINDOW:
. A ns w e r: B. MCC-R v i a a s t e p do w n tr ans f o r m e r an d t h e inve rt e r cab i ne t s t a ti c s w it ch. Explanation:
INVERTER 1A C-4/E-7 VOLT FAILURE What is the current source of power to the NBPP?
System Operating Procedure 2.2.22, describes NBPP automatic transfer to MCC
A. MCC-L via a step down transformer and the inverter cabinet static switch.
-R. See Attachment 1 Steps 1.2.5 and 2.3.
B. MCC-R via a step down transformer and the inverter cabinet static switch.
Power to the No
C. MCC-L via a step down transformer and then directly to the power panel.
-Break Power Panel (NBPP) #1 is normally supplied from 250 VDC bus 1A through inverter 1A and a static switch. The inverter failure alarm indicates that the power into or out of the inverter is failed which causes the NBPP to transfer to MCC
D. MCC-R via a step down transformer and then directly to the power panel.
-R. MCC-R powers the NBPP through a step down (115V AC) transformer to the static switch in the inverter cabinet.
Answer:
B. MCC-R via a step down transformer and the inverter cabinet static switch.
Explanation:
System Operating Procedure 2.2.22, describes NBPP automatic transfer to MCC-R. See Steps 1.2.5 and 2.3.
Power to the No-Break Power Panel (NBPP) #1 is normally supplied from 250 VDC bus 1A through inverter 1A and a static switch. The inverter failure alarm indicates that the power into or out of the inverter is failed which causes the NBPP to transfer to MCC-R. MCC-R powers the NBPP through a step down (115V AC) transformer to the static switch in the inverter cabinet.
The static switch can also be operated with the NBPP PWR TRANSFER switch on Panel C (MCC or IVTR) or by pressing the ALTERNATE SOURCE SUPPLYING LOAD or INVERTER SUPPLYING LOAD button on the inverter. The NBPP power can also be transferred by placing the MANUAL BYPASS SWITCH on the inverter to ALTERNATE SOURCE TO LOAD or 45


The static switch can also be operated with the NBPP PWR TRANSFER switch on Panel C (MCC or IVTR) or by pressing the ALTERNATE SOURCE SUPPLYING LOAD or INVERTER SUPPLYING LOAD button on the inverter. The NBPP power can also be transferred by placing the MANUAL BYPASS SWITCH on the inverter to ALTERNATE SOURCE TO LOAD or 45 NORMAL OPERATION per SOP 2.2.22. The static switch and manual bypass switch transfer the NBPP power supply in a make before break logic.
NORMAL OPERATION per SOP 2.2.22. The static switch and manual bypass switch transfer the NBPP power supply in a make before break logic.
Distracters:
Distracters:
 
A. This answer is incorrect because MCC-L powers the PMIS-UPS inverter as an alternate supply. This answer is plausible because the PMIS-UPS is a different uninterruptable power supply at the station. The candidate who confuses the NBPP and PMIS-UPS panels would select this answer.
A. This answer is incorrect because MCC
C. This answer is incorrect because MCC-L powers the PMIS-UPS inverter as an alternate supply. This answer is plausible because the PMIS-UPS is a different uninterruptable power supply at the station. The candidate who confuses the NBPP and PMIS-UPS panels would select this answer.
-L powers the PMIS
-UPS inverter as an alternate supply. This answer is plausible because the PMIS
-UPS is a different uninterruptable power supply at the station. The candidate who confuses the NBPP and PMIS
-UPS panels would select this answer.
 
C. This answer is incorrect because MCC
-L powers the PMIS
-UPS inverter as an alternate supply. This answer is plausible because the PMIS
-UPS is a different uninterruptable power supply at the station. The candidate who confuses the NBPP and PMIS
-UPS panels would select this answer.
 
D. MCC-R automatically powers NBPP through the Static switch. To feed the NBPP directly requires a MANUAL BYPASS SWITCH to be manipulated at the inverter cabinet. This answer is plausible because powering NBPP by bypassing the static switch is a means of powering the panel. The candidate who cannot recall the different configuration arrangements in the inverter cabinet would select this answer.
D. MCC-R automatically powers NBPP through the Static switch. To feed the NBPP directly requires a MANUAL BYPASS SWITCH to be manipulated at the inverter cabinet. This answer is plausible because powering NBPP by bypassing the static switch is a means of powering the panel. The candidate who cannot recall the different configuration arrangements in the inverter cabinet would select this answer.
Technical Reference(s):
Technical Reference(s):     Procedure 2.3_C-4 (Panel C - Annunciator C-4), Rev. 31 Proposed references to be provided to applicants during examination:         None Learning Objective: COR0010102 AC Electrical Distribution COR0010102001090G Describe the AC Electrical Distribution System design feature(s) and/or interlock(s) that provide for the following: Transfer from preferred power to alternate power supplies Question Source:          Bank #                    25667 Modified Bank #
Procedure 2.3_C
New Question History:        Last NRC Exam        2012 Question Cognitive Level:          Memory or Fundamental Knowledge Comprehension or Analysis                    X 10 CFR Part 55 Content:        55.41    7 55.43 Comments:
-4 (Panel C  
LOD 3 46
- Annunciator C
-4), Rev. 31 Proposed references to be provided to applicants during examination:
None Learning Objective:
COR0010102 AC Electrical Distribution COR0010 102001090G Describe the AC Electrical Distribution System design feature(s) and/or interlock(s) that provide for the following: Transfer from preferred power to alternate power supplies


Question Source:
47 48 49 ES-401                             Sample Written Examination                       Form ES-401-5 Question Worksheet Examination Outline Cross-
Bank #  25667    Modified Bank #
New      Question History:
Last NRC Exam 2012 Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41  7  55.43    Comments:  LOD 3  46 47 48 49 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
-


==Reference:==
==Reference:==
 
Level                 RO             SRO Tier #                 __2__         _____
Level   RO SRO Tier #     __2__ _____ Group # __1__ _____ K/A #   _263000 K5.01 _
Group #               __1__         _____
__ Importance Rating
K/A #                 _263000 K5.01 ___
_ 2.6_ _____ 263000 DC Electrical Distribution  
Importance Rating     _ 2.6_         _____
- Knowledge of the operational implications of the following concepts as they apply to D.C. electrical distribution: (CFR: 41.5) K5.01 Hydrogen generation during battery charging.
263000 DC Electrical Distribution - Knowledge of the operational implications of the following concepts as they apply to D.C. electrical distribution: (CFR: 41.5) K5.01 Hydrogen generation during battery charging.
Question: 46 The plant is operating at 100% power with the following conditions:
Question: 46 The plant is operating at 100% power with the following conditions:
Battery charge in progress following the replacement of a cell in the Division II 250 VDC station battery.
* Battery charge in progress following the replacement of a cell in the Division II 250 VDC station battery.
A complete loss of Battery room ventilation occurs.
* A complete loss of Battery room ventilation occurs.
What is the immediate operational concern for the present conditions?
What is the immediate operational concern for the present conditions?
A. Hydrogen buildup in the battery room is a fire hazard.
A. Hydrogen buildup in the battery room is a fire hazard.
B. Hydrogen buildup in the battery room displaces oxygen.
B. Hydrogen buildup in the battery room displaces oxygen.
C. Battery room temperature rise causes cell overheating and loss of electrolyte level.
C. Battery room temperature rise causes cell overheating and loss of electrolyte level.
D. Battery room temperature rise leads to cell reversal conditions in the new replacement battery cell.
D. Battery room temperature rise leads to cell reversal conditions in the new replacement battery cell.
Answer: A. Hydrogen buildup in the battery room is a fire hazard.
Answer:
 
A. Hydrogen buildup in the battery room is a fire hazard.
Explanation:
Explanation:
Battery room ventilation is required to maintain room temperature and disperse hydrogen generated from battery charging. In the case of a battery charge in progress, the hydrogen removal function is the operational concern. The hydrogen buildup is a fire/explosive hazard and concern. Although hydrogen can displace oxygen in a space, it becomes a fire hazard at much lower concentrations.
Battery room ventilation is required to maintain room temperature and disperse hydrogen generated from battery charging. In the case of a battery charge in progress, the hydrogen removal function is the operational concern. The hydrogen buildup is a fire/explosive hazard and concern. Although hydrogen can displace oxygen in a space, it becomes a fire hazard at much lower concentrations.
The candidate should understand that hydrogen removal is the concern and that the buildup of hydrogen is a fire/explosion hazard.
The candidate should understand that hydrogen removal is the concern and that the buildup of hydrogen is a fire/explosion hazard.
Distracters: B. This option is incorrect because even though hydrogen buildup is a concern in the battery room, it is the fire/explosive hazard that is the concern and not displacement of oxygen as an explosive or fire hazard would exist for a significant period of time before displacement of sufficient oxygen to cause a problem could occur, if at all. Since operators do deal with confined spaces and habitable environments the operator could believe that the battery room 50 fan is there to prevent displacement of oxygen. However this is plausible because displacement of oxygen is an issue with other gases.
Distracters:
B. This option is incorrect because even though hydrogen buildup is a concern in the battery room, it is the fire/explosive hazard that is the concern and not displacement of oxygen as an explosive or fire hazard would exist for a significant period of time before displacement of sufficient oxygen to cause a problem could occur, if at all. Since operators do deal with confined spaces and habitable environments the operator could believe that the battery room 50
 
fan is there to prevent displacement of oxygen. However this is plausible because displacement of oxygen is an issue with other gases.
C. This option is incorrect because even though battery room temperature may rise the concern with the loss of ventilation is not the temperature but the hydrogen. But because there could be a room temperature rise associated with the loss of ventilation the candidate may believe that the reason is temperature and a high rate of electrolyte loss.
C. This option is incorrect because even though battery room temperature may rise the concern with the loss of ventilation is not the temperature but the hydrogen. But because there could be a room temperature rise associated with the loss of ventilation the candidate may believe that the reason is temperature and a high rate of electrolyte loss.
D. This option is incorrect because even though battery room temperature may rise the concern with the loss of ventilation is not the temperature but the hydrogen. A candidate may not know the contributory factors associated with cell reversal but may know that it is a serious battery operational concern. That candidate may also believe that high cell temperature could contribute to cell reversal and would choose this option.
D. This option is incorrect because even though battery room temperature may rise the concern with the loss of ventilation is not the temperature but the hydrogen. A candidate may not know the contributory factors associated with cell reversal but may know that it is a serious battery operational concern. That candidate may also believe that high cell temperature could contribute to cell reversal and would choose this option.
Technical Reference(s):                TS Basis B.3.8.1                          _____________
(Attach if not previously provided)    SOP 2.2.24.1, R13 250VDC Electrical System (Div. 1)__
(including version/revision number) 0.39 R50 Hot Work, ACD0150507R04-L-Batteries &
Current Converters.
Procedure 2.2.38 (HVAC Control Building), Rev. 40 Proposed references to be provided to applicants during examination:            None _
Learning Objective:
Per COR002-07-02, DC Electrical Distribution
: 10. Briefly describe the following concepts as they apply to DC Electrical Distribution System.
: a.      Hydrogen generation during battery charging.
Question Source:                Bank #                _      _
Modified Bank #        _21329_ (See attached)
New                    ___ ___
Question History:              Last NRC Exam          __CNS 2005__
Question Cognitive Level:      Memory or Fundamental Knowledge              _____
Comprehension or Analysis                    __X__
10 CFR Part 55 Content:        55.41 _(5)_
Comments:
LOD 2 51


Technical Reference(s):
52 53 54 55 56 57 ES-401                             Sample Written Examination                       Form ES-401-5 Question Worksheet Examination Outline Cross-
TS Basis B.3.8.1
_____________
(Attach if not previously provided)
SOP 2.2.24.1, R13 250VDC Electrical System (Div. 1)
__ (including version/revision number) 0.39 R50 Hot Work , ACD0150507R04
-L-Batteries & Current Converters.
Procedure 2.2.38 (HVAC Control Building), Rev. 40 Proposed references to be provided to applicants during examination:
None  _  Learning Objective:
Per COR002 02, DC Electrical Distribution
: 10. Briefly describe the following concepts as they apply to DC Electrical Distribution System. a. Hydrogen generation during battery charging.
Question Source:
Bank #  _ _ Modified Bank #
_21329_ (See attached)
New  ___  ___  Question History:
Last NRC Exam __CNS  2005__  Question Cognitive Level:
Memory or Fundamental Knowledge
_____ Comprehension or Analysis  __X__  10 CFR Part 55 Content:
55.41 _(5)_
Comments: LOD 2 51 52 53 54 55 56 57 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
-


==Reference:==
==Reference:==
 
Level                 RO               SRO Tier #                 __2__           _____
Level   RO SRO Tier #     __2__ _____ Group # __1__ _____ K/A #   _ 264000 K6.03 _
Group #               __1__           _____
__ Importance Rating
K/A #                 _ 264000 K6.03 ___
_ 3.5 _ _____ 264000 EDGs  
Importance Rating     _ 3.5 _         _____
- Knowledge of the effect that a loss or malfunction of the following will have on the emergency generators (diesel): (CFR: 41.7) K6.03 Lube oil pumps
264000 EDGs - Knowledge of the effect that a loss or malfunction of the following will have on the emergency generators (diesel): (CFR: 41.7) K6.03 Lube oil pumps Question: 47 DG1 is manually started for post maintenance testing:
 
* DIESEL GEN 1 BKR EG1 is open.
Question: 47
* Engine driven lube oil pump fails (no oil flow).
 
DG1 is manually started for post maintenance testing:
DIESEL GEN 1 BKR EG1 is open.
Engine driven lube oil pump fails (no oil flow).
What condition FIRST trips the Diesel Generator?
What condition FIRST trips the Diesel Generator?
A. Low Lube Oil Pressure B. High Lube Oil Temperature C. Low Turbocharger Oil Pressure D. High Thrust Bearing Oil Temperature Answer: A. Low Lube Oil Pressure Explanation:
A. Low Lube Oil Pressure B. High Lube Oil Temperature C. Low Turbocharger Oil Pressure D. High Thrust Bearing Oil Temperature Answer:
When the diesel generator is manually started all the diesel generator trips are in effect. With the loss of the engine driven oil pump and the loss of all lube oil pressure, the diesel generator trips at <20 psig lube oil pressure. This trip is only bypassed on an automatic start. This signal is the first that trips the DG and anything beyond that cannot trip the DG because it is already tripped. Distractors:
A. Low Lube Oil Pressure Explanation:
B. This option is incorrect because the diesel generator does not trip on high oil temperature. The DG will trip on high bearing or connecting rod temperature(s). With the loss of lube oil pressure, it is reasonable to assume a relationship between a loss of oil flow and elevated metal temperatures. A low oil pressure trip would occur before engine oil temperature became elevated. This selection is plausible if the candidate does not remember the relationship between the trip set points of the low lube oil pressure in relation to the relative length of time it would take the bearing metal temperature to rise.
When the diesel generator is manually started all the diesel generator trips are in effect. With the loss of the engine driven oil pump and the loss of all lube oil pressure, the diesel generator trips at <20 psig lube oil pressure. This trip is only bypassed on an automatic start. This signal is the first that trips the DG and anything beyond that cannot trip the DG because it is already tripped.
58 C. This option is incorrect because the turbocharger oil pressure trip occurs at a lower pressure than does the engine oil pressure trip. So as oil pressure falls, the engine oil trip point would be reached first because the low Turbocharger Oil Pressure setpoint varies with load of the machine. The Turbocharger Low Oil trip setpoint is calculated based on the following formula: 4 psig + (0.49 x turbocharger discharge air pressure [in Hg]). This selection is plausible because a candidate may not recognize the relatively low turbocharger discharge air pressures or the configuration of the oil supply to the turbocharge
Distractors:
: r.
B. This option is incorrect because the diesel generator does not trip on high oil temperature.
D. This option is incorrect because this type of catastrophic condition would occur after the oil pressure had tripped the diesel generator. This selection is plausible because the loss of lube oil could cause the turbocharger thrust bearing temperature to elevate to the bearing failure point but typically only on the supply line to the turbocharger. Thus, a candidate could reason that this would trip the diesel generator.
The DG will trip on high bearing or connecting rod temperature(s). With the loss of lube oil pressure, it is reasonable to assume a relationship between a loss of oil flow and elevated metal temperatures. A low oil pressure trip would occur before engine oil temperature became elevated. This selection is plausible if the candidate does not remember the relationship between the trip set points of the low lube oil pressure in relation to the relative length of time it would take the bearing metal temperature to rise.
58


Technical Reference(s):
C. This option is incorrect because the turbocharger oil pressure trip occurs at a lower pressure than does the engine oil pressure trip. So as oil pressure falls, the engine oil trip point would be reached first because the low Turbocharger Oil Pressure setpoint varies with load of the machine. The Turbocharger Low Oil trip setpoint is calculated based on the following formula: 4 psig + (0.49 x turbocharger discharge air pressure [in Hg]). This selection is plausible because a candidate may not recognize the relatively low turbocharger discharge air pressures or the configuration of the oil supply to the turbocharger.
SOP 2.2.20, Standby AC Power System (Diesel (Attach if not previously provided)
D. This option is incorrect because this type of catastrophic condition would occur after the oil pressure had tripped the diesel generator. This selection is plausible because the loss of lube oil could cause the turbocharger thrust bearing temperature to elevate to the bearing failure point but typically only on the supply line to the turbocharger. Thus, a candidate could reason that this would trip the diesel generator.
Generators). Rev. 92  
Technical Reference(s):                 SOP 2.2.20, Standby AC Power System (Diesel (Attach if not previously provided)     Generators). Rev. 92 (including version/revision number) 2.3_DG1 R21              _________________________
Proposed references to be provided to applicants during examination:          __None___________
Learning Objective:
Per COR002-08-02R32, Diesel Generators
: 11. Predict the consequences a malfunction of the following would have on the Diesel Generators:
: c. Lube Oil pumps Question Source:                Bank #                  _______
Modified Bank #        _______
New                    ___X___
Question History:                Last NRC Exam          ____________
Question Cognitive Level:        Memory or Fundamental Knowledge              __X __
Comprehension or Analysis                    __ __
10 CFR Part 55 Content:          55.41 _(7)_
Comments:
LOD 3 59


(including version/revision number) 2.3_DG1 R21
60 61 62 ES-401                           Sample Written Examination                       Form ES-401-5 Question Worksheet Examination Outline Cross-
_________________________
Proposed references to be provided to applicants during examination:
__None___________
Learning Objective:  Per COR002 02R32, Diesel Generators           
: 11. Predict the consequences a malfunction of the following would have on the Diesel Generators:
: c. Lube Oil pumps Question Source:
Bank #  _______ Modified Bank #
_______  New  ___X___
Question History:
Last NRC Exam
____________
Question Cognitive Level:
Memory or Fundamental Knowledge
__X __ Comprehension or Analysis  __  __
10 CFR Part 55 Content:
55.41 _(7)_  Comments: LOD 3 59 60 61 62 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
-


==Reference:==
==Reference:==
 
Level                 RO             SRO Tier #               __2__         _____
Level   RO SRO Tier #     __2__ _____ Group # __1__ _____ K/A #   _ 300000 A3.02 _
Group #               __1__         _____
__ Importance Rating
K/A #                 _ 300000 A3.02 ___
_ 2.9 _ _____ 300000 Instrument Air  
Importance Rating     _ 2.9 _       _____
- Ability to monitor automatic operations of the instrument air system including: (CFR: 41.7) A3.02 Air Temperature
300000 Instrument Air - Ability to monitor automatic operations of the instrument air system including: (CFR: 41.7) A3.02 Air Temperature Question: 48 The air compressors are operating with the Compressor Control Module (CCM) in LOCAL with the following lineup:
 
* Air Compressor 1A is in Lead and running.
Question: 48
* Air Compressor 1B is First Backup and is idle.
 
* Air Compressor 1C is Second Backup and is idle.
The air compressors are operating with the Compressor Control Module (CCM) in LOCAL with the following lineup:
Air Compressor 1A is in Lead and running.
Air Compressor 1B is First Backup and is idle.
Air Compressor 1C is Second Backup and is idle.
Reactor Equipment Cooling (REC) to the air compressors is lost due to an REC pipe rupture.
Reactor Equipment Cooling (REC) to the air compressors is lost due to an REC pipe rupture.
What condition automatically trips Air Compressor 1A?
What condition automatically trips Air Compressor 1A?
What automatically occurs following the trip of Air Compressor 1A?
What automatically occurs following the trip of Air Compressor 1A?
A. High air temperature.
A. High air temperature.
Air Compressor 1B starts and continuously supplies air loads.
Air Compressor 1B starts and continuously supplies air loads.
B. High air temperature.
B. High air temperature.
Air Compressor 1B starts but trips soon after it starts.  
Air Compressor 1B starts but trips soon after it starts.
 
C. Low cooling water pressure.
C. Low cooling water pressure.
Air Compressor 1B starts and continuously supplies air loads.
Air Compressor 1B starts and continuously supplies air loads.
D. Low cooling water pressure.
D. Low cooling water pressure.
Air Compressor 1B starts but trips soon after it starts.
Air Compressor 1B starts but trips soon after it starts.
 
Answer:
Answer: A. High air temperature.
A. High air temperature.
Air Compressor 1B starts and continuously supplies air loads.
Air Compressor 1B starts and continuously supplies air loads.
Explanation:
Explanation:
The normal cooling water lineup is REC to Air Compressor 1A and TEC to Air Compressors 1B and 1C. When REC is lost to Air Compressor 1A, air temperatures will rise until Air Compressor 1A trips. Lowering air pressure will start the 1 st Backup compressor (which is cooled by TEC) so it continues to operate supplying system air loads. All air loads should be supplied. Also the sequence with compressors with CCM in LOCAL is lead cycles 110 to 100 psig, 1 st Backup starts at 93 psig and 2 nd Backup starts at 90 psig.
The normal cooling water lineup is REC to Air Compressor 1A and TEC to Air Compressors 1B and 1C. When REC is lost to Air Compressor 1A, air temperatures will rise until Air Compressor 1A trips. Lowering air pressure will start the 1st Backup compressor (which is cooled by TEC) so it continues to operate supplying system air loads. All air loads should be supplied. Also the sequence with compressors with CCM in LOCAL is lead cycles 110 to 100 psig, 1st Backup starts at 93 psig and 2nd Backup starts at 90 psig.
63 If power is lost to the REC
63
-TEC cross-tie valves (which is not the case here), then the REC and TEC alignment to the compressors when power is restored is Air Compressor 1A and Air Compressor 1B supplied by REC and Air Compressor 1C supplied by TEC. This response makes choices B and D highly plausible. The compressor protection from loss of cooling water is high temperature trips, not low cooling water pressure.
B. This option is incorrect because Air Compressor 1B does not trip on high air temperature but continues to operate as its cooling water supply is from TEC. The candidate who knows that Air Compressor 1A trips due to high temperature but who also believes that air compressor 1B is cooled by REC may choose this answer. This is a likely misconception that candidates could have, as two of the compressors are supplied by one closed cooling water system and one is supplied by a different system.


If power is lost to the REC-TEC cross-tie valves (which is not the case here), then the REC and TEC alignment to the compressors when power is restored is Air Compressor 1A and Air Compressor 1B supplied by REC and Air Compressor 1C supplied by TEC. This response makes choices B and D highly plausible. The compressor protection from loss of cooling water is high temperature trips, not low cooling water pressure.
B. This option is incorrect because Air Compressor 1B does not trip on high air temperature but continues to operate as its cooling water supply is from TEC. The candidate who knows that Air Compressor 1A trips due to high temperature but who also believes that air compressor 1B is cooled by REC may choose this answer. This is a likely misconception that candidates could have, as two of the compressors are supplied by one closed cooling water system and one is supplied by a different system.
C. This option is incorrect because Air Compressor 1A does not trip due to low cooling water pressure. This selection is plausible since many plant components that require cooling water do trip or isolate due to low cooling water pressure so a candidate may choose this option. Additionally, Air Compressor 1B does start and carry the load which adds to this selections plausibility.
C. This option is incorrect because Air Compressor 1A does not trip due to low cooling water pressure. This selection is plausible since many plant components that require cooling water do trip or isolate due to low cooling water pressure so a candidate may choose this option. Additionally, Air Compressor 1B does start and carry the load which adds to this selections plausibility.
D. This option is incorrect because the Air Compressor 1A does not trip due to low cooling water pressure nor will Air Compressor 1B trip as it remains supplied with cooling water from the TEC system.
D. This option is incorrect because the Air Compressor 1A does not trip due to low cooling water pressure nor will Air Compressor 1B trip as it remains supplied with cooling water from the TEC system. This selection is plausible since many plant components that require cooling water do trip or isolate due to low cooling water pressure so a candidate may choose this option. Additionally, Air Compressor 1B does start and carry the load which adds to this selections plausibility.
This selection is plausible since many plant components that require cooling water do trip or isolate due to low cooling water pressure so a candidate may choose this option. Additionally, Air Compressor 1B does start and carry the load which adds to this selections plausibility.
Technical Reference(s):                 SOP 2.2.59, Plant Air System. Rev. 74 (Attach if not previously provided)
Technical Reference(s):
(including version/revision number)                 ___________________
SOP 2.2.59, Plant Air System. Rev. 74 (Attach if not previously provided)
Proposed references to be provided to applicants during examination:       __None___________
        (including version/revision number)
___________________
Proposed references to be provided to applicants during examination:
__None___________
Learning Objective:
Learning Objective:
COR001-17-01 6 Predict the consequences the following would have on the Plant Air system:
COR001-17-01 6 Predict the consequences the following would have on the Plant Air system:
a.REC failure b.TEC failure
a.REC failure b.TEC failure
: 10. Given plant conditions, determine if any of the following should occur:
: 10. Given plant conditions, determine if any of the following should occur:
c.Air Compressor automatic trip Question Source:
c.Air Compressor automatic trip Question Source:               Bank #               _______
Bank #   _______ Modified Bank #
Modified Bank #       _______
_______  New   ___X___
New                   ___X___
Question History:
Question History:               Last NRC Exam         ____________
Last NRC Exam
Question Cognitive Level:       Memory or Fundamental Knowledge             _____
____________
Comprehension or Analysis                   __X__
Question Cognitive Level:
10 CFR Part 55 Content:         55.41 _(7)_
Memory or Fundamental Knowledge
64
_____ Comprehension or Analysis   __X__ 10 CFR Part 55 Content:
 
55.41 _(7)_ 64 Comments: LOD 3   65 66 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
Comments:
-
LOD 3 65
 
66 ES-401                             Sample Written Examination                   Form ES-401-5 Question Worksheet Examination Outline Cross-


==Reference:==
==Reference:==
 
Level               RO             SRO Tier #             __2__         _____
Level   RO SRO Tier #     __2__ _____ Group # __1__ _____ K/A #   _400000 A2.01 _
Group #             __1__         _____
__ Importance Rating
K/A #               _400000 A2.01 ___
_ 3.3 _ _____ 400000 Component Cooling Water  
Importance Rating   _ 3.3 _       _____
- Ability to (a) predict the impacts of the following on the CCWS and (b) based on those predictions use procedures to correct, control or mitigate the consequences of those abnormal operations: (CFR: 41.5) A2.01 Loss of CCW pump. Question: 49 The Unit is operating at 50% power with REC Pump A breaker tagged OPEN.
400000 Component Cooling Water - Ability to (a) predict the impacts of the following on the CCWS and (b) based on those predictions use procedures to correct, control or mitigate the consequences of those abnormal operations: (CFR: 41.5) A2.01 Loss of CCW pump.
Question: 49 The Unit is operating at 50% power with REC Pump A breaker tagged OPEN.
REC Heat Exchanger B and 3 REC pumps are in service.
REC Heat Exchanger B and 3 REC pumps are in service.
One REC pump trips and is unavailable.
* One REC pump trips and is unavailable.
M-1/A-1, REC SYSTEM LOW PRESSURE alarms.
* M-1/A-1, REC SYSTEM LOW PRESSURE alarms.
  (1) How does the system respond with no operator action?
(1) How does the system respond with no operator action?
(2) What Immediate Action(s) is
(2) What Immediate Action(s) is/are required upon receipt of this alarm with system pressure at 60 psig IAW Procedure 5.2REC (Loss of REC)?
/are required upon receipt of this alarm with system pressure at 60 psig IAW Procedure 5.2REC (Loss of REC)?
A. (1) Non-critical loads isolate after a 20 second time delay.
A. (1) Non-critical loads isolate after a 20 second time delay.
(2) Isolate Augmented Radwaste cooling ONLY.
(2) Isolate Augmented Radwaste cooling ONLY.
 
B. (1) Non-critical loads isolate after a 20 second time delay.
B. (1) Non-critical loads isolate after a 20 second time delay.
  (2) Isolate Augmented Radwaste cooling and RWCU Non
(2) Isolate Augmented Radwaste cooling and RWCU Non-Regen HX Inlet.
-Regen HX Inlet.
 
C. (1) Non-critical loads isolate after a 40 second time delay.
C. (1) Non-critical loads isolate after a 40 second time delay.
  (2) Isolate Augmented Radwaste cooling ONLY.
(2) Isolate Augmented Radwaste cooling ONLY.
D. (1) Non-critical loads isolate after a 40 second time delay.
D. (1) Non-critical loads isolate after a 40 second time delay.
  (2) Isolate Augmented Radwaste cooling and RWCU Non
(2) Isolate Augmented Radwaste cooling and RWCU Non-Regen HX Inlet.
-Regen HX Inlet.
Answer:
Answer: D. (1) Non-critical loads isolate after a 40 second time delay.
D. (1) Non-critical loads isolate after a 40 second time delay.
  (2) Isolate Augmented Radwaste cooling and RWCU Non
(2) Isolate Augmented Radwaste cooling and RWCU Non-Regen HX Inlet.
-Regen HX Inlet.
Explanation:
Explanation:
The REC system contains 4 pumps with 3 pumps normally in operation. The standby pump is currently unavailable to be started. The REC heat exchanger outlet piping contains pressure switches that isolate the non
The REC system contains 4 pumps with 3 pumps normally in operation. The standby pump is currently unavailable to be started. The REC heat exchanger outlet piping contains pressure switches that isolate the non-critical loads and Augmented Radwaste after a 40 second time delay. Without a standby REC pump available, 5.2REC Immediate Actions require MANUALLY isolating Augmented Radwaste and the RWCU NRHX in an attempt to restore system pressure PRIOR to automatic isolation.
-critical loads and Augmented Radwaste after a 40 second time delay. Without a standby REC pump available, 5.2REC Immediate Actions require MANUALLY isolating Augmented Radwaste and the RWCU NRHX in an attempt to restore system pressure PRIOR to automatic isolation.
67
67 A. This answer is incorrect because non
 
-critical loads isolate after heat exchanger outlet pressure remains below the isolation setting for 40 seconds and additionally isolating RWCU NRHX is required as part of procedure 5.2REC immediate operator actions. This answer is plausible because the non
A. This answer is incorrect because non-critical loads isolate after heat exchanger outlet pressure remains below the isolation setting for 40 seconds and additionally isolating RWCU NRHX is required as part of procedure 5.2REC immediate operator actions. This answer is plausible because the non-critical loads of the REC system do isolate on a system low pressure condition with the time delay being easily confused with the Standby REC pump auto start which occurs 20 second on DG sequential load and isolating Augmented RW happens automatically after the 40 seconds but is often forgotten as a 5.2REC immediate action. The candidate who cannot recall the correct time delay on the Non-critical load valve isolation and does not remember all 5.2REC immediate operator actions would choose this answer.
-critical loads of the REC system do isolate on a system low pressure condition with the time delay being easily confused with the Standby REC pump auto start which occurs 20 second on DG sequential load and isolating Augmented RW happens automatically after the 40 seconds but is often forgotten as a 5.2REC immediate action. The candidate who cannot recall the correct time delay on the Non
B. This answer is incorrect because non-critical loads isolate after heat exchanger outlet pressure remains below the isolation setting for 40 seconds This answer is plausible because the non-critical loads of the REC system do isolate on a system low pressure condition with the time delay being easily confused with the Standby REC pump auto start which occurs 20 second on DG sequential load. The candidate who cannot recall the correct time delay on the Non-critical load valve isolation and remembers all 5.2REC immediate operator actions would choose this answer.
-critical load valve isolation and does not remember all 5.2REC immediate operator actions would choose this answer. B. This answer is incorrect because non
C. This answer is incorrect because isolating RWCU NRHX is required as part of procedure 5.2REC immediate operator actions. This answer is plausible because isolating Augmented RW happens automatically after the 40 seconds and can be forgotten as a 5.2REC immediate action. The candidate who recalls the correct time delay on the Non-critical load valve isolation and does not remember all 5.2REC immediate operator actions would choose this answer.
-critical loads isolate after heat exchanger outlet pressure remains below the isolation setting for 40 seconds This answer is plausible because the non
Technical Reference(s): Procedure 2.3_M-1, Panel M, Rev. 14.
-critical loads of the REC system do isolate on a system low pressure condition with the time delay being easily confused with the Standby REC pump auto start which occurs 20 second on DG sequential load. The candidate who cannot recall the correct time delay on the Non
-critical load valve isolation and remembers all 5.2REC immediate operator actions would choose this answer.
C. This answer is incorrect because isolating RWCU NRHX is required as part of procedure 5.2REC immediate operator actions. This answer is plausible because isolating Augmented RW happens automatically after the 40 seconds and can be forgotten as a 5.2REC immediate action. The candidate who recalls the correct time delay on the Non
-critical load valve isolation and does not remember all 5.2REC immediate operator actions would choose this answer.
Technical Reference(s): Procedure 2.3_M
-1, Panel M, Rev. 14.
Procedure 5.2REC (Loss of REC), Rev. 16.
Procedure 5.2REC (Loss of REC), Rev. 16.
Proposed references to be provided to applicants during examination:
Proposed references to be provided to applicants during examination:       __None___________
__None___________
Learning Objective:
Per COR002-19-02
: 6. Given a specific REC malfunction, determine the effect on any of the following:
: a.      REC header pressure
: d.      Standby REC pump operation Question Source:              Bank #                _      _
Modified Bank #        _______
New                    ___X___
Question History:            Last NRC Exam          _        _
Question Cognitive Level:    Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content:      55.41 _(5)_
Comments:
LOD 3 68


Learning Objective:
69 70 71 72 ES-401                           Sample Written Examination                     Form ES-401-5 Question Worksheet Examination Outline Cross-
Per COR002 02 6. Given a specific REC malfunction, determine the effect on any of the following:
: a. REC header pressure
: d. Standby REC pump operation Question Source:
Bank #  _
_ Modified Bank #
_______  New  ___X___  Question History:
Last NRC Exam
_
_
Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X    10 CFR Part 55 Content:
55.41 _(5)_  Comments: LOD 3  68 69 70 71 72 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
-


==Reference:==
==Reference:==
 
Level               RO           SRO Tier #             __2__         _____
Level   RO SRO Tier #     __2__ _____ Group # __1__ _____ K/A #   _ 211000 A1.07 _
Group #             __1__         _____
__ Importance Rating
K/A #               _ 211000 A1.07 ___
_ 4.3_ _____ 211000 SLC  
Importance Rating   _ 4.3_       _____
- Ability to predict and/or monitor changes in parameters associated with operating the standby liquid control system controls including: (CFR: 41.5) A1.07 Reactor power Question: 50 The plant operating at 100% rated power with SLC pump 1A out of service when an ATWS occurs: No control rods insert.
211000 SLC - Ability to predict and/or monitor changes in parameters associated with operating the standby liquid control system controls including: (CFR: 41.5) A1.07 Reactor power Question: 50 The plant operating at 100% rated power with SLC pump 1A out of service when an ATWS occurs:
The MSIVs are closed.
* No control rods insert.
Average torus temperature is just below BIIT and rising slowly.
* The MSIVs are closed.
RPV level remains at +35 inches.
* Average torus temperature is just below BIIT and rising slowly.
* RPV level remains at +35 inches.
SLC pump 1B is placed to START and the following conditions are present:
SLC pump 1B is placed to START and the following conditions are present:
Reactor power is 40% and slowly lowering.
* Reactor power is 40% and slowly lowering.
Boron is injecting with an initial tank level of 80%.
* Boron is injecting with an initial tank level of 80%.
Assuming the SLC pump is operating at its design flow rate and RPV level remains steady, what is the approximate SLC tank level and status of the Average Power Range Monitor (APRM) downscales after 25 minutes?
Assuming the SLC pump is operating at its design flow rate and RPV level remains steady, what is the approximate SLC tank level and status of the Average Power Range Monitor (APRM) downscales after 25 minutes?
SLC Tank Level APRM Downscale Alarm (9-5-1/C-8) A. 54%         OFF B. 54%
SLC Tank Level           APRM Downscale Alarm (9-5-1/C-8)
ON C. 26%
A.             54%                           OFF B.             54%                           ON C.             26%                           OFF D.             26%                           ON Answer:
O FF D. 26%
B.             54%                           ON Explanation:
ON   Answer: B. 54%
Per COR002-29-02 The SLC pumps are triplex plunger type pumps with a design flow rate of 53 gpm. Based on this design flow rate, the minimum TS flow rate of 38.2 gpm will always be met. At the design flow rate of 53 gpm, each positive displacement pump is capable of injecting Hot Shutdown Boron Weight into the RPV within approximately 22 minutes. Normal SLC tank volume is maintained at approximately 80% and EOP 7A specifies that a 26% drop of SLC tank level will be HSD boron weight. This volume may be calculated as follows:
ON Explanation:
73
Per COR002 02 The SLC pumps are triplex plunger type pumps with a design flow rate of 53 gpm. Based on will always be met. At the design flow rate of 53 gpm, each positive displacement pump is capable of injecting Hot Shutdown Boron Weight into the RPV within approximately 22 minutes. Normal SLC tank volume is maintained at approximately 80% and EOP 7A specifies that a 26% drop of SLC tank level will be HSD boron weight. This volume may be calculated as follows:
 
73 SLC tank overflow = 4565 Gallons (vertical cylindrical tank) 80% SLC tank level = 4565 x 0.8 = 3652 gallons (approximate normal volume) 4565 gallons/100 % tank = 45.65 gallons/% tank 45.65 gal/% tank x 26 (HSDBW) = 1186.9 gallons Assumed electrical supply frequency is 60 Hz.
SLC tank overflow = 4565 Gallons (vertical cylindrical tank) 80% SLC tank level = 4565 x 0.8 = 3652 gallons (approximate normal volume) 4565 gallons/100 % tank = 45.65 gallons/% tank 45.65 gal/% tank x 26 (HSDBW) = 1186.9 gallons Assumed electrical supply frequency is 60 Hz.
1187 gal 53 gpm   = 22.4 minutes 1 SLC pump at 53 gpm injection flow will inject 26% (HSBW) of the SLC tank in approximately 22 minutes
1187 gal 53 gpm = 22.4 minutes 1 SLC pump at 53 gpm injection flow will inject 26% (HSBW) of the SLC tank in approximately 22 minutes.
. The candidate should predict that 25 minutes of SLC injection will inject HSBW (drop tank level by 26 percent) which in turn will drop reactor power below 3%. An extra margin of slightly over two minutes is added for conservatism.
The candidate should predict that 25 minutes of SLC injection will inject HSBW (drop tank level by 26 percent) which in turn will drop reactor power below 3%. An extra margin of slightly over two minutes is added for conservatism.
Normal ATWS power reduction strategies are to intentionally lower RPV level which lowers reactor power. The stem of the question maintains RPV level steady in order for the candidate to answer the question solely from a boron injection standpoint.
Normal ATWS power reduction strategies are to intentionally lower RPV level which lowers reactor power. The stem of the question maintains RPV level steady in order for the candidate to answer the question solely from a boron injection standpoint.
A. This option is incorrect because, with 26% of the SLC injected, hot shutdown boron weight is injected and so reactor power will be less than 3% (APRM downscale). A candidate who believes that HSBW has not yet been injected (the candidate who confuses HSBW with Cold shutdown boron weight) may choose this answer believing that until 60% of the tank is injected that the APRM downscale will not be in.
A. This option is incorrect because, with 26% of the SLC injected, hot shutdown boron weight is injected and so reactor power will be less than 3% (APRM downscale). A candidate who believes that HSBW has not yet been injected (the candidate who confuses HSBW with Cold shutdown boron weight) may choose this answer believing that until 60% of the tank is injected that the APRM downscale will not be in.
C. This option is incorrect because SLC tank level would not be at 26% after 25 minutes with only one pump in operation. This would be the approximate level had two pumps been in operation so a candidate who fails to evaluate that only one pump is in operation may choose this option. Additionally the APRMs would be downscale but a candidate may believe that more than 60% of the tank must be injected in order to get reactor power less than 3%. D. This option is incorrect because SLC tank level would not be at 26% after 25 minutes with only one pump in operation. This would be the approximate level had two pumps been in operation however so a candidate who fails to evaluate that only one pump is in operation may choose this option.
C. This option is incorrect because SLC tank level would not be at 26% after 25 minutes with only one pump in operation. This would be the approximate level had two pumps been in operation so a candidate who fails to evaluate that only one pump is in operation may choose this option. Additionally the APRMs would be downscale but a candidate may believe that more than 60% of the tank must be injected in order to get reactor power less than 3%.
Technical Reference(s):
D. This option is incorrect because SLC tank level would not be at 26% after 25 minutes with only one pump in operation. This would be the approximate level had two pumps been in operation however so a candidate who fails to evaluate that only one pump is in operation may choose this option.
SOP 2.2.74, Standby Liquid Control, Rev. 52
Technical Reference(s):               SOP 2.2.74, Standby Liquid Control, Rev. 52     _____
_____ (Attach if not previously provided) 6.SLC.101 R23 SLC Pump Operability Test
(Attach if not previously provided)   6.SLC.101 R23 SLC Pump Operability Test       ______
______ (including version/revision number)
(including version/revision number)                             _________________________
_________________________
Proposed references to be provided to applicants during examination:       __None___________
Proposed references to be provided to applicants during examination:
__None___________
Learning Objective:
Learning Objective:
Per INT008-06-10 R27
: 6. State the SLC tank level at which RPV level may be slowly raised to above the scram setpoint.
Question Source:              Bank #                _______
Modified Bank #        _______
74


Per INT008 10 R27      6. State the SLC tank level at which RPV level may be slowly raised to above the scram setpoint. Question Source:
New               ___X ___
Bank #  _______ Modified Bank #
Question History:         Last NRC Exam     ____________
_______  74 New   ___X ___ Question History:
Question Cognitive Level: Memory or Fundamental Knowledge _____
Last NRC Exam
Comprehension or Analysis      __ X__
____________
10 CFR Part 55 Content:  55.41 _ 7 _
Comments:
LOD 3 75


Question Cognitive Level:
ES-401                             Sample Written Examination                       Form ES-401-5 Question Worksheet Examination Outline Cross-
Memory or Fundamental Knowledge
_____ Comprehension or Analysis  __ X__
10 CFR Part 55 Content:
55.41 _ 7  _
Comments: LOD 3        75 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
-


==Reference:==
==Reference:==
 
Level                 RO Tier #                 2 Group #               1 K/A #                 215003 A4.03 Importance Rating     3.6 215003 IRM - Ability to manually operate and/or monitor in the control room:
Level   RO   Tier #     2   Group # 1 K/A #   215003 A4.03 Importance Rating
(CFR: 41.7 / 45.5 to 45.8) - A4.03 IRM range switches Question: 51 A Plant startup is in progress with reactor power at or near the point of adding heat.
 
===3.6 215003===
IRM  
- Ability to manually operate and/or monitor in the control room:  
(CFR: 41.7 / 45.5 to 45.8)  
- A4.03 IRM range switches
 
Question: 51 A Plant startup is in progress with reactor power at or near the point of adding heat.
 
Reactor period is near infinity with the Reactor Mode Select switch in STARTUP / HOT STANDBY position.
Reactor period is near infinity with the Reactor Mode Select switch in STARTUP / HOT STANDBY position.
IRM C is on Range 7 and indicates 35 on the Panel 9
IRM C is on Range 7 and indicates 35 on the Panel 9-5 Recorder, when its range switch is taken to Range 6.
-5 Recorder, when its range switch is taken to Range 6.
What is the new indication for IRM C and what automatic action(s) occur(s)?
What is the new indication for IRM C and what automatic action(s) occur(s)?
Recorder Indication Automatic Action A. 97   Rod Block Only.
Recorder Indication           Automatic Action A.     97                             Rod Block Only.
 
B.     97                             Rod Block and Half Scram.
B. 97   Rod Block and Half Scram.
C.     110                           Rod Block Only.
C. 110   Rod Block Only.
D.     110                           Rod Block and Half Scram.
D. 110   Rod Block and Half Scram.
Answer:
Answer: C. 110   Rod Block Only.
C.     110                           Rod Block Only.
Explanation:
Explanation:
Placing the IRM range switch to the next lower scale will increase the current IRM recorder reading by approximately 3.125 (125/40 = 3.125). This will result in 35 x 3.125 = 109.4 which is above the TRM Control Rod Block Instrumerange correlation is calculated incorrectly (40/125 = .36) and 35 / .36 = 97.22.
Placing the IRM range switch to the next lower scale will increase the current IRM recorder reading by approximately 3.125 (125/40 = 3.125). This will result in 35 x 3.125 = 109.4 which is above the TRM Control Rod Block Instrumentation setpoint of 108/125 of Full Scale. If the range correlation is calculated incorrectly (40/125 = .36) and 35 / .36 = 97.22.
Distracters:
Distracters:
A. This option is incorrect because IRM C would indicate approximately 110. This choice is plausible due to incorrectly calculating the range correlation and not correctly recalling IRM Rod Block setpoints. The candidate that incorrectly calculates the IRM range correlation and does not recall the IRM Rod block setpoint would select this answer.
A. This option is incorrect because IRM C would indicate approximately 110. This choice is plausible due to incorrectly calculating the range correlation and not correctly recalling IRM Rod Block setpoints. The candidate that incorrectly calculates the IRM range correlation and does not recall the IRM Rod block setpoint would select this answer.
B. This option is incorrect because IRM C would indicate approximately 110 and a Rod Block AND Half Scram would not occur at this value of the scale. This choice is plausible due to incorrectly calculating the range correlation and not correctly recalling IRM Rod Block and 76


B. This option is incorrect because IRM C would indicate approximately 110 and a Rod Block AND Half Scram would not occur at this value of the scale. This choice is plausible due to incorrectly calculating the range correlation and not correctly recalling IRM Rod Block and 76 Scram setpoints. The candidate that incorrectly calculates the IRM range correlation and does not recall IRM Scram and Rod block setpoints would select this answer.
Scram setpoints. The candidate that incorrectly calculates the IRM range correlation and does not recall IRM Scram and Rod block setpoints would select this answer.
D. This option is incorrect because IRM C indicating 110 causes a Rod Block ONLY. This choice is plausible due to not correctly recalling IRM Rod Block and Scram setpoints. The candidate that correctly calculates the IRM range correlation and does not recall IRM Scram and Rod block setpoints would select this answer.
Technical Reference(s):      IOP 4.1.2 Intermediate Range Monitoring System, Rev. 23 Major Design Change DC-76-1 CNS ESAR Proposed references to be provided to applicants during examination:        none Learning Objective:    OPS Intermediate Range Monitor / COR002-12-02 LO-01: State the purpose of the following items related to Intermediate Range Monitoring:
: h. Range switch LO-05: Describe the IRM system design features and/or interlocks that provide the following:
d: Varying system sensitivity levels using Range switches Question Source:              Bank #
Modified Bank #
New                    X Question History:            Last NRC Exam Question Cognitive Level:    Memory or Fundamental Knowledge Comprehension or Analysis                      X 10 CFR Part 55 Content:      55.41 (7)
Comments:
LOD 4 77


D. This option is incorrect because IRM C indicating 110 causes a Rod Block ONLY.
Panel 9-5 IRM Recorder example showing 1-125% scale.
This choice is plausible due to not correctly recalling IRM Rod Block and Scram setpoints. The candidate that correctly calculates the IRM range correlation and does not recall IRM Scram and Rod block setpoints would select this answer.
78


Technical Reference(s): IOP 4.1.2 Intermediate Range Monitoring System, Rev. 23 Major Design Change DC 1    CNS ESAR  Proposed references to be provided to applicants during examination:
The IRM indicator below is located on back panel 9-12 and illustrates the red 0-40% and black 0-125% scale.
none Learning Objective:
IRM Range Switch example. It should be noted that these switches located on panel 9-5 are mechanically pinned to prohibit going above range 9.
OPS Intermediate Range Monitor / COR002 02    LO-01: State the purpose of the following items related to Intermediate Range Monitoring:
79
: h. Range switch LO-05: Describe the IRM system design features and/or interlocks that provide the following:
 
d: Varying system sensitivity levels using Range switches  Question Source:
SETPOINT                                       CIC                            9-5-1/E-7 (2354) 102.5/125 of scale (TRM  108/125 of   MNI-NAM-41A through MNI-NAM-41H full scale)
Bank #    Modified Bank #
SETPOINT                                  CIC                           9-5-1/D-7 (2353) Upscale trip at 117.5/125 of scale NMI-NAM-41A, NMI-NAM-41C, (Tech Spec  121/125 of scale) or inop   NMI-NAM-41E, or NMI-NAM-41G due to:
New  X  Question History:
: 1. IRM module unplugged
Last NRC Exam Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis  X  10 CFR Part 55 Content:
55.41 (7)  Comments:
LOD 4  77 Panel 9-5 IRM Recorder example showing 1
-125% scale.
78 The IRM indicator below is located on back panel 9
-12 and illustrates the red 0
-40% and black 0-125% scale.
IRM Range Switch example. It should be noted that these switches located on panel 9
-5 are mechanically pinned to prohibit going above range 9.
79 SETPOINT (2354) 102.5/125 of scale (TRM  108/125 of full scale)
CIC 9-5-1/E-7 MNI-NAM-41A through MNI
-NAM-41H    SETPOINT (2353) Upscale trip at 117.5/125 of scale (Tech Spec  121/125 of scale) or inop due to: 1. IRM module unplugged
: 2. High voltage low
: 2. High voltage low
: 3. MODE switch not in operate
: 3. MODE switch not in operate
: 4. Loss of negative supply voltage CIC 9-5-1/D-7 NMI-NAM-41A, NMI-NAM-41C, NMI-NAM-41E, or NMI
: 4. Loss of negative supply voltage 80
-NAM-41G    80 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
 
-
ES-401                           Sample Written Examination                         Form ES-401-5 Question Worksheet Examination Outline Cross-


==Reference:==
==Reference:==
 
Level                 RO Tier #                 2 Group #               1 K/A #                 215004 2.4.3 Importance Rating     3.7 215004 Source Range Monitor - 2.4.3 Ability to identify post-accident instrumentation.
Level   RO   Tier #     2   Group # 1 K/A #   215004 2.4.3 Importance Rating
(CFR: 41.6 / 45.4)
 
Question: 52 Which one of the following is a Post Accident Monitoring (PAM) instrument?
===3.7 215004===
A. Source Range Monitor (SRM)
Source Range Monitor  
B. Traversing In-core Probe (TIP)
- 2.4.3 Ability to identify post
C. Condensate Storage Tank Level Indicator D. Reactor Building Ventilation Exhaust Plenum Radiation Monitors Answer:
-accident instrumentation.
A. Source Range Monitor (SRM)
(CFR: 41.6 / 45.4)
Explanation:
Q ues ti on: 5 2  W h i c h one of the following is a Pos t Acc iden t M on it o ri n g (PAM) i ns tr u m en t?
TLCO 3.3.3 Post Accident Instrumentation (PAM) Instrumentation, Table 3.3.3-1 specifies the instruments that are post accident instrumentation. The neutron monitoring systems in the table are the SRMs, IRMs, and APRMs. The only instrument listed in the options that is a PAM instrument is the SRM.
A. S ou r c e Rang e M on it o r (SR M)   B. T rave r s i n g I n-co r e P r ob e (T I P) C. Condensate Storage Tank Level Indicator D. Reactor Building Ventilation Exhaust Plenum Radiation Monitors A ns w e r: A. S ou r c e Rang e M on it o r (SR M)   Explanation:
TLCO 3.3.3 Post Accident Instrumentation (PAM) Instrumentation, Table 3.3.3
-1 specifies the instruments that are post accident instrumentation. The neutron monitoring systems in the table are the SRMs, IRMs, and APRMs. The only instrument listed in the options that is a PAM instrument is the SRM.
Distracters:
Distracters:
B. This option is incorrect because the TIP system is not a PAM instrument. This choice is plausible due to TIPs being the ONLY neutron monitoring instrument not being part of PAM. Because the TIP can enter the core and provide data for different core locations a candidate could believe that this system function is required post
B. This option is incorrect because the TIP system is not a PAM instrument. This choice is plausible due to TIPs being the ONLY neutron monitoring instrument not being part of PAM.
-accident and choose this option. Validation data indicated that candidates thought that this may be a plausible alternative for a post-accident monitor. The TIP system would not be a good selection as a PAM, because it would only be truly effective if there were a significant neutron flux detected. The SRMs are calibrated for lower neutron flux levels and do not physically communicate outside of the primary containment like the TIPs do. If the TIPs were used with any significant pressure on the RPV and there was a penetration of the TIP tubing, then this would be a direct path from the core to the secondary containment.
Because the TIP can enter the core and provide data for different core locations a candidate could believe that this system function is required post-accident and choose this option.
Validation data indicated that candidates thought that this may be a plausible alternative for a post-accident monitor. The TIP system would not be a good selection as a PAM, because it would only be truly effective if there were a significant neutron flux detected.
The SRMs are calibrated for lower neutron flux levels and do not physically communicate outside of the primary containment like the TIPs do. If the TIPs were used with any significant pressure on the RPV and there was a penetration of the TIP tubing, then this would be a direct path from the core to the secondary containment.
C. This option is incorrect because the Condensate Storage Tank level instrument is not a required PAM instrument. Because the tank level instruments provide control room alarms and the tank is an alternate suction source for low pressure ECCS a candidate may choose this option.
C. This option is incorrect because the Condensate Storage Tank level instrument is not a required PAM instrument. Because the tank level instruments provide control room alarms and the tank is an alternate suction source for low pressure ECCS a candidate may choose this option.
81 D: This option is incorrect because the reactor building ventilation radiation monitoring instruments are not used for post
81
-accident monitoring. Because other ventilation radiation monitoring instruments (Turbine Building) are PAM required instruments a candidate may choose this option.
 
Technical Reference(s):
D: This option is incorrect because the reactor building ventilation radiation monitoring instruments are not used for post-accident monitoring. Because other ventilation radiation monitoring instruments (Turbine Building) are PAM required instruments a candidate may choose this option.
TRM 3.3.3 Proposed references to be provided to applicants during examination:
Technical Reference(s):       TRM 3.3.3 Proposed references to be provided to applicants during examination:         NONE Learning Objective:
NONE Learning Objective:
OPS Source Range Monitor/COR002-30-02 LO-02         Given conditions and/or parameters associated with the SRM system, determine if related Technical Specification and Technical Requirements Manual Limiting Conditions for Operation are met.
OPS Source Range Monitor/COR002 02 LO-02 Given conditions and/or parameters associated with the SRM system, determine if related Technical Specification and Technical Requirements Manual Limiting Conditions for Operation are met.
D.      Technical Requirements Manual
: 2. T 3.3.3, Non-Type A, Non-Category 1 Post Accident Monitoring (PAM) Instrumentation INT007-06-02 TRM - Instrumentation
: 1.        Given plant conditions, determine if the following TRM Limiting Conditions for Operation (TLCOs) are met:
: c. T 3.3.3 Non-Type A, Non-Category 1 Post Accident Monitoring PAM Instrumentation Question Source:              Bank #
Modified Bank #
New                    X Question History:            Last NRC Exam Question Cognitive Level:    Memory or Fundamental Knowledge                X Comprehension or Analysis 10 CFR Part 55 Content:      55.41 (6)
Comments:
LOD 4 82


D. Technical Requirements Manual
83 ES-401                           Sample Written Examination                     Form ES-401-5 Question Worksheet Examination Outline Cross-
: 2. T 3.3.3, Non
-Type A, Non
-Category 1 Post Accident Monitoring (PAM) Instrumentation INT007-06-02 TRM - Instrumentation
: 1. Given plant conditions, determine if the following TRM Limiting Conditions for Operation (TLCOs) are met:
: c. T 3.3.3 Non-Type A, Non
-Category 1 Post Accident      Monitoring PAM Instrumentation Question Source:
Bank #    Modified Bank #
New  X Question History:
Last NRC Exam Question Cognitive Level:
Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content:
55.41 (6)  Comments:
LOD 4  8 2 83 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
-


==Reference:==
==Reference:==
 
Level                 RO Tier #                 2 Group #               1 K/A #                 217000 K1.07 Importance Rating     3.1 Knowledge of the physical connections and/or cause-effect relationships between REACTOR CORE ISOLATION COOLING SYSTEM (RCIC) and the following: (CFR: 41.2 to 41.9 / 45.7 to 45.8) - K1.07 Leak detection Question: 53 Following a plant transient the following conditions are present:
Level   RO   Tier #     2   Group # 1 K/A #   217000 K1.07 Importance Rating
* Reactor Core Isolation Cooling (RCIC) is injecting at 400 gpm following an automatic initiation.
 
* A steam leak develops in the RCIC Room.
===3.1 Knowledge===
* RCIC Room area temperature is 195&deg;F and rising.
of the physical connections and/or cause
-effect relationships between REACTOR CORE ISOLATION COOLING SYSTEM (RCIC) and the following: (CFR: 41.2 to 41.9 / 45.7 to 45.8)  
- K1.07 Leak detection Question: 53 Following a plant transient the following conditions are present:
Reactor Core Isolation Cooling (RCIC) is injecting at 400 gpm following an automatic initiation.
A steam leak develops in the RCIC Room.
RCIC Room area temperature is 195&deg;F and rising.  
 
What is the effect on RCIC?
What is the effect on RCIC?
 
A.     Only RCIC-MO-16 (RCIC STEAM SUPPLY OUTBOARD ISOLATION VALVE) closes and the RCIC turbine trips.
A. Only RCIC-MO-16 (RCIC STEAM SUPPLY OUTBOARD ISOLATION VALVE) closes and the RCIC turbine trips.
B.     Only RCIC-MO-15 (RCIC STEAM SUPPLY INBOARD ISOLATION VALVE) closes and the RCIC turbine coasts down (no trip).
B. Only RCIC-MO-15 (RCIC STEAM SUPPLY INBOARD ISOLATION VALVE) closes and the RCIC turbine coasts down (no trip).
C.     RCIC-MO-15 and RCIC-MO-16 (RCIC STEAM SUPPLY INBOARD and OUTBOARD ISOLATION VALVE) close and the RCIC turbine trips.
C. RCIC-MO-15 and RCIC
D.     RCIC-MO-15 and RCIC-MO-16 (RCIC STEAM SUPPLY INBOARD and OUTBOARD ISOLATION VALVE) close and the RCIC turbine coasts down (no trip).
-MO-16 (RCIC STEAM SUPPLY INBOARD and OUTBOARD ISOLATION VALVE) close and the RCIC turbine trips.
Answer:
D. RCIC-MO-15 and RCIC
C.     RCIC-MO-15 and RCIC-MO-16 (RCIC STEAM SUPPLY INBOARD and OUTBOARD ISOLATION VALVE) close and the RCIC turbine trips.
-MO-16 (RCIC STEAM SUPPLY INBOARD and OUTBOARD ISOLATION VALVE) close and the RCIC turbine coasts down (no trip).
 
Answer: C. RCIC-MO-15 and RCIC
-MO-16 (RCIC STEAM SUPPLY INBOARD and OUTBOARD ISOLATION VALVE) close and the RCIC turbine trips.
 
Explanation:
Explanation:
RCIC Room temperature is part of the leak detection system. The following conditions will cause an automatic RCIC system isolation (Group 5 isolation):
RCIC Room temperature is part of the leak detection system. The following conditions will cause an automatic RCIC system isolation (Group 5 isolation):
psig)   6 sec. TD) RCIC steam line high space tempe When a Group 5 isolation occurs the RCIC Steam Supply Line Inboard and Outboard Isolation valves (MO
* RCIC steam supply low pressure ( 61 psig)
-15 & 16) close, the RCIC turbine trips and the Minimum Flow valve (MO
* RCIC steam supply line high flow ( 288% of rated + 6 sec. TD)
-27) closes.
* RCIC steam line high space temperature (195&deg;F)
When a Group 5 isolation occurs the RCIC Steam Supply Line Inboard and Outboard Isolation valves (MO-15 & 16) close, the RCIC turbine trips and the Minimum Flow valve (MO-27) closes.
Distracters:
Distracters:
84 A. This option is incorrect because RCIC-MO-15 also closes. The actions specified in this option are the actions that occur when a half group 5 isolation occurs on channel A. These are also the automatic actions that occur when the manual isolation button is depressed which only functions when an initiation is present. Since these actions are very specific and occur only when an initiation signal is present, as is the case here, a candidate may choose this option believing that with the initiation signal present the isolation is only on channel A. B. This option is incorrect because RCIC
84
-MO-16 also closes and the RCIC turbine trips. The actions specified in this option are those that occur with a half group 5 isolation on the B channel. A candidate may believe that with the initiation signal present the turbine coasts down due to the governor valve closing, rather than trip due to the trip throttle valve rapidly closing. 


D. This option is incorrect because the RCIC turbine trips when the isolation signal is present. The isolation signal is a direct turbine trip signal. The candidate may believe the governor valve closes allowing the turbine to coast down rather than the trip throttle valve rapidly closing. Technical Reference(s):
A. This option is incorrect because RCIC-MO-15 also closes. The actions specified in this option are the actions that occur when a half group 5 isolation occurs on channel A. These are also the automatic actions that occur when the manual isolation button is depressed which only functions when an initiation is present. Since these actions are very specific and occur only when an initiation signal is present, as is the case here, a candidate may choose this option believing that with the initiation signal present the isolation is only on channel A.
SOP 2.2.67, Reactor Core Isolation Cooling System, Rev. 70   Proposed references to be provided to applicants during examination:
B. This option is incorrect because RCIC-MO-16 also closes and the RCIC turbine trips. The actions specified in this option are those that occur with a half group 5 isolation on the B channel. A candidate may believe that with the initiation signal present the turbine coasts down due to the governor valve closing, rather than trip due to the trip throttle valve rapidly closing.
NONE Learning Objective:
D. This option is incorrect because the RCIC turbine trips when the isolation signal is present.
OPS Reactor Core Isolation Cooling/COR002 02 R25     10. Predict the consequences of the following on the RCIC system:
The isolation signal is a direct turbine trip signal. The candidate may believe the governor valve closes allowing the turbine to coast down rather than the trip throttle valve rapidly closing.
: n. Steam line break (Steam Tunnel/RCIC Room)
Technical Reference(s):       SOP 2.2.67, Reactor Core Isolation Cooling System, Rev. 70 Proposed references to be provided to applicants during examination:           NONE Learning Objective:   OPS Reactor Core Isolation Cooling/COR002-18-02 R25
OPS Containment COR002 02 R30   6. Describe the interrelationship between PCIS and the following:
: 10. Predict the consequences of the following on the RCIC system:
: f. RCIC Question Source:
: n.     Steam line break (Steam Tunnel/RCIC Room)
Bank #   Modified Bank #
OPS Containment COR002-03-02 R30
New X Question History:
: 6.     Describe the interrelationship between PCIS and the following:
: f.     RCIC Question Source:               Bank #
Modified Bank #
New X Question History:
Question Cognitive Level:      Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content:        55.41 (7)
Comments:
LOD 3 85


Question Cognitive Level:
86 87 ES-401                             Sample Written Examination                     Form ES-401-5 Question Worksheet Examination Outline Cross-
Memory or Fundamental Knowledge      Comprehension or Analysis X    10 CFR Part 55 Content:
55.41 (7)  Comments:
LOD 3  85 86 87 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
-


==Reference:==
==Reference:==
 
Level                 RO Tier #               2 Group #               2 K/A #                 201001 A3.08 Importance Rating     3.0 201001 CRD Hydraulic - Ability to monitor automatic operations of the control rod drive hydraulic system including: (CFR: 41.7) A3.08 Drive water flow Question: 54 The Plant is performing a Reactor Startup.
Level   RO   Tier #     2   Group # 2 K/A #   201001 A3.08 Importance Rating
* Control Rod 26-27 is selected.
 
* A single notch withdrawal signal is applied to Control Rod 26-27.
===3.0 201001===
What is the flow indication on Panel 9-5 CRD-FI-305, (Drive Water Flow) while the rod is withdrawing?
CRD Hydraulic  
A.     2 gpm B.     4 gpm C.     6 gpm D.     8 gpm Answer:
- Ability to monitor automatic operations of the control rod drive hydraulic system including: (CFR: 41.7) A3.08 Drive water flow Question: 54 The Plant is performing a Reactor Startup.
A.     2 gpm Explanation:
Control Rod 26
DRIVE WATER LINE: This line connects the drive water header to the HCU manifold. It is normally pressurized to 265 psi above reactor pressure.
-27 is selected.
WITHDRAWAL LINE: The withdrawal line connects the HCU manifold to the CRDM above piston area. The withdrawal line is pressurized with drive water when the associated CRDM is being withdrawn.
A single notch withdrawal signal is applied to Control Rod 26
-27. What is the flow indication on Panel 9
-5 CRD-FI-305, (Drive Water Flow) while the rod is withdrawing
?  A. 2 gpm                      
 
B. 4 gpm C. 6 gpm                   D. 8 gpm                   Answer: A. 2 gpm Explanation:
DRIVE WATER LINE: This line connects the drive water header to the HCU manifold. It is normally pressurized to 265 psi above reactor pressure.
WITHDRAWAL LINE: The withdrawal line connects the HCU manifold to the CRDM above piston area. The withdrawal line is pressurized with drive water when the associated CRDM is being withdrawn.
The insert solenoid valve has a throttle valve set to allow the normal insertion flow rate through that stabilizing valve, when it is open (energized). When there is no rod movement, the flow through the insert solenoid valve will be approximately 4 gpm. When there is an insert rod signal from the REACTOR MANUAL CONTROL SYSTEM (RMCS), the insert solenoid valve will close, balancing the 4 gpm flow directed to an HCU for normal insertion of a control rod.
The insert solenoid valve has a throttle valve set to allow the normal insertion flow rate through that stabilizing valve, when it is open (energized). When there is no rod movement, the flow through the insert solenoid valve will be approximately 4 gpm. When there is an insert rod signal from the REACTOR MANUAL CONTROL SYSTEM (RMCS), the insert solenoid valve will close, balancing the 4 gpm flow directed to an HCU for normal insertion of a control rod.
The withdrawal solenoid valve has a throttle valve set to allow the normal withdrawal flow rate through the stabilizing valve when it is open (energized). When there is no rod movement, the flow through the withdrawal solenoid valve will be approximately 2 gpm. A withdraw signal from RMCS will close the valve balancing the 2 gpm flow directed to an HCU for normal withdrawal of a control rod.
The withdrawal solenoid valve has a throttle valve set to allow the normal withdrawal flow rate through the stabilizing valve when it is open (energized). When there is no rod movement, the flow through the withdrawal solenoid valve will be approximately 2 gpm. A withdraw signal from RMCS will close the valve balancing the 2 gpm flow directed to an HCU for normal withdrawal of a control rod.
Flow from the stabilizing valves passes through a local flow indicator. This local indication is used to adjust the throttle valves and verify proper operation. The normal reading, with no rod movement, is 6 gpm.
Flow from the stabilizing valves passes through a local flow indicator. This local indication is used to adjust the throttle valves and verify proper operation. The normal reading, with no rod movement, is 6 gpm.
88 When the insert valve shuts, the withdrawal valve is still open; therefore the flow indicator will show 2 gpm flow through the stabilizing valves. When the withdrawal valve shuts, the insert valve is still open; therefore the flow indicator will show 4 gpm flow through the stabilizing valves. Distracters:
88
 
When the insert valve shuts, the withdrawal valve is still open; therefore the flow indicator will show 2 gpm flow through the stabilizing valves. When the withdrawal valve shuts, the insert valve is still open; therefore the flow indicator will show 4 gpm flow through the stabilizing valves.
Distracters:
B. This answer is incorrect due to CRD withdraw flow being equal to 2 gpm. This choice is plausible due to confusing insert the flow which is 4 gpm. The candidate who confuses the expected insert flow with withdraw flow would choose this option.
B. This answer is incorrect due to CRD withdraw flow being equal to 2 gpm. This choice is plausible due to confusing insert the flow which is 4 gpm. The candidate who confuses the expected insert flow with withdraw flow would choose this option.
C. This answer is incorrect due to CRD withdraw flow being equal to 2 gpm.
C. This answer is incorrect due to CRD withdraw flow being equal to 2 gpm.
This choice is plausible due to confusing total stabilizing flow which is 6 gpm. The  
This choice is plausible due to confusing total stabilizing flow which is 6 gpm. The candidate who confuses the total stabilizing flow with withdraw flow would choose this option.
D. This answer is incorrect due to CRD withdraw flow being equal to 2 gpm. This choice is plausible due to confusing total stabilizing flow which is 6 gpm and withdraw flow of 2 gpm.
A candidate may know that 6 gpm flow is the normal stabilizing flow and that when a withdrawal occurs that this flow is added to the 6 gpm (yielding 8 gpm) and choose this option.
Technical Reference(s): SOP 2.2.8 (Control Rod Drive Hydraulic System), Rev. 90 Proposed references to be provided to applicants during examination:          NONE Learning Objective:      LO-9 Given a CRDH system component manipulation, predict and explain the changes in the following parameters:
: h.      CRD drive water flow Question Source:                Bank #
Modified Bank #
New X Question History:                Last NRC Exam          n\a Question Cognitive Level:        Memory or Fundamental Knowledge              X Comprehension or Analysis 10 CFR Part 55 Content:          55.41 (6)
Comments:
LOD 3 89


candidate who confuses the total stabilizing flow with withdraw flow would choose this option.
90 91 92 ES-401                             Sample Written Examination                     Form ES-401-5 Question Worksheet Examination Outline Cross-
D. This answer is incorrect due to CRD withdraw flow being equal to 2 gpm. This choice is plausible due to confusing total stabilizing flow which is 6 gpm and withdraw flow of 2 gpm. A candidate may know that 6 gpm flow is the normal stabilizing flow and that when a withdrawal occurs that this flow is added to the 6 gpm (yielding 8 gpm) and choose this option. Technical Reference(s):  SOP 2.2.8 (Control Rod Drive Hydraulic System), Rev. 90 Proposed references to be provided to applicants during examination:
NONE  Learning Objective:
LO-9 Given a CRDH system component manipulation, predict and explain the changes in the following parameters:
: h. CRD drive water flow Question Source:
Bank #    Modified Bank #
New X  Question History:
Last NRC Exam n\a  Question Cognitive Level:
Memory or Fundamental Knowledge X Comprehension or Analysis    10 CFR Part 55 Content:
55.41 (6)  Comments: LOD 3  89 90 91 92 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
-


==Reference:==
==Reference:==
 
Level                   RO Tier #                   2 Group #                 2 K/A #                   201006 A4.04 Importance Rating       3.3 201006 RWM - Ability to manually operate and/or monitor in the control room:
Level   RO   Tier #     2   Group # 2 K/A #   201006 A4.04 Importance Rating
(CFR: 41.7 / 45.5 to 45.8) A4.04 Rod withdrawal error indication Question: 55 The Plant is conducting a Reactor startup with the following conditions present:
 
* Group 3 Rod Movement Sheet Insert Limit: 04
===3.3 201006===
* Group 3 Rod Movement Sheet Withdraw Limit: 08
RWM  
* Control Rod 26-27 is selected and is the first of four rods in the Group.
- Ability to manually operate and/or monitor in the control room:
* When Control Rod 26-27 is withdrawn from 06 to 08, it double notches.
(CFR: 41.7 / 45.5 to 45.8) A4.04 Rod withdrawal error indication Question: 55
 
The Plant is conducting a Reactor startup with the following conditions present:
Group 3 Rod Movement Sheet Insert Limit: 04 Group 3 Rod Movement Sheet Withdraw Limit: 08 Control Rod 26
-27 is selected and is the first of four rods in the Group.
When Control Rod 26
-27 is withdrawn from 06 to 08, it double notches.
Which RWM IDT targets turn RED?
Which RWM IDT targets turn RED?
A. WITHDRAW BLOCK only. B. WITHDRAW BLOCK and OUT OF SEQUENCE only. C. WITHDRAW BLOCK, OUT OF SEQUENCE and SELECT ERROR only. D. WITHDRAW BLOCK, INSERT BLOCK, OUT OF SEQUENCE and SELECT ERROR only. Answer: B. WITHDRAW BLOCK and OUT OF SEQUENCE only. Explanation:
A. WITHDRAW BLOCK only.
B. WITHDRAW BLOCK and OUT OF SEQUENCE only.
C. WITHDRAW BLOCK, OUT OF SEQUENCE and SELECT ERROR only.
D. WITHDRAW BLOCK, INSERT BLOCK, OUT OF SEQUENCE and SELECT ERROR only.
Answer:
B. WITHDRAW BLOCK and OUT OF SEQUENCE only.
Explanation:
Rod 26-27 has moved beyond the Withdraw Limit of Group 3 (08) and is at position 10.
Rod 26-27 has moved beyond the Withdraw Limit of Group 3 (08) and is at position 10.
The RWM allows only 1 withdrawal error and will impose rod blocks to prevent further progress until the withdraw error is corrected. A withdraw error below the LPSP causes WITHDRAW BLOCK and OUT OF SEQUENCE to turn RED.
The RWM allows only 1 withdrawal error and will impose rod blocks to prevent further progress until the withdraw error is corrected. A withdraw error below the LPSP causes WITHDRAW BLOCK and OUT OF SEQUENCE to turn RED.
Line 2,463: Line 2,363:
Whenever the RWM is OPERATING < LPSP, the existence of a WITHDRAW ERROR causes OUT OF SEQUENCE (red) and WITHDRAW BLOCK (red) to display.
Whenever the RWM is OPERATING < LPSP, the existence of a WITHDRAW ERROR causes OUT OF SEQUENCE (red) and WITHDRAW BLOCK (red) to display.
If, under these same conditions, the selection of any other control rod (other than the one causing the WITHDRAW ERROR) will cause OUT OF SEQUENCE (red), SELECT ERROR (red), INSERT BLOCK (red) and WITHDRAW BLOCK (red) to display.
If, under these same conditions, the selection of any other control rod (other than the one causing the WITHDRAW ERROR) will cause OUT OF SEQUENCE (red), SELECT ERROR (red), INSERT BLOCK (red) and WITHDRAW BLOCK (red) to display.
Whenever a WITHDRAW ERROR occurs in the mode of enforcement (OPERATING < LPSP), the RWM will ONLY permit error rods to be repositioned.
Whenever a WITHDRAW ERROR occurs in the mode of enforcement (OPERATING < LPSP),
the RWM will ONLY permit error rods to be repositioned.
Distracters:
Distracters:
93 A. This option is incorrect because Control Rod 26-27 is beyond the allowable withdrawal position, therefore, RWM IDT Withdraw Block AND Out of Sequence targets turn red. This choice is plausible due to being partially correct. The candidate that does not remember the In Sequence target also turns red would select this answer.
93


C. This option is incorrect because with Control Rod 26
A. This option is incorrect because Control Rod 26-27 is beyond the allowable withdrawal position, therefore, RWM IDT Withdraw Block AND Out of Sequence targets turn red. This choice is plausible due to being partially correct. The candidate that does not remember the In Sequence target also turns red would select this answer.
-27 beyond the allowable withdrawal position only RWM IDT Withdraw Block AND Out of Sequence targets are red. The candidate may choose this if they believe the Select Error turns red also because control rod 26-27 is selected.
C. This option is incorrect because with Control Rod 26-27 beyond the allowable withdrawal position only RWM IDT Withdraw Block AND Out of Sequence targets are red. The candidate may choose this if they believe the Select Error turns red also because control rod 26-27 is selected.
D. This option is incorrect because Control Rod 26
D. This option is incorrect because Control Rod 26-27 is beyond the allowable withdrawal position, therefore, only RWM IDT Withdraw Block AND Out of Sequence targets will turn red. The candidate may choose this if they believe the Select Error is Red due to the rod remaining selected and the insert error is present due to the rod double notching in which case this would be the correct answer.
-27 is beyond the allowable withdrawal position, therefore, only RWM IDT Withdraw Block AND Out of Sequence targets will turn red. The candidate may choose this if they believe the Select Error is Red due to the rod remaining selected and the insert error is present due to the rod double notching in which case this would be the correct answer.
Technical Reference(s): Procedure 4.2 (Rod Worth Minimizer), Rev. 29 Proposed references to be provided to applicants during examination:        NONE Learning Objective:
COR002-26-02, OPS Rod Worth Minimizer LO-01                  State the purpose of the following items related to the Rod Worth Minimizer:
: g. IDT Display Console LO-05          Briefly describe the following concepts as they apply to the RWM:
: g.      Withdraw error Question Source:              Bank #
Modified Bank #
New X Question History:              Last NRC Exam          n\a Question Cognitive Level:      Memory or Fundamental Knowledge Comprehension or Analysis            X 10 CFR Part 55 Content:        55.41 (7)
Comments:
LOD 3 94


Technical Reference(s):  Procedure 4.2 (Rod Worth Minimizer), Rev. 29 Proposed references to be provided to applicants during examination:
95 96 ES-401                           Sample Written Examination                     Form ES-401-5 Question Worksheet Examination Outline Cross-
NONE Learning Objective:
COR002-26-02, OPS Rod Worth Minimizer LO-01 State the purpose of the following items related to the Rod Worth Minimizer:
: g. IDT Display Console LO-05  Briefly describe the following concepts as they apply to the RWM:    g. Withdraw error Question Source:
Bank #    Modified Bank #
New X  Question History:
Last NRC Exam n\a  Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis  X  10 CFR Part 55 Content:
55.41 (7)  Comments: LOD 3  94 95 96 ES-401 Sample Written Examination Fo rm ES-401-5 Question Worksheet Examination Outline Cross
-


==Reference:==
==Reference:==
 
Level                 RO Tier #                 2 Group #               2 K/A #         202001 2.2.22 Importance Rating     4.0 202001 Recirculation G2.2.22- Knowledge of limiting conditions for operations and safety limits.
Level   RO   Tier #     2   Group # 2 K/A # 202001 2.2.22 Importance Rating
(CFR: 41.10 / 43.2 / 45.13)
 
Question: 56 Reactor power steady at 26%.
===4.0 202001===
Recirculation G2.2.22- Knowledge of limiting conditions for operations and safety limits. (CFR: 41.10 / 43.2 / 45.13)
Q ues ti on: 5 6  Reactor power steady at 26%.
What is the HIGHEST Core Flow which is a TS Safety Limit Violation?
What is the HIGHEST Core Flow which is a TS Safety Limit Violation?
CORE FLOW     A. 9% B. 10%
CORE FLOW A.     9%
C. 11% D. 13% A ns w e r: A. 9% E xp l ana ti on:   TS requires if either reactor steam dome pressure is < 785 psig or core flow < 10%, THERMAL POWER shall be  
B.     10%
< 25%. With reactor power at 26%, core flow at 9% violates the safety limit.
C.     11%
D.     13%
Answer:
A.     9%
Explanation:
TS requires if either reactor steam dome pressure is < 785 psig or core flow < 10%, THERMAL POWER shall be < 25%. With reactor power at 26%, core flow at 9% violates the safety limit.
Distracters:
Distracters:
B. This answer is incorrect because at 10% the safety limit is satisfied. This answer is plausible due to the MCPR safety limit being applicable at  
B. This answer is incorrect because at 10% the safety limit is satisfied. This answer is plausible due to the MCPR safety limit being applicable at > 10% core flow The candidate that confuses the MCPR safety limit with the Low Pressure of Flow safety limit would select this answer.
> 10% core flow The candidate that confuses the MCPR safety limit with the Low Pressure of Flow safety limit would select this answer. C. This answer is incorrect because at 11% the safety limit is satisfied. This answer is plausible due to the MCPR safety limit being > 1.11 for two loop operation and can be easily confused with 11. The candidate that confuses the two loop MCPR limit with the Low Pressure of Flow safety limit would select this answer.
C. This answer is incorrect because at 11% the safety limit is satisfied. This answer is plausible due to the MCPR safety limit being > 1.11 for two loop operation and can be easily confused with 11. The candidate that confuses the two loop MCPR limit with the Low Pressure of Flow safety limit would select this answer.
D. This answer is incorrect because at 13% the safety limit is satisfied. This answer is plausible due to the MCPR safety limit being > 1.13 for single loop operation and can be easily confused with 13. The candidate that confuses the single loop MCPR limit with the Low Pressure of Flow safety limit would select this answer.
D. This answer is incorrect because at 13% the safety limit is satisfied. This answer is plausible due to the MCPR safety limit being > 1.13 for single loop operation and can be easily confused with 13. The candidate that confuses the single loop MCPR limit with the Low Pressure of Flow safety limit would select this answer.
97 Technical Reference(s):
97
Technical Specifications 2.0 Safety Limits Proposed references to be provided to applicants during examination: None Learning Objective:
INT00705010010800 From memory, state each CNS Safety Limit and discuss the basis for each of the Safety Limits.
Question Source:
Bank #    Modified Bank #
New X  Question Cognitive Level:
Memory or Fundamental Knowledge  X Comprehension or Analysis    10 CFR Part 55 Content:
55.41 (10)
Comments: LOD 3  98 99 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
-


==Reference:==
Technical Reference(s):     Technical Specifications 2.0 Safety Limits Proposed references to be provided to applicants during examination:    None Learning Objective:  INT00705010010800 From memory, state each CNS Safety Limit and discuss the basis for each of the Safety Limits.
Question Source:            Bank #
Modified Bank #
New X Question Cognitive Level:    Memory or Fundamental Knowledge            X Comprehension or Analysis 10 CFR Part 55 Content:      55.41 (10)
Comments:
LOD 3 98


Level  RO  Tier #    2  Group #  2  K/A #  204000 K1.03 Importance Rating
99 ES-401                            Sample Written Examination                        Form ES-401-5 Question Worksheet Examination Outline Cross-


===3.1 204000===
==Reference:==
RWCU  
Level                  RO Tier #                  2 Group #                2 K/A #                  204000 K1.03 Importance Rating      3.1 204000 RWCU - Knowledge of the physical connections and/or cause-effect relationships between reactor water cleanup system and the following: (CFR: 41.2 to 41.9 / 45.7 to 45.8)
- Knowledge of the physical connections and/or cause
- K1.03 Reactor feedwater system Question: 57 Where does the Reactor Water Cleanup System (RWCU) return water to the reactor?
-effect relationships between reactor water cleanup system and the following: (CFR: 41.2 to 41.9 / 45.7 to 45.8)  
Via the A. suction side of A reactor recirculation pump.
- K1.03 Reactor feedwater system
 
Question: 57 W he r e doe s t h e R eac t o r W a t e r C l eanu p S ys t e m (R W CU) r e t u r n w a t e r t o t h e r eac t or?   Via the-  A. suction side of A reactor recirculation pump.
B. suction side of B reactor recirculation pump.
B. suction side of B reactor recirculation pump.
C. A feedwater line downstream of the outboard check valve. D. B feedwater line downstream of the outboard check valve.
C. A feedwater line downstream of the outboard check valve.
Answer: C. A feedwater line downstream of the outboard check valve.
D. B feedwater line downstream of the outboard check valve.
Answer:
C. A feedwater line downstream of the outboard check valve.
Explanation:
When the return isolation valve is in its normal (open) position it allows return flow from the RHX to the Reactor Feedwater piping, where the water is returned to the reactor vessel via feedwater line "A".
Distractors:
A. This option is incorrect because RWCU returns water to the A FW line. RWCU does however get its water supply form the A RR loop so a candidate who only knows that there is a physical tie to the RR loop but not the purpose of that connection would choose this option.
B. This option is incorrect because the RWCU water returns to the reactor via the A FW line.
The RWCU system does interconnect to the RR system so a candidate who only knows there is a connection but neither specific location nor the purpose of that connection would choose this option.
D: This option is incorrect because the RWCU system return water to the reactor is via the A FW line not the B FW line. A candidate who knows that the return is via the FW system but not which line would choose this option.
100


Explanation:
Technical Reference(s):     Procedure 2.2.66 (Reactor Water Cleanup System), Rev. 104 Proposed references to be provided to applicants during examination:      NONE Learning Objective:  OPS Reactor Water Cleanup COR001-20-01 LO-4  Briefly describe the interrelationship between the RWCU system and the following:
When the return isolation valve is in its normal (open) position it allows return flow from the RHX to the Reactor Feedwater piping, where the water is returned to the reactor vessel via feedwater line "A". Distractors:
: j.     Reactor Feedwater system Question Source:            Bank #
A. This option is incorrect because RWCU returns water to the A FW line. RWCU does however get its water supply form the A RR loop so a candidate who only knows that there is a physical tie to the RR loop but not the purpose of that connection would choose this option. B. This option is incorrect because the RWCU water returns to the reactor via the A FW line. The RWCU system does interconnect to the RR system so a candidate who only knows there is a connection but neither specific location nor the purpose of that connection would choose this option.
Modified Bank #
D: This option is incorrect because the RWCU system return water to the reactor is via the A FW line not the B FW line. A candidate who knows that the return is via the FW system but not which line would choose this option.
New X Question History:          Last NRC Exam          n\a Question Cognitive Level:  Memory or Fundamental Knowledge              X Comprehension or Analysis 10 CFR Part 55 Content:    55.41 (4)
Comments:
LOD 2 101


100 Technical Reference(s):
102 103 ES-401                             Sample Written Examination                     Form ES-401-5 Question Worksheet Examination Outline Cross-
Procedure 2.2.66 (Reactor Water Cleanup System), Rev. 104 Proposed references to be provided to applicants during examination:
NONE Learning Objective:
OPS Reactor Water Cleanup COR001 01      LO-4 Briefly describe the interrelationship between the RWCU system and the following:
: j. Reactor Feedwater system Question Source:
Bank #    Modified Bank #
New X  Question History:
Last NRC Exam n\a  Question Cognitive Level:
Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content:
55.41 (4)  Comments:
LOD 2  101 102 103 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
-


==Reference:==
==Reference:==
Level                RO Tier #                2 Group #              2 K/A #                216000 K2.01 Importance Rating    2.8 216000 Nuclear Boiler Inst. - Knowledge of electrical power supplies to the following:
(CFR: 41.7) K2.01 Analog trip system: Plant-Specific Question: 58 What supplies power to the switch contact logic of Wide Range Level Indicating Switch, NBI-LIS-57A (PCIS initiation logic)?
A. AA-1 B. NBPP A C. RPSPP1A D. 24V Power Panel DC-A Answer:
C. RPSPP1A Explanation:
The Wide Range level instrument NBI-LIS-57A is a Yarway type level indicating switch located on Local Rack LRP-PNL-(25-5A) in the Reactor Building. This switch is powered from 120 VAC RPSPP1A. This switch provides partial initiation to the following functions: Alternate Rod Insertion/Recirculation Pump Trip (ARI/RPT), Groups 1, 3, 6 and 7 isolations as well as Control Room Emergency Filtration System (CREFS) initiation.
Distracters:
A. This answer is incorrect because AA-1 does not supply power to any NBI level instrument or the logic power to any NBI level switch contact. This answer is plausible because panel AA-2 supplies power to narrow range level instruments and the candidate may confuse these two panels and choose this answer.
B. This answer is incorrect because NBPP-1A supplies RPV level Indicators NBI-LI-94A and NBI-LR/PR-97. This power supply is in the same division of power as NBI-LIS-57A, and is therefore plausible.
D: This answer is incorrect because the 24 VDC A supplies neutron monitoring and radiation monitor trip auxiliaries. This answer is plausible because it powers instruments on the same panel and close proximity to the RPV level instruments.
Technical Reference(s):        OPS Nuclear Boiler Instrumentation/COR002-15-02 Rev. 26 SOP 2.2.22 Rev. 70 Vital Instrument Power System 104


Level  RO  Tier #    2  Group #  2  K/A #  216000 K2.01 Importance Rating
Proposed references to be provided to applicants during examination:   NONE Learning Objective:
 
===2.8 216000===
Nuclear Boiler Inst.
- Knowledge of electrical power supplies to the following:
(CFR: 41.7) K2.01 Analog trip system: Plant
-Specific  Q ues ti on: 5 8  W ha t supplies po w e r to the switch contact logic of Wide Range Level Indicating Switch, NB I-L I S-57 A (PCIS initiation logic)
?  A. AA-1  B. NBPP A  C. RPSPP 1 A  D. 24V Power Panel DC
-A A ns w e r: C. RPSPP 1 A  Explanation:
The Wide Range level instrument NBI
-LIS-57A is a Yarway type level indicating switch located on Local Rack LRP
-PNL-(25-5A) in the Reactor Building. This switch is powered from 120 VAC RPSPP1A. This switch provides partial initiation to the following functions:  Alternate Rod Insertion/Recirculation Pump Trip (ARI/RPT), Groups 1, 3, 6 and 7 isolations as well as Control Room Emergency Filtration System (CREFS) initiation.
Distracters:
A. This answer is incorrect because AA
-1 does not supply power to any NBI level instrument or the logic power to any NBI level switch contact. This answer is plausible because panel AA
-2 supplies power to narrow range level instruments and the candidate may confuse these two panels and choose this answer.
B. This answer is incorrect because NBPP
-1A supplies RPV level Indicators NBI
-LI-94A and NBI-LR/PR-97. This power supply is in the same division of power as NBI
-LIS-57A, and is therefore plausible.
D:  This answer is incorrect because the 24 VDC A supplies neutron monitoring and radiation monitor trip auxiliaries. This answer is plausible because it powers instruments on the same panel and close proximity to the RPV level instruments.
Technical Reference(s):
OPS Nuclear Boiler Instrumentation/COR002 02 Rev. 26 SOP 2.2.22 Rev. 70 Vital Instrument Power System 104 Proposed references to be provided to applicants during examination:
NONE Learning Objective:
COR002-15-02 Rev. 26, OPS Nuclear Boiler Instrumentation LO-05 Predict the consequences of the following on the NBI:
COR002-15-02 Rev. 26, OPS Nuclear Boiler Instrumentation LO-05 Predict the consequences of the following on the NBI:
: k. Loss of AC power
: k. Loss of AC power Question Source:            Bank #
Modified Bank #
New X Question History:          N/A Question Cognitive Level:  Memory or Fundamental Knowledge            X Comprehension or Analysis 10 CFR Part 55 Content:    55.41 (7)
Comments:
LOD 4 105


Question Source:
106 107 ES-401                           Sample Written Examination                     Form ES-401-5 Question Worksheet Examination Outline Cross-
Bank #    Modified Bank #
New X  Question History:
N/A  Question Cognitive Level:
Memory or Fundamental Knowledge X Comprehension or Analysis    10 CFR Part 55 Content:
55.41 (7)  Comments: LOD 4    105 106 107 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
-


==Reference:==
==Reference:==
 
Level                 RO Tier #               2 Group #               2 K/A #                 226001 K3.02 Importance Rating     3.5 226001 RHR/LPCI: CTMT Spray Mode - Knowledge of the effect that a loss or malfunction of the RHR/LPCI: containment spray system mode will have on following: (CFR: 41.7 /
Level   RO   Tier #     2   Group # 2 K/A #   226001 K3.02 Importance Rating
45.4) - K3.02 Containment/drywell/suppression chamber temperature Question: 59 The Plant is operating at power when a steam leak occurs in the Drywell.
 
* A reactor scram occurs.
===3.5 226001===
* RHR is spraying the torus and drywell.
RHR/LPCI: CTMT Spray Mode  
* Torus pressure is 11 psig lowering 1 psig/minute.
- Knowledge of the effect that a loss or malfunction of the RHR/LPCI: containment spray system mode will have on following: (CFR: 41.7 / 45.4) - K3.02 Containment/drywell/suppression chamber temperature Question: 59 The Plant is operating at power when a steam leak occurs in the Drywell.
* Drywell temperature is 245&deg;F lowering slowly.
A reactor scram occurs.
RHR is spraying the torus and drywell.
Torus pressure is 11 psig lowering 1 psig/minute.
Drywell temperature is 245&deg;F lowering slowly.
A logic failure has caused the loss of spray valve control permissive.
A logic failure has caused the loss of spray valve control permissive.
What is the effect on drywell temperature?
What is the effect on drywell temperature?
Drywell temperature -
Drywell temperature A. rises due to the loss of Drywell FCUs.
A. rises due to the loss of Drywell FCUs.
B. rises due to the loss of Drywell Spray.
B. rises due to the loss of Drywell Spray.
C. lowers due to the start of Drywell FCUs.
C. lowers due to the start of Drywell FCUs.
D. lowers due to the Drywell Spray valves going full open.
D. lowers due to the Drywell Spray valves going full open.
Answer: B. rises due to the loss of Drywell Spray.
Answer:
B. rises due to the loss of Drywell Spray.
Explanation:
Explanation:
RHR must have spray valve control in order to spray containment. With a loss of the permissive, the spray valves close and the loss of containment cooling occurs. Drywell temperature rises due to the steam leak continuing.
RHR must have spray valve control in order to spray containment. With a loss of the permissive, the spray valves close and the loss of containment cooling occurs. Drywell temperature rises due to the steam leak continuing.
A. This answer is incorrect because under the current conditions loss of DW FCUs does not cause DW Temperature to rise. The DW FCUs are secured prior to initiating DW Spray, therefore with temperature lowering, the only option for DW temperature to rise is due to the leak and loss of DW Spray ONLY. This choice is plausible because loss of DW FCUs normally would cause DW temperature to rise, but due to being secured prior to spray would have no impact on the rising DW temperature due to loss of DW Spray. The candidate that does not recognize DW FCUs being secured prior to initiating DW spray would select this answer. 108 C. This answer is incorrect because the Drywell FCUs do not start following loss of spray. This answer is plausible due to the DW FCUs having to be secured (control Switch placed in OFF overrides auto start) prior to spraying the DW. The FCUs have a STANDBY feature which auto starts if DW Temperature reaches 145&deg;F. The candidate who does not understand the condition of the DW FCUs would select this answer.
A. This answer is incorrect because under the current conditions loss of DW FCUs does not cause DW Temperature to rise. The DW FCUs are secured prior to initiating DW Spray, therefore with temperature lowering, the only option for DW temperature to rise is due to the leak and loss of DW Spray ONLY. This choice is plausible because loss of DW FCUs normally would cause DW temperature to rise, but due to being secured prior to spray would have no impact on the rising DW temperature due to loss of DW Spray. The candidate that does not recognize DW FCUs being secured prior to initiating DW spray would select this answer.
108
 
C. This answer is incorrect because the Drywell FCUs do not start following loss of spray. This answer is plausible due to the DW FCUs having to be secured (control Switch placed in OFF overrides auto start) prior to spraying the DW. The FCUs have a STANDBY feature which auto starts if DW Temperature reaches 145&deg;F. The candidate who does not understand the condition of the DW FCUs would select this answer.
D. This answer is incorrect because the Drywell Spray valves receive a closure signal on the loss of spray valve permissive. This answer is plausible because some RHR valves receive an open signal with a LPCI signal present. The candidate who does not know the effects of loss of spray valve control would select this answer.
D. This answer is incorrect because the Drywell Spray valves receive a closure signal on the loss of spray valve permissive. This answer is plausible because some RHR valves receive an open signal with a LPCI signal present. The candidate who does not know the effects of loss of spray valve control would select this answer.
Technical Reference(s):        OPS Residual Heat Removal System/COR002-23-02 EOP-5.8.7 Rev.29 Primary Containment Flooding/Spray Systems SOP 2.2.69.3 Rev.46 EOP-5.8 Attachment 2 EOP/SAG Graphs Rev.15 Proposed references to be provided to applicants during examination:      None Learning Objective:
Per COR002-23-02, OPS Residual Heat Removal System 7      Given a specific RHR system malfunction, determine the effect on any of the following:
: d.      Drywell parameters (pressure, temperature)
Question Source:                Bank #
Modified Bank #
New X Question History:              Last NRC Exam        n\a Question Cognitive Level:      Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content:        55.41 (7)
Comments:        LOD 3 109


Technical Reference(s):
110 111 112 ES-401                             Sample Written Examination                   Form ES-401-5 Question Worksheet Examination Outline Cross-
OPS Residual Heat Removal System/COR002 02                                                EOP-5.8.7 Rev.29 Primary Containment Flooding/Spray Systems SOP 2.2.69.3 Rev.46 EOP-5.8 Attachment 2 EOP/SAG Graphs Rev.15 Proposed references to be provided to applicants during examination:
None  Learning Objective:
Per COR002-23-02, OPS Residual Heat Removal System 7 Given a specific RHR system malfunction, determine the effect on any of the following:
: d. Drywell parameters (pressure, temperature)
Question Source:
Bank #    Modified Bank #
New X  Question History:
Last NRC Exam n\a Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X  10 CFR Part 55 Content:
55.41 (7)  Comments: LOD 3  109 110 111 112 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
-


==Reference:==
==Reference:==
 
Level                 RO Tier #               2 Group #               2 K/A #                 230000 K4.05 Importance Rating     2.8 230000 RHR/LPCI: Torus/Pool Spray Mode - Knowledge of RHR/LPCI: torus/suppression pool spray mode design feature(s) and/or interlocks which provide for the following:
Level   RO   Tier #     2   Group # 2 K/A #   230000 K4.05 Importance Rating
(CFR: 41.7) - K4.05 Pump minimum flow protection Question: 60 The plant is operating at 100% power when a small break LOCA occurs. The following conditions are present:
 
* RHR Loop A is in Torus Spray and Suppression Pool Cooling.
===2.8 230000===
* Torus pressure is 2.3 psig and lowering slowly.
RHR/LPCI: Torus/Pool Spray Mode  
What condition FIRST causes RHR-MO-16A (LOOP A MIN FLOW BYP VLV) to automatically OPEN as flow on RHR Loop A is being lowered?
- Knowledge of RHR/LPCI: torus/suppression pool spray mode design feature(s) and/or interlocks which provide for the following:
RHR-MO-16A opens A. 3.5 seconds after flow lowers to less than 490 gpm.
(CFR: 41.7)  
B. 3.5 seconds after flow lowers to less than 2107 gpm.
- K4.05 Pump minimum flow protection Q ues ti on: 6 0  Th e p l an t i s ope r a ti n g a t 100% po w e r w he n a s m a l l b r ea k LOC A occu r s. T h e f o ll o w i ng conditi on s a r e p resen t:   RH R Loo p A is i n T o r u s S p r a y an d S upp ress i o n P oo l C oo li n g. To r u s p ressu r e i s 2.3 ps i g an d lowering s l o w l y. W ha t conditi o n F I RS T cause s RH R-MO-16A (LOOP A MIN FLOW BYP VLV) t o automatically OPEN a s fl o w o n RH R L oo p A i s be i n g lowered? RH R-MO-16 A opens -  A. 3.5 seconds after flow lowers to less than 4 90 gp m. B. 3.5 seconds after flow lowers to less than 210 7 gp m. C. as soon as flow is less than 490 gpm with no time delay
C. as soon as flow is less than 490 gpm with no time delay.
. D. as soon as flow is less than 210 7 gpm with no time delay. A ns w e r: B. 3.5 seconds after flow lowers to less than 210 7 gp m. Explanation:
D. as soon as flow is less than 2107 gpm with no time delay.
Minimum Flow Valves RHR
Answer:
-MO-16A/B - The pump minimum flow valves, one in each loop, provide the necessary flow through the pump in order to prevent pump overheating. The RHR pump minimum flow control valves are normally open. The RHR
B. 3.5 seconds after flow lowers to less than 2107 gpm.
-MO-16A valve closes whe n -MO-20 (RHR Cross
Explanation:
-tie) is not either Loop A or B for 3.5 seconds if RHR
Minimum Flow Valves RHR-MO-16A/B - The pump minimum flow valves, one in each loop, provide the necessary flow through the pump in order to prevent pump overheating. The RHR pump minimum flow control valves are normally open. The RHR-MO-16A valve closes when flow is 2107 gpm for 3.5 seconds in the associated loop if RHR-MO-20 (RHR Cross-tie) is not full open, or 2107 gpm in either Loop A or B for 3.5 seconds if RHR-MO-20 is full open. The valve opens 3.5 seconds after flow is less than 2107 gpm on system shutdown.
-MO-20 is full open. The valve opens 3.5 seconds after flow is less than 210 7 gp m on system shutdown.
Distracters:
Distracters:
A. This answer is incorrect because the RHR minimum flow control valve opens when flow is less than 2107 gpm for 3.5 seconds in the associated loop. The flow value of 490 is plausible because it is the HPCI system minimum flow setting. The time delay of 3.5 seconds was used because it is the setting for the RHR minimum flow time delay. If the candidate confuses the HPCI system minimum flow value with the RHR minimum flow value, then this answer may be selected.
A. This answer is incorrect because the RHR minimum flow control valve opens when flow is less than 2107 gpm for 3.5 seconds in the associated loop. The flow value of 490 is plausible because it is the HPCI system minimum flow setting. The time delay of 3.5 seconds was used because it is the setting for the RHR minimum flow time delay. If the candidate confuses the HPCI system minimum flow value with the RHR minimum flow value, then this answer may be selected.
113 C. This answer is incorrect because the RHR minimum flow control valve opens when flow is less than 2107 gpm for 3.5 seconds in the associated loop. The flow value of 490 is plausible because it is the HPCI system minimum flow setting. The no time delay is als o associated with the HPCI system minimum flow valve. If the candidate confuses the HPCI system minimum flow value with the RHR minimum flow value, then this answer may be selected.
113
 
C. This answer is incorrect because the RHR minimum flow control valve opens when flow is less than 2107 gpm for 3.5 seconds in the associated loop. The flow value of 490 is plausible because it is the HPCI system minimum flow setting. The no time delay is also associated with the HPCI system minimum flow valve. If the candidate confuses the HPCI system minimum flow value with the RHR minimum flow value, then this answer may be selected.
D. This answer is incorrect because the minimum flow control valve opens when flow is less than 2107 gpm for 3.5 seconds in the associated loop. The valve responds after a time delay. If the candidate does not remember the time delay, then this answer may be selected.
D. This answer is incorrect because the minimum flow control valve opens when flow is less than 2107 gpm for 3.5 seconds in the associated loop. The valve responds after a time delay. If the candidate does not remember the time delay, then this answer may be selected.
Technical Reference(s):
Technical Reference(s):             SOP 2.2.69 (Residual Heat Removal System), Rev. 97 Proposed references to be provided to applicants during examination:    NONE Learning Objective:  OPS Residual Heat Removal System/COR002-23-02 LO-3 Describe RHR system design feature(s) and/or interlocks which provide for the following:
SOP 2.2.69 (Residual Heat Removal System), Rev. 97
: c. Pump minimum flow protection Question Source:              Bank #
Modified Bank #
New X Question History:            Last NRC Exam          n\a Question Cognitive Level:    Memory or Fundamental Knowledge            X Comprehension or Analysis 10 CFR Part 55 Content:      55.41 (7)
Comments LOD 3 114


Proposed references to be provided to applicants during examination:
115 116 117 ES-401                             Sample Written Examination                       Form ES-401-5 Question Worksheet Examination Outline Cross-
NONE Learning Objective:
OPS Residual Heat Removal System/COR002 02    LO-3  Describe RHR system design feature(s) and/or interlocks which provide for the following:
: c. Pump minimum flow protection Question Source:
Bank #    Modified Bank #
New X  Question History:
Last NRC Exam n\a  Question Cognitive Level:
Memory or Fundamental Knowledge X Comprehension or Analysis    10 CFR Part 55 Content:
55.41 (7)  Comments LOD 3  114 115 116 117 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
-


==Reference:==
==Reference:==
 
Level                 RO Tier #                 2 Group #               2 K/A #                 241000 K5.04 Importance Rating     3.3 241000 Reactor/Turbine Pressure Regulator - Knowledge of the operational Implications of the following concepts as they apply to reactor/turbine pressure regulating system:
Level   RO   Tier #     2 Group # 2 K/A #   241000 K5.04 Importance Rating
(CFR: 41.5) K5.04 Turbine inlet pressure vs. reactor pressure.
 
Question: 61 During a startup the turbine is synchronized to the grid with the following conditions present:
===3.3 241000===
* Equalizing header pressure is 935 psig.
Reactor/Turbine Pressure Regulator  
* Reactor Pressure is 940 psig.
- Knowledge of the operational Implications of the following concepts as they apply to reactor/turbine pressure regulating system: (CFR: 41.5) K5.04 Turbine inlet pressure vs. reactor pressure.
* Reactor power is 25%.
Q ues ti on: 6 1  During a startup the turbine is synchronized to the grid with the following conditions present:
Equalizing header pressure is 935 psig.
Reactor Pressure is 940 psig.
Reactor power is 25%.
As power is raised from 25% to 100%, how does the pressure at the equalizing header and the reactor change, if at all, during the power ascension?
As power is raised from 25% to 100%, how does the pressure at the equalizing header and the reactor change, if at all, during the power ascension?
Equalizing header pressure -
Equalizing header pressure A. remains at 935 psig and reactor pressure rises to about 958 psig.
A. remains at 935 psig and reactor pressure rises to about 958 psig.
B. remains at 935 psig and reactor pressure rises to about 990 psig.
B. remains at 935 psig and reactor pressure rises to about 990 psig.
C. rises to about 958 psig and reactor pressure rises to about 958 psig.
C. rises to about 958 psig and reactor pressure rises to about 958 psig.
D. rises to about 958 psig and reactor pressure rises to about 990 psig. A ns w e r: D. rises to about 958 psig and reactor pressure rises to about 990 psig.
D. rises to about 958 psig and reactor pressure rises to about 990 psig.
When the main turbine is on
Answer:
-line, Throttle Pressure is controlled by modulating the governor valves. The pressure target is set via the Human Machine Interface (HMI). The DEH system gain is configured in the controller calculations is set to 3.33%. This gain results in a 3.33% flow demand change for every 1 psi of error sensed at the main steam equalizing header. A 30 psi error change results in a 100% flow demand change. This feature makes the response of the Throttle Pressure linear. RPV Steam Dome pressure varies based on the response of the DEH system to power and is a function of steam flow head loss. Since the head loss is a function of volume flow rate squared, the rise in reactor pressure is greater that the rise in the equalizing header. Distractors: A. This option is incorrect because the equalizing header pressure rises as power is raised. Even though pressure setpoint remains the same the offset between pressure setpoint and the equalizing header pressure provides the error to drive the governor valves to the new position. Reactor pressure does rise as indicated but rises more than to 958 psig. A 118 candidate may very well believe that since pressure setpoint is constant that equalizing header is constant and if that same candidate understands that reactor pressure rises due to head loss they would choose this option.
D. rises to about 958 psig and reactor pressure rises to about 990 psig.
 
When the main turbine is on-line, Throttle Pressure is controlled by modulating the governor valves. The pressure target is set via the Human Machine Interface (HMI). The DEH system gain is configured in the controller calculations is set to 3.33%. This gain results in a 3.33% flow demand change for every 1 psi of error sensed at the main steam equalizing header. A 30 psi error change results in a 100% flow demand change. This feature makes the response of the Throttle Pressure linear. RPV Steam Dome pressure varies based on the response of the DEH system to power and is a function of steam flow head loss. Since the head loss is a function of volume flow rate squared, the rise in reactor pressure is greater that the rise in the equalizing header.
B. This option is incorrect because even though reactor pressure does rise to about 990 psig, the equalizing header pressure does not remain at 935 psig as power is raised. Eve n
Distractors:
though pressure setpoint remains the same the offset between pressure setpoint and the equalizing header pressure provides the error to drive the governor valves to the new position so the offset between the setpoint and the actual equalizing header gets larger as power is raised. A candidate who understands the true response of reactor pressure but who does not understand that it is the offset between the pressure setpoint and equalizing header that provides the signal to open the governor valves would choose this option.
A. This option is incorrect because the equalizing header pressure rises as power is raised.
Even though pressure setpoint remains the same the offset between pressure setpoint and the equalizing header pressure provides the error to drive the governor valves to the new position. Reactor pressure does rise as indicated but rises more than to 958 psig. A 118


candidate may very well believe that since pressure setpoint is constant that equalizing header is constant and if that same candidate understands that reactor pressure rises due to head loss they would choose this option.
B. This option is incorrect because even though reactor pressure does rise to about 990 psig, the equalizing header pressure does not remain at 935 psig as power is raised. Even though pressure setpoint remains the same the offset between pressure setpoint and the equalizing header pressure provides the error to drive the governor valves to the new position so the offset between the setpoint and the actual equalizing header gets larger as power is raised. A candidate who understands the true response of reactor pressure but who does not understand that it is the offset between the pressure setpoint and equalizing header that provides the signal to open the governor valves would choose this option.
C. This option is incorrect because reactor pressure rises to greater than the new equalizing header pressure. This choice is a plausible misconception of reactor pressure and equalizing header differential pressure lowering (and equalizing) as reactor power is raised to rated (head losses become negligible at such a high steam flow). The candidate who confuses the head loss between the reactor and the equalizing header as power is raised would choose this option.
C. This option is incorrect because reactor pressure rises to greater than the new equalizing header pressure. This choice is a plausible misconception of reactor pressure and equalizing header differential pressure lowering (and equalizing) as reactor power is raised to rated (head losses become negligible at such a high steam flow). The candidate who confuses the head loss between the reactor and the equalizing header as power is raised would choose this option.
Technical Reference(s):              SOP 2.2.77.1 Rev.35, Digital Electro-Hydraulic Control System SOP 2.2.77 Rev.111, Turbine Generator Proposed references to be provided to applicants during examination:        None Learning Objective:    COR002-09-02 Rev.17
: 5. Explain the interrelation between the following parameter sets, and describe how their interrelationship affects operation of the DEH Control system.
: b. Turbine inlet pressure vs. reactor pressure Question Source:              Bank #
Modified Bank #
New                    X Question History:            Last NRC Exam Question Cognitive Level:    Memory or Fundamental Knowledge Comprehension or Analysis                    X 10 CFR Part 55 Content:      55.41    (5)
Comments:
LOD 3 119


Technical Reference(s):  SOP 2.2.77.1 Rev.35, Digital Electro
120 121 122 ES-401                             Sample Written Examination                     Form ES-401-5 Question Worksheet Examination Outline Cross-
-Hydraulic Control System      SOP 2.2.77 Rev.111, Turbine Generator
 
Proposed references to be provided to applicants during examination:
None  Learning Objective:
COR002-09-02 Rev.17    5. Explain the interrelation between the following parameter sets,    and describe how their interrelationship affects operation of the DEH Control system.
: b. Turbine inlet pressure vs. reactor pressure Question Source:
Bank #    Modified Bank #
N ew  X  Question History:
Last NRC Exam Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis  X  10 CFR Part 55 Content:
55.41    (5)
 
Comments: LOD 3  119 120 121 122 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
-


==Reference:==
==Reference:==
 
Level                 RO Tier #                 2 Group #               2 K/A #                 259001 Reactor Feedwater K6.04 Importance Rating     2.8 259001 Reactor Feedwater - Knowledge of the effect that a loss or malfunction of the following will have on the reactor feedwater system: (CFR: 41.7 / 45.7) - K6.04 Extraction steam Question: 62 The plant is operating at 95% power on the 100% rod line when the Non-Return Isolation Check Valve for Feedwater Heater 1-A-5 goes full closed due to a short in its motor operator.
Level   RO   Tier #     2   Group # 2 K/A #   259001 Reactor Feedwater K6.04 Importance Rating
(1) How does feed water temperature respond?
 
(2) Where does operation stabilize on the power-to-flow map?
===2.8 259001===
(1)                                               (2)
Reactor Feedwater  
Feedwater temperature                                 100%Rod Line A.     Lowers and returns to near the previous temperature.       Below B.     Stabilizes at a lower temperature ONLY.                     Below C.     Lowers and returns to near the previous temperature.       Above D.     Stabilizes at a lower temperature ONLY.                     Above Answer:
- Knowledge of the effect that a loss or malfunction of the following will have on the reactor feedwater system: (CFR: 41.7 / 45.7)  
D.     Stabilizes at a lower temperature ONLY.                     Above Explanation:
- K6.04 Extraction steam Question: 62 Th e p l an t i s ope r a ti n g a t 95% po w e r o n t h e 100% r o d li n e w he n t h e N on-R e t u r n I so l a ti o n C hec k V a l v e for Feed w a t e r H ea t e r 1-A-5 goe s f u l l c lose d du e to a sho r t i n it s m o t o r ope r a t o r. (1) How does feed water temperature respond?
The Extraction Steam system conducts steam from Main Turbine connections to two parallel Feedwater heater strings to improve the overall efficiency of the reactor by preheating the incoming feedwater and reducing the reactor heat load. The NON-RETURN ISOLATION CHECK VALVES (NRVs) are installed in the extraction steam supply lines to the Feedwater Heaters. NRV-1 is designed with an integral motor operator and when closed will essentially function as a stop check valve. When NRV-1 goes closed, feedwater Heater 1-A-5 heat transfer rate significantly lowers. This causes the FW temperature to lower.
(2) Where does operation stabilize on the power
-to-flow map?         (1)   (2) Feedwater temperature 100%Rod Line A. Lowers and returns to near the previous temperature.
Below       B. Stabilizes at a lower temperature ONLY.
Below C. Lowers and returns to near the previous temperature.
Above D. Stabilizes at a lower temperature ONLY.
Above   Answer: D. Stabilizes at a lower temperature ONLY.
Above Explanation:
The Extraction Steam system conducts steam from Main Turbine connections to two parallel Feedwater heater strings to improve the overall efficiency of the reactor by preheating the incoming feedwater and reducing the reactor heat load. The NON
-RETURN ISOLATION CHECK VALVES (NRVs) are installed in the extraction steam supply lines to the Feedwater Heaters. NRV
-1 is designed with an integral motor operator and when closed will essentially function as a stop check valve. When NRV
-1 goes closed, feedwater Heater 1
-A-5 heat transfe r rate significantly lowers. This causes the FW temperature to lower.
ESAR Vol. V., Section XIV, Part 5.2.1; Loss of Feedwater Heating A decrease in feedwater temperature due to loss of feedwater heating results in a core power increase. This power rise (at the same reactor recirculation flow) raises the operating point above the 100% rod line. So at a constant flow, power is higher and therefore the operating point is higher on the power to flow map.
ESAR Vol. V., Section XIV, Part 5.2.1; Loss of Feedwater Heating A decrease in feedwater temperature due to loss of feedwater heating results in a core power increase. This power rise (at the same reactor recirculation flow) raises the operating point above the 100% rod line. So at a constant flow, power is higher and therefore the operating point is higher on the power to flow map.
Distractors:
Distractors:
123 A. This answer is incorrect because with the loss of extraction steam, feedwater is not being preheated so feedwater temperature lowers and remains lower. The operating point on the power-to-flow map when compared to the rod line is above the 100% rod line due to the now higher reactor power. A candidate who believes that the malfunctioning NRV causes more steam to be admitted to the heater (as is the case with closure of a heater steam dump valve) or who does not know the plant response on the power
123
-to-flow map when feedwater temperature lowers would choose this answer.
B. This answer is incorrect because the operating point on the power
-to-flow map is higher not lower after conditions stabilize. The candidate who understands the effect of the NRV malfunction but not the final impact on the reactor may choose this answer.
C. This answer is incorrect because with the loss of extraction steam, feedwater is not being preheated so feedwater temperature lowers and remains lower. The candidate who does not understand the extraction steam system and evaluates the effect of the NRV as that of a dump valve would choose this answer.


Technical Reference(s):
A. This answer is incorrect because with the loss of extraction steam, feedwater is not being preheated so feedwater temperature lowers and remains lower. The operating point on the power-to-flow map when compared to the rod line is above the 100% rod line due to the now higher reactor power. A candidate who believes that the malfunctioning NRV causes more steam to be admitted to the heater (as is the case with closure of a heater steam dump valve) or who does not know the plant response on the power -to-flow map when feedwater temperature lowers would choose this answer.
2.4EX-STM, Rev. 18 2.1.10 Station Power Changes, Attachment 1 Power to Flow Map, Rev. 107   Proposed references to be provided to applicants during examination:
B. This answer is incorrect because the operating point on the power-to-flow map is higher not lower after conditions stabilize. The candidate who understands the effect of the NRV malfunction but not the final impact on the reactor may choose this answer.
NONE Learning Objective:
C. This answer is incorrect because with the loss of extraction steam, feedwater is not being preheated so feedwater temperature lowers and remains lower. The candidate who does not understand the extraction steam system and evaluates the effect of the NRV as that of a dump valve would choose this answer.
OPS Extraction Steam and Heater Drains COR001 01   LO-06 Given a specific extraction steam and heater drains malfunction, determine the effect on any of the following:
Technical Reference(s):       2.4EX-STM, Rev. 18 2.1.10 Station Power Changes, Attachment 1 Power to Flow Map, Rev. 107 Proposed references to be provided to applicants during examination:       NONE Learning Objective:   OPS Extraction Steam and Heater Drains COR001-04-01 LO-06 Given a specific extraction steam and heater drains malfunction, determine the effect on any of the following:
: d. Reactor Feedwater
: d.     Reactor Feedwater Question Source:              Bank #
Modified Bank #
New X Question History:            Last NRC Exam        n\a Question Cognitive Level:    Memory or Fundamental Knowledge Comprehension or Analysis                    X 10 CFR Part 55 Content:      55.41 (7)
Comments:
LOD 3 124


Question Source:
NRV that closes 125
Bank #    Modified Bank #
New X  Question History:
Last NRC Exam n\a  Question Cognitive Level:
Memory or Fundamental Knowledge  Comprehension or Analysis  X  10 CFR Part 55 Content:
55.41 (7)  Comments:
LOD 3 124 NRV that closes 125 126 USAR Chapter XIV 127 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
-


==Reference:==
126 USAR Chapter XIV 127


Level  RO  Tier #    2  Group #  2  K/A #  239001 A1.08 Importance Rating
ES-401                            Sample Written Examination                      Form ES-401-5 Question Worksheet Examination Outline Cross-


===3.8 239001===
==Reference:==
Main and Reheat Steam System: Ability to predict and/or monitor changes in parameters associated with operating the MAIN AND REHEAT STEAM SYSTEM controls including: A1.08 Reactor pressure (CFR: 41.5 / 45.5)
Level            RO Tier #            2 Group #          2 K/A #            239001 A1.08 Importance Rating 3.8 239001 Main and Reheat Steam System: Ability to predict and/or monitor changes in parameters associated with operating the MAIN AND REHEAT STEAM SYSTEM controls including: A1.08 Reactor pressure (CFR: 41.5 / 45.5)
Q ues ti on: 6 3  Reactor power is reduced to 65% for performance of Surveillance Procedure 6.MS.201 (Main Steam Isolation Valve Operability Test). The MSIV Test Pushbutton is depressed and sticks in the depressed position.
Question: 63 Reactor power is reduced to 65% for performance of Surveillance Procedure 6.MS.201 (Main Steam Isolation Valve Operability Test).
The MSIV Test Pushbutton is depressed and sticks in the depressed position.
How long does it take the MSIV to fully close from the time the pushbutton is first depressed and what is the final reactor pressure when conditions stabilize?
How long does it take the MSIV to fully close from the time the pushbutton is first depressed and what is the final reactor pressure when conditions stabilize?
Time to Close Final Reactor Pressure A.   < 5 second s    > Initial Pressure B.   < 5 seconds < Initial Pressure
Time to Close               Final Reactor Pressure A.       < 5 seconds                    > Initial Pressure B.       < 5 seconds                   < Initial Pressure C.     > 20 seconds                    > Initial Pressure D.     > 20 seconds                    < Initial Pressure.
 
Answer:
C.   > 2 0 second s  > Initial Pressure D.   > 2 0 second s    < Initial Pressure
C.     > 20 seconds                    > Initial Pressure Explanation: Requires operating a Main Steam Isolation Valve and predicting the reactor pressure change.
. A ns w e r: C.   > 2 0 second s  > Initial Pressure Explanation: Requires operating a Main Steam Isolation Valve and predicting the reactor pressure change. MSIV Spring Only Closure Test vents air off the MSIV operator to allow closure by spring pressure only which takes greater than 20 seconds to reach full closure. With the pushbutton in the depressed position and when conditions stabilize with the MSIV closed, reactor pressure will be higher due to the increased head loss of 3 Main Steam Lines vs. 4 open Main Steam Lines. Distracters:
MSIV Spring Only Closure Test vents air off the MSIV operator to allow closure by spring pressure only which takes greater than 20 seconds to reach full closure. With the pushbutton in the depressed position and when conditions stabilize with the MSIV closed, reactor pressure will be higher due to the increased head loss of 3 Main Steam Lines vs. 4 open Main Steam Lines.
Distracters:
A. This answer is incorrect due to MSIV slow closure being greater than 5 seconds. This choice is plausible if confused with TS maximum MSIV closure time of 5 seconds. The candidate who confuses MSIV slow closure time and correctly identifies reactor pressure response would choose this answer.
A. This answer is incorrect due to MSIV slow closure being greater than 5 seconds. This choice is plausible if confused with TS maximum MSIV closure time of 5 seconds. The candidate who confuses MSIV slow closure time and correctly identifies reactor pressure response would choose this answer.
128 B. This answer is incorrect due to MSIV slow closure being greater than 5 seconds and reactor pressure rising greater than the original pre
128
-closure pressure. This choice is plausible if confused with TS maximum MSIV closure time of 5 and if confused with equalizing header pressure which would be maintained at the original pressure by the DEH system. The candidate who confuses MSIV slow closure time and confuses equalizing header pressure or DEH response would choose this answer.
 
D. This answer is incorrect due to reactor pressure rising greater than the original pre
-closure pressure. This choice is plausible if confused with equalizing header pressure which would be maintained at the original pressure by the DEH system. The candidate who correctly identifies MSIV slow closure time and confuses equalizing header pressure or DEH response would choose this answer.


B. This answer is incorrect due to MSIV slow closure being greater than 5 seconds and reactor pressure rising greater than the original pre-closure pressure. This choice is plausible if confused with TS maximum MSIV closure time of 5 and if confused with equalizing header pressure which would be maintained at the original pressure by the DEH system. The candidate who confuses MSIV slow closure time and confuses equalizing header pressure or DEH response would choose this answer.
D. This answer is incorrect due to reactor pressure rising greater than the original pre-closure pressure. This choice is plausible if confused with equalizing header pressure which would be maintained at the original pressure by the DEH system. The candidate who correctly identifies MSIV slow closure time and confuses equalizing header pressure or DEH response would choose this answer.
Technical Reference(s):
Technical Reference(s):
Procedure 6.MS.201 (Main Steam Isolation Valve Operability Test), Rev. 15 Procedure 2.2.56 (Main Steam System), Rev. 49
Procedure 6.MS.201 (Main Steam Isolation Valve Operability Test), Rev. 15 Procedure 2.2.56 (Main Steam System), Rev. 49 Proposed references to be provided to applicants during examination:      None Learning Objective:    COR0021402001070F - Predict the consequences of the following items on the MAIN STEAM SYSTEM: Closure of one or more MSIV's at power Question Source:              Bank #
Modified Bank #
New X Question History:              Last NRC Exam Question Cognitive Level:      Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content:        55.41 (5)
Comments:
LOD 3 129


Proposed references to be provided to applicants during examination
130 131 ES-401                           Sample Written Examination                   Form ES-401-5 Question Worksheet Examination Outline Cross-
: None Learning Objective:
COR0021402001070F
- Predict the consequences of the following items on the MAIN STEAM SYSTEM:  Closure of one or more MSIV's at power
 
Question Source:
Bank #    Modified Bank #
New X  Question History:
Last NRC Exam Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X    10 CFR Part 55 Content:
55.41 (5)  Comments: LOD 3 129 130 131 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
-


==Reference:==
==Reference:==
Level                RO Tier #                2 Group #              2 K/A #                272000 A2.15 Importance Rating    2.5 272000 Radiation Monitoring - Ability to (a) predict the impacts of the following on the radiation monitoring system; and (b) based on those predictions, use procedures to correct, control or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5) A2.15 Maintenance operations Question: 64 Maintenance is being performed on RMP-RM-150B (OFFGAS RAD MONITOR B) while operating at rated power with the following annunciator in alarm:
OFFGAS RAD                PANEL/WINDOW:
MON DOWNSCALE OR INOP                  9-4-1/D-4 (1) What is the impact on the Off-Gas system if radiation levels rise causing RMP-RM-150A (OFF GAS RAD MONITOR A) to reach the Hi Hi Trip Setpoint?
(2) What action is required following the Off-Gas System isolation IAW Procedure 2.4OG (Off-Gas Abnormal)?
A. (1) Off-Gas isolates IMMEDIATELY.
(2) SCRAM and enter Procedure 2.1.5 (Reactor Scram).
B. (1) Off-Gas isolates IMMEDIATELY.
(2) Lower reactor power per Procedure 2.1.10 (Station Power Changes).
C. (1) Off-Gas isolates after a 15 minute time delay.
(2) SCRAM and enter Procedure 2.1.5 (Reactor Scram).
D. (1) Off-Gas isolates after a 15 minute time delay.
(2) Lower reactor power per Procedure 2.1.10 (Station Power Changes).
Answer:
C. (1) Off-Gas isolates after a 15 minute time delay.
(2) SCRAM and enter Procedure 2.1.5 (Reactor Scram).
Explanation:
Requires knowledge of maintenance activity impact on the OG system and Scram actions of 2.4OG. The off-gas stream is monitored by two radiation monitors: RMP-RM-150A and RMP-RM-150B. Both monitors must be in a tripped condition to start a 15 minute timer which will 132


Level  RO  Tier #    2 Group #  2 K/A #  272000 A2.15 Importance Rating
isolate off-gas after the timer times out. With one channel inoperable for maintenance (mode switch not in operate) half of the trip logic is actuated and (9-4-1/D-4) sounds. When RM-150A goes upscale, ( 9-4-1/C-4) sounds. At this point a 15 minute timer begins counting down. After 15 minutes, an off-gas isolation signal will occur. OG-254 and AOG-902 close after 15 minutes which isolates the off-gas system. IAW 2.4OG - if the OG system isolates due to Hi Radiation, Scram and enter Procedure 2.1.5.
 
===2.5 272000===
Radiation Monitoring
- Ability to (a) predict the impacts of the following on the radiation monitoring system; and (b) based on those predictions, use procedures to correct, control or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5) A2.15 Maintenance operations Question:  64
 
Maintenance is being performed on RMP
-RM-150B (OFFGAS RAD MONITOR B) while operating at rated power with the following annunciator in alarm:
OFFGAS RAD MON DOWNSCA LE OR INOP    PANEL/WINDOW:
9-4-1/D-4  (1)  What is the impact on the Off
-Gas system if radiation levels rise causing RMP
-RM-150A (OFF GAS RAD MONITOR A) to reach the Hi Hi Trip Setpoint?
  (2)  What action is required following the Off
-Gas System isolation IAW Procedure 2.4OG (Off
-Gas Abnormal)?
 
A.  (1) Off-Gas isolates IMMEDIATELY.
(2) SCRAM and enter Procedure 2.1.5 (Reactor Scram).
B.  (1) Off-Gas isolates IMMEDIATELY.
(2) Lower reactor power per Procedure 2.1.10 (Station Power Changes).
C.  (1) Off-Gas isolates after a 15 minute time delay.
(2) SCRAM and enter Procedure 2.1.5 (Reactor Scram).
D.  (1) Off-Gas isolates after a 15 minute time delay.
(2) Lower reactor power per Procedure 2.1.10 (Station Power Changes).
Answer: C.  (1) Off-Gas isolates after a 15 minute time delay.
(2) SCRAM and enter Procedure 2.1.5 (Reactor Scram).
Explanation:
Requires knowledge of maintenance activity impact on the OG system and Scram actions of 2.4OG. The off
-gas stream is monitored by two radiation monitors: RMP
-RM-150A and RMP
-RM-150B. Both monitors must be in a tripped condition to start a 15 minute timer which will 132 isolate off
-gas after the timer times out. With one channel inoperable for maintenance (mode switch not in operate) half of the trip logic is actuated and (9 1/D-4) sounds. When RM
-150A goes upscale, ( 9 1/C-4) sounds. At this point a 15 minute timer begins counting down. After 15 minutes, an off
-gas isolation signal will occur. OG
-254 and AOG
-902 close after 15 minutes which isolates the off
-gas system. IAW 2.4OG  
- if the OG system isolates due to Hi Radiation, Scram and enter Procedure 2.1.5.
Distracters:
Distracters:
A. This answer is incorrect due Off
A. This answer is incorrect due Off-gas does NOT isolate immediately under these conditions.
-gas does NOT isolate immediately under these conditions. This choice is plausible due to not recalling OG having a 15 min time delay  
This choice is plausible due to not recalling OG having a 15 min time delay - if stem were change to reflect RB HVAC RM - immediately isolate would be correct. The candidate that confuses OG isolation time delay and correctly identifies a scram is required following isolation due to valid radiation levels would choose this answer.
- if stem were change to reflect RB HVAC RM  
B. This answer is incorrect due Off-gas does NOT isolate immediately under these conditions and reducing power not required under the given conditions. This choice is plausible due to not recalling OG having a 15 min time delay - if stem were change to reflect RB HVAC RM
- immediately isolate would be correct. The candidate that confuses OG isolation time delay and correctly identifies a scram is required following isolation due to valid radiation levels would choose this answer.
    - immediately isolate would be correct and power reduction being required to maintain main condenser vacuum and to reduce the OG radiation levels prior to OG isolation. The candidate that confuses OG isolation timeframe and does not recognize AOP Scram action would choose this answer.
 
B. This answer is incorrect due Off
-gas does NOT isolate immediately under these conditions and reducing power not required under the given conditions. This choice is plausible due to not recalling OG having a 15 min time delay  
- if stem were change to reflect RB HVAC RM  
- immediately isolate would be correct and power reduction being required to maintain main condenser vacuum and to reduce the OG radiation levels prior to OG isolation. The candidate that confuses OG isolation timeframe and does not recognize AOP Scram action would choose this answer.
D. This answer is incorrect reducing power not required under the given conditions. This choice is plausible due to power reduction being required to maintain main condenser vacuum and to reduce the OG radiation levels prior to OG isolation. The candidate that correctly identifies OG isolation time delay and does not recognize AOP Scram action would choose this answer.
D. This answer is incorrect reducing power not required under the given conditions. This choice is plausible due to power reduction being required to maintain main condenser vacuum and to reduce the OG radiation levels prior to OG isolation. The candidate that correctly identifies OG isolation time delay and does not recognize AOP Scram action would choose this answer.
Technical Reference(s):
Technical Reference(s):
Procedure 2.4OG (Off
Procedure 2.4OG (Off-Gas Abnormal), Rev. 22 Proposed references to be provided to applicants during examination:         None Learning Objective: COR001-18-02
-Gas Abnormal), Rev. 22 Proposed references to be provided to applicants during examination:
: 7.     Given a specific Radiation Monitoring system malfunction, determine the effect on any of the following:
None Learning Objective:
COR001-18-02   7. Given a specific Radiation Monitoring system malfunction, determine the effect on any of the following:
: b. Station Area Radiation monitoring
: b. Station Area Radiation monitoring
: 11. Predict the consequences of the following items on the Radiation Monitoring system. g. Maintenance/Surveillance operations Question Source:
: 11. Predict the consequences of the following items on the Radiation Monitoring system.
Bank #   Modified Bank #
: g. Maintenance/Surveillance operations Question Source:               Bank #
New   X Question History:
Modified Bank #
Last NRC Exam 133 Question Cognitive Level:
New                     X Question History:               Last NRC Exam 133
Memory or Fundamental Knowledge Comprehension or Analysis  X  10 CFR Part 55 Content:
55.41 (5)  Comments: LOD 3    OFF GAS RAD MON DOWNSCALE OR INOP, 9-4-1/D-4 RMP-RM-150A(B), 1 mR/hr or monitor inop. (Mode switch not in operate or high voltage level low)
Upon a downscale signal from both monitors an Off Gas system isolation will occur after a 15
-min. time delay. Upon an inoperative signal from both monitors, an Off Gas system isolation w ill occur after a 15
-min. time delay. ALARM PROCEDURE 2.3_9 1 PANEL 9 ANNUNCIATOR 9 1 D-4 OFFGAS RAD MON DOWNSCALE OR INOP (Page 134)  OFFGAS RAD MON DOWNSCALE OR INOP    PANEL/WINDOW:
9-4-1/D-4  AUTOMATIC ACTIONS Off-gas timer initiates upon a simultaneous trip signal (High
-High, Inoperable, or Downscale) from both off-gas monitors.
134 OFFGAS TIMER INITIATED    PANEL/WINDOW:
9-4-1/C-4 SETPOINT (1758) OFFGAS TIMER INITIATED on any simultaneous combination of an A and B trip due to: 1. Channel inoperable
: 2. Downscale at 1.05 mR/hr
: 3. Hi-Hi trip at 1.58E3 for Channel A or Channel B (Tech Spec  1 Ci/sec) CIC 9-4-1/C-4 RMP-RM-150A and RMP
-RM-150B  AUTOMATIC ACTIONS After 14 minutes (ODAM  15 minutes) of continuous alarm condition, following valves close:  OG-AO-254, OFF/GAS SYSTEM ISOLATION.
OPERATOR OBSERVATION AND ACTION Reduce power, as necessary, per Procedure 2.1.10 to clear alarm.
Enter Procedures 2.4OG and 5.2FUEL.
135 136 137 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
-


==Reference:==
Question Cognitive Level:     Memory or Fundamental Knowledge Comprehension or Analysis                  X 10 CFR Part 55 Content:      55.41 (5)
Comments:
LOD 3 OFF GAS RAD MON                      RMP-RM-150A(B),            Upon a downscale signal DOWNSCALE OR INOP,                  1 mR/hr or monitor        from both monitors an Off 9-4-1/D-4                            inop. (Mode switch        Gas system isolation will not in operate or high    occur after a 15-min. time voltage level low)        delay.
Upon an inoperative signal from both monitors, an Off Gas system isolation will occur after a 15-min. time delay.
ALARM PROCEDURE 2.3_9-4-1 PANEL 9 ANNUNCIATOR 9-4-1 D-4    OFFGAS RAD MON DOWNSCALE OR INOP (Page 134)
OFFGAS RAD PANEL/WINDOW:
MON DOWNSCALE 9-4-1/D-4 OR INOP AUTOMATIC ACTIONS Off-gas timer initiates upon a simultaneous trip signal (High-High, Inoperable, or Downscale) from both off-gas monitors.
134


Level  RO  Tier #    2   Group # 2 K/A #  288000 A3.01 Importance Rating
OFFGAS TIMER            PANEL/WINDOW:
INITIATED                9-4-1/C-4 SETPOINT                                          CIC                                    9-4-1/C-4 (1758) OFFGAS TIMER INITIATED on any              RMP-RM-150A and RMP-RM-150B simultaneous combination of an A and B trip due to:
: 1. Channel inoperable
: 2. Downscale at 1.05 mR/hr
: 3. Hi-Hi trip at 1.58E3 for Channel A or Channel B (Tech Spec 1 Ci/sec)
AUTOMATIC ACTIONS After 14 minutes (ODAM 15 minutes) of continuous alarm condition, following valves close:
* OG-AO-254, OFF/GAS SYSTEM ISOLATION.
OPERATOR OBSERVATION AND ACTION
* Reduce power, as necessary, per Procedure 2.1.10 to clear alarm.
* Enter Procedures 2.4OG and 5.2FUEL.
135


===3.8 288000===
136 137 ES-401                              Sample Written Examination                      Form ES-401-5 Question Worksheet Examination Outline Cross-
Plant Ventilation
- Ability to monitor automatic operations of the plant ventilation systems including: (CFR: 41.7 / 45.7)
- A3.01 Isolation/initiation signals Question: 65


What automatically starts the Essential Control Building Ventilation system?
==Reference:==
Level                RO Tier #                2 Group #              2 K/A #                288000 A3.01 Importance Rating    3.8 288000 Plant Ventilation - Ability to monitor automatic operations of the plant ventilation systems including: (CFR: 41.7 / 45.7) - A3.01 Isolation/initiation signals Question: 65 What automatically starts the Essential Control Building Ventilation system?
A. Battery room exhaust fan trip.
A. Battery room exhaust fan trip.
B. Low Critical Switchgear room temperature.
B. Low Critical Switchgear room temperature.
C. High Critical Switchgear room temperature.
C. High Critical Switchgear room temperature.
D. Non-essential Control Building supply fan trip.
D. Non-essential Control Building supply fan trip.
Answer: C. High Critical Switchgear room temperature.
Answer:
C. High Critical Switchgear room temperature.
Explanation:
Explanation:
The Essential Control Building Ventilation System is located in the Critical AC Switchgear Rooms and is comprised of two 100% capacity redundant supply and exhaust fans. These fans are powered from Critical Division 1 and Division 2 power supplies. When EF
The Essential Control Building Ventilation System is located in the Critical AC Switchgear Rooms and is comprised of two 100% capacity redundant supply and exhaust fans. These fans are powered from Critical Division 1 and Division 2 power supplies. When EF-SWGR-1F or EF-SWGR-1G, EXHAUST FAN, is running, the isolation dampers to the RHR SWBP Room and from the Non-Essential Control Building HVAC System close. Either Essential Control Building Ventilation Subsystem can remove any potential hydrogen buildup in the Battery Rooms.
-SWGR-1F or EF-SWGR-1G, EXHAUST FAN, is running, the isolation dampers to the RHR SWBP Room and from the Non
The Essential Control Building Ventilation system functions to ensure the 4160V Critical Switchgear Rooms do not overheat. Essential Control Building HV controls Battery Room temperatures between 50&deg;F and 120&deg;F. On a critical switchgear room high temperature (110&deg;F for room 1F and 115&deg;F for room 1G) the respective system automatically starts.
-Essential Control Building HVAC System close. Either Essential Control Building Ventilation Subsystem can remove any potential hydrogen buildup in the Battery Rooms.
The Essential Control Building Ventilation system functions to ensure the 4160V Critical Switchgear Rooms do not overheat. Essential Control Building HV controls Battery Room temperatures betw een 50F and 120F. On a critical switchgear room high temperature (110&deg;F for room 1F and 115F for room 1G) the respective system automatically starts.
Distracters:
Distracters:
A. This answer is incorrect because there is no interlock between the battery room exhaust fan operation and the essential control building ventilation system. One design criteria of the essential control building ventilation system is to remove any potential hydrogen buildup in the Battery Rooms should a Battery Room exhaust fan not be in operation. Because the essential control building ventilation system services the Battery Rooms, this answer is plausible.
A. This answer is incorrect because there is no interlock between the battery room exhaust fan operation and the essential control building ventilation system. One design criteria of the essential control building ventilation system is to remove any potential hydrogen buildup in the Battery Rooms should a Battery Room exhaust fan not be in operation. Because the essential control building ventilation system services the Battery Rooms, this answer is plausible.
B. This answer is incorrect because a low temperature of 50F causes the essential control building ventilation system to trip not start. This answer is plausible because the essential control building ventilation system does have an interlock with the critical switchgear room temperature. If the interlock is not correctly determined, this answer may be chosen by a candidate.
B. This answer is incorrect because a low temperature of 50&deg;F causes the essential control building ventilation system to trip not start. This answer is plausible because the essential control building ventilation system does have an interlock with the critical switchgear room temperature. If the interlock is not correctly determined, this answer may be chosen by a candidate.
138 D. This answer is incorrect because the trip of the non
138
-essential fan does not have an interlock with the essential control building ventilation system. However, this answer is plausible because the initiation of the essential control building ventilation system will cause the non
-essential control building fans to trip. The candidate who knows one of the system causes the other to trip would select this answer. The normal convention of HVAC fans is for one to be in standby and automatically start if the running fan trips. Because this area is served by both systems, the interlock may be blurred.


Technical Reference(s):
D. This answer is incorrect because the trip of the non-essential fan does not have an interlock with the essential control building ventilation system. However, this answer is plausible because the initiation of the essential control building ventilation system will cause the non-essential control building fans to trip. The candidate who knows one of the system causes the other to trip would select this answer. The normal convention of HVAC fans is for one to be in standby and automatically start if the running fan trips. Because this area is served by both systems, the interlock may be blurred.
SOP 2.2.38, HVAC CONTROL BUILDING, Rev. 40 Proposed references to be provided to applicants during examination:
Technical Reference(s):       SOP 2.2.38, HVAC CONTROL BUILDING, Rev. 40 Proposed references to be provided to applicants during examination:         NONE Learning Objective:   OPS Heating, Ventilation and Air Conditioning COR001-08-01 R28
NONE Learning Objective:
: 11. Describe the HVAC design features and interlocks that provide for the following:
OPS Heating, Ventilation and Air Conditioning COR001 01 R28   11. Describe the HVAC design features and interlocks that provide for the following:
: c. Automatic starting and stopping of fans Question Source:              Bank #
: c. Automatic starting and stopping of fans
Modified Bank #
New X Question History:              Last NRC Exam          n\a Question Cognitive Level:      Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content:        55.41 (7)
Comments:
LOD 3 139


Question Source:
140 141 ES-401                             Sample Written Examination                     Form ES-401-5 Question Worksheet Examination Outline Cross-
Bank #    Modified Bank
#  New X  Question History:
Last NRC Exam n\a  Question Cognitive Level:
Memory or Fundamental Knowledge X  Comprehension or Analysis 10 CFR Part 55 Content:
55.41 (7)
Comments: LOD 3    139 140 141 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
-


==Reference:==
==Reference:==
 
Level               RO Tier #               3 Group #
Level   RO   Tier #     3   Group #   K/A #   2.1.8 Importance Rating 3.4   2.1.8 Ability to coordinate personnel activities outside the control room. (CFR: 41.10 / 45.5 / 45.12 / 45.13)
K/A #               2.1.8 Importance Rating   3.4 2.1.8 Ability to coordinate personnel activities outside the control room. (CFR: 41.10 /
 
45.5 / 45.12 / 45.13)
Q ues ti on: 6 6  With the plant at power, a manual valve is closed in order to maintain Primary Containment OPERABLE IAW TS LCO 3.6.1.3 (PCIVs).
Question: 66 With the plant at power, a manual valve is closed in order to maintain Primary Containment OPERABLE IAW TS LCO 3.6.1.3 (PCIVs).
The valve must be opened for maintenance purposes. The NLO opens the valve and remains at the valve.
The valve must be opened for maintenance purposes. The NLO opens the valve and remains at the valve.
Which of the following identifies the MINIMUM additional actions required IAW Procedure 2.0.1 (Plant Operations Policy)?
Which of the following identifies the MINIMUM additional actions required IAW Procedure 2.0.1 (Plant Operations Policy)?
Line 2,902: Line 2,698:
B. Instruct the NLO to close the valve in event of an accident condition ONLY.
B. Instruct the NLO to close the valve in event of an accident condition ONLY.
C. Establish continuous communication with the Control Room and instruct the NLO to close the valve in event of an accident condition ONLY.
C. Establish continuous communication with the Control Room and instruct the NLO to close the valve in event of an accident condition ONLY.
D. Establish continuous communication with the Control Room, instruct the NLO to close the valve in event of an accident condition and document the instructions in the Control Room log. Answer: D. Establish continuous communication with the Control Room, instruct the NLO to close the valve in event of an accident condition and document the instructions in the Control Room log. Explanation:
D. Establish continuous communication with the Control Room, instruct the NLO to close the valve in event of an accident condition and document the instructions in the Control Room log.
Answer:
D. Establish continuous communication with the Control Room, instruct the NLO to close the valve in event of an accident condition and document the instructions in the Control Room log.
Explanation:
This question requires the Reactor Operator to coordinate personnel activities outside the control room (Plant Operator local valve operations under TS Administrative Controls).
This question requires the Reactor Operator to coordinate personnel activities outside the control room (Plant Operator local valve operations under TS Administrative Controls).
Procedure 2.0.1 (Plant Operations Policy) provides the following guidance:
Procedure 2.0.1 (Plant Operations Policy) provides the following guidance:
Isolation valves closed to satisfy LCO 3.6.1.3 may be re
Isolation valves closed to satisfy LCO 3.6.1.3 may be re-opened on an intermittent basis following administrative controls:
-opened on an intermittent basis following administrative controls:
* A person shall be stationed at valve controls while valve is open.
A person shall be stationed at valve controls while valve is open.
* If valve is being controlled outside of Control Room, person at valve controls shall be in continuous communication with Control Room.
If valve is being controlled outside of Control Room, person at valve controls shall be in continuous communication with Control Room.
* Person at valve controls shall be instructed to close valve in event of an accident condition. These instructions shall be documented (the Control Room log satisfies this requirement).
Person at valve controls shall be instructed to close valve in event of an accident condition. These instructions shall be documented (the Control Room log satisfies this requirement).
142
142 Distracters:
 
Distracters:
A. This answer is incorrect; specifically continuous communication and direction for valve operation during accident conditions is not included. There is an allowance for opening the containment isolation valve on an intermittent basis under administrative controls. This choice is plausible if the candidate does not remember all of the specific requirements for opening a PCIV under administrative controls.
A. This answer is incorrect; specifically continuous communication and direction for valve operation during accident conditions is not included. There is an allowance for opening the containment isolation valve on an intermittent basis under administrative controls. This choice is plausible if the candidate does not remember all of the specific requirements for opening a PCIV under administrative controls.
B. This answer is incorrect due to not providing ALL the procedural requirements; specifically direction for valve operation during emergency conditions is not included in the selection.
B. This answer is incorrect due to not providing ALL the procedural requirements; specifically direction for valve operation during emergency conditions is not included in the selection.
This choice is plausible if the candidate does not remember all of the requirements for opening a PCIV under administrative controls.
This choice is plausible if the candidate does not remember all of the requirements for opening a PCIV under administrative controls.
C. This answer is incorrect due to not providing ALL the procedural requirements; specifically the requirement to document the instructions in the Control Room log. This choice is plausible if all the requirements for opening a PCIV under administrative controls are not specifically known. The candidate does not recall the requirement to document valve closure directions in the CR log would select this answer.
C. This answer is incorrect due to not providing ALL the procedural requirements; specifically the requirement to document the instructions in the Control Room log. This choice is plausible if all the requirements for opening a PCIV under administrative controls are not specifically known. The candidate does not recall the requirement to document valve closure directions in the CR log would select this answer.
Technical Reference(s):
Technical Reference(s):
Procedure 2.0.1 (Plant Operations Policy), Rev. 62 Proposed references to be provided to applicants during examination:
Procedure 2.0.1 (Plant Operations Policy), Rev. 62 Proposed references to be provided to applicants during examination:       None Learning Objective:
None Learning Objective:
SKL0080102 Ops Watchstanding Principles for Licensed Operators 0010400 Briefly describe the administrative controls for primary or secondary containment manual valve opening or associated cap removal when primary or secondary containment is required.
SKL0080102 Ops Watchstanding Principles for Licensed Operators 0010400 Briefly describe the administrative controls for primary or secondary containment manual valve opening or associated cap removal when primary or secondary containment is required.
Question Source:                Bank #
Modified Bank #
New X Question History:              Last NRC Exam        n\a Question Cognitive Level:      Memory or Fundamental Knowledge            X Comprehension or Analysis 10 CFR Part 55 Content:        55.41 (10)
Comments:
LOD 2 143


Question Source:
144 ES-401                               Sample Written Examination                   Form ES-401-5 Question Worksheet Examination Outline Cross-
Bank #    Modified Bank #
New X  Question History:
Last NRC Exam n\a  Question Cognitive Level:
Memory or Fundamental Knowledge X Comprehension or Analysis    10 CFR Part 55 Content:
55.41 (10)
Comments:
LOD 2 143 144 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet   Examination Outline Cross
-


==Reference:==
==Reference:==
 
Level                 RO Tier #                 3 Group #
Level   RO   Tier #     3   Group #   K/A #   2.1.18 Importance Rating 3.6   2.1.18 Ability to make accurate, clear, and concise logs, records, status boards, and reports. (CFR: 41.10 / 45.12 / 45.13)
K/A #                 2.1.18 Importance Rating     3.6 2.1.18 Ability to make accurate, clear, and concise logs, records, status boards, and reports.
Q ues ti on: 6 7  Th e R eac t o r ha s j us t bee n dec l a r e d CR I T I CA L du ri n g p l an t s t a rtup. I n additi o n t o: D a t e/T i m e  C on tr o l R o d N u m be r  C on tr o l R o d P ositi o n  Sequenc e  R eac t o r Pressure W ha t additi ona l criti calit y i n f o r m a ti o n i s requ ir e d t o b e logge d i n t h e C on tr o l R oo m O pe r a t o r s Lo g I A W P rocedu r e 2.1.1 (S t a rt u p P rocedu r e)?
(CFR: 41.10 / 45.12 / 45.13)
A. SR M C oun t s AN D I R M O ve rl ap. B. SR M C oun t s AN D M ode r a t o r Te m pe r a t u r e. C. R eac t o r P e ri o d AN D I R M O ve rl ap. D. R eac t o r P e ri o d AN D M ode r a t o r Te m pe r a t u r e. A ns w e r: D. R eac t o r P e ri o d AN D M ode r a t o r T e m pe r a t u r e. Explanation:
Question: 67 The Reactor has just been declared CRITICAL during plant startup. In addition to:
This answer requires knowledge/ability of required accurate criticality log entries. Procedure  
* Date/Time
 
* Control Rod Number
====2.1.1 requires====
* Control Rod Position
the following narrative log criticality data  
* Sequence
- Log control rod number, control rod position, moderator temperature, reactor pressure, sequence, reactor period, and time on Procedure 10.13, Attachment 1, and in Control Room Operator's log.
* Reactor Pressure What additional criticality information is required to be logged in the Control Room Operators Log IAW Procedure 2.1.1 (Startup Procedure)?
 
A. SRM Counts AND IRM Overlap.
B. SRM Counts AND Moderator Temperature.
C. Reactor Period AND IRM Overlap.
D. Reactor Period AND Moderator Temperature.
Answer:
D. Reactor Period AND Moderator Temperature.
Explanation:
This answer requires knowledge/ability of required accurate criticality log entries. Procedure 2.1.1 requires the following narrative log criticality data - Log control rod number, control rod position, moderator temperature, reactor pressure, sequence, reactor period, and time on Procedure 10.13, Attachment 1, and in Control Room Operator's log.
Distracters:
Distracters:
A. This answer is incorrect due to SRM Counts and IRM overlap not required to be logged in the narrative log for criticality data. This choice is plausible due to SRM & IRM chart recorders are required to be annotated and IRM overlap required to be document in Procedure 2.1.1. The candidate who confuses annotating charts and documenting IRM overlap would choose this answer.
A. This answer is incorrect due to SRM Counts and IRM overlap not required to be logged in the narrative log for criticality data. This choice is plausible due to SRM & IRM chart recorders are required to be annotated and IRM overlap required to be document in Procedure 2.1.1. The candidate who confuses annotating charts and documenting IRM overlap would choose this answer.
B. This answer is incorrect due to SRM Counts not required to be logged in the narrative log for criticality data. This choice is plausible due to SRM & IRM chart recorders are required to 145 be annotated IAW Procedure 2.1.1.
B. This answer is incorrect due to SRM Counts not required to be logged in the narrative log for criticality data. This choice is plausible due to SRM & IRM chart recorders are required to 145
The candidate who confuses annotating charts and correctly identifies Moderator Temperature would choose this answer.
 
be annotated IAW Procedure 2.1.1. The candidate who confuses annotating charts and correctly identifies Moderator Temperature would choose this answer.
C. This answer is incorrect because IRM overlap is not required to be logged in the narrative log for criticality data. This choice is plausible due to IRM overlap is required to be documented in Procedure 2.1.1 but not at the time of criticality. The candidate who correctly identifies Period and confuses documenting IRM overlap would choose this answer.
C. This answer is incorrect because IRM overlap is not required to be logged in the narrative log for criticality data. This choice is plausible due to IRM overlap is required to be documented in Procedure 2.1.1 but not at the time of criticality. The candidate who correctly identifies Period and confuses documenting IRM overlap would choose this answer.
Technical Reference(s): GOP 2.1.1, Startup Procedure, Rev. 178 Proposed references to be provided to applicants during examination:
Technical Reference(s): GOP 2.1.1, Startup Procedure, Rev. 178 Proposed references to be provided to applicants during examination:         NONE Learning Objective:
NONE Learning Objective:
INT0320104 CNS Administrative Procedures General Operating Procedures (Startup and Shutdown) Procedures 00A0800         Describe the required actions to be completed upon achieving criticality as described in Procedure 2.1.1, Startup Procedure.
INT0320104 CNS Administrative Procedures General Operating Procedures (Startup and Shutdown) Procedures 00A0800 Describe the required actions to be completed upon achieving criticality as described in Procedure 2.1.1, Startup Procedure.
Question Source:               Bank #
Question Source:
Modified Bank #2377 New Question History:               Last NRC Exam         n\a Question Cognitive Level:       Memory or Fundamental Knowledge               X Comprehension or Analysis 10 CFR Part 55 Content:         55.41 (10)
Bank #   Modified Bank #
Comments:
2377 New   Question History:
LOD 3 146
Last NRC Exam n\a Question Cognitive Level:
 
Memory or Fundamental Knowledge X Comprehension or Analysis     10 CFR Part 55 Content:
147 148 149 150 ES-401                             Sample Written Examination                       Form ES-401-5 Question Worksheet Examination Outline Cross-
55.41 (10) Comments:
LOD 3 146 147 148 149 150 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
-


==Reference:==
==Reference:==
 
Level                 RO Tier #                 3 Group #
Level   RO   Tier #     3   Group #   K/A #   2.1.31 Importance Rating 4.6   2.1.31 Ability to locate control room switches, controls, and indications, and to determine that they correctly reflect the desired plant lineup. (CFR: 41.10 / 45.12)
K/A #                 2.1.31 Importance Rating     4.6 2.1.31 Ability to locate control room switches, controls, and indications, and to determine that they correctly reflect the desired plant lineup. (CFR: 41.10 / 45.12)
Q ues ti on: 68   Which of the following identifies the indication and location that confirms the reactor is in Hot Shutdown when the Mode Switch is in the Shutdown position during a reactor cooldown?
Question: 68 Which of the following identifies the indication and location that confirms the reactor is in Hot Shutdown when the Mode Switch is in the Shutdown position during a reactor cooldown?
A. Feedwater Temperature indicates 200&deg;F on Panel A
A. Feedwater Temperature indicates 200&deg;F on Panel A.
. B. RR Suction Temperature indicates 240&deg;F on Panel 9
B. RR Suction Temperature indicates 240&deg;F on Panel 9-4.
-4. C. RHR HX Inlet Temperature indicates 200&deg;F on Panel 9
C. RHR HX Inlet Temperature indicates 200&deg;F on Panel 9-3.
-3. D. Vessel Head Adjacent to Flange temperature indicates 240&deg;F on Panel 9
D. Vessel Head Adjacent to Flange temperature indicates 240&deg;F on Panel 9-21.
-21. A ns w e r: B. RR Suction Temperature indicates 240&deg;F on Panel 9
Answer:
-4. Explanation:
B. RR Suction Temperature indicates 240&deg;F on Panel 9-4.
Requires knowledge of indication location and indication which supports the plant being in Mode 3, IAW TS Definitions  
Explanation:
- Hot Shutdown (Mode 3) is determined by Mode Switch position (Shutdown) and Average Reactor Coolant Temperature (>212&deg;F) with all reactor vessel head closure bolts fully tensioned. The RR suction temperature is an accurate measure of average reactor coolant temperature and one of the temperatures monitored on reactor cooldown. 200&deg;F is a plausible distractor due to confusing the FPC Time to 200&deg;F (was previously Time to Boil) curve.
Requires knowledge of indication location and indication which supports the plant being in Mode 3, IAW TS Definitions - Hot Shutdown (Mode 3) is determined by Mode Switch position (Shutdown) and Average Reactor Coolant Temperature (>212&deg;F) with all reactor vessel head closure bolts fully tensioned. The RR suction temperature is an accurate measure of average reactor coolant temperature and one of the temperatures monitored on reactor cooldown.
200&deg;F is a plausible distractor due to confusing the FPC Time to 200&deg;F (was previously Time to Boil) curve.
Distracters:
Distracters:
A. This answer is incorrect because FW Temperature indicating 200&deg;F is the temperature of the water entering the reactor and not average reactor coolant. This choice is plausible due to Panel A being the location for FW Temperature (correct location) and confusing FW Temperature as a valid indication of Average Reactor Coolant Temperature and temperature required for HOT shutdown confused with COLD shutdown of  
A. This answer is incorrect because FW Temperature indicating 200&deg;F is the temperature of the water entering the reactor and not average reactor coolant. This choice is plausible due to Panel A being the location for FW Temperature (correct location) and confusing FW Temperature as a valid indication of Average Reactor Coolant Temperature and temperature required for HOT shutdown confused with COLD shutdown of < 212&deg;F. The candidate who correctly identifies FW Temperature indication is located on Panel A and confuses Cold Shutdown with Hot Shutdown temperature requirements would choose this answer.
< 212&deg;F. The candidate who correctly identifies FW Temperature indication is located on Panel A and confuses Cold Shutdown with Hot Shutdown temperature requirements would choose this answer. C. This answer is incorrect because RHR Hx inlet Temperature indicating 200&deg;F is indication of cold shutdown conditions. This choice is plausible due to Hx inlet Temperature being located on Panel 9
C. This answer is incorrect because RHR Hx inlet Temperature indicating 200&deg;F is indication of cold shutdown conditions. This choice is plausible due to Hx inlet Temperature being located on Panel 9-3 (correct temperature) and RHR Hx inlet temperature being reactor coolant temperature if indicating 200&deg;F. The candidate who correctly identifies RHR Hx inlet 151
-3 (correct temperature) and RHR Hx inlet temperature being reactor coolant temperature if indicating 200&deg;F. The candidate who correctly identifies RHR Hx inlet 151 temperature but confuses Cold Shutdown with Hot Shutdown temperature requirements would choose this answer.


temperature but confuses Cold Shutdown with Hot Shutdown temperature requirements would choose this answer.
D. This answer is incorrect because RPV metal temperatures are not specified on T.S. Table 1.1-1 when determining Reactor Modes. This answer is plausible because of the heat transport mechanism and temperature gradients involved across the RPV shell wall. The candidate who incorrectly believes that RPV metal temperatures instead of average reactor coolant temperature would provide indication of Reactor Mode conditions would choose this answer.
D. This answer is incorrect because RPV metal temperatures are not specified on T.S. Table 1.1-1 when determining Reactor Modes. This answer is plausible because of the heat transport mechanism and temperature gradients involved across the RPV shell wall. The candidate who incorrectly believes that RPV metal temperatures instead of average reactor coolant temperature would provide indication of Reactor Mode conditions would choose this answer.
Technical Reference(s):
Technical Reference(s):     TS Table 1.1-1 Proposed references to be provided to applicants during examination:   None Learning Objective: INT00705010010400 From memory, given a set of plant conditions, determine the plant MODE.
TS Table 1.1
Question Source:             Bank #
-1 Proposed references to be provided to applicants during examination:
Modified Bank #
None Learning Objective:   INT00705010010400 From memory, given a set of plant conditions, determine the plant MODE.
New X Question History:           Last NRC Exam         n\a Question Cognitive Level:   Memory or Fundamental Knowledge           X Comprehension or Analysis 10 CFR Part 55 Content:     55.41 (10)
Question Source:
Comments:
Bank # Modified Bank #
LOD 3 152
New X   Question History:
 
Last NRC Exam n\a Question Cognitive Level:
153 ES-401                             Sample Written Examination                       Form ES-401-5 Question Worksheet Examination Outline Cross-
Memory or Fundamental Knowledge X Comprehension or Analysis     10 CFR Part 55 Content:
55.41 (10) Comments: LOD 3   152 153 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
-


==Reference:==
==Reference:==
 
Level                 RO Tier #                 3 Group #
Level   RO   Tier #     3   Group #   K/A #   2.2.13 Importance Rating 4.1   2.2.13 Knowledge of tagging and clearance procedures. (CFR: 41.10 / 45.13)
K/A #                 2.2.13 Importance Rating     4.1 2.2.13 Knowledge of tagging and clearance procedures. (CFR: 41.10 / 45.13)
Question: 69 Which of the following identifies the FIRST two tagging order steps for placing a system pump under clearance, and the reason for this order IAW Procedure 0.9 (Tagout)? OPEN the pump breaker and then close the-A. suction valve to minimize draining time.
Question: 69 Which of the following identifies the FIRST two tagging order steps for placing a system pump under clearance, and the reason for this order IAW Procedure 0.9 (Tagout)?
B. discharge valve to minimize draining time
OPEN the pump breaker and then close the A. suction valve to minimize draining time.
.
B. discharge valve to minimize draining time.
C. suction valve to protect low pressure components.
C. suction valve to protect low pressure components.
D. discharge valve to protect low pressure components.
D. discharge valve to protect low pressure components.
Answer: D. discharge valve to protect low pressure components.
Answer:
D. discharge valve to protect low pressure components.
Explanation:
Explanation:
Requires knowledge of tagging procedure 0.9. Procedure 0.9 provides guidance for pump tagging to remove the power source first. If isolating the pump, the discharge valve is closed before the suction valve to prevent possible over
Requires knowledge of tagging procedure 0.9. Procedure 0.9 provides guidance for pump tagging to remove the power source first. If isolating the pump, the discharge valve is closed before the suction valve to prevent possible over-pressurization of low pressure components on the suction side.
-pressurization of low pressure components on the suction side.
 
Distracters:
Distracters:
A. This answer is incorrect due to the suction valve being closed prior closing the discharge valve and minimizing draining time. This choice is plausible due to confusing the reason for closing the discharge valve prior to the suction and minimizing draining time is desired but not the reason. The candidate who confuses valve closure order and reason for the order would choose this answer.
A. This answer is incorrect due to the suction valve being closed prior closing the discharge valve and minimizing draining time. This choice is plausible due to confusing the reason for closing the discharge valve prior to the suction and minimizing draining time is desired but not the reason. The candidate who confuses valve closure order and reason for the order would choose this answer.
B. This answer is incorrect because minimizing draining time is not the reason for the order of operations. This choice is plausible due to confusing the reason for closing the discharge valve prior to the suction and minimizing draining time is desired but not the reason. The candidate who correctly identifies the valve closure order and confuses the reason for the order would choose this answer.
B. This answer is incorrect because minimizing draining time is not the reason for the order of operations. This choice is plausible due to confusing the reason for closing the discharge valve prior to the suction and minimizing draining time is desired but not the reason. The candidate who correctly identifies the valve closure order and confuses the reason for the order would choose this answer.
C. This answer is incorrect due to the suction valve being closed prior closing the discharge valve. This choice is plausible due to confusing the reason for closing the discharge valve prior to the suction. The candidate who confuses valve closure order and correctly identifies the reason for the order would choose this answer.
C. This answer is incorrect due to the suction valve being closed prior closing the discharge valve. This choice is plausible due to confusing the reason for closing the discharge valve prior to the suction. The candidate who confuses valve closure order and correctly identifies the reason for the order would choose this answer.
154 Technical Reference(s):
154
Procedure 0.9, Tagout, Rev. 85 Proposed references to be provided to applicants during examination:
* Technical Reference(s):     Procedure 0.9, Tagout, Rev. 85 Proposed references to be provided to applicants during examination:     None Learning Objective: SKL00803020010600 Describe the proper sequence for hanging and picking up tags with regards to Tagging Orders.
None Learning Objective: SKL00803020010600 Describe the proper sequence for hanging and picking up tags with regards to Tagging Orders.
Question Source:           Bank #
Question Source:
Modified Bank #
Bank # Modified Bank #
New X Question History:           Last NRC Exam Question Cognitive Level:   Memory or Fundamental Knowledge             X Comprehension or Analysis 10 CFR Part 55 Content:     55.41 (10)
New X   Question History:
Comments:
Last NRC Exam Question Cognitive Level:
LOD 3 155
Memory or Fundamental Knowledge X Comprehension or Analysis     10 CFR Part 55 Content:
 
55.41 (10)
ES-401                           Sample Written Examination                       Form ES-401-5 Question Worksheet Examination Outline Cross-
Comments: LOD 3       155 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
-


==Reference:==
==Reference:==
 
Level                   RO Tier #                 3 Group #
Level   RO   Tier #     3   Group #   K/A #   2.2.41 Importance Rating 3.5   2.2.41 Ability to obtain and interpret station electrical and mechanical drawings.
K/A #                   2.2.41 Importance Rating       3.5 2.2.41 Ability to obtain and interpret station electrical and mechanical drawings.
(CFR: 41.10 / 45.12 / 45.13)
(CFR: 41.10 / 45.12 / 45.13)
Question: 70 An Operator is preparing to write a Clearance Order on a Solenoid Valve.
Question: 70 An Operator is preparing to write a Clearance Order on a Solenoid Valve.
  (1) Where can the Operator obtain Controlled Copies of the electrical and mechanical drawings needed to prepare the Clearance
(1) Where can the Operator obtain Controlled Copies of the electrical and mechanical drawings needed to prepare the Clearance?
?  (2) When interpreting the electrical drawing
(2) When interpreting the electrical drawing, what is the status of Contact 3-4 when the switch is in the CLOSE position?
, what is the status of Contact 3
(1)                                 (2)
-4 when the switch is in the CLOSE position?  
A.         Control Room                             Open B.         Operations Support Center                 Open C.         Control Room                             Closed D.         Operations Support Center                 Closed Answer:
        (1)     (2)
A.         Control Room                             Open Explanation:
A. Control Room Open B. Operations Support Center Open C. Control Room Closed D. Operations Support Center Closed Answer: A. Control Room Open Explanation:
GENERATING A TAGOUT SECTION Controlled Distribution drawings are located in the areas identified below:
GENERATING A TAGOUT SECTION Controlled Distribution drawings are located in the areas identified below:
156  
156
: a. Aperture cards located in Technical Support Center (TSC), Information Resource Center (IRC), AND Central Alarm Station (CAS).
: a. Aperture cards located in Technical Support Center (TSC), Information Resource Center (IRC), AND Central Alarm Station (CAS).
: b. Full size drawings with copies located in the TSC, EOF, Control Room, Simulator Control Room, I&C Shop, E
: b. Full size drawings with copies located in the TSC, EOF, Control Room, Simulator Control Room, I&C Shop, E-Shop, Work Control Center (WCC), and Planning.
-Shop, Work Control Center (WCC), and Planning.
The embedded switch development depicts a switch with two positions: Close (1) and Open (2). An x in the column is used to determine the status of the contacts (1-2, 3-4) associated with the switch.
The embedded switch development depicts a switch with two positions: Close (1) and Open (2). An "x" in the column is used to determine the status of the contacts (1
In the given switch development matrix, an x is in column (2), which corresponds to the Open position. The contacts 3-4 will be in the condition as drawn (i.e. closed). Should the switch be in the Close position, column (1), it does not have an x in it; therefore contacts 3-4 would be open.
-2, 3-4) associated with the switch. In the given switch development matrix, an "x" is in column (2), which corresponds to the Open position. The contacts 3
Distractors:
-4 will be in the condition as drawn (i.e. closed). Should the switch be in the Close position, column (1), it does not have an "x" in it; therefore contacts 3
B. This option is incorrect because the Operations Support Center is not on the Controlled Drawing distribution list. Contact 3-4 is open when the switch is in the Close position. A candidate may believe that because the operations support center uses many drawing that this would be where controlled copies exist would choose this option. It should also be noted that this may be confused with the Operational Support Center which is in the same area as the TSC and does maintain controlled documents for emergency purposes.
-4 would be open. Distractors:
C. This option is incorrect because contact 3-4 is open when the switch is in the Close position. A candidate may believe that the position of the x in the switch development means open and would choose this option. This answer is plausible because the drawing location is correct.
B. This option is incorrect because the Operations Support Center is not on the Controlled Drawing distribution list. Contact 3
D. This option is incorrect because the Operations Support Center is not on the Controlled Drawing distribution list and contact 3-4 is open when the switch is in the Close position.
-4 is open when the switch is in the Close position. A candidate may believe that because the operations support center uses many drawing that this would be where controlled copies exist would choose this option. It should also be noted that this may be confused with the Operational Support Center which is in the same area as the TSC and does maintain controlled documents for emergency purposes.
A candidate may believe that the position of the x in the switch development means open and that the OSC is on the distribution list and choose this option. This answer is plausible because the switch contact 3-4 do exist. It should also be noted that this may be confused with the Operational Support Center which is in the same area as the TSC and does maintain controlled documents for emergency purposes.
 
Technical Reference(s): 3.DRAWING, Drawing Control, Rev. 4 0.9, Tagout, Rev. 85 OTH015-09-08, Electrical Print Reading for Clearance Order and Circuit                                       Evaluation Applications, Rev. 0 Proposed references to be provided to applicants during examination:         none Learning Objective:     OTH015-09-08/Electrical Print Reading for Clearance Order and Circuit Evaluation Applications
C. This option is incorrect because contact 3
: 5.       Describe the basic types of electrical drawings.
-4 is open when the switch is in the Close position. A candidate may believe that the position of the x in the switch development means open and would choose this option. This answer is plausible because the drawing location is correct.
D. This option is incorrect because the Operations Support Center is not on the Controlled Drawing distribution list and contact 3
-4 is open when the switch is in the Close position. A candidate may believe that the position of the x in the switch development means open and that the OSC is on the distribution list and choose this option. This answer is plausible because the switch contac t 3-4 do exist. It should also be noted that this may be confused with the Operational Support Center which is in the same area as the TSC and does maintain controlled documents for emergency purposes.
Technical Reference(s):
3.DRAWING, Drawing Control, Rev. 4 0.9, Tagout, Rev. 85 OTH015-09-08, Electrical Print Reading for Clearance Order and Circuit   Evaluation Applications, Rev. 0 Proposed references to be provided to applicants during examination:
none Learning Objective:
OTH015-09-08/Electrical Print Reading for Clearance Order and Circuit Evaluation Applications
: 5. Describe the basic types of electrical drawings.
: a. Determine the status of prints and actions required when generating a clearance order.
: a. Determine the status of prints and actions required when generating a clearance order.
: b. Determine where electrical prints are available and how to retrieve them.
: b. Determine where electrical prints are available and how to retrieve them.
Question Source:                Bank #
Modified Bank #
New X Question History:                Last NRC Exam 157
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content:  55.41 (10)
Comments:
LOD 2 158


Question Source:
ES-401                           Sample Written Examination                       Form ES-401-5 Question Worksheet Examination Outline Cross-
Bank #  Modified Bank #
New X  Question History:
Last NRC Exam 157 Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X    10 CFR Part 55 Content:
55.41 (10)
Comments:  LOD 2    158 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
-


==Reference:==
==Reference:==
Level   RO   Tier #     3   Group #   K/A #   2.3.4 Importance Rating 3.2   2.3.4 Knowledge of radiation exposure limits under normal or emergency conditions.
Level               RO Tier #               3 Group #
(CFR: 41.12 / 43.4 / 45.10)
K/A #               2.3.4 Importance Rating   3.2 2.3.4 Knowledge of radiation exposure limits under normal or emergency conditions.
Question: 71
(CFR: 41.12 / 43.4 / 45.10)
 
Question: 71 Which of the following is the FIRST accumulated dose value ABOVE WHICH a tour member is required to have an NRC Form 5 or equivalent issued IAW 9.ALARA.1 (Personnel Dosimetry and Occupational Radiation Exposure Program)?
Which of the following is the FIRST accumulated dose value ABOVE WHICH a tour member is required to have an NRC Form 5 or equivalent issued IAW 9.ALARA.1 (Personnel Dosimetry and Occupational Radiation Exposure Program)?
A. 100 mrem B. 500 mrem C. 1000 mrem D. 5000 mrem Answer:
 
A. 100 mrem Explanation:
A. 10 0 mrem B. 50 0 mrem C. 100 0 mrem D. 500 0 mrem Answer: A. 10 0 mrem Explanation:
 
9.ALARA.1 Personnel Dosimetry and Occupational Radiation Exposure Program step 5.5 specifies the dose value which requires an NRC Form 5 or equivalent to be issued. The accumulated dose value specified is >100 mrem. 9.ALARA.13 provides the instructions for completing the NRC Form 5.
9.ALARA.1 Personnel Dosimetry and Occupational Radiation Exposure Program step 5.5 specifies the dose value which requires an NRC Form 5 or equivalent to be issued. The accumulated dose value specified is >100 mrem. 9.ALARA.13 provides the instructions for completing the NRC Form 5.
Distractors:
Distractors:
B. This option is incorrect because the selection is above the >100 mrem accumulated value. This choice is plausible because it is the accumulated dose allowed for a pregnant female per 10CFR20.1208. If the candidate confuses these two values this option would be chosen due to 500 mrem being a familiar dose value.
B. This option is incorrect because the selection is above the >100 mrem accumulated value.
C. This option is incorrect because the selection is above the >100 mrem accumulated value. This choice is plausible because it is the CNS administrative dose limit for the year. If the candidate confuses these two values this option would be chosen due to 1000 mrem being a familiar dose value.
This choice is plausible because it is the accumulated dose allowed for a pregnant female per 10CFR20.1208. If the candidate confuses these two values this option would be chosen due to 500 mrem being a familiar dose value.
C. This option is incorrect because the selection is above the >100 mrem accumulated value.
This choice is plausible because it is the CNS administrative dose limit for the year. If the candidate confuses these two values this option would be chosen due to 1000 mrem being a familiar dose value.
D. This option is incorrect because the selection is above the >100 mrem accumulated value.
This choice is plausible because it is the annual allowed 10CFR20.1201 TEDE dose. If the candidate confuses these two values this option would be chosen due to its potential familiarity.
159
 
Technical Reference(s):      9.ALARA.1 (Personnel Dosimetry and Occupational Radiation Exposure Program), Rev. 43 Proposed references to be provided to applicants during examination: NONE Learning Objective: INT032-01-15R06 OPS CNS Administrative Procedures Radiation Protection and Chemistry Procedures.
D. Procedure 9.ALARA.1, Personnel Dosimetry and Occupational Radiation Exposure            Program
: g.      Monitoring Tour Groups Question Source:            Bank #
Modified Bank #
New X Question History:            Last NRC Exam Question Cognitive Level:    Memory or Fundamental Knowledge        X Comprehension or Analysis 10 CFR Part 55 Content:      55.41 (12)
Comments:
LOD: 4 160


D. This option is incorrect because the selection is above the >100 mrem accumulated value. This choice is plausible because it is the annual allowed 10CFR20.1201 TEDE dose. If the candidate confuses these two values this option would be chosen due to its potential familiarity.
161 162 163 ES-401                             Sample Written Examination                     Form ES-401-5 Question Worksheet Examination Outline Cross-
159 Technical Reference(s):
9.ALARA.1 (Personnel Dosimetry and Occupational Radiation Exposure Program), Rev. 43 Proposed references to be provided to applicants during examination:
NONE  Learning Objective:
INT032-01-15R06 OPS CNS Administrative Procedures Radiation Protection and Chemistry Procedures.
D. Procedure 9.ALARA.1, Personnel Dosimetry and Occupational Radiation Exposure  Program    g. Monitoring Tour Groups Question Source:
Bank #  Modified Bank #
New X  Question History:
Last NRC Exam Question Cognitive Level:
Memory or Fundamental Knowledge  X Comprehension or Analysis    10 CFR Part 55 Content:
55.41 (12)  Comments:  LOD: 4  160 161 162 163 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
-


==Reference:==
==Reference:==
 
Level                 RO Tier #               3 Group #
Level   RO   Tier #     3   Group #     K/A #   2.3.5 Importance Rating 2.9
K/A #                 2.3.5 Importance Rating     2.9 2.3.5 Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.
 
Question: 72 What is/are the MINIMUM personnel monitoring requirement(s) for exiting a contaminated area in the Reactor Building and dressing in street clothing IAW 9.EN-RP-100 (Radiation Worker Expectations)?
====2.3.5 Ability====
to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.
Question: 72 What is/are the MINIMUM personnel monitoring requirement(s) for exiting a contaminated area in the Reactor Building and dressing in street clothing IAW 9.EN
-RP-100 (Radiation Worker Expectations)?
A. Perform a hand and foot frisk with a frisker ONLY.
A. Perform a hand and foot frisk with a frisker ONLY.
B. Perform a whole body contamination monitor scan ONLY.
B. Perform a whole body contamination monitor scan ONLY.
C. Perform a whole body frisk with a frisker then a whole body contamination monitor scan.
C. Perform a whole body frisk with a frisker then a whole body contamination monitor scan.
D. Perform a hand and foot frisk with a frisker then a whole body contamination monitor scan.
D. Perform a hand and foot frisk with a frisker then a whole body contamination monitor scan.
Answer: D. Perform a hand and foot frisk with a frisker then a whole body contamination monitor scan.
Answer:
D. Perform a hand and foot frisk with a frisker then a whole body contamination monitor scan.
Explanation:
Explanation:
When exiting a contaminated area within the Reactor Building, personnel are required as a minimum to perform a hand and foot frisk (with a frisker) as soon as practical upon exiting the CA (RB airlock has friskers to support as soon as practical) and then proceed to a PCM for whole body contamination monitoring. Clothing can then be changed within the RCA. If Exiting the RCA personnel are required monitor themselves for contamination with a whole body contamination monitor and a gamma portal monitor.
When exiting a contaminated area within the Reactor Building, personnel are required as a minimum to perform a hand and foot frisk (with a frisker) as soon as practical upon exiting the CA (RB airlock has friskers to support as soon as practical) and then proceed to a PCM for whole body contamination monitoring. Clothing can then be changed within the RCA. If Exiting the RCA personnel are required monitor themselves for contamination with a whole body contamination monitor and a gamma portal monitor.
The candidate should recognize that contamination control requires a minimum of a hand and foot frisk to reduce the potential for spread of contamination. The requirement to exit the RCA via a whole body monitor is always a requirement.
The candidate should recognize that contamination control requires a minimum of a hand and foot frisk to reduce the potential for spread of contamination. The requirement to exit the RCA via a whole body monitor is always a requirement.
A. This answer is incorrect because it does not include the whole body monitor. The candidate could choose this distractor if he/she does not equate the whole body monitor with contamination control. This answer is plausible because performing hand and foot frisk is required. B. This answer is incorrect because it does not include hand and foot frisk. The candidate could choose this distractor if he/she does not equate hand and foot frisk with contamination control. This answer is plausible because using the whole body monitor is required.
A. This answer is incorrect because it does not include the whole body monitor. The candidate could choose this distractor if he/she does not equate the whole body monitor with contamination control. This answer is plausible because performing hand and foot frisk is required.
164 C. This answer is incorrect because it does not properly show the allowance for a hand and foot frisk. The candidate could choose this distractor if he/she did not remember the minimum requirement of the procedure. This answer is plausible because using the whole body monitor is required.
B. This answer is incorrect because it does not include hand and foot frisk. The candidate could choose this distractor if he/she does not equate hand and foot frisk with contamination control. This answer is plausible because using the whole body monitor is required.
164
 
C. This answer is incorrect because it does not properly show the allowance for a hand and foot frisk. The candidate could choose this distractor if he/she did not remember the minimum requirement of the procedure. This answer is plausible because using the whole body monitor is required.
Technical Reference(s):      9.EN-RP-100, Radiation Worker Expectations, Rev. 4 Proposed references to be provided to applicants during examination:        None Learning Objective:
Question Source:              Bank #
Modified Bank #
New X Question History:            Last NRC Exam Question Cognitive Level:    Memory or Fundamental Knowledge              X Comprehension or Analysis 10 CFR Part 55 Content:      55.41 (11)
Comments:
LOD 2 165


Technical Reference(s):
166 167 ES-401                             Sample Written Examination                     Form ES-401-5 Question Worksheet Examination Outline Cross-
9.EN-RP-100, Radiation Worker Expectations, Rev. 4 Proposed references to be provided to applicants during examination:
None Learning Objective:
Question Source:
Bank #    Modified Bank #
New X Question History:
Last NRC Exam Question Cognitive Level:
Memory or Fundamental Knowledge X Comprehension or Analysis    10 CFR Part 55 Content: 55.41 (11)
Comments: LOD 2 165 166 167 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
-


==Reference:==
==Reference:==
 
Level                 RO Tier #                 3 Group #
Level   RO   Tier #     3   Group #   K/A #   2.3.13 Importance Rating 3.4   2.3.13 Knowledge of radiological safety procedures pertaining to licensed operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high
K/A #                 2.3.13 Importance Rating     3.4 2.3.13 Knowledge of radiological safety procedures pertaining to licensed operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.
-radiation areas, aligning filters, etc.
(CFR: 41.12 / 43.4 / 45.9 / 45.10)
(CFR: 41.12 / 43.4 / 45.9 / 45.10)
Question: 73 An Operator has to enter a room to close a valve to stop a small water leak.
Question: 73 An Operator has to enter a room to close a valve to stop a small water leak.
The affected work area general radiation level is 1600 mrem/hour.
* The affected work area general radiation level is 1600 mrem/hour.
What type of entry permit is required?
What type of entry permit is required?
Is continuous RP coverage required during the entry?
Is continuous RP coverage required during the entry?
Entry Permit Continuous RP Coverage A. SWP NOT Required
Entry Permit             Continuous RP Coverage A.     SWP                         NOT Required B.      SWP                        Required C.      RWP                        Required D.      RWP                        NOT Required Answer:
B.      SWP                        Required Explanation:
Requires knowledge of LHRA entry requirements. IAW 9.EN-RP-101, entry into LHRAs require a SWP and continuous HP coverage. A Locked High Radiation Area is an area accessible to individuals in which radiation levels from sources external to the body could result in an individual receiving a deep dose equivalent > 1 rem (10 mSv) in 1 hour at 30 cm (~ 12") from the radiation source or from any surface that the radiation penetrates. A Special Work Permit Area (SWP) is an area where a SWP has been issued to control access to, and work within, which involves any one or combination of the conditions: a Very High Radiation Area, Locked High Radiation Area, a High Radiation Area, High Contamination Area, Discrete Radioactive Particle Area, or an Airborne Radiation Area.
Distracters:
A. This answer is incorrect because continuous RP coverage is required. This answer is plausible because an SWP permit is required for entry and under some conditions 168


B. SWP Required C. RWP Required  D. RWP NOT Required
Operations personnel can take action to protect the health and safety of the public without continuous RP coverage. The isolation of a small leak is not an instance where health and safety of the public is an issue. The candidate who believes continuous RP coverage is NOT required would select this answer.
C. This answer is incorrect because an SWP is required for entry. This answer is plausible because there are instances where Operations personnel can take actions to protect the health and safety of the public without signing on a permit or signing on the SWP. However, the action to isolate a small leak is not an instance where the health and safety of the public is an issue. The candidate who understands continuous RP coverage is required but doesnt realize an SWP is required would select this answer.
D. This answer is incorrect because an SWP and Continuous RP coverage are required to enter a LHRA. This answer is plausible because there are instances where Operations personnel can take actions to protect the health and safety of the public without signing on a permit or signing on the SWP and continuous HP coverage not being required. The candidate who believes the action can be taken and doesnt realize the requirements of a locked high radiation area entry would select this answer.
Technical Reference(s): Procedure 9.EN-RP-101 (Access Control For Radiologically Controlled Areas), Rev 15 Proposed references to be provided to applicants during examination:        NONE Learning Objective:    INT032-01-100 OPS CNS Administrative Procedures Radiation Protection H.      9-EN-RP-101, Access Control for Radiologically Controlled Areas
: 1.      Precautions and limitations
: 2.      RCA access and egress Question Source:              Bank #
Modified Bank #
New X Question History:              N/A Question Cognitive Level:      Memory or Fundamental Knowledge              X Comprehension or Analysis 10 CFR Part 55 Content:        55.41 (12)
Comments:
LOD 2 169


Answer:  B. SWP Required  Explanation:
170 ES-401                             Sample Written Examination                       Form ES-401-5 Question Worksheet Examination Outline Cross-
Requires knowledge of LHRA entry requirements. IAW 9.EN
-RP-101, entry into LHRAs require a SWP and continuous HP coverage. A Locked High Radiation Area is an area accessible to individuals in which radiation levels from sources external to the body could result in an individual receiving a deep dose equivalent > 1 rem (10 mSv) in 1 hour at 30 cm (~ 12") from the radiation source or from any surface that the radiation penetrates. A Special Work Permit Area (SWP) is an area where a SWP has been issued to control access to, and work within, which involves any one or combination of the conditions: a Very High Radiation Area, Locked High Radiation Area, a High Radiation Area, High Contamination Area, Discrete Radioactive Particle Area, or an Airborne Radiation Area.
Distracters:
A. This answer is incorrect because continuous RP coverage is required. This answer is plausible because an SWP permit is required for entry and under some conditions 168 Operations personnel can take action to protect the health and safety of the public without continuous RP coverage. The isolation of a small leak is not an instance where health and safety of the public is an issue. The candidate who believes continuous RP coverage is NOT required would select this answer.
C. This answer is incorrect because an SWP is required for entry. This answer is plausible because there are instances where Operations personnel can take actions to protect the health and safety of the public without signing on a permit or signing on the SWP. However, the action to isolate a small leak is not an instance where the health and safety of the public is an issue. The candidate who understands continuous RP coverage is required but doesn't realize an SWP is required would select this answer.
D. This answer is incorrect because an SWP and Continuous RP coverage are required to enter a LHRA. This answer is plausible because there are instances where Operations personnel can take actions to protect the health and safety of the public without signing on a permit or signing on the SWP and continuous HP coverage not being required. The candidate who believes the action can be taken and doesn't realize the requirements of a locked high radiation area entry would select this answer.
Technical Reference(s):  Procedure 9.EN
-RP-101 (Access Control For Radiologically Controlled Areas), Rev 15 Proposed references to be provided to applicants during examination:
NONE  Learning Objective:
INT032-01-100 OPS CNS Administrative Procedures Radiation Protection H. 9-EN-RP-101, Access Control for Radiologically Controlled Areas
: 1. Precautions and limitations
: 2. RCA access and egress Question Source:
Bank #  Modified Bank #
New X  Question History:
N/A  Question Cognitive Level:
Memory or Fundamental Knowledge X Comprehension or Analysis    10 CFR Part 55 Content:
55.41 (12)  Comments:  LOD 2  169 170 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
-


==Reference:==
==Reference:==
 
Level                 RO Tier #               3 Group #
Level   RO   Tier #     3   Group #   K/A #   2.4.45 Importance Rating 4.1 2.4.45 Ability to prioritize and interpret the significance of each annunciator or alarm.
K/A #                 2.4.45 Importance Rating     4.1 2.4.45 Ability to prioritize and interpret the significance of each annunciator or alarm.
(CFR: 41.10 / 43.5 / 45.3 / 45.12)
(CFR: 41.10 / 43.5 / 45.3 / 45.12)
 
Question: 74 The Plant is at power when a transient occurs.
Q ues ti on: 7 4  Th e P l an t i s a t po w e r w he n a tr ans i en t occu r s. M ulti p l e B l ac k an d Y e ll o w outli ne d annunc i a t o r s a r e i n a l a r m. W h i c h co l o r e d a l a r m t ak e p recedenc e an d w hy IAW Procedure 2.3.1 (General Alarm Procedure)
* Multiple Black and Yellow outlined annunciators are in alarm.
?  A. BLACK; an EOP entry is required.
Which colored alarm take precedence and why IAW Procedure 2.3.1 (General Alarm Procedure)?
A. BLACK; an EOP entry is required.
B. YELLOW; a plant shutdown condition may be present.
B. YELLOW; a plant shutdown condition may be present.
C. BLACK; BOTH an EOP entry is required AND a Plant Shutdown condition may be present. D. YELLOW; BOTH an EOP entry is required AND a Plant Shutdown condition may be present. Answer: B. YELLOW; a Plant Shutdown condition may be present.
C. BLACK; BOTH an EOP entry is required AND a Plant Shutdown condition may be present.
D. YELLOW; BOTH an EOP entry is required AND a Plant Shutdown condition may be present.
Answer:
B. YELLOW; a Plant Shutdown condition may be present.
Explanation:
Explanation:
The window box assembly is a matrix of divided lamp windows with engraved legend plates and multi-colored window bezels. Each window has been given a "PRIORITY" signifying the importance of the alarm:
The window box assembly is a matrix of divided lamp windows with engraved legend plates and multi-colored window bezels. Each window has been given a "PRIORITY" signifying the importance of the alarm:
 
Priority I - RED; alarms that alert of EOP entry conditions or conditions requiring or causing an automatic or manual plant shutdown, or significant system setpoints.
Priority I  
Priority II - YELLOW; alarm conditions which may require or rapidly cause a plant shutdown or radiation release.
- RED; alarms that alert of EOP entry conditions or conditions requiring or causing an automatic or manual plant shutdown, or significant system setpoints.
Priority III - BLACK; alarms that indicate off normal plant conditions that affect plant or component         operability but should not lead to plant shutdown or radiation release.
Priority II  
- YELLOW; alarm conditions which may require or rapidly cause a plant shutdown or radiation release.
Priority III  
- BLACK; alarms that indicate off normal plant conditions that affect plant or component operability but should not lead to plant shutdown or radiation release.
Distracters:
Distracters:
A. This option is incorrect because black outlined alarms do not represent EOP entry conditions and do not have a higher priority than yellow outlines alarms. The candidate may 171 choose this if he/she does not understand the color of EOP entry tiles. This option is plausible because black outlines alarms do indicate off normal conditions.
A. This option is incorrect because black outlined alarms do not represent EOP entry conditions and do not have a higher priority than yellow outlines alarms. The candidate may 171


C. This option is incorrect because yellow outlined alarms do not indicate an EOP entry condition is present. The candidate may choose this if he/she does not understand the meaning of color
choose this if he/she does not understand the color of EOP entry tiles. This option is plausible because black outlines alarms do indicate off normal conditions.
-coded alarms. This option is plausible because yellow outlined alarms do indicate off normal conditions and do have a higher priority than black outline alarms.
C. This option is incorrect because yellow outlined alarms do not indicate an EOP entry condition is present. The candidate may choose this if he/she does not understand the meaning of color-coded alarms. This option is plausible because yellow outlined alarms do indicate off normal conditions and do have a higher priority than black outline alarms.
D. This option is incorrect because the black outlined alarms do not indicate a plant shutdown condition may be present. The candidate may choose this if he/she does not understand the meaning of color-coded alarms. This option is plausible because black outlined alarms do indicate off normal conditions.
Technical Reference(s):      Procedure 2.3.1 (General Alarm Procedure), Rev. 63 Proposed references to be provided to applicants during examination:      NONE Learning Objective:    COR002-35-02, Plant Annunciator System LO-02 State the purpose of the following components related to the Plant Annunciator System.:
: k. Alarm Window Boxes Question Source:              Bank #
Modified Bank #
New X Question History:
Question Cognitive Level:    Memory or Fundamental Knowledge              X Comprehension or Analysis 10 CFR Part 55 Content:      55.41 (10)
Comments:
LOD 3 172


D. This option is incorrect because the black outlined alarms do not indicate a plant shutdown condition may be present. The candidate may choose this if he/she does not understand the meaning of col or-coded alarms. This option is plausible because black outlined alarms do indicate off normal conditions.
173 ES-401                           Sample Written Examination                   Form ES-401-5 Question Worksheet Examination Outline Cross-
Technical Reference(s):
Procedure 2.3.1 (General Alarm Procedure), Rev. 63 Proposed references to be provided to applicants during examination:
NONE  Learning Objective:
COR002-35-02, Plant Annunciator System LO-02 State the purpose of the following components related to the Plant Annunciator System.:
: k. Alarm Window Boxes
 
Question Source:
Bank #  Modified Bank #
New X  Question History:    Question Cognitive Level:
Memory or Fundamental Knowledge X Comprehension or Analysis    10 CFR Part 55 Content:
55.41 (10)  Comments:  LOD 3    172 17 3 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
-


==Reference:==
==Reference:==
Level                RO Tier #                3 Group #
K/A #                2.4.49 Importance Rating    4.6 2.4.49 Ability to perform without reference to procedures those actions that require immediate operation of system components and controls.
(CFR: 41.10 / 43.2 / 45.6)
Question: 75 There are operational circumstances when operators must perform immediate operator actions without reference to procedures.
Which statement represents one of those circumstances?
A. When an EOP directs performing the action.
B. When a scram is directed by Technical Specifications.
C. When an alarm procedure directs performing the action.
D. When an abnormal procedure directs performing the action.
Answer:
D. When an abnormal procedure directs performing the action.
Explanation:
Abnormal and Emergency (Non-EOP) procedures contain Immediate Operator Actions which the control room operator has committed to memory. Should a condition exist that requires Immediate Operator Actions, Procedure 2.0.1.2 directs performing the action without use of procedures.
Distractors:
A: This answer is incorrect because EOPs do not contain immediate operator actions. Actions are taken per the EOPs without the control room operator having the procedure in hand, but the actions are directed by the CRS. This answer is plausible because actions are directed to be taken without use of the procedure but they are not immediately performed from memory. The candidate who recalls immediately performing actions directed from EOPs would select this answer.
B: This answer is incorrect because no actions are taken immediately from Technical Specifications. This answer is plausible because Technical Specifications have completion times to be taken immediately, but per TS immediate means to pursue without delay and in a controlled manner.
C: This answer is incorrect because alarm procedures do not contain immediate operator actions. This answer is plausible because actions can be performed immediately after 174


Level  RO  Tier #    3  Group #    K/A #  2.4.49 Importance Rating 4.6  2.4.49 Ability to perform without reference to procedures those actions that require immediate operation of system components and controls.
entering the procedure. The candidate who recalls scram actions contained in alarm procedures may believe the action can be taken prior to entering the alarm procedure would select this answer.
(CFR: 41.10 / 43.2 / 45.6)
Technical Reference(s): Procedure 2.0.1.2 (Operations Procedure Policy), Rev. 44 Proposed references to be provided to applicants during examination:     none Learning Objective:   INT032-01-03 (OPS CNS Administrative Procedure Conduct of Operations and General Alarm Procedures (Formal Classroom/Pre-OJT Training)
Q ues ti on: 7 5  The r e a r e ope r a ti ona l c ir cu m s tance s w he n ope r a t o r s m us t perform immediate operator actions w it hou t r e f e renc e t o p rocedu res. W h i c h s t a t e m en t r ep resen t s on e o f thos e c ir cu m s tances?  A. W he n a n EO P d ir ec t s pe rf o r m i n g t h e acti on. B. W he n a sc r a m is d ir ec t e d b y Techn i ca l Specifi cati on s. C. W he n a n a l a r m p rocedu r e d ir ec t s pe rf o r m i n g t h e acti on. D. W he n a n abno r m a l p rocedu r e d ir ec t s pe rf o r m i n g th e action. A ns w e r: D. W he n a n abno r m a l p rocedu r e d ir ec t s pe rf o r m i n g t h e acti on. Explanation:
G. Procedure 2.0.1.2, Operations Procedure Policy
Abnormal and Emergency (Non
-EOP) procedures contain Immediate Operator Actions which the control room operator has committed to memory. Should a condition exist that requires Immediate Operator Actions, Procedure 2.0.1.2 directs performing the action without use of procedures.
 
Distractors:
A:  This answer is incorrect because EOPs do not contain immediate operator actions. Actions are taken per the EOPs without the control room operator having the procedure in hand, but the actions are directed by the CRS. This answer is plausible because actions are directed to be taken without use of the procedure but they are not immediately performed from memory. The candidate who recalls immediately performing actions directed from EOPs would select this answer.
B:  This answer is incorrect because no actions are taken immediately from Technical Specifications. This answer is plausible because Technical Specifications have completion times to be taken immediately, but per TS immediate means to pursue without delay and in a controlled manner.
C:  This answer is incorrect because alarm procedures do not contain immediate operator actions. This answer is plausible because actions can be performed immediately after 174 entering the procedure. The candidate who recalls scram actions contained in alarm procedures may believe the action can be taken prior to entering the alarm procedure would select this answer.
Technical Reference(s): Procedure 2.0.1.2 (Operations Procedure Policy), Rev. 44 Proposed references to be provided to applicants during examination:
none Learning Objective:
INT032-01-03 (OPS CNS Administrative Procedure Conduct of Operations and General Alarm Procedures (Formal Classroom/Pre
-OJT Training)   G. Procedure 2.0.1.2, Operations Procedure Policy
: 1. Discuss the following as described in Procedure 2.0.1.2, Operations Procedure Policy.
: 1. Discuss the following as described in Procedure 2.0.1.2, Operations Procedure Policy.
: f. Attachment 1, Immediate Operator Actions Question Source:
: f. Attachment 1, Immediate Operator Actions Question Source:             Bank #
Bank # Modified Bank #
Modified Bank #
New X   Question History:
New X Question History:
Question Cognitive Level:
Question Cognitive Level:   Memory or Fundamental Knowledge Comprehension or Analysis                   X 10 CFR Part 55 Content:     55.41 (10)
Memory or Fundamental Knowledge Comprehension or Analysis   X 10 CFR Part 55 Content:
Comments:
55.41 (10)
LOD 2 175
Comments: LOD 2   175 176 177 178 179 180 RO References
 
T o r u s p r e s s u r e (p s i g)______________H y d r o s t a t i c h e a d (p s i g)P C w a t e r l e v e l (f t.)____________S t r a i n e r l e v e l (f t.)____________0.4 3 x____________=_____________T o r u s o v e r p r e s s u r e (p s i g)______________+-4 N O T E 3 T o r u s o v e r p r e s s u r e i s s u m o f t o r u s p r e s s u r e a n d h y d r o s t a t i c h e a d a b o v e s u c t i o n s t r a i n e r T o r u s p r e s s u r e (p s i g)______________H y d r o s t a t i c h e a d (p s i g)P C w a t e r l e v e l (f t.)____________S t r a i n e r l e v e l (f t.)____________0.4 3 x____________=_____________T o r u s o v e r p r e s s u r e (p s i g)______________+-4 N O T E 3 T o r u s o v e r p r e s s u r e i s s u m o f t o r u s p r e s s u r e a n d h y d r o s t a t i c h e a d a b o v e s u c t i o n s t r a i n e r


Figure 1 - TIME TO BOILING
176 177 178 179 180 RO References NOTE 3 Torus overpressure is sum of torus pressure and hydrostatic head above suction strainer Torus pressure (psig)                      ______________
- WATER LEVEL AT HIGH LEVEL TRIP
Hydrostatic head (psig)
PC water level (ft.)  ____________
Strainer level (ft.)      -4 0.43 x    ____________ =    +
Torus overpressure (psig)                      ______________


Figure 2 - TIME TO BOILING  
Figure 1 - TIME TO BOILING - WATER LEVEL AT HIGH LEVEL TRIP Figure 2 - TIME TO BOILING - WATER LEVEL AT FLANGE Figure 3 - TIME TO BOILING - WATER TO LEVEL FLOODED TO 1001
- WATER LEVEL AT FLANGE


Figure 3 - TIME TO BOILING
U.S. Nuclear Regulatory Commission Site-Specific SRO Written Examination Applicant Information Name:
- WATER TO LEVEL FLOODED TO 1001'
Date:                                             Facility/Unit: Cooper Nuclear Station Region:         I     II     III   IV           Reactor Type: W       CE     BW     GE Start Time:                                       Finish Time:
 
Instructions Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. To pass the examination you must achieve a final grade of at least 80.00 percent overall, with 70.00 percent or better on the SRO-only items if given in conjunction with the RO exam; SRO-only exams given alone require a final grade of 80.00 percent to pass. You have 8 hours to complete the combined examination, and 3 hours if you are only taking the SRO portion.
U.S. Nuclear Regulatory Commission Site-Specific SRO Written Examination Applicant Information Name: Date: Facility/Unit:
Cooper Nuclear Station Region: I   II   III   IV   Reactor Type: W CE BW GE   Start Time:
Finish Time:
Instructions Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. To pass the examination you must achieve a final grade of at least 80.00 percent overall, with 70.00 percent or better on the SRO-only items if given in conjunction with the RO exam; SRO
-only exams given alone require a final grade of 80.00 percent to pass. You have 8 hours to complete the combined examination, and 3 hours if you are only taking the SRO portion.
Applicant Certification All work done on this examination is my own. I have neither given nor received aid.
Applicant Certification All work done on this examination is my own. I have neither given nor received aid.
______________________________________
Applicants Signature Results RO/SRO-Only/Total Examination Values                                 /      /        Points Applicants Scores                                                  /      /        Points Applicants Grade                                                    /      /          Percent 1
Applicant's Signature Results RO/SRO-Only/Total Examination Values


  /            /              Points    Applicant's Scores
ES-401                           Sample Written Examination                   Form ES-401-5 Question Worksheet Examination Outline Cross-
 
  /            /              Points    Applicant's Grade
 
  /            /                Percent    1 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross
-


==Reference:==
==Reference:==


Level SRO Tier #   1 Group # 1 K/A # 295006G2.1.6 Importance Rating
Level                   SRO Tier #                   1 Group #                 1 K/A #             295006G2.1.6 Importance Rating       4.8 295006 SCRAM 2.1.6: Ability to manage control room crew during plant transients.
Question: 76 Which of the following completes the statement below identifying the CRS direction to the Reactor Operators following an Automatic Scram from rated power and the procedure which provides guidance for this direction?
Assign the RO Procedure 2.1.5 (Reactor Scram) Attachments ____(1)____ IAW Procedure
____(2)____.
Attachment 1  Mitigating Task Scram Actions Attachment 2  Reactor Power Control Attachment 3  Reactor Water Level Control Attachment 4  Reactor Pressure Control Attachment 5  Balance of Plant Actions A. (1) 1 & 2 ONLY AND assign the BOP Attachments 3, 4, and 5.
(2) 2.0.3 (Conduct of Operations)
B. (1) 1 & 2 ONLY AND assign the BOP Attachments 3, 4, and 5.
(2) 2.0.1.3 (Time Critical Operator Action Control and Maintenance)
C. (1) 1, 2 & 3 ONLY AND assign the BOP Attachments 4 and 5.
(2) 2.0.1.3 (Time Critical Operator Action Control and Maintenance)
D. (1) 1, 2 & 3 ONLY AND assign the BOP Attachments 4 and 5.
(2) 2.0.3 (Conduct of Operations)
Answer:
D. (1) 1, 2 & 3 ONLY AND assign the BOP Attachments 4 and 5.
(2) 2.0.3 (Conduct of Operations).
2


===4.8 295006===
Explanation:
SCRAM 2.1.6: Ability to manage control room crew during plant transients.
Requires knowledge and coordination of procedure 2.1.5 (Reactor Scram) Attachments and selecting the procedure which provides for division of operator responsibilities during a transient. IAW procedure 2.0.3, The CRO-RO is normally responsible for reactivity control, safe reactor shutdown, and mitigating task scram actions and post-scram reactor level control (Attachments 1, 2, and 3 of Procedure 2.1.5). The CRO-BOP is normally responsible for post-scram Pressure Control and BOP System operation (Attachments 4 and 5 of Procedure 2.1.5).
Question:  76 Which of the following completes the statement below identifying the CRS direction to the Reactor Operators following an Automatic Scram from rated power and the procedure which provides guidance for this direction?
Assign the RO Procedure 2.1.5 (Reactor Scram) Attachments ____(1)____ IAW Procedure ____(2)____.
Attachment 1 Mitigating Task Scram Actions Attachment 2 Reactor Power Control Attachment 3 Reactor Water Level Control Attachment 4 Reactor Pressure Control Attachment 5 Balance of Plant Actions A.  (1)  1 & 2 ONLY AND assign the BOP Attachments 3, 4, and 5.
  (2)  2.0.3 (Conduct of Operations)
B. (1)  1 & 2 ONLY AND assign the BOP Attachments 3, 4, and 5.
  (2)  2.0.1.3 (Time Critical Operator Action Control and Maintenance)
 
C. (1)  1, 2 & 3 ONLY AND assign the BOP Attachments 4 and 5.  (2)  2.0.1.3 (Time Critical Operator Action Control and Maintenance)
 
D. (1)  1, 2 & 3 ONLY AND assign the BOP Attachments 4 and 5.  (2)  2.0.3 (Conduct of Operations)
Answer:  D. (1)  1, 2 & 3 ONLY AND assign the BOP Attachments 4 and 5.  (2)  2.0.3 (Conduct of Operations).
2 Explanation:
Requires knowledge and coordination of procedure 2.1.5 (Reactor Scram) Attachments and selecting the procedure which provides for division of operator responsibilities during a transient. IAW procedure 2.0.3, The CRO
-RO is normally responsible for reactivity control, safe reactor shutdown, and mitigating task scram actions and post
-scram reactor level control (Attachments 1, 2, and 3 of Procedure 2.1.5). The CRO
-BOP is normally responsible for post
-scram Pressure Control and BOP System operation (Attachments 4 and 5 of Procedure 2.1.5).
Assign the RO Procedure 2.1.5 (Reactor Scram) Attachments 1, 2, & 3 AND the BOP Attachments 4& 5 IAW Procedure 2.0.3 (Conduct of Operations).
Assign the RO Procedure 2.1.5 (Reactor Scram) Attachments 1, 2, & 3 AND the BOP Attachments 4& 5 IAW Procedure 2.0.3 (Conduct of Operations).
Distracters:
Distracters:
A. This answer is incorrect due to RPV Level control being assigned to the RO. This answer is plausible if ATWS conditions were present, assigning level control to the BOP would be correct. (One of four validators chose this answer due to a common misconception). The candidate who confuses reactor conditions following a scram and correctly identifies the procedure providing guidance would select this answer.
A. This answer is incorrect due to RPV Level control being assigned to the RO. This answer is plausible if ATWS conditions were present, assigning level control to the BOP would be correct. (One of four validators chose this answer due to a common misconception). The candidate who confuses reactor conditions following a scram and correctly identifies the procedure providing guidance would select this answer.
B. This answer is incorrect due to RPV Level control being assigned to the RO and the procedure not providing division of operator responsibilities.
B. This answer is incorrect due to RPV Level control being assigned to the RO and the procedure not providing division of operator responsibilities. This answer is plausible if ATWS conditions were present, assigning level control to the BOP would be correct and scram mitigating actions being a plausible misconception with time critical operator actions.
This answer is plausible if ATWS conditions were present, assigning level control to the BOP would be correct and scram mitigating actions being a plausible misconception with time critical operator actions. The candidate who confuses reactor conditions following a scram and confuses time critical operator actions with division of operator responsibilities would select this answer.
The candidate who confuses reactor conditions following a scram and confuses time critical operator actions with division of operator responsibilities would select this answer.
C. This answer is incorrect due to the procedure not providing division of operator responsibilities. (One of four validators chose this answer due to a common misconception). This answer is plausible due to scram mitigating actions being a plausible misconception with time critical operator actions. The candidate who correctly identifies Attachments 1, 2, & 3 and confuses time critical operator actions with division of operator responsibilities would select this answer.
C. This answer is incorrect due to the procedure not providing division of operator responsibilities. (One of four validators chose this answer due to a common misconception). This answer is plausible due to scram mitigating actions being a plausible misconception with time critical operator actions. The candidate who correctly identifies Attachments 1, 2, & 3 and confuses time critical operator actions with division of operator responsibilities would select this answer.
Technical Reference(s):
Technical Reference(s):
Procedure 2.0.3 (Conduct of Operations), Rev. 87 Procedure 2.0.1.3 (Time Critical Operator Action Control and Maintenance), Rev. 02   Proposed references to be provided to applicants during examination:
Procedure 2.0.3 (Conduct of Operations), Rev. 87 Procedure 2.0.1.3 (Time Critical Operator Action Control and Maintenance), Rev. 02 Proposed references to be provided to applicants during examination:         NONE Learning Objective:
NONE Learning Objective:
INT032010400D0400 Describe the general sequence of events performed in the Reactor Scram section of procedure 2.1.5, Reactor Scram.
INT032010400D0400 Describe the general sequence of events performed in the Reactor Scram section of procedure 2.1.5, Reactor Scram.
Question Source:
Question Sourc e:   Bank #     Modified Bank #
Bank #
New X   Question Cognitive Level:
Modified Bank #
Memory or Fundamental Knowledge Comprehension or Analysis X   10 CFR Part 55 Content:
New X Question Cognitive Level:     Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content:         55.41 55.43 5 LOD 3 3
55.41     55.43 5 LOD 3 3 SRO Only - 10CFR55.43 b (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Requires coordination of procedure attachments and selection of procedure which provides this guidance. 4 5
6 Examination Outline Cross
-


==Reference:==
SRO Only - 10CFR55.43 b (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Requires coordination of procedure attachments and selection of procedure which provides this guidance.
4


Level  SRO Tier #    1 Group #  1 K/A # 295018AA2.03 Importance Rating
5 6
Examination Outline Cross-


===3.5 295018===
==Reference:==
Partial or Complete Loss of Component Cooling Water AA2 Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER:
AA2.03:  Cause for partial or complete loss


Question: 77 The plant is operating at 100% power when the following conditions occur:
Level                      SRO Tier #                    1 Group #                    1 K/A #              295018AA2.03 Importance Rating          3.5 295018 Partial or Complete Loss of Component Cooling Water AA2      Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER:
SW Pressure on both Divisions has risen but is still in the green band.
AA2.03: Cause for partial or complete loss Question: 77 The plant is operating at 100% power when the following conditions occur:
REC system pressure is steady and is in the gree n band. REC Surge Tank Level High alarms.
* SW Pressure on both Divisions has risen but is still in the green band.
RR MG Set Oil Temperatures are rising.
* REC system pressure is steady and is in the green band.
Drywell temperature and pressure are rising.
* REC Surge Tank Level High alarms.
RWCU F/D Inlet Temp High alarms.
* RR MG Set Oil Temperatures are rising.
* Drywell temperature and pressure are rising.
* RWCU F/D Inlet Temp High alarms.
Which one of the following identifies the cause of these conditions and the required action to correct the problem?
Which one of the following identifies the cause of these conditions and the required action to correct the problem?
The REC Heat Exchanger-A. SW Outlet valve has closed, shift REC heat exchangers IAW 5.2REC (Loss of REC)
The REC Heat Exchanger A. SW Outlet valve has closed, shift REC heat exchangers IAW 5.2REC (Loss of REC).
. B. REC Outlet valve has closed, shift REC heat exchangers IAW 5.2REC (Loss of REC)
B. REC Outlet valve has closed, shift REC heat exchangers IAW 5.2REC (Loss of REC).
. C. SW Outlet valve has closed, shift REC heat exchangers IAW 5.2SW (Service Water Casualties
C. SW Outlet valve has closed, shift REC heat exchangers IAW 5.2SW (Service Water Casualties).
).
D. REC Outlet valve has closed, shift REC heat exchangers IAW 5.2SW (Service Water Casualties).
D. REC Outlet valve has closed, shift REC heat exchangers IAW 5.2SW (Service Water Casualties
Answer:
). Answer:     A. SW Outlet valve has closed, shift REC heat exchangers IAW 5.2REC (Loss of REC)
A. SW Outlet valve has closed, shift REC heat exchangers IAW 5.2REC (Loss of REC).
. Explanation:
Explanation:
7 A loss of SW flow to the REC Heat Exchanger recovery is covered in both 5.2REC and 5.2SW (assumes loss of SW pumps or piping). If the pressure of the service water system lowers to < 38 psig the system will isolate non
7
-critical loads. The subsequent steps will have the Operators place the other loops REC Heat Exchanger in service. For the given condition, SW Pressure rising indicates that there was some restriction in the flow path. 5.2SW shift is to use a good SW loop vs. 5.2REC due to REC cooling issues. The actions for shifting REC HXs in 5.2SW allow for bypassing Group 6 signal to SW
 
-MO-650 (REC HX A SERVICE OUTLET), opening REC-19 & 21 (REC HX INLETs), and transferring REC
A loss of SW flow to the REC Heat Exchanger recovery is covered in both 5.2REC and 5.2SW (assumes loss of SW pumps or piping). If the pressure of the service water system lowers to <
-TIC-451B to MANUAL. These steps differ from the steps in 5.2REC due to shifting heat exchangers for different reasons and would not be appropriate (see highlighted differences provided).
38 psig the system will isolate non-critical loads. The subsequent steps will have the Operators place the other loops REC Heat Exchanger in service. For the given condition, SW Pressure rising indicates that there was some restriction in the flow path. 5.2SW shift is to use a good SW loop vs. 5.2REC due to REC cooling issues. The actions for shifting REC HXs in 5.2SW allow for bypassing Group 6 signal to SW-MO-650 (REC HX A SERVICE OUTLET), opening REC-19 & 21 (REC HX INLETs), and transferring REC-TIC-451B to MANUAL. These steps differ from the steps in 5.2REC due to shifting heat exchangers for different reasons and would not be appropriate (see highlighted differences provided).
Distracters:
Distracters:
B. This answer is incorrect because the REC outlet valve closing will not cause SW pressure to rise and 5.2SW not being the procedure utilized to shift heat exchangers under the provided conditions. This answer is plausible because closure of this valve would provide the other indications and if the stem were changed to reflect REC pressure rising would be correct and 5.2SW provides guidance to shift (use a good SW loop vs. 5.2REC due to REC cooling issues). The candidate who confuses indications provided and which procedure provides the correct guidance would select this answer.
B. This answer is incorrect because the REC outlet valve closing will not cause SW pressure to rise and 5.2SW not being the procedure utilized to shift heat exchangers under the provided conditions. This answer is plausible because closure of this valve would provide the other indications and if the stem were changed to reflect REC pressure rising would be correct and 5.2SW provides guidance to shift (use a good SW loop vs. 5.2REC due to REC cooling issues). The candidate who confuses indications provided and which procedure provides the correct guidance would select this answer.
C. This answer is incorrect because 5.2SW is not the procedure utilized to shift heat exchangers under the provided conditions. This answer is plausible because 5.2SW provides guidance to shift (use a good SW loop vs. 5.2REC due to REC cooling issues). The candidate who correctly identifies the cause and confuses which procedure provides the correct guidance would select this answer.
C. This answer is incorrect because 5.2SW is not the procedure utilized to shift heat exchangers under the provided conditions. This answer is plausible because 5.2SW provides guidance to shift (use a good SW loop vs. 5.2REC due to REC cooling issues).
The candidate who correctly identifies the cause and confuses which procedure provides the correct guidance would select this answer.
D. This answer is incorrect because the REC outlet valve closing will not cause SW pressure to rise. This answer is plausible because closure of this valve would provide the other indications and if the stem were changed to reflect REC pressure rising would be correct along. The candidate who confuses indications provided and correctly identifies the procedure providing the correct guidance would select this answer.
D. This answer is incorrect because the REC outlet valve closing will not cause SW pressure to rise. This answer is plausible because closure of this valve would provide the other indications and if the stem were changed to reflect REC pressure rising would be correct along. The candidate who confuses indications provided and correctly identifies the procedure providing the correct guidance would select this answer.
Technical Reference(s):
Technical Reference(s):
Emergency Procedure 5.2REC, Loss of REC, Rev. 16 Emergency Procedure 5.2SW, Service Water Casualties, Rev. 24.
Emergency Procedure 5.2REC, Loss of REC, Rev. 16 Emergency Procedure 5.2SW, Service Water Casualties, Rev. 24.
Proposed references to be provided to applicants during examination:
Proposed references to be provided to applicants during examination:         None Learning Objective:
None Learning Objective:
INT0320126L0L0100 Given plant condition(s), determine from memory the appropriate Abnormal/Emergency Procedure(s) to be utilized to mitigate the event(s).
INT0320126L0L0100 Given plant condition(s), determine from memory the appropriate Abnormal/Emergency Procedure(s) to be utilized to mitigate the event(s).
Question Source:
Question Source:
Bank #     Modified Bank #
Bank #
X (See attached)
Modified Bank #     X (See attached)
New     Question History Last NRC Exam: 2011 Question 78 Question Cognitive Level:
New Question History               Last NRC Exam: 2011 Question 78 Question Cognitive Level:     Memory or Fundamental Knowledge X 8
Memory or Fundamental Knowledge X 8 Comprehension or Analysis     10 CFR Part 55 Content:
 
55.41     55.43 5 LOD 3 SRO Only - 10CFR55.43 b (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Comprehension or Analysis 10 CFR Part 55 Content:       55.41 55.43 5 LOD 3 SRO Only - 10CFR55.43 b (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Requires assessment of plant conditions and selection of procedure to mitigate the conditions.
Requires assessment of plant conditions and selection of procedure to mitigate the conditions.
QUESTION: S 3 78 The plant is operating at 100% power when the following conditions occur:
QUESTION: S 3         78 The plant is operating at 100% power when the following conditions occur:
SW Pressure on both Divisions have risen but are still in the green band REC system pressure is steady and is in the green band REC Surge Tank Level High alarms RR MG Set Oil Temperatures are risin g  Drywell temperature and pressure are rising RWCU F/D Inlet Temp High alarms
* SW Pressure on both Divisions have risen but are still in the green band
 
* REC system pressure is steady and is in the green band
What is the cause and which procedure should be entered to correct the problem?
* REC Surge Tank Level High alarms
: a. REC Heat Exchanger SW Outlet valve failed closed; enter 5.2REC to correct the problem. b. Service water non
* RR MG Set Oil Temperatures are rising
-critical loop isolated; enter 5.2SW to correct the problem.
* Drywell temperature and pressure are rising
* RWCU F/D Inlet Temp High alarms What is the cause and which procedure should be entered to correct the problem?
: a. REC Heat Exchanger SW Outlet valve failed closed; enter 5.2REC to correct the problem.
: b. Service water non-critical loop isolated; enter 5.2SW to correct the problem.
: c. REC system has developed a leak; enter 5.2REC to correct the problem.
: c. REC system has developed a leak; enter 5.2REC to correct the problem.
: d. Service water pump tripped; enter 5.2SW to correct the problem.
: d. Service water pump tripped; enter 5.2SW to correct the problem.
ANSWER:        S3      78
: a. REC Heat Exchanger SW Outlet valve failed closed; enter 5.2REC to correct the problem.
9


ANSWER:  S 3 78    a. REC Heat Exchanger SW Outlet valve failed closed; enter 5.2REC to correct the problem. 9 10 11 12 13 14 15 Examination Outline Cross
10 11 12 13 14 15 Examination Outline Cross-
-


==Reference:==
==Reference:==


Level SRO Tier #   1 Group # 1 K/A # 295019G2.4.4 Importance Rating
Level                     SRO Tier #                     1 Group #                   1 K/A #               295019G2.4.4 Importance Rating         4.7 295019 Partial or Complete Loss of Instrument Air 2.4.4     Ability to recognize abnormal indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures.
 
Question: 78 An air leak occurs with the plant operating at rated power causing Instrument Air (IA) pressure to lower.
===4.7 295019===
Partial or Complete Loss of Instrument Air
 
====2.4.4 Ability====
to recognize abnormal indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures.
Question: 78   An air leak occurs with the plant operating at rated power causing Instrument Air (IA) pressure to lower.
At what pressure is 5.2AIR, Attachment 2 (IA Pressure Loss) FIRST required to be implemented?
At what pressure is 5.2AIR, Attachment 2 (IA Pressure Loss) FIRST required to be implemented?
 
A. 90 psig B. 85 psig C. 75 psig D. 60 psig Answer:
A. 90 psig B. 85 psig C. 75 psig D. 60 psig Answer: C. 75 psig Explanation:
C. 75 psig Explanation:
The Control Room Operator must recognize the Emergency Procedure entry due to abnormal system parameter indications. The Control Room Supervisor must evaluate plant conditions and determine when to implement attachments of the procedure. 5.2AIR entry is required if SA or IA pressure is below green band and does not recover back into green band. Per 5.2AIR Subsequent Operator Actions procedure section, when Instrument Air header pressure lowers below 77 psig, Attachment 2 entry is required and the attachment provides instructions that are performed when system pressure is considered to be too low to support continued operation.
The Control Room Operator must recognize the Emergency Procedure entry due to abnormal system parameter indications. The Control Room Supervisor must evaluate plant conditions and determine when to implement attachments of the procedure. 5.2AIR entry is required if SA or IA pressure is below green band and does not recover back into green band. Per 5.2AIR Subsequent Operator Actions procedure section, when Instrument Air header pressure lowers below 77 psig, Attachment 2 entry is required and the attachment provides instructions that are performed when system pressure is considered to be too low to support continued operation.
Distracters:
Distracters:
A. This option is incorrect because Attachment 2 does not require entry at this high a pressure. This option is plausible because this is the pressure that starts the second standby air compressor. At this pressure all 3 air compressors are operating loaded. The candidate who believes this pressure is the pressure requiring Attachment 2 entry would select this option. B. This option is incorrect because 85 psig does not require Attachment 2 performance. This option is plausible because 5.2AIR subsequent operator actions do require actions to be taken if this pressure is reached and 85 psig is the setpoint for SA low pressure annunciator. The candidate who recalls the 85 psig pressure but does not recognize that Attachment 2 entry is not required would select this option.
A. This option is incorrect because Attachment 2 does not require entry at this high a pressure.
16 D. This option is incorrect because 60 psig is significantly below the FIRST pressure requiring Attachment 2 performance. This pressure is plausible because this is the pressure which triggers low air pressure alarms in the main control room. The candidate who recalls the 60 psig pressure being related to low air pressure alarm and recognizes 60 psig is abnormally low would select this option.
This option is plausible because this is the pressure that starts the second standby air compressor. At this pressure all 3 air compressors are operating loaded. The candidate who believes this pressure is the pressure requiring Attachment 2 entry would select this option.
B. This option is incorrect because 85 psig does not require Attachment 2 performance. This option is plausible because 5.2AIR subsequent operator actions do require actions to be taken if this pressure is reached and 85 psig is the setpoint for SA low pressure annunciator.
The candidate who recalls the 85 psig pressure but does not recognize that Attachment 2 entry is not required would select this option.
16
 
D. This option is incorrect because 60 psig is significantly below the FIRST pressure requiring Attachment 2 performance. This pressure is plausible because this is the pressure which triggers low air pressure alarms in the main control room. The candidate who recalls the 60 psig pressure being related to low air pressure alarm and recognizes 60 psig is abnormally low would select this option.
Technical Reference(s):
Technical Reference(s):
Procedure 5.2Air (Loss of Instrument Air), Rev. 19.
Procedure 5.2Air (Loss of Instrument Air), Rev. 19.
Proposed references to be provided to applicants during examination:
Proposed references to be provided to applicants during examination:         NONE Learning Objective:
NONE Learning Objective:
COR0011702001070A Given a specific Plant Air system malfunction, determine the effect on any of the following: Plant Operation Question Source:
COR0011702001070A Given a specific Plant Air system malfunction, determine the effect on any of the following:   Plant Operation Question Source:
Bank #
Bank #     Modified Bank #
Modified Bank #
New X   Question Cognitive Level:
New X Question Cognitive Level:     Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content:         55.41 10 55.43 2 55.45 6 Difficulty: 3 SRO Only - 10CFR55.43 b (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Memory or Fundamental Knowledge Comprehension or Analysis X   10 CFR Part 55 Content:
55.41 10     55.43 2     55.45 6 Difficulty: 3 SRO Only - 10CFR55.43 b (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
This meets SRO ONLY due to requiring knowledge of when to implement attachments associated with emergency procedures.
This meets SRO ONLY due to requiring knowledge of when to implement attachments associated with emergency procedures.
17 18 19 20 21 
17


22 Examination Outline Cross
18 19 20 21 22 Examination Outline Cross-
-


==Reference:==
==Reference:==


Level SRO Tier #   1 Group # 1 K/A # 295023A2.05 Importance Rating
Level                   SRO Tier #                   1 Group #                 1 K/A #             295023A2.05 Importance Rating       4.6 295023 Refueling Accidents AA2       Ability to determine and/or interpret the following as they apply to REFUELING ACCIDENTS:
 
A2.05 Entry conditions of emergency plan Question: 79 The plant is in MODE 5 with refueling in progress when an irradiated fuel bundle is dropped over the core.
===4.6 295023===
The Refueling Floor ARM (RA-1) is 5.5 x 104 mR/hr and rising.
Refueling Accidents AA2 Ability to determine and/or interpret the following as they apply to REFUELING ACCIDENTS:
A2.05 Entry conditions of emergency plan
 
Question: 79 The plant is in MODE 5 with refueling in progress when an irradiated fuel bundle is dropped over the core.
The Refueling Floor ARM (RA
-1) is 5.5 x 10 4 mR/hr and rising.
The ERP Kaman is indicating 1.80E +06 &#xb5;Ci/sec and rising.
The ERP Kaman is indicating 1.80E +06 &#xb5;Ci/sec and rising.
  (1) What is the current MINIMUM required Emergency Classification?
(1) What is the current MINIMUM required Emergency Classification?
(2) What ERP Kaman reading requires escalation to the next higher classification?
(2) What ERP Kaman reading requires escalation to the next higher classification?
A. (1) Unusual Event B.   (1) Unusual Event
A. (1) Unusual Event (2) 3.70E+06 Ci/sec B. (1) Unusual Event (2) 3.70E+07 Ci/sec C (1) Alert (2) 3.70E+06 Ci/sec D. (1) Alert (2) 3.70E+07 Ci/sec Answer:
 
D. (1) Alert (2) 3.70E+07 Ci/sec Explanation:
C   (1) Alert D.   (1) Alert
 
Answer: D. (1) Alert (2)
Explanation:
This question was on the 2011 CNS NRC Written exam as Question #77 and included the appropriate EAL handout. An Alert declaration is required if irradiated fuel is damaged and refuel floor radiation levels exceed 50 R/hr. If refuel floor ARM is rising combined with reactor cavity or spent fuel pool level lowering, the threshold for an Unusual Event declaration is met.
This question was on the 2011 CNS NRC Written exam as Question #77 and included the appropriate EAL handout. An Alert declaration is required if irradiated fuel is damaged and refuel floor radiation levels exceed 50 R/hr. If refuel floor ARM is rising combined with reactor cavity or spent fuel pool level lowering, the threshold for an Unusual Event declaration is met.
The ERP Kaman indication is also a NOUE. The Site Area Emergency for release rate from the ERP is 3.50E+07 2 for the EAL Category A, states not to wait for the 15 minute time to elapse before declaring if the start time is unknown. No times were given in the question so this requirement is met.
The ERP Kaman indication is also a NOUE. The Site Area Emergency for release rate from the ERP is 3.50E+07 Ci/sec. With the information given, an Alert is required to be declared. Note 2 for the EAL Category A, states not to wait for the 15 minute time to elapse before declaring if the start time is unknown. No times were given in the question so this requirement is met.
23 Distracters:
23
A. This answer is incorrect because an Alert is required. A UE would be correct if radiation level was provided with a lowering SFP level. All other SAE values in Table A
-1 are +06 is greater than this value, it is possible to consider the threshold being exceeded. The candidate who chooses an incorrect release point would select this answer. This answer is plausible because the UE threshold is met.
B. This answer is incorrect because an Alert is required. A UE would be correct if radiation level was provided with a lowering SFP level. 3.70E+07 is correct for this release point to require a higher EAL classification. The candidate who misses the damaged fuel EAL woul d select this answer. This answer is plausible because the UE threshold is met.


C. This answer is incorrect because the listed radiation release is too low for escalation to the next higher EAL. All other SAE values in Table A
Distracters:
-1 are established at 3.70Since 3.7E+06 is greater than this value, it is possible to consider the threshold being exceeded. If the candidate misreads the table, then this answer would be chosen. This answer is plausible because the release rates are contained in Table A-1. Technical Reference(s):
A. This answer is incorrect because an Alert is required. A UE would be correct if radiation level was provided with a lowering SFP level. All other SAE values in Table A-1 are established at 3.50E+06 Ci/sec. Since 3.70E+06 is greater than this value, it is possible to consider the threshold being exceeded. The candidate who chooses an incorrect release point would select this answer. This answer is plausible because the UE threshold is met.
B. This answer is incorrect because an Alert is required. A UE would be correct if radiation level was provided with a lowering SFP level. 3.70E+07 is correct for this release point to require a higher EAL classification. The candidate who misses the damaged fuel EAL would select this answer. This answer is plausible because the UE threshold is met.
C. This answer is incorrect because the listed radiation release is too low for escalation to the next higher EAL. All other SAE values in Table A-1 are established at 3.70E+06 Ci/sec.
Since 3.7E+06 is greater than this value, it is possible to consider the threshold being exceeded. If the candidate misreads the table, then this answer would be chosen. This answer is plausible because the release rates are contained in Table A-1.
Technical Reference(s):
Procedure EPIP 5.7.1 Attachment 4, (Emergency Action Level Matrix), Rev. 11 Procedure EPIP 5.7.1 (Emergency Classification), Rev. 50.
Procedure EPIP 5.7.1 Attachment 4, (Emergency Action Level Matrix), Rev. 11 Procedure EPIP 5.7.1 (Emergency Classification), Rev. 50.
Proposed references to be provided to applicants during examination:
Proposed references to be provided to applicants during examination:             EAL Matrix Category A Learning Objective:
EAL Matrix Category A Learning Objective:
GEN0030401C0C050E Concerning event classification: Given a copy of EPIP 5.7.1 and hypothetical abnormal plant symptoms, indications, or events, determine any and all EALs which have been exceeded and specify the appropriate emergency classification.
GEN0030401C0C050E Concerning event classification: Given a copy of EPIP 5.7.1 and hypothetical abnormal plant symptoms, indications, or events, determine any and all EALs which have been exceeded and specify the appropriate emergency classification.
Question Source:
Question Source:
Bank #     Modified Bank # 19335 New     Question History:
Bank #
2011 NRC exam Question Cognitive Level:
Modified Bank # 19335 New Question History:           2011 NRC exam Question Cognitive Level:     Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content:         55.41 55.43 5 Difficulty: 3 SRO Only - 10CFR55.43 b (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Memory or Fundamental Knowledge Comprehension or Analysis X   10 CFR Part 55 Content:
55.41     55.43 5 Difficulty: 3 SRO Only - 10CFR55.43 b (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
The SRO is responsible for EAL classification declarations.
The SRO is responsible for EAL classification declarations.
24 REFUEL AREA HIGH RAD    PANEL/WINDOW:
24
9-3-1/A-10 SETPOINT 1. (1448) RX BLDG FUEL POOL (HR) AREA RAD HIGH at 500 mR/Hr 2. (1449) RX BLDG FUEL POOL (LR) AREA RAD HIGH at 10 mR/Hr CIC 9-3-1/A-10 1. RMA-RA-1  2. RMA-RA-2  25 26 27 28 Examination Outline Cross
 
-
REFUEL AREA       PANEL/WINDOW:
HIGH RAD          9-3-1/A-10 SETPOINT                             CIC        9-3-1/A-10
: 1. (1448) RX BLDG FUEL POOL (HR) AREA 1. RMA-RA-1 RAD HIGH at 500 mR/Hr
: 2. (1449) RX BLDG FUEL POOL (LR) AREA 2. RMA-RA-2 RAD HIGH at 10 mR/Hr 25
 
26 27 28 Examination Outline Cross-


==Reference:==
==Reference:==


Level SRO Tier #   1 Group # 1 K/A # 295028G2.4.11 Importance Rating
Level                       SRO Tier #                     1 Group #                     1 K/A #               295028G2.4.11 Importance Rating           4.2 295028 High Drywell Temperature 2.4.11 Knowledge of abnormal condition procedures.
 
Question: 80 While operating at 80% power, the following indications are observed:
===4.2 295028===
* Drywell Temperature is 148&deg;F and rising 1&deg;F/5 min.
High Drywell Temperature 2.4.11 Knowledge of abnormal condition procedures.
* Drywell Pressure is 0.35 psig and rising slowly (0.05 psig/min).
Question: 80 While operating at 80% power, the following indications are observed:
Drywell Temperature is 148&deg;F and rising 1&deg;F/5 min.
Drywell Pressure is 0.35 psig and rising slowly (0.05 psig/min).
What action is required to mitigate these conditions?
What action is required to mitigate these conditions?
Ensure all available DW FCU control switches are in-  A. RUN IAW 2.4PC (Primary Containment Control).
Ensure all available DW FCU control switches are in A. RUN IAW 2.4PC (Primary Containment Control).
B. RUN IAW H-1/A-2 (Drywell Zone 1 High Temp).
B. RUN IAW H-1/A-2 (Drywell Zone 1 High Temp).
C. OVERRIDE IAW H
C. OVERRIDE IAW H-1/A-2 (Drywell Zone 1 High Temp).
-1/A-2 (Drywell Zone 1 High Temp).
 
D. OVERRIDE IAW 2.4PC (Primary Containment Control).
D. OVERRIDE IAW 2.4PC (Primary Containment Control).
Answer: A. RUN IAW 2.4PC (Primary Containment Control). Explanation: Requires knowledge of abnormal procedure 2.4PC supplemental actions because DW temperature and pressure are rising. 2.4PC requires verification of drywell FCU control switches in RUN vs placing switches in OVERRIDE. The EOPs are the only procedures that allow placing the switches in OVERRIDE. The Drywell Zone 1 alarm provides guidance to check the FCUs are operating and ensuring the cooling to the FCU cooling coils is correctly aligned. The CRS must know that with the temperature and pressure rising if the FCUs are performing their function so it is appropriate to ensure their control switches are in RUN.
Answer:
A. RUN IAW 2.4PC (Primary Containment Control).
Explanation: Requires knowledge of abnormal procedure 2.4PC supplemental actions because DW temperature and pressure are rising. 2.4PC requires verification of drywell FCU control switches in RUN vs placing switches in OVERRIDE. The EOPs are the only procedures that allow placing the switches in OVERRIDE. The Drywell Zone 1 alarm provides guidance to check the FCUs are operating and ensuring the cooling to the FCU cooling coils is correctly aligned. The CRS must know that with the temperature and pressure rising if the FCUs are performing their function so it is appropriate to ensure their control switches are in RUN.
29


29 Distracters:
Distracters:
B. This answer is incorrect because the alarm procedure does not provide direction to ensure all available FCU control switches are in RUN. This answer is plausible because the alarm procedure does contain a step to check they are operating. The candidate who believes the procedure guidance is adequate for the given conditions would select this answer.
B. This answer is incorrect because the alarm procedure does not provide direction to ensure all available FCU control switches are in RUN. This answer is plausible because the alarm procedure does contain a step to check they are operating. The candidate who believes the procedure guidance is adequate for the given conditions would select this answer.
C. This answer is incorrect because the procedure is wrong and the action is incorrect. This answer is plausible because the FCU control switches are taken to run but only after drywell temper or pressure rise to EOP 3A levels. The candidate who recalls placing the FCU switches in OVERRIDE is a mitigating strategy but does not recognize that the EOPs are not entered would select this answer.
C. This answer is incorrect because the procedure is wrong and the action is incorrect. This answer is plausible because the FCU control switches are taken to run but only after drywell temper or pressure rise to EOP 3A levels. The candidate who recalls placing the FCU switches in OVERRIDE is a mitigating strategy but does not recognize that the EOPs are not entered would select this answer.
Line 3,438: Line 3,193:
Procedure 2.4PC (Primary Containment Control), Rev. 17.
Procedure 2.4PC (Primary Containment Control), Rev. 17.
ARP 2.3_H-1, Rev. 10.
ARP 2.3_H-1, Rev. 10.
Proposed references to be provided to applicants during examination:
Proposed references to be provided to applicants during examination:         None Learning Objective:
None Learning Objective:
INT0320128K0K0100 Given plant condition(s), and the applicable Abnormal/Emergency Procedure, determine the correct subsequent actions required to mitigate the event(s).
INT0320128K0K0100 Given plant condition(s), and the applicable Abnormal/Emergency Procedure, determine the correct subsequent actions required to mitigate the event(s).
Question Source:
Question Source:
Bank #     Modified Bank #
Bank #
New X Question Cognitive Level:
Modified Bank #
Memory or Fundamental Knowledg e    Comprehension or Analysis X   10 CFR Part 55 Content:
New X Question Cognitive Level:     Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content:         55.41 10 55.43 5 55.45 13 Difficulty: 3 SRO Only - 10CFR55.43 b (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
55.41 10     55.43 5     55.45 13 Difficulty: 3 SRO Only - 10CFR55.43 b (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
This meets SRO ONLY due to requiring knowledge supplemental AOP actions and when to implement the actions associated with abnormal procedures.
This meets SRO ONLY due to requiring knowledge supplemental AOP actions and when to implement the actions associated with abnormal procedures.
30 31 
30


32 33 Examination Outline Cross
31 32 33 Examination Outline Cross-
-


==Reference:==
==Reference:==


Level SRO Tier #   1 Group # 1 K/A # 295005EA2.04 Importance Rating
Level                     SRO Tier #                     1 Group #                   1 K/A #               295005EA2.04 Importance Rating         3.3 295005 Main Turbine Generator Trip EA2       Ability to determine and/or interpret the following as they apply to MAIN TURBINE GENERATOR TRIP:
 
EA2.08 Electrical distribution status Question: 81 The plant is operating at rated power with the Startup Transformer out of service for maintenance.
===3.3 295005===
The main generator trips on load reject causing ALL 4160 VAC Buses to de-energize.
Main Turbine Generator Trip EA2 Ability to determine and/or interpret the following as they apply to MAIN TURBINE GENERATOR TRIP:
(1) What power source automatically reenergizes the Critical Busses?
EA2.08 Electrical distribution status
(2) What action is required IAW 5.3EMPWR (Emergency Power During MODES 1, 2, or 3)?
 
A. (1) Diesel Generators (2) Direct DCC to perform the CNS-Black Plant Procedure.
Question: 81 The plant is operating at rated power with the Startup Transformer out of service for maintenance.
B. (1) Diesel Generators (2) Coordinate with DCC to backfeed through the Normal Transformer.
The main generator trips on load reject causing ALL 4160 VAC Buses to de
C. (1) Emergency Transformer (2) Direct DCC to perform the CNS-Black Plant Procedure.
-energize. (1) What power source automatically reenergizes the Critical Busses?
D. (1) Emergency Transformer (2) Coordinate with DCC to backfeed through the Normal Transformer.
(2) What action is required IAW 5.3EMPWR (Emergency Power During MODES 1, 2, or 3)?
Answer:
A. (1) Diesel Generators (2) Direct DCC to perform the CNS
C. (1) Emergency Transformer (2) Direct DCC to perform the CNS-Black Plant Procedure.
-Black Plant Procedure.
B. (1) Diesel Generators (2) Coordinate with DCC to backfeed through the Normal Transformer.
C. (1) Emergency Transformer (2) Direct DCC to perform the CNS
-Black Plant Procedure.
D. (1) Emergency Transformer (2) Coordinate with DCC to backfeed through the Normal Transformer.
Answer: C. (1) Emergency Transformer (2) Direct DCC to perform the CNS-Black Plant Procedure.
Explanation:
Explanation:
With the Startup Station Service Transformer out of service, the Normal Station Transformer is providing power to Critical Buses 1F and 1G through 4160V Buses 1A and 1B. When the main generator trips, the Normal Station Service Transformer becomes de
With the Startup Station Service Transformer out of service, the Normal Station Transformer is providing power to Critical Buses 1F and 1G through 4160V Buses 1A and 1B. When the main generator trips, the Normal Station Service Transformer becomes de-energized. Buses 4160 1A and 1B become de-energized which for one second de-energizes 1F and 1G. The Emergency Station Service Transformer repowers 1F and 1G directly. The Diesel Generators receive a start signal because of the short-lived (1 second) de-energization of 4160V buses1F and 1G. With all 4160V buses are de-energized for a short period of time a Station Blackout condition exists. However, due to the short lived duration, the proper procedure to enter is 5.3EMPWR. A common misconception is that only procedure 5.3SBO has guidance for directing DCC to enter the CNS Black Plant procedure. Procedure 5.3EMPWR Attachment 3 has been recently revised (within the last two years) to direct DCC to enter the CNS Black Plant procedure.
-energized. Buses 4160 1A and 1B become de
Distracters:
-energized which for one second de
A. This answer is incorrect because the Critical Buses are energized from the Emergency Service Station Transformer. This answer is plausible if the order in which emergency 34
-energizes 1F and 1G. The Emergency Station Service Transformer repowers 1F and 1G directly. The Diesel Generators receive a start signal because of the short
-lived (1 second) de
-energization of 4160V buses1F and 1G. With all 4160V buses are de
-energized for a short period of time a Station Blackout condition exists. However, due to the short lived duration, the proper procedure to enter is 5.3EMPWR. A common misconception is that only procedure 5.3SBO has guidance for directing DCC to enter the CNS Black Plant procedure. Procedure 5.3EMPWR Attachment 3 has been recently revised (within the last two years) to direct DCC to enter the CNS Black Plant procedure.


Distracters:
power supplies energize the Critical Buses is confused or if the stem were changed to reflect a Loss of Offsite Power (LOOP - the emergency transformer is unavailable). The candidate who confuses the order in which emergency power supplies energize the Critical Buses and correctly identifies 5.3EMPWR directs the DCC to enter the Black Plant procedure would select this option.
A. This answer is incorrect because the Critical Buses are energized from the Emergency Service Station Transformer. This answer is plausible if the order in which emergency 34 power supplies energize the Critical Buses is confused or if the stem were changed to reflect a Loss of Offsite Power (LOOP  
B. This answer is incorrect because the Critical Buses are energized from the Emergency Service Station Transformer and backfeed is not directed in 5.3 EMPWR. This answer is plausible if the order in which emergency power supplies energize the Critical Buses is confused or if the stem were changed to reflect a Loss of Offsite Power (LOOP - the emergency transformer is unavailable) AND due to the Normal transformer being available for backfeed (Off Site power remains available). The candidate who confuses the order in which emergency power supplies energize the Critical Buses and does not know backfeed through the Normal transformer is only directed in 5.3SBO would select this answer.
- the emergency transformer is unavailable). The candidate who confuses the order in which emergency power supplies energize the Critical Buses and correctly identifies 5.3EMPWR directs the DCC to enter the Black Plant procedure would select this option.
B. This answer is incorrect because the Critical Buses are energized from the Emergency Service Station Transformer and backfeed is not directed in 5.3 EMPWR. This answer i s plausible if the order in which emergency power supplies energize the Critical Buses is confused or if the stem were changed to reflect a Loss of Offsite Power (LOOP  
- the emergency transformer is unavailable) AND due to the Normal transformer being available for backfeed (Off Site power remains available). The candidate who confuses the order in which emergency power supplies energize the Critical Buses and does not know backfeed through the Normal transformer is only directed in 5.3SBO would select this answer.
D. This answer is incorrect because backfeed is not directed in 5.3 EMPWR. This answer is plausible to the Normal transformer being available for backfeed (Off Site power remains available). The candidate who correctly identifies the order in which emergency power supplies energize the Critical Buses and does not know backfeed through the Normal transformer is only directed in 5.3SBO would select this answer.
D. This answer is incorrect because backfeed is not directed in 5.3 EMPWR. This answer is plausible to the Normal transformer being available for backfeed (Off Site power remains available). The candidate who correctly identifies the order in which emergency power supplies energize the Critical Buses and does not know backfeed through the Normal transformer is only directed in 5.3SBO would select this answer.
Technical Reference(s):
Technical Reference(s):
Procedure 5.3EMPWR (Emergency Power During Modes 1, 2, or 3), Rev. 48.
Procedure 5.3EMPWR (Emergency Power During Modes 1, 2, or 3), Rev. 48.
Procedure 5.3SBO (Station Blackout) Rev. 33 Proposed references to be provided to applicants during examination:
Procedure 5.3SBO (Station Blackout) Rev. 33 Proposed references to be provided to applicants during examination:           NONE Learning Objective: INT0320126Q0Q0100, Given plant condition(s), and the applicable Abnormal/Emergency Procedure, determine the correct subsequent actions required to mitigate the event(s).
NONE Learning Objective:   INT0320126Q0Q0100, Given plant condition(s), and the applicable Abnormal/Emergency Procedure, determine the correct subsequent actions required to mitigate the event(s).
Question Source:
Question Source:
Bank #     Modified Bank #
Bank #
New X   Question Cognitive Level:
Modified Bank #
Memory or Fundamental Knowledge Comprehension or Analysis X   10 CFR Part 55 Content:
New X Question Cognitive Level:       Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content:         55.41 10 55.43 5 55.45 13 Difficulty: 3 SRO Only - 10 CFR 55.43(b)(5) - Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
55.4 1 10     55.43 5     55.45 13 Difficulty: 3 SRO Only - 10 CFR 55.43(b)(5)  
- Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Requires assessing plant conditions to determine if supplemental actions contained in Attachment 3 (Electrical Systems Guideline) of 5.3EMPWR (Emergency Power During Modes 1, 2, OR 3) are required.
Requires assessing plant conditions to determine if supplemental actions contained in Attachment 3 (Electrical Systems Guideline) of 5.3EMPWR (Emergency Power During Modes 1, 2, OR 3) are required.
35 36 37 38 Examination Outline Cross
35
-
 
36 37 38 Examination Outline Cross-


==Reference:==
==Reference:==


Level SRO Tier #   1 Group # 1 K/A # 600000G2.2.44 Importance Rating
Level                   SRO Tier #                 1 Group #                 1 K/A #             600000G2.2.44 Importance Rating       4.4 600000 Plant Fire On Site 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.
 
Question: 82 While the Fire Brigade is combatting a Temporary Trailer fire (south of the Training Building) with fire hoses, the following indications are observed on the Main Control Room Fire Protection Panel:
===4.4 600000===
G                        R  G                R            G                        R ELECTRIC FIRE          DIESEL FIRE                        ELECTRIC FIRE PUMP C                PUMP D                                PUMP E AUTO                  AUTO                                AUTO PULL TO PULL TO LOCK LOCK The operator then places the control switch for Fire Pump E to START and the pump starts.
Plant Fire On Site 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.
(1) What is the known MINIMUM pressure reached in the fire protection header prior to the operator taking the above action?
Question: 82   While the Fire Brigade is combatting a Temporary Trailer fire (south of the Training Building) with fire hoses, the following indications are observed on the Main Control Room Fire Protection Panel:    
(2) What is the operability status of the Fire Suppression Water System?
 
A. (1) 68 psig (2) Operable B. (1) 68 psig (2) Inoperable C. (1) 141 psig (2) Operable D. (1) 141 psig (2) Inoperable 39
The operator then places the control switch for Fire Pump E to START and the pump starts.
  (1) What is the known MINIMUM pressure reached in the fire protection header prior to the operator taking the above action?
(2) What is the operability status of the Fire Suppression Water System? A. (1) 68 psig (2) Operable
 
B. (1) 68 psig (2) Inoperable C. (1) 141 psig (2) Operable
 
D. (1) 141 psig


(2)  Inoperable ELECTRIC FIRE PUMP C AUTO PULL TO LOCK G R R DIESEL FIRE PUMP D AUTO G ELECTRIC FIRE PUMP E AUTO  PULL TO LO CK G R  39 Answer: B. (1) 68 psig (2) Inoperable Explanation:
Answer:
B. (1) 68 psig (2) Inoperable Explanation:
Requires candidate to determine Fire Protection header pressure which automatically starts ALL Fire Pumps and recognize failure of Fire Pump E to auto start (interpret control room indications). Based on the information given to the candidate, it is known the header pressure lowered at least to 68 psig which is the automatic start setpoint of Fire Pump C. 141 psig is the automatic start setpoint for Fire Pump D plus a 10 second time delay. With Fire Pump D running this choice is plausible. Second part requires knowledge of TRM bases (requires Fire Pump E to support operability) and TSR 3.11.2.13 which requires each pump to sequentially start based upon lowering system pressure. Fire Pump E failed to auto start on low system pressure and is therefore Inoperable even if manually started (operator action impact) from the Control Room.
Requires candidate to determine Fire Protection header pressure which automatically starts ALL Fire Pumps and recognize failure of Fire Pump E to auto start (interpret control room indications). Based on the information given to the candidate, it is known the header pressure lowered at least to 68 psig which is the automatic start setpoint of Fire Pump C. 141 psig is the automatic start setpoint for Fire Pump D plus a 10 second time delay. With Fire Pump D running this choice is plausible. Second part requires knowledge of TRM bases (requires Fire Pump E to support operability) and TSR 3.11.2.13 which requires each pump to sequentially start based upon lowering system pressure. Fire Pump E failed to auto start on low system pressure and is therefore Inoperable even if manually started (operator action impact) from the Control Room.
Distracters:
Distracters:
A. This answer is incorrect due to the Fire Suppression system being inoperable. This answer is plausible if the SR for this pump to auto start on low system pressure is not known or Fire Pumps D & E are required to support system operability. This is a plausible misconception because there are 2 pumps required and the C pump does not count. The candidate who knows fire pump auto start setpoints and does not know pump auto start on low system pressure is required for operability or which pumps are required for operability would choose this answer.
A. This answer is incorrect due to the Fire Suppression system being inoperable. This answer is plausible if the SR for this pump to auto start on low system pressure is not known or Fire Pumps D & E are required to support system operability. This is a plausible misconception because there are 2 pumps required and the C pump does not count. The candidate who knows fire pump auto start setpoints and does not know pump auto start on low system pressure is required for operability or which pumps are required for operability would choose this answer.
C. This answer is incorrect due to 141psig being above the auto start setpoint of Fire pump C and the Fire Suppression system being inoperable. This answer is plausible if the Fire pump auto start setpoint are confused (If the stem were changed to indicate Fire Pump C not running, 141 psig would be correct) and the SR for this pump to auto start on low system pressure is not known or Fire Pumps D & E are required to support system operability. This is a plausible misconception because there are 2 pumps required and the C pump does not count. The candidate who confuses fire pump auto start setpoints and does not know pump auto start on low system pressure is required for operability or which pumps are required for operability would choose this answer.
C. This answer is incorrect due to 141psig being above the auto start setpoint of Fire pump C and the Fire Suppression system being inoperable. This answer is plausible if the Fire pump auto start setpoint are confused (If the stem were changed to indicate Fire Pump C not running, 141 psig would be correct) and the SR for this pump to auto start on low system pressure is not known or Fire Pumps D & E are required to support system operability. This is a plausible misconception because there are 2 pumps required and the C pump does not count. The candidate who confuses fire pump auto start setpoints and does not know pump auto start on low system pressure is required for operability or which pumps are required for operability would choose this answer.
 
D. This answer is incorrect due to 141psig being above the auto start setpoint of Fire pump C.
D. This answer is incorrect due to 141psig being above the auto start setpoint of Fire pump C. This answer is plausible if the Fire pump auto start setpoint are confused (If the stem were changed to indicate Fire Pump C not running, 141 psig would be correct). The candidate who confuses fire pump auto start setpoints and knows pump auto start on low system pressure is required for operability and which pumps are required for operability would choose this answer.
This answer is plausible if the Fire pump auto start setpoint are confused (If the stem were changed to indicate Fire Pump C not running, 141 psig would be correct). The candidate who confuses fire pump auto start setpoints and knows pump auto start on low system pressure is required for operability and which pumps are required for operability would choose this answer.
 
Technical Reference(s):
Technical Reference(s):
Technical Requirements Manual (TRM), Rev. 8/27/2014.
Technical Requirements Manual (TRM), Rev. 8/27/2014.
Procedure 2.3_FP-4 (Fire Protection  
Procedure 2.3_FP-4 (Fire Protection - Annunciator 4), Rev. 11.
- Annunciator 4), Rev. 11.
Procedure 2.2.30 (Fire Protection System), Rev. 62.
Procedure 2.2.30 (Fire Protection System), Rev. 62.
Surveillance Procedure 6.FP.102 (Annual Testing of Fire Pumps), Rev. 33 Procedure 0.23 (CNS Fire Protection Plan), Rev. 71 Proposed references to be provided to applicants during examination:
Surveillance Procedure 6.FP.102 (Annual Testing of Fire Pumps), Rev. 33 Procedure 0.23 (CNS Fire Protection Plan), Rev. 71 Proposed references to be provided to applicants during examination:       NONE Learning Objective:
NONE Learning Objective:
40
40 COR00105020010200 Given condition(s) and/or parameters associated with the Fire Protection system, determine if related Technical Requirements Manual Limiting Condition for Operation are met.
 
COR00105020010200 Given condition(s) and/or parameters associated with the Fire Protection system, determine if related Technical Requirements Manual Limiting Condition for Operation are met.
Question Source:
Question Source:
Bank #     Modified Bank #   New X   Question Cognitive Level:
Bank #
Memory or Fundamental Knowledge Comprehension or Analysis X   10 CFR Part 55 Content:
Modified Bank #
55.41.4     55.43.2 55.45.12 Difficulty: 3 SRO Only - 10CFR55.43 b (2) Facility operating limitations in the TS and their bases.
New X Question Cognitive Level:           Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content:     55.41.4 55.43.2 55.45.12 Difficulty: 3 SRO Only - 10CFR55.43 b (2) Facility operating limitations in the TS and their bases.
Requires knowledge of Fire Suppression Water System surveillance requirements, bases and CNS Fire Protection Plan.
Requires knowledge of Fire Suppression Water System surveillance requirements, bases and CNS Fire Protection Plan.
41 42 43 44 45 Examination Outline Cross
41
-
 
42 43 44 45 Examination Outline Cross-


==Reference:==
==Reference:==


Level SRO Tier #   1 Group # 2 K/A # 295007A2.01 Importance Rating
Level                     SRO Tier #                     1 Group #                   2 K/A #               295007A2.01 Importance Rating         4.1 295007 High Reactor Pressure AA2       Ability to determine and/or interpret the following as they apply to HIGH REACTOR PRESSURE:
 
A2.01 Reactor pressure Question: 83 The plant is operating in Mode 1, End-of-Cycle with pressure being maintained at 1015 psig.
===4.1 295007===
A pressure adjustment is made at 0815 on September 1st.
High Reactor Pressure AA2 Ability to determine and/or interpret the following as they apply to HIGH REACTOR PRESSURE:
The RO observes STABLE pressure on the following 3 indicators at the specified time:
A2.01 Reactor pressure
Time 0817           RFC-PI-90A, RX PRESS is indicating 1015 psig.
 
Time 0818           RFC-PI-90C, RX PRESS is indicating 1020 psig.
Question: 83
Time 0819           RFC-PI-90B, RX PRESS is indicating 1025 psig.
 
The plant is operating in Mode 1, End
-of-Cycle with pressure being maintained at 1015 psig.
A pressure adjustment is made at 0815 on September 1 st. The RO observes STABLE pressure on the following 3 indicators at the specified time:
Time 0817 RFC-PI-90A, RX PRESS is indicating 1015 psig.
Time 0818 RFC-PI-90C, RX PRESS is indicating 1020 psig.
Time 0819 RFC-PI-90B, RX PRESS is indicating 1025 psig.
What is the LATEST time the reactor is required to be in MODE 3 IAW TS 3.4.10, Reactor Steam Dome Pressure if these conditions remain unchanged?
What is the LATEST time the reactor is required to be in MODE 3 IAW TS 3.4.10, Reactor Steam Dome Pressure if these conditions remain unchanged?
A. 2033 B. 2034 C. 2118 D. 2119 Answer: B. 2034 Explanation:
A. 2033 B. 2034 C. 2118 D. 2119 Answer:
RFC-PI-90A and RFC
B. 2034 Explanation:
-PI-90C are within specification limits. RFC
RFC-PI-90A and RFC-PI-90C are within specification limits. RFC-PI-90B is noted to be exceeding TS LCO 3.4.10 requirements at 0819. At this time, the CRS is notified and the CRS enters LCO 3.4.10 Condition A. Technical Specification LCO 3.4.10 requires the reactor steam dome pressure be  1020 psig when in Modes 1 and 2. Condition A requires the pressure to be restored within limits within 15 minutes. Condition B requires the unit to be in Mode 3 within 12 hours if Condition A cannot be met. 0819 + 15 minutes (Condition A) + 12 hours = 0834 +12 hours = 2034.
-PI-90B is noted to be exceeding TS LCO 3.4.10 requirements at 0819. At this time, the CRS is notified and the CRS enters LCO 3.4.10 Condition A. Technical Specification LCO 3.4.10 requires the reactor steam restored within limits within 15 minutes. Condition B requires the unit to be in Mode 3 within 12 hours if Condition A cannot be met. 0819 + 15 minutes (Condition A) + 12 hours = 0834 +12 hours = 2034.
Distracters:
Distracters:
A. This answer is incorrect because 2033 does not incorporate the 15 min Condition A completion time from the time 1020 psig was exceeded. This answer is plausible if 15 minutes is applied to the time 1020 psig is reached. The candidate that applies the 15 min completion time of Condition A from when pressure reaches 1020 psig would select this answer.
A.     This answer is incorrect because 2033 does not incorporate the 15 min Condition A completion time from the time 1020 psig was exceeded. This answer is plausible if 15 minutes is applied to the time 1020 psig is reached. The candidate that applies the 15 min completion time of Condition A from when pressure reaches 1020 psig would select this answer.
C. This answer is incorrect because 2118 does not incorporate the 15 min Condition A completion time from the time 1020 psig was exceeded. This answer is plausible due to a 1 hour completion time being applicable to other TSs and from memory can be easily 46 confused for Condition A completion time and LCO entry is confused with pressure at 1020 psig vs. >1020. The candidate that applies a 1 hour completion time of Condition A and confuses the actual LCO entry pressure would select this answer.
C.     This answer is incorrect because 2118 does not incorporate the 15 min Condition A completion time from the time 1020 psig was exceeded. This answer is plausible due to a 1 hour completion time being applicable to other TSs and from memory can be easily 46
D. This answer is incorrect because 2119 does not incorporate the 15 min Condition A completion time from the correct time 1020 psig was exceeded. This answer is plausible due to 1 hour completion time being applicable to other TSs and from memory can be easily confused for Condition A completion time. The candidate that applies a 1 hour completion time to Condition A and correctly identifies the actual LCO entry pressure would select this answer.
 
confused for Condition A completion time and LCO entry is confused with pressure at 1020 psig vs. >1020. The candidate that applies a 1 hour completion time of Condition A and confuses the actual LCO entry pressure would select this answer.
D.       This answer is incorrect because 2119 does not incorporate the 15 min Condition A completion time from the correct time 1020 psig was exceeded. This answer is plausible due to 1 hour completion time being applicable to other TSs and from memory can be easily confused for Condition A completion time. The candidate that applies a 1 hour completion time to Condition A and correctly identifies the actual LCO entry pressure would select this answer.
Technical Reference(s):
Technical Reference(s):
Technical Specifications LCO 3.4.10
Technical Specifications LCO 3.4.10 Proposed references to be provided to applicants during examination: LCO 3.4.10 with LCO Pressure removed and Condition A Completion Time blanked.
 
Proposed references to be provided to applicants during examination: LCO 3.4.10 with LCO Pressure removed and Condition A Completion Time blanked.
Learning Objective:
Learning Objective:
INT007-05-05, OPS Tech Specs 3.4 Reactor Coolant System (RCS)
INT007-05-05, OPS Tech Specs 3.4 Reactor Coolant System (RCS)
: 3. Given a set of plant conditions that constitutes noncompliance with a section 3.4 LCO, determine the ACTIONS that are required.
: 3. Given a set of plant conditions that constitutes noncompliance with a section 3.4 LCO, determine the ACTIONS that are required.
: 10. From memory, in MODES 1 and 2, state the actions required in one hour if the reactor steam dome pressure LCO is not met (LCO 3.4.10).
: 10. From memory, in MODES 1 and 2, state the actions required in one hour if the reactor steam dome pressure LCO is not met (LCO 3.4.10).
Question Source:
Question Source:
Bank #     Modified Bank #
Bank #
New X   Question Cognitive Level:
Modified Bank #
Memory or Fundamental Knowledge Comprehension or Analysis X   10 CFR Part 55 Content:
New X Question Cognitive Level:     Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content:       55.41.10 55.43.5 55.45.13 Difficulty: 3 SRO Only - 10CFR55.43 b (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
55.41.10     55.43.5 55.45.13 Difficulty: 3 SRO Only - 10CFR55.43 b (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Application of Required Actions (Section 3) IAW with rules of application requirements (Section 1).
Application of Required Actions (Section 3) IAW with rules of application requirements (Section 1). 47 48 49 50 51 52 Examination Outline Cross
47
-
 
48 49 50 51 52 Examination Outline Cross-


==Reference:==
==Reference:==


Level SRO Tier #   1 Group # 2 K/A # 295020G2.2.14 Importance Rating
Level                     SRO Tier #                     1 Group #                   2 K/A #               295020G2.2.14 Importance Rating         4.3 295020 Inadvertent Containment Isolation 2.2.14 Knowledge of the process for controlling equipment configuration or status.
 
Question: 84 A loss of Division 1 RPSPP1A power occurs while operating at power. Troubleshooting the problem is about to commence.
===4.3 295020===
Inadvertent Containment Isolation 2.2.14 Knowledge of the process for controlling equipment configuration or status.
Question: 84 A loss of Division 1 RPSPP1A power occurs while operating at power. Troubleshooting the problem is about to commence.
What is the response of the RWCU PCIV(s)?
What is the response of the RWCU PCIV(s)?
What procedure controls the configuration of the RWCU PCIV(s) during troubleshooting?
What procedure controls the configuration of the RWCU PCIV(s) during troubleshooting?
 
A. (1) ONLY RWCU-MO-15 (INBOARD ISOLATION VALVE) closes.
A. (1) ONLY RWCU
(2) 0.31(Equipment Status Control) is the controlling procedure.
-MO-15 (INBOARD ISOLATION VALVE) closes.
B. (1) ONLY RWCU-MO-15 (INBOARD ISOLATION VALVE) closes.
  (2) 0.31(Equipment Status Control) is the controlling procedure.
(2) 0-CNS-WM-102 (Work Implementation and Closeout) is the controlling procedure.
B. (1) ONLY RWCU-MO-15 (INBOARD ISOLATION VALVE) closes.
C. (1) BOTH RWCU-MO-15 and RWCU-MO-18 (OUTBOARD ISOLATION VALVE) close.
  (2) 0-CNS-WM-102 (Work Implementation and Closeout) is the controlling procedure.
(2) 0.31(Equipment Status Control) is the controlling procedure.
 
D. (1) BOTH RWCU-MO-15 and RWCU-MO-18 (OUTBOARD ISOLATION VALVE) close.
C. (1) BOTH RWCU
(2) 0-CNS-WM-102 (Work Implementation and Closeout) is the controlling procedure.
-MO-15 and RWCU
Answer:
-MO-18 (OUTBOARD ISOLATION VALVE) close.
C. (1) BOTH RWCU-MO-15 and RWCU-MO-18 (OUTBOARD ISOLATION VALVE) close.
  (2) 0.31(Equipment Status Control) is the controlling procedure.
(2) 0.31(Equipment Status Control) is the controlling procedure.
D. (1) BOTH RWCU
-MO-15 and RWCU
-MO-18 (OUTBOARD ISOLATION VALVE) close.
  (2) 0-CNS-WM-102 (Work Implementation and Closeout) is the controlling procedure.
Answer: C. (1) BOTH RWCU
-MO-15 and RWCU
-MO-18 (OUTBOARD ISOLATION VALVE) close. (2) 0.31(Equipment Status Control) is the controlling procedure.
Explanation:
Explanation:
The loss of Div 1 RPS causes both RWCU PCIVs (Inboard and Outboard) to close due to loss power to the Non
The loss of Div 1 RPS causes both RWCU PCIVs (Inboard and Outboard) to close due to loss power to the Non-regenerative heat exchanger outlet temperature instrument. The loss of Div 2 RPS power only closes the OUTBD isolation valve for RWCU. Second part requires knowledge
-regenerative heat exchanger outlet temperature instrument. The loss of Div 2 RPS power only closes the OUTBD isolation valve for RWCU. Second part requires knowledge & selection of procedures which provide guidance for configuration control during troubleshooting activities following a plant transient. Procedure 0.31 provides direction for configuration control of Motor operated isolation valves.
& selection of procedures which provide guidance for configuration control during troubleshooting activities following a plant transient. Procedure 0.31 provides direction for configuration control of Motor operated isolation valves.
Distracters:
Distracters:
A. This option is incorrect because both RWCU
A. This option is incorrect because both RWCU-MO-15 and 18 PCIVs close. The procedure controlling configuration is correct. This answer is plausible because generally an RPS bus de-energizing causes one division valve only to close. The candidate who does not recall the power supply to the Non-regenerative heat exchanger outlet temperature switch and that it solely causes both PCIVs to close may select this option.
-MO-15 and 18 PCIVs close. The procedure controlling configuration is correct. This answer is plausible because generally an RPS bus de-energizing causes one division valve only to close. The candidate who does not recall the power supply to the Non
B. This option is incorrect because both RWCU-MO-15 and 18 PCIVs close. The procedure controlling configuration is not correct. This answer is plausible because generally an RPS bus de-energizing causes one division valve only to close. The candidate who does not recall the power supply to the Non-regenerative heat exchanger outlet temperature switch 53
-regenerative heat exchanger outlet temperature switch and that it solely causes both PCIVs to close may select this option. B. This option is incorrect because both RWCU
 
-MO-15 and 18 PCIVs close. The procedure controlling configuration is not correct. This answer is plausible because generally an RPS bus de-energizing causes one division valve only to close. The candidate who does not recall the power supply to the Non
and that it solely causes both PCIVs to close may select this option. The candidate who does not recall the procedure that controls configuration in this instance may select this answer.
-regenerative heat exchanger outlet temperature switch 53 and that it solely causes both PCIVs to close may select this option. The candidate who does not recall the procedure that controls configuration in this instance may select this answer. D. This answer is incorrect because the controlling procedure is 0.31 not 0-CNS-WM-102 The correct valve response is listed. This answer is plausible because the correct valve response is listed. The candidate who does not correctly recall the proper configuration control procedure but does recall the correct valve response may select this answer.
D. This answer is incorrect because the controlling procedure is 0.31 not 0-CNS-WM-102 The correct valve response is listed. This answer is plausible because the correct valve response is listed. The candidate who does not correctly recall the proper configuration control procedure but does recall the correct valve response may select this answer.
Technical Reference(s):
Technical Reference(s):
Procedure 2.2.22 (Vital Instrument Power System), Rev. 71 Procedure 0.31(Equipment Status Control), Rev. 71.
Procedure 2.2.22 (Vital Instrument Power System), Rev. 71 Procedure 0.31(Equipment Status Control), Rev. 71.
Procedure 0-CNS-WM-102 (Work Implementation and Closeout), Rev. 01.
Procedure 0-CNS-WM-102 (Work Implementation and Closeout), Rev. 01.
Proposed references to be provided to applicants during examination:
Proposed references to be provided to applicants during examination:             None Learning Objective:
None Learning Objective:
COR0022102001080C Given a specific RPS malfunction, determine the effect on any of the following: c. PCIS INT032010100H010D Discuss the following as described in Administrative Procedure 0.31, Equipment Status Control: System component checklist requirements Question Source:
COR0022102001080C Given a specific RPS malfunction, determine the effect on any of the following:
Bank #
: c. PCIS INT032010100H010D Discuss the following as described in Administrative Procedure 0.31, Equipment Status Control: System component checklist requirements Question Source:
Modified Bank #
Bank #     Modified Bank #
New X Question Cognitive Level:       Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content:           55.41.10 55.43.3 55.45.13 Difficulty: 3 SRO Only - 10CFR55.43 b (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
New X   Question Cognitive Level:
Requires assessment of plant conditions and selection of procedure which provides guidance for configuration control during troubleshooting.
Memory or Fundamental Knowledge Comprehension or Analysis X   10 CFR Part 55 Content:
54
55.41.10     55.43.3 55.45.13 Difficulty: 3 SRO Only - 10CFR55.43 b (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
 
Requires assessment of plant conditions and selection of procedure which provides guidance for configuration control during troubleshooting. 54 5 5 56 57 58 59 60 Examination Outline Cross
55 56 57 58 59 60 Examination Outline Cross-
-


==Reference:==
==Reference:==


Level SRO Tier #   1 Group # 2 K/A # 295034A2.01 Importance Rating
Level                       SRO Tier #                     1 Group #                     2 K/A #             295034A2.01 Importance Rating         4.2 295034 Secondary Containment Ventilation High Radiation EA2       Ability to determine and/or interpret the following as they apply to SECONDARY CONTAINMENT VENTILATION HIGH RADIATION:
 
EA2.01 Ventilation radiation levels Question: 85 The plant is in Mode 1 and annunciator 9-4-1/E-4, RX BLDG VENT HI-HI RAD alarms due to a valid signal.
===4.2 295034===
(1) What is the LOWEST radiation level which causes this alarm to actuate?
Secondary Containment Ventilation High Radiation EA2 Ability to determine and/or interpret the following as they apply to SECONDARY CONTAINMENT VENTILATION HIGH RADIATION:
(2) What is the TS Bases for the allowable value of this instrument setpoint?
EA2.01 Ventilation radiation levels Question: 85 The plant is in Mode 1 and annunciator 9 1/E-4, RX BLDG VENT HI
A. (1) 5 mR/hr (2) Detect a steam leak in Secondary Containment.
-HI RAD alarms due to a valid signal.
B. (1) 5 mR/hr (2) Detect gross fuel cladding failure.
  (1) What is the LOWEST radiation level which causes this alarm to actuate?
C. (1) 10 mR/hr (2) Detect a steam leak in Secondary Containment.
(2) What is the TS Bases for the allowable value of this instrument setpoint?
D. (1) 10 mR/hr (2) Detect gross fuel cladding failure.
A. (1) 5 mR/hr (2) Detect a steam leak in Secondary Containment.
Answer:
B. (1) 5 mR/hr (2) Detect gross fuel cladding failure. C. (1) 10 mR/hr (2) Detect a steam leak in Secondary Containment.
D. (1) 10 mR/hr (2) Detect gross fuel cladding failure.
D. (1) 10 mR/hr (2) Detect gross fuel cladding failure.
Answer: D. (1) 10 mR/hr (2) Detect gross fuel cladding failure.
Explanation:
Explanation:
Requires candidate to first determine the annunciator setpoint (radiation level) for Rx Bldg Vent Hi-Hi Rad (10 vs. 5 mR/hr which is the Hi Rad). The second part requires knowledge of TS 3.3.6.2 (Secondary Containment Isolation Instrumentation) Bases for the setpoint. The setpoint is based upon detecting gross fuel cladding failure.
Requires candidate to first determine the annunciator setpoint (radiation level) for Rx Bldg Vent Hi-Hi Rad (10 vs. 5 mR/hr which is the Hi Rad). The second part requires knowledge of TS 3.3.6.2 (Secondary Containment Isolation Instrumentation) Bases for the setpoint. The setpoint is based upon detecting gross fuel cladding failure.
Distracters:
Distracters:
A. This option is incorrect as the setpoint for the Reactor Building Vent HI Hi rad is 10 mr/hr NOT 5 mr/hr. The basis for the setpoint is to detect gross clad fuel failure and NOT a steam leak in Secondary Containment. This choice is plausible if the Hi Rad is confused with the Hi HI Rad setpoint (change stem to reflect Hi Rad annunciator answer becomes correct) and steam leak in Secondary Containment would provide elevated radiation levels but is not the bases. A candidate may select this answer if they confuse the Hi setpoint of 5 mr/hr with the Hi Hi and believe the bases for the setpoint is a steam leak in Secondary Containment
A.     This option is incorrect as the setpoint for the Reactor Building Vent HI Hi rad is 10 mr/hr NOT 5 mr/hr. The basis for the setpoint is to detect gross clad fuel failure and NOT a steam leak in Secondary Containment. This choice is plausible if the Hi Rad is confused with the Hi HI Rad setpoint (change stem to reflect Hi Rad annunciator answer becomes correct) and steam leak in Secondary Containment would provide elevated radiation levels but is not the bases. A candidate may select this answer if they confuse the Hi setpoint of 5 mr/hr with the Hi Hi and believe the bases for the setpoint is a steam leak in Secondary Containment.
.
B.     This option is incorrect as the setpoint for the Reactor Building Vent HI Hi rad is 10 mr/hr NOT 5 mr/hr. This choice is plausible if the Hi Rad is confused with the Hi HI Rad 61
B. This option is incorrect as the setpoint for the Reactor Building Vent HI Hi rad is 10 mr/hr NOT 5 mr/hr. This choice is plausible if the Hi Rad is confused with the Hi HI Rad 61 setpoint (change stem to reflect Hi Rad annunciator). A candidate may select this answer if they confuse the Hi setpoint of 5 mr/hr with the Hi Hi and know the bases for the setpoint gross fuel cladding failure.
 
C. This option is incorrect as the setpoint is to detect gross clad fuel failure and NOT a steam leak in Secondary Containment. This choice is plausible due to a steam leak in Secondary Containment would provide elevated radiation levels but is not the bases. A candidate may select this answer if they correctly identify the Hi Hi setpoint and believe the bases for the setpoint is a steam leak in Secondary Containment
setpoint (change stem to reflect Hi Rad annunciator). A candidate may select this answer if they confuse the Hi setpoint of 5 mr/hr with the Hi Hi and know the bases for the setpoint gross fuel cladding failure.
. Technical Reference(s):
C.       This option is incorrect as the setpoint is to detect gross clad fuel failure and NOT a steam leak in Secondary Containment. This choice is plausible due to a steam leak in Secondary Containment would provide elevated radiation levels but is not the bases. A candidate may select this answer if they correctly identify the Hi Hi setpoint and believe the bases for the setpoint is a steam leak in Secondary Containment.
Procedure 2.3_9 1 (PANEL 9 ANNUNCIATOR 9 1), Rev. 46.
Technical Reference(s):
Technical Specification LCO 3.3.6.2, Secondary Containment Isolation Instrumentation Proposed references to be provided to applicants during examination:
Procedure 2.3_9-4-1 (PANEL 9 ANNUNCIATOR 9-4-1), Rev. 46.
NONE Learning Objective:
Technical Specification LCO 3.3.6.2, Secondary Containment Isolation Instrumentation Proposed references to be provided to applicants during examination:               NONE Learning Objective:
COR00118020010200 Given condition(s) and/or parameters associated with the Radiation Monitoring System, determine if related Technical Specification, Technical Requirements Manual, and Off Site Dose Assessment Manual Limiting conditions for Operation are met.
COR00118020010200 Given condition(s) and/or parameters associated with the Radiation Monitoring System, determine if related Technical Specification, Technical Requirements Manual, and Off Site Dose Assessment Manual Limiting conditions for Operation are met.
Question Source:
Question Source:
Bank #     Modified Bank #
Bank #
New X Question Cognitive Level:
Modified Bank #
Memory or Fundamental Knowledge Comprehension or Analysis X   10 CFR Part 55 Content:
New X Question Cognitive Level:     Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content:         55.41.10 55.43.3 55.45.13 Difficulty: 3 SRO Only - 10CFR55.43 b (2) Facility operating limitations in the TS and their bases.
55.41.10     55.43.3 55.45.13 Difficulty: 3 SRO Only - 10CFR55.43 b (2) Facility operating limitations in the TS and their bases.
Requires knowledge of TS Bases for the alarm setpoint.
Requires knowledge of TS Bases for the alarm setpoint.
62 RX BLDG VENT HIGH RAD    PANEL/WINDOW:
62
9-4-1/E-5 SETPOINT 5 mR/hr 1. (1777) RX BLDG VENT MONITOR DIV I HIGH RAD, Monitor A or C
 
: 2. (1778) RX BLDG VENT MONITOR DIV II HIGH RAD, Monitor B or D CIC 9-4-1/E  1. RMP-RR-455 CH-1 or CH-3  2. RMP-RR-455 CH-2 or CH-4 PROBABLE CAUSES Refueling activities.
RX BLDG VENT               PANEL/WINDOW:
HIGH RAD                9-4-1/E-5 SETPOINT                                       CIC                                    9-4-1/E 5 mR/hr
: 1. (1777) RX BLDG VENT MONITOR DIV I           1. RMP-RR-455 CH-1 or CH-3 HIGH RAD, Monitor A or C
: 2. (1778) RX BLDG VENT MONITOR DIV II         2. RMP-RR-455 CH-2 or CH-4 HIGH RAD, Monitor B or D PROBABLE CAUSES Refueling activities.
REFERENCES Technical Specification LCO 3.3.6.2, Secondary Containment Isolation Instrumentation.
REFERENCES Technical Specification LCO 3.3.6.2, Secondary Containment Isolation Instrumentation.
Off-Site Dose Assessment Manual DLCO 3.2.1, Gaseous Effluents Concentration.
Off-Site Dose Assessment Manual DLCO 3.2.1, Gaseous Effluents Concentration.
Line 3,671: Line 3,384:
Off-Site Dose Assessment Manual DLCO 3.2.3, Iodine and Particulates.
Off-Site Dose Assessment Manual DLCO 3.2.3, Iodine and Particulates.
Emergency Procedure 5.1RAD, Building Radiation Trouble.
Emergency Procedure 5.1RAD, Building Radiation Trouble.
RX BLDG VENT HI-HI RAD    PANEL/WINDOW:
RX BLDG VENT               PANEL/WINDOW:
9-4-1/E-4 SETPOINT 1. (1763) RX BLDG VENT MONITOR A HI-HI RAD at 10 mR/hr 2. (1764) RX BLDG VENT MONITOR B HI-HI RAD at 10 mR/hr
HI-HI RAD                9-4-1/E-4 SETPOINT                                       CIC                                    9-4-1/E
: 3. (1779) RX BLDG VENT MONITOR C HI-HI RAD at 10 mR/hr
: 1. (1763) RX BLDG VENT MONITOR A               1. RMP-RM-452A HI-HI RAD at 10 mR/hr
: 4. (1780) RX BLDG VENT MONITOR D HI-HI RAD at 10 mR/hr CIC 9-4-1/E 1. RMP-RM-452A  2. RMP-RM-452B 
: 2. (1764) RX BLDG VENT MONITOR B               2. RMP-RM-452B HI-HI RAD at 10 mR/hr
: 3. RMP-RM-452C  4. RMP-RM-452D  PROBABLE CAUSES Refueling floor accident.
: 3. (1779) RX BLDG VENT MONITOR C               3. RMP-RM-452C HI-HI RAD at 10 mR/hr
: 4. (1780) RX BLDG VENT MONITOR D               4. RMP-RM-452D HI-HI RAD at 10 mR/hr PROBABLE CAUSES Refueling floor accident.
Line external to primary containment breaks.
Line external to primary containment breaks.
REFERENCES General Operating Procedure 2.1.22, Recovering from a Group Isolation.
REFERENCES General Operating Procedure 2.1.22, Recovering from a Group Isolation.
Emergency Procedure 5.1RAD, Building Radiation Trouble.
Emergency Procedure 5.1RAD, Building Radiation Trouble.
63 64 Examination Outline Cross
63
-
 
64 Examination Outline Cross-


==Reference:==
==Reference:==


Level SRO Tier #   2 Group # 1 K/A # 203000A2.04 Importance Rating
Level                       SRO Tier #                     2 Group #                     1 K/A #               203000A2.04 Importance Rating         3.6 203000 Residual Heat Removal /Low Pressure Coolant Injection: Injection Mode A2        Ability to (a) predict the impacts of the following on the RHR/LPCI: INJECTION MODE (PLANT SPECIFIC); and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
A2.04 A.C. failures Question: 86 The following conditions exist during a large break LOCA from rated power:
* No Off-Site is power available.
* DG1 is unavailable.
* RHR Pump D is unavailable.
* Core Spray Pump B is unavailable.
* Reactor Building is inaccessible.
* RPV Pressure is 35 psig and steady.
(1) Which RHR Loop is available for LPCI injection?
(2) What action is procedurally required if additional injection is required to assure adequate core cooling?
A. (1) A (2) Enter 5.3EMPWR (Emergency Power During MODES 1, 2, OR 3) and use the Supplemental Diesel Generator (SDG) to re-energize 4160V 1F Bus.
B. (1) A (2) Enter 5.3ALT-STRATEGY (Alternate Core Cooling Mitigating Strategies) and inject using Fire Protection to RHR.
C. (1) B (2) Enter 5.3ALT-STRATEGY (Alternate Core Cooling Mitigating Strategies) and inject using Fire Protection to RHR.
D. (1) B (2) Enter 5.3EMPWR (Emergency Power During MODES 1, 2, OR 3) and use the Supplemental Diesel Generator (SDG) to re-energize 4160V 1F Bus.
ANSWER:
A. (1) A (2) Enter 5.3EMPWR (Emergency Power During MODES 1, 2, OR 3) and use the Supplemental Diesel Generator (SDG) to re-energize 4160V 1F Bus.
Explanation of answer:
65


===3.6 203000===
RHR Loop A consists of RHR pumps A & C. RHR Loop B consists of RHR pumps B & D. Loss of Div 1 4160V Bus 1F de-energizes RHR pumps A & B and AC valves in Loop A. The only pump with power is RHR Pump C (Loop A) with RHR-MO-27A (Outboard Injection Valve) de-energized in the OPEN position provides LPCI flow to the RPV with RHR-MO25A (Inboard Injection Valve) being DC powered. No RHR pumps are available in RHR Loop B (RHR pump D unavailable and B with no power). The connection of Fire Protection to SW Emergency Core Flooding ties into RHR Loop A piping. This is accomplished by tying in Fire Protection to the suction side of the RHRSW Booster pump via a 3 inch fire hose. The Diesel Fire pump discharge pressure is approximately 160 psig. With the friction loss going through the fire hose and the pressure drop across the RHRSW Booster pump the Fire Protection pressure is too low to inject with the RHR system running. At full injection flow, the RHR discharge pressure is approximately 175 psig. In order to connect Fire Protection to RHR Loop B requires opening the normally closed MO-20 valve which has no power and access to the Reactor Building is not allowed. Loss of all off-site power with only DG2 supplying Div 2 4160V Bus 1G with a LOCA requires entry into EOP-1A (RPV Control) and 5.3EMPWR. 5.3EMPWR provides guidance for RPV and Containment, Balance of Plant, and Electrical Systems guidance via Attachments 1, 2,
Residual Heat Removal /Low Pressure Coolant Injection: Injection Mode A2 Ability to (a) predict the impacts of the following on the RHR/LPCI: INJECTION MODE (PLANT SPECIFIC); and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
& 3. Requires SRO to determine the most effective and timely means of gaining additional RPV injection under conditions challenging Adequate Core Cooling. Utilizing the SDG will provide 2 additional RHR pumps for injection. EOP-1A provides guidance to utilize procedure 5.3ALT-STRATEGY to align required/desired systems, but due to extreme system/plant conditions, normal operation is not possible. RCIC is the preferred injection system, but with RPV at 35 psig RCIC will not provide sufficient makeup (injecting fire protection is the next preferred but requires Reactor Building access). NOTE: The RHR System operating pressure listed here was obtained from Surveillance Procedure 6.1RHR.101 data and the Diesel Fire Protection system operating pressure was obtained from Surveillance Procedure 6.FP.101 data.
A2.04 A.C. failures Question:  86 The following conditions exist during a large break LOCA from rated power:
Distracters:
No Off-Site is power available.
DG1 is unavailable.
RHR Pump D is unavailable.
Core Spray Pump B is unavailable.
Reactor Building is inaccessible.
RPV Pressure is 35 psig and steady.
  (1) Which RHR Loop is available for LPCI injection?
(2) What action is procedurally required if additional injection is required to assure adequate core cooling?
A.  (1)  A (2)  Enter 5.3EMPWR (Emergency Power During MODES 1, 2, OR 3) and use the Supplemental Diesel Generator (SDG) to re
-energize 4160V 1F Bus.
 
B.  (1)  A (2)  Enter 5.3ALT-STRATEGY (Alternate Core Cooling Mitigating Strategies) and inject using Fire Protection to RHR.
C.  (1)  B (2)  Enter 5.3ALT-STRATEGY (Alternate Core Cooling Mitigating Strategies) and inject using Fire Protection to RHR.
D.  (1)  B (2)  Enter 5.3EMPWR (Emergency Power During MODES 1, 2, OR 3) and use the Supplemental Diesel Generator (SDG) to re
-energize 4160V 1F Bus.
ANSWER:  A.  (1)  A (2)  Enter 5.3EMPWR (Emergency Power During MODES 1, 2, OR 3) and use the Supplemental Diesel Generator (SDG) to re
-energize 4160V 1F Bus.
Explanation of answer:
65 RHR Loop A consists of RHR pumps A & C. RHR Loop B consists of RHR pumps B & D. Loss of Div 1 4160V Bus 1F de
-energizes RHR pumps A & B and AC valves in Loop A. The only pump with power is RHR Pump C (Loop A) with RHR-MO-27A (Outboard Injection Valve) de
-energized in the OPEN position provides LPCI flow to the RPV with RHR
-MO25A (Inboard Injection Valve) being DC powered. No RHR pumps are available in RHR Loop B (RHR pump D unavailable and B with no power). The connection of Fire Protection to SW Emergency Core Flooding ties into RHR Loop A piping. This is accomplished by tying in Fire Protection to the suction side of the RHRSW Booster pump via a 3 inch fire hose. The Diesel Fire pump discharge pressure is approximately 160 psig. With the friction loss going through the fire hose and the pressure drop across the RHRSW Booster pump the Fire Protection pressure is too low to inject with the RHR system running. At full injection flow, the RHR discharge pressure is approximately 175 psig. In order to connect Fire Protection to RHR Loop B requires opening the normally closed MO
-20 valve which has no power and access to the Reactor Building is not allowed. Loss of all off
-site power with only DG2 supplying Div 2 4160V Bus 1G with a LOCA requires entry into EOP
-1A (RPV Control) and 5.3EMPWR. 5.3EMPWR provides guidance for RPV and Containment, Balance of Plant, and Electrical Systems guidance via Attachments 1, 2,  
& 3. Requires SRO to determine the most effective and timely means of gaining additional RPV injection under conditions challenging Adequate Core Cooling. Utilizing the SDG will provide 2 additional RHR pumps for injection. EOP
-1A provides guidance to utilize procedure 5.3ALT
-STRATEGY to align required/desired systems, but due to extreme system/plant conditions, normal operation is not possible. RCIC is the preferred injection system, but with RPV at 35 psig RCIC will not provide sufficient makeup (injecting fire protection is the next preferred but requires Reactor Building access). NOTE: The RHR System operating pressure listed here was obtained from Surveillance Procedure 6.1RHR.101 data and the Diesel Fire Protection system operating pressure was obtained from Surveillance Procedure 6.FP.101 dat
: a. Distracters:
B. This answer is incorrect because aligning Fire Protection to RHR will not add water due to RHR system pressure being much higher than Fire Protection header pressure while RHR Pump C is operating. This choice is plausible because the alignment of FP can be performed. The candidate who correctly identifies the available RHR loop and confuses the differences of system driving head for injection would choose this answer.
B. This answer is incorrect because aligning Fire Protection to RHR will not add water due to RHR system pressure being much higher than Fire Protection header pressure while RHR Pump C is operating. This choice is plausible because the alignment of FP can be performed. The candidate who correctly identifies the available RHR loop and confuses the differences of system driving head for injection would choose this answer.
C. This answer is incorrect because B Loop RHR is unavailable and the reactor building is inaccessible to manually open crosstie to Loop B (MO
C. This answer is incorrect because B Loop RHR is unavailable and the reactor building is inaccessible to manually open crosstie to Loop B (MO-20). This choice is plausible due to the diversity of power supplies to the RHR pumps & valves and 5.3ALT-STRATEGY provides guidance for Fire Protection connection to RHR but this will not work due to RHR system pressure and Fire Protection pressure differences. The candidate who incorrectly identifies the available RHR loop and confuses the Service Water availability for injection would choose this answer.
-20). This choice is plausible due to the diversity of power supplies to the RHR pumps & valves and 5.3ALT
-STRATEGY provides guidance for Fire Protection connection to RHR but this will not work due to RHR system pressure and Fire Protection pressure differences. The candidate who incorrectly identifies the available RHR loop and confuses the Service Water availability for injection would choose this answer.
 
D. This answer is incorrect because B Loop RHR is unavailable. This choice is plausible due to the diversity of power supplies to the RHR pumps & valves. The candidate who incorrectly identifies the available RHR loop and correctly identifies the need to energize Bus 1F for the SDG would choose this answer.
D. This answer is incorrect because B Loop RHR is unavailable. This choice is plausible due to the diversity of power supplies to the RHR pumps & valves. The candidate who incorrectly identifies the available RHR loop and correctly identifies the need to energize Bus 1F for the SDG would choose this answer.
Technical Reference(s):
Technical Reference(s):
Procedure 5.8 Attachment 1(RPV Control EOP
Procedure 5.8 Attachment 1(RPV Control EOP-1A), Rev. 17.
-1A), Rev. 17.
Procedure 5.3ALT-STRATEGY Alternate Core Cooling Mitigating Strategies, Rev. 42.
Procedure 5.3ALT-STRATEGY Alternate Core Cooling Mitigating Strategies, Rev. 42.
Procedure 5.3EMPWR (Emergency Power During MODES 1, 2, OR 3), Rev. 43.
Procedure 5.3EMPWR (Emergency Power During MODES 1, 2, OR 3), Rev. 43.
66 Proposed references to be provided to applicants during examination:
66
NONE Learning Objective:
 
COR0022302001080A Predict the consequences a malfunction of the following will have on the RHR system: A.C. electrical power (including RPS)
Proposed references to be provided to applicants during examination:   NONE Learning Objective:
COR0022302001080A         Predict the consequences a malfunction of the following will have on the RHR system: A.C. electrical power (including RPS)
Question Source:
Question Source:
Bank #     Modified Bank # 4029 (Attached)
Bank #
New Question Cognitive Level:
Modified Bank # 4029     (Attached)
Memory or Fundamental Knowledge Comprehension or Analysis X   10 CFR Part 55 Content:
New Question Cognitive Level:   Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content:     55.41.5 55.45.6 Difficulty: 3 SRO Only - 10CFR55.43 b (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
55.41.5     55.45.6 Difficulty: 3 SRO Only - 10CFR55.43 b (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Requires assessment of conditions and selection of procedure with which to proceed.
Requires assessment of conditions and selection of procedure with which to proceed.
67 68 69 70 71 72 73 74 75 Examination Outline Cross
67
-
 
68 69 70 71 72 73 74 75 Examination Outline Cross-


==Reference:==
==Reference:==


Level SRO Tier #   2 Group # 1 K/A # S212000G2.1.20 Importance Rating
Level                     SRO Tier #                     2 Group #                   1 K/A #               S212000G2.1.20 Importance Rating         4.6 212000 Reactor Protection System 2.1.20 Ability to interpret and execute procedure steps.
 
Question: 87 The plant is in Mode 5 with all OPERABLE Control Rods (Core Cell contains fuel) fully inserted.
===4.6 212000===
Procedure 4.5 (Reactor Protection/Alternate Rod Insertion Systems), Section 7 (Bypass Reactor Mode Switch - Shutdown Position Scram and transfer MODE switch), directs the following Step:
Reactor Protection System 2.1.20 Ability to interpret and execute procedure steps.
7.3.2.2     Inform Shift Manager that Reactor Mode Switch - Shutdown Position Scram (LCO 3.3.1.1, Function 10) is inoperable due to being bypassed.
Question: 87 The plant is in Mode 5 with all OPERABLE Control Rods (Core Cell contains fuel) fully inserted.
: a. Enter applicable Conditions and Required Actions for following LCOs as required:
Procedure 4.5 (Reactor Protection/Alternate Rod Insertion Systems), Section 7 (Bypass Reactor Mode Switch  
: 1. LCO 3.3.1.1.
- Shutdown Position Scram and transfer MODE switch), directs the following Step:
: 2. LCO 3.10.3.
7.3.2.2 Inform Shift Manager that Reactor Mode Switch  
: 3. LCO 3.10.4.
- Shutdown Position Scram (LCO 3.3.1.1, Function 10) is inoperable due to being bypassed.
: a. Enter applicable Conditions and Required Actions for following LCOs as required: 1. LCO 3.3.1.1.
: 2. LCO 3.10.3. 3. LCO 3.10.4.
What is the MINIMUM Required Action(s) for this step IAW TS LCO 3.3.1.1 (RPS Instrumentation) if an OPERABLE Control Rod is fully withdrawn with the DIV 1 Reactor Mode Switch Scram bypass jumper installed?
What is the MINIMUM Required Action(s) for this step IAW TS LCO 3.3.1.1 (RPS Instrumentation) if an OPERABLE Control Rod is fully withdrawn with the DIV 1 Reactor Mode Switch Scram bypass jumper installed?
Place the bypassed channel in trip...
Place the bypassed channel in trip...
A. within 12 hours ONLY.
A. within 12 hours ONLY.
B. OR Insert the Control Rod within 1 hour.
B. OR Insert the Control Rod within 1 hour.
C. OR Insert the Control Rod within 6 hours.
C. OR Insert the Control Rod within 6 hours.
D. OR Insert the Control Rod within 12 hours.
D. OR Insert the Control Rod within 12 hours.
Answer: B. OR Insert the Control Rod within 1 hour.
Answer:
B. OR Insert the Control Rod within 1 hour.
Explanation:
Explanation:
Requires the step to be interpreted and executed to determine the TS Required actions. Requires knowledge of TS Bases and what is being bypassed by the jumpers (ONLY Mode Switch SHUTDOWN position is in RPS Channels A3 & B3). LCO is applicable in MODEs 1, 2 and 5 (With any control rod withdrawn from a core cell containing one or more fuel assemblies). Requires application of multiple TS ACTIONS to determine correct action with knowledge of TS 3.3.1.1 Bases (Loss of trip capability). Condition C requires restoration of RPS trip capability within 1 hour. If not completed within 1 hr, enter Condition H (as directed by TABLE 3.3.1.1
Requires the step to be interpreted and executed to determine the TS Required actions.
-1) which requires initiating action to insert the Control Rod immediately. Options are to exit the 76 MODE of applicability (all Control Rods inserted), place bypassed channel in trip or remove jumper to restore trip capability. With the bypassed channel in trip, trip capability is restored.
Requires knowledge of TS Bases and what is being bypassed by the jumpers (ONLY Mode Switch SHUTDOWN position is in RPS Channels A3 & B3). LCO is applicable in MODEs 1, 2 and 5 (With any control rod withdrawn from a core cell containing one or more fuel assemblies).
Requires application of multiple TS ACTIONS to determine correct action with knowledge of TS 3.3.1.1 Bases (Loss of trip capability). Condition C requires restoration of RPS trip capability within 1 hour. If not completed within 1 hr, enter Condition H (as directed by TABLE 3.3.1.1-1) which requires initiating action to insert the Control Rod immediately. Options are to exit the 76
 
MODE of applicability (all Control Rods inserted), place bypassed channel in trip or remove jumper to restore trip capability. With the bypassed channel in trip, trip capability is restored.
Distracters:
Distracters:
A. This answer is incorrect due to placing the bypassed channel in trip within 12 hours not being the only additional TS Action. This choice is plausible if loss of function in one channel and RPS trip (full scram) capability is not recognized and not understanding the impact of inserting the withdrawn control rod (no longer in the MODE of applicability). The candidate who does not recognize a loss of function & trip capability would choose this answer.
A. This answer is incorrect due to placing the bypassed channel in trip within 12 hours not being the only additional TS Action. This choice is plausible if loss of function in one channel and RPS trip (full scram) capability is not recognized and not understanding the impact of inserting the withdrawn control rod (no longer in the MODE of applicability). The candidate who does not recognize a loss of function & trip capability would choose this answer.
C. This answer is incorrect due to inserting the Control Rod within 6 hours not being the only additional TS Action. This choice is plausible if loss of function in one channel and RPS trip (full scram) capability is not recognized and not understanding the impact of inserting the withdrawn control rod (no longer in the MODE of applicability). The candidate who does not recognize a loss of RPS trip capability would choose this answer. Placing the bypassed channel in trip OR Inserting the Control Rod within 6 hours is plausible if both jumpers were installed with failure to recognize loss of trip (full scram) capability.
C. This answer is incorrect due to inserting the Control Rod within 6 hours not being the only additional TS Action. This choice is plausible if loss of function in one channel and RPS trip (full scram) capability is not recognized and not understanding the impact of inserting the withdrawn control rod (no longer in the MODE of applicability). The candidate who does not recognize a loss of RPS trip capability would choose this answer. Placing the bypassed channel in trip OR Inserting the Control Rod within 6 hours is plausible if both jumpers were installed with failure to recognize loss of trip (full scram) capability.
D. This answer is incorrect due to inserting the Control Rod within 12 hours not meeting the additional TS Action. This choice is plausible if loss of function in one channel and RPS trip (full scram) capability is not recognized and not understanding the impact of inserting the withdrawn control rod (no longer in the MODE of applicability). The candidate who does not recognize a loss of function & RPS trip capability but does understand exiting the Mode of applicability (insert Control Rod) would choose this answer. Technical Reference(s):
D. This answer is incorrect due to inserting the Control Rod within 12 hours not meeting the additional TS Action. This choice is plausible if loss of function in one channel and RPS trip (full scram) capability is not recognized and not understanding the impact of inserting the withdrawn control rod (no longer in the MODE of applicability). The candidate who does not recognize a loss of function & RPS trip capability but does understand exiting the Mode of applicability (insert Control Rod) would choose this answer.
Technical Reference(s):
Procedure 4.5 (Reactor Protection/Alternate Rod Insertion Systems), Rev. 31.
Procedure 4.5 (Reactor Protection/Alternate Rod Insertion Systems), Rev. 31.
LCO 3.3.1.1 RPS Instrumentation LCO 3.10.4 Single Control Rod Withdrawal
LCO 3.3.1.1         RPS Instrumentation LCO 3.10.4           Single Control Rod Withdrawal--Cold Shutdown Proposed references to be provided to applicants during examination:             LCO 3.3.1.1 &
--Cold Shutdown
 
Proposed references to be provided to applicants during examination:
LCO 3.3.1.1 &
Table with Mode Switch info only Learning Objective:
Table with Mode Switch info only Learning Objective:
COR00221020010200 Given conditions and/or parameters associated with the RPS, determine if related Technical Specification and Technical Requirements Manual Limiting Condition for Operations are met.
COR00221020010200 Given conditions and/or parameters associated with the RPS, determine if related Technical Specification and Technical Requirements Manual Limiting Condition for Operations are met.
COR0022102001050D Briefly describe the following concepts as they apply to RPS:   Mode switch position Question Source:
COR0022102001050D Briefly describe the following concepts as they apply to RPS: Mode switch position Question Source:
Bank #     Modified Bank #
Bank #
New X Question Cognitive Level:
Modified Bank #
Memory or Fundamental Knowledge Comprehension or Analysis X   10 CFR Part 55 Content:
New X Question Cognitive Level:       Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content:           55.41.10 55.43.5 55.45.12 Difficulty: 3 77
55.41.10     55.43.5     55.45.12 Difficulty: 3 77 SRO Only - 10CFR55.43 b (2) Facility operating limitations in the TS and their bases.
 
SRO Only - 10CFR55.43 b (2) Facility operating limitations in the TS and their bases.
Requires application of TS action statements with knowledge to TS Bases. The SRO is required to determine "loss of function" for a given equipment condition.
Requires application of TS action statements with knowledge to TS Bases. The SRO is required to determine "loss of function" for a given equipment condition.
78 79 80 81 82 83 84 Examination Outline Cross
78
-
 
79 80 81 82 83 84 Examination Outline Cross-


==Reference:==
==Reference:==


Level SRO Tier #   2 Group # 1 K/A # 215003A2.05 Importance Rating
Level                     SRO Tier #                   2 Group #                   1 K/A #               215003A2.05 Importance Rating       3.5 215003 Intermediate Range Monitor System A2        Ability to (a) predict the impacts of the following on the INTERMEDIATE RANGE MONITOR (IRM) SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
A2.05 Faulty or erratic operation of detectors/system Question: 88 A reactor startup is in progress with the following conditions present:
* IRM G is Inoperable and bypassed.
* Reactor power is below the point of adding heat.
The following annunciator alarms and clears multiple times within 30 seconds and repeats for 3 minutes due to IRM A spiking.
IRM RPS CH A            PANEL/WINDOW:
UPSCALE TRIP OR INOP                9-5-1/D-7 (1)      What is the result of IRM A spiking?
(2)      What is the MOST Restrictive required Technical Specification action?
A.    (1) Rod block ONLY (2) Place Channel or Associated Trip system in Trip within 12 hours.
B.    (1) Rod block and 1/2 scram (2) Place Channel or Associated Trip system in Trip within 12 hours.
C.    (1) Rod block ONLY (2) Place Channel in one trip system OR one trip system in trip within 6 hours.
D.    (1) Rod block and 1/2 scram (2) Place Channel in one trip system OR one trip system in trip within 6 hours.
Answer:
B.    (1) Rod block and 1/2 scram (2) Place Channel or Associated Trip system in Trip within 12 hours.
Explanation:
Procedure 4.1.2 provides Abnormal IRM Readings on Attachment 1 to determine IRM operability due to Spiking. Since Rod Block & Scram setpoints (102.5 & 117.5/125 of scale) are reached (>59/125 scale) IRM A is inoperable. The Trip System is made up of 4 Channels (IRMs A, C, E, and G are in Division 1 and B, D, F, and H are in Division 2). TS 3.3.1.1 Condition A is entered for IRMs A and G being inoperable. Table 3.3.1.1-1 requires 3 operable 1


===3.5 215003===
IRMs per trip system. With IRM A &G Inoperable, Table 3.3.1-1 Required Channels per Trip System requirements are not met. LCO 3.3.1.1, Condition A applies and must be met. If 2 IRMs were INOP in two different trip systems, then Condition B would apply.
Intermediate Range Monitor System A2 Ability to (a) predict the impacts of the following on the INTERMEDIATE RANGE MONITOR (IRM) SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
A2.05 Faulty or erratic operation of detectors/system Question:  88 A reactor startup is in progress with the following conditions present:
IRM G is Inoperable and bypassed.
Reactor power is below the point of adding heat.
The following annunciator alarms and clears multiple times within 30 seconds and repeats for 3 minutes due to IRM A spiking.      IRM RPS CH A UPSCALE TRIP OR INOP    PANEL/WINDOW:
9-5-1/D-7  (1)  What is the result of IRM A spiking?
(2) What is the MOST Restrictive required Technical Specification action?
A. (1)  Rod block ONLY (2)  Place Channel or Associated Trip system in Trip within 12 hours.
B. (1)  Rod block and 1/2 scram (2)  Place Channel or Associated Trip system in Trip within 12 hours.
C. (1)  Rod block ONLY (2)  Place Channel in one trip system OR one trip system in trip within 6 hours.
D. (1)  Rod block and 1/2 scram (2)  Place Channel in one trip system OR one trip system in trip within 6 hours.
Answer:  B. (1)  Rod block and 1/2 scram (2)  Place Channel or Associated Trip system in Trip within 12 hours.
Explanation:
Procedure 4.1.2 provides Abnormal IRM Readings on Attachment 1 to determine IRM operability due to Spiking. Since Rod Block & Scram setpoints (102.5 & 117.5/125 of scale) are reached (>59/125 scale) IRM A is inoperable. The Trip System is made up of 4 Channels (IRMs A, C, E, and G are in Division 1 and B, D, F, and H are in Division 2). TS 3.3.1.1 Condition A is entered for IRMs A and G being inoperable. Table 3.3.1.1
-1 requires 3 operable 1
IRMs per trip system. With IRM A &G Inoperable, Table 3.3.1
-1 Required Channels per Trip System requirements are not met. LCO 3.3.1.1, Condition A applies and must be met. If 2 IRMs were INOP in two different trip systems, then Condition B would apply.
Distracters:
Distracters:
A. This answer is incorrect because an RPS trip would also occur. The required action is correct. This answer is plausible because an IRM does cause a Rod Block and the required action is correct. The candidate who does not realize the RPS trip occurs would select this answer.
A. This answer is incorrect because an RPS trip would also occur. The required action is correct. This answer is plausible because an IRM does cause a Rod Block and the required action is correct. The candidate who does not realize the RPS trip occurs would select this answer.
Line 3,816: Line 3,512:
Procedure 4.1.2 (Intermediate Range Monitoring System), Rev. 21.
Procedure 4.1.2 (Intermediate Range Monitoring System), Rev. 21.
Procedure 4.5 (Reactor Protection/Alternate Rod Insertion Systems), Rev. 31.
Procedure 4.5 (Reactor Protection/Alternate Rod Insertion Systems), Rev. 31.
TS LCO 3.3.1.1, Reactor Protection System Instrumentation Proposed references to be provided to applicants during examination:
TS LCO 3.3.1.1, Reactor Protection System Instrumentation Proposed references to be provided to applicants during examination:           LCO 3.3.1.1 Learning Objective:
LCO 3.3.1.1 Learning Objective:
COR00212020010200 Given conditions and/or parameters associated with the IRM's, determine if related Technical Specification and Technical Requirements Manual Limiting Condition for Operation are met.
COR00212020010200 Given conditions and/or parameters associated with the IRM's, determine if related Technical Specification and Technical Requirements Manual Limiting Condition for Operation are met.
COR0021202001090A Given plant conditions, determine if the following IRM actions should occur: Rod Block.
COR0021202001090A Given plant conditions, determine if the following IRM actions should occur: Rod Block.
COR0021202001090B Given plant conditions, determine if the following IRM actions should occur: Reactor Scram.
COR0021202001090B Given plant conditions, determine if the following IRM actions should occur: Reactor Scram.
INT007-05-05, OPS CNS Tech Specs 3.3, Instrumentation
INT007-05-05, OPS CNS Tech Specs 3.3, Instrumentation
: 3. Given a set of plant conditions that constitutes non
: 3. Given a set of plant conditions that constitutes non-compliance with a Section 3.3 LCO, determine the ACTIONS that are required.
-compliance with a Section 3.3 LCO, determine the ACTIONS that are required.
Question Source:
Question Source:
Bank #     Modified Bank #
Bank #
New X   Question Cognitive Level:
Modified Bank #
Memory or Fundamental Knowledge X   Comprehension or Analysis     10 CFR Part 55 Content:
New X Question Cognitive Level:       Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content:         55.41.50 55.43.2 Difficulty: 3 2
55.41.50     55.43.2 Difficulty: 3 2
 
SRO Only - Facility operating limitations in the technical specifications and their bases. (10 CFR 55.43(b)(2)).
SRO Only - Facility operating limitations in the technical specifications and their bases.
(10 CFR 55.43(b)(2)).
The SRO is responsible for TS LCO required action determination.
The SRO is responsible for TS LCO required action determination.
IRM UPSCALE    PANEL/WINDOW:
IRM               PANEL/WINDOW:
9-5-1/E-7 SETPOINT (2354) 102.5/125 of scale (TRM 108/125 of full scale)
UPSCALE                9-5-1/E-7 SETPOINT                                     CIC                                    9-5-1/E-7 (2354) 102.5/125 of scale (TRM             MNI-NAM-41A through MNI-NAM-41H 108/125 of full scale)
CIC 9-5-1/E-7 MNI-NAM-41A through MNI
IRM RPS CH A             PANEL/WINDOW:
-NAM-41H IRM RPS CH A UPSCALE TRIP OR INOP    PANEL/WINDOW:
UPSCALE TRIP OR INOP                9-5-1/D-7 SETPOINT                                     CIC                                    9-5-1/D-7 (2353) Upscale trip at 117.5/125 of scale NMI-NAM-41A, NMI-NAM-41C, (Tech Spec  121/125 of scale) or inop       NMI-NAM-41E, or NMI-NAM-41G due to:
9-5-1/D-7 SETPOINT (2353) Upscale trip at 117.5/125 of scale (Tech Spec  121/125 of scale) or inop due to: 1. IRM module unplugged
: 1. IRM module unplugged
: 2. High voltage low
: 2. High voltage low
: 3. MODE switch not in operate
: 3. MODE switch not in operate
: 4. Loss of negative supply voltage CIC 9-5-1/D-7 NMI-NAM-41A, NMI-NAM-41C, NMI-NAM-41E, or NMI
: 4. Loss of negative supply voltage 3
-NAM-41G  3 4
 
5 6
4 5
7 8
6 7
9 10 Examination Outline Cross
8 9
-
10 Examination Outline Cross-


==Reference:==
==Reference:==


Level SRO Tier #   2 Group # 1 K/A # 259002G2.2.40 Importance Rating
Level                       SRO Tier #                       2 Group #                     1 K/A #               259002G2.2.40 Importance Rating           4.5 259002 Reactor Water Level Control System 2.2.40 Ability to apply Technical Specifications for a system Question: 89 The Plant is operating at rated power on March 10th.
At 1300 the RO observes the following:
* RFC-LI-94A indicates 36 inches and is slowly rising.
* RFC-LI-94B indicates 32 inches and is slowly lowering.
* RFC-LI-94C indicates 37 inches and is slowly rising.
* Feedwater flow is slowly rising.
If these conditions continue, when is the LATEST reactor power is required to be less than 25%
RTP IAW TS 3.3.2.2, Feedwater and Main Turbine High Water Level Trip??
A. March 10 at 1500 B. March 10 at 1900 C. March 17 at 1300 D. March 17 at 1700 Answer:
D. March 17 at 1700 Explanation:
Requires knowledge of TS instrument surveillance requirements and application of required actions associated with the RVLC System. The RVLC system utilizes NR level and FW flow instruments to properly control RPV water level. If one level instrument differs from the average mean by 8 inches or more, the RVLCS will remove the anomalous input and control utilizing the remaining valid instruments. The RVLC halts the vessel rise after it removes LI-94B from processing. SR 3.3.2.2.1 requires a channel check which is documented in 6.LOG.601 requiring these Narrow Range level indicators to indicate within 2 inches of each other. Since the B NR is greater than 2 inches from A & C, this channel is considered inoperable (failed channel check). TS 3.3.2.2, Condition A is entered at the time the failure is identified and the required time to place the channel in a tripped condition is 7 days (March 17 at 1300). Once the completion time is expired, reactor power must be lowered below 25% RTP IAW Condition C within 4 hours. 7 days + 4 hours = March 17 at 1700.
11


===4.5 259002===
Distracters:
Reactor Water Level Control System 2.2.40 Ability to apply Technical Specifications for a system Question:  89 The Plant is operating at rated power on March 10 th. At 1300 the RO observes the following:
A. This answer is incorrect because only 1 NR instrument being inoperable. This answer is plausible due to 2 NR level indicators being higher than normal setpoint (35) and reflects a 2 hour completion time. The candidate who incorrectly determines 2 level NR Level instruments are inoperable and confuses completion time to restore trip capability vs.
RFC-LI-94A indicates 36 inches and is slowly rising.
reducing power to <25% would select this answer.
RFC-LI-94B indicates 32 inches and is slowly lowering.
B. This answer is incorrect because only 1 NR instrument being inoperable. This answer is plausible due to 2 NR level indicators being higher than normal setpoint (35) and reflects a 2 + 4 hour (6 hrs) completion time. The candidate who incorrectly determines 2 level NR Level instruments are inoperable but correctly identifies the additional 4 hours to reducing power to <25% would select this answer.
RFC-LI-94C indicates 37 inches and is slowly rising.
C. This answer is incorrect due not applying the 4 additional hours allowed to reduce power below 25%. This answer is plausible due to 1 NR level indicator being lower than normal setpoint (35) with higher FW flow and reflects the 7 day completion time (if stem were changed to ask when required to be in trip - would be correct). The candidate who correctly determines 1 level NR Level instrument is inoperable but confuses placing the channel in trip vs. reducing power to <25% would select this answer.
Feedwater flow is slowly rising.
If these conditions continue, when is the LATEST reactor power is required to be less than 25% RTP IAW TS 3.3.2.2, Feedwater and Main Turbine High Water Level Trip??    A. March 10 at 1500
 
B. March 10 at 1900 C. March 17 at 1300 D. March 17 at 1700 Answer: D. March 17 at 1700 Explanation:
Requires knowledge of TS instrument surveillance requirements and application of required actions associated with the RVLC System. The RVLC system utilizes NR level and FW flow instruments to properly control RPV water level. If one level instrument differs from the average mean by 8 inches or more, the RVLCS will remove the anomalous input and control utilizing the remaining valid instruments. The RVLC halts the vessel rise after it removes LI
-94B from processing. SR 3.3.2.2.1 requires a channel check which is documented in 6.LOG.601 requiring these Narrow Range level indicators to indicate within 2 inches of each other. Since the B NR is greater than 2 inches from A & C, this channel is considered inoperable (failed channel check).
TS 3.3.2.2, Condition A is entered at the time the failure is identified and the required time to place the channel in a tripped condition is 7 days (March 17 at 1300). Once the completion time is expired, reactor power must be lowered below 25% RTP IAW Condition C within 4 hours. 7 days + 4 hours = March 17 at 1700.
11 Distracters:
A. This answer is incorrect because only 1 NR instrument being inoperable. This answer is plausible due to 2 NR level indicators being higher than normal setpoint (35") and reflects a 2 hour completion time. The candidate who incorrectly determines 2 level NR Level instruments are inoperable and confuses completion time to restore trip capability vs. reducing power to <25% would select this answer.
B. This answer is incorrect because only 1 NR instrument being inoperable. This answer is plausible due to 2 NR level indicators being higher than normal setpoint (35") and reflects a 2 + 4 hour (6 hrs) completion time. The candidate who incorrectly determines 2 level NR Level instruments are inoperable but correctly identifies the additional 4 hours to reducing power to <25% would select this answer.
C. This answer is incorrect due not applying the 4 additional hours allowed to reduce power below 25%. This answer is plausible due to 1 NR level indicator being lower than normal setpoint (35") with higher FW flow and reflects the 7 day completion time (if stem were changed to ask when required to be in trip  
- would be correct). The candidate who correctly determines 1 level NR Level instrument is inoperable but confuses placing the channel in trip vs. reducing power to <25% would select this answer.
Technical Reference(s):
Technical Reference(s):
Technical Specifications LCO 3.3.2.2, Feedwater and Main Turbine High Water Level Trip Instrumentation.
Technical Specifications LCO 3.3.2.2, Feedwater and Main Turbine High Water Level Trip Instrumentation.
Procedure 4.4.1 (Reactor Vessel Level Control System), Rev. 7
Procedure 4.4.1 (Reactor Vessel Level Control System), Rev. 7 Proposed references to be provided to applicants during examination:           TS 3.3.2.2 Learning Objective:
 
Proposed references to be provided to applicants during examination:
TS 3.3.2.2
 
Learning Objective:
INT007-05-04, OPS CNS Tech Specs 3.3, Instrumentation
INT007-05-04, OPS CNS Tech Specs 3.3, Instrumentation
: 3. Given a set of plant conditions that constitutes non
: 3. Given a set of plant conditions that constitutes non-compliance with a Section 3.3 LCO, determine the ACTIONS that are required.
-compliance with a Section 3.3 LCO, determine the ACTIONS that are required.
Question Source:
Question Source:
Bank #     Modified Bank #
Bank #
New X Question Cognitive Level:
Modified Bank #
Memory or Fundamental Knowledge Comprehension or Analysis X   10 CFR Part 55 Content:
New X Question Cognitive Level:     Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content:         55.41.10 55.43.2 55.43.5 55.45.13 Difficulty: 3 SRO Only - 10CFR55.43 b (2) Facility operating limitations in the technical specifications and their bases.
55.41.10 55.43.2 55.43.5     55.45.13 Difficulty: 3 SRO Only - 10CFR55.43 b (2) Facility operating limitations in the technical specifications and their bases.
Requires assessing instrument / plant response and application of TS required actions.
Requires assessing instrument / plant response and application of TS required actions.
12 13 14 Examination Outline Cross
12
-
 
13 14 Examination Outline Cross-


==Reference:==
==Reference:==


Level SRO Tier #   2 Group # 1 K/A # 262001A2.06 Importance Rating
Level                       SRO Tier #                       2 Group #                     1 K/A #               262001A2.06 Importance Rating           2.9 262001 A.C. Electrical Distribution A2        Ability to (a) predict the impacts of the following on the A.C. ELECTRICAL DISTRIBUTION; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
A2.06 Deenergizing a Plant Bus Question: 90 The plant is in Mode 3.
An event requires IMMEDIATELY de-energizing 4160V Bus 1A.
(1) What is the impact of IMMEDIATELY de-energizing 4160V Bus 1A?
(2) How many Off Site Power Circuits are Inoperable IMMEDIATELY following power transfer?
A. (1) A momentary loss of Bus 1F occurs.
(2) One B. (1) A momentary loss of Bus 1F occurs.
(2) Two C. (1) A momentary loss of Bus 1G occurs.
(2) One D. (1) A momentary loss of Bus 1G occurs.
(2) Two Answer:
A. (1) A momentary loss of Bus 1F occurs.
(2) One Explanation:
Requires predicting the impact of de-energizing Bus 1A and knowledge of TS 3.8.1 Bases for Off Site power source OPERABILITY which would then be utilized to control the consequences of this Bus loss IAW TSs. Conditions provided that the plant is in Mode 3. De-energizing Bus 1A without manually transferring Bus 1F prior to de-energizing Bus 1A results in 1 second time delay for auto transfer to the Emergency Station Service Transformer (ESST) (Impact = short duration de-energization). De-energizing Bus 1A results in loss of 1 offsite circuit which is the Startup Station Service Transformer (SSST) because it no longer has fast transfer capability to both 4160V Critical Buses (1F and 1G). The ESST remains operable because it is powering the 4160V Critical Bus 1F (fast transfer complete) and it maintains its fast transfer capability to the other division Critical Bus 1G. When a 4160V critical bus is manually transferred to the ESST IAW Procedure 2.2.18, the opposite divisions ESST supply breaker is placed in pull-to-lock requiring declaring both Off Site power circuits inoperable. The difference between manual transfer and automatic transfer is a common misconception for operability of the Off Site power circuits. Procedurally de-energizing Bus 1A is only done with the plant in Mode 4 or 5, but due 15


===2.9 262001===
to the emergent nature of this event, the bus would be de-energized by opening supply breakers and then followed up by performing applicable steps of procedure 2.2.18.
A.C. Electrical Distribution A2 Ability to (a) predict the impacts of the following on the A.C. ELECTRICAL DISTRIBUTION; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
Distracters:
A2.06 Deenergizing a Plant Bus Question:  90
B. This answer is incorrect because ONLY the SSST is inoperable. This choice is plausible due to the common misconception of OPERABLE Off Site circuits and if de-energized IAW Procedure 2.2.18 requires declaring TWO offsite circuits INOPERABLE. The candidate that correctly recognizes momentary loss of power to Bus 1F and knows both Off Site circuits are declared inoperable due to breaker alignment during manual transfer per procedure would select this option.
 
C. This answer is incorrect due to Bus 1G not being impacted by de-energizing Bus 1A. This choice is plausible if electric plant alignment is not known or confused. Candidate that confuses momentary loss of power to Bus 1F and correctly identifies the number of INOPERABLE Off Site circuits would select this option.
The plant is in Mode 3.
D. This answer is incorrect due to Bus 1G not being impacted by de-energizing Bus 1A and ONLY the SSST is inoperable. This choice is plausible if electric plant alignment is not known or confused and the common misconception of OPERABLE Off Site circuits and if de-energized IAW Procedure 2.2.18 requires declaring TWO offsite circuits INOPERABLE.
An event requires IMMEDIATELY de
Candidate that confuses momentary loss of power to Bus 1F and does not recognize the correct number of OPERABLE offsite circuits or knows both Off Site circuits are declared inoperable due to breaker alignment during manual transfer per procedure would select this option.
-energizing 4160V Bus 1A.
Technical Reference(s):
  (1)  What is the impact of IMMEDIATELY de
-energizing 4160V Bus 1A?
(2)  How many Off Site Power Circuits are Inoperable IMMEDIATELY following power transfer?
 
A.  (1) A momentary loss of Bus 1F occurs.
  (2) One  B.  (1) A momentary loss of Bus 1F occurs.
  (2) Two C.  (1) A momentary loss of Bus 1G occurs.
  (2) One  D.  (1) A momentary loss of Bus 1G occurs.
  (2) Two  Answer: A.  (1) A momentary loss of Bus 1F occurs.  (2) One  Explanation: 
 
Requires predicting the impact of de
-energizing Bus 1A and knowledge of TS 3.8.1 Bases for Off Site power source OPERABILITY which would then be utilized to control the consequences of this Bus loss IAW TSs. Conditions provided that the plant is in Mode 3. De
-energizing Bus 1A without manually transferring Bus 1F prior to de
-energizing Bus 1A results in 1 second time delay for auto transfer to the Emergency Station Service Transformer  (ESST) (Impact = short duration de
-energization). De
-energizing Bus 1A results in loss of 1 offsite circuit which is the Startup Station Service Transformer (SSST) because it no longer has fast transfer capability to both 4160V Critical Buses (1F and 1G). The ESST remains operable because it is powering the 4160V Critical Bus 1F (fast transfer complete) and it maintains its fast transfer capability to the other division Critical Bus 1G. When a 4160V critical bus is manually transferred to the ESST IAW Procedure 2.2.18, the opposite division's ESST supply breaker is placed in pull
-to-lock requiring declaring both Off Site power circuits inoperable. The difference between manual transfer and automatic transfer is a common misconception for operability of the Off Site power circuits. Procedurally de-energizing Bus 1A is only done with the plant in Mode 4 or 5, but due 15 to the emergent nature of this event, the bus would be de
-energized by opening supply breakers and then followed up by performing applicable steps of procedure 2.2.18.
Distracters: B. This answer is incorrect because ONLY the SSST is inoperable. This choice is plausible due to the common misconception of OPERABLE Off Site circuits and if de
-energized IAW Procedure 2.2.18 requires declaring TWO offsite circuits INOPERABLE. The candidate that correctly recognizes momentary loss of power to Bus 1F and knows both Off Site circuits are declared inoperable due to breaker alignment during manual transfer per procedure would select this option.
C. This answer is incorrect due to Bus 1G not being impacted by de
-energizing Bus 1A. This choice is plausible if electric plant alignment is not known or confused. Candidate that confuses momentary loss of power to Bus 1F and correctly identifies the number of INOPERABLE Off Site circuits would select this option.
 
D. This answer is incorrect due to Bus 1G not being impacted by de
-energizing Bus 1A and ONLY the SSST is inoperable. This choice is plausible if electric plant alignment is not known or confused and the common misconception of OPERABLE Off Site circuits and if de-energized IAW Procedure 2.2.18 requires declaring TWO offsite circuits INOPERABLE. Candidate that confuses momentary loss of power to Bus 1F and does not recognize the correct number of OPERABLE offsite circuits or knows both Off Site circuits are declared inoperable due to breaker alignment during manual transfer per procedure would select this option. Technical Reference(s):
Procedure 2.2.18 (4160V Auxiliary Power Distribution System), Revision 178.
Procedure 2.2.18 (4160V Auxiliary Power Distribution System), Revision 178.
TS LCO 3.8.1 Proposed references to be provided to applicants during examination:
TS LCO 3.8.1 Proposed references to be provided to applicants during examination:           NONE Learning Objective:
NONE Learning Objective:
INT00705090010300 Given a set of plant conditions that constitutes non-compliance with a Section 3.8 LCO, determine the ACTIONS that are required.
INT00705090010300 Given a set of plant conditions that constitutes non
Question Source:
-compliance with a Section 3.8 LCO, determine the ACTIONS that are required.
Bank #
Question Source:   Bank #     Modified Bank #
Modified Bank #
New X   Question Cognitive Level:
New X Question Cognitive Level:       Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content:         55.41.5 55.45.6 Difficulty: 3 SRO Only - 10 CFR 55.43(b)(2) - Facility operating limitations in the TS and their bases.
Memory or Fundamental Knowledge Comprehension or Analysis X   10 CFR Part 55 Content:
55.41.5 55.45.6 Difficulty: 3 SRO Only - 10 CFR 55.43(b)(2)  
- Facility operating limitations in the TS and their bases.
Requires knowledge of TS Bases for determination of Off Site circuit INOPERABILITY.
Requires knowledge of TS Bases for determination of Off Site circuit INOPERABILITY.
16 17 18 19 20 21 22 Examination Outline Cross
16
-
 
17 18 19 20 21 22 Examination Outline Cross-


==Reference:==
==Reference:==


Level SRO Tier #   2 Group # 2 K/A # 201006G2.1.23 Importance Rating
Level                     SRO Tier #                   2 Group #                   2 K/A #             201006G2.1.23 Importance Rating       4.4 201006 Rod Worth Minimizer System 2.1.23 Ability to perform specific system and integrated plant procedures during all modes of plant operation.
 
Question: 91 6.RWM.301 (Rod Worth Minimizer Functional Test For Startup) is being performed.
===4.4 201006===
Which of the following completes the statements below regarding how the Select Error function of the RWM is tested IAW 6.RWM.301 AND the significance of maintaining the RWM operable during startup IAW TS Bases?
Rod Worth Minimizer System 2.1.23 Ability to perform specific system and integrated plant procedures during all modes of plant operation.
The operator is required to verify the SELECT ERROR indicator turns red while selecting
Question: 91 6.RWM.301 (Rod Worth Minimizer Functional Test For Startup) is being performed.
____(1)____ from each RWM group (except Group 1).
 
Which of the following completes the statements below regarding how the "Select Error" function of the RWM is tested IAW 6.RWM.301 AND the significance of maintaining the RWM operable during startup IAW TS Bases?
The operator is required to verify the SELECT ERROR indicator turns red while selecting ____(1)____ from each RWM group (except Group 1).
 
The RWM enforces compliance with BPWS which ensures that the initial conditions of the analysis for a ____(2)____ is NOT violated.
The RWM enforces compliance with BPWS which ensures that the initial conditions of the analysis for a ____(2)____ is NOT violated.
 
A. (1) a single rod (2) control rod drop accident B. (1) a single rod (2) single control rod withdrawal error C. (1) ALL individual rods (2) control rod drop accident D. (1) ALL individual rods (2) single control rod withdrawal error Answer:
A. (1) a single rod (2) control rod drop accident B. (1) a single rod (2) single control rod withdrawal error  
A. (1) a single rod (2) control rod drop accident Explanation: Requires knowledge of RWM Functional Test and TS 3.3.2.1 Bases. 6.RWM.301 (Rod Worth Minimizer Functional Test For Startup) requires the operator to verify the SELECT ERROR functions (indicator on PMIS turns red) when selecting a single rod from each RWM group (except Group 1). The RWM enforces BPWS which ensures that the initial conditions of the CRDA analysis are not violated.
 
C. (1) ALL individual rods (2) control rod drop accident D. (1) ALL individual rods (2) single control rod withdrawal error Answer: A. (1) a single rod (2) control rod drop accident Explanation: Requires knowledge of RWM Functional Test and TS 3.3.2.1 Bases. 6.RWM.301 (Rod Worth Minimizer Functional Test For Startup) requires the operator to verify the SELECT ERROR functions (indicator on PMIS turns red) when selecting a single rod from each RWM group (except Group 1). The RWM enforces BPWS which ensures that the initial conditions of the CRDA analysis are not violated.
Distracters:
Distracters:
B. This answer is incorrect due to a single control rod withdrawal error (RWE) not being the TS bases for the RWM. This choice is plausible due to a CRDA being easily confused with a RWE which is the bases for the Rod Block Monitor. The candidate that correctly identifies selecting only 1 rod in the remaining groups and confuses RBM vs. RWM bases would select this choice.
B. This answer is incorrect due to a single control rod withdrawal error (RWE) not being the TS bases for the RWM. This choice is plausible due to a CRDA being easily confused with a RWE which is the bases for the Rod Block Monitor. The candidate that correctly identifies selecting only 1 rod in the remaining groups and confuses RBM vs. RWM bases would select this choice.
23 C. This answer is incorrect due to selecting all rods in the remaining groups not being required to support SELECT ERROR function verification. This choice is plausible due having many groups requiring single rod selection being easily confused with all the rods within each group. The candidate that confuses only 1 rod in the remaining groups vs. all rods and correctly identifies the RWM bases would select this choice.
23


C. This answer is incorrect due to selecting all rods in the remaining groups not being required to support SELECT ERROR function verification. This choice is plausible due having many groups requiring single rod selection being easily confused with all the rods within each group. The candidate that confuses only 1 rod in the remaining groups vs. all rods and correctly identifies the RWM bases would select this choice.
D. This answer is incorrect due to selecting all rods in the remaining groups not being required to support SELECT ERROR function verification and a single control rod withdrawal error (RWE) not being the TS bases for the RWM. This choice is plausible due having many groups requiring single rod selection being easily confused with all the rods within each group and a CRDA being easily confused with a RWE which is the bases for the Rod Block Monitor. The candidate that confuses only 1 rod in the remaining groups vs. all rods and confuses RBM vs. RWM bases would select this choice.
D. This answer is incorrect due to selecting all rods in the remaining groups not being required to support SELECT ERROR function verification and a single control rod withdrawal error (RWE) not being the TS bases for the RWM. This choice is plausible due having many groups requiring single rod selection being easily confused with all the rods within each group and a CRDA being easily confused with a RWE which is the bases for the Rod Block Monitor. The candidate that confuses only 1 rod in the remaining groups vs. all rods and confuses RBM vs. RWM bases would select this choice.
Technical Reference(s):
Technical Reference(s):
Procedure RWM.301 (Rod Worth Minimizer Functional Test For Startup), Rev. 10.
Procedure RWM.301 (Rod Worth Minimizer Functional Test For Startup), Rev. 10.
TS Bases   Proposed references to be provided to applicants during examination:
TS Bases Proposed references to be provided to applicants during examination:         None Learning Objective:
None Learning Objective:
INT007-05-04:
INT007-05-04: 2. Discuss the applicable Safety Analysis in the Bases associated with each Section 3.3 Specification.
: 2. Discuss the applicable Safety Analysis in the Bases associated with each Section 3.3 Specification.
Question Source:
Question Source:
Bank #     Modified Bank #
Bank #
New X   Question Cognitive Level:
Modified Bank #
Memory or Fundamental Knowledge   Comprehension or Analysis X   10 CFR Part 55 Content:
New X Question Cognitive Level:       Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content:         55.41.10 55.43.5 55.45.2 55.45.6 Difficulty: 3 SRO Only - 10CFR55.43 b (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
55.41.10 55.43.5 55.45.2 55.45.6 Difficulty: 3 SRO Only - 10CFR55.43 b (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Requires knowledge of TS Surveillance Requirements and Bases. It is the SRO who determines how much of a system is required to be tested. In this case, the SRO must know that testing only the first control rod of each RWM group must be tested.
Requires knowledge of TS Surveillance Requirements and Bases. It is the SRO who determines how much of a system is required to be tested. In this case, the SRO must know that testing only the first control rod of each RWM group must be tested. 24 25 26 27 Examination Outline Cross
24
-
 
25 26 27 Examination Outline Cross-


==Reference:==
==Reference:==


Level SRO Tier #   2 Group # 2 K/A # 202002A2.06 Importance Rating
Level                   SRO Tier #                   2 Group #                 2 K/A #             202002A2.06 Importance Rating       3.3 202002 Recirculation flow Control System A2         Ability to (a) predict the impacts of the following on the RECIRCULATION FLOW CONTROL SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
 
A2.06 Low reactor water level: Plant-Specific Question: 92 The plant is operating in MODE 1 at 97% power when RFP A trips.
===3.3 202002===
(1) How is the Recirculation Flow Control System affected if RPV level lowers to 25 inches on NR Level instruments?
Recirculation flow Control System A2 Ability to (a) predict the impacts of the following on the RECIRCULATION FLOW CONTROL SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
(2) If RRMG B speed does not change and CANNOT be reduced, what attachment provides direction to trip RRMG B with RPV level stable at 20 inches on NR Level instruments IAW 2.4RR (Reactor Recirculation Abnormal)?
A2.06 Low reactor water level: Plant
A. (1) Both Recirculation Pumps run back to 22% speed.
-Specific Question: 92
(2) Attachment 1 {Trip of Reactor Recirculation Pump(s)}.
 
B. (1) Both Recirculation Pumps run back to 22% speed.
The plant is operating in MODE 1 at 97% power when RFP A trips.
(2) Attachment 4 (Reactor Recirculation Flow Control Failure/RRMG Scoop Tube Lockout).
  (1) How is the Recirculation Flow Control System affected if RPV level lowers to 25 inches on NR Level instruments?  
C. (1) Both Recirculation Pumps run back towards 45% speed.
(2) If RRMG B speed does not change and CANNOT be reduced, what attachment provides direction to trip RRMG B with RPV level stable at 20 inches on NR Level instruments IAW 2.4RR (Reactor Recirculation Abnormal)?
(2) Attachment 1 {Trip of Reactor Recirculation Pump(s)}.
A. (1) Both Recirculation Pumps run back to 22% speed.
D. (1) Both Recirculation Pumps run back towards 45% speed.
  (2) Attachment 1 {Trip of Reactor Recirculation Pump(s)}.
(2) Attachment 4 (Reactor Recirculation Flow Control Failure/RRMG Scoop Tube Lockout).
B. (1) Both Recirculation Pumps run back to 22% speed.
Answer:
  (2) Attachment 4 (Reactor Recirculation Flow Control Failure/RRMG Scoop Tube Lockout).
D. (1) Both Recirculation Pumps run back towards 45% speed.
 
(2) Attachment 4 (Reactor Recirculation Flow Control Failure/RRMG Scoop Tube Lockout).
C. (1) Both Recirculation Pumps run back towards 45% speed.
  (2) Attachment 1
{Trip of Reactor Recirculation Pump(s)}.
D. (1) Both Recirculation Pumps run back towards 45% speed.
  (2) Attachment 4 (Reactor Recirculation Flow Control Failure/RRMG Scoop Tube Lockout).
Answer: D. (1) Both Recirculation Pumps run back towards 45% speed.
  (2) Attachment 4 (Reactor Recirculation Flow Control Failure/RRMG Scoop Tube Lockout).
Explanation:
Explanation:
lf both Reactor Recirculation (RR) pumps are running and not locked out, RR pumps run back towards 45% speed if the following condition is met:
lf both Reactor Recirculation (RR) pumps are running and not locked out, RR pumps run back towards 45% speed if the following condition is met:
Total steam flow >
Total steam flow > 9 Mlbm/hr with at least 1 RFP tripped/flow < 1 Mlbm/hr and selected reactor water level < 27.5 inches. Both RR pumps runback towards 45% and stops when the condition causing the runback is no longer true or no other 45% runback conditions exist. With power at 97%, FW flow is around 9.27 Mlbm/hr. As the RR pumps start running back FW flow will lower below 9 Mlbm/hr and runback will stop. If the stem were changed to reflect Discharge valve closure or FW Flow <20% - 22% runback would be correct.
9 Mlbm/hr with at least 1 RFP tripped/flow < 1 Mlbm/hr and selected reactor water level < 27.5 inches. Both RR pumps runback towards 45% and stops when the condition causing the runback is no longer true or no other 45% runback conditions exist. With power at 97%, FW flow is around 9.27 Mlbm/hr. As the RR pumps start running back FW flow will lower below 9 Mlbm/hr and runback will stop. If the stem were changed to reflect Discharge valve closure or FW Flow <20%  
On a trip of a RFP entry into Abnormal Procedure 2.4MC-RF, Condensate and Feedwater Abnormal is required. Additionally the inability to change speed of RRMG B requires entry into 2.4RR (Reactor Recirculation Abnormal). With RPV level not recovering, Attachment 4 (Reactor Recirculation Flow Control Failure/RRMG Scoop Tube Lockout) is required to be performed which directs tripping the RRMG B if a faster power reduction is needed.
- 22% runback would be correct.
28
On a trip of a RFP entry into Abnormal Procedure 2.4MC
 
-RF, Condensate and Feedwater Abnormal is required. Additionally the inability to change speed of RRMG B requires entry into 2.4RR (Reactor Recirculation Abnormal). With RPV level not recovering, Attachment 4 (Reactor Recirculation Flow Control Failure/RRMG Scoop Tube Lockout) is required to be performed which directs tripping the RRMG B if a faster power reduction is needed.
Distractors:
28 Distractors:
A. This answer is incorrect because both RR pumps run back towards 45% speed and Attachment 1 not providing guidance to trip RRMG B. Once the condition that caused the runback (lowering FW flow) is clear the RR pumps stop running back. This answer is plausible because the RR pumps would runback to 22% if the Discharge valve closes or FW Flow lowers below 20% and Attachment 1 titled Trip of RR Pump(s). Attachment 1 would be required to be entered following the trip of RRMG B. The candidate that confuses RR Runback logic and is not familiar with 2.4RR attachments and thinks tripping RRMG B IAW Attachment 1 would select this answer.
A. This answer is incorrect because both RR pumps run back towards 45% speed and Attachment 1 not providing guidance to trip RRMG B. Once the condition that caused the runback (lowering FW flow) is clear the RR pumps stop running back. This answer is plausible because the RR pumps would runback to 22% if the Discharge valve closes or FW Flow lowers below 20% and Attachment 1 titled Trip of RR Pump(s). Attachment 1 would be required to be entered following the trip of RRMG B. The candidate that confuses RR Runback logic and is not familiar with 2.4RR attachments and thinks tripping RRMG B IAW Attachment 1 would select this answer.
 
B. This answer is incorrect because both RR pumps run back towards 45% speed. Once the condition that caused the runback (lowering FW flow) is clear the RR pumps stop running back. This answer is plausible because the RR pumps would runback to 22% if the Discharge valve closes or FW Flow lowers below 20%. The candidate that confuses RR Runback logic and is familiar with 2.4RR attachments would select this answer.
B. This answer is incorrect because both RR pumps run back towards 45% speed. Once the condition that caused the runback (lowering FW flow) is clear the RR pumps stop running back. This answer is plausible because the RR pumps would runback to 22% if the Discharge valve closes or FW Flow lowers below 20%. The candidate that confuses RR Runback logic and is familiar with 2.4RR attachments would select this answer.  
C. This answer is incorrect because Attachment 1 does not provide guidance to trip RRMG B.
 
This answer is plausible because Attachment 1 titled Trip of RR Pump(s). Attachment 1 would be required to be entered following the trip of RRMG B. The candidate that correctly identifies RR Runback logic and is not familiar with 2.4RR attachments and thinks tripping RRMG B IAW Attachment 1 would select this answer.
C. This answer is incorrect because Attachment 1 does not provide guidance to trip RRMG B. This answer is plausible because Attachment 1 titled Trip of RR Pump(s). Attachment 1 would be required to be entered following the trip of RRMG B. The candidate that correctly identifies RR Runback logic and is not familiar with 2.4RR attachments and thinks tripping RRMG B IAW Attachment 1 would select this answer.
Technical Reference(s):
Technical Reference(s):
Procedure 2.2.68, Reactor Recirculation System Operations, Rev 77.
Procedure 2.2.68, Reactor Recirculation System Operations, Rev 77.
Procedure 2.4MC
Procedure 2.4MC-RF, Condensate and Feedwater Abnormal, Rev 14 Procedure 2.4RR, Reactor Recirculation Abnormal, Rev 40.
-RF, Condensate and Feedwater Abnormal, Rev 14 Procedure 2.4RR, Reactor Recirculation Abnormal, Rev 40.
Proposed references to be provided to applicants during examination:         None Learning Objective:
Proposed references to be provided to applicants during examination:
None Learning Objective:  
: 4. Describe the interrelationships between the Reactor Recirculation system or the Recirculation Flow Control system and the following:
: 4. Describe the interrelationships between the Reactor Recirculation system or the Recirculation Flow Control system and the following:
: j. Reactor water level/pressure
: j. Reactor water level/pressure
Line 4,021: Line 3,679:
Bank #
Bank #
Modified Bank #
Modified Bank #
New X   Question Cognitive Level:
New X Question Cognitive Level:               Memory or Fundamental Knowledge           X Comprehension or Analysis 10 CFR Part 55 Content:         41.5 45.6 Difficulty: 3 SRO Only - 10CFR55.43 b (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Memory or Fundamental Knowledge X Comprehension or Analysis     10 CFR Part 55 Content:
29
41.5 45.6 Difficulty:
 
3 SRO Only - 10CFR55.43 b (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. 29 Assessing plant conditions and selecting a procedure with which to proceed. The SRO assesses plant conditions and determines whether RPV level will or will not recover and make priority decisions about responding to a plant transient.
Assessing plant conditions and selecting a procedure with which to proceed. The SRO assesses plant conditions and determines whether RPV level will or will not recover and make priority decisions about responding to a plant transient.
30 31 32 33 34 Examination Outline Cross
30
-
 
31 32 33 34 Examination Outline Cross-


==Reference:==
==Reference:==


Level SRO Tier #   2 Group # 2 K/A # 271000 G2.4.20 Importance Rating
Level                     SRO Tier #                   2 Group #                   2 K/A #             271000G2.4.20 Importance Rating       4.3 271000 Offgas System 2.4.20 Knowledge of operational implications of EOP warnings, cautions, and notes.
 
(CFR: 41.10 / 43.5 / 45.13)
===4.3 271000===
Question: 93 The following conditions exist following an earthquake:
Offgas System 2.4.20 Knowledge of operational implications of EOP warnings, cautions, and notes.
* The reactor is scrammed and all control rods fully insert.
(CFR: 41.10 / 43.5 / 45.13) Question: 93 The following conditions exist following an earthquake:
* SRVs cannot to be opened for pressure control.
The reactor is scrammed and all control rods fully insert.
* No means of RPV injection are available.
SRVs cannot to be opened for pressure control.
* RPV water level is -150 inches corrected fuel zone and steady.
No means of RPV injection are available.
* One MSL remains open.
RPV water level is  
* RPV pressure is 950 psig.
-150 inches corrected fuel zone and steady.
* Primary Containment pressure is 0.45 psig.
One MSL remains open.
* Torus water level is 9.5 feet and lowering 0.2 feet/minute.
RPV pressure is 950 psig.
* The TSC is not operational The CRS intends to begin using the Steam Jet Air Ejectors (SJAEs).
Primary Containment pressure is 0.45 psig.
(1) What is the required CRS response?
Torus water level is 9.5 feet and lowering 0.2 feet/minute.
(2) What CAUTION is of concern when carrying out this response?
The TSC is not operational The CRS intends to begin using the Steam Jet Air Ejectors (SJAEs).  
  (1) What is the required CRS response?
(2) What CAUTION is of concern when carrying out this response?
A. (1) Transition to EOP 2A and direct Emergency Depressurization.
A. (1) Transition to EOP 2A and direct Emergency Depressurization.
  (2) When placing SJAEs in service, caution of sending personnel through potentially high radiation areas must be addressed.
(2) When placing SJAEs in service, caution of sending personnel through potentially high radiation areas must be addressed.
B. (1) Transition to EOP 2A and direct Emergency Depressurization.
B. (1) Transition to EOP 2A and direct Emergency Depressurization.
  (2) When placing SJAEs in service, caution of forcing the shutter slide on the breaker could cause damage to the breaker preventing closure.
(2) When placing SJAEs in service, caution of forcing the shutter slide on the breaker could cause damage to the breaker preventing closure.
 
C. (1) Transition to EOP 2A and direct Steam Cooling.
C. (1) Transition to EOP 2A and direct Steam Cooling.
  (2) When placing SJAEs in service, caution of sending personnel through potentially high radiation areas must be addressed.
(2) When placing SJAEs in service, caution of sending personnel through potentially high radiation areas must be addressed.
D. (1) Transition to EOP 2A and direct Steam Cooling.
D. (1) Transition to EOP 2A and direct Steam Cooling.
  (2) When placing SJAEs in service, caution of forcing the shutter slide on the breaker could cause damage to the breaker preventing closure.
(2) When placing SJAEs in service, caution of forcing the shutter slide on the breaker could cause damage to the breaker preventing closure.
Answer: A. (1) Transition to EOP 2A and direct Emergency Depressurization.
Answer:
  (2) When placing SJAEs in service, caution of sending personnel through potentially high radiation areas must be addressed.
A. (1) Transition to EOP 2A and direct Emergency Depressurization.
35 Explanation:
(2) When placing SJAEs in service, caution of sending personnel through potentially high radiation areas must be addressed.
The CRS must direct Emergency Depressurization due to EOP 3A direction on torus water level being below 9.6' of water. When the downcomers become uncovered, the drywell and torus pressure suppression function is lost because any possible future steam leakage into the drywell will not be directed below the torus water level and condensed. The result could be primary containment failure due to overpressure. Emergency depressurization is the primary action to take with the given conditions. To accomplish the ED, the steam jet air ejectors are going to be used. The CRS is knowledgeable of all the plant conditions and can prioritize which system(s) is/are to be used and the priority placed on the order to use the systems. EOPs allow using one or all of the listed systems and the CRS directs placing as many in service that is required to perform the task of lowering RPV pressure less than 50 psig below tours space pressure. The CRS must be knowledgeable about the CAUTIONS applicable to which system to use and prioritize based upon procedure guidance and plant conditions. At this point ED is complete and the RPV will not again pressurize.
35


Explanation:
The CRS must direct Emergency Depressurization due to EOP 3A direction on torus water level being below 9.6' of water. When the downcomers become uncovered, the drywell and torus pressure suppression function is lost because any possible future steam leakage into the drywell will not be directed below the torus water level and condensed. The result could be primary containment failure due to overpressure. Emergency depressurization is the primary action to take with the given conditions. To accomplish the ED, the steam jet air ejectors are going to be used. The CRS is knowledgeable of all the plant conditions and can prioritize which system(s) is/are to be used and the priority placed on the order to use the systems. EOPs allow using one or all of the listed systems and the CRS directs placing as many in service that is required to perform the task of lowering RPV pressure less than 50 psig below tours space pressure. The CRS must be knowledgeable about the CAUTIONS applicable to which system to use and prioritize based upon procedure guidance and plant conditions. At this point ED is complete and the RPV will not again pressurize.
Distracters:
Distracters:
B, This answer is incorrect because the CAUTION of concern is not correct. There are no breaker trip actions to take when placing the SJAEs in service. This caution is based upon placing the AOG third stage SJAEs in service which is contained in the same procedure.
The guidance is contained in the section of EOP procedure 5.8.2 which is knowledge the SRO is required to know but is not general RO knowledge based upon the placement being well inside the procedure guidance. This answer is plausible because the action to transfer to EOP 2A and emergency depressurize is required. The second part is plausible because it is a caution that is in the procedure for placing a like system in service.
C. This answer is incorrect because the requirement to steam cool is not met. RPV level is steady and as long as level can be maintained above -183 inches with no means of injection available. The second part of the answer is correct because there is a potential of sending personnel through a high radiation field when performing the task. It is the SROs responsibility to determine if personnel safety risk merits performing the task. The SRO has determined it safe as inferred in the question stem. This answer is plausible because the correct caution is stated D. This answer is incorrect because both parts of the answer are incorrect. The correct action is to emergency depressurize and the correct caution to regard is sending the personnel through the potential high radiation field. This answer is plausible because the actions given are correct if given different circumstances. The same procedure is used for using the AOG third stage SJAEs in service and this system can be used as an emergency depressurization system. The SRO must know detailed procedure guidance to realize there are no breaker trips with the slide shutter when using the SJAEs.
36


B, This answer is incorrect because the CAUTION of concern is not correct. There are no breaker trip actions to take when placing the SJAEs in service. This caution is based upon placing the AOG third stage SJAEs in service which is contained in the same procedure. The guidance is contained in the section of EOP procedure 5.8.2 which is knowledge the SRO is required to know but is not general RO knowledge based upon the placement being well inside the procedure guidance. This answer is plausible because the action to transfer to EOP 2A and emergency depressurize is required. The second part is plausible because it is a caution that is in the procedure for placing a like system in service.
Technical Reference(s):
C. This answer is incorrect because the requirement to steam cool is not met. RPV level is steady and as long as level can be maintained above
PSTG/SATG AMP-TBD00 rev 8 APDX B EOP 5.8.2, RPV Depressurization Systems (Table 2), Rev. 40 Proposed references to be provided to applicants during examination:       NONE Learning Objective:
-183 inches with no means of injection available. The second part of the answer is correct because there is a potential of sending personnel through a high radiation field when performing the task. It is the SROs responsibility to determine if personnel safety risk merits performing the task. The SRO has determined it safe as inferred in the question stem. This answer is plausible because the correct caution is stated D. This answer is incorrect because both parts of the answer are incorrect. The correct action is to emergency depressurize and the correct caution to regard is sending the personnel through the potential high radiation field. This answer is plausible because the actions given are correct if given different circumstances. The same procedure is used for using the AOG third stage SJAEs in service and this system can be used as an emergency depressurization system. The SRO must know detailed procedure guidance to realize there are no breaker trips with the slide shutter when using the SJAEs.
36 Technical Reference(s):   PSTG/SATG AMP-TBD00 rev 8 APDX B EOP 5.8.2, RPV Depressurization Systems (Table 2), Rev. 40 Proposed references to be provided to applicants during examination:
NONE Learning Objective:
: 6. Identify any EOP support procedure addressed in Flowchart 2A and apply any associated special operating instructions or cautions.
: 6. Identify any EOP support procedure addressed in Flowchart 2A and apply any associated special operating instructions or cautions.
: 7. Given plant conditions and EOP flowchart 2A, EMERGENCY RPV DEPRESSURIZATION / STEAM COOLING, determine required actions.
: 7. Given plant conditions and EOP flowchart 2A, EMERGENCY RPV DEPRESSURIZATION / STEAM COOLING, determine required actions.
Question Source:
Question Source:
Bank #     Modified Bank #
Bank #
New X   Question Cognitive Level:
Modified Bank #
Memory or Fundamental Knowledge Comprehension or Analysis X   10 CFR Part 55 Content:
New X Question Cognitive Level:   Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content:       55.41.10 55.43.5 55.45.13 LOD 3 SRO Only - 10CFR55.43 b (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
55.41.10 55.43.5 55.45.13 LOD 3 SRO Only - 10CFR55.43 b (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Knowledge of diagnostic steps and decision points in the emergency operating procedures (EOP) that involve transitions to event specific sub-procedures or emergency contingency procedures.
Knowledge of diagnostic steps and decision points in the emergency operating procedures (EOP) that involve transitions to event specific sub
37
-procedures or emergency contingency procedures.
 
37 38 39 40 41 42 43 44 Examination Outline Cross
38 39 40 41 42 43 44 Examination Outline Cross-
-


==Reference:==
==Reference:==


Level SRO Tier #   3 Group # 1 K/A # 2.1.4 Importance Rating 3.8     2.1.4 Knowledge of individual licensed operator responsibilities related to shift staffing, such as medical requirements, "no
Level                       SRO Tier #                       3 Group #                     1 K/A #                       2.1.4 Importance Rating           3.8 2.1.4     Knowledge of individual licensed operator responsibilities related to shift staffing, such as medical requirements, no-solo operation, maintenance of active license status, 10CFR55, etc.
-solo" operation, maintenance of active license status, 10CFR55, etc.
Question: 94 Which of the following completes the statements below regarding SRO License requirements per CNS procedures?
Question: 94 Which of the following completes the statements below regarding SRO License requirements per CNS procedures?
A MINIMUM of ____(1)____ 12-hour shifts under instruction is required to allow taking the watch after a four month absence from watchstanding.
A MINIMUM of ____(1)____ 12
A Special Prescription Respirator Glasses Verification is required IAW ____(2)____?
-hour shifts under instruction is required to allow taking the watch after a four month absence from watchstanding.
A. (1) four (2) NTP8.1 (Administration of Licensed Operator Medical Examination Program)
A "Special Prescription Respirator Glasses Verification" is required IAW ____(2)____?
B. (1) four (2) 2.0.7 (Licensed Operator Active/Reactivation/Medical Status Maintenance Program)
A. (1) four (2) NTP8.1 (Administration of Licensed Operator Medical Examination Program)
C. (1) five (2) NTP8.1 (Administration of Licensed Operator Medical Examination Program)
D. (1) five (2) 2.0.7 (Licensed Operator Active/Reactivation/Medical Status Maintenance Program)
Answer:
B. (1) four (2) 2.0.7 (Licensed Operator Active/Reactivation/Medical Status Maintenance Program)
Explanation:
Requires knowledge that four 12 hour shift watches are required to support license reactivation.
In order to maintain a license active, the operator is required to perform five twelve hour shifts (seven 8 hour shifts) per calendar quarter. With the license away from watchstanding duties for four months the license is no longer active. The license must be reactivated which requires four 12 hour shift watches under instruction. It is plausible to choose five since this is required to support license maintenance (five 12 hour shifts). The Special Prescription Respirator Glasses Verification is contained on the SRO On-Shift Time & Reactivation Attachments in procedure 2.0.7. If a change in medical condition were to occur, Procedure 2.0.7 directs licensee to complete Attachment 2 of Procedure NTP8.1. It is plausible to choose procedure NTP8.1 due 45


B.  (1)  four (2)  2.0.7 (Licensed Operator Active/Reactivation/Medical Status Maintenance Program)
being titled Medical Status Maintenance Program and addressing prescription medication changes..
C.  (1)  five (2)  NTP8.1 (Administration of Licensed Operator Medical Examination Program)
Distracters:
 
D.  (1)  five (2)  2.0.7 (Licensed Operator Active/Reactivation/Medical Status Maintenance Program)  Answer:  B.  (1)  four (2)  2.0.7 (Licensed Operator Active/Reactivation/Medical Status Maintenance Program)  Explanation:
Requires knowledge that four 12 hour shift watches are required to support license reactivation. In order to maintain a license active, the operator is required to perform five twelve hour shifts (seven 8 hour shifts) per calendar quarter. With the license away from watchstanding duties for four months the license is no longer active. The license must be reactivated which requires four 12 hour shift watches under instruction. It is plausible to choose five since this is required to support license maintenance (five 12 hour shifts). The Special Prescription Respirator Glasses Verification is contained on the SRO On
-Shift Time & Reactivation Attachments in procedure 2.0.7. If a change in medical condition were to occur, Procedure 2.0.7 directs licensee to complete Attachment 2 of Procedure NTP8.1. It is plausible to choose procedure NTP8.1 due 45 being titled "Medical Status Maintenance Program" and addressing prescription medication changes.. Distracters:
A. This answer is incorrect because NTP8.1 does not contain the Special Prescription Respirator Glasses Verification. This choice is plausible because this procedure is required to support change in licensed operator medical status. The candidate who correctly identifies the number of watches required to reactivate and confuses Respirator Glasses Verification with a change in license medical status would select this answer.
A. This answer is incorrect because NTP8.1 does not contain the Special Prescription Respirator Glasses Verification. This choice is plausible because this procedure is required to support change in licensed operator medical status. The candidate who correctly identifies the number of watches required to reactivate and confuses Respirator Glasses Verification with a change in license medical status would select this answer.
C. This answer is incorrect because the number of shifts stated is incorrect and NTP8.1 does not contain the Special Prescription Respirator Glasses Verification guidance. This choice is plausible due 5 shifts being required for initial license activation and quarterly license maintenance and this procedure being required to support change in licensed operator medical status. The candidate who confuses the number of watches required to activate/maintain vs. reactivate and confuses Respirator Glasses Verification with a change in license medical status would select this answer.
C. This answer is incorrect because the number of shifts stated is incorrect and NTP8.1 does not contain the Special Prescription Respirator Glasses Verification guidance. This choice is plausible due 5 shifts being required for initial license activation and quarterly license maintenance and this procedure being required to support change in licensed operator medical status. The candidate who confuses the number of watches required to activate/maintain vs. reactivate and confuses Respirator Glasses Verification with a change in license medical status would select this answer.
D. This answer is incorrect because the number of shifts required is incorrect. This choice is plausible because 5 shifts is required for initial license activation. The candidate who confuses the number of watches required to activate/maintain vs. reactivate and correctly identifies the procedure requiring Special Prescription Respirator Glasses Verification would select this answer.
D. This answer is incorrect because the number of shifts required is incorrect. This choice is plausible because 5 shifts is required for initial license activation. The candidate who confuses the number of watches required to activate/maintain vs. reactivate and correctly identifies the procedure requiring Special Prescription Respirator Glasses Verification would select this answer.
Technical Reference(s):
Technical Reference(s):
Procedure 2.0.7 (Licensed Operator Active/Reactivation/Medical Status Maintenance Program) Rev. 09 Procedure NTP8.1 (Administration of Licensed Operator Medical Examination Program), Rev. 17 Proposed references to be provided to applicants during examination:
Procedure 2.0.7 (Licensed Operator Active/Reactivation/Medical Status Maintenance Program) Rev. 09 Procedure NTP8.1 (Administration of Licensed Operator Medical Examination Program), Rev. 17 Proposed references to be provided to applicants during examination:             None Learning Objective:
None Learning Objective:
INT00705130010100 Given a set of plant conditions, recognize non-compliance with a Chapter 5.0 Requirement.
INT00705130010100 Given a set of plant conditions, recognize non
-compliance with a Chapter 5.0 Requirement.
Question Source:
Question Source:
Bank #     Modified Bank #
Bank #
New X   Question Cognitive Level:
Modified Bank #
Memory or Fundamental Knowledge X   Comprehension or Analysis   10 CFR Part 55 Content:
New X Question Cognitive Level:       Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content:         55.41.10 55.43.2 Difficulty: 4 SRO Only - 10CFR55.43 b (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
55.41.10 55.43.2   Difficulty: 4 SRO Only - 10CFR55.43 b (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Requires knowledge of NRC license maintenance requirements and selection of procedure requiring Special Prescription Respirator Glasses Verification.
Requires knowledge of NRC license maintenance requirements and selection of procedure requiring Special Prescription Respirator Glasses Verification.
46 47 48 Examination Outline Cross
46
-
 
47 48 Examination Outline Cross-


==Reference:==
==Reference:==


Level S RO Tier #   3 Group # 1 K/A # 2.1.41 Importance Rating 3.7     2.1.41 Knowledge of the refueling process.
Level                     SRO Tier #                     3 Group #                   1 K/A #                     2.1.41 Importance Rating         3.7 2.1.41     Knowledge of the refueling process.
Question: 95 Which activity REQUIRES Refuel Floor Supervisor permission during refueling operations in Mode 5? A. Allowing under vessel access.
Question: 95 Which activity REQUIRES Refuel Floor Supervisor permission during refueling operations in Mode 5?
A. Allowing under vessel access.
B. Allowing access to the refuel floor.
B. Allowing access to the refuel floor.
C. Re-commencing fuel handling operations.
C. Re-commencing fuel handling operations.
D. Using greater than 50 gallons of demineralized water on the refuel floor.
D. Using greater than 50 gallons of demineralized water on the refuel floor.
Answer: C. Re-commencing fuel handling operations.
Answer:
C. Re-commencing fuel handling operations.
Explanation:
Explanation:
Requires knowledge of Refuel Floor SRO responsibilities during refueling operations. Refuel Floor Supervisor permission is required to re
Requires knowledge of Refuel Floor SRO responsibilities during refueling operations. Refuel Floor Supervisor permission is required to re-commence fuel handling operations IAW Attachment 4 (Reset Checklist) which shall be used each time the normal fuel handling process is stopped/interrupted. This includes, but is not limited to, Shift Turnover, Fuel Mover/Spotter mid-shift role change, or following a distraction which interrupts the normal fuel handling process flow. Putting the applicable procedure in the stem would eliminate under vessel access plausibility due to title being Fuel Handling - Refueling Platform.
-commence fuel handling operations IAW Attachment 4 (Reset Checklist) which shall be used each time the normal fuel handling process is stopped/interrupted. This includes, but is not limited to, Shift Turnover, Fuel Mover/Spotter mid-shift role change, or following a distraction which interrupts the normal fuel handling process flow.
Distracters:
Putting the applicable procedure in the stem would eliminate under vessel access plausibility due to title being "Fuel Handling  
- Refueling Platform
". Distracters:
A. This answer is incorrect because Refuel Floor Supervisor permission is not required to allow access to the under vessel area. This answer is plausible because under vessel area gets posted to prohibit access without Shift Manager's permission. The candidate who confuses access permission authority would select this choice.
A. This answer is incorrect because Refuel Floor Supervisor permission is not required to allow access to the under vessel area. This answer is plausible because under vessel area gets posted to prohibit access without Shift Manager's permission. The candidate who confuses access permission authority would select this choice.
B. This answer is incorrect because Refuel Floor Supervisor permission is not required to allow access to the refuel floor. This choice is plausible due to the Refuel floor SRO permission is required to access the fuel handling area  
B. This answer is incorrect because Refuel Floor Supervisor permission is not required to allow access to the refuel floor. This choice is plausible due to the Refuel floor SRO permission is required to access the fuel handling area - the fuel handling area is located within the refueling floor. The candidate who confuses refuel floor with fuel handling area would choose this answer.
- the fuel handling area is located within the refueling floor. The candidate who confuses refuel floor with fuel handling area would choose this answer.
D. This answer is incorrect because Refuel Floor Supervisor permission is not required to use greater than 50 gallons of demineralized water on the refuel floor. This choice is plausible due to the Refuel floor SRO is required to brief available refueling floor personnel on limiting demineralized water usage and requirement to notify Control Room if using > 50 gallons demineralized water each shift. The candidate who confuses briefing vs. giving permission would choose this answer.
D. This answer is incorrect because Refuel Floor Supervisor permission is not required to use greater than 50 gallons of demineralized water on the refuel floor. This choice is plausible due to the Refuel floor SRO is required to brief available refueling floor personnel on limiting demineralized water usage and requirement to notify Control Room if using > 50 gallons demineralized water each shift. The candidate who confuses briefing vs. giving permission would choose this answer.
Technical Reference(s):
Technical Reference(s):
Procedure 2.1.20.3, RPV Refueling Preparation (Wet Lift of Dryer and Separator), Rev. 45 49 Procedure 10.25 (Refueling  
Procedure 2.1.20.3, RPV Refueling Preparation (Wet Lift of Dryer and Separator), Rev. 45 49
- Core Unload, Reload, and Shuffle), Rev. 59 Procedure 2.2.31 (Fuel Handling  
 
- Refueling Platform), Rev. 47 Proposed references to be provided to applicants during examination:
Procedure 10.25 (Refueling - Core Unload, Reload, and Shuffle), Rev. 59 Procedure 2.2.31 (Fuel Handling - Refueling Platform), Rev. 47 Proposed references to be provided to applicants during examination:       None Learning Objective:
None Learning Objective:
INT0231002001160A Identify the administrative duties and responsibilities of the each of the following: Refueling Floor Supervisor Question Source:
INT0231002001160A Identify the administrative duties and responsibilities of the each of the following: Refueling Floor Supervisor Question Source:
Bank #     Modified Bank #
Bank #
New X   Question Cognitive Level:
Modified Bank #
Memory or Fundamental Knowledge X   Comprehension or Analysis   10 CFR Part 55 Content:
New X Question Cognitive Level:     Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content:       55.41.2 55.41.10 55.43.6 55.45.13 Difficulty: 3 10 CFR 55.43(b)(7) - Fuel handling facilities and procedures.
55.41.2 55.41.10 55.43.6 55.45.13   Difficulty: 3 10 CFR 55.43(b)(7)  
- Fuel handling facilities and procedures.
Requires knowledge of Refuel floor SRO responsibilities.
Requires knowledge of Refuel floor SRO responsibilities.
50 51 52 53 54 Examination Outline Cross
50
-
 
51 52 53 54 Examination Outline Cross-


==Reference:==
==Reference:==


Level SRO Tier #   3 Group # 2 K/A # 2.2.11 Importance Rating 3.3     2.2.11 Knowledge of the process for controlling temporary design changes.
Level                   SRO Tier #                   3 Group #                 2 K/A #                   2.2.11 Importance Rating       3.3 2.2.11     Knowledge of the process for controlling temporary design changes.
Question: 96 The plant is operating at power in Mode 1.
Question: 96 The plant is operating at power in Mode 1.
What is the MAXIMUM time a Temporary Alteration In Support of Maintenance (TASM) can be installed on plant equipment WITHOUT performing a 10CFR50.59 Review IAW Procedure 3.4.4 (Temporary Configuration Change)?
What is the MAXIMUM time a Temporary Alteration In Support of Maintenance (TASM) can be installed on plant equipment WITHOUT performing a 10CFR50.59 Review IAW Procedure 3.4.4 (Temporary Configuration Change)?
A. 30 days B. 60 days   C. 90 days D. 120 days ANSWER: C. 90 days Explanation:
A. 30 days B. 60 days C. 90 days D. 120 days ANSWER:
A temporary alteration is necessary to support maintenance if it makes the maintenance activity easier, or the maintenance activity has been planned to allow prompt restoration. TASMs have regulatory considerations specific to duration under 10CFR50.59. Engineering Procedure 3.4.4, Temporary Configuration Change, Attachment 7, Step 4.1.1 and 4.1.2 describe the requirements for a 10CFR50.59 review prior to installation if it is expected to be in place > 90 days, or if after installation, it is going to be installed > 90 days. The procedure states that if a TASM that was installed and originally not expected to exceed 90 day that a 50.59 review should be performed. Procedure 0-EN-HU-106, Procedure and Work Instruction Use and Adherence, Step 3.12 defines should and states: Should - Denotes strong recommendation and indicates an action that is expected to be performed as described unless there is a compelling reason not to do so. Distracters:
C. 90 days Explanation:
A. This answer is incorrect because the time listed is not the maximum time a TASM can be installed without a 10CFR50.59 Review being performed. The maximum time is 90 days as allowed by federal regulations. This answer is plausible because some non
A temporary alteration is necessary to support maintenance if it makes the maintenance activity easier, or the maintenance activity has been planned to allow prompt restoration. TASMs have regulatory considerations specific to duration under 10CFR50.59. Engineering Procedure 3.4.4, Temporary Configuration Change, Attachment 7, Step 4.1.1 and 4.1.2 describe the requirements for a 10CFR50.59 review prior to installation if it is expected to be in place > 90 days, or if after installation, it is going to be installed > 90 days. The procedure states that if a TASM that was installed and originally not expected to exceed 90 day that a 50.59 review should be performed. Procedure 0-EN-HU-106, Procedure and Work Instruction Use and Adherence, Step 3.12 defines should and states: Should - Denotes strong recommendation and indicates an action that is expected to be performed as described unless there is a compelling reason not to do so.
-emergency events require NRC notification reports and the candidate may recall the 30 days without tying it to the TASM requirements.
Distracters:
A. This answer is incorrect because the time listed is not the maximum time a TASM can be installed without a 10CFR50.59 Review being performed. The maximum time is 90 days as allowed by federal regulations. This answer is plausible because some non-emergency events require NRC notification reports and the candidate may recall the 30 days without tying it to the TASM requirements.
B. This answer is incorrect because the time listed is not the maximum time a TASM can be installed without a 10CFR50.59 Review being performed. The maximum time is 90 days as allowed by federal regulations. This answer is plausible because configuration change affected documents must be processed within 60 days of CED installation and the candidate may remember that time frame.
55


B. This answer is incorrect because the time listed is not the maximum time a TASM can be installed without a 10CFR50.59 Review being performed. The maximum time is 90 days as allowed by federal regulations. This answer is plausible because configuration change affected documents must be processed within 60 days of CED installation and the candidate may remember that time frame.
D. This answer is incorrect because the time listed exceeds the maximum time a TASM can be installed without a 10CFR50.59 Review being performed. The maximum time is 90 days as allowed by federal regulations. This answer is plausible because there is a time limit of 120 days in Tech Specs for time shutdown which requires scram timing on startup prior to 40%
55 D. This answer is incorrect because the time listed exceeds the maximum time a TASM can be installed without a 10CFR50.59 Review being performed. The maximum time is 90 days as allowed by federal regulations. This answer is plausible because there is a time limit of 120 days in Tech Specs for time shutdown which requires scram timing on startup prior to 40% power. The candidate may recall the 120 days but be unsure of its source.
power. The candidate may recall the 120 days but be unsure of its source.
Technical Reference(s):
Technical Reference(s):
Procedure 3.4.4 (Temporary Configuration Change), Rev. 15.
Procedure 3.4.4 (Temporary Configuration Change), Rev. 15.
Proposed references to be provided to applicants during examination:
Proposed references to be provided to applicants during examination:       NONE Learning Objective:
NONE Learning Objective:
INT0320109A0A010A Procedure 3.4.4, Temporary Configuration Change - Discuss the following as described in Engineering Procedure 3.4, Configuration Change Control: Temporary Configuration Changes (TCCs)
INT0320109A0A010A Procedure 3.4.4, Temporary Configuration Change  
- Discuss the following as described in Engineering Procedure 3.4, Configuration Change Control: Temporary Configuration Changes (TCCs)
Question Source:
Question Source:
Bank #     Modified Bank #
Bank #
New X   Question Cognitive Level:
Modified Bank #
Memory or Fundamental Knowledge X   Comprehension or Analysis     10 CFR Part 55 Content:
New X Question Cognitive Level:     Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content:       55.41.10 55.43.3 55.45.13 Difficulty: 3 SRO Only - 10CFR55.43 b (3) Facility licensee procedure required to obtain authority for design and operating changes in the facility.
55.41.10 55.43.3 55.45.13   Difficulty: 3 SRO Only - 10CFR55.43 b (3) Facility licensee procedure required to obtain authority for design and operating changes in the facility.
Requires knowledge of Administrative processes for temporary modifications/configuration changes.
Requires knowledge of Administrative processes for temporary modifications/configuration changes. SRO Task: 200001W0303 Approve Installation of a Plant Temporary Configuration Change (TCC) Order. The SRO is responsible for knowledge of the TCC process.
SRO Task: 200001W0303 Approve Installation of a Plant Temporary Configuration Change (TCC) Order. The SRO is responsible for knowledge of the TCC process.
56 57 Examination Outline Cross
56
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57 Examination Outline Cross-


==Reference:==
==Reference:==


Level SRO Tier #   3 Group # 2 K/A # 2.2.19 Importance Rating 3.4     2.2.19 Knowledge of maintenance work order requirements.
Level                       SRO Tier #                     3 Group #                     2 K/A #                     2.2.19 Importance Rating         3.4 2.2.19     Knowledge of maintenance work order requirements.
Question: 97 Which of the following identifies an item required to be addressed by a SRO performing a Work Order Standard Plant Impact Statement IAW 0-CNS-WM-102 (Work Implementation and Closeout)?
Question: 97 Which of the following identifies an item required to be addressed by a SRO performing a Work Order Standard Plant Impact Statement IAW 0-CNS-WM-102 (Work Implementation and Closeout)?
A. Verification of required parts availability
A. Verification of required parts availability.
. B. Verifying the Work Instructions are accurate.
B. Verifying the Work Instructions are accurate.
 
C. Identification of single valve isolation requirements.
C. Identification of single valve isolation requirements.
D. Identifying required walkdowns by Shop/Responsible Work Center.
D. Identifying required walkdowns by Shop/Responsible Work Center.
Answer: C. Identification of single valve isolation requirements.
Answer:
Explanation: Requires knowledge of SRO/FIN SRO/WCCA/WCC Supervisor impact review of maintenance work orders. The Standard Plant Impact Statement addresses the following items:
C. Identification of single valve isolation requirements.
Explanation: Requires knowledge of SRO/FIN SRO/WCCA/WCC Supervisor impact review of maintenance work orders. The Standard Plant Impact Statement addresses the following items:
: 1. Does the work activity affect SSCs identified in TSs?
: 1. Does the work activity affect SSCs identified in TSs?
: 2. Will the maintenance activity require the equipment to be declared inoperable/unavailable?
: 2. Will the maintenance activity require the equipment to be declared inoperable/unavailable?
: 3. Is a power reduction required?
: 3. Is a power reduction required?
: 4. Any other special plant conditions required to perform this work activity?
: 4. Any other special plant conditions required to perform this work activity?
: 5. Is there a potential to affect the Operability of systems, structures, or components required to be operable?
: 5. Is there a potential to affect the Operability of systems, structures, or components required to be operable?
: 6. Does the maintenance activity create the potential for inadvertent actuations (RPS trip, turbine trip, ECCS actuation, SDC isolation, system discharge valve closure)?
: 6. Does the maintenance activity create the potential for inadvertent actuations (RPS trip, turbine trip, ECCS actuation, SDC isolation, system discharge valve closure)?
: 7. Does the maintenance activity have an actual or potential reactivity impact? (CRD, recirc flow, feed flow, feed temperature, etc.)?
: 7. Does the maintenance activity have an actual or potential reactivity impact? (CRD, recirc flow, feed flow, feed temperature, etc.)?
: 8. Does the work activity increase the potential for loss of off
: 8. Does the work activity increase the potential for loss of off-site power?
-site power?
: 9. Will single valve isolation be required?
: 9. Will single valve isolation be required?
: 10. Are contingency plans or compensatory actions necessary?
: 10. Are contingency plans or compensatory actions necessary?
Distracters:
Distracters:
A. This answer is incorrect due to verification of required parts availability not being part of the impact review by operations. This answer is plausible due to SROs normally performing this item. Parts availability is the most common reason for removing work from the schedule and is highly scrutinized by Operations to maintain schedule stability and minimize unnecessary burden on the operators and site workers. The candidate who recognizes the impact of parts availability on planning & scheduling during the operations impact review would select this answer.
A. This answer is incorrect due to verification of required parts availability not being part of the impact review by operations. This answer is plausible due to SROs normally performing this item. Parts availability is the most common reason for removing work from the schedule and is highly scrutinized by Operations to maintain schedule stability and minimize unnecessary burden on the operators and site workers. The candidate who recognizes the impact of parts availability on planning & scheduling during the operations impact review would select this answer.
B. This answer is incorrect due to verifying the accuracy of the Work Instructions not being part of the impact review by operations. This answer is plausible due to SROs often reviewing 58 the instructions for impact on Operations but is not required to be verified by the operations organization. The candidate who incorrectly believes reviewing Work Instructions is verifying their accuracy as part of the operations impact review would select this answer.
B. This answer is incorrect due to verifying the accuracy of the Work Instructions not being part of the impact review by operations. This answer is plausible due to SROs often reviewing 58
 
the instructions for impact on Operations but is not required to be verified by the operations organization. The candidate who incorrectly believes reviewing Work Instructions is verifying their accuracy as part of the operations impact review would select this answer.
D. This answer is incorrect due to determining if a Shop/Responsible Work Center walkdown is required not being part of the impact review by operations. This answer is plausible due operators being required to walkdown clearance orders in support of work orders but is not required to be determined by the operations organization. The candidate who incorrectly believes determining clearance order walkdowns are part of the operations impact review would select this answer.
D. This answer is incorrect due to determining if a Shop/Responsible Work Center walkdown is required not being part of the impact review by operations. This answer is plausible due operators being required to walkdown clearance orders in support of work orders but is not required to be determined by the operations organization. The candidate who incorrectly believes determining clearance order walkdowns are part of the operations impact review would select this answer.
Technical Reference(s):
Technical Reference(s):
Procedure CNS-WM-102, Work Implementation and Closeout, Rev.01. Proposed references to be provided to applicants during examination:
Procedure CNS-WM-102, Work Implementation and Closeout, Rev.01.
None Learning Objective:   SKL0110101001290A, 0.40, Work Control Program, Discuss the following as described in Administrative Procedure 0.40, Work Control Program: 1) Precautions and limitations 2) Minor maintenance 3) Emergency MWR processing.
Proposed references to be provided to applicants during examination:           None Learning Objective: SKL0110101001290A, 0.40, Work Control Program, Discuss the following as described in Administrative Procedure 0.40, Work Control Program: 1)
Precautions and limitations 2) Minor maintenance 3) Emergency MWR processing.
Question Source:
Question Source:
Bank #     Modified Bank #
Bank #
New X   Question Cognitive Level:
Modified Bank #
Memory or Fundamental Knowledge X   Comprehension or Analysis     10 CFR Part 55 Content:
New X Question Cognitive Level:               Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content:         55.41.10 55.43.5 55.45.13 Difficulty: 3 SRO Only - 10CFR55.43 b (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
55.41.10 55.43.5 55.45.13   Difficulty: 3 SRO Only - 10CFR55.43 b (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Knowledge of administrative procedures that specify implementation of plant normal procedures.
Knowledge of administrative procedures that specify implementation of plant normal procedures.
59 60 61 62 63 Examination Outline Cross
59
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60 61 62 63 Examination Outline Cross-


==Reference:==
==Reference:==


Level SRO Tier #   3 Group # 3 K/A # 2.3.6 Importance Rating 3.8     2.3.6 Ability to approve release permits.
Level                     SRO Tier #                   3 Group #                   3 K/A #                   2.3.6 Importance Rating       3.8 2.3.6     Ability to approve release permits.
Question: 98 Waste Sample Tank A is required to be discharged.
Question: 98 Waste Sample Tank A is required to be discharged.
What is the MINIMUM recirculation time which allows approval of the release by the Shift Manager IAW Procedure 8.8.11 (Liquid Radioactive Waste Discharge Authorization)?
What is the MINIMUM recirculation time which allows approval of the release by the Shift Manager IAW Procedure 8.8.11 (Liquid Radioactive Waste Discharge Authorization)?
A. 1/2 hour prior to being sampled.
A. 1/2 hour prior to being sampled.
B. 2 hour prior to being sampled.
B. 2 hour prior to being sampled.
 
C. 3 hours prior to being sampled.
C. 3 hours prior to being sample
D. 4 hours prior to being sampled.
: d. D. 4 hours prior to being sampled.
Answer:
 
B. 2 hour prior to being sampled.
Answer: B. 2 hour prior to being sampled.
 
Explanation:
Explanation:
The following items are verified by the Shift Manager prior to approving a liquid RW discharge:
The following items are verified by the Shift Manager prior to approving a liquid RW discharge:
Tank recirculation time > 2 hours prior to being sampled Proper dilution flow (159,000 gpm)
* Tank recirculation time > 2 hours prior to being sampled
Tank volume sampled matches volume to be released Distracters:
* Proper dilution flow (159,000 gpm)
* Tank volume sampled matches volume to be released Distracters:
A. This answer is incorrect due to tank recirculation not meeting the 2 hour procedural requirement. This answer is plausible due to 30 minutes is the normal tank recirculation time prior to sampling to support transfer vs. discharge. The candidate who confuses tank sample recirc time for transfer would select this answer.
A. This answer is incorrect due to tank recirculation not meeting the 2 hour procedural requirement. This answer is plausible due to 30 minutes is the normal tank recirculation time prior to sampling to support transfer vs. discharge. The candidate who confuses tank sample recirc time for transfer would select this answer.
C. This answer is incorrect due to tank recirculation being greater than the 2 hour procedural minimum requirement. This answer is plausible due to 3 hrs satisfying the required recirulation time but is not the minimum. The candidate who does not recognize the minimum tank sample recirc time for discharge would select this answer.
C. This answer is incorrect due to tank recirculation being greater than the 2 hour procedural minimum requirement. This answer is plausible due to 3 hrs satisfying the required recirulation time but is not the minimum. The candidate who does not recognize the minimum tank sample recirc time for discharge would select this answer.
D. This answer is incorrect due to tank recirculation being greater than the 2 hour procedural minimum requirement. This answer is plausible due to 4 hours satisfying the required recirulation time but is not the minimum and can be confused with Phase Separator decanting time limitations based upon turbidity results being > 10 FTU requiring stopping 64 decanting for at least 4 hours before initiating another decant. The candidate who does not recognize the minimum tank sample recirc time for discharge would select this answer.
D. This answer is incorrect due to tank recirculation being greater than the 2 hour procedural minimum requirement. This answer is plausible due to 4 hours satisfying the required recirulation time but is not the minimum and can be confused with Phase Separator decanting time limitations based upon turbidity results being > 10 FTU requiring stopping 64
 
decanting for at least 4 hours before initiating another decant. The candidate who does not recognize the minimum tank sample recirc time for discharge would select this answer.
Technical Reference(s):
Technical Reference(s):
Procedure 8.8.11 Liquid Radioactive Waste Discharge Authorization, Rev. 32.
Procedure 8.8.11 Liquid Radioactive Waste Discharge Authorization, Rev. 32.
Procedure 2.5.1.6 (RW Low Conductivity Liquid Waste Sample Tank Fluid Transfer), Rev. 43 Procedure 2.5.1.7 (RWCU Phase Separator Tank Transfer), Rev. 27 Proposed references to be provided to applicants during examination:
Procedure 2.5.1.6 (RW Low Conductivity Liquid Waste Sample Tank Fluid Transfer), Rev. 43 Procedure 2.5.1.7 (RWCU Phase Separator Tank Transfer), Rev. 27 Proposed references to be provided to applicants during examination:         NONE Learning Objective:
NONE Learning Objective:
INT0320115B0B0200 State who, by title, is required to grant permission to perform a liquid discharge with the liquid radwaste monitor inoperable.
INT0320115B0B0200 State who, by title, is required to grant permission to perform a liquid discharge with the liquid radwaste monitor inoperable.
Question Source:
Question Source:
Bank #     Modified Bank #
Bank #
New   X   Question Cognitive Level:
Modified Bank #
Memory or Fundamental Knowledge x   Comprehension or Analysis     10 CFR Part 55 Content:
New X Question Cognitive Level:     Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content:         55.41.13 55.43.4 55.45.10 Difficulty: 3 SRO Only - 10CFR55.43 b (4) Radiation hazard that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.
55.41.13 55.43.4 55.45.10   Difficulty: 3 SRO Only - 10CFR55.43 b (4) Radiation hazard that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.
Requires knowledge of the process for liquid release approval.
Requires knowledge of the process for liquid release approval.
65 66 67 Examination Outline Cross
65
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66 67 Examination Outline Cross-


==Reference:==
==Reference:==


Level SRO Tier #   3 Group # 4 K/A # 2.4.29 Importance Rating 4.4     2.4.29 Knowledge of the Emergency Plan.
Level                     SRO Tier #                     3 Group #                   4 K/A #                     2.4.29 Importance Rating         4.4 2.4.29   Knowledge of the Emergency Plan.
Question: 99 Given the following:
Question: 99 Given the following:
At 1200 the threshold for a NOUE is exceeded.
* At 1200 the threshold for a NOUE is exceeded.
At 1210 the Emergency Director declared classification of a NOUE.
* At 1210 the Emergency Director declared classification of a NOUE.
At 1215 the threshold for an ALERT is exceeded.
* At 1215 the threshold for an ALERT is exceeded.
At 1220 the Emergency Director declared classification of an ALERT. What is the LATEST time that State/Local agency notifications of the NOUE classification is required to be performed IAW EPIP 5.7.6 (Notification)?
* At 1220 the Emergency Director declared classification of an ALERT.
A. 1215   B. 1225   C. 1230 D. 1235 Answer: B. 1225 Explanation:
What is the LATEST time that State/Local agency notifications of the NOUE classification is required to be performed IAW EPIP 5.7.6 (Notification)?
Requires knowledge of E
A.     1215 B.     1225 C.     1230 D.     1235 Answer:
-Plan notification requirements. Initial notification to State/local agencies of E
B.     1225 Explanation:
-plan classification is required to be performed within 15 minutes of emergency declaration. There is a common misapplication of start times when staggering (subsequent) EALs are met. The SRO must keep separate times running for each EAL as it is entered and not reset an earlier time due to subsequent EALs. This is the responsibility of the Emergency Director (SRO) until relieved by another Emergency Director.
Requires knowledge of E-Plan notification requirements. Initial notification to State/local agencies of E-plan classification is required to be performed within 15 minutes of emergency declaration. There is a common misapplication of start times when staggering (subsequent)
 
EALs are met. The SRO must keep separate times running for each EAL as it is entered and not reset an earlier time due to subsequent EALs. This is the responsibility of the Emergency Director (SRO) until relieved by another Emergency Director.
Distracters:
Distracters:
A. This answer is incorrect due to not being the latest time requiring notification. This answer is plausible due to being 15 minutes from exceeding an EAL threshold. The candidate who cannot differentiate between the time to notify State/Local from exceeding EAL threshold vs. Declaration would choose this answer.
A. This answer is incorrect due to not being the latest time requiring notification. This answer is plausible due to being 15 minutes from exceeding an EAL threshold. The candidate who cannot differentiate between the time to notify State/Local from exceeding EAL threshold vs. Declaration would choose this answer.
C. This answer is incorrect due exceeding the latest time requiring notification. This answer is plausible due to being 15 hour from exceeding an Alert EAL threshold. The candidate who cannot differentiate between the time to notify State/Local from Alert EAL threshold vs. NOUE Declaration would choose this answer.
C. This answer is incorrect due exceeding the latest time requiring notification. This answer is plausible due to being 15 hour from exceeding an Alert EAL threshold. The candidate who cannot differentiate between the time to notify State/Local from Alert EAL threshold vs.
D. This answer is incorrect due exceeding the latest time requiring notification. This answer is plausible due to being 15 minutes from the ALERT declaration. The candidate who cannot 68 differentiate between the time to notify State/Local from Alert EAL Declaration vs. NOUE Declaration would choose this answer.
NOUE Declaration would choose this answer.
D. This answer is incorrect due exceeding the latest time requiring notification. This answer is plausible due to being 15 minutes from the ALERT declaration. The candidate who cannot 68
 
differentiate between the time to notify State/Local from Alert EAL Declaration vs. NOUE Declaration would choose this answer.
Technical Reference(s):
Technical Reference(s):
EPIP 5.7.2 (Emergency Director EPIP), Rev. 32 Emergency Procedure 5.7.6 (Notification), Rev 24.
EPIP 5.7.2 (Emergency Director EPIP), Rev. 32 Emergency Procedure 5.7.6 (Notification), Rev 24.
Proposed references to be provided to applicants during examination:
Proposed references to be provided to applicants during examination:               NONE Learning Objective:
NONE Learning Objective:
GEN0030401B0B030B Emergency Notifications and Communications Systems: State the time requirements for initial and/or follow-up notifications to offsite agencies.
GEN0030401B0B030B Emergency Notifications and Communications Systems: State the time requirements for initial and/or follow
Question Source:
-up notifications to offsite agencies. Question Source:
Bank #
Bank #     Modified Bank # 6082 New     Question Cognitive Level:
Modified Bank # 6082 New Question Cognitive Level:     Memory or Fundamental Knowledge                 X Comprehension or Analysis 10 CFR Part 55 Content:         55.41.10 55.43.5 55.45.13 Difficulty: 2 SRO Only - 10CFR55.43 b (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Memory or Fundamental Knowledge X Comprehension or Analysis     10 CFR Part 55 Content:
The SRO is responsible for task 344022O0303, Direct Emergency Response as Emergency Director (Emergency Plan) 69
55.41.10 55.43.5 55.45.13   Difficulty: 2 SRO Only - 10CFR55.43 b (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
 
The SRO is responsible for task 344022O0303, Direct Emergency Response as Emergency Director (Emergency Plan) 69 70 71 72  
70 71 72 Question 100 and associated references redacted due to SUNSI considerations
 
SRO References INFORMATION ONLY INFORMATION ONLY
 
INFORMATION ONLY INFORMATION ONLY
 
INFORMATION ONLY INFORMATION ONLY
 
INFORMATION ONLY INFORMATION ONLY
 
INFORMATION ONLY INFORMATION ONLY


Question 100 and associated references redacted due to SUNSI considerations
AG1.1          1      2    3        4      5      DEF              AS1.1      1      2        3      4        5      DEF          AA1.1      1      2      3      4      5    DEF                AU1.1        1      2      3      4        5    DEF Any valid gaseous monitor reading > Table A-1 column GE            Any valid gaseous monitor reading > Table A-1 column SAE        Any valid gaseous monitor reading > Table A-1 column              Any valid gaseous monitor reading > Table A-1 column UE for 15 min. (Note 1)                                                for  15 min. (Note 1)                                            Alert for  15 min. (Note 2)                                    for  60 min. (Note 2)
AG1.2          1      2    3        4      5      DEF              AS1.2      1      2        3      4        5      DEF          AA1.2      1      2      3      4      5    DEF                AU1.2        1      2      3      4        5    DEF Dose assessment using actual meteorology indicates doses              Dose assessment using actual meteorology indicates doses          Any valid liquid effluent monitor reading > Table A-1 column      Any valid liquid effluent monitor reading > Table A-1 column 1          > 1 Rem TEDE or > 5 Rem thyroid CDE at or beyond the site boundary
                                                                                          > 0.1 Rem TEDE or > 0.5 Rem thyroid CDE at or beyond the site boundary Alert for  15 min. (Note 2)                                    UE for  60 min. (Note 2)
Offsite Rad Conditions AG1.3          1      2    3        4      5      DEF              AS1.3      1      2        3      4        5      DEF          AA1.3      1      2      3      4      5    DEF                AU1.3        1      2      3      4        5    DEF Field survey results indicate closed window dose rates                Field survey indicates closed window dose rate > 0.1 Rem/hr        Confirmed sample analyses for gaseous or liquid releases          Confirmed sample analyses for gaseous or liquid releases
                    > 1 Rem/hr expected to continue for  60 min. at or beyond            that is expected to continue for  60 min. at or beyond the site  indicate concentrations or release rates > 200 x ODAM limits      indicate concentrations or release rates > 2 x ODAM the site boundary (Note 1)                                            boundary (Note 1)                                                  for  15 min. (Note 2)                                            limits for  60 min. (Note 2)
OR                                                                    OR Analyses of field survey samples indicate thyroid CDE                Field survey sample analysis indicates thyroid CDE > 0.5 Rem
                    > 5 Rem for 1 hr of inhalation at or beyond the site boundary        for 1 hr of inhalation at or beyond the site boundary
: m.                                                                                                                                                          AA2.1      1      2      3      4      5    DEF                AU2.1        1      2        3      4      5      DEF Table A-1 Effluent Monitor Classification Thresholds e                                                                                                                                                          Damage to irradiated fuel OR loss of water level (uncovering      Unplanned water level drop in the reactor cavity or spent fuel GE                SAE                  ALERT                    UE            irradiated fuel outside the RPV) that causes EITHER of the        pool as indicated by any of the following:
Monitor nt                                                                    for  15 min.      for  15 min.          for  15 min.          for  60 min.      following:
* LI-86 (calibrated to 1001' elev.)
2 Valid RMA-RA-1 Fuel Pool Area Rad reading > 50 R/hr
* Spent fuel pool low level alarm OR
* Visual observation ERP                  3.50E+08 &#xb5;Ci/sec  3.50E+07 &#xb5;Ci/sec    2.80E+06 &#xb5;Ci/sec          2.24E+05 &#xb5;Ci/sec AND Valid RMP-RM-452 A-D Rx Bldg Vent Exhaust Plenum                Valid area radiation monitor reading rise on RMA-RA-1 or Onsite Rad                                                                                                                                                Hi-Hi alarm Conditions                                                                                                                                                                                                                RMA-RA-2 Rx Bldg Vent              3.50E+07 &#xb5;Ci/sec  3.50E+06 &#xb5;Ci/sec    5.45E+05 &#xb5;Ci/sec          8.48E+04 &#xb5;Ci/sec GASEOUS Spent Fuel Pool                            Turb Bldg Vent                                                                                                      AA2.2      1      2      3      4      5    DEF                AU2.2        1      2      3      4        5    DEF 3.50E+07 &#xb5;Ci/sec  3.50E+06 &#xb5;Ci/sec    5.62E+05 &#xb5;Ci/sec          9.02E+04 &#xb5;Ci/sec Events A water level drop in the reactor refueling cavity or spent fuel  Unplanned valid area radiation monitor reading or survey RW / ARW Bldg Vent            3.50E+07 &#xb5;Ci/sec  3.50E+06 &#xb5;Ci/sec    5.64E+05 &#xb5;Ci/sec          9.08E+04 &#xb5;Ci/sec      pool that will result in irradiated fuel becoming uncovered        results rise by a factor of 1,000 over normal levels*
* Normal levels can be considered as the highest reading in the past 24 The lesser of *:          The lesser of *:                                                                          hours excluding the current peak value 200 x calculated          2 x calculated LIQUID Rad Waste Effluent                  -----              -----          alarm values              alarm values        AA3.1      1      2      3      4      5    DEF 3                                                                                                        OR monitor upscale OR monitor upscale      Dose rates > 15 mRem/hr in EITHER of the following areas requiring continuous occupancy to maintain plant safety MCR/CAS                      Service Water Effluent                  -----              -----          4.80E-04 &#xb5;Ci/cc          4.80E-06 &#xb5;Ci/cc      functions:
Rad                                                                                                                                                    Main Control Room (RM-RA-20)
OR
* with effluent discharge not isolated        CAS


PreparedforNPPDby:OperationsSupportServices,Inc.-www.ossi-net.com(Rev.84/2/10)Modes: 1PowerOperationDefueledDEF 2Startup 5Refueling 3HotShutdown 4ColdShutdownCooperNuclearStationEmergencyActionLevelMatrixEPIP5.7.1Attachment4,Rev.11Sequentialnumberwithinsubcategory/classificationSubcategorynumber(1ifnosubcategory)XXX.XCategory(A,H,S,F,C,E)Emergencyclassification(G,S,A,U)EALIdentifierMODE4,5orDEF GENERALEMERGENCY SITEAREAEMERGENCY ALERT UNUSUALEVENTOnsiteRadConditions
N for possible escalation above the Unusual Event due inoperable, or out of service, before the event d as it will have no adverse impact on the ability of the ond that already allowed by Technical Specifications at likely longer than the specified time interval. If off-ed the backfeed, its power to the safety-related buses}}
&SpentFuelPoolEventsAbnorm.RadRelease/RadEffluent A 2OffsiteRadConditions 1Hazards&OtherCondi-tionsAffect-ingPlantSafety HNaturalorDestructivePhenomenaFireorExplosionControlRoomEvacuationHazardousGasSecurityJudgment 1 2 3 4 5 6ISFSI ETableC-1RPVLeakageIndications*Drywellequipmentdrainsumplevelrise*Drywellfloordrainsumplevelrise*ReactorBuildingequipmentdrainsumplevelrise*ReactorBuildingfloordrainsumplevelrise*SuppressionPoolwaterlevelrise*RPVmake-upraterise*ObservationofunisolableRCSleakageMCR/CASRad 3AU1.1AU1.2AA1.1AS1.1AS1.2Doseassessmentusingactualmeteorologyindicatesdoses>0.1RemTEDEor>0.5RemthyroidCDEatorbeyondthesiteboundaryAA1.2AA1.3Confirmedsampleanalysesforgaseousorliquidreleasesindicateconcentrationsorreleaserates>200xODAMlimitsfor&#xb3;15min.(Note2)AU2.1Unplannedwaterleveldropinthereactorcavityorspentfuelpoolasindicatedbyanyofthefollowing:*LI-86(calibratedto1001'elev.)*Spentfuelpoollowlevelalarm*VisualobservationANDValidarearadiationmonitorreadingriseonRMA-RA-1orRMA-RA-2AA2.1Doseassessmentusingactualmeteorologyindicatesdoses>1RemTEDEor>5RemthyroidCDEatorbeyondthesiteboundaryAG1.2AU1.3Confirmedsampleanalysesforgaseousorliquidreleasesindicateconcentrationsorreleaserates>2xODAMlimitsfor&#xb3;60min.(Note2)AS1.3Fieldsurveyindicatesclosedwindowdoserate>0.1Rem/hrthatisexpectedtocontinuefor
&#xb3;60min.atorbeyondthesiteboundary(Note1)ORFieldsurveysampleanalysisindicatesthyroidCDE>0.5Remfor1hrofinhalationatorbeyondthesiteboundaryFieldsurveyresultsindicateclosedwindowdoserates>1Rem/hrexpectedtocontinuefor
&#xb3;60min.atorbeyondthesiteboundary(Note1)ORAnalysesoffieldsurveysamplesindicatethyroidCDE>5Remfor1hrofinhalationatorbeyondthesiteboundaryAG1.3NoneHS4.1AhostileactionisoccurringorhasoccurredwithintheProtectedAreaasreportedbytheSecurityShiftSupervisorHA1.1Seismicevent>0.1gasindicatedbySMA-3StrongMotionAccelographorAlarmB-3/A-1EMERGENCYSEISMICHIGHLEVELANDEarthquakeconfirmedbyanyofthefollowing:*Earthquakefeltinplant*NationalEarthquakeInformationCenter
*ControlRoomindicationofdegradedperformanceofsystemsrequiredforthesafeshutdownoftheplantHG4.1AhostileactionhasoccurredsuchthatplantpersonnelareunabletooperateequipmentrequiredtomaintainsafetyfunctionsORAhostileactionhascausedfailureofSpentFuelCoolingSystemsandimminentfueldamageislikelyforafreshlyoff-loadedreactorcoreinpoolAhostileactionisoccurringorhasoccurredwithintheOwnerControlledAreaasreportedbytheSecurityShiftSupervisorORAvalidatednotificationfromNRCofanairlinerattackthreatwithin30min.ofthesiteHU4.1AsecurityconditionthatdoesnotinvolveahostileactionasreportedbytheSecurityShiftSupervisorORAcrediblesite-specificsecuritythreatnotificationORAvalidatednotificationfromNRCprovidinginformationofanaircraftthreatHA4.1HU1.1Seismiceventidentifiedbyanytwoofthefollowing:*SMA-3StrongMotionAccelographactuatedorAlarmB-3/B-1SEISMICEVENT*Earthquakefeltinplant*NationalEarthquakeInformationCenterHU2.1FireinanyTableH-1areanotextinguishedwithin15min.ofControlRoomnotificationorreceiptofavalidControlRoomalarmduetofire(Note3)HA2.1 FireorexplosionresultinginEITHER:VisibledamagetoanyTableH-1areacontainingsafetysystemsorcomponentsORControlRoomindicationofdegradedperformanceofsafetysystemsHU3.1Toxic,corrosive,asphyxiantorflammablegasesinamountsthathaveorcouldaffectnormalplantoperationsHA3.1AccesstoanyTableH-1areaisprohibitedduetotoxic,corrosive,asphyxiantorflammablegaseswhichjeopardizeoperationofsystemsrequiredtomaintainsafeoperationsorsafelyshutdownthereactor(Note7)HU1.2TornadostrikingwithinProtectedAreaboundaryORSustainedhighwinds>100mphHU2.2ExplosionwithintheProtectedAreaHU1.3MainturbinefailureresultingincasingpenetrationordamagetoturbineorgeneratorsealsHU1.4FloodinginanyTableH-1areathathasthepotentialtoaffectsafety-relatedequipmentrequiredbyTechnicalSpecificationsforthecurrentoperatingmodeHA1.2Tornadostrikingorhighwinds>100mphresultinginEITHER:VisibledamagetoanyTableH-1areastructurecontainingsafetysystemsorcomponentsORControlRoomindicationofdegradedperformanceofsafetysystemsHA1.6VehiclecrashresultinginEITHER:VisibledamagetoanyTableH-1areastructurecontainingsafetysystemsorcomponentsORControlRoomindicationofdegradedperformanceofsafetysystemsHA1.3Mainturbinefailure-generatedprojectilesresultinEITHER:VisibledamagetoorpenetrationofanyTableH-1areastructurecontainingsafetysystemsorcomponentsORControlRoomindicationofdegradedperformanceofsafetysystemsHA1.4FloodinginanyTableH-1arearesultinginEITHER:AnelectricalshockhazardthatprecludesaccesstooperateormonitorsafetyequipmentORControlRoomindicationofdegradedperformanceofsafetysystemsHU3.2Recommendationbylocal,countyorstateofficialstoevacuateorsheltersitepersonnelbasedonanoffsiteeventHU6.1OtherconditionsexistwhichinthejudgmentoftheEmergencyDirectorindicatethatEITHER:EventsareinprogressorhaveoccurredwhichindicateapotentialdegradationofthelevelofsafetyoftheplantORAsecuritythreattofacilityprotectionhasbeeninitiatedNoreleasesofradioactivematerialrequiringoffsiteresponseormonitoringareexpectedunlessfurtherdegradationofsafetysystemsoccursProcedure5.1ASD,AlternateShutdown,orProcedure5.4FIRE-S/D,FireInducedShutdownFromOutsidetheControlRoom,requiresControlRoomevacuationHA5.1HS5.1ControlRoomevacuationhasbeeninitiatedANDControloftheplantcannotbeestablishedwithin15min.OtherconditionsexistwhichinthejudgmentoftheEmergencyDirectorindicatethateventsareinprogressorhaveoccurredwhichinvolveEITHER:Anactualorpotentialsubstantialdegradationofthe levelofsafetyoftheplantORAsecurityeventthatinvolvesprobablelifethreateningrisktositepersonnelordamagetositeequipmentbecauseofhostileactionAnyreleasesareexpectedtobelimitedtosmallfractionsoftheEPAProtectiveActionGuidelineexposurelevelsbeyondthesiteboundaryHA6.1HS6.1Otherconditionsexistwhichinthejudgmentofthe EmergencyDirectorindicatethateventsareinprogressorhaveoccurredwhichinvolveEITHER:AnactualorlikelymajorfailuresofplantfunctionsneededforprotectionofthepublicORHostileactionthatresultsinintentionaldamageormaliciousacts;1)towardsitepersonnelorequipmentthatcouldleadtothelikelyfailureofor;2)thatpreventeffectiveaccesstoequipmentneededfortheprotection ofthepublicAnyreleasesarenotexpectedtoresultinexposurelevelswhichexceedEPAProtectiveActionGuidelineexposurelevels(1RemTEDEand5RemthyroidCDE)beyondthesiteboundaryHG6.1OtherconditionsexistwhichinthejudgmentoftheEmergencyDirectorindicatethateventsareinprogressorhaveoccurredwhichinvolveEITHER:ActualorimminentsubstantialcoredegradationormeltingwithpotentialforlossofcontainmentintegrityORHostileactionthatresultsinanactuallossofphysicalcontrolofthefacilityReleasescanbereasonablyexpectedtoexceedEPAProtectiveActionGuidelineexposurelevels(1RemTEDEand5RemthyroidCDE)beyondthesiteboundaryNone5DEF12345DEF12345DEF12345DEF12345DEF12345DEF12345DEF12345DEF12345DEF12345DEF12345DEF12345DEF12345DEF12345DEF12345DEF12345DEF1234HU1.5Highriver/forebaywaterlevel>899'MSLORLowriverlevel/forebay<870'MSLHA1.5Highriver/forebaywaterlevel>902'MSLORLowriver/forebaylevel<865'MSLAnyvalidliquideffluentmonitorreading>TableA-1column"UE"for&#xb3;60min.(Note2)Anyvalidgaseousmonitorreading>TableA-1column"UE"for&#xb3;60min.(Note2)Anyvalidgaseousmonitorreading>TableA-1column"SAE"for&#xb3;15min.(Note1)Anyvalidgaseousmonitorreading>TableA-1column"Alert"for
&#xb3;15min.(Note2)Anyvalidgaseousmonitorreading>TableA-1column"GE"for&#xb3;15min.(Note1)AG1.1Anyvalidliquideffluentmonitorreading>TableA-1column"Alert"for
&#xb3;15min.(Note2)5DEF12345DEF12345DEF12345DEF12345DEF12345DEF12345DEF12345DEF12345DEF12345DEF12345DEF1234DamagetoirradiatedfuelORlossofwaterlevel(uncoveringirradiatedfueloutsidetheRPV)thatcausesEITHERofthefollowing:ValidRMA-RA-1FuelPoolAreaRadreading>50R/hrORValidRMP-RM-452A-DRxBldgVentExhaustPlenumHi-HialarmAA3.1Doserates>15mRem/hrinEITHERofthefollowingareasrequiringcontinuousoccupancytomaintainplantsafetyfunctions:MainControlRoom(RM-RA-20)ORCASAU2.2Unplannedvalidarearadiationmonitorreadingorsurveyresultsrisebyafactorof1,000overnormallevels**Normallevelscanbeconsideredasthehighestreadinginthepast24hoursexcludingthecurrentpeakvalueAA2.2AwaterleveldropinthereactorrefuelingcavityorspentfuelpoolthatwillresultinirradiatedfuelbecominguncoveredNoneNoneNoneDamagetoaloadedcaskconfinementboundaryEU1.1N/ANoneNoneNoneNoneColdSD/RefuelSystemMalfunct.LossofACPowerCU1.1ACpowercapabilitytocritical4160Vbuses1Fand1Greducedtoasinglepowersource(TableC-4)for
&#xb3;15min.suchthatanyadditionalsinglefailurewouldresultinlossofallACpowertocriticalbuses(Note3)<105VDCbusvoltageindicationsonallTechnicalSpecificationrequired125VDCbusesfor
&#xb3;15min.(Note3)CU6.1RPVlevel<-158in.for
&#xb3;30min.(Note3)ANDAnyContainmentChallengeindication,TableC-5CA2.1RPVlevel<-42in.ORRPVlevelcannotbemonitoredfor
&#xb3;15min.(Note3)withanyunexplainedRPVleakageindication,TableC-1CG2.1CS2.1RPVLevelComm.CU4.1CU2.1RPVlevelcannotberestoredandmaintained>+3in.for&#xb3;15min.(Note3)duetoRCSleakageCA1.1LossofalloffsiteandallonsiteACpower(TableC-4)tocritical4160Vbuses1Fand1Gfor
&#xb3;15min.(Note3)UnplannedRPVleveldropfor
&#xb3;15min(Note3)belowEITHER:RPVflange(LI-86:206in.normalcalibration,113.75in.elevatedcalibration)ORRPVlevelbandwhentheRPVlevelbandisestablishedbelowtheRPVflangeCU2.2RPVlevelcannotbemonitoredwithanyunexplainedRPVleakageindication,TableC-1CU2.3CS2.2CS2.3RCSTemp.AnyunplannedeventresultsinRCStemperature>212&deg;FduetolossofdecayheatremovalcapabilityCU3.1LossofallRCStemperatureandRPVlevelindicationfor&#xb3;15min.(Note3)CU3.2CA3.1AnyunplannedeventresultsinEITHER:RCStemperature>212&deg;Ffor>TableC-3duration(Note4)ORRPVpressureincrease>10psigduetolossofRCScoolingNoneInadvertentCriticality C 2 3 4 5Anunplannedsustainedpositiveperiodobservedonnuclearinstrumentation 1 GENERALEMERGENCY SITEAREAEMERGENCY ALERT UNUSUALEVENT45DEFLossofallTableC-2onsite(internal)communicationmethodsaffectingtheabilitytoperformroutineoperationsORLossofallTableC-2offsite(external)communicationmethodsaffectingtheabilitytoperformoffsitenotifications 5 4CU5.14545WithContainmentClosurenotestablished,RPVlevel<-48in.(Note4)WithContainmentClosureestablished,RPVlevel<-158in.(Note4)4545NoneNoneNone4545RPVlevelcannotbemonitoredfor
&#xb3;30min.(Note3)withalossofinventoryasindicatedbyEITHER:UnexplainedRPVleakageindication,TableC-1ORErraticSourceRangeMonitorindication45RPVlevelcannotbemonitoredfor
&#xb3;30min.(Note3)withcoreuncoveryindicatedbyEITHER:UnexplainedRPVleakageindication,TableC-1ORErraticSourceRangeMonitorindicationANDAnyContainmentChallengeindication,TableC-5CG2.2NoneNoneNone 54545454545None5DEF12345DEF12345DEF12345DEF12345DEF12345DEF12345DEF12345DEF12345DEF12345DEF12345DEF12345DEF12345DEF12345DEF12345DEF12345DEF12345DEF1234LossofDCPower 645 5DEFTableC-4ACPowerSourcesOffsiteOnsite*StartupStationServiceTransformer*EmergencyStationServiceTransformer*Backfeed345kvlinethroughMainPowerTransformertotheNormalStationServiceTransformer(Note8)*DG-1*DG-2TableC-5ContainmentChallengeIndications*ContainmentClosurenotestablished(Note4)*DeflagrationconcentrationsexistinsidePC
&#xb3;6%H 2indrywellortorusAND&#xb3;5%O 2indrywellortorus*UnplannedriseinPCpressure*SecondaryContainmentarearadiation>1000mR/hr(EOP-5ATable10)NoneStationIntercomSystem"Gaitronics"SiteUHFRadioConsolesRadioPagingSystemAlternateIntercomSoundPowerSystemCNSOn-SiteCellPhoneSystem Telephonesystem(PBX)FederalTelecommunicationsSystem(FTS2001)LocalTelephones(C.O.Lines)CNSStateNotificationTelephonesOnsite(internal)Offsite(external)
X X X
X X
X XTableC-2CommunicationsSystemsSystem X X
X X
X1.RCSintact(ContainmentClosureN/A)2.ContainmentClosureestablishedANDRCSnotintact3.ContainmentClosurenotestablishedANDRCSnotintact60min.*20min.*0min.TableC-3RCSReheatDurationThresholds*IfanRCSheatremovalsystemisinoperationwithinthistimeframeandRCStemperatureisbeingreduced,theEALisnotapplicableTableH-1SafeShutdownAreas*ReactorBuilding*ControlBuilding*ServiceWaterPumpRoom*DieselGeneratorBuilding*CableExpansionRoomGASEOUSLIQUIDTableA-1EffluentMonitorClassificationThresholdsMonitorGEfor&#xb3;15min.SAEfor&#xb3;15min.ALERTfor&#xb3;15min.UEfor&#xb3;60min.ERP3.50E+07&#xb5;Ci/sec2.24E+05&#xb5;Ci/sec
*witheffluentdischargenotisolatedRxBldgVent8.48E+04&#xb5;Ci/sec3.50E+08&#xb5;Ci/secTurbBldgVent9.02E+04&#xb5;Ci/secRW/ARWBldgVent9.08E+04&#xb5;Ci/secRadWasteEffluent----------Thelesserof*:2xcalculatedalarmvaluesORmonitorupscaleServiceWaterEffluent----------4.80E-06&#xb5;Ci/cc4.80E-04&#xb5;Ci/cc2.80E+06&#xb5;Ci/sec5.45E+05&#xb5;Ci/sec5.62E+05&#xb5;Ci/sec5.64E+05&#xb5;Ci/secThelesserof*:200xcalculatedalarmvaluesORmonitorupscale3.50E+06&#xb5;Ci/sec3.50E+07&#xb5;Ci/sec3.50E+06&#xb5;Ci/sec3.50E+07&#xb5;Ci/sec3.50E+06&#xb5;Ci/sec3.50E+07&#xb5;Ci/secNotes1.TheEmergencyDirectorshouldnotwaituntiltheapplicabletimehaselapsed,butshoulddeclaretheeventassoonasitisdeterminedthattheconditionwilllikelyexceedtheapplicabletime.Ifdoseassessmentresultsareavailable,declarationshouldbebasedondoseassessmentinsteadofradiationmonitorvalues.(SeeEALAS1.2/AG1.2.)Donotdelaydeclarationawaitingdoseassessmentresults.2.TheEmergencyDirectorshouldnotwaituntiltheapplicabletimehaselapsed,butshoulddeclaretheeventassoonasitisdeterminedthatthereleasedurationhasexceeded,orwilllikelyexceed,theapplicabletime.Intheabsenceofdatatothecontrary,assumethatthereleasedurationhasexceededtheapplicabletimeifanongoingreleaseisdetectedandthereleasestarttimeisunknown.3.TheEmergencyDirectorshouldnotwaituntiltheapplicabletimehaselapsed,butshoulddeclaretheeventassoonasitisdeterminedthattheconditionwilllikelyexceedtheapplicabletime.4.ContainmentClosureistheactiontakentosecureprimaryorsecondarycontainmentanditsassociatedstructures,systems,andcomponentsasafunctionalbarriertofissionproductreleaseunderexistingplantconditions.ContainmentClosurerequirementsarespecifiedinAdministrativeProcedure0.50.5,OutageShutdownSafety.5.ManualscrammethodsforEALSA2.1andEALSS2.1arethefollowing:*ReactorScrampushbuttons*ReactorModeswitchinSHUTDOWN*ManualorautoactuationofARI6.SeeTableF-1,FissionProductBarrierMatrix,forpossibleescalationabovetheUnusualEventduetoRCSLeakage.7.Iftheequipmentinthestatedareawasalreadyinoperable,oroutofservice,beforetheeventoccurred,thenthisEALshouldnotbedeclaredasitwillhavenoadverseimpactontheabilityoftheplanttosafelyoperateorsafelyshutdownbeyondthatalreadyallowedbyTechnicalSpecificationsatthetimeoftheevent.8.Thetimerequiredtoestablishthebackfeedislikelylongerthanthespecifiedtimeinterval.Ifoff-normalplantconditionshavealreadyestablishedthebackfeed,itspowertothesafety-relatedbusesmaybeconsideredanoffsitepowersource.
PreparedforNPPDby:OperationsSupportServices,Inc.-www.ossi-net.com(Rev.84/2/10)Modes: 1PowerOperationDefueledDEF 2Startup 5Refueling 3HotShutdown 4ColdShutdownCooperNuclearStationEmergencyActionLevelMatrixEPIP5.7.1Attachment4,Rev.11Sequentialnumberwithinsubcategory/classificationSubcategorynumber(1ifnosubcategory)XXX.XCategory(A,H,S,F,C,E)Emergencyclassification(G,S,A,U)EALIdentifierMODE4,5orDEF GENERALEMERGENCY SITEAREAEMERGENCY ALERT UNUSUALEVENTOnsiteRadConditions
&SpentFuelPoolEventsAbnorm.RadRelease/RadEffluent A 2OffsiteRadConditions 1Hazards&OtherCondi-tionsAffect-ingPlantSafety HNaturalorDestructivePhenomenaFireorExplosionControlRoomEvacuationHazardousGasSecurityJudgment 1 2 3 4 5 6ISFSI ETableC-1RPVLeakageIndications*Drywellequipmentdrainsumplevelrise*Drywellfloordrainsumplevelrise*ReactorBuildingequipmentdrainsumplevelrise*ReactorBuildingfloordrainsumplevelrise*SuppressionPoolwaterlevelrise*RPVmake-upraterise*ObservationofunisolableRCSleakageMCR/CASRad 3AU1.1AU1.2AA1.1AS1.1AS1.2Doseassessmentusingactualmeteorologyindicatesdoses>0.1RemTEDEor>0.5RemthyroidCDEatorbeyondthesiteboundaryAA1.2AA1.3Confirmedsampleanalysesforgaseousorliquidreleasesindicateconcentrationsorreleaserates>200xODAMlimitsfor&#xb3;15min.(Note2)AU2.1Unplannedwaterleveldropinthereactorcavityorspentfuelpoolasindicatedbyanyofthefollowing:*LI-86(calibratedto1001'elev.)*Spentfuelpoollowlevelalarm*VisualobservationANDValidarearadiationmonitorreadingriseonRMA-RA-1orRMA-RA-2AA2.1Doseassessmentusingactualmeteorologyindicatesdoses>1RemTEDEor>5RemthyroidCDEatorbeyondthesiteboundaryAG1.2AU1.3Confirmedsampleanalysesforgaseousorliquidreleasesindicateconcentrationsorreleaserates>2xODAMlimitsfor&#xb3;60min.(Note2)AS1.3Fieldsurveyindicatesclosedwindowdoserate>0.1Rem/hrthatisexpectedtocontinuefor
&#xb3;60min.atorbeyondthesiteboundary(Note1)ORFieldsurveysampleanalysisindicatesthyroidCDE>0.5Remfor1hrofinhalationatorbeyondthesiteboundaryFieldsurveyresultsindicateclosedwindowdoserates>1Rem/hrexpectedtocontinuefor
&#xb3;60min.atorbeyondthesiteboundary(Note1)ORAnalysesoffieldsurveysamplesindicatethyroidCDE>5Remfor1hrofinhalationatorbeyondthesiteboundaryAG1.3NoneHS4.1AhostileactionisoccurringorhasoccurredwithintheProtectedAreaasreportedbytheSecurityShiftSupervisorHA1.1Seismicevent>0.1gasindicatedbySMA-3StrongMotionAccelographorAlarmB-3/A-1EMERGENCYSEISMICHIGHLEVELANDEarthquakeconfirmedbyanyofthefollowing:*Earthquakefeltinplant*NationalEarthquakeInformationCenter
*ControlRoomindicationofdegradedperformanceofsystemsrequiredforthesafeshutdownoftheplantHG4.1AhostileactionhasoccurredsuchthatplantpersonnelareunabletooperateequipmentrequiredtomaintainsafetyfunctionsORAhostileactionhascausedfailureofSpentFuelCoolingSystemsandimminentfueldamageislikelyforafreshlyoff-loadedreactorcoreinpoolAhostileactionisoccurringorhasoccurredwithintheOwnerControlledAreaasreportedbytheSecurityShiftSupervisorORAvalidatednotificationfromNRCofanairlinerattackthreatwithin30min.ofthesiteHU4.1AsecurityconditionthatdoesnotinvolveahostileactionasreportedbytheSecurityShiftSupervisorORAcrediblesite-specificsecuritythreatnotificationORAvalidatednotificationfromNRCprovidinginformationofanaircraftthreatHA4.1HU1.1Seismiceventidentifiedbyanytwoofthefollowing:*SMA-3StrongMotionAccelographactuatedorAlarmB-3/B-1SEISMICEVENT*Earthquakefeltinplant*NationalEarthquakeInformationCenterHU2.1FireinanyTableH-1areanotextinguishedwithin15min.ofControlRoomnotificationorreceiptofavalidControlRoomalarmduetofire(Note3)HA2.1 FireorexplosionresultinginEITHER:VisibledamagetoanyTableH-1areacontainingsafetysystemsorcomponentsORControlRoomindicationofdegradedperformanceofsafetysystemsHU3.1Toxic,corrosive,asphyxiantorflammablegasesinamountsthathaveorcouldaffectnormalplantoperationsHA3.1AccesstoanyTableH-1areaisprohibitedduetotoxic,corrosive,asphyxiantorflammablegaseswhichjeopardizeoperationofsystemsrequiredtomaintainsafeoperationsorsafelyshutdownthereactor(Note7)HU1.2TornadostrikingwithinProtectedAreaboundaryORSustainedhighwinds>100mphHU2.2ExplosionwithintheProtectedAreaHU1.3MainturbinefailureresultingincasingpenetrationordamagetoturbineorgeneratorsealsHU1.4FloodinginanyTableH-1areathathasthepotentialtoaffectsafety-relatedequipmentrequiredbyTechnicalSpecificationsforthecurrentoperatingmodeHA1.2Tornadostrikingorhighwinds>100mphresultinginEITHER:VisibledamagetoanyTableH-1areastructurecontainingsafetysystemsorcomponentsORControlRoomindicationofdegradedperformanceofsafetysystemsHA1.6VehiclecrashresultinginEITHER:VisibledamagetoanyTableH-1areastructurecontainingsafetysystemsorcomponentsORControlRoomindicationofdegradedperformanceofsafetysystemsHA1.3Mainturbinefailure-generatedprojectilesresultinEITHER:VisibledamagetoorpenetrationofanyTableH-1areastructurecontainingsafetysystemsorcomponentsORControlRoomindicationofdegradedperformanceofsafetysystemsHA1.4FloodinginanyTableH-1arearesultinginEITHER:AnelectricalshockhazardthatprecludesaccesstooperateormonitorsafetyequipmentORControlRoomindicationofdegradedperformanceofsafetysystemsHU3.2Recommendationbylocal,countyorstateofficialstoevacuateorsheltersitepersonnelbasedonanoffsiteeventHU6.1OtherconditionsexistwhichinthejudgmentoftheEmergencyDirectorindicatethatEITHER:EventsareinprogressorhaveoccurredwhichindicateapotentialdegradationofthelevelofsafetyoftheplantORAsecuritythreattofacilityprotectionhasbeeninitiatedNoreleasesofradioactivematerialrequiringoffsiteresponseormonitoringareexpectedunlessfurtherdegradationofsafetysystemsoccursProcedure5.1ASD,AlternateShutdown,orProcedure5.4FIRE-S/D,FireInducedShutdownFromOutsidetheControlRoom,requiresControlRoomevacuationHA5.1HS5.1ControlRoomevacuationhasbeeninitiatedANDControloftheplantcannotbeestablishedwithin15min.OtherconditionsexistwhichinthejudgmentoftheEmergencyDirectorindicatethateventsareinprogressorhaveoccurredwhichinvolveEITHER:Anactualorpotentialsubstantialdegradationofthe levelofsafetyoftheplantORAsecurityeventthatinvolvesprobablelifethreateningrisktositepersonnelordamagetositeequipmentbecauseofhostileactionAnyreleasesareexpectedtobelimitedtosmallfractionsoftheEPAProtectiveActionGuidelineexposurelevelsbeyondthesiteboundaryHA6.1HS6.1Otherconditionsexistwhichinthejudgmentofthe EmergencyDirectorindicatethateventsareinprogressorhaveoccurredwhichinvolveEITHER:AnactualorlikelymajorfailuresofplantfunctionsneededforprotectionofthepublicORHostileactionthatresultsinintentionaldamageormaliciousacts;1)towardsitepersonnelorequipmentthatcouldleadtothelikelyfailureofor;2)thatpreventeffectiveaccesstoequipmentneededfortheprotection ofthepublicAnyreleasesarenotexpectedtoresultinexposurelevelswhichexceedEPAProtectiveActionGuidelineexposurelevels(1RemTEDEand5RemthyroidCDE)beyondthesiteboundaryHG6.1OtherconditionsexistwhichinthejudgmentoftheEmergencyDirectorindicatethateventsareinprogressorhaveoccurredwhichinvolveEITHER:ActualorimminentsubstantialcoredegradationormeltingwithpotentialforlossofcontainmentintegrityORHostileactionthatresultsinanactuallossofphysicalcontrolofthefacilityReleasescanbereasonablyexpectedtoexceedEPAProtectiveActionGuidelineexposurelevels(1RemTEDEand5RemthyroidCDE)beyondthesiteboundaryNone5DEF12345DEF12345DEF12345DEF12345DEF12345DEF12345DEF12345DEF12345DEF12345DEF12345DEF12345DEF12345DEF12345DEF12345DEF12345DEF1234HU1.5Highriver/forebaywaterlevel>899'MSLORLowriverlevel/forebay<870'MSLHA1.5Highriver/forebaywaterlevel>902'MSLORLowriver/forebaylevel<865'MSLAnyvalidliquideffluentmonitorreading>TableA-1column"UE"for&#xb3;60min.(Note2)Anyvalidgaseousmonitorreading>TableA-1column"UE"for&#xb3;60min.(Note2)Anyvalidgaseousmonitorreading>TableA-1column"SAE"for&#xb3;15min.(Note1)Anyvalidgaseousmonitorreading>TableA-1column"Alert"for
&#xb3;15min.(Note2)Anyvalidgaseousmonitorreading>TableA-1column"GE"for&#xb3;15min.(Note1)AG1.1Anyvalidliquideffluentmonitorreading>TableA-1column"Alert"for
&#xb3;15min.(Note2)5DEF12345DEF12345DEF12345DEF12345DEF12345DEF12345DEF12345DEF12345DEF12345DEF12345DEF1234DamagetoirradiatedfuelORlossofwaterlevel(uncoveringirradiatedfueloutsidetheRPV)thatcausesEITHERofthefollowing:ValidRMA-RA-1FuelPoolAreaRadreading>50R/hrORValidRMP-RM-452A-DRxBldgVentExhaustPlenumHi-HialarmAA3.1Doserates>15mRem/hrinEITHERofthefollowingareasrequiringcontinuousoccupancytomaintainplantsafetyfunctions:MainControlRoom(RM-RA-20)ORCASAU2.2Unplannedvalidarearadiationmonitorreadingorsurveyresultsrisebyafactorof1,000overnormallevels**Normallevelscanbeconsideredasthehighestreadinginthepast24hoursexcludingthecurrentpeakvalueAA2.2AwaterleveldropinthereactorrefuelingcavityorspentfuelpoolthatwillresultinirradiatedfuelbecominguncoveredNoneNoneNoneDamagetoaloadedcaskconfinementboundaryEU1.1N/ANoneNoneNoneNoneColdSD/RefuelSystemMalfunct.LossofACPowerCU1.1ACpowercapabilitytocritical4160Vbuses1Fand1Greducedtoasinglepowersource(TableC-4)for
&#xb3;15min.suchthatanyadditionalsinglefailurewouldresultinlossofallACpowertocriticalbuses(Note3)<105VDCbusvoltageindicationsonallTechnicalSpecificationrequired125VDCbusesfor
&#xb3;15min.(Note3)CU6.1RPVlevel<-158in.for
&#xb3;30min.(Note3)ANDAnyContainmentChallengeindication,TableC-5CA2.1RPVlevel<-42in.ORRPVlevelcannotbemonitoredfor
&#xb3;15min.(Note3)withanyunexplainedRPVleakageindication,TableC-1CG2.1CS2.1RPVLevelComm.CU4.1CU2.1RPVlevelcannotberestoredandmaintained>+3in.for&#xb3;15min.(Note3)duetoRCSleakageCA1.1LossofalloffsiteandallonsiteACpower(TableC-4)tocritical4160Vbuses1Fand1Gfor
&#xb3;15min.(Note3)UnplannedRPVleveldropfor
&#xb3;15min(Note3)belowEITHER:RPVflange(LI-86:206in.normalcalibration,113.75in.elevatedcalibration)ORRPVlevelbandwhentheRPVlevelbandisestablishedbelowtheRPVflangeCU2.2RPVlevelcannotbemonitoredwithanyunexplainedRPVleakageindication,TableC-1CU2.3CS2.2CS2.3RCSTemp.AnyunplannedeventresultsinRCStemperature>212&deg;FduetolossofdecayheatremovalcapabilityCU3.1LossofallRCStemperatureandRPVlevelindicationfor&#xb3;15min.(Note3)CU3.2CA3.1AnyunplannedeventresultsinEITHER:RCStemperature>212&deg;Ffor>TableC-3duration(Note4)ORRPVpressureincrease>10psigduetolossofRCScoolingNoneInadvertentCriticality C 2 3 4 5Anunplannedsustainedpositiveperiodobservedonnuclearinstrumentation 1 GENERALEMERGENCY SITEAREAEMERGENCY ALERT UNUSUALEVENT45DEFLossofallTableC-2onsite(internal)communicationmethodsaffectingtheabilitytoperformroutineoperationsORLossofallTableC-2offsite(external)communicationmethodsaffectingtheabilitytoperformoffsitenotifications 5 4CU5.14545WithContainmentClosurenotestablished,RPVlevel<-48in.(Note4)WithContainmentClosureestablished,RPVlevel<-158in.(Note4)4545NoneNoneNone4545RPVlevelcannotbemonitoredfor
&#xb3;30min.(Note3)withalossofinventoryasindicatedbyEITHER:UnexplainedRPVleakageindication,TableC-1ORErraticSourceRangeMonitorindication45RPVlevelcannotbemonitoredfor
&#xb3;30min.(Note3)withcoreuncoveryindicatedbyEITHER:UnexplainedRPVleakageindication,TableC-1ORErraticSourceRangeMonitorindicationANDAnyContainmentChallengeindication,TableC-5CG2.2NoneNoneNone 54545454545None5DEF12345DEF12345DEF12345DEF12345DEF12345DEF12345DEF12345DEF12345DEF12345DEF12345DEF12345DEF12345DEF12345DEF12345DEF12345DEF12345DEF1234LossofDCPower 645 5DEFTableC-4ACPowerSourcesOffsiteOnsite*StartupStationServiceTransformer*EmergencyStationServiceTransformer*Backfeed345kvlinethroughMainPowerTransformertotheNormalStationServiceTransformer(Note8)*DG-1*DG-2TableC-5ContainmentChallengeIndications*ContainmentClosurenotestablished(Note4)*DeflagrationconcentrationsexistinsidePC
&#xb3;6%H 2indrywellortorusAND&#xb3;5%O 2indrywellortorus*UnplannedriseinPCpressure*SecondaryContainmentarearadiation>1000mR/hr(EOP-5ATable10)NoneStationIntercomSystem"Gaitronics"SiteUHFRadioConsolesRadioPagingSystemAlternateIntercomSoundPowerSystemCNSOn-SiteCellPhoneSystem Telephonesystem(PBX)FederalTelecommunicationsSystem(FTS2001)LocalTelephones(C.O.Lines)CNSStateNotificationTelephonesOnsite(internal)Offsite(external)
X X X
X X
X XTableC-2CommunicationsSystemsSystem X X
X X
X1.RCSintact(ContainmentClosureN/A)2.ContainmentClosureestablishedANDRCSnotintact3.ContainmentClosurenotestablishedANDRCSnotintact60min.*20min.*0min.TableC-3RCSReheatDurationThresholds*IfanRCSheatremovalsystemisinoperationwithinthistimeframeandRCStemperatureisbeingreduced,theEALisnotapplicableTableH-1SafeShutdownAreas*ReactorBuilding*ControlBuilding*ServiceWaterPumpRoom*DieselGeneratorBuilding*CableExpansionRoomGASEOUSLIQUIDTableA-1EffluentMonitorClassificationThresholdsMonitorGEfor&#xb3;15min.SAEfor&#xb3;15min.ALERTfor&#xb3;15min.UEfor&#xb3;60min.ERP3.50E+07&#xb5;Ci/sec2.24E+05&#xb5;Ci/sec
*witheffluentdischargenotisolatedRxBldgVent8.48E+04&#xb5;Ci/sec3.50E+08&#xb5;Ci/secTurbBldgVent9.02E+04&#xb5;Ci/secRW/ARWBldgVent9.08E+04&#xb5;Ci/secRadWasteEffluent----------Thelesserof*:2xcalculatedalarmvaluesORmonitorupscaleServiceWaterEffluent----------4.80E-06&#xb5;Ci/cc4.80E-04&#xb5;Ci/cc2.80E+06&#xb5;Ci/sec5.45E+05&#xb5;Ci/sec5.62E+05&#xb5;Ci/sec5.64E+05&#xb5;Ci/secThelesserof*:200xcalculatedalarmvaluesORmonitorupscale3.50E+06&#xb5;Ci/sec3.50E+07&#xb5;Ci/sec3.50E+06&#xb5;Ci/sec3.50E+07&#xb5;Ci/sec3.50E+06&#xb5;Ci/sec3.50E+07&#xb5;Ci/secNotes1.TheEmergencyDirectorshouldnotwaituntiltheapplicabletimehaselapsed,butshoulddeclaretheeventassoonasitisdeterminedthattheconditionwilllikelyexceedtheapplicabletime.Ifdoseassessmentresultsareavailable,declarationshouldbebasedondoseassessmentinsteadofradiationmonitorvalues.(SeeEALAS1.2/AG1.2.)Donotdelaydeclarationawaitingdoseassessmentresults.2.TheEmergencyDirectorshouldnotwaituntiltheapplicabletimehaselapsed,butshoulddeclaretheeventassoonasitisdeterminedthatthereleasedurationhasexceeded,orwilllikelyexceed,theapplicabletime.Intheabsenceofdatatothecontrary,assumethatthereleasedurationhasexceededtheapplicabletimeifanongoingreleaseisdetectedandthereleasestarttimeisunknown.3.TheEmergencyDirectorshouldnotwaituntiltheapplicabletimehaselapsed,butshoulddeclaretheeventassoonasitisdeterminedthattheconditionwilllikelyexceedtheapplicabletime.4.ContainmentClosureistheactiontakentosecureprimaryorsecondarycontainmentanditsassociatedstructures,systems,andcomponentsasafunctionalbarriertofissionproductreleaseunderexistingplantconditions.ContainmentClosurerequirementsarespecifiedinAdministrativeProcedure0.50.5,OutageShutdownSafety.5.ManualscrammethodsforEALSA2.1andEALSS2.1arethefollowing:*ReactorScrampushbuttons*ReactorModeswitchinSHUTDOWN*ManualorautoactuationofARI6.SeeTableF-1,FissionProductBarrierMatrix,forpossibleescalationabovetheUnusualEventduetoRCSLeakage.7.Iftheequipmentinthestatedareawasalreadyinoperable,oroutofservice,beforetheeventoccurred,thenthisEALshouldnotbedeclaredasitwillhavenoadverseimpactontheabilityoftheplanttosafelyoperateorsafelyshutdownbeyondthatalreadyallowedbyTechnicalSpecificationsatthetimeoftheevent.8.Thetimerequiredtoestablishthebackfeedislikelylongerthanthespecifiedtimeinterval.Ifoff-normalplantconditionshavealreadyestablishedthebackfeed,itspowertothesafety-relatedbusesmaybeconsideredanoffsitepowersource.}}

Latest revision as of 12:51, 9 March 2020

2015-04-DRAFT Written Exam
ML15127A501
Person / Time
Site: Cooper Entergy icon.png
Issue date: 04/21/2015
From: Vincent Gaddy
Operations Branch IV
To:
References
Download: ML15127A501 (541)
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Text

U.S. Nuclear Regulatory Commission Site-Specific RO Written Examination Applicant Information Name:

Date: Facility/Unit: Cooper Nuclear Station Region: I II III IV Reactor Type: W CE BW GE Start Time: Finish Time:

Instructions Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. To pass the examination, you must achieve a final grade of at least 80.00 percent. Examination papers will be collected 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after the examination begins.

Applicant Certification All work done on this examination is my own. I have neither given nor received aid.

Applicants Signature Results Examination Value __________ Points Applicants Score __________ Points Applicants Grade __________ Percent 1

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO Tier # __1__

Group # __1__

K/A # _295001.AK3.04___

Importance Rating _3.4__

295001 Partial or Complete Loss of Forced Core Flow Circulation- Knowledge of the reasons for the following responses as they apply to partial or complete loss of forced core flow circulation:

AK3.04 Reactor SCRAM Question: 1 With the Reactor initially operating at 72% power the following conditions exist:

  • Reactor Recirculation MG A trips on motor generator high air temperature.
  • PMIS shows the plant is operating in the Stability Exclusion Region (red region) of the Power to Flow map.
  • SRM period swings with a fluctuation time of <3 seconds.

Why does Procedure 2.4RR (Reactor Recirculation Abnormal) require a scram?

A. To protect the fuel and ensure fuel clad integrity.

B. RPV water level control is unpredictable due to power swings.

C. It is the fastest way to exit the area of operation while in single loop.

D. To ensure the reactor is shut down before exceeding the RPV pressure limit.

Answer:

A. To protect the fuel and ensure fuel clad integrity.

Explanation:

The reactor is designed such that thermal hydraulic oscillations are prevented or can be readily detected and suppressed without exceeding specified fuel design limits. Specific directions are provided to avoid operation in this region and to immediately exit upon entry. Entries into the Stability Exclusion Region (SER) are not part of normal operation. Although operator action can prevent the occurrence of and protect the reactor from an instability, the APRM Neutron Flux-High (Flow Biased) scram function will suppress oscillations prior to exceeding the Safety Limit MCPR. A manual scram is inserted as an Immediate Operator Action if instability is observed while operating in the SER to ensure MCPR Safety Limit is not challenged, therefore protecting the fuel and ensure fuel clad integrity.

Distracters:

2

B. This option is incorrect because there is no requirement within 2.4RR to scram based upon RPV level swings. The reactor feedwater pumps have anticipatory circuitry which attempts to maintain RPV level in a suitable band so the operators will not think level control is unreliable. This option is plausible because abnormal condition procedure 2.4RXLVL directs scramming the reactor if RPV level cannot be maintained between 12 inches and 50 inches. Due to the very nature of reactor pressure oscillations during neutron flux oscillations, Reactor Feed Pump Turbine speed may be affected due to variations in Reactor Feed Pump output to the reactor. The candidate who believes that power oscillations would cause unreliable or uncontrollable RPV level oscillations per 2.4RR would select this option.

C. This option is incorrect because the direction to scram is not based on the fastest way of exiting the stability exclusion area. The scram is required because the core is exhibiting thermal hydraulic instabilities. Technical Specifications states to immediately exit the region but does not specifically state to scram the reactor. Procedure 2.4RR requires the operator to insert a manual reactor scram based on observed abnormal neutron flux oscillations.

This option is plausible because scramming is an immediate way of exiting the exclusion region.

D. This option is incorrect because RPV pressure limits are not threatened relatively small power swings. The RPV high pressure scram is designed to shut down the reactor before the thermal power transferred to the reactor coolant increases and challenges the integrity of the fuel cladding and the reactor coolant pressure boundary. Reactor power protection from the APRM flow biased high neutron flux will occur before the pressure spike reaches the RPV pressure limit scram. This option is plausible because power swings will cause pressure swings.

Technical Reference(s): 2.4RR Reactor Recirculation Abnormal, Rev. 40 Proposed references to be provided to applicants during examination: ___NONE_______

Learning Objective: COR002-22-02, OPS Reactor Recirculation System, Rev. 32

5. Briefly describe the following concepts as they apply to the Reactor Recirculation system, or to the Recirculation Flow Control system:
i. Power to Flow Map (including normal operation/startup/shutdown and Stability Exclusion Region)
l. Thermal limits Question Source: Bank # _______

Modified Bank # _______

New __X_____

Question History: Last NRC Exam _________

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 (10) 3

Comments:

LOD 2 4

5 6

From 2.4RR procedure change 7

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO Tier # __1__

Group # __1__

K/A # _295003 AA1.03___

Importance Rating _4.3__

295003 Partial or Complete Loss of AC:

Ability to operate and/or monitor the following as they apply to partial or complete loss of A.C. power:

AA1.03 Systems necessary to assure safe plant shutdown Question: 2 With the plant operating at 100% power, a loss of offsite power occurs.

  • Both diesel generators fail to start and CANNOT be started.
  • The Supplemental Diesel Generator is NOT available.
  • HPCI and RCIC recover RPV water level to +35 inches (Narrow Range).

What operational restriction applies to the continued use of HPCI in response to this event, until onsite or offsite electrical power is restored?

HPCI must be secured...

A. after one cycle of operation and must remain off.

B. just prior to the division's battery being exhausted and must remain off.

C. NO LATER THAN 15 minutes from the time that injection flow was reduced and must remain off.

D. after one cycle of operation and must remain off unless RPV level lowers to +3 inches narrow range.

Answer:

A. after one cycle of operation and must remain off.

Explanation:

This procedure assumes that RPV water level and pressure is initially controlled by High Pressure Coolant Injection (HPCI), as directed by the EOPs. CNS has committed to secure HPCI after one cycle of operation, even if EOPs allow HPCI use, in order to extend station battery life during station blackout (SBO). One cycle of HPCI operation is ~ 10 minutes. SBO analysis assumes Reactor Core Isolation Cooling (RCIC) is operable and maintains RPV level 8

and pressure until on-site or off-site electrical power can be restored. If RCIC is unable to perform this function, compensatory actions must be taken to ensure adequate core cooling.

This could include starting HPCI. Since RPV level is restored to +35 inches, HPCI is not needed for adequate core cooling and must be secured after one cycle and must remain off.

Should RPV level subsequently lower to -150 inches (meaning RCIC cannot maintain RPV level as assumed in the analysis), then EOPs allow HPCI to be used as operation is outside the SBO analysis. If level lowers to +3 inches, HPCI cannot be restarted as adequate core cooling is not threatened (RPV level is still approximately 168 inches above top of active fuel).

Distracters:

B. This option is incorrect because HPCI is manually secured after approximately 10 minutes of operation. Operation until the battery is exhausted would be inconsistent with the commitment to secure HPCI after 1 Cycle of operation. The operator who does not correctly recall the restriction in 5.3SBO to extend station battery life would select this option. This option is plausible because HPCI may be utilized without significantly draining battery power as HPCI turbine speed is generally not changed too much to control injection. Significant draining of battery power would be an issue if cycling HPCI valves or allowing the HPCI Auxiliary Oil Pump to run at low turbine speeds or following shutdown of the turbine.

C. This option is incorrect because CNS has a commitment that HPCI should be secured after approximately 10 minutes of operation vs. 15 minutes from the time HPCI flow is reduced. There is no requirement based on when flow is reduced, the only basis is one cycle. Waiting 15 minutes is inconsistent with the commitment to operate HPCI for no more than one cycle (~10 minutes) during a Station Black Out event. The operator who does not correctly recall the restriction in 5.3SBO to extend station battery life would select this option. This option is plausible because the utilization of HPCI and/or RCIC during this event may be required and 15 minutes is a common number for emergency procedure usage. Examples would be 15 minutes to classify an event, 15 minutes below TAF for significant fuel damage to occur, loss of electrical power to classify an event, etc.

D. This option is incorrect because HPCI is secured after one cycle of operation and no procedural step allows restarting per 5.3SBO. If the operator does not remember the restriction in 5.3SBO they would select this option. This option is plausible because EOPs allow the use of HPCI to maintain adequate core cooling at a much lower RPV water level. There is no reason to be concerned with RPV water level at this point because it is at +35 inches (normal band) and RPV water level would have to drop ~200 inches to reach the point where adequate core cooling is challenged. In the event that adequate core cooling is challenged, the EOPs override the emergency procedure 5.3SBO. If level lowers to +3 inches, HPCI cannot be restarted as adequate core cooling is not threatened (RPV level is still approximately 168 inches above top of active fuel).

Technical Reference(s): 5.3SBO Station Blackout, Rev 33.

Proposed references to be provided to applicants during examination: ___NONE_______

9

Learning Objective: INT032-01-31 CNS Abnormal Procedures (RO) Electrical W. Given plant condition(s) and the applicable Abnormal/Emergency Procedure, determine the correct subsequent actions required to mitigate the event(s).

Question Source: Bank # _13338_

Modified Bank # _ _

New _______

Question History: Last NRC Exam __2002___

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 (10)

Comments:

LOD 4 10

11 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO Tier # __1__

Group # __1__

K/A # _295004.AA2.01___

Importance Rating _3.2__

295004 Partial or Total Loss of DC Pwr- Ability to determine and/or interpret the following as they apply to partial or complete loss of D.C. power:

AA2.01 Cause of partial or complete loss of D.C. power Question: 3 The plant is operating at power with the following conditions:

  • Breaker 1FE red indicating light is illuminated.
  • Breaker 1GE red indicating light is illuminated.
  • 4160V buses A, C, and E indicating lights are off.

What is causing the above conditions?

A. Panel BB1 has a blown fuse.

B. Panel BB3 has a blown fuse.

C. Panel AA1 has a blown fuse.

D. Panel AA3 has a blown fuse.

Answer:

C. Panel AA1 has a blown fuse.

Explanation Panel AA1 provides DC power to the 4160V buses A, C, and E indication. Breaker 1FE indication is supplied by Panel AA3 and breaker 1GE indication is supplied by Panel BB3. In order to determine which power supply has been lost the operator must know which power supply is providing the power to the breakers listed. Not all control room breaker indication is powered from the same DC power supply. So the candidate must determine the power supply to the breaker indication. With the breakers listed the candidate determines that Breaker 1FE has indicating lights, so DC Power Panel AA3 has power. Knowing that breaker 1GE has its indicating light illuminated, the candidate knows that DC Power Panel BB3 has power. Knowing that indicating lights for breakers associated with buses A, C, and E NOT being illuminated, the candidate knows that DC Power Panel AA1 has become de-energized.

Distracters:

12

A. This option is incorrect because Panel BB1 is not the power supply to any of the breakers listed in the question. This option is plausible because Panel BB1 does provide power to other 4160V breaker indications in the control room and other non-critical division components. The candidate who does not correctly recall the power supplies listed would select this option.

B. This option is incorrect because breaker 1GE has indication lights which are powered from BB3. This option is plausible because BB3 does supply other breaker indication in the control room. The candidate who does not correctly recall the power supplies listed would select this option.

D. This option is incorrect because breaker 1FE has indication lights which are powered from AA3. This option is plausible because AA3 is in the same division power supply supplies other breaker indications in the control room. The candidate who does recognize the divisional power of the breakers but does not correctly recall the power supplies listed would select this option.

Technical Reference(s): APP 2.3_9-5-2 Rev. 43 Proposed references to be provided to applicants during examination: ___NONE_______

Learning Objective: COR0020702 OPS DC ELECTRICAL DISTRIBUTION

8. Given a specific DC Electrical Distribution system malfunction, determine the effect on any of the following:
p. AC Electrical Distribution Question Source: Bank # _ _

Modified Bank # _______

New __X____

Question History: Last NRC Exam _ _

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X __

10 CFR Part 55 Content: 55.41 (7)

Comments:

LOD 4 13

14 15 16 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO Tier # __1__

Group # __1__

K/A # _295005.2.1.23___

Importance Rating _4.3__

295005 Main Turbine Generator Trip 2.1.23 Ability to perform specific system and integrated plant procedures during all modes of plant operation.

Question: 4 Reactor power is 35% during a startup.

Main Turbine bearing vibrations are as follows:

  • Bearing 5 vibration rises rapidly to 15 mils and steadies out.
  • Vibration on bearings 4 and 6 are 7 mils and rising.

What action(s) is/are required?

A. Trip the Main Turbine ONLY.

B. Scram the Reactor AND trip the Main Turbine.

C. Lower reactor power until bearing vibration lowers to <14 mils.

D. Suspend the startup to allow raising the turbine casing temperature.

Answer:

B. Scram the Reactor AND trip the Main Turbine.

Explanation:

The Main Turbine (One High Pressure and two Low Pressure), Generator and Exciter are on a single shaft and incorporate 9 bearings. Bearing 1 is on the High Pressure Turbine end of the shaft and Bearing 9 is on the Exciter end of the shaft. There are three bearings (4, 5, and 6) that have shown a rise in vibration, and one of them (5) rises above the action point for tripping the Main Turbine. These bearing are associated with the #2 Low Pressure Turbine. There is no indication given that the number 9 bearing is rising. This bearing is associated with the Exciter.

With reactor power >29.5% (164.5 psig first stage pressure), Annunciator 9-5-2/C-4 is clear.

These conditions require the operator to scram the reactor per 2.4TURB.

Since there is an unexpected turbine or generator vibration rise, there is an entry condition for procedure 2.4TURB. 2.4TURB Attachment 1 High Vibration is applicable.

17

Distracters:

A. This option is incorrect because a reactor scram is required before tripping the turbine. With the reactor power level given, the reactor is scrammed and then the turbine is tripped.

Tripping the turbine and not scramming the reactor would force a reactor scram and operators should not force an automatic RPS trip. The candidate who potentially focuses on during a startup and does not realize the reactor power level is high enough that a turbine trip would cause a reactor scram would select this option.

C. This option is incorrect because lowering reactor power is only taken if bearing 9 vibration is rising above 14 mils. Because this is the action to take for bearing 9, this option is plausible.

The candidate who recalls lowering power to bring bearing vibration down but does not recall it being only for bearing 9 would select this option. Additionally, it would most likely require a significant amount of rod insertion and time to reduce power to the necessary value. This would most likely result in bearing damage in the interim.

D. This option is incorrect because a reactor scram is required. This is an action to be taken if the rotor becomes long during startup which makes this answer plausible due to rotor vibration may be present under this condition. There is no indication turbine expansion is excessive. The candidate who recalls rotor long actions and believes it is causing high vibrations would select this option.

Technical Reference(s): 2.4TURB Main Turbine Abnormal, Rev. 30 2.3_9-5-2 (Panel 9-5-2 Annunciator Response Procedure), Rev.

43 Proposed references to be provided to applicants during examination: ___NONE_______

Learning Objective:

INT0320127, CNS Abnormal Procedures (RO) Turbine/Generator O. Given plant condition(s), determine from memory if a manual reactor scram or an emergency shutdown from power is required due to the event(s).

P. Given plant condition(s), determine from memory if a Main Turbine trip is required due to the event(s).

Question Source: Bank # _______

Modified Bank # _24663_ (See attached)

New _______

Question History: Last NRC Exam __ _____

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 (10)

Comments:

LOD: 3 18

19 20 21 22 23 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO Tier # __1__

Group # __1__

K/A # _295006.AK1.01___

Importance Rating _3.7__

295006 SCRAM- Knowledge of the operational implications of the following concepts as they apply to SCRAM:

AK1.01 Decay heat generation and removal Question: 5 The plant is operating at 100% power on the 221st day of continuous operation when all outboard MSIVs go closed.

  • At time T=0, the reactor automatically scrams.
  • At time T=3 seconds, reactor pressure spikes to 1100 psig.
  • At time T=10 minutes, reactor pressure is in its expected band.

What is the status of the SRVs at T= 10 minutes?

A. 1 SRV is cycling.

B. No SRVs are open or cycling.

C. 1 SRV is open and another SRV is cycling.

D. 2 SRVs are open and another SRV is cycling.

Answer:

A. 1 SRV is cycling.

Explanation:

For the first hour after the Reactor Scram decay generation in the reactor is exponentially decreased from approximately from 7% to 1%. Each of the 8 relief valves are designed to flow 9% (power) when at rated pressure. With the MSIVs shut, the decay heat removal is via the Nuclear Pressure Relief system to the torus. Low-Low Set (LLS) arms under these conditions and RV-71 D controls RPV pressure between 875 and 1010 psig.

Distracters:

B. This answer is incorrect because the decay heat load at 10 minutes following a reactor scram requires at least one SRV (RV-71D) to be periodically cycling to control RPV pressure between 875 psig to 1010 psig. This option is plausible because several hours to one day after a scram if containment conditions are normal then the heat loss to containment would approach that of decay heat and reactor pressure would fall with SRVs all closed. Also if the candidate does not construct an accurate mental model regarding the status of the MSIVs or 24

if the candidate does not remember the amount of decay heat generated following a reactor scram they may very well believe that after 10 minutes decay heat is less than ambient heat loss and SRV actuation is no longer required.

C. This option is incorrect because RV-71D is controlling reactor pressure between 875 psig to 1010 psig. If one LLS set valve is not enough to maintain Reactor Pressure then the second LLS valve opens at 1040 psig also cycle to between 1040 and 875. Since there is no indication given of an ATWS, a single relief valve is capable of maintaining reactor pressure.

This selection is plausible if the candidate does not recall that a single SRV is capable of maintaining RPV pressure based on an initial decay heat rate of 7%. Additionally, the volumetric flow rate across the SRV is reduced with a reduced reactor pressure. This reduction in flow may lead the candidate to an inaccurate mental model of RPV pressure control.

D. This option is incorrect because RV-71D is controlling RPV pressure between 875 psig to 1010 psig. If one LLS set valve is not enough to maintain Reactor Pressure in the required band, then the second LLS valve opens at 1040 psig and then a 3rd SRV opens at 1090 psig and cycle to maintain RPV pressure. This option is plausible during an ATWS event or if the candidate does not recall that 1 SRV is capable of maintaining reactor pressure based on the decay heat rate of 7%.

Technical Reference(s): 2.2.1 Nuclear Pressure Relief System, Rev.38 Proposed references to be provided to applicants during examination: ___NONE_______

Learning Objective: COR0021602 Nuclear Pressure Relief

8. Predict the consequences a malfunction of the following would have on the NPR system:
i. Main Steam system
12. Given plant conditions, determine if the following should occur:
c. SRV/SV opening on safety function.

Question Source: Bank # _18065______

Modified Bank # _______

New _______

Question History: Last NRC Exam __ ___

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 (8)

Comments:

LOD 2 25

26 27 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO Tier # __1__

Group # __1__

K/A # _295016 AK2.01___

Importance Rating _4.4__

295016 Control Room Abandonment:

Knowledge of the interrelations between control room abandonment and the following:

(CFR: 41.7 / 45.8)

AK2.01 Remote shutdown panel: Plant-Specific Question: 6 Following a toxic gas event requiring the control room to be abandoned, the following conditions exist:

  • BOTH Low-Low Set (LLS) valves are cycling.

The ADS ISOLATION switch in the Alternate Shutdown Room is now placed in ISOLATE.

Which LLS valve is able to be controlled from the ASD room?

Which LLS valve continues to cycle?

Controlled from LLS Valve that ASD Room Continues to Cycle A. RV-71D RV-71F B RV-71F RV-71F C. RV-71D RV-71D D. RV-71F RV-71D Answer:

D. RV-71F RV-71D Explanation:

There are two LLS valves that automatically control reactor pressure once LLS is activated (RV-71D and RV-71F). In the ASD room, the bottom section of the ADS/REC panel contains two isolation switches. One switch is for Safety Relief Valves 71E, 71F, and 71G. The isolation switch removes valve position status indication from the Control Room and prevents automatic valve operation from Low-Low Set Logic so RV-71F will no longer function in the LLS mode.

28

However, RV-71F will function by valve spring pressure and when manually actuated from the ASD room. No fire exists which would cause spurious equipment operation, therefore LLS valve 71D operation is not effected by the ASD panel switch operation and continues to operate in the LLS mode.

Distracters:

A. This option is incorrect because Low-Low set valve 71D can continue to cycle. This selection is plausible since ASD switch manipulation is an uncommon occurrence and the candidate may not internalize that the LLS logic has been removed from 71F.

B. This option is incorrect because Low-Low set valve controlled from the ASD room is not able to cycle on Low-Low set. This selection is plausible since ASD switch manipulation is an uncommon occurrence and the candidate may not remember that LLS logic has been removed from 71F with the ASD switch.

C. This option is incorrect because Low-Low set valve 71D can continue to cycle. This selection is plausible since ASD switch manipulation is an uncommon occurrence and the candidate may have the inaccurate mental model that LLS logic is removed from both 71D and 71F under these conditions.

Technical Reference(s): GE Electrical Drawing 753E253 Sheet 2.

Proposed references to be provided to applicants during examination: ___NONE_______

Learning Objective: COR002-34-02 Ops Alternate Shutdown

2. Describe the interrelationship between ASD and the following:
a. Nuclear Pressure Relief (NPR) system Question Source: Bank # _ _

Modified Bank # 21372 _ (See attached)

New __ _____

Question History: Last NRC Exam ___2006___

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 (10)

Comments:

LOD 3 29

From CNS 2006 NRC Exam 30

31 32 33 34 791E253 Sheet 35

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO Tier # __1__

Group # __1__

K/A # _295018 AK3.03___

Importance Rating _3.1__

295018 Partial or Total Loss of CCW:

Knowledge of the reasons for the following responses as they apply to partial or complete loss of component cooling water:

(CFR: 41.5 / 45.6)

AK3.03 Securing individual components (prevent equipment damage)

Question: 7 The plant is operating at rated power.

  • A loss of all REC pumps occurs.
  • 5.2REC, LOSS of REC is entered.
  • All attempts to restore REC have failed.
  • The reactor is manually scrammed.

Why is the running CRD pump required to be secured?

A. Prevent reactor vessel stratification.

B. Prevent reactor water level overfill.

C. Prevent pump bearing overheating.

D. Prevent pump motor bearing overheating.

Answer:

C. Prevent pump bearing overheating.

Explanation:

The REC system supplies cooling to the CRD pump bearing and oil cooler via the non-critical supply loop. RECs only safety function is to provide cooling to the Reactor Building Quadrant room Fan Coil Units for RHR pump operation. With the loss of REC cooling the running CRD pump must be secured to prevent pump damage due to overheating. While the quadrant fan coil units provide cooling for the RHR pumps, they are not credited for keeping the CRD pump motor cool for operation. The CRD pump injects water in the bottom head region of the reactor vessel which can be a part of the reason for stratification events which occur with low core flow and cool CRD water amassing in the bottom of the reactor vessel. With the loss of REC the reactor recirculation pumps are secured so core flow is low. Also with continued CRD injection during a scram, CRD is supplying RPV injection around 140 gpm which can eventually lead to overfilling 36

the RPV if the scram is not reset or CRD-29 valve is not closed. With the new reactor vessel level control system and setpoint setdown overfill events are not as likely, but can occur if the operator does not keep track of RPV level and take the actions described here.

From COR002-19-02 OPS Reactor Equipment Cooling

5. A sustained loss of REC to the CRD pumps would cause damage to the pumps since the bearing and oil coolers are supplied by the REC non-critical equipment supply loop.

Distracters:

A. This option is incorrect because vessel stratification is not an issue requiring removing the CRD pump from service. The CRD system injects into the bottom head region via the CRD mechanisms. If core flow is low and the CRD system is injecting, the colder water can collect in the bottom head region and cause bottom head metal temperatures to be much colder than the rest of the vessel metal temperatures. The vessel is then considered stratified.

Stopping CRD flow is a means of precluding stratification, if the core flow is low because the RR pumps are tripped. However, with normal reactor scram conditions, and reactor feedwater pumps controlling RPV level, the stratification can be prevented by closing CRD-29, charging water isolation valve. The CRD pump is tripped because of a lack of REC cooling and not because of stratification issues.

D. This option is incorrect because REC does not provide cooling to the motor bearing. REC is lost to the quad fan coil units so the quadrant temperatures rise but there are no restrictions for CRD pump operation based on the quadrant temperatures. If the candidate does not know that CRD pumps are not restricted from operating due to increased room temperature they would select this option. This option is plausible because the quadrant temperature rises and the CRD pump motor operating temperature also rises.

B. This option is incorrect because vessel overfill is not an issue requiring removing the CRD pump from service. With a scram present, CRD system flow into the RPV is approximately 140 gpm. This relatively cool water expands after it is injected into the RPV. With reactor feedwater pumps controlling RPV level, the overfill can be prevented by closing CRD-29, charging water isolation valve. With the old reactor vessel level control system it was common to overfill the RPV due to CRD injection. The new reactor vessel level control system minimizes these events but they can still happen if the operators are slow to reset the reactor scram so this option is plausible for this reason.

Technical Reference(s): 5.2REC Loss of REC, Rev. 16 Proposed references to be provided to applicants during examination: ___NONE_______

Learning Objective: COR002-19-02 OPS Reactor Equipment Cooling

6. Given a specific REC malfunction, determine the effect on any of the following:
g. CRDH system Question Source: Bank # _ _

Modified Bank # _ _

New ____X _

Question History: Last NRC Exam ___ ___

Question Cognitive Level: Memory or Fundamental Knowledge _ __

37

Comprehension or Analysis __ X _

10 CFR Part 55 Content: 55.41 (5)

Comments:

LOD 3 38

39 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO Tier # __1__

Group # __1__

K/A # _295019.AA1.02___

Importance Rating _3.3__

295019 Partial or Total Loss of Inst. Air- Ability to operate and/or monitor the following as they apply to partial or complete loss of instrument air:

AA1.02 Instrument air system valves: Plant-Specific Question: 8 The plant is operating near rated power.

  • A station operator has reported that a leak has developed in the Augmented Radwaste Building basement air system.
  • Instrument air header pressure has lowered to 75 psig and has stabilized.
  • A reactor scram has been inserted.

What action is required IAW Procedure 5.2AIR (Loss of Instrument Air)?

A. Close IA-SOV-21, Drywell Instrument Air Supply Valve.

B. Close IA-MO-80, Non Critical Instrument Air Isolation Valve.

C. Open SA-MO-81, Service Air to Instrument Air Crosstie Valve.

D. Ensure SA-AO-PCV-609, Service Air System Isolation Valve is open.

Answer:

B. Close IA-MO-80, Non Critical Instrument Air Isolation Valve.

Explanation:

A leak is present in the non-critical instrument air system. This leak is large enough to lower the instrument air header pressure and has stabilized. Since the station operator has located the source of the leak in a timely manner, closing IA-MO-80 from the control room isolates the leak and allows the instrument air header pressure to recover.

Distracters:

A. This option is incorrect because IA-SOV-21, Drywell IA Supply Valve is used to supply back up IA to the inboard MSIV in the event Nitrogen to the DW is lost. Although this valve is normally closed while operating at power, the procedure provides guidance to ensure this valve is closed which would be applicable during plant shutdown conditions and would help preserve air required to operate other plant components. The candidate would chose this answer if they believe this valve to be normally open and that by shutting the valve it would preserve pneumatics to the drywell.

40

C. This option is incorrect because SA-MO-81 is required to be opened by the control room operator when Air pressure lowers to 85 psig and it has been determined that a clogged instrument air dryer is the cause of lowering air pressure. The candidate would select this if he/she remembered it is an action that is identified for IA pressure below 85 psig (but does not remember the other requirement for confirmation of a clogged dryer).

D. This option is incorrect because SA-AO-PCV-609, Service Air System Isolation Valve automatically closes at >77 psig to isolate the Service Air Header from the Instrument Air Header. The procedure has an action to reopen this valve once it has been verified there is no leak in SA header. The candidate would chose this if he/she does not remember this valve automatically closes on low air pressure or believes the valve must be opened to recover SA pressure. Since recovery of SA is not the goal of the procedure then this action is not correct. This action, if it were able to be accomplished, would jeopardize the IA system which is why it is important for a candidate to discriminate against this option Technical Reference(s): 5.2Air Loss of Instrument Air, Rev. 19 Proposed references to be provided to applicants during examination: ___NONE_______

Learning Objective:

7 Given a specific Plant Air system malfunction, determine the effect on any of the following:

a. Plant operation Question Source: Bank # _______

Modified Bank # _3979_ (See attached)

New _______

Question History: Last NRC Exam ___ ____

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 (7)

Comments:

LOD: 2 41

42 43 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO Tier # __1__

Group # __1__

K/A # _295021 AA2.01___

Importance Rating _3.5__

295021 Loss of Shutdown Cooling:

Ability to determine and/or interpret the following as they apply to loss of shutdown cooling:

(CFR: 41.10 / 43.5 / 45.13)

AA2.01 Reactor water heatup/cooldown rate Question: 9 The plant is shutdown for refueling with the following conditions:

  • The reactor has been shutdown for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
  • RPV water level is being maintained 50 to 60 inches.
  • RHR HX inlet temperature indicates 100ºF.

What is MINIMUM time for the reactor coolant to reach 212ºF if shutdown cooling is lost?

A. 2.1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> B. 2.9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> C. 4.9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> D. 13.1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Answer:

A. 2.1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Explanation:

Determining the Time to boil is an interpretation of heatup rate (T final -T initial )/Time This requires knowledge of time after shutdown, initial reactor coolant temperature, and reactor coolant inventory available. Based upon the reactor being shutdown for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, initial Reactor Coolant Temperature at 100°F, and RPV level at the high level trip results in an estimated time to boil of 2.1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. Abnormal Procedure 2.4SDC, Shutdown Cooling Abnormal, Attachment 5 contains a family of curves based upon RPV (or cavity) water level. Using the water level at the high level trip curves and hours after shutdown (not Days after shutdown) and the given reactor coolant temperature, the answer can be determined. Interpolation on the family of curves is allowed.

Distracters:

44

B. This option is incorrect because the time to boil under the given conditions is 2.1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. This option is plausible if the Water Level at the flange graph is confused with the water level at the high level trip curve. The candidate who uses the 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> on the Water Level at the flange graph would select this option.

C. This option is incorrect because the time to boil under the given conditions is 2.1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. This option is plausible if the time since shutdown is confused with 48 days vs. hours and water level at the flange. The candidate who uses the 48 days on the Water Level at high level trip graph would select this option.

D. This option is incorrect due to time to boil under the given conditions is 2.1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. This option is plausible if the time since shutdown is confused with 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> with water level flooded to 1001. The candidate who uses the 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> on the Water Level to Level Flooded to 1001 would select this option.

Technical Reference(s): 2.4SDC, Shutdown Cooling Abnormal, Rev 14 Proposed references to be provided to applicants during examination: 2.4SDC Attachment 5 Learning Objective: INT0231002001170A Give a set of plant conditions and time of the reactor shutdown, determine: Time to core boiling.

Question Source: Bank # _ _

Modified Bank # 9181 _ (See attached)

New __ ____

Question History: Last NRC Exam ___ ___

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 (14)

Comments:

LOD 3 45

46 47 48 49 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO Tier # __1__

Group # __1__

K/A # _295023 2.2.40___

Importance Rating _3.4__

295023 Refueling Accidents:

2.2.40 Ability to apply Technical Specifications for a system.

(CFR: 41.10 / 43.2 / 43.5 / 45.3)

Question: 10 The plant is in day 12 of a refueling outage.

  • Refueling operations are in progress.

What is the required MINIMUM Technical Specifications water level above the top of the RPV flange that ensures off site dose is maintained within the allowable limits?

A. 6 feet B. 12 feet C. 21 feet D. 37 feet Answer:

C. 21 feet Explanation:

TS 3.9.6 (REFUELING OPERATIONS) - Reactor Pressure Vessel (RPV) Water Level requires RPV water level to be 21 ft above the top of the RPV flange during movement of irradiated fuel assemblies with in the RPV, during movement of new fuel assemblies or handling of control rods within the RPV, when irradiate fuel assemblies are seated within the RPV. This minimum level retains iodine fission product activity to limit offsite doses in the event of a refueling accident.

Distracters:

A. This option is incorrect because the listed water level is too low. The required level is a minimum of 21 feet above the flange. This option is plausible because it is the minimum water level above a suspended fuel bundle on the refuel bridge and the candidate may recall this number and select this option.

50

B. This option is incorrect because the listed water level is too low. The required level is a minimum of 21 feet above the flange. This option is plausible because it is the height of a fuel bundle or a control rod and the candidate may recall this number and select this option.

D. This option is incorrect because the listed water level is too high. The required level is a minimum of 21 feet above the flange. This option is plausible because it is the normal spent fuel water level and the candidate may recall this number and select this option.

Technical Reference(s): Technical Specifications 3.9.6 Proposed references to be provided to applicants during examination: ___NONE_______

Learning Objective: INT007-05-10 OPS CNS Tech Specs 3.9, Refueling Operations

1. Given a set of plant conditions, recognize non-compliance with a Section 3.9 LCO.

Question Source: Bank # _ _

Modified Bank # _ _

New ___X___

Question History: Last NRC Exam ___ ___

Question Cognitive Level: Memory or Fundamental Knowledge __X__

Comprehension or Analysis __ __

10 CFR Part 55 Content: 55.41 (10)

Comments:

LOD 2 51

52 53 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO Tier # __1__

Group # __1__

K/A # _295024EK1.01___

Importance Rating _4.1__

295024 High Drywell Pressure- Knowledge of the operational implications of the following concepts as they apply to high drywell pressure:

EK1.01 Drywell integrity: Plant-Specific Question: 11 During an ATWS, a LOCA occur s wit h the following conditions present:

  • EOP 6A and 7A have been entered.
  • Average drywell temperature is 210°F and steady.
  • Drywell pressure is 19 psig and rising.
  • Average suppression pool temperature is 205°F and rising.
  • Torus water level is 14 feet and rising slowly.
  • Reactor pressure is 1000 psig and steady.

Why is an Emergency Depressurization required at this time IAW EOP 3A (Primary Containment Control)?

To prevent A. chugging, and possible loss of the pressure suppression function.

B. raising torus pressure above the PCPL A which may result in failure of containment.

C. excessive torus to drywell vacuum breaker operation and possible vacuum breaker failure.

D. the torus water level rise which may cause loss of containment on SRV actuation due to a water column in the system discharge piping.

Answer:

B. raising torus pressure above the PCPL A which may result in failure of containment.

Explanation:

Operation is on the wrong side of the Heat Capacity Temperature limit curve for the given reactor pressure. Emergency depressurizing too far above this point has the potential to lead to exceeding the PCPL A limit which could result in loss of containment integrity.

54

Current conditions place the plant on the unsafe side of the HCTL Curve.

A. Heat Capacity Temperature Limit (HCTL) (GRAP07)

1. Definition - the highest torus water temperature from which emergency RPV depressurization will not raise suppression chamber pressure above Primary Containment Pressure Limit A, while the rate of energy transfer from the RPV to the containment is greater than the capacity of the containment vent.
2. Use - The HCTL is a function of RPV pressure and primary containment water level. It is utilized to preclude failure of the containment and equipment in the containment necessary for the safe shutdown of the plant.

Flowchart 3A requires emergency RPV depressurization when RPV pressure and torus water temperature cannot be maintained within HCTL.

B. Primary Containment Pressure Limits A/B (GRAP11)

1. Definition - the lesser of either:
a. The pressure capability of the primary containment, or
b. The maximum primary containment pressure at which vent valves sized to reject all decay heat from the containment can be opened and closed.

55

c. For Primary Containment Pressure Limit A, the maximum primary containment pressure at which SRVs can be opened and will remain open.
2. Use - Each PCPL is a function of primary containment water level and primary containment pressure. The limits are utilized to avoid challenges to primary containment vent valve operability, SRV operability, and primary containment integrity.

RPV vent valve operability is not a concern in derivation of the PCPL because RPV venting can be accomplished using the motor operated main steam line drain valves.

Operability of these valves are not affected by containment atmospheric pressure.

Distracters:

A. This option is incorrect because chugging is not an issue with emergency depressurizing.

Chugging is an issue when drywell sprays are initiated when suppression chamber pressure exceeds the Suppression Chamber Spray Initiation Pressure (SCSIP) to preclude chuggingthe cyclic condensation of steam at the downcomer openings of the drywell vents.

When a steam bubble collapses at the exit of the downcomers, the rush of water drawn into the downcomers to fill the void induces stresses at the junction of the downcomers and the vent header. Repeated application of such stresses could cause fatigue failure of these joints, thereby creating a direct path between the drywell and suppression chamber. Steam discharged through the downcomers could then bypass the suppression pool and directly pressurize the primary containment. This option is plausible because the chugging phenomenon is a real consideration in certain conditions and this is a common misconception among operators.

C. This option is incorrect because excessive torus to drywell vacuum breaker operation is not occurring so valve failure is not an issue. This option is plausible because all the energy added to the suppression pool causes suppression pool level and air space pressure to rise.

Vacuum breaker operation may occur but not excessively. If the candidate believes adding the energy to the suppression pool at the given conditions may affect vacuum breaker integrity would select this option.

D. This option is incorrect because the primary containment water level of 16 ft is defined as the SRV Tail Pipe Level Limit (SRVTPLL). It is the highest primary containment water level at which opening of an SRV will not result in exceeding the capability of the SRV tail pipe, tail pipe supports, T quencher, or T quencher supports. By maintaining primary containment water level below this limit, SRV system damage and containment failure may be precluded.

This option is plausible because had the torus level been higher than the conditions specified in the stem then this would have been the correct answer. So a candidate who does not recall the SRVTPLL may very well choose this option.

Technical Reference(s): INT0080613, OPS EOP FLOWCHART 3A - PRIMARY CONTAINMENT CONTROL, Rev 17 AMT-TB00 Appendix B (PSTG) for Step RC/P-2, Rev 8 Proposed references to be provided to applicants during examination: _HCTL Graph Learning Objective:

4. State the basis for primary containment control actions as they apply to the following.

56

a. Specific setpoints
b. Primary Containment Control Systems
c. Graphs referenced on Flowchart 3A Question Source: Bank # _21460______

Modified Bank # _______

New _______

Question History: Last NRC Exam ____ ____

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 (8)

Comments:

LOD: 4 57

58 59 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO Tier # __1__

Group # __1__

K/A # _295025.EK2.07___

Importance Rating _3.7__

295025 High Reactor Pressure- Knowledge of the interrelations between high reactor pressure and the following:

EK2.07 RCIC: Plant-Specific Question: 12 The plant is operating at power and HPCI is inoperable due to maintenance on its Auxiliary Lube Oil Pump. An inadvertent PCIS Group 1 isolation occurs. RPV pressure rises until Low-Low Set arms.

What system is capable of maintaining adequate core cooling IAW EOPs as LLS actuates?

A. SLC B. RCIC C. MC/RF D. Core Spray Answer:

B. RCIC Explanation:

The high pressure injection systems per EOP 1A, Table 3 are MC/RF, High Pressure Coolant Injection (HPCI), Reactor Core Isolation Cooling (RCIC), and the smaller Control Rod Drive hydraulic (CRD) system. The MSIVs closing cause a reactor scram and reactor pressure spike above the reactor scram setpoint of 1045 psig because the reactor energy is now fully contained within the RPV. The Low-Low Set (LLS) valves actuate and control RPV pressure between 835 psig and 1040 psig. The RFPs are steam driven and lose their motive force when the MSIVs close so they are not available for injection. The Main Condensate (MC) system utilizes booster pumps that can inject below 550 psig RPV pressure. The RCIC system is the only available injection system for the conditions given in this question.

Distracters:

A. This option is incorrect because SLC is not a listed injection system per Table 3. This selection is plausible because SLC is a viable high pressure system per Table 4 of EOP 1A.

Table 4 systems are not utilized unless Table 3 systems are unavailable or ineffective and RPV water level has dropped below the ADS timer actuation setpoint.

60

C. This option is incorrect because the RFPs are not available for injection due to MSIV closure.

The RFPs are the normal injection system at this reactor pressure so this is a plausible answer. The candidate must realize the Group 1 closes the MSIVs and this causes the RFP turbine motive force to go away after the residual steam decays from the steam supply piping. In order to utilize main condensate, reactor pressure will have to be reduced well below the setpoint of the Low-Low Set value. If this is not realized, then this option would be chosen.

D. This option is incorrect because Core Spray is a lower pressure injection system listed on Table 3. The Core Spray shutoff head value is ~350 psig so RPV pressure would have to be lowered well below the Low-Low Set value to begin injection. This is plausible because Core Spray is a listed injection system per EOP 1A, Table 3, but will not provide flow at the present RPV pressure.

Technical Reference(s): EOP 1A (RPV Control), Rev. 18 Vendor Manual 1869 COR0021802 OPS Reactor Core Isolation Cooling, Rev. 25 5.9SAMG Attachment 2 Plant Condition Assessment 1 Proposed references to be provided to applicants during examination: ___NONE_______

Learning Objective: COR0021802 OPS Reactor Core Isolation Cooling

10. Predict the consequences of the following on the RCIC system:
a. High/Low Reactor Pressure Question Source: Bank #

Modified Bank #

New X Question History: Last NRC Exam ____________

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 (7)

Comments:

LOD 3 61

62 63 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO Tier # __1__

Group # __1__

K/A # _295026EK3.04___

Importance Rating _3.6__

295026 Suppression Pool High Water Temp- Knowledge of the reasons for the following responses as they apply to suppression pool high water temperature:

EK3.04 SBLC injection Question: 13 The plant is experiencing an ATWS with the following conditions:

  • Reactor Power is 15% and steady.
  • Suppression Pool temperature is 96° F and rising.

Why is boron injection required to be initiated before average suppression pool water temperature reaches 140° F IAW EOP 6A {Reactor Power (Failure to Scram)}?

A. Prevents exceeding the 25% peak-to-peak periodic neutron flux oscillations.

B. Prevents violating Technical Specification Limit for Suppression Pool Temperature.

C. Ensures the reactor will be shutdown under all conditions before the suppression pool is heated beyond its design limits.

D. Ensures the reactor will be shutdown under hot-standby conditions before the suppression pool reaches the Heat Capacity Temperature Limit.

Answer:

D. Ensures the reactor will be shutdown under hot-standby conditions before the suppression pool reaches the Heat Capacity Temperature Limit.

Explanation:

The BIIT for 15% power is 140° F. Boron injection before this temperature ensures the Reactor will be shut down before the HCTL is exceeded.

The Boron Injection Initiation Temperature (BIIT) is the greater of:

  • The highest suppression pool temperature at which initiation of boron injection will permit injection of the Hot Shutdown Boron Weight of boron before suppression pool temperature exceeds the Heat Capacity Temperature Limit.

64

  • The suppression pool temperature at which a reactor scram is required by Technical Specifications.

The BIIT is a function of reactor power. If boron injection is initiated before suppression pool temperature reaches the BIIT, emergency RPV depressurization may be precluded at lower reactor power levels. At higher reactor power levels, however, the suppression pool heatup rate may become so high that the Hot Shutdown Boron Weight of boron cannot be injected before suppression pool temperature reaches the Heat Capacity Temperature Limit even if boron injection is initiated early in the event. Refer to Section 16 of this appendix for a detailed discussion of the BIIT.

Distracters:

A. This option is incorrect because boron is injected after large neutron oscillations are observed not before. This option is plausible because boron injection is directly tied to large neutron oscillations and boron is injected to minimize them. The candidate who does not know the boron injection is performed in response to power oscillations would select this option.

B. This option is incorrect because the Technical Specification limit for suppression pool temperature is already violated. This option is plausible because there is a Technical Specification limit on pool temperature. The candidate who recalls the Technical Specification limit and not the BIIT limit would select this option.

C. This option is incorrect because the suppression pool design temperature limit is not tied to the start of boron injection bases upon pool temperature. Getting the reactor shutdown before HCTL is reached ensures the pool can absorb the energy released to the pool during emergency depressurization. This option is plausible because shutting the reactor down will lessen the energy being transferred to the suppression pool. The candidate who cannot recall the reasons for the BIIT curve would select this option.

Technical Reference(s): AMP-TB00 (CNS PSTGs) Rev. 8 Appendix B Proposed references to be provided to applicants during examination: NONE Learning Objective: INT008-06-06. OPS EOP Flowchart 6A - RPV Pressure & Power (Failure-to-Scram)

3. Describe the conditions that require boron injection, and when boron injection can be secured.

INT008-06-18, OPS EOP and SAG Graphs and Cautions

1. Using the graphs provided in the EOP and SAG Graphs Flowchart, explain how the shape of each curve or family of curves was determined.

Question Source: Bank #

Modified Bank # 5334 New Question History: Last NRC Exam ____________

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 65

10 CFR Part 55 Content: 55.41 (5)

Comments:

LOD 3 66

67 68 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO Tier # __1__

Group # __1__

K/A # _295028EA1.01___

Importance Rating _3.8__

295028 High Drywell Temperature- Ability to operate and/or monitor the following as they apply to high drywell temperature:

EA1.01 Drywell spray: Mark-I&II Question: 14 Drywell spray has been placed in service during a LOCA due to high Drywell temperature.

  • Drywell temperature and pressure are lowering.

When is Drywell spray required to be stopped IAW EOP 3A (Primary Containment Control)?

Drywell spray is REQUIRED to be stopped before A. drywell pressure lowers to zero psig.

B. the suppression chamber to drywell vacuum breakers open.

C. the reactor building to suppression chamber vacuum breakers open.

D. drywell temperature and pressure lower to the UNSAFE (Red) region of the Drywell Spray Initiation Limit (DWISL) curve.

Answer:

A. drywell pressure lowers to zero psig.

Explanation:

EOP-3A override DS-1 directs if drywell sprays have been started, then before drywell pressure drops to 0 psig ensure drywell sprays are stopped to ensure that primary containment pressure is not reduced below atmospheric. Drywell pressure can be reduced below suppression chamber pressure causing the suppression chamber to DW vacuum breakers to open (non-condensables return to DW). DW pressure would have to be significantly negative to cause the suppression chamber pressure to lower sufficiently to cause the reactor building to suppression chamber vacuum breakers to open due to the reactor building being maintained at a negative pressure.

69

Distracters:

B. This answer is incorrect because Drywell spray is not required to be stopped before the suppression chamber to drywell vacuum breakers open. This choice is plausible due the common misconception that operation of the suppression chamber to drywell vacuum breakers is abnormal and not desired. Candidates that have the misconception of suppression chamber to drywell vacuum breaker operation would select this answer.

C. This answer is incorrect because Drywell spray is not required to be stopped before the reactor building to suppression chamber vacuum breakers open. This choice is plausible due the operation of the reactor building to suppression chamber vacuum breakers introducing air into the torus being abnormal and not desired. Candidates that have the misconception of the reactor building to suppression chamber vacuum breaker operation and securing drywell sprays would select this answer.

D. This answer is incorrect because Drywell spray is not required to be stopped before the drywell temperature and pressure lower to the RED region of the Drywell Spray Initiation Limit (DWISL) curve. This choice is a common misconception because drywell spray cannot be initiated in the unsafe region of the curve but if spray is already in progress it may continue. Candidates that have the misconception of always being in the SAFE region of DWSIL would select this answer Technical Reference(s): EOP-3A Primary Containment Control, Rev. 15 AMP-TBD00 (EOP/PSTG Technical Basis), Rev. 08 Proposed references to be provided to applicants during examination: None Learning Objective: INT00806130010300 OPS EOP FLOWCHART 3A - PRIMARY CONTAINMENT CONTROL Explain why torus and drywell sprays must be secured before torus/drywell pressure lowers to 0 psig.

Question Source: Bank # _______

Modified Bank # ___ (See attached)

New ___X__

Question History: Last NRC Exam ____________

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 (7)

Comments:

LOD 3 70

71 72 73 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO Tier # __1__

Group # __1__

K/A # _295030EA2.02___

Importance Rating _3.8__

295030 Low Suppression Pool Water Level- Ability to determine and/or interpret the following as they apply to low suppression pool water level:

EA2.02 Suppression pool temperature Question: 15 Following a LOCA, the following conditions are present:

  • RPV water level above TAF and slowly rising
  • Torus pressure is 4.5 psig (stable).
  • Only RHR pump C is injecting at the NPSH and Vortex limit of 6500 gpm.

What is the CURRENT Torus average water temperature based upon the above conditions?

A. 140°F B. 175°F C. 185°F D. 2 1 0 °F Answer:

C. 185°F Explanation:

Requires determining Suppression Pool temperature which limits RHR flow to 6500 gpm with low SP water level. With 4.5 psig pressure and 5.5 feet of water level in the torus, there is 5.14 psig overpressure in the torus (see attached calculation). The NPSH and vortex limit for RHR is 6500 gpm. With RPV level rising and above the top of active fuel, the CRS is correct in limiting flow to the NPSH and vortex limits. When using EOP graphs, Procedure 5.8 (EMERGENCY OPERATING PROCEDURES), Step 3.11 allows interpolation of graphs if SPDS is unavailable.

SPDS is unavailable because the curves and graphs are given. Step 3.11 also states interpolation is not required even though it is preferred. The question is asking for the Current Torus average water temperature so interpolation is required in this question.

74

Distracters:

A. This option is incorrect because the listed temperature is too low. This option is plausible if torus overpressure is miscalculated for 0 psig overpressure. The candidate who incorrectly calculates torus overpressure would select this answer.

B. This option is incorrect because the listed temperature is too low. This option is plausible if torus overpressure is miscalculated and interpolated for 3.9 psig overpressure. The candidate that incorrectly calculates torus overpressure would select this answer.

D. This option is incorrect because the listed temperature is too high. This option is plausible if torus overpressure is miscalculated for 10 psig overpressure. The candidate that incorrectly calculates torus overpressure would select this answer.

Technical Reference(s): CNS PSTG AMP TB00, Section 16, Rev. 8 Proposed references to be provided to applicants during examination: _EOP Graphs 4 and 5, EOP Note 3_

Learning Objective:

INT0080613 OPS EOP FLOWCHART 3A - PRIMARY CONTAINMENT CONTROL

4. State the basis for primary containment control actions as they apply to the following.
c. Graphs referenced on Flowchart 3A Question Source: Bank # _______

Modified Bank # _______

New ____X___

Question History: Last NRC Exam _________

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 (10)

Comments:

LOD: 3 75

76 77 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO Tier # __1__

Group # __1__

K/A # _295031 2.4.2___

Importance Rating _3.8__

295031 Reactor Low Water Level:

2.4.2 Knowledge of system set points, interlocks and automatic actions associated with EOP entry conditions.

Question: 16 RPV water level lowers to the value requiring entry into EOP-1A (RPV Control) and remains steady at that level.

No other EOP entry conditions are satisfied.

What group isolation(s) automatically occur(s)?

A. Group 2 only.

B. Group 3 only.

C. Groups 3 and 6 only.

D. Groups 2, 3, and 6.

Answer:

A. Group 2 only.

Explanation:

Instrument zero RPV level at CNS is +165 inches above the top of active fuel. Normal RPV water level at CNS is +35 inches above instrument zero. EOP-1A entry condition on low RPV level is +3 inches above instrument zero. The PCIS Group 2 isolation actuates at

+3 inches RPV level. The PCIS Group 3 and 6 isolations both actuate at -42 inches wide range which is 45 inches RPV level below the Group 2 actuation point. Providing an actual RPV water level in the stem would reduce the difficulty to a point in which the question would no longer be discriminatory therefore the water level is intentionally omitted to comply with the NUREG requirement to avoid use of questions with a difficulty of 1.

Distracters:

B. This option is incorrect because a Group 3 isolation does not occur until RPV level is at

-42 inches. The candidate who does not recall this setpoint would select this option.

This option is plausible because Group 3 does occur on a low RPV level condition.

78

C. This option is incorrect because the Group 3 & 6 isolations do not occur until the -42 inches level is reached. The candidate who does not recall this setpoint would select this option. This option is plausible because the Groups 3 & 6 isolation do occur on a low RPV level.

D. This option is incorrect because the Group 3 and 6 isolations do not occur until -42 inches RPV level. Group 2 isolation does actually occur at this RPV level. The candidate who does not recall these Group isolations actuation points would select this option. This option is plausible because the group 2, 3 and 6 isolations occur on a low RPV level condition.

Technical Reference(s): 2.1.22 Recovering From A Group Isolation, Rev. 59 Proposed references to be provided to applicants during examination: ___NONE_______

Learning Objective: COR0020302 OPS CONTAINMENT

21. Given plant conditions, determine if the following should have occurred:
b. Any of the PCIS group isolations Question Source: Bank # _19694_

Modified Bank # _______

New _______

Question History: Last NRC Exam ___ ___

Question Cognitive Level: Memory or Fundamental Knowledge ____

Comprehension or Analysis __ X__

10 CFR Part 55 Content: 55.41 (7)

Comments:

LOD 2 EOP 1A Entry Conditions 79

Group 2 Isolations Group 3 Isolations 80

Group 6 Isolations 81

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO Tier # __1__

Group # __1__

K/A # _295037 EK1.05___

Importance Rating _3.4__

295037 SCRAM Condition Present and Reactor Power above APRM Downscale or Unknown:

Knowledge of the operational implications of the following concepts as they apply to scram condition present and reactor power above APRM downscale or unknown:

(CFR: 41.8 to 41.10)

EK1.05 Cold shutdown boron weight: Plant-Specific Question: 17 The plant is operating at full power when the following occurs:

  • All MSIVs close resulting in a Reactor Scram with 27 rods failing to fully insert.
  • RPV water level is intentionally lowered due to level/power conditions being met.

When is a normal reactor cooldown FIRST allowed to commence IAW EOP 6A {RPV Pressure (Failure to Scram)}?

A. When all APRMs indicate downscale.

B. When Hot Shutdown Boron Weight is injected.

C. When Cold Shutdown Boron Weight is injected.

D. When one control rod is at 48 and all other control rods are at position 02.

Answer:

C. When Cold Shutdown Boron Weight is injected.

Explanation:

When the reactor is not shutdown and lowering RPV level to lower reactor power is required, boron is being injected into the RPV. When hot shutdown boron weight (26% of the Standby Liquid Control tank level) has been injected into the RPV, the RPV level is raised via outside the shroud injection systems. RPV level is raised to bring more of the boron into fuel region to aid in keeping reactor power low. The normal RPV level band of +3 inches to +54 inches is directed at this point. Once cold shutdown boron weight is injected (60% of the Standby Liquid Control tank level) or it is determined the reactor will remain shutdown without relying on the 82

boron concentration in the RPV, reactor pressure can be lowered so the plant can be placed in a cold shutdown condition.

From training material INT0080606 OPS EOP Flowchart 6A - RPV Pressure & Power (Failure-to-Scram):

FS/Q Injection normally continues until the entire usable contents of the SLC tank have been injected. Actions in other EOP steps, however, are conditioned upon lesser amounts of boron:

a. When the Hot Shutdown Boron Weight has been injected, RPV water level may be restored above the low level scram setpoint. The second override in Flowchart 7A Step FS/L-11 becomes active.
b. When the Cold Shutdown Boron Weight has been injected, RPV cooldown may be initiated in accordance with Step FS/P-6.

FS/P RPV depressurization and cooldown may not proceed until the reactor is shutdown with no boron injected or the amount of boron injected into the RPV is sufficient to keep the reactor shut down.

A volume of boron solution equivalent to 60% of the SLC tank (or 2258 lbs of boric acid and 2321 lbs of borax) is called the Cold Shutdown Boron Weight (CSBW). The CSBW is defined to be the amount of soluble boron which, if injected into the RPV and mixed uniformly, will maintain the reactor shutdown under all conditions.

The CSBW is determined assuming:

a. All rods are full out
b. Core is at most reactive exposure
c. No Xenon in the core
d. No voids in the core
e. Water is at most reactive temperature (68F)
f. Shutdown cooling and RWCU are in service
g. RPV water level at 54 in.

NOTE 1 of EOP 6A provides a method of calculating the SLC tank level resulting from injection of the CSBW. Weights of borax and boric acid are provided if boron injection must be performed by preparing the borate solution at the RWCU precoat tank.

If any amount of boron less than the CSBW has been injected into the RPV, cooldown is not permitted unless it can be determined that control rod insertion alone assures the reactor will remain shutdown under all conditions. The core reactivity response from cooldown in a partially borated core is unpredictable and subsequent steps may not prescribe the correct actions for such conditions if criticality were to occur.

The reactor is shutdown under all conditions without boron if all control rods are inserted to or beyond position 02, the shutdown margin is met (theoretically strongest control rod withdrawn and all other control rods fully inserted), or engineering determination.

83

Distracters:

A. This option is incorrect because APRM downscales are not used to determine when the cooldown can commence. If the candidate does not understand that cold shutdown boron weight (CSBW) must be injected prior to cooling down the reactor, he/she would select this option. This option is plausible because APRM downscale indication is used to determine if EOPs are entered.

B. This option is incorrect because HSBW only allows water level to be restored to +3 inches to

+54 inches to promote boron mixing in the core. If the candidate does not understand that cold shutdown boron weight (CSBW) must be injected prior to cooling down the reactor, he/she would select this option. This option is plausible because HSBW does allow a major parameter change to be made during ATWS conditions.

D. This option is incorrect because with 1 control rod not at 02 and all other control rods at 02, the reactor cannot be considered shut down under all conditions without boron. The RE would be required to evaluate this and inform the control room. If the candidate thinks he/she can make this call without RE evaluation they would select this option. This option is plausible because 1 control rod at position 48 and all other control rods at position 00 is the definition of shutdown under all conditions without boron. The reactor can be considered shutdown at this point and the cooldown could commence.

Technical Reference(s): EOP-6A, Rev 16/EOP-7A, Rev 17-Failure to Scram EOPs Proposed references to be provided to applicants during examination: ___NONE_______

Learning Objective: INT0080606 OPS EOP Flowchart 6A - RPV Pressure & Power (Failure-to-Scram)

12. Given an EOP flowchart 6A, RPV PRESSURE/POWER step, state the reason for the actions contained in the step.

Question Source: Bank # _ _

Modified Bank # _ _

New ___X___

Question History: Last NRC Exam ___ ___

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 (10)

Comments:

LOD 3 84

85 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO Tier # __1__

Group # __1__

K/A # _295038 EK2.02___

Importance Rating _3.6__

295038 High Off-site Release Rate:

Knowledge of the interrelations between high off-site release rate and the following:

(CFR: 41.7 / 45.8)

EK2.02 Offgas system Question: 18 The plant is operating at rated power when the following annunciator is received:

MAIN STM LINE PANEL/WINDOW:

HIGH RAD 9-4-1/A-5 1 minute later the following annunciator is received:

OFFGAS PANEL/WINDOW:

HIGH RAD 9-4-1/C-5 What automatic action minimizes off-site release rates if the ERP, MSL and OG radiation monitors continue rising?

What procedure(s) is/are required to be entered due to these alarms?

A. MSIV closure 5.2FUEL (Fuel Failure) ONLY.

B. MSIV closure 2.4OG (Off-Gas Abnormal) and 5.2FUEL (Fuel Failure).

C. Offgas isolation 5.2FUEL (Fuel Failure) ONLY.

D. Offgas isolation 2.4OG (Off-Gas Abnormal) and 5.2FUEL (Fuel Failure).

Answer:

D. Offgas isolation 2.4OG (Off-Gas Abnormal) and 5.2FUEL (Fuel Failure).

86

Explanation:

The interrelationship between high off-site release rate and the Offgas system is that the Offgas system automatically isolates to minimize/terminate off-site releases.

During fuel failure events, MSL radiation level rise along with RB area radiation levels followed by SJAE radiation. MSL Radiation Hi and OG Radiation Hi annunciators are received long before the ERP rad alarms. As radiation levels continue to rise, with no operator action, the OG system will isolate 15 minutes after reaching the Hi-Hi radiation value thus terminating the release from the main condenser. If MSL rad Hi-Hi is received, the reactor is required to be Scrammed and if shutdown, the MSIVs and MSL drains are required to be closed. Per OFFGAS High Rad annunciator, procedures 2.4OG & 5.2Fuel are required to be entered. Per MSL Rad High annunciator, Procedure 5.2FUEL is required to be entered if raised radiation levels are not due to hydrogen injection. The Control Room Operator is expected to recognize entry conditions from memory in Abnormal and Emergency procedures.

Distracters:

A. This answer is incorrect because MSIVs do not auto close due to MSL Rad Hi or Hi Hi and 2.4OG is not the ONLY procedure required to be entered. This answer is plausible because the original design of MSL Hi Hi was to isolate the MSIVs. If the stem were changed to reflect Manual action - MSIV closure would be correct and MSL Rad Hi only requires entry into 5.2FUEL. The candidate that confuses Auto vs. Manual action or the original design of MSL Rad Hi Hi and does not remember the entry conditions to AOP 5.2FUEL and 2.4OG would select this answer.

B. This answer is incorrect because MSIVs do not auto close due to MSL Rad Hi or Hi Hi.

This answer is plausible because the original design of MSL Hi Hi was to isolate the MSIVs.

If the stem were changed to reflect Manual action - MSIV closure would be correct. The candidate that confuses Auto vs. Manual action or the original design of MSL Rad Hi Hi and correctly identifies the entry conditions to AOP 5.2FUEL and 2.4OG would select this answer.

C. This answer is incorrect because 2.4OG is not the ONLY procedure required to be entered.

This answer is plausible because and MSL Rad Hi only requires entry into 5.2FUEL. The candidate that correctly identifies Offgas isolation and does not remember the entry conditions to AOP 5.2FUEL and 2.4OG would select this answer.

Technical Reference(s): SOP 2.2.62 Off Gas System, Rev. 27 Proposed references to be provided to applicants during examination: ___NONE_______

Learning Objective: COR001-161-1 OPS Off Gas, Rev 27

11. Given an Off Gas system component manipulation, predict and explain the change in the following parameters:
b. Radioactive release rate 87

Question Source: Bank # _ _

Modified Bank # _ _

New ___X___

Question History: Last NRC Exam ___ ___

Question Cognitive Level: Memory or Fundamental Knowledge __ __

Comprehension or Analysis __ X__

10 CFR Part 55 Content: 55.41 (7)

Comments:

LOD 2 88

89 90 91 92 93 94 Per 2.0.3 R88 Conduct Of Operations the following recognition requirements apply:

Per 2.0.1.2 R44 Operations Procedure Policy the following apply:

95

\

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO Tier # __1__

Group # __1__

K/A # _600000AK3.04___

Importance Rating _2.8__

600000 Plant Fire On Site Knowledge of the reasons for the following responses as they apply to plant fire on site:

AK3.04 Actions contained in the abnormal procedure for plant fire on site Question: 19 The plant is operating at 100% when the Shift Manager directs a Control Room evacuation due to a fire in the control room.

Why are all the AC powered Reactor Feedwater Pump lube oil pump control switches placed in PULL-TO-LOCK IAW Procedure 5.4FIRE-S/D (Fire Induced Shutdown from Outside Control Room)?

To ensure A. a reactor water overfill event is prevented.

B. reactor water level is intentionally lowered to aid in FW preheating.

C. automatic start of DC lube oil pumps to maintain RFP bearing lubrication during pump operation.

D. automatic start of DC lube oil pumps to maintain RFP bearing lubrication during pump coast down.

Answer:

A. a reactor water overfill event is prevented.

Explanation:

Prior to evacuation of the Control Room, the RFP AC Lube oil pumps are placed in PULL-TO-LOCK to prevent an overfill event from occurring. This action results in the RFPs tripping due to low bearing oil pressure with DC Lube oil pumps auto starting to provide bearing lubrication during pump coast down. Reactor water level lowers and HPCI may be used for RPV level control from the ASD Panel.

Distracters:

1

B. This answer is incorrect because the intent of the procedure guidance is to prevent an overfill event. RPV level is intentionally lowered during ATWS events to preheat the feedwater and lower reactor power which makes this answer plausible. The candidate who creates an incorrect mental model would recall the requirement for intentionally lowering RPV level by tripping the RFPs.

C. This answer is incorrect because the intent of the procedure guidance is to prevent an overfill event. The action of placing the AC lube oil pumps to Pull-to-Lock automatically starts the DC lube oil pumps for the RFPs but these pumps are only designed to protect the RFPs during coast down and are not of sufficient capacity to support continued operation and automatic level control. Additionally, with the MSIVs closed, there will be no motive force for the RFPs to maneuver. The candidate selects this if it is believed the RFPs can continue to be controlled by RVLC in automatic with only the DC lube oil pumps. This answer is plausible because the DC oil pumps are provided to ensure proper lubrication.

D. This answer is incorrect because the intent of the procedure guidance is to prevent an RPV overfill event by securing the control oil and allowing the RFPTs to trip. Stopping the AC lube oil pumps will auto start the DC lube oil pumps to support RFP bearing lubrication during coast down following RFP trip, but this is not the reason that the AC lube oil pumps are secured IAW 5.4FIRE. This selection is plausible and the candidate may select it knowing how the RFLO system operates and not the reason for performance IAW 5.4FIRE.

Technical Reference(s): 5.4 FIRE-SD FIRE INDUCED SHUTDOWN FROM OUTSIDE CONTROL ROOM, Rev. 62.

Proposed references to be provided to applicants during examination: ___NONE_______

Learning Objective: INT032-01-34 OPS CNS Abnormal Procedures (RO) Fire H. Given plant condition(s) and the applicable Abnormal/Emergency Procedure, determine the correct subsequent actions required to mitigate the event(s).

Question Source: Bank # _ _

Modified Bank # _ _

New ___X___

Question History: Last NRC Exam ___ ___

Question Cognitive Level: Memory or Fundamental Knowledge __X__

Comprehension or Analysis __ __

10 CFR Part 55 Content: 55.41 (10)

Comments:

LOD 3 2

3 4

5 6

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO Tier # __1__

Group # __1__

K/A # _700000 AA1.01___

Importance Rating _3.6__

700000 Generator Voltage and Electric Grid Disturbances:

Ability to operate and/or monitor the following as they apply to generator voltage and electric grid disturbances:

AA1.01 Grid frequency and voltage Question: 20 The plant is operating at 70% power when Doniphan Control Center (DCC) System Operator notifies the Control Room that a grid disturbance near Cooper is occurring.

The following conditions exist:

  • NPPD System frequency is 59.8 hertz and steady.

What is causing the above indications?

What is the required position of the GEN VOLTAGE REGULATOR switch under these conditions IAW Procedure 5.3GRID (Degraded Grid Voltage)?

A. The Grid OFF B. The Grid ON C. CNS OFF D. CNS ON Answer:

C. CNS OFF 7

Explanation:

Actions in this procedure to place the voltage regulator to OFF are mandated under conditions indicative of a CNS voltage regulator oscillation causing 345 kV voltage variations. If the grid voltage is oscillating and voltage regulator is working as designed, as grid voltage lowers regulator will raise field amps in an attempt to maintain terminal voltage. If regulator is causing the oscillations, field amps rising will cause 345 kV volts to rise. So, if 345 kV voltage rises as field amps lower, regulator is working properly and there is no need to transfer to the base adjuster. If 345 kV voltage rises as our field amps rise, then the CNS voltage regulator is driving the voltage oscillations and transferring to the base adjuster should stabilize conditions.

Distracters:

A. This answer is incorrect because the CNS voltage regulator is causing the problem. This answer is plausible because the voltage regulator switch position is correct. The candidate who does not realize rising voltage regulator output and corresponding grid voltage rising is caused by the voltage regulator would select this answer.

B. This answer is incorrect because the CNS voltage regulator is causing the problem and 5.3GRID directs placing the voltage regulator switch in OFF. If the candidate does not understand that the current condition on the grid is occurring due to the CNS main generator voltage regulator failing high in automatic they may chose this answer.

D. This answer is incorrect because the CNS voltage regulator is causing the problem so its control switch must be placed in OFF. The candidate that does not recall the correct procedure guidance from 5.3GRID would select this answer.

Technical Reference(s): 5.3GRID Degraded Grid Voltage, Rev. 41 Proposed references to be provided to applicants during examination: ___NONE_______

Learning Objective: INT032-01-31 CNS Abnormal Procedures (RO) Electrical S. Given plant condition(s) and the applicable Abnormal/Emergency Procedure, determine the correct subsequent actions required to mitigate the event(s).

Question Source: Bank # _ _

Modified Bank # _ _

New ___X___

Question History: Last NRC Exam ___ ___

Question Cognitive Level: Memory or Fundamental Knowledge __ __

Comprehension or Analysis __ X_

10 CFR Part 55 Content: 55.41 (10)

Comments:

LOD 3 8

9 10 11 12 13 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO Tier # __1__

Group # __1__

K/A # _295002 2.4.50___

Importance Rating _4.2__

295002 Loss of Main Condenser Vacuum:

2.4.50 Ability to verify system alarm setpoints and operate controls identified in the alarm response manual.

Question: 21 The Plant is operating at full power with the following conditions:

  • Alarm B-1/B-3 TG LOW VACUUM PRE TRIP sounds.
  • Main Generator load is 805 MWe.

(1) Where is the alarm setpoint validated?

(2) What action is directed by B-1/B-3 TG LOW VACUUM PRE TRIP, when vacuum CANNOT be maintained 23" Hg?

A. (1) CONDENSER PRESSURE TRIP GRAPH on DEH HMI.

(2) Perform Rapid Power reduction.

B. (1) CONDENSER PRESSURE TRIP GRAPH on RVLC HMI.

(2) Scram and trip the turbine.

C. (1) CONDENSER PRESSURE TRIP GRAPH on RVLC HMI.

(2) Perform Rapid Power reduction.

D. (1) CONDENSER PRESSURE TRIP GRAPH on DEH HMI.

(2) Scram and trip the turbine.

Answer:

D. (1) CONDENSER PRESSURE TRIP GRAPH on DEH HMI.

(2) Scram and trip the turbine.

Explanation:

The main condenser low vacuum alarms (Pre-trip and Trip) are received from the ALARMS program module of the Trip Tricon unit. The DEH HMI contains a dynamic display, CONDENSER PRESSURE TRIP GRAPH. The control room operator verifies the alarm by noting the current plant operating status. With the condition given, the CONDENSER PRESSURE TRIP GRAPH indicates the operating point very close to the 8.25 absolute pressure trip point. Alarm B-1/B-3 directs tripping the main turbine if condenser vacuum cannot be maintained 23" Hg.

14

Distracters:

A. This answer is incorrect because rapid power reduction is not the correct action to take.

This answer is plausible because the correct HMI is listed and lowering reactor power aids in condenser vacuum recovery. The candidate who knows the correct location to verify the alarm and knows reactor power reduction helps in mitigating the low vacuum would select this answer.

B. This answer is incorrect because the Main Turbine low vacuum pre trip alarm cannot be verified on the RVLCS HMI. This answer is plausible because the RVLCS HMI can display condenser vacuum and the RFP turbine receives a trip on loss of vacuum. The candidate who believes the vacuum pre-trip alarm can be verified on the RVLCS HMI would select this answer.

C. This answer is incorrect because rapid power reduction is not the correct action to take and the RVLCS HMI is not the correct place to verify the alarm. The abnormal procedure 2.4VAC, Loss of Condenser Vacuum provides direction to lower reactor power but does not direct rapid power reduction so this answer is plausible. The alarm procedure, B-1/B-3 does not direct lowering reactor power, but does provide direction to scram the reactor if vacuum is below 23 Hg. If the candidate cannot recall the scram action from B-1/B-3 would select this answer. This answer is plausible because rapidly lowering reactor power aids in condenser vacuum recovery.

Technical Reference(s): 2.4VAC LOSS OF CONDENSER VACUUM, Rev. 25 ALARM PROCEDURE 2.3_B-1 PANEL B - ANNUNCIATOR B-1, Rev. 34 Proposed references to be provided to applicants during examination: ___NONE_______

Learning Objective: INT032-01-32 CNS Abnormal Procedures (RO) Off Gas/Vacuum J. Given plant condition(s), determine from memory if a manual reactor scram or an emergency shutdown from power is required due to the event(s).

K. Given plant condition(s), determine from memory if a Main Turbine trip is required due to the event(s).

Question Source: Bank #

Modified Bank #

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 (10)

Comments:

LOD 3 15

16 17

\

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO Tier # __1__

Group # __1__

K/A # _295013 AK1.03___

Importance Rating _3.0__

295013 High Suppression Pool Temp.:

Knowledge of the operational implications of the following concepts as they apply to high suppression pool temperature:

(CFR: 41.8 to 41.10)

AK1.03 Localized heating Question: 22 The plant is operating at 100% power.

  • The operator takes manual control of RPV pressure using the SRVs.

Why are the SRVs alternated while controlling pressure IAW Procedure 2.2.1 (Nuclear Pressure Relief System)?

To prevent A. localized torus overheating.

B. SRV failure due to excessive cycling.

C. SRV failure due to excessive valve temperature.

D. inaccurate average torus temperature indication.

Answer:

A. localized torus overheating.

Explanation:

The SRV discharge is into the suppression pool. The energy is transported thru the SRV tailpipe to a T-Quencher located approximately 6 feet below normal Suppression Pool water level. If the same valve is re-opened repeatedly, the area in the Suppression Pool does not have a chance of mixing and a localized high temperature results.

Distracters:

B. This answer is incorrect because the procedure requires a minimum of 3 seconds between re-opening the same SRV. If the SRV was re-opened without the 3 second wait, then the SRV tailpipe could sustain damage due to the water slug being drawn into the tailpipe 18

because the vacuum breakers are not given enough time to clear the column of water. The candidate may pick this if he/she does not recall the wait period between valve re-opening.

C. This answer is incorrect because the valve is designed to withstand the temperature of the fluid flowing through it. The candidate may select this answer if he/she does not understand that using a single SRV would not result in valve overheating.

D. This answer is incorrect because alternating the location of the heat addition aids in torus mixing and allows an accurate average torus water temperature but this is not the reason for alternating the opening of the SRVs. The candidate may pick this if he/she believes the reason for alternating the opening sequence may invalidate an accurate average temperature. This answer is plausible because the energy is added to the same body of water at various locations.

Technical Reference(s): SOP 2.2.1 Nuclear Pressure Relief System, Rev. 38 Proposed references to be provided to applicants during examination: ___NONE_______

Learning Objective: COR002-16-02 OPS Nuclear Pressure Relief

4. Given a Nuclear Pressure Relief system component manipulation, predict and explain the changes in the following parameters:
e. Suppression pool temperature Question Source: Bank # _ _

Modified Bank # _ _

New ___X __

Question History: Last NRC Exam ___ ___

Question Cognitive Level: Memory or Fundamental Knowledge __X__

Comprehension or Analysis __ __

10 CFR Part 55 Content: 55.41 (10)

Comments:

LOD 2 19

20 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO Tier # __1__

Group # __1__

K/A # _295002 AK2.02___

Importance Rating _3.1__

295022 Loss of CRD Pumps:

Knowledge of the interrelations between loss of CRD pumps and the following:

AK2.02 CRD mechanism Question: 23 A plant startup is in progress with pressure set at 926 psig and reactor power at 8%.

  • CRD Pump A trips and cannot be restarted.
  • All attempts by the crew to start CRD Pump B are unsuccessful.

What is the consequence to the Control Rod Drive Mechanism (CRDM) if operation continues with these conditions?

A. Reduced seal life.

B. Unlatching of the collet fingers.

C. Cooling water orifice plugging.

D. Failure of the piston tube assembly.

Answer:

A. Reduced seal life.

Explanation:

From 2.4CRD Loss of or inadequate cooling water to the CRDMs causes the inability to move rods and elevated CRDM temperatures. The CRDMs can operate without cooling water flow but seal life may be shortened by exposure to reactor operating temperatures. CRDM temperatures over 350°F may result in a measurable delay in scram response times. A rise to 400°F could result in up to a 0.150 second rise in the 90% insertion time for an otherwise normally performing CRD. The evaluation of Scram time Tau correction is performed by Procedure 10.35 when temperature exceeds 350°F.

Distracters:

21

B This answer is incorrect because unlatching of the collet fingers is not caused by loss of CRD flow. This answer is plausible because loss of CRD flow through the mechanism would allow more weight of the drive to rest on the collet fingers. The candidate that misinterprets the CRDM internal flow paths and effects would select this answer.

C. This answer is incorrect because cooling water plugging is not an immediate issue. This answer is plausible because suddenly stopping cooling water flow could cause impurities settled in the mechanism to become disturbed. The screen at the mechanism cooling water inlet should filter out the larger debris that could cause orifice plugging which makes this answer plausible.

D. This answer is incorrect because there is no concern with the piston tube assembly.

Elevated drive temperatures result in elevated seal temperatures and the candidate may believe seal failure could result in elevated d/p across the piston tube assembly and damage to the piston tube. This answer is plausible because piston tube assembly damage can occur in CRDMs.

Technical Reference(s): 2.4CRD CRD Trouble, Rev 15 Proposed references to be provided to applicants during examination: ___NONE_______

Learning Objective: COR002-04-02 Control Rod Drive Hydraulics

12. Given a specific CRDH system malfunction, determine the effect on any of the following:
c. Control Rod Drive Mechanisms (CRDMs)

Question Source: Bank # 19962 Modified Bank #

New Question History: Last NRC Exam ___ ___

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 (7)

Comments:

LOD 3 22

23 24 25 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO Tier # __1__

Group # __1__

K/A # _295033 EK3.02___

Importance Rating _3.5__

295033 High Secondary Containment Area Radiation Levels:

Knowledge of the reasons for the following responses as they apply to high secondary containment area radiation levels:

EK3.02 Reactor SCRAM Question: 24 The plant is operating at 100% power when an un-isolable RCIC steam line failure results in high radiation in the secondary containment.

What is the reason for inserting a reactor scram before secondary containment radiation levels reach the Max Safe value?

A. To preclude flooding of the Northeast Quadrant.

B. To prevent ANY component environmental qualification (EQ) limit from being exceeded.

C. To reduce the driving head of the primary system discharging into secondary containment.

D. To ensure the reactor is shut down before radiation release rates exceed the values for a General Emergency.

Answer:

C. To reduce the driving head of the primary system discharging into secondary containment.

Explanation:

With a primary system discharging into secondary containment, the EPGs give the basis for when to scram the plant and conduct an emergency depressurization. Scramming reduces the driving head of a primary system that is discharging into the secondary containment and in anticipation of performing an emergency depressurization if radiation levels continue to rise.

Distracters:

A. This answer is incorrect because according to the 0-BARRIER series of procedures, the NE quad (S1 Stair area on attached drawing) is not an area of flooding concern from the steam tunnel or for the RCIC area. Water from this area is normally routed to the 1B Sump which is located in the NE Quad. Due to the nature of a steam leak, the fire detection system will actuate and close RW-AO-771 which redirects any water from the 26

steam tunnel drainage to an open discharge in the torus area for holding, therefore the NE Quad area is not jeopardized. This selection is plausible due to the Northeast Quads location to the MSL tunnel, its potential flooding path and the potential radiological contribution to the surrounding environment. Additionally, there are secondary containment level values associated with Max Normal and Max Safe which require SCRAM and/or ED.

B. This answer is incorrect because the scram is not based on protecting EQ equipment in secondary containment. Secondary containment has numerous EQ equipment that are designed to withstand DBA, and Special Events so the candidate may believe these actions are need to protect the EQ rating of equipment. This answer is plausible because protecting safe shutdown equipment is a prudent action to take.

D. This answer is incorrect because the scram is required to reduce the driving head of the leak into secondary containment. Additionally, significant fuel damage would be required to reach the values of a General Emergency. This answer is plausible because it is the reason for a scram if the leak were outside primary and secondary containment. In other words, if the candidate were on the Radioactive Release Control leg of EOP-5A and a radioactive release to the environment were occurring, then a scram and ED would be required. The candidate who is aware of the reason for the scram in EOP 5A, Radioactive Release Control, and not realize the question does not pertain to radioactive release would select this option.

Technical Reference(s): EOP-5A Secondary Containment Control, Rev. 15 0-BARRIER-MAPS Rev. 4 Attachment 3 B&R Drawing 2038 Sheet 1 Rev. 54 Proposed references to be provided to applicants during examination: ___NONE_______

Learning Objective: INT008-06-17 EOP Flowchart 5A Secondary Containment and Radioactivity Release Control

7. Given plant conditions and EOP flowchart 5A, SECONDARY CONTAINMENT CONTROL and RADIOACTIVITY RELEASE CONTROL, state the reasons for the actions contained in the steps.

Question Source: Bank # _ _

Modified Bank # _ _

New __X____

Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge __X__

Comprehension or Analysis __ __

10 CFR Part 55 Content: 55.41 (10)

Comments:

LOD 3 27

28 Technical Reference redacted due to SUNSI considerations

Section of B&R Drawing 2038 Sheet 1 Rev.54 30

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # __1__ _____

Group # __2__ _____

K/A # _295034 EA1.03___

Importance Rating _4.0 _ _____

295034 Secondary Containment Ventilation High Radiation / 9 - Ability to operate and / or monitor the following as they apply to secondary containment ventilation high radiation:

(CFR: 41.7) EA1.03 Secondary containment ventilation Question: 25 The following Reactor Building Ventilation parameters are observed while operating at power:

  • Exhaust Rad Channel A 11 mr/hr
  • Exhaust Rad Channel B 6 mr/hr
  • Exhaust Rad Channel C 5 mr/hr
  • Exhaust Rad Channel D 10 mr/hr
  • Reactor Building DP -0.20 wg NO plant system responses occur.

Which of the following is a required Control Room Operator action for Secondary Containment Ventilation systems?

A. Start ONE SGT train ONLY.

B. Start BOTH SGT trains.

C. Start the Standby RB supply fan.

D. Start the Standby RB exhaust fan.

Answer:

B. Start BOTH SGT trains.

Explanation:

Radiation Monitors A and C are in division 1 while B and D are in division 2. Either division will trip upon receipt of one (1) Radiation monitor upscale trip at 10 mr/hr. With a trip signal present in both divisions, a full Group 6 isolation will be initiated. The Group 6 signal isolates Reactor BLDG Ventilation which causes the initiation of both trains of Standby Gas Treatment.

Distracters:

A. This answer is incorrect because the action is to ensure both SGT trains is required.

This answer is plausible because only 1 SGT is required to maintain Secondary Containment negative pressure and there is direction to shutdown 1 SGT per Procedure 31

2.2.73 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of receiving Group 6. The candidate that recognizes a Group 6 signal is present and does not know the requirement to start BOTH SGT trains and knows 1 SGT is required to be shut down within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> would select this answer.

C. This is incorrect because starting the Standby RB Supply fan is not required with a Group 6 isolation signal present. This option is plausible because RB DP is below the procedural minimum (2.2.47) which would require starting an additional fan. The candidate that does not recognize the Group 6 isolation and confuses RB DP as being high would select this answer.

D. This is incorrect because starting the Standby RB Exhaust fan is not required with a Group 6 isolation signal present. This option is plausible because RB DP is below the procedural minimum (2.2.47) which would require starting an additional fan. The candidate that does not recognize the Group 6 isolation and recognizes RB DP as being low would select this answer.

Technical Reference(s): Procedure 2.1.22 (Recovering from a Group Isolation), Rev. 59 Procedure 2.2.47, HVAC Reactor Building, Rev. 51 Procedure 4.7.5, Reactor Building Vent Exhaust Radiation Monitoring System, Rev 18 Proposed references to be provided to applicants during examination: ___None_________

Learning Objective:

COR001-08-01

6. Describe the interrelationships between HVAC systems and the following:
e. Process Radiation Monitoring system
11. Describe the HVAC design features and interlocks that provide for the following:
b. Secondary containment isolation COR001-18-01
5. Describe the interrelationship between the RM system and the following:
r. Reactor Building Ventilation system
8. Describe the Radiation Monitoring system design feature(s) and/or interlock(s) that provide for the following:
b. Automatic action to contain the radioactive release in the event that the predetermined release rates are exceeded.

Question Source: Bank # __ _

Modified Bank # _6021__ (See attached)

New _______

Question History: Last NRC Exam ____________

Question Cognitive Level: Memory or Fundamental Knowledge __X __

Comprehension or Analysis __ __

10 CFR Part 55 Content: 55.41 _(7) _

Comments:

LOD 2 32

33 34 35 36 37 38 39 40 41 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # __1__ _____

Group # __2__ _____

K/A # _295035 EA2.02 __

Importance Rating _2.8 _ _____

295035 Secondary Containment High Differential Pressure /5 - Ability to determine and/or interpret the following as they apply to secondary containment high differential pressure: (CFR: 41.8) EA2.02 Off-site release rate: Plant-Specific Question: 26 The plant is conducting refueling operations involving the movement of recently irradiated fuel assemblies in the secondary containment.

  • A fuel bundle is dropped on the top of the core.
  • A Group 6 isolation occurs due to bundle damage.
  • Calm winds are indicated on the SPDS Weather Display.
  • HV-MO-262 MG SET-1A INLET and HV-AO-263 MG SET-1A INLET fail to fully isolate.
  • Reactor Building Average DP indicates -0.04 wg and stable.

Which of the following identifies the actual Off-Site radiological release rate response?

A. ERP release rate rises.

B. ERP release rate lowers then rises.

C. Reactor Building monitored release rate rises.

D. Reactor Building unmonitored release rate rises.

Answer:

A. ERP release rate rises.

Explanation:

A fuel handling accident involving handling of recently irradiated fuel inside of the secondary containment is one of the two principal accident scenarios for which credit is taken for secondary containment operability. Typically the Secondary Containment requires 0.25 inches of vacuum water gauge to maintain OPERABILITY. In this case however, the Reactor Building Average DP is below 0 inches water gauge (negative) and stable so both Standby Gas Trains are able to maintain negative building pressure under calm wind conditions and no RB unmonitored release is indicated. Since the Group 6 isolation occurred, the RB exhaust fans have terminated the release from the reactor building exhaust plenum. The ERP release rate rises because both SGTs are now operating and providing increased flow of airborne radiation due to the dropped fuel bundle damage and failure of the HV-MO-262 MG SET-1A INLET and 42

HV-AO-263 MG SET-1A INLET to fully isolate. This flow that was being processed through the RB exhaust plenum is now routed to the ERP via the SGT system.

Distracters:

B. This answer is incorrect because the ERP release rate will not lower but only rise. The candidate may incorrectly assume that the start of the SGT causes the ERP release rate to lower because of added dilution to the ERP KAMAN calculation is based on the increased ERP flow. However, the plant is in a refueling outage so there is no radiation exiting the ERP. Once the SGT stream containing the radioactive gases from the damaged fuel bundle reach the ERP, the release rate will rise but it will not go down (for quite some time).

C. This answer is incorrect because the RB ventilation system has isolated. This choice is plausible if the candidate thinks that the failure of the HV-MO-262 MG SET-1A INLET and HV-AO-263 MG SET-1A INLET to fully isolate contribute to a rise in Reactor Building ventilation release rate. In this case, the building is being maintained negative by the SGT system and not the RB ventilation system.

D. This answer is incorrect in this instance, because the SGT system in conjunction with the partial integrity of the Secondary Containment is adequate to maintain a negative RB pressure. This answer is plausible if the candidate does not recognize that the RB pressure is sufficiently negative to prevent an unmonitored release for the current wind conditions.

Technical Reference(s): SOP 2.2.47, HVAC Reactor Building, Rev 51 (Attach if not previously provided) SOP2.2.73, Standby Gas Treatment System, Rev 52.

Proposed references to be provided to applicants during examination: _None__________

Learning Objective:

COR001-08-01

13. Briefly describe the following concepts as they apply to HVAC:
a. Airborne contamination control COR002-28-02R22
1. State the purpose of the following items related to the Standby Gas Treatment (SGT)

System:

e. High efficiency inlet filter (HEPA)
f. Activated carbon iodine adsorber (charcoal filter)
e. High efficiency final filter
7. Given a specific Standby Gas Treatment System malfunction, determine the effect on any of the following:
a. Secondary Containment differential pressure
b. Off-Site release rate COR002-03-02R30
7. Describe the interrelationship between Secondary Containment and the following:
a. Reactor Building Ventilation
c. SGT
d. ERP
19. Predict the consequences of the following items on Secondary Containment:

43

c. High airborne radiation
25. Predict the consequences of a malfunction of the following on Secondary Containment:
a. Reactor Building Ventilation Question Source: Bank # _______

Modified Bank # _______

New ___X___

Question History: Last NRC Exam ____________

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 _(9)_

Comments:

LOD 4 44

45 46 47 48 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # _1___ _____

Group # _2___ _____

K/A # _500000 2.4.21 ___

Importance Rating _4.0 _ _____

500000 High CTMT Hydrogen Conc. / 5 - 2.4.21 Knowledge of parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactive release control, etc. (CFR: 41.7)

Question: 27 The plant has experienced a LOCA.

Which of the following identifies the MINIMUM Drywell Hydrogen concentration that requires venting and purging Primary Containment (PC)?

During this venting are ODAM Release Rates allowed to be exceeded IAW Procedure 5.8.21 {PC Venting AND Hydrogen Control (Less Than Combustible Limits)}?

PC H2 concentration at...

A. 0.34%; Exceeding ODAM limits IS NOT allowed.

B. 0.34%; Exceeding ODAM limits IS allowed.

C. 1%; Exceeding ODAM limits IS NOT allowed.

D. 1%; Exceeding ODAM limits IS allowed.

Answer:

C. 1%; Exceeding ODAM limits IS NOT allowed.

Explanation:

Requires knowledge of containment H 2 concentration which requires venting (Rad Release Control) and the impact on offsite release. EOP 3A (PCCP) requires venting & purging PC when H 2 concentration reaches 1% only if offsite radioactivity release rate is expected to remain below the offsite release rate limits specified in ODAM. PC H 2 concentration above 1% is an entry condition per EOP-3A and is required to be memorized by the CRO.

Distracters:

A. This answer is incorrect due to H 2 concentration being less than 1%. This choice is plausible if the H 2 Hi & Hi Hi alarm setpoints are confused (0.34% is 10% of the Hi Hi alarm setpoint). The candidate who confuses Hi & Hi Hi H 2 alarm setpoints and correctly recognizes release within ODAM limits would select this option.

49

B. This answer is incorrect due to H 2 concentration being less than 1% and having release above ODAM limits. This choice is plausible if the H 2 Hi & Hi Hi alarm setpoints are confused (0.34% is 10% of the Hi Hi alarm setpoint) and if venting PC is confused with the emergency release rate which requires a General Emergency (above ODAM limit). The candidate who confuses Hi & Hi Hi H 2 alarm setpoints and confuses emergency release above ODAM limits would select this option.

D. This answer is incorrect due to having releases above ODAM limits. This choice is plausible if venting PC is confused with the emergency release rate which requires a General Emergency (above ODAM limit). The candidate who confuses Hi & Hi Hi H 2 alarm setpoints and confuses emergency release above ODAM limits would select this option.

Technical Reference(s):

EOP 5.8.21 {PC Venting AND Hydrogen Control (Less Than Combustible Limits)}, Rev 18 EOP-3A (Primary Containment Control), Rev 15 5.9 H2O2 { Primary Containment Combustible Gas Control (SAG 3)}, Rev 8 Procedure 5.8.22 Proposed references to be provided to applicants during examination: __None _________

Learning Objective:

COR002-03-02

12. Describe the Containment design features and/or interlocks that provide for the following:
d. Hydrogen control Question Source: Bank # _ _

Modified Bank # _______

New ___X___

Question History: Last NRC Exam _ _

Question Cognitive Level: Memory or Fundamental Knowledge __ __

Comprehension or Analysis __ X __

10 CFR Part 55 Content: 55.41 _(7)__

Comments:

DIF 4 50

51 52 53 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # __2__ _____

Group # __1__ _____

K/A # _203000 A3.03____

Importance Rating _3.7_ _____

203000 RHR/LPCI: Injection Mode - Ability to monitor automatic operation of the RHR/LPCI: injection mode (plant specific) including: (CFR: 41.7) A3.03 Pump discharge pressure Question: 28 With the plant operating at rated power a LOCA occurs.

The following conditions exist:

  • Reactor pressure is 600 psig and lowering.
  • Drywell pressure is 5.2 psig and rising slowly.
  • Torus pressure is 4.0 psig and rising slowly.

When does LPCI injection flow into the RPV FIRST occur for the listed pressures below?

A. 550 psig B. 435 psig C. 230 psig D. 105 psig Answer:

C. 230 psig Explanation:

With drywell pressure greater than 1.84 psig a LPCI initiation signal is present. The RHR pumps receive a start signal and operate on minimum flow. As reactor pressure continues to lower the RHR inboard injection valves open when reactor pressure reaches 436 psig. Flow however does not occur until reactor pressure falls below the shutoff head of the RHR pumps at 230 psig. At or below this pressure indications of flow from the RHR would first occur. It should be noted that Plant Condition Assessment (PCA-1) chart indicates that RHR Pump shutoff head is ~300 psig, thus making 230 psig the only plausible choice.

Distracters:

A. This option is incorrect because, with reactor pressure at 550 psig, the RHR pumps are not injecting, as this pressure is greater than the shutoff head of the RHR pumps. As reactor 54

pressure lowers to this value, there would be condensate and condensate booster pump flow if they are operating. A candidate who has seen these conditions during training in the in the simulator may confuse the where the source of injection came from and choose this option.

B. This option is incorrect because, with reactor pressure at 435 psig, the RHR pumps are not injecting, as this pressure is greater than the shutoff head of the RHR pumps. The RHR Inboard injection valves however are interlocked to open at this pressure establishing a flow path from the RHR pumps to the reactor vessel. A candidate may choose this answer however believing that flow begins when the inboard injection valve opens.

D. This option is incorrect because with reactor pressure at 105 psig this is not the FIRST pressure at which the RHR pumps are injecting. This is the pressure where pressure main condensate pumps (only) inject. This value is plausible because the candidate may recall the condensate pump injection pressure and confuse this with the discharge pressure where RHR injects.

Technical Reference(s): USAR Section VI, Table VI-5-4. Plant ECCS Parameters (Attach if not previously provided) 5.9SAMG Attachment 2 Plant Condition Assessment 1 R7 (including version/revision number)

Proposed references to be provided to applicants during examination: ___None _______

Learning Objective:

COR002-23-02, OPS Residual Heat Removal System

4. Describe the interrelationship between the RHR system and the following:
n. Reactor pressure Question Source: Bank # _______

Modified Bank # _______

New ___X___

Question History: Last NRC Exam ____________

Question Cognitive Level: Memory or Fundamental Knowledge __ __

Comprehension or Analysis ___X__

10 CFR Part 55 Content: 55.41 _(7)_

Comments:

LOD 3 55

56 57 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # __2__ _____

Group # __1__ _____

K/A # _205000 A.4.07 ___

Importance Rating _3.7_ _____

205000Shutdown Cooling - Ability to manually operate and/or monitor in the control room: (CFR: 41.7) A4.07 Reactor temperatures (moderator, vessel, flange)

Question: 29 Residual Heat Removal (RHR) loop A is aligned for shutdown cooling mode of operation with RHR pump A operating. The following conditions exist:

  • The vessel head is tensioned.
  • SDC system flow is 5000 gpm.
  • Temperatures are logged as follows:

RHR-TR-131, CH 9 NBI-TR-89, CH 9 (reactor NBI-TR-89, CH6 (reactor TIME (RHR HX inlet temp) vessel metal temp) vessel flange temp) 0100 210ºF 222ºF 332ºF 0115 188ºF 196ºF 331ºF 0130 166ºF 173ºF 331ºF 0145 143ºF 147ºF 330ºF What RHR action is required and why?

A. Throttle CLOSED RHR-MO-27A, Inboard Injection Valve, to reduce the cooldown rate.

B. Throttle OPEN RHR-MO-66, RHR Heat Exchanger Bypass Valve, to reduce the cooldown rate.

C. Throttle CLOSED RHR-MO-27A, Inboard Injection Valve, to ensure accurate temperature indication at the RHR HX Inlet.

D. Throttle OPEN RHR-MO-66, RHR Heat Exchanger Bypass Valve, to ensure accurate temperature indication at the RHR HX Inlet.

Answer:

B. Throttle OPEN RHR-MO-66, RHR Heat Exchanger Bypass Valve, to reduce the cooldown rate.

58

Explanation:

When the RHR system is placed in operation for SDC the heat up and cooldown rates are adjusted to average heatup/cooldown rate 90°F/hr averaged over any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period. RHR-MO-27A, RHR-MO-66A and RHR-MO-12A are the valves that are procedurally adjusted to manipulate cooldown rate. From the data provided, if the cooldown rate is continued at the current rate, the 90°F/hr administrative limit for cooldown will be exceeded. The cooldown rate for the reactor vessel metal temperature is currently excessive and if the current rate continues the cooldown rate will exceed the limit.

Opening RHR-MO-66A will reduce the cooldown rate. Although both RHR-MO-12 and 27 could be closed to reduce cooldown the SDC flow is already low and cannot be lowered to less than 5000 gpm and closing either of these valves would reduce SDC flow.

NBI-TR-89, CH 6 is measured at the vessel flange which is in the air space above the coolant level. The RTD is on the outer flange surface so the flange temperature rate change does not directly follow the coolant rate of change. If the coolant rate of change raised by 10°F one hour, the flange rate of change would most likely not change. (NOTE: this data was gathered from actual plant parameters)

There is a NOTE in Procedure 2.2.69.2 that alerts the operator that vessel metal temperatures changing at rates different than reactor coolant rates may indicate core flow is too low and the temperatures may not be as accurate as needed and that coolant temperature may be higher than indicated. The options C and D A. This option is incorrect because even though the cooldown rate is excessive, the conditions given have SDC flow at 5000 gpm and so throttling closed either RHR-MO-27A or 12A would reduce flow rate and would therefore be inappropriate with flow at the low limit. The candidate who fails to completely evaluate all the provided conditions may choose this option because this is a procedural method for reducing the cooldown rate.

C. This option is incorrect because the cooldown rate is excessive. The temperatures are trending together correctly so the change needed is to reduce the cooldown rate.

Surveillance Procedure 6.RCS.601 requires monitoring of various temperatures during the cooldown to ensure proper trending so it a candidate may believe that the trends are not correct and that a SDC manipulation is required for accurate indication. MO-27 is a valve that can be manipulated to adjust SDC flow. However since flow is already at 5000 gpm this action is inappropriate because it would cause flow to go below 5000 gpm.

D. This option is incorrect because the cooldown rate is excessive which requires action.

Surveillance Procedure 6.RCS.601 requires monitoring of various temperatures during the cooldown and if the temperatures are not trending together then it requires the manipulation of MO-66 to ensure that the water temperature measured is accurate. Because this is a possible manipulation that may be required, a candidate may believe that these values are not trending correctly and choose this answer.

Technical Reference(s):

SP 6.RCS.601, RCS Heatup/Cooldown Rate Monitoring, Rev 21 SOP 2.2.69.2, RHR System Shutdown Operations, Rev 89 Proposed references to be provided to applicants during examination: __None_________

59

Learning Objective:

Per COR002-23-02

6. Given an RHR control manipulation, predict and explain changes in the following:
a. Heat exchanger temperature and flow
d. Reactor parameters (level, pressure, temperature)

Question Source: Bank # _______

Modified Bank # _______

New ___X___

Question History: Last NRC Exam ____________

Question Cognitive Level: Memory or Fundamental Knowledge __ __

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 _(7)_

Comments:

LOD 3 60

31A (Drywell 26A Spray) 20 61 PRESS To 'B' MAINT LOOP RHR (Drywell) (Loop Crosstie) 25A 27A 26CV (LPCI) 38A (Torus 39A 81A 274A Spray) FI (HX Outlet) 12A From Rx To Rx (Torus HX 34A LT Cooling) 16A TE A

RHR SW (A Only) 18 RR'A' 66A (HX Bypass)

(Torus) PS PS 17 13A FPC 15A (SDC PS PS Suction) RHR SW CST 13C To 'B' LOOP 15C 65A RHR (HX Inlet)

TE RHR LOOP 'A' Figure 1, Rev. 9 f:\home\jyknapp\figures\cxa05157\co022302.fig\fig1.r09 COR002-23-02 CXA05157

62 63 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # __2__ _____

Group # __1__ _____

K/A # _ 206000 2.2.39 ___

Importance Rating _ 3.9 _ _____

206000 HPCI - 2.2.39 Knowledge of less than or equal to one hour Technical Specification action statements for systems. (CFR: 41.7)

Question: 30 The plant is operating at 10% of rated power.

HPCI-MO-15, STM SUPP INBD ISOL VLV is found closed and cannot be opened with its control switch.

HPCI is declared Inoperable.

What Technical Specification ACTION is required?

A. Enter Technical Specification 3.0.3 immediately.

B. Verify all ADS SRVs are Operable within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

C. Verify RCIC system is Operable by administrative means within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

D. Isolate HPCI by deactivating HPCI-MO-15 and HPCI-MO-16 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

Answer:

C. Verify RCIC system is Operable by administrative means within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

Explanation:

Per TS 3.5.1, CONDITION C; If the HPCI System is inoperable and the RCIC System is verified to be OPERABLE, the HPCI System must be restored to OPERABLE status within 14 days. In this condition, adequate core cooling is ensured by the OPERABILITY of the redundant and diverse low-pressure ECCS injection/spray subsystems in conjunction with ADS. Also, the RCIC System will automatically provide makeup water at most reactor operating pressures.

Verification of RCIC OPERABILITY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is therefore required when HPCI is inoperable.

RCIC is required to be determined to be operable by administrative means within one hour.

Distracters:

A. This option is incorrect because entry into 3.0.3 is not required because only one ECCS system is inoperable. If two were inoperable then this option would be correct. The candidate may remember that there is an immediate entry into 3.0.3 required for an inoperable ECCS. Since it is a less than 1-hour specification associated with HPCI a candidate may choose this option.

64

B. This option is incorrect because verifying ADS operability is not required by this specification. Since functionally ADS provides a backup to the HPCI system (in conjunction with low pressure ECCS) and because RCIC is not an ECCS system, a candidate may believe that verifying that ADS is operable is the required action and would therefore choose this option. This option is a common misconception.

D. This option is incorrect because this action is not required. The actions for a primary containment isolation valve inoperable are contained in 3.6.1.3. But since only one valve is inoperable and it is closed this specification does not apply. However this would be the action required if both HPCI isolation valves had failed which is why a candidate may choose this answer.

Technical Reference(s): Technical Specification 3.5.1, ECCS-Operating ________

(Attach if not previously provided) SOP 2.2.33, High Pressure Coolant Injection System, Rev 77 Proposed references to be provided to applicants during examination: __None___________

Learning Objective:

INT007-05-06, OPS Tech Specs 3.5, Emergency Core Cooling systems (ECCS) and Reactor Core Isolation Cooling (RCIC) System

3. Given a set of plant conditions that constitutes non-compliance with a Section 3.5 LCO, determine the Actions that are required.
4. From memory, in MODES 1, 2, and 3, state the actions required in one hour if HPCI System is inoperable or two or more low pressure ECCS injection/spray subsystems inoperable or HPCI System and one or more ADS valves are inoperable (LCO 3.5.1).

Question Source: Bank # _______

Modified Bank # _______

New ___X___

Question History: Last NRC Exam ____________

Question Cognitive Level: Memory or Fundamental Knowledge _ X__

Comprehension or Analysis __ __

10 CFR Part 55 Content: 55.41 _10_

Comments:

LOD 3 65

66 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # __2__ _____

Group # __1__ _____

K/A # _ 209001 K1.01___

Importance Rating _3.1_ _____

209001 LPCS - Knowledge of the physical connection and/or cause-effect relationships between low pressure core spray and the following: (CFR: 41.2 - 41.9) K1.01 Condensate storage tank: Plant-Specific Question: 31 What is the alternate suction source for Core Spray Pump A and when is it aligned to this source?

A. ECST, to raise torus level in Modes 1/2/3.

B. CST A, to raise torus level in Modes 1/2/3.

C. ECST, for core reflood capability when the torus is drained in Modes 4/5.

D. CST A, for core reflood capability when the torus is drained in Modes 4/5.

Answer:

D. CST A, for core reflood capability when the torus is drained in Modes 4/5.

Explanation:

Per USAR Chapter VI Section 4 the CST provides an alternate suction source to CS. The suction to the CS pumps can also be lined up to Condensate Storage Tank (CST) 1A.

CNS Technical Specifications allow refueling operations to be conducted with the suppression pool drained provided an operable CS or LPCI subsystem is aligned to take a suction on CST 1A, containing at least 150,000 gallons. In this condition, the reactor vessel is depressurized and the CS subsystem provides core reflooding capability.

Distracters:

A. This option is incorrect because CS A is not capable of being aligned to the ECST. This is a plausible selection since other core cooling systems (HPCI and RCIC) may be aligned to the ECST, which is why a candidate may believe that CS is capable of this alignment.

Additionally, the purpose of the alternate source for the CS is to provide for core reflood when in modes 4 and 5. But under certain conditions, the Core Spray system can be used to fill the torus (with pressure maintenance) so a candidate may choose this option.

B. This option is incorrect because the purpose of the alternate suction source for CS A is not to provide the capability to fill the torus it is to allow reflood in modes 4 and 5 when the torus is not available. This selection is plausible since the torus can be filled with the CS system (from pressure maintenance not the alternate suction path).

67

C. This option is incorrect because the alternate suction source is from the CST not the ECST.

Other core cooling systems (HPCI and RCIC) may be aligned to the ECST, which makes this a plausible selection if the candidate believes that CS is capable of this alignment.

Technical Reference(s): SOP 2.2.9, Core Spray System, Rev 76 ____

(Attach if not previously provided) COR002-06-02, Core Spray System, Rev 87 __

(including version/revision number) __ USAR Chapter VI Section 4________ _____________

Proposed references to be provided to applicants during examination: __None _______

Learning Objective:

COR002-06-02, Core Spray System

3. Describe the interrelationships between the Core Spray and the following:
a. Condensate Storage Tank Question Source: Bank # _______

Modified Bank # _______

New __ X __

Question History: Last NRC Exam ____________

Question Cognitive Level: Memory or Fundamental Knowledge __X _

Comprehension or Analysis ____

10 CFR Part 55 Content: 55.41 8 Comments:

LOD 3 68

69 70 71 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # __2__ _____

Group # __1__ _____

K/A # _259002 K3.02 ___

Importance Rating _3.7 _ _____

259002 Reactor Water Level Control - Knowledge of the effect that a loss or malfunction of the reactor water level control system will have on following: (CFR: 41.7 / 45.4) (CFR:

41.7 / 45.5 to 45.8) - K3.02 Reactor feedwater system Question: 32 The plant is operating at near rated power when a loss of both RVLC/RFPT CORE switches causes RFPT control to transfer to MDEM.

How are the RFPs affected?

A. Speed lowers to idle speed.

B. Speed rises to upper automatic clamp.

C. Speed is held constant at current speed.

D. Speed lowers to minimum governor speed.

Answer:

C. Speed is held constant at current speed.

Explanation:

The loss of RVLC/RFPT CORE switches, which are part of the Reactor Vessel Level Control System, result in the RFPs transferring to MDEM. With the loss of the core switches, there is no control available from HMIs. Since the RFPs transfer to MDEM the RFP speed is held constant and no longer modulates to control level.

Distracters:

A. This option is incorrect because, with the failure of the core switches the controllers transfer to MDEM and now instead of RFP speed modulating it now locks at its current speed. A candidate could believe that with the loss of the switches that the output would be low as is the case with many analog controllers and that speed would therefore lower to idle speed (1000 RPM). This selection is plausible as idle speed is an operationally significant speed during a RFP start a candidate may choose this option.

B. This option is incorrect because, when the RFPs transfer to MDEM, they do so at their current speed. But a candidate may believe that with the loss of the core switches RFPT speed would rise to the upper limit of 5800 RPM. This selection is plausible since some traditional control systems that suffer loss of input indicating a lowering or loss of speed do 72

go to their maximum values so a candidate who does not understand this system may choose this option.

D. This option is incorrect because, with the failure of the core switches the controllers transfer to MDEM and now instead of RFP speed modulating it now locks at its current speed. A candidate could believe that with the loss of the switches that the output would be low as is the case with many analog controllers and that speed would therefore lower to the minimum clamp on the governor (2000 RPM). This is a different speed than the idle choice listed in option A so the candidate who believes that the speed control functions as do many analog controllers may choose this options lending plausibility for this selection as the minimum governor speed is an operationally significant value.

Technical Reference(s): Instrument Procedure 4.4.1, Reactor Vessel Level Control, Rev 7 COR002-32-02 Reactor Vessel Level Control, Rev 14 Proposed references to be provided to applicants during examination: __None__________

Learning Objective:

COR002-32-02 Reactor Vessel Level Control

9. Given a specific RVLC system malfunction, determine the effect on any of the following:
c. Feedwater System Question Source: Bank # _ _

Modified Bank # _______

New __X ___

Question History: Last NRC Exam ___ ______

Question Cognitive Level: Memory or Fundamental Knowledge __X__

Comprehension or Analysis ___

10 CFR Part 55 Content: 55.41 (7)

Comments:

LOD: 3 73

74 75 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # __2__ _____

Group # __1__ _____

K/A # 211000 K2.02 _

Importance Rating _3.1 _ _____

211000 SLC - Knowledge of electrical power supplies to the following: (CR: 41.7) K2.02 Explosive valves Question: 33 What is the power supply that is used to fire the A Standby Liquid Control (SLC) squib valve?

A. MCC K B. MCC S C. MCC M D. CPP Answer:

A. MCC K Explanation:

MCC K is the power supply to the A SLC pump and the squib valve receives its power from the pump supply breaker.

Distracters:

B. This option is incorrect as this is the power supply to the B SLC pump and the B squib valve.

A candidate may confuse which power supply is associated with which pump and may therefore choose this option. This answer is plausible because this power supply does power a SLC squib valve.

C. This option is incorrect because this is the listed power supply for SLC heat tracing. A candidate who knows SLC loads are powered from MCC M but is not certain of the squib valve power supply may choose this answer because of its association with the SLC system.

This answer is plausible because this power supply does power components in the SLC system.

D. This option is incorrect because this is the power supply for the squib valve ready lights. A candidate who does not recall that the squib valves are fired by an auxiliary contact in the pump breaker may believe this circuit also powers the squib valves since it does provide power to a squib related component (squib ready lights). Note: that even though this power supply appears different than the other options it is highly plausible because it powers actual squib components just not to power to fire the squib.

76

Technical Reference(s):

Procedure 2.2.74A (Standby Liquid Control System Component Checklist), Rev.

11 Procedure 7.2.25 {SLC System Explosive (Squib) Valve Trigger/Primer Chamber Assembly Replacement}, Rev. 19 Proposed references to be provided to applicants during examination: None Learning Objective:

COR0022902 R20

13. State the electrical power supply to the following SLC components:
b. Squib valves Question Source: Bank #

Modified Bank # 32 on 2014 NRC Exam (See attached)

New Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 41.7 Difficulty: 2 77

78 79 80 81 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # __2__ _____

Group # __1__ _____

K/A # _212000 K4.02 __

Importance Rating _3.5_ _____

212000 RPS - Knowledge of the reactor protection system design feature(s) and/or interlocks which provide for the following: (CFR: 41.7) K4.02 The prevention of a reactor SCRAM following a single component failure Question: 34 The reactor is operating at 100% rated power.

The RPS MG Set A motor supply fused disconnect fuse blows.

(1) What is the status of the RPS A Electrical Protection Assemblies (EPA)?

(2) What is the status of RPS?

A. (1) Two (2) RPS A EPA Breakers opened.

(2) Full reactor Scram.

B. (1) Two (2) RPS A EPA Breakers opened.

(2) 1/2 Scram ONLY on RPSPP1A.

C. (1) Four (4) RPS A EPA Breakers opened.

(2) Full reactor Scram.

D. (1) Four (4) RPS A EPA Breakers opened.

(2) 1/2 Scram ONLY on RPSPP1A.

Answer:

B. (1) Two (2) RPS A EPA Breakers opened.

(2) 1/2 Scram ONLY on RPSPP1A.

Explanation:

RPS MG Set A generator supplies RPSPP1A and only RPSPP1A becomes de-energized. RPS MG Set B supplies RPSPP1B. The MG sets supply power to their respective RPS power panel via a pair of EPA in series. The EPAs ensure a pure and consistent power source for the sensitive RPS instrumentation which are energized and fail safe in the de-energized state.

There is no automatic power transfer in the RPS system. Loss of power to the RPS MG Set causes two EPAs to trip on under-voltage and or under frequency which causes a half scram on the A side. Since MG set B remains energized a full scram does NOT occur.

There are four EPA breakers for each RPS power supply. TWO series breakers for the MG Set and two series breakers for the alternate power supply. Only the two breakers associated with the degraded power supply will trip allowing power to be manually transferred without operation of any EPA Breakers.

82

Distracters:

A. This option is incorrect because RPSPP1B remains energized. The candidate could choose this option if he/she did not correctly identify that only one side of RPS was impacted. This answer is plausible because there are some single components that cause a full scram vice a divisional partial scram and the number of EPAs that trip is correct. The correct combination of 2 EPA breakers opening (A1 & B1) will cause a full scram. The RPS Shorting Link Switches are in CLOSE during power operation thus inhibiting a full reactor scram from a single channel trip.

C. This option is incorrect because only the two breakers associated with the MG Set will trip because RPSPP1B remains energized. The candidate may choose this option if he/she did not identify that only two EPAs were impacted and did not correctly identify that only one side of RPS was impacted. This answer is plausible because there are some single components that cause a full scram vice a divisional partial scram. The RPS Shorting Link Switches are in CLOSE during power operation thus inhibiting a full reactor scram from a single channel trip.

D. This option is incorrect because only the two breakers associated with the MG Set will trip.

The candidate may choose this option if he/she did not identify that only two EPAs were impacted and did not correctly identify that only one side of RPS was impacted. This answer is plausible because the partial scram is correct. The correct combination of 4 EPA breakers opening (A1, 2, 3 & 4) will only cause a 1/2 scram. The RPS Shorting Link Switches are in CLOSE during power operation thus inhibiting a full reactor scram from a single channel trip.

Technical Reference(s): OPS Reactor Protection System COR002-21-02, Rev 23 (Attach if not previously provided) SOP-2.2.22, Vital Instrument Power Supply, Rev 71____

(including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: None Learning Objective:

COR002-21-02, OPS Reactor Protection System

4. Describe the RPS design features and/or interlocks that provide for the following:
b. Scram prevention following single component failure
l. Under/over voltage and frequency protection
8. Given a specific RPS malfunction, determine the effect on any of the following:
f. RPS logic channels Question Source: Bank # _ _

Modified Bank # _5210__ (See attached)

New _______

Question History: Last NRC Exam ____________

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 _(7)_

Comments:

LOD 2 83

Original Bank Question:

84

85 86 87 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # __2__ _____

Group # __1__ _____

K/A # _ 215003 K5.01 _

Importance Rating _2.6 _ _____

215003 IRM - Knowledge of the operational implications of the following concepts as they apply to intermediate range monitor (IRM) system: (CFR: 41.5) K5.01 Detector operation Question: 35 IRM A detector is installed with twice the argon fill pressure of the other IRM detectors.

How does this affect the operation of the IRM A versus the others when subjected to the same neutron field?

IRM A High High trip is A. less conservative and the downscale rod block is less conservative.

B. less conservative and the downscale rod block is more conservative.

C. more conservative and the downscale rod block is less conservative.

D. more conservative and the downscale rod block is more conservative.

Answer:

C. more conservative and the downscale rod block is less conservative.

Explanation:

The IRM detectors operate in the ionization region and their output (in a constant neutron flux) is effected primarily by the Argon pressure. The IRMs are not sensitive to small voltage changes because of their operation in the ionization region. Fission events in the detector cause ionizations as the fission fragments ionize the detector fill gas. If there is more argon in the detector then there will be more ions created by a given fission event so for IRM A each fission event causes a greater detector output. This means that the high high trip for that IRM occurs at a lower neutron flux (conservative) and because the IRM output is higher it also means that the downscale trip does not occur until a lower (non-conservative) neutron flux is reached.

A. This option is incorrect because the high high trip is more conservative because the output will reach the high high trip at a lower neutron flux. This selection is plausible because a candidate may believe that the additional argon in the detector would shield the ions from the anode (cathode), and would also believe that any reduction in detector output causes all associated trips to be less conservative.

88

B. This option is incorrect because the high high trip is more conservative because the output will reach the high high trip at a lower neutron flux. A candidate who believes that the additional argon in the detector would shield the ions from the anode (cathode) may choose this answer. This choice is plausible since it would be the correct answer if an event occurred that overall reduced the output (such as a very low detector voltage or loss of argon gas pressure).

D. This option is incorrect because the downscale rod block would be less conservative. With the additional argon gas the downscale rod block would occur at a lower neutron level than that of the other detectors so as neutron flux lowers the downscale for IRM A is delayed. A candidate may believe that, because there is more argon gas pressure, the raised detector output makes all the associated actions more conservative and would therefore choose this option.

Technical Reference(s): Lesson Plan COR002-12-02 (IRM), Rev 15 (Attach if not previously provided)

(including version/revision number) ______________________________________________

Proposed references to be provided to applicants during examination: ___None_________

Learning Objective:

COR001-10-02

5. Describe how changes in each of the following affect detector sensitivity:
b. Detector gas pressure Question Source: Bank # ________

Modified Bank # _23352_ (See attached)

New _______

Question History: Last NRC Exam _CNS 2008__

Question Cognitive Level: Memory or Fundamental Knowledge ____

Comprehension or Analysis __X___

10 CFR Part 55 Content: 55.41 _(5)__

Comments:

LOD 3 89

Original Bank question:

90

From COR002-12-02, Slide 31 91

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # _2___ _____

Group # _1___ _____

K/A # _215004 K6.01 ___

Importance Rating _3.2 _ _____

215004 Source Range Monitor - Knowledge of the effect that a loss or malfunction of the following will have on the source range monitor (SRM) system: (CFR: 41.7) K6.01 RPS Question: 36 The plant is performing a reactor startup requiring the SRM Shorting Link Switches to be placed in the OPEN position for testing.

All Shorting Link Switches have been placed in OPEN, however while placing switches for the B Channel to OPEN, the contacts remain CLOSED (switch position changes while switch contacts do not change state).

What is the impact of this RPS Logic malfunction if SRM A reaches the Hi Hi setpoint?

A. A full Reactor Scram occurs.

B. Only a half scram from "A" RPS trip system occurs.

C. Only a half scram from "B" RPS trip system occurs.

D. A full Reactor Scram occurs only if a second SRM trip signal is received.

Answer:

B. Only a half scram from "A" RPS trip system occurs.

Explanation:

The shorting link switches being OPEN allow the neutron monitoring non-coincident trips (any single neutron monitoring trip causes a FULL scram) in the A3 and B3 scram logics. If the RPS B3 shorting link contacts remaining closed, the B3 channel will not allow the SRM upscale high-high trip (SRM trip is only available in the A/B3 logic) and prevents the RPS B trip from occurring. This failure will not impact the ability of the SRM channel to input a trip signal. The candidate would choose this answer based on the shorting links being open would cause the B3 RPS Manual Scram Trip logic and A3 RPS Manual Scram Trip Logic to become a two-out-of-two taken once logic. By removing the ability of the B3 logic to trip, a 1/2 scram could not occur on the B side and a full scam due to SRM inputs would not be possible. This question is a K/A match because the RPS logic failure bypasses SRM scram inputs. If the contacts were not closed, one SRM upscale trip causes both A3 and B3 logics to trip which is a full reactor scram.

A. This option is incorrect because the RPS failure prevents any and all SRM induced full Reactor scrams (but not half scrams). The candidate could choose this distractor if he/she 1

did not know the relationship between RPS Manual Scram Trip logic and the non-coincident neutron monitoring trips. Since the RPS failure is on the B3 system, it is a common misconception that this failure would only prevent SRM B and D from initiating a full scram.

C. This option is incorrect because the failure present in RPS actually prevents an RPS B trip.

The candidate who does not understand that the trip condition of the B3 logic is energized would choose this option.

D. This option is incorrect because the failure prevents any and all SRM induced full Reactor scrams. The candidate could choose this distractor if he/she did not understand the non-coincident input of the SRMs through the shorting links or does not recall the Shorting Links switch positions. This answer is plausible because a non-coincident scram from the SRMs does occur on the upscale trip of any SRM.

Technical Reference(s): _ ________

(Attach if not previously provided) 791E256 Reactor Protection System Elementary Electrical (including version/revision number) _ __________

Proposed references to be provided to applicants during examination: ___None _______

Learning Objective:

Per COR002-30-02, Source Range Monitor

8. Predict the consequences a malfunction of the following would have on the SRM system:
a. RPS (including shorting switches)

Question Source: Bank # _ _

Modified Bank # ___23354_____

New ____ ___

Question History: Last NRC Exam ____________

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 _(7)_

Comments:

LOD 4 2

3 4

5 6

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # __2__ _____

Group # __1__ _____

K/A # __215005 A1.03___

Importance Rating _3.6__ _____

215005 APRM / LPRM - Ability to predict and/or monitor changes in parameters associated with operating the average power range monitor/local power range monitor systems controls including: (CFR: 41.5) A1.03 Control rod block status Question: 37 The reactor is operating at 15% rated power when the following occurs:

  • IRM "B" fails upscale.

Then, before any operator action is taken:

  • APRMs "B" and "E" both fail downscale.

What are the minimum action(s) that will allow all rod blocks and/or scrams to be cleared?

A. Bypass IRM "B" only.

B. Bypass APRM "E" only.

C. Bypass APRM "B" and IRM "B".

D. Bypass APRM "B" and APRM "E".

Answer:

D. Bypass APRM "B" and APRM "E".

Explanation:

Since the reactor is at 15% RTP, it is implied that the Reactor Mode Switch is in the RUN position. The upscale IRM B or IRM H and the downscale APRM B together generates an RPS trip on the "B" RPS. APRM "B" or APRM E" failed downscale cause a rod block.

To clear the RPS trip, APRM "B" OR IRM B and IRM H must be bypassed. To clear the rod block, both APRM B AND APRM E must be bypassed.

Bypassing both APRM B and APRM E will clear the rod block, and bypassing APRM B also clears the RPS trip signal.

7

Distracters:

A. This option is incorrect because this action will not clear the rod block or 1/2 scram. This answer is plausible because if the Mode switch were in STARTUP, this action would clear the 1/2 scram and rod block. The candidate who does not recognize the Mode switch being in RUN and understands IRM upscale causes a 1/2 scram & rod block in startup would choose this option.

B. This option is incorrect because this action will not clear either the rod block or the half scram. The candidate could choose this distractor if he/she did not understand that a rod block is generated by EITHER APRM B OR APRM E failed downscale and he/she did not know the 1/2 scram was generated by the upscale IRM B or IRM H and the downscale on APRM B. This answer is plausible because the APRM can be bypassed and its direct trip is also bypassed clearing the 1/2 scram.

C. This option is incorrect because this action will clear the half scram but will not clear the rod block. The candidate could choose this option if he/she did not know the 1/2 scram was generated by both the upscale IRM B and the downscale APRM B. This answer is plausible because the APRM and IRM can be bypassed and their direct trips are also bypassed.

Technical Reference(s): IOP 4.1.3, Average Power Range Monitoring System, Rev. 25 Proposed references to be provided to applicants during examination: ___None________

Learning Objective:

Per COR002-01-02, Average Power Range Monitor

5. Describe the interrelationships between the Average Power Range Monitor system and the following:
a. Reactor Protection System (RPS)
b. Intermediate Range Monitoring system (IRM) f Reactor Manual Control System (RMCS)

Question Source: Bank # _23450_

Modified Bank # _______

New _______

Question History: Last NRC Exam _CNS 2006__

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 _(5)_

Comments:

LOD 3 8

9 10 11 12 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # __2__ _____

Group # __1__ _____

K/A # __217000 A2.16___

Importance Rating _3.5__ _____

217000 RCIC - Ability to (a) predict the impacts of the following on the reactor core isolation cooling system (RCIC) ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5) A2.16 Low condensate storage tank level Question: 38 Given the following conditions:

  • RCIC is injecting to the reactor at 400 gpm.
  • RCIC suction is from the ECST.
  • Torus temperature is 125°F and steady.
  • ECST level is 25 inches and lowering.
  • Torus level is 12 9 and rising.

(1) If the conditions persist, how is RCIC impacted?

(2) What action is required IAW 9-4-1/F-2 (RCIC Suction Transfer)?

(1) (2)

RCIC Suction aligns to torus Action required A. at 23 ECST Level. Secure RCIC B. at 23 ECST Level. Makeup to the ECSTs C. at 13 1 Torus Level . Secure RCIC D. at 13 1 Torus Level . Makeup to the ECSTs Answer:

B. at 23 ECST Level. Makeup to the ECSTs Explanation:

The alternate source of water for the RCIC pump is the Suppression Pool (torus). This source of water is used if the Emergency Condensate Storage Tank levels are low. The valve (MO-41) for RCIC pump suction from the Suppression Pool will automatically open on a low level of 23" from the bottom of either Emergency Condensate Storage Tank. Per AP 9-4-1/F-2 (RCIC Suction Transfer), when the suction transfer alarm is received, action should be taken to provide makeup to the ECST per Procedure 2.2.7, Condensate Storage and Transfer System.

Distracters:

13

A. This option is incorrect because there is no need to secure the RCIC system following the swap. If the torus temperature were 145°F instead of 125°F, RCIC would need to be secured in order to prevent overheating the lube oil. The candidate who understands the RCIC suction swap but who does not remember the maximum torus temperature that supports RCIC lube oil cooling would choose this answer.

C. This option is incorrect because the RCIC suction swap does not occur on high torus level and the torus water temperature is not high enough to prohibit RCIC operation. The HPCI system does have a suction swap on high level, as well as the low ECST, so a candidate could easily confuse which system has only one parameter that swaps the suction and what that condition causes the swap. If the torus temperature were higher as well the RCIC system would need to be secured due to high oil temperature concerns.

D. This option is incorrect because the RCIC suction swap does not occur on high torus level.

The HPCI system does have a suction swap on high level, as well as the low ECST, so a candidate could easily confuse which system has only one parameter that swaps the suction and what condition causes the swap. This answer option is a plausible misconception because candidates often confuse the interlocks associated with HPCI and RCIC which are quite similar and this option would be correct if the system asked in the stem were HPCI.

Makeup to the ECSTs is correct.

Technical Reference(s):

(Attach if not previously provided) SOP 2.2.67,Reactor Core Isolation Cooling Ops, Rev. 70 (including version/revision number) ARP 9-4-1/F-2, Rev. 51___________________________

Proposed references to be provided to applicants during examination: ___None________

Learning Objective:

COR002-18-02, Reactor Core Isolation Cooling

10. Describe the interrelationship between RCIC system and the following:
h. ECSTs
8. Describe the RCIC system design features and/or interlocks that provide for the following:
a. Alternate water supplies
10. Predict the consequences of the following on the RCIC system:
c. Low ECST level
11. State the reason for the following:
b. RCIC suction transfer on Low ECST water level
12. Given plant conditions, determine if the following RCIC actions should occur:
c. ECST suction transfer Question Source: Bank # _ _

Modified Bank # _ __

New ___X___

Question History: Last NRC Exam ____________

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 _(5)_

14

Comments:

LOD 3 15

16 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # _ 2__ _____

Group # _ 1__ _____

K/A # _ 218000 A3.02 ___

Importance Rating _ 3.6_ _____

218000 ADS - Ability to monitor automatic operations of the automatic depressurization system including: (CFR: 41.7) A3.02 ADS valve tail pipe temperatures Question: 39 With reactor pressure at 900 psig an automatic initiation of ADS occurs.

What provides positive indication that each ADS valve is open?

A. The red light for each ADS valve is illuminated on Panel 9-3.

B. The green light for each ADS valves is extinguished on panel 9-3.

C. PC-TR-24 and 25 (SUPPR COOL TEMP RECORDER), Point 4 reading 90°F.

D. Temperature readings 285ºF to 300ºF for each ADS valve on temperature recorder MS-TR-166 (SAFETY AND RELIEF VALVE LEAKAGE TEMPS).

Answer:

D. Temperature readings 285ºF to 300ºF for each ADS valve on temperature recorder MS-TR-166 (SAFETY AND RELIEF VALVE LEAKAGE TEMPS).

Explanation:

When an SRV lifts, the tailpipe temperature and pressure increase to approximately 285-300ºF and 30 psig in a constant enthalpy process that superheats the SRV exhaust steam. The temperature recorder provides a positive indication that the ADS SRVs are open. The red indicating light associated with each SRV illuminates when the solenoid for the associated valve is energized meaning the circuitry is supposed to reposition the pilot valve allowing the SRV to open. The red light does not indicate SRV position however as the SRV may fail to open even with the solenoid energized. The green light for each SRV extinguish on an ADS signal, but do not indicate SRV position.

The annunciators are tied to the ADS logic being satisfied. They do not indicate SRV position.

The SRV tailpipe temperatures actually indicate steam is flowing from the SRV, hence they indicate that the individual SRVs are open and passing steam.

The SPDS computer program monitors Suppression Pool Temperatures at various locations around the pool and averages the temperature readings. The SPDS indicated temperature only means that energy is being transferred to the Suppression Pool but is not an indication all six ADS valves are open.

17

A. This option is incorrect because the red lights only indicate that the solenoids are energized but do not provide positive indication that the valves are open. However the red lights associated with each ADS valve are energized every time ADS automatically actuates. A candidate would see the red lights illuminated whenever he/she sees an ADS actuation in the simulator. And while this does occur for every valve it is not positive indication of valve position.

B. This option is incorrect because the green lights extinguishing only indicate that the ADS logic is satisfied. It does not provide positive indication of SRV position. However, this occurs with every ADS actuation so a candidate who does not completely understand the system may associate that indication with positive indication of valve position and choose this option.

C. This option is incorrect because the average suppression pool temperature rising at one point is not positive indication all the ADS valves are open, only that SRVs are open. This answer is plausible because the average temperature rises when heat is added to the suppression pool. The candidate who knows the average temperature utilizes all different area temperatures in the suppression pool and doesnt fully internalize the average temperature will rise no matter how many valves are open would select this option.

Technical Reference(s): Procedure 2.4SRV (Stuck Open Relief Valve), Rev. 15 Procedure 2.2.1 (Nuclear Pressure Relief System), Rev. 38 Proposed references to be provided to applicants during examination: __None _________

Learning Objective:

Per COR002-16-02, Nuclear Pressure Relief

6. Briefly describe the following concepts as they apply to NPR:
d. Tail pipe temperature monitoring Question Source: Bank # _______

Modified Bank # _1860__ (See attached)

New _______

Question History: Last NRC Exam ____________

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 _(7) _

Comments:

LOD 3 18

19 20 21 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # __2__ _____

Group # __1__ _____

K/A # _ 223002 A4.06___

Importance Rating _ 3.6 _ _____

223002 PCIS/Nuclear Steam Supply Shutoff - Ability to manually operate and/or monitor in the control room: (CFR: 41.7) A.4.06 Confirm initiation to completion Question: 40 A manual reactor Scram is inserted due to a transient while operating at 50% power.

The following conditions occur after the Mitigating Task Scram Actions are complete:

  • RPV pressure lowered to 600 psig and is currently rising slowly.
  • RPV water level lowered to -42 inches and is currently rising slowly.
  • Drywell pressure rises to 1.2 psig and stabilizes.
  • RHR-MO-920 {(Div 1), AOG STEAM SUPPLY VLV} is open.
  • RCIC-MO-15 (INBD ISOL VLV) is open.

Which valve(s) is/are required to be CLOSED to ensure PCIS initiation is complete?

A. MSIVs B. RR-AO-741 C. RCIC-MO-15 D. RHR-MO-920 Answer:

D. RHR-MO-920 Explanation:

Requires knowledge of Mitigating Task Scram Actions and PCIS initiation signals. The Mitigating Task Scram Actions are to depress the manual scram pushbuttons and then take the Reactor Mode Switch out of RUN and place it in REFUEL. The action of taking the Reactor MODE Switch out of RUN ensures the MSIVs do NOT close due to low equalizing header pressure. The interlock is MODE Switch in RUN and equalizing header pressure 835 psig.

Reactor water level lowered below the level (+3 inches) that initiates a Group 2 isolation. When a Group 2 isolation occurs RHR-MO-920, AOG STEAM SUPPLY VLV is required to close and if the isolation does not complete the valve closure the operator is to close the valve. No 22

conditions other than those that would result in the initiation of a Group 2 isolation have occurred.

Distracters:

A. This option is incorrect because RPV level did not get low enough for the Group 1 Isolation nor did steam pressure go low enough with the mode switch in RUN to cause a Group 1 isolation. The candidate who does not recall the water level for the Group 1 isolation or who fails to analyze the effect of the mode switch position may choose this answer. If the mode switch were not in SHUTDOWN then the MSIVs would be required to be closed. This option would be correct if the mode switch were in RUN. This answer is plausible because MSIVs will close on a low RPV level.

B. This option is incorrect because RPV level did not get low enough for the Group 7 Isolation.

Water level would have to lower to -113 for this isolation to occur. A candidate may confuse the Group 7 isolation with a Group 6 isolation which would occur at the water level provided and would therefore choose this option.

C. This option is incorrect because reactor pressure is not low enough to cause a Group 5 isolation. The candidate who is unsure of the Group 5 isolation setpoint on reactor pressure may choose this option particularly if they confuse the isolation and initiation conditions as a RCIC initiation signal is present. This answer is plausible because the listed valve closes on a low RPV pressure signal.

Technical Reference(s): GOP 2.1.22, Recovery From a Group Isolation, Rev. 59 Proposed references to be provided to applicants during examination: __None___________

Learning Objective:

COR002-03-02, Containment

21. Given plant conditions, determine if the following should have occurred:
a. Any of the PCIS group isolations Question Source: Bank # _______

Modified Bank # _______

New ___X___

Question History: Last NRC Exam ____________

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 _(7)_

Comments:

LOD 3 23

24 GROUP 1 25

GROUP 2 26

27 28 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # _ 2__ _____

Group # __1 _ _____

K/A # _ 239002 2.2.12___

Importance Rating _3.7_ _____

239002 SRVs - 2.2.12 Knowledge of surveillance procedures. (CFR: 41.10 / 45.13)

Question: 41 While performing Surveillance 6.ADS.201, ADS Manual Valve Actuation (IST) during startup following a refuel outage, the BOP Operator reports that SRV RV-71 E Control Switch has been placed to OPEN.

When conditions stabilize, which of the following indications validate that SRV RV-71 E is full open IAW 6.ADS.201?

A. Main Generator output lowers.

B. Total indicated steam flow rises.

C. PMIS temperatures within MAX T.

D. Bypass valves throttle in the closed direction.

Answer:

D. Bypass valves throttle in the closed direction.

Explanation:

The acceptance criteria for 6.ADS.201 specify the valid parameters for verifying that an SRV has properly opened. The operability limit specified is a change of BPV Position 2%.

Distracters:

A. This option is incorrect because the Main Generator is off-line during SRV testing following a refuel outage. This is plausible if the candidate only thinks of total plant effect without realizing the Main Generator is off-line.

B. This option is incorrect because total steam line flow is not an approved method of verifying SRV position for surveillance purposes. This answer is plausible because the SRVs are located upstream of the main steam line flow measurement devices, therefore total indicated steam flow will actually lower. The candidate who selects this answer could have an incorrect mental model of where SRVs are located in relation to the MS Flow elements.

C. This option is incorrect because PMIS temperatures are not acceptance criteria for SRV opening during this surveillance. These are actually listed as the closing criteria for this surveillance. This answer is plausible because this is the opening criteria per surveillance procedure 6.SRV.302, Safety Valve and Relief Valve Position Indication Instrument and 29

ADS Pneumatic Supply Check. The candidate who recalls this surveillance criteria and not the surveillance criteria in 6.ADS.201 would select this option.

Technical Reference(s): 6.ADS.201 ADS Manual Valve Actuation Rev.11 Proposed references to be provided to applicants during examination: __None___________

Learning Objective:

Per COR002-16-02, Nuclear Pressure Relief

4. Given a Nuclear Pressure Relief system component manipulation, predict and explain the changes in the following parameters:
c. Reactor pressure
f. Reactor power
g. Turbine load Question Source: Bank # _______

Modified Bank # _______

New ___X___

Question History: Last NRC Exam ____________

Question Cognitive Level: Memory or Fundamental Knowledge _ __

Comprehension or Analysis _ X__

10 CFR Part 55 Content: 55.41 _(7)_

Comments:

LOD 3 30

FROM 6.ADS.201 31

32 Shaded Blocks ANNUNCIATOR STATUS LIGHT STATUS TEMPERATURE (ALM/CLR) (ON/OFF) (°F) BPV** OPERABILITY VALVE VALVE 9-3-1/C- TRM BPV POS POS LIMITS NUMBER STATUS 9-3-1/A-2 1 GREEN AMBER RED LIMIT MS-TR-166 PMIS PID* (%) (%) (%)

1 CLOSED OFF MS-RV-OPEN ON 71A 2

CLOSED OFF 1

CLOSED OFF MS-RV-OPEN ON 71B 2

CLOSED OFF 1

CLOSED OFF MS-RV-OPEN ON 71C 2

CLOSED OFF 1

CLOSED OFF BPV POS MS-RV-OPEN ON CLOSED1 =

71D CLOSED OFF 2 BPV POS 1 CLOSED2 CLOSED OFF MS-RV-OPEN ON 71E CLOSED OFF 2 BPV POS 2%

1 CLOSED OFF MS-RV-OPEN ON 71F 2

CLOSED OFF 1

CLOSED OFF MS-RV-OPEN ON 71G 2

CLOSED OFF 1

CLOSED OFF MS-RV-OPEN ON 71H 2

CLOSED OFF

  • Use PMIS Points T142 through T149 for MS-RV-71A through MS-RV-71H, respectively.
    • BPV POSITION = BPV POSITION CLOSED1 - BPV POSITION OPEN.

33

34 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # __2__ _____

Group # __1__ _____

K/A # _ 259002 K1.07 ___

Importance Rating _ 2.6 _ _____

259002 Reactor Water Level Control - Knowledge of the physical connections and/or cause-effect relationships between reactor water level control system and the following:

(CFR: 41.2) K1.07 Rod worth minimizer: Plant-Specific Question: 42 In order to determine its mode of operation, what information must the Rod Worth Minimizer (RWM) receive from the Reactor Vessel Level Control System (RVLCS)?

A. Total main steam flow and feedwater flow rates.

B. Total main steam flow and feedwater flow rate mismatch.

C. Time that total main steam flow and feedwater flow are above limits.

D. Reactor power level calculated from total main steam flow and feedwater flow rates.

Answer:

A. Total main steam flow and feedwater flow rates.

Explanation:

The RWM operates during plant startup or shutdown. Based upon steam flow and feedwater flow rates, the mode of operation of the RWM is determined. The algorithms come from the Reactor Vessel Level Control System and determine when the RWM starts or stops enforcing predetermined control rod movements. When operating in the Low Power Alarm Point (LPAP) mode, the RVLCS sends the total main steam flow signal to the RWM. The RWM algorithm determines if the total main steam signal has been above 35% for 60 seconds. If these conditions are met, then operation is above the LPAP. Operation in the Transition Zone (TZ)Error! Bookmark not defined. is operation between LPSP and LPAP. Occurs when > 20%

total Main Steam flow AND > 20% total feedwater flow (with each condition present for at least 60 seconds) AND 35% total Main Steam flow (for any amount of time) as sensed by the LPSP and LPAP algorithms in the Reactor Vessel Level Control System, respectively. LPSP is a variable used by the RWM program and the RWM mode will be OPERATING < LPSP when either total Main Steam flow is at or below 20% or total feedwater flow is at or below 20%

(this condition being determined by the LPSP algorithm in the RVLCS program) for any period of time.

Distracters:

35

B. This option is incorrect because the RWM does not use main steam flow and feedwater flow mismatch for determining its operating mode. This answer is plausible because the RVLCS utilizes steam flow/feedwater flow mismatch algorithms for hardware error signals to be initiated and the candidate may confuse the two.

C. This option is incorrect because the time measurement comes from the RWM algorithm and not the RVLCS. This option is plausible because main steam and feedwater flow rates are timed to determine operating mode. The candidate who does not know which system is measuring the time that flows are at a given level would select this option.

D. This option is incorrect because reactor power is not an input to the RWM. This option is plausible because the RWM will utilize main steam flow and feedwater flow rate to approximate reactor power since it is required to be in service when below 9.85% rated power. If a RFP controller or feedwater controller were to fail high, then this would create a false indicated reactor power signal, therefore algorithms must be utilized to adequately justify the current reactor power. The RWM utilizes raw feedwater flow and steam flow data, thus reactor power is not a direct input. The candidate who knows the Technical Specification requirements for BPWS and knows the reactor power level where BPWS constraints must be met would select this answer.

Technical Reference(s): IOP 4.2, Rod Worth Minimizer, Rev. 29 Proposed references to be provided to applicants during examination: __None___________

Learning Objective:

Per COR002-32-02, Reactor Vessel Level Control

2. Describe the interrelationship between RVLC and the following:
j. RWM Question Source: Bank #

Modified Bank #

New X Question History: Last NRC Exam ____________

Question Cognitive Level: Memory or Fundamental Knowledge __X__

Comprehension or Analysis __ __

10 CFR Part 55 Content: 55.41 _(7)_

Comments:

LOD 3 36

37 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # _2__ _____

Group # _1 __ _____

K/A # _261000 K3.02 ___

Importance Rating _3.6_ _____

261000 SGTS - Knowledge of the effect that a loss or malfunction of the standby gas treatment system will have on the following: (CFR: 41.7) K3.02 Off-site release rate Question: 43 An accident occurred that resulted in fuel failure and a breach of the reactor coolant system boundary. The A SGT train is in service to support Primary containment venting when the following alarms:

  • Annunciator K-1/A-2 SGT A HIGH MOISTURE What SGT system component is primarily affected and how is the offsite release rate affected?

A. The Charcoal Filter and release rate rises primarily due to iodine activity.

B. The Charcoal Filter and release rate rises primarily due to particulate activity.

C. The High Efficiency Inlet Filter and the release rate rises primarily due to iodine activity.

D. The High Efficiency Inlet Filter and the release rate rises primarily due to particulate activity.

Answer:

A. The charcoal filter and release rate rises primarily due to iodine activity.

Explanation:

This matches the KA due to moisture content within the SGT impacts the amount of Iodine released. The quantity of radioactive airborne contaminants is reduced as the air passes through the SGT train. The SGT has a design flow rate of 1780 cfm. The train is comprised of multiple compartments including a moisture separator and electric heating element upstream of the HEPA filters and activated carbon iodine adsorber. The heating system is designed to reduce the relative humidity of the inlet stream from 100% to 70% when the SGT system is operating. The charcoal filters are iodide-impregnated activated carbon filters capable of removing in excess of 97.5% of the methyl iodide in the air stream under entering conditions of 70% relative humidity. As relative humidity rises above 70%, the efficiency of the carbon filters decreases thus allowing more iodine components to pass. The decreased hold-up time from adsorption on the carbon filters result in elevated radiation values due to the lack of radioactive decay of the iodine. The HEPA filter is designed to remove particulate components of >0.30 microns which are impinged on the filter and is virtually unaffected by relative humidity.

At high humidity values above 70%, the charcoal filter becomes less efficient for adsorbing the iodine thus raising the Committed Dose Equivalent (CDE) for off-site release rates.

38

Distracters:

B. This option is incorrect because the primary SGT system component that is affected by high humidity is the charcoal filter. The charcoal filter needs the SGT heater to ensure that the relative humidity of the gas stream entering the charcoal filter is sufficiently low to allow the charcoal filter to function efficiency with a relatively high adsorption rate for iodine. The roughing filter and HEPA filter are the primary filtering units for particulate material. This is a plausible selection if the candidate believes that the charcoal filter is used primarily for particulates.

C. This option is incorrect because although iodine would be a concern with the high humidity the high efficiency filter is capable of performing its function of removing particulates with high humidity. The HEPA filter has no discernable effect on iodides.

This is a plausible selection if the candidate confuses the primary function of the high efficiency inlet filter and the charcoal filter.

D. This option is incorrect because the primary component affected is the charcoal filter which adsorbs the iodine components, vs. the HEPA filter which traps particulates. Very little particulate activity will pass through the entire SGT because after the HEPA filter, there is a carbon filter and then a second HEPA filter. This selection is plausible if a candidate believes that the HEPA filter efficiency is affected by the high humidity.

Technical Reference(s): USAR V, Section 3.3.4 Procedure 2.2.73, Standby Gas Treatment System, Rev. 52.

Proposed references to be provided to applicants during examination: ___None__________

Learning Objective:

COR002-28-2, Standby Gas Treatment System

7. Given a specific Standby Gas Treatment System malfunction, determine the effect on any of the following:
b. Off-site release rate Question Source: Bank #

Modified Bank #

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 _(7)_

Comments:

LOD 3 39

40 From Flanders FFI Vendor manual data:

HFATS Test Los Alamos National Laboratory developed an alternate test method in the 1980s under contract to the U.S. Department of Energy (DOE). It is often referred to as the HFATS test (High Flow Alternative Test System). It was developed specifically to test filters rated at airflows higher than 100 CFM, but it can be used for lower flows. It is only limited by the size of the system fan and the aerosol generator output. This method was later standardized in the publication of a recommended practice, IEST=-RP-CC007.1, Testing ULPA Filters, published by the Institute of Environmental Sciences and Technology. Currently, ASME AG-1 Section FC allows for testing by this method. The filter is challenged with an acceptable polydispersed oil aerosol and the penetration through the filter is measured with a Laser Particle Counter. The Particle Counter counts and sizes individual droplets in a size range from 0.1 to 3.0 micrometers in diameter. The ratio of the downstream counts to the upstream counts in each size range is the penetration. Although this value is not equal to the penetration measured by the Q-107, research performed by Los Alamos National Laboratory verified it to be very similar and the method to be an acceptable alternative to the penetration measured by Mil-Std-282 Test Method.

Humidity and Water Resistance.

HEPA filter media will tolerate high humidity (95% +/-5%) and some direct wetting, but excessive moisture, either from air borne droplets or condensation, can plug the filter and result in failure by over-pressure. Metal case filters are more suitable for moisture laden atmospheres.

Because aluminum separators can corrode in some environments and slough particles downstream of the filter, Separatorless filters are also recommended for l=moist conditions, except in high-temperature or caustic application.

41

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # __2__ _____

Group # __1__ _____

K/A # _262001 K2.01 ___

Importance Rating _ 3.3 _ _____

262001 AC Electrical Distribution - Knowledge of electrical power supplies to the following: (CFR: 41.7) K2.01 Off-site sources of power.

Question: 44 What is a power source to the Emergency Service Station Transformer?

A. Directly from the Auburn line.

B. Directly from the Cornfield Substation.

C. 161 KV substation via the T6 transformer.

D. 161 KV substation via the T7 transformer.

Answer:

C. 161 KV substation via the T6 transformer.

Explanation:

During normal station operation, the Emergency Service Station Transformer (ESST) is energized by the 69 kV transmission line from the 69 kV Bay of the 161 kV Substation through Air Break Switch 5298. The Emergency Transformer supply can be aligned to either the Cooper 161 kV System via Transformer T6 or to the OPPD 69 kV line.

Distracters:

A. This option is incorrect because the Auburn line connects with the 161 kV Substation. From the 161 kV Substation the power must go through the 69 kV Bay and the T6 transformer to connect with the ESST. The Auburn line does not connect directly with the ESST. The 161 kV switchyard has recently gone through a major design change and the candidate may not fully understand the new configuration. This answer is plausible because the Auburn line does supply the ESST, just not directly.

B. This option is incorrect because the 69 kV Cornfield Substation has been removed during the 161 kV Substation major design change. The candidate who recalls the old arrangement would select this answer. This answer is plausible because the Cornfield Substation was previously used to directly feed the ESST.

D. This option is incorrect because the T7 transformer does not feed the ESST. The T7 transformer is part of the AC distribution system and is a new addition to the station so the candidate who confuses it with the T6 transformer would select this answer.

42

Technical Reference(s): SOP 2.2.17, ESST, Rev. 64 Proposed references to be provided to applicants during examination: __None___________

Learning Objective:

COR001-01-01, OPS AC Distribution

7. State the electrical power supplies to the following:
a. Off-Site Sources of Power Question Source: Bank # _ _

Modified Bank # __ _

New ___ X ___

Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge __X__

Comprehension or Analysis __ __

10 CFR Part 55 Content: 55.41 _(7)_

Comments:

LOD 3 43

44 ES-401 Sample Written Examination Question Worksheet Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 262002 K4.01 Importance Rating 3.1 262002 Uninterruptible Power Supply (A.C. / D.C.)

Ability to monitor automatic operations of the UNINTERRUPTABLE POWER SUPPLY (A.C./D.C.) including:

K4.01Transfer from preferred to alternate source.

Question: 45 The plant is in a normal full power electrical lineup. The following alarm is received:

NO BREAK PANEL/WINDOW:

INVERTER 1A C-4/E-7 VOLT FAILURE What is the current source of power to the NBPP?

A. MCC-L via a step down transformer and the inverter cabinet static switch.

B. MCC-R via a step down transformer and the inverter cabinet static switch.

C. MCC-L via a step down transformer and then directly to the power panel.

D. MCC-R via a step down transformer and then directly to the power panel.

Answer:

B. MCC-R via a step down transformer and the inverter cabinet static switch.

Explanation:

System Operating Procedure 2.2.22, describes NBPP automatic transfer to MCC-R. See Steps 1.2.5 and 2.3.

Power to the No-Break Power Panel (NBPP) #1 is normally supplied from 250 VDC bus 1A through inverter 1A and a static switch. The inverter failure alarm indicates that the power into or out of the inverter is failed which causes the NBPP to transfer to MCC-R. MCC-R powers the NBPP through a step down (115V AC) transformer to the static switch in the inverter cabinet.

The static switch can also be operated with the NBPP PWR TRANSFER switch on Panel C (MCC or IVTR) or by pressing the ALTERNATE SOURCE SUPPLYING LOAD or INVERTER SUPPLYING LOAD button on the inverter. The NBPP power can also be transferred by placing the MANUAL BYPASS SWITCH on the inverter to ALTERNATE SOURCE TO LOAD or 45

NORMAL OPERATION per SOP 2.2.22. The static switch and manual bypass switch transfer the NBPP power supply in a make before break logic.

Distracters:

A. This answer is incorrect because MCC-L powers the PMIS-UPS inverter as an alternate supply. This answer is plausible because the PMIS-UPS is a different uninterruptable power supply at the station. The candidate who confuses the NBPP and PMIS-UPS panels would select this answer.

C. This answer is incorrect because MCC-L powers the PMIS-UPS inverter as an alternate supply. This answer is plausible because the PMIS-UPS is a different uninterruptable power supply at the station. The candidate who confuses the NBPP and PMIS-UPS panels would select this answer.

D. MCC-R automatically powers NBPP through the Static switch. To feed the NBPP directly requires a MANUAL BYPASS SWITCH to be manipulated at the inverter cabinet. This answer is plausible because powering NBPP by bypassing the static switch is a means of powering the panel. The candidate who cannot recall the different configuration arrangements in the inverter cabinet would select this answer.

Technical Reference(s): Procedure 2.3_C-4 (Panel C - Annunciator C-4), Rev. 31 Proposed references to be provided to applicants during examination: None Learning Objective: COR0010102 AC Electrical Distribution COR0010102001090G Describe the AC Electrical Distribution System design feature(s) and/or interlock(s) that provide for the following: Transfer from preferred power to alternate power supplies Question Source: Bank # 25667 Modified Bank #

New Question History: Last NRC Exam 2012 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

LOD 3 46

47 48 49 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # __2__ _____

Group # __1__ _____

K/A # _263000 K5.01 ___

Importance Rating _ 2.6_ _____

263000 DC Electrical Distribution - Knowledge of the operational implications of the following concepts as they apply to D.C. electrical distribution: (CFR: 41.5) K5.01 Hydrogen generation during battery charging.

Question: 46 The plant is operating at 100% power with the following conditions:

  • Battery charge in progress following the replacement of a cell in the Division II 250 VDC station battery.
  • A complete loss of Battery room ventilation occurs.

What is the immediate operational concern for the present conditions?

A. Hydrogen buildup in the battery room is a fire hazard.

B. Hydrogen buildup in the battery room displaces oxygen.

C. Battery room temperature rise causes cell overheating and loss of electrolyte level.

D. Battery room temperature rise leads to cell reversal conditions in the new replacement battery cell.

Answer:

A. Hydrogen buildup in the battery room is a fire hazard.

Explanation:

Battery room ventilation is required to maintain room temperature and disperse hydrogen generated from battery charging. In the case of a battery charge in progress, the hydrogen removal function is the operational concern. The hydrogen buildup is a fire/explosive hazard and concern. Although hydrogen can displace oxygen in a space, it becomes a fire hazard at much lower concentrations.

The candidate should understand that hydrogen removal is the concern and that the buildup of hydrogen is a fire/explosion hazard.

Distracters:

B. This option is incorrect because even though hydrogen buildup is a concern in the battery room, it is the fire/explosive hazard that is the concern and not displacement of oxygen as an explosive or fire hazard would exist for a significant period of time before displacement of sufficient oxygen to cause a problem could occur, if at all. Since operators do deal with confined spaces and habitable environments the operator could believe that the battery room 50

fan is there to prevent displacement of oxygen. However this is plausible because displacement of oxygen is an issue with other gases.

C. This option is incorrect because even though battery room temperature may rise the concern with the loss of ventilation is not the temperature but the hydrogen. But because there could be a room temperature rise associated with the loss of ventilation the candidate may believe that the reason is temperature and a high rate of electrolyte loss.

D. This option is incorrect because even though battery room temperature may rise the concern with the loss of ventilation is not the temperature but the hydrogen. A candidate may not know the contributory factors associated with cell reversal but may know that it is a serious battery operational concern. That candidate may also believe that high cell temperature could contribute to cell reversal and would choose this option.

Technical Reference(s): TS Basis B.3.8.1 _____________

(Attach if not previously provided) SOP 2.2.24.1, R13 250VDC Electrical System (Div. 1)__

(including version/revision number) 0.39 R50 Hot Work, ACD0150507R04-L-Batteries &

Current Converters.

Procedure 2.2.38 (HVAC Control Building), Rev. 40 Proposed references to be provided to applicants during examination: None _

Learning Objective:

Per COR002-07-02, DC Electrical Distribution

10. Briefly describe the following concepts as they apply to DC Electrical Distribution System.
a. Hydrogen generation during battery charging.

Question Source: Bank # _ _

Modified Bank # _21329_ (See attached)

New ___ ___

Question History: Last NRC Exam __CNS 2005__

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 _(5)_

Comments:

LOD 2 51

52 53 54 55 56 57 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # __2__ _____

Group # __1__ _____

K/A # _ 264000 K6.03 ___

Importance Rating _ 3.5 _ _____

264000 EDGs - Knowledge of the effect that a loss or malfunction of the following will have on the emergency generators (diesel): (CFR: 41.7) K6.03 Lube oil pumps Question: 47 DG1 is manually started for post maintenance testing:

  • DIESEL GEN 1 BKR EG1 is open.
  • Engine driven lube oil pump fails (no oil flow).

What condition FIRST trips the Diesel Generator?

A. Low Lube Oil Pressure B. High Lube Oil Temperature C. Low Turbocharger Oil Pressure D. High Thrust Bearing Oil Temperature Answer:

A. Low Lube Oil Pressure Explanation:

When the diesel generator is manually started all the diesel generator trips are in effect. With the loss of the engine driven oil pump and the loss of all lube oil pressure, the diesel generator trips at <20 psig lube oil pressure. This trip is only bypassed on an automatic start. This signal is the first that trips the DG and anything beyond that cannot trip the DG because it is already tripped.

Distractors:

B. This option is incorrect because the diesel generator does not trip on high oil temperature.

The DG will trip on high bearing or connecting rod temperature(s). With the loss of lube oil pressure, it is reasonable to assume a relationship between a loss of oil flow and elevated metal temperatures. A low oil pressure trip would occur before engine oil temperature became elevated. This selection is plausible if the candidate does not remember the relationship between the trip set points of the low lube oil pressure in relation to the relative length of time it would take the bearing metal temperature to rise.

58

C. This option is incorrect because the turbocharger oil pressure trip occurs at a lower pressure than does the engine oil pressure trip. So as oil pressure falls, the engine oil trip point would be reached first because the low Turbocharger Oil Pressure setpoint varies with load of the machine. The Turbocharger Low Oil trip setpoint is calculated based on the following formula: 4 psig + (0.49 x turbocharger discharge air pressure [in Hg]). This selection is plausible because a candidate may not recognize the relatively low turbocharger discharge air pressures or the configuration of the oil supply to the turbocharger.

D. This option is incorrect because this type of catastrophic condition would occur after the oil pressure had tripped the diesel generator. This selection is plausible because the loss of lube oil could cause the turbocharger thrust bearing temperature to elevate to the bearing failure point but typically only on the supply line to the turbocharger. Thus, a candidate could reason that this would trip the diesel generator.

Technical Reference(s): SOP 2.2.20, Standby AC Power System (Diesel (Attach if not previously provided) Generators). Rev. 92 (including version/revision number) 2.3_DG1 R21 _________________________

Proposed references to be provided to applicants during examination: __None___________

Learning Objective:

Per COR002-08-02R32, Diesel Generators

11. Predict the consequences a malfunction of the following would have on the Diesel Generators:
c. Lube Oil pumps Question Source: Bank # _______

Modified Bank # _______

New ___X___

Question History: Last NRC Exam ____________

Question Cognitive Level: Memory or Fundamental Knowledge __X __

Comprehension or Analysis __ __

10 CFR Part 55 Content: 55.41 _(7)_

Comments:

LOD 3 59

60 61 62 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # __2__ _____

Group # __1__ _____

K/A # _ 300000 A3.02 ___

Importance Rating _ 2.9 _ _____

300000 Instrument Air - Ability to monitor automatic operations of the instrument air system including: (CFR: 41.7) A3.02 Air Temperature Question: 48 The air compressors are operating with the Compressor Control Module (CCM) in LOCAL with the following lineup:

  • Air Compressor 1A is in Lead and running.
  • Air Compressor 1B is First Backup and is idle.
  • Air Compressor 1C is Second Backup and is idle.

Reactor Equipment Cooling (REC) to the air compressors is lost due to an REC pipe rupture.

What condition automatically trips Air Compressor 1A?

What automatically occurs following the trip of Air Compressor 1A?

A. High air temperature.

Air Compressor 1B starts and continuously supplies air loads.

B. High air temperature.

Air Compressor 1B starts but trips soon after it starts.

C. Low cooling water pressure.

Air Compressor 1B starts and continuously supplies air loads.

D. Low cooling water pressure.

Air Compressor 1B starts but trips soon after it starts.

Answer:

A. High air temperature.

Air Compressor 1B starts and continuously supplies air loads.

Explanation:

The normal cooling water lineup is REC to Air Compressor 1A and TEC to Air Compressors 1B and 1C. When REC is lost to Air Compressor 1A, air temperatures will rise until Air Compressor 1A trips. Lowering air pressure will start the 1st Backup compressor (which is cooled by TEC) so it continues to operate supplying system air loads. All air loads should be supplied. Also the sequence with compressors with CCM in LOCAL is lead cycles 110 to 100 psig, 1st Backup starts at 93 psig and 2nd Backup starts at 90 psig.

63

If power is lost to the REC-TEC cross-tie valves (which is not the case here), then the REC and TEC alignment to the compressors when power is restored is Air Compressor 1A and Air Compressor 1B supplied by REC and Air Compressor 1C supplied by TEC. This response makes choices B and D highly plausible. The compressor protection from loss of cooling water is high temperature trips, not low cooling water pressure.

B. This option is incorrect because Air Compressor 1B does not trip on high air temperature but continues to operate as its cooling water supply is from TEC. The candidate who knows that Air Compressor 1A trips due to high temperature but who also believes that air compressor 1B is cooled by REC may choose this answer. This is a likely misconception that candidates could have, as two of the compressors are supplied by one closed cooling water system and one is supplied by a different system.

C. This option is incorrect because Air Compressor 1A does not trip due to low cooling water pressure. This selection is plausible since many plant components that require cooling water do trip or isolate due to low cooling water pressure so a candidate may choose this option. Additionally, Air Compressor 1B does start and carry the load which adds to this selections plausibility.

D. This option is incorrect because the Air Compressor 1A does not trip due to low cooling water pressure nor will Air Compressor 1B trip as it remains supplied with cooling water from the TEC system. This selection is plausible since many plant components that require cooling water do trip or isolate due to low cooling water pressure so a candidate may choose this option. Additionally, Air Compressor 1B does start and carry the load which adds to this selections plausibility.

Technical Reference(s): SOP 2.2.59, Plant Air System. Rev. 74 (Attach if not previously provided)

(including version/revision number) ___________________

Proposed references to be provided to applicants during examination: __None___________

Learning Objective:

COR001-17-01 6 Predict the consequences the following would have on the Plant Air system:

a.REC failure b.TEC failure

10. Given plant conditions, determine if any of the following should occur:

c.Air Compressor automatic trip Question Source: Bank # _______

Modified Bank # _______

New ___X___

Question History: Last NRC Exam ____________

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 _(7)_

64

Comments:

LOD 3 65

66 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # __2__ _____

Group # __1__ _____

K/A # _400000 A2.01 ___

Importance Rating _ 3.3 _ _____

400000 Component Cooling Water - Ability to (a) predict the impacts of the following on the CCWS and (b) based on those predictions use procedures to correct, control or mitigate the consequences of those abnormal operations: (CFR: 41.5) A2.01 Loss of CCW pump.

Question: 49 The Unit is operating at 50% power with REC Pump A breaker tagged OPEN.

REC Heat Exchanger B and 3 REC pumps are in service.

  • One REC pump trips and is unavailable.
  • M-1/A-1, REC SYSTEM LOW PRESSURE alarms.

(1) How does the system respond with no operator action?

(2) What Immediate Action(s) is/are required upon receipt of this alarm with system pressure at 60 psig IAW Procedure 5.2REC (Loss of REC)?

A. (1) Non-critical loads isolate after a 20 second time delay.

(2) Isolate Augmented Radwaste cooling ONLY.

B. (1) Non-critical loads isolate after a 20 second time delay.

(2) Isolate Augmented Radwaste cooling and RWCU Non-Regen HX Inlet.

C. (1) Non-critical loads isolate after a 40 second time delay.

(2) Isolate Augmented Radwaste cooling ONLY.

D. (1) Non-critical loads isolate after a 40 second time delay.

(2) Isolate Augmented Radwaste cooling and RWCU Non-Regen HX Inlet.

Answer:

D. (1) Non-critical loads isolate after a 40 second time delay.

(2) Isolate Augmented Radwaste cooling and RWCU Non-Regen HX Inlet.

Explanation:

The REC system contains 4 pumps with 3 pumps normally in operation. The standby pump is currently unavailable to be started. The REC heat exchanger outlet piping contains pressure switches that isolate the non-critical loads and Augmented Radwaste after a 40 second time delay. Without a standby REC pump available, 5.2REC Immediate Actions require MANUALLY isolating Augmented Radwaste and the RWCU NRHX in an attempt to restore system pressure PRIOR to automatic isolation.

67

A. This answer is incorrect because non-critical loads isolate after heat exchanger outlet pressure remains below the isolation setting for 40 seconds and additionally isolating RWCU NRHX is required as part of procedure 5.2REC immediate operator actions. This answer is plausible because the non-critical loads of the REC system do isolate on a system low pressure condition with the time delay being easily confused with the Standby REC pump auto start which occurs 20 second on DG sequential load and isolating Augmented RW happens automatically after the 40 seconds but is often forgotten as a 5.2REC immediate action. The candidate who cannot recall the correct time delay on the Non-critical load valve isolation and does not remember all 5.2REC immediate operator actions would choose this answer.

B. This answer is incorrect because non-critical loads isolate after heat exchanger outlet pressure remains below the isolation setting for 40 seconds This answer is plausible because the non-critical loads of the REC system do isolate on a system low pressure condition with the time delay being easily confused with the Standby REC pump auto start which occurs 20 second on DG sequential load. The candidate who cannot recall the correct time delay on the Non-critical load valve isolation and remembers all 5.2REC immediate operator actions would choose this answer.

C. This answer is incorrect because isolating RWCU NRHX is required as part of procedure 5.2REC immediate operator actions. This answer is plausible because isolating Augmented RW happens automatically after the 40 seconds and can be forgotten as a 5.2REC immediate action. The candidate who recalls the correct time delay on the Non-critical load valve isolation and does not remember all 5.2REC immediate operator actions would choose this answer.

Technical Reference(s): Procedure 2.3_M-1, Panel M, Rev. 14.

Procedure 5.2REC (Loss of REC), Rev. 16.

Proposed references to be provided to applicants during examination: __None___________

Learning Objective:

Per COR002-19-02

6. Given a specific REC malfunction, determine the effect on any of the following:
a. REC header pressure
d. Standby REC pump operation Question Source: Bank # _ _

Modified Bank # _______

New ___X___

Question History: Last NRC Exam _ _

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 _(5)_

Comments:

LOD 3 68

69 70 71 72 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # __2__ _____

Group # __1__ _____

K/A # _ 211000 A1.07 ___

Importance Rating _ 4.3_ _____

211000 SLC - Ability to predict and/or monitor changes in parameters associated with operating the standby liquid control system controls including: (CFR: 41.5) A1.07 Reactor power Question: 50 The plant operating at 100% rated power with SLC pump 1A out of service when an ATWS occurs:

  • Average torus temperature is just below BIIT and rising slowly.
  • RPV level remains at +35 inches.

SLC pump 1B is placed to START and the following conditions are present:

  • Reactor power is 40% and slowly lowering.
  • Boron is injecting with an initial tank level of 80%.

Assuming the SLC pump is operating at its design flow rate and RPV level remains steady, what is the approximate SLC tank level and status of the Average Power Range Monitor (APRM) downscales after 25 minutes?

SLC Tank Level APRM Downscale Alarm (9-5-1/C-8)

A. 54% OFF B. 54% ON C. 26% OFF D. 26% ON Answer:

B. 54% ON Explanation:

Per COR002-29-02 The SLC pumps are triplex plunger type pumps with a design flow rate of 53 gpm. Based on this design flow rate, the minimum TS flow rate of 38.2 gpm will always be met. At the design flow rate of 53 gpm, each positive displacement pump is capable of injecting Hot Shutdown Boron Weight into the RPV within approximately 22 minutes. Normal SLC tank volume is maintained at approximately 80% and EOP 7A specifies that a 26% drop of SLC tank level will be HSD boron weight. This volume may be calculated as follows:

73

SLC tank overflow = 4565 Gallons (vertical cylindrical tank) 80% SLC tank level = 4565 x 0.8 = 3652 gallons (approximate normal volume) 4565 gallons/100 % tank = 45.65 gallons/% tank 45.65 gal/% tank x 26 (HSDBW) = 1186.9 gallons Assumed electrical supply frequency is 60 Hz.

1187 gal 53 gpm = 22.4 minutes 1 SLC pump at 53 gpm injection flow will inject 26% (HSBW) of the SLC tank in approximately 22 minutes.

The candidate should predict that 25 minutes of SLC injection will inject HSBW (drop tank level by 26 percent) which in turn will drop reactor power below 3%. An extra margin of slightly over two minutes is added for conservatism.

Normal ATWS power reduction strategies are to intentionally lower RPV level which lowers reactor power. The stem of the question maintains RPV level steady in order for the candidate to answer the question solely from a boron injection standpoint.

A. This option is incorrect because, with 26% of the SLC injected, hot shutdown boron weight is injected and so reactor power will be less than 3% (APRM downscale). A candidate who believes that HSBW has not yet been injected (the candidate who confuses HSBW with Cold shutdown boron weight) may choose this answer believing that until 60% of the tank is injected that the APRM downscale will not be in.

C. This option is incorrect because SLC tank level would not be at 26% after 25 minutes with only one pump in operation. This would be the approximate level had two pumps been in operation so a candidate who fails to evaluate that only one pump is in operation may choose this option. Additionally the APRMs would be downscale but a candidate may believe that more than 60% of the tank must be injected in order to get reactor power less than 3%.

D. This option is incorrect because SLC tank level would not be at 26% after 25 minutes with only one pump in operation. This would be the approximate level had two pumps been in operation however so a candidate who fails to evaluate that only one pump is in operation may choose this option.

Technical Reference(s): SOP 2.2.74, Standby Liquid Control, Rev. 52 _____

(Attach if not previously provided) 6.SLC.101 R23 SLC Pump Operability Test ______

(including version/revision number) _________________________

Proposed references to be provided to applicants during examination: __None___________

Learning Objective:

Per INT008-06-10 R27

6. State the SLC tank level at which RPV level may be slowly raised to above the scram setpoint.

Question Source: Bank # _______

Modified Bank # _______

74

New ___X ___

Question History: Last NRC Exam ____________

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __ X__

10 CFR Part 55 Content: 55.41 _ 7 _

Comments:

LOD 3 75

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO Tier # 2 Group # 1 K/A # 215003 A4.03 Importance Rating 3.6 215003 IRM - Ability to manually operate and/or monitor in the control room:

(CFR: 41.7 / 45.5 to 45.8) - A4.03 IRM range switches Question: 51 A Plant startup is in progress with reactor power at or near the point of adding heat.

Reactor period is near infinity with the Reactor Mode Select switch in STARTUP / HOT STANDBY position.

IRM C is on Range 7 and indicates 35 on the Panel 9-5 Recorder, when its range switch is taken to Range 6.

What is the new indication for IRM C and what automatic action(s) occur(s)?

Recorder Indication Automatic Action A. 97 Rod Block Only.

B. 97 Rod Block and Half Scram.

C. 110 Rod Block Only.

D. 110 Rod Block and Half Scram.

Answer:

C. 110 Rod Block Only.

Explanation:

Placing the IRM range switch to the next lower scale will increase the current IRM recorder reading by approximately 3.125 (125/40 = 3.125). This will result in 35 x 3.125 = 109.4 which is above the TRM Control Rod Block Instrumentation setpoint of 108/125 of Full Scale. If the range correlation is calculated incorrectly (40/125 = .36) and 35 / .36 = 97.22.

Distracters:

A. This option is incorrect because IRM C would indicate approximately 110. This choice is plausible due to incorrectly calculating the range correlation and not correctly recalling IRM Rod Block setpoints. The candidate that incorrectly calculates the IRM range correlation and does not recall the IRM Rod block setpoint would select this answer.

B. This option is incorrect because IRM C would indicate approximately 110 and a Rod Block AND Half Scram would not occur at this value of the scale. This choice is plausible due to incorrectly calculating the range correlation and not correctly recalling IRM Rod Block and 76

Scram setpoints. The candidate that incorrectly calculates the IRM range correlation and does not recall IRM Scram and Rod block setpoints would select this answer.

D. This option is incorrect because IRM C indicating 110 causes a Rod Block ONLY. This choice is plausible due to not correctly recalling IRM Rod Block and Scram setpoints. The candidate that correctly calculates the IRM range correlation and does not recall IRM Scram and Rod block setpoints would select this answer.

Technical Reference(s): IOP 4.1.2 Intermediate Range Monitoring System, Rev. 23 Major Design Change DC-76-1 CNS ESAR Proposed references to be provided to applicants during examination: none Learning Objective: OPS Intermediate Range Monitor / COR002-12-02 LO-01: State the purpose of the following items related to Intermediate Range Monitoring:

h. Range switch LO-05: Describe the IRM system design features and/or interlocks that provide the following:

d: Varying system sensitivity levels using Range switches Question Source: Bank #

Modified Bank #

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 (7)

Comments:

LOD 4 77

Panel 9-5 IRM Recorder example showing 1-125% scale.

78

The IRM indicator below is located on back panel 9-12 and illustrates the red 0-40% and black 0-125% scale.

IRM Range Switch example. It should be noted that these switches located on panel 9-5 are mechanically pinned to prohibit going above range 9.

79

SETPOINT CIC 9-5-1/E-7 (2354) 102.5/125 of scale (TRM 108/125 of MNI-NAM-41A through MNI-NAM-41H full scale)

SETPOINT CIC 9-5-1/D-7 (2353) Upscale trip at 117.5/125 of scale NMI-NAM-41A, NMI-NAM-41C, (Tech Spec 121/125 of scale) or inop NMI-NAM-41E, or NMI-NAM-41G due to:

1. IRM module unplugged
2. High voltage low
3. MODE switch not in operate
4. Loss of negative supply voltage 80

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO Tier # 2 Group # 1 K/A # 215004 2.4.3 Importance Rating 3.7 215004 Source Range Monitor - 2.4.3 Ability to identify post-accident instrumentation.

(CFR: 41.6 / 45.4)

Question: 52 Which one of the following is a Post Accident Monitoring (PAM) instrument?

A. Source Range Monitor (SRM)

B. Traversing In-core Probe (TIP)

C. Condensate Storage Tank Level Indicator D. Reactor Building Ventilation Exhaust Plenum Radiation Monitors Answer:

A. Source Range Monitor (SRM)

Explanation:

TLCO 3.3.3 Post Accident Instrumentation (PAM) Instrumentation, Table 3.3.3-1 specifies the instruments that are post accident instrumentation. The neutron monitoring systems in the table are the SRMs, IRMs, and APRMs. The only instrument listed in the options that is a PAM instrument is the SRM.

Distracters:

B. This option is incorrect because the TIP system is not a PAM instrument. This choice is plausible due to TIPs being the ONLY neutron monitoring instrument not being part of PAM.

Because the TIP can enter the core and provide data for different core locations a candidate could believe that this system function is required post-accident and choose this option.

Validation data indicated that candidates thought that this may be a plausible alternative for a post-accident monitor. The TIP system would not be a good selection as a PAM, because it would only be truly effective if there were a significant neutron flux detected.

The SRMs are calibrated for lower neutron flux levels and do not physically communicate outside of the primary containment like the TIPs do. If the TIPs were used with any significant pressure on the RPV and there was a penetration of the TIP tubing, then this would be a direct path from the core to the secondary containment.

C. This option is incorrect because the Condensate Storage Tank level instrument is not a required PAM instrument. Because the tank level instruments provide control room alarms and the tank is an alternate suction source for low pressure ECCS a candidate may choose this option.

81

D: This option is incorrect because the reactor building ventilation radiation monitoring instruments are not used for post-accident monitoring. Because other ventilation radiation monitoring instruments (Turbine Building) are PAM required instruments a candidate may choose this option.

Technical Reference(s): TRM 3.3.3 Proposed references to be provided to applicants during examination: NONE Learning Objective:

OPS Source Range Monitor/COR002-30-02 LO-02 Given conditions and/or parameters associated with the SRM system, determine if related Technical Specification and Technical Requirements Manual Limiting Conditions for Operation are met.

D. Technical Requirements Manual

2. T 3.3.3, Non-Type A, Non-Category 1 Post Accident Monitoring (PAM) Instrumentation INT007-06-02 TRM - Instrumentation
1. Given plant conditions, determine if the following TRM Limiting Conditions for Operation (TLCOs) are met:
c. T 3.3.3 Non-Type A, Non-Category 1 Post Accident Monitoring PAM Instrumentation Question Source: Bank #

Modified Bank #

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 (6)

Comments:

LOD 4 82

83 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO Tier # 2 Group # 1 K/A # 217000 K1.07 Importance Rating 3.1 Knowledge of the physical connections and/or cause-effect relationships between REACTOR CORE ISOLATION COOLING SYSTEM (RCIC) and the following: (CFR: 41.2 to 41.9 / 45.7 to 45.8) - K1.07 Leak detection Question: 53 Following a plant transient the following conditions are present:

  • A steam leak develops in the RCIC Room.
  • RCIC Room area temperature is 195°F and rising.

What is the effect on RCIC?

A. Only RCIC-MO-16 (RCIC STEAM SUPPLY OUTBOARD ISOLATION VALVE) closes and the RCIC turbine trips.

B. Only RCIC-MO-15 (RCIC STEAM SUPPLY INBOARD ISOLATION VALVE) closes and the RCIC turbine coasts down (no trip).

C. RCIC-MO-15 and RCIC-MO-16 (RCIC STEAM SUPPLY INBOARD and OUTBOARD ISOLATION VALVE) close and the RCIC turbine trips.

D. RCIC-MO-15 and RCIC-MO-16 (RCIC STEAM SUPPLY INBOARD and OUTBOARD ISOLATION VALVE) close and the RCIC turbine coasts down (no trip).

Answer:

C. RCIC-MO-15 and RCIC-MO-16 (RCIC STEAM SUPPLY INBOARD and OUTBOARD ISOLATION VALVE) close and the RCIC turbine trips.

Explanation:

RCIC Room temperature is part of the leak detection system. The following conditions will cause an automatic RCIC system isolation (Group 5 isolation):

  • RCIC steam supply low pressure ( 61 psig)
  • RCIC steam supply line high flow ( 288% of rated + 6 sec. TD)
  • RCIC steam line high space temperature (195°F)

When a Group 5 isolation occurs the RCIC Steam Supply Line Inboard and Outboard Isolation valves (MO-15 & 16) close, the RCIC turbine trips and the Minimum Flow valve (MO-27) closes.

Distracters:

84

A. This option is incorrect because RCIC-MO-15 also closes. The actions specified in this option are the actions that occur when a half group 5 isolation occurs on channel A. These are also the automatic actions that occur when the manual isolation button is depressed which only functions when an initiation is present. Since these actions are very specific and occur only when an initiation signal is present, as is the case here, a candidate may choose this option believing that with the initiation signal present the isolation is only on channel A.

B. This option is incorrect because RCIC-MO-16 also closes and the RCIC turbine trips. The actions specified in this option are those that occur with a half group 5 isolation on the B channel. A candidate may believe that with the initiation signal present the turbine coasts down due to the governor valve closing, rather than trip due to the trip throttle valve rapidly closing.

D. This option is incorrect because the RCIC turbine trips when the isolation signal is present.

The isolation signal is a direct turbine trip signal. The candidate may believe the governor valve closes allowing the turbine to coast down rather than the trip throttle valve rapidly closing.

Technical Reference(s): SOP 2.2.67, Reactor Core Isolation Cooling System, Rev. 70 Proposed references to be provided to applicants during examination: NONE Learning Objective: OPS Reactor Core Isolation Cooling/COR002-18-02 R25

10. Predict the consequences of the following on the RCIC system:
n. Steam line break (Steam Tunnel/RCIC Room)

OPS Containment COR002-03-02 R30

6. Describe the interrelationship between PCIS and the following:
f. RCIC Question Source: Bank #

Modified Bank #

New X Question History:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 (7)

Comments:

LOD 3 85

86 87 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO Tier # 2 Group # 2 K/A # 201001 A3.08 Importance Rating 3.0 201001 CRD Hydraulic - Ability to monitor automatic operations of the control rod drive hydraulic system including: (CFR: 41.7) A3.08 Drive water flow Question: 54 The Plant is performing a Reactor Startup.

What is the flow indication on Panel 9-5 CRD-FI-305, (Drive Water Flow) while the rod is withdrawing?

A. 2 gpm B. 4 gpm C. 6 gpm D. 8 gpm Answer:

A. 2 gpm Explanation:

DRIVE WATER LINE: This line connects the drive water header to the HCU manifold. It is normally pressurized to 265 psi above reactor pressure.

WITHDRAWAL LINE: The withdrawal line connects the HCU manifold to the CRDM above piston area. The withdrawal line is pressurized with drive water when the associated CRDM is being withdrawn.

The insert solenoid valve has a throttle valve set to allow the normal insertion flow rate through that stabilizing valve, when it is open (energized). When there is no rod movement, the flow through the insert solenoid valve will be approximately 4 gpm. When there is an insert rod signal from the REACTOR MANUAL CONTROL SYSTEM (RMCS), the insert solenoid valve will close, balancing the 4 gpm flow directed to an HCU for normal insertion of a control rod.

The withdrawal solenoid valve has a throttle valve set to allow the normal withdrawal flow rate through the stabilizing valve when it is open (energized). When there is no rod movement, the flow through the withdrawal solenoid valve will be approximately 2 gpm. A withdraw signal from RMCS will close the valve balancing the 2 gpm flow directed to an HCU for normal withdrawal of a control rod.

Flow from the stabilizing valves passes through a local flow indicator. This local indication is used to adjust the throttle valves and verify proper operation. The normal reading, with no rod movement, is 6 gpm.

88

When the insert valve shuts, the withdrawal valve is still open; therefore the flow indicator will show 2 gpm flow through the stabilizing valves. When the withdrawal valve shuts, the insert valve is still open; therefore the flow indicator will show 4 gpm flow through the stabilizing valves.

Distracters:

B. This answer is incorrect due to CRD withdraw flow being equal to 2 gpm. This choice is plausible due to confusing insert the flow which is 4 gpm. The candidate who confuses the expected insert flow with withdraw flow would choose this option.

C. This answer is incorrect due to CRD withdraw flow being equal to 2 gpm.

This choice is plausible due to confusing total stabilizing flow which is 6 gpm. The candidate who confuses the total stabilizing flow with withdraw flow would choose this option.

D. This answer is incorrect due to CRD withdraw flow being equal to 2 gpm. This choice is plausible due to confusing total stabilizing flow which is 6 gpm and withdraw flow of 2 gpm.

A candidate may know that 6 gpm flow is the normal stabilizing flow and that when a withdrawal occurs that this flow is added to the 6 gpm (yielding 8 gpm) and choose this option.

Technical Reference(s): SOP 2.2.8 (Control Rod Drive Hydraulic System), Rev. 90 Proposed references to be provided to applicants during examination: NONE Learning Objective: LO-9 Given a CRDH system component manipulation, predict and explain the changes in the following parameters:

h. CRD drive water flow Question Source: Bank #

Modified Bank #

New X Question History: Last NRC Exam n\a Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 (6)

Comments:

LOD 3 89

90 91 92 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO Tier # 2 Group # 2 K/A # 201006 A4.04 Importance Rating 3.3 201006 RWM - Ability to manually operate and/or monitor in the control room:

(CFR: 41.7 / 45.5 to 45.8) A4.04 Rod withdrawal error indication Question: 55 The Plant is conducting a Reactor startup with the following conditions present:

  • Group 3 Rod Movement Sheet Insert Limit: 04
  • Group 3 Rod Movement Sheet Withdraw Limit: 08

Which RWM IDT targets turn RED?

A. WITHDRAW BLOCK only.

B. WITHDRAW BLOCK and OUT OF SEQUENCE only.

C. WITHDRAW BLOCK, OUT OF SEQUENCE and SELECT ERROR only.

D. WITHDRAW BLOCK, INSERT BLOCK, OUT OF SEQUENCE and SELECT ERROR only.

Answer:

B. WITHDRAW BLOCK and OUT OF SEQUENCE only.

Explanation:

Rod 26-27 has moved beyond the Withdraw Limit of Group 3 (08) and is at position 10.

The RWM allows only 1 withdrawal error and will impose rod blocks to prevent further progress until the withdraw error is corrected. A withdraw error below the LPSP causes WITHDRAW BLOCK and OUT OF SEQUENCE to turn RED.

The RWM will permit the operator to correct the displayed withdraw error only, even when more than one withdraw error conditions are actually present. You cannot select and insert just any withdraw error you wish to choose; you can only insert the displayed withdraw error rod.

Whenever the RWM is OPERATING < LPSP, the existence of a WITHDRAW ERROR causes OUT OF SEQUENCE (red) and WITHDRAW BLOCK (red) to display.

If, under these same conditions, the selection of any other control rod (other than the one causing the WITHDRAW ERROR) will cause OUT OF SEQUENCE (red), SELECT ERROR (red), INSERT BLOCK (red) and WITHDRAW BLOCK (red) to display.

Whenever a WITHDRAW ERROR occurs in the mode of enforcement (OPERATING < LPSP),

the RWM will ONLY permit error rods to be repositioned.

Distracters:

93

A. This option is incorrect because Control Rod 26-27 is beyond the allowable withdrawal position, therefore, RWM IDT Withdraw Block AND Out of Sequence targets turn red. This choice is plausible due to being partially correct. The candidate that does not remember the In Sequence target also turns red would select this answer.

C. This option is incorrect because with Control Rod 26-27 beyond the allowable withdrawal position only RWM IDT Withdraw Block AND Out of Sequence targets are red. The candidate may choose this if they believe the Select Error turns red also because control rod 26-27 is selected.

D. This option is incorrect because Control Rod 26-27 is beyond the allowable withdrawal position, therefore, only RWM IDT Withdraw Block AND Out of Sequence targets will turn red. The candidate may choose this if they believe the Select Error is Red due to the rod remaining selected and the insert error is present due to the rod double notching in which case this would be the correct answer.

Technical Reference(s): Procedure 4.2 (Rod Worth Minimizer), Rev. 29 Proposed references to be provided to applicants during examination: NONE Learning Objective:

COR002-26-02, OPS Rod Worth Minimizer LO-01 State the purpose of the following items related to the Rod Worth Minimizer:

g. IDT Display Console LO-05 Briefly describe the following concepts as they apply to the RWM:
g. Withdraw error Question Source: Bank #

Modified Bank #

New X Question History: Last NRC Exam n\a Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 (7)

Comments:

LOD 3 94

95 96 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO Tier # 2 Group # 2 K/A # 202001 2.2.22 Importance Rating 4.0 202001 Recirculation G2.2.22- Knowledge of limiting conditions for operations and safety limits.

(CFR: 41.10 / 43.2 / 45.13)

Question: 56 Reactor power steady at 26%.

What is the HIGHEST Core Flow which is a TS Safety Limit Violation?

CORE FLOW A. 9%

B. 10%

C. 11%

D. 13%

Answer:

A. 9%

Explanation:

TS requires if either reactor steam dome pressure is < 785 psig or core flow < 10%, THERMAL POWER shall be < 25%. With reactor power at 26%, core flow at 9% violates the safety limit.

Distracters:

B. This answer is incorrect because at 10% the safety limit is satisfied. This answer is plausible due to the MCPR safety limit being applicable at > 10% core flow The candidate that confuses the MCPR safety limit with the Low Pressure of Flow safety limit would select this answer.

C. This answer is incorrect because at 11% the safety limit is satisfied. This answer is plausible due to the MCPR safety limit being > 1.11 for two loop operation and can be easily confused with 11. The candidate that confuses the two loop MCPR limit with the Low Pressure of Flow safety limit would select this answer.

D. This answer is incorrect because at 13% the safety limit is satisfied. This answer is plausible due to the MCPR safety limit being > 1.13 for single loop operation and can be easily confused with 13. The candidate that confuses the single loop MCPR limit with the Low Pressure of Flow safety limit would select this answer.

97

Technical Reference(s): Technical Specifications 2.0 Safety Limits Proposed references to be provided to applicants during examination: None Learning Objective: INT00705010010800 From memory, state each CNS Safety Limit and discuss the basis for each of the Safety Limits.

Question Source: Bank #

Modified Bank #

New X Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 (10)

Comments:

LOD 3 98

99 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO Tier # 2 Group # 2 K/A # 204000 K1.03 Importance Rating 3.1 204000 RWCU - Knowledge of the physical connections and/or cause-effect relationships between reactor water cleanup system and the following: (CFR: 41.2 to 41.9 / 45.7 to 45.8)

- K1.03 Reactor feedwater system Question: 57 Where does the Reactor Water Cleanup System (RWCU) return water to the reactor?

Via the A. suction side of A reactor recirculation pump.

B. suction side of B reactor recirculation pump.

C. A feedwater line downstream of the outboard check valve.

D. B feedwater line downstream of the outboard check valve.

Answer:

C. A feedwater line downstream of the outboard check valve.

Explanation:

When the return isolation valve is in its normal (open) position it allows return flow from the RHX to the Reactor Feedwater piping, where the water is returned to the reactor vessel via feedwater line "A".

Distractors:

A. This option is incorrect because RWCU returns water to the A FW line. RWCU does however get its water supply form the A RR loop so a candidate who only knows that there is a physical tie to the RR loop but not the purpose of that connection would choose this option.

B. This option is incorrect because the RWCU water returns to the reactor via the A FW line.

The RWCU system does interconnect to the RR system so a candidate who only knows there is a connection but neither specific location nor the purpose of that connection would choose this option.

D: This option is incorrect because the RWCU system return water to the reactor is via the A FW line not the B FW line. A candidate who knows that the return is via the FW system but not which line would choose this option.

100

Technical Reference(s): Procedure 2.2.66 (Reactor Water Cleanup System), Rev. 104 Proposed references to be provided to applicants during examination: NONE Learning Objective: OPS Reactor Water Cleanup COR001-20-01 LO-4 Briefly describe the interrelationship between the RWCU system and the following:

j. Reactor Feedwater system Question Source: Bank #

Modified Bank #

New X Question History: Last NRC Exam n\a Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 (4)

Comments:

LOD 2 101

102 103 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO Tier # 2 Group # 2 K/A # 216000 K2.01 Importance Rating 2.8 216000 Nuclear Boiler Inst. - Knowledge of electrical power supplies to the following:

(CFR: 41.7) K2.01 Analog trip system: Plant-Specific Question: 58 What supplies power to the switch contact logic of Wide Range Level Indicating Switch, NBI-LIS-57A (PCIS initiation logic)?

A. AA-1 B. NBPP A C. RPSPP1A D. 24V Power Panel DC-A Answer:

C. RPSPP1A Explanation:

The Wide Range level instrument NBI-LIS-57A is a Yarway type level indicating switch located on Local Rack LRP-PNL-(25-5A) in the Reactor Building. This switch is powered from 120 VAC RPSPP1A. This switch provides partial initiation to the following functions: Alternate Rod Insertion/Recirculation Pump Trip (ARI/RPT), Groups 1, 3, 6 and 7 isolations as well as Control Room Emergency Filtration System (CREFS) initiation.

Distracters:

A. This answer is incorrect because AA-1 does not supply power to any NBI level instrument or the logic power to any NBI level switch contact. This answer is plausible because panel AA-2 supplies power to narrow range level instruments and the candidate may confuse these two panels and choose this answer.

B. This answer is incorrect because NBPP-1A supplies RPV level Indicators NBI-LI-94A and NBI-LR/PR-97. This power supply is in the same division of power as NBI-LIS-57A, and is therefore plausible.

D: This answer is incorrect because the 24 VDC A supplies neutron monitoring and radiation monitor trip auxiliaries. This answer is plausible because it powers instruments on the same panel and close proximity to the RPV level instruments.

Technical Reference(s): OPS Nuclear Boiler Instrumentation/COR002-15-02 Rev. 26 SOP 2.2.22 Rev. 70 Vital Instrument Power System 104

Proposed references to be provided to applicants during examination: NONE Learning Objective:

COR002-15-02 Rev. 26, OPS Nuclear Boiler Instrumentation LO-05 Predict the consequences of the following on the NBI:

k. Loss of AC power Question Source: Bank #

Modified Bank #

New X Question History: N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 (7)

Comments:

LOD 4 105

106 107 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO Tier # 2 Group # 2 K/A # 226001 K3.02 Importance Rating 3.5 226001 RHR/LPCI: CTMT Spray Mode - Knowledge of the effect that a loss or malfunction of the RHR/LPCI: containment spray system mode will have on following: (CFR: 41.7 /

45.4) - K3.02 Containment/drywell/suppression chamber temperature Question: 59 The Plant is operating at power when a steam leak occurs in the Drywell.

  • RHR is spraying the torus and drywell.
  • Torus pressure is 11 psig lowering 1 psig/minute.
  • Drywell temperature is 245°F lowering slowly.

A logic failure has caused the loss of spray valve control permissive.

What is the effect on drywell temperature?

Drywell temperature A. rises due to the loss of Drywell FCUs.

B. rises due to the loss of Drywell Spray.

C. lowers due to the start of Drywell FCUs.

D. lowers due to the Drywell Spray valves going full open.

Answer:

B. rises due to the loss of Drywell Spray.

Explanation:

RHR must have spray valve control in order to spray containment. With a loss of the permissive, the spray valves close and the loss of containment cooling occurs. Drywell temperature rises due to the steam leak continuing.

A. This answer is incorrect because under the current conditions loss of DW FCUs does not cause DW Temperature to rise. The DW FCUs are secured prior to initiating DW Spray, therefore with temperature lowering, the only option for DW temperature to rise is due to the leak and loss of DW Spray ONLY. This choice is plausible because loss of DW FCUs normally would cause DW temperature to rise, but due to being secured prior to spray would have no impact on the rising DW temperature due to loss of DW Spray. The candidate that does not recognize DW FCUs being secured prior to initiating DW spray would select this answer.

108

C. This answer is incorrect because the Drywell FCUs do not start following loss of spray. This answer is plausible due to the DW FCUs having to be secured (control Switch placed in OFF overrides auto start) prior to spraying the DW. The FCUs have a STANDBY feature which auto starts if DW Temperature reaches 145°F. The candidate who does not understand the condition of the DW FCUs would select this answer.

D. This answer is incorrect because the Drywell Spray valves receive a closure signal on the loss of spray valve permissive. This answer is plausible because some RHR valves receive an open signal with a LPCI signal present. The candidate who does not know the effects of loss of spray valve control would select this answer.

Technical Reference(s): OPS Residual Heat Removal System/COR002-23-02 EOP-5.8.7 Rev.29 Primary Containment Flooding/Spray Systems SOP 2.2.69.3 Rev.46 EOP-5.8 Attachment 2 EOP/SAG Graphs Rev.15 Proposed references to be provided to applicants during examination: None Learning Objective:

Per COR002-23-02, OPS Residual Heat Removal System 7 Given a specific RHR system malfunction, determine the effect on any of the following:

d. Drywell parameters (pressure, temperature)

Question Source: Bank #

Modified Bank #

New X Question History: Last NRC Exam n\a Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 (7)

Comments: LOD 3 109

110 111 112 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO Tier # 2 Group # 2 K/A # 230000 K4.05 Importance Rating 2.8 230000 RHR/LPCI: Torus/Pool Spray Mode - Knowledge of RHR/LPCI: torus/suppression pool spray mode design feature(s) and/or interlocks which provide for the following:

(CFR: 41.7) - K4.05 Pump minimum flow protection Question: 60 The plant is operating at 100% power when a small break LOCA occurs. The following conditions are present:

  • RHR Loop A is in Torus Spray and Suppression Pool Cooling.
  • Torus pressure is 2.3 psig and lowering slowly.

What condition FIRST causes RHR-MO-16A (LOOP A MIN FLOW BYP VLV) to automatically OPEN as flow on RHR Loop A is being lowered?

RHR-MO-16A opens A. 3.5 seconds after flow lowers to less than 490 gpm.

B. 3.5 seconds after flow lowers to less than 2107 gpm.

C. as soon as flow is less than 490 gpm with no time delay.

D. as soon as flow is less than 2107 gpm with no time delay.

Answer:

B. 3.5 seconds after flow lowers to less than 2107 gpm.

Explanation:

Minimum Flow Valves RHR-MO-16A/B - The pump minimum flow valves, one in each loop, provide the necessary flow through the pump in order to prevent pump overheating. The RHR pump minimum flow control valves are normally open. The RHR-MO-16A valve closes when flow is 2107 gpm for 3.5 seconds in the associated loop if RHR-MO-20 (RHR Cross-tie) is not full open, or 2107 gpm in either Loop A or B for 3.5 seconds if RHR-MO-20 is full open. The valve opens 3.5 seconds after flow is less than 2107 gpm on system shutdown.

Distracters:

A. This answer is incorrect because the RHR minimum flow control valve opens when flow is less than 2107 gpm for 3.5 seconds in the associated loop. The flow value of 490 is plausible because it is the HPCI system minimum flow setting. The time delay of 3.5 seconds was used because it is the setting for the RHR minimum flow time delay. If the candidate confuses the HPCI system minimum flow value with the RHR minimum flow value, then this answer may be selected.

113

C. This answer is incorrect because the RHR minimum flow control valve opens when flow is less than 2107 gpm for 3.5 seconds in the associated loop. The flow value of 490 is plausible because it is the HPCI system minimum flow setting. The no time delay is also associated with the HPCI system minimum flow valve. If the candidate confuses the HPCI system minimum flow value with the RHR minimum flow value, then this answer may be selected.

D. This answer is incorrect because the minimum flow control valve opens when flow is less than 2107 gpm for 3.5 seconds in the associated loop. The valve responds after a time delay. If the candidate does not remember the time delay, then this answer may be selected.

Technical Reference(s): SOP 2.2.69 (Residual Heat Removal System), Rev. 97 Proposed references to be provided to applicants during examination: NONE Learning Objective: OPS Residual Heat Removal System/COR002-23-02 LO-3 Describe RHR system design feature(s) and/or interlocks which provide for the following:

c. Pump minimum flow protection Question Source: Bank #

Modified Bank #

New X Question History: Last NRC Exam n\a Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 (7)

Comments LOD 3 114

115 116 117 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO Tier # 2 Group # 2 K/A # 241000 K5.04 Importance Rating 3.3 241000 Reactor/Turbine Pressure Regulator - Knowledge of the operational Implications of the following concepts as they apply to reactor/turbine pressure regulating system:

(CFR: 41.5) K5.04 Turbine inlet pressure vs. reactor pressure.

Question: 61 During a startup the turbine is synchronized to the grid with the following conditions present:

  • Equalizing header pressure is 935 psig.
  • Reactor Pressure is 940 psig.
  • Reactor power is 25%.

As power is raised from 25% to 100%, how does the pressure at the equalizing header and the reactor change, if at all, during the power ascension?

Equalizing header pressure A. remains at 935 psig and reactor pressure rises to about 958 psig.

B. remains at 935 psig and reactor pressure rises to about 990 psig.

C. rises to about 958 psig and reactor pressure rises to about 958 psig.

D. rises to about 958 psig and reactor pressure rises to about 990 psig.

Answer:

D. rises to about 958 psig and reactor pressure rises to about 990 psig.

When the main turbine is on-line, Throttle Pressure is controlled by modulating the governor valves. The pressure target is set via the Human Machine Interface (HMI). The DEH system gain is configured in the controller calculations is set to 3.33%. This gain results in a 3.33% flow demand change for every 1 psi of error sensed at the main steam equalizing header. A 30 psi error change results in a 100% flow demand change. This feature makes the response of the Throttle Pressure linear. RPV Steam Dome pressure varies based on the response of the DEH system to power and is a function of steam flow head loss. Since the head loss is a function of volume flow rate squared, the rise in reactor pressure is greater that the rise in the equalizing header.

Distractors:

A. This option is incorrect because the equalizing header pressure rises as power is raised.

Even though pressure setpoint remains the same the offset between pressure setpoint and the equalizing header pressure provides the error to drive the governor valves to the new position. Reactor pressure does rise as indicated but rises more than to 958 psig. A 118

candidate may very well believe that since pressure setpoint is constant that equalizing header is constant and if that same candidate understands that reactor pressure rises due to head loss they would choose this option.

B. This option is incorrect because even though reactor pressure does rise to about 990 psig, the equalizing header pressure does not remain at 935 psig as power is raised. Even though pressure setpoint remains the same the offset between pressure setpoint and the equalizing header pressure provides the error to drive the governor valves to the new position so the offset between the setpoint and the actual equalizing header gets larger as power is raised. A candidate who understands the true response of reactor pressure but who does not understand that it is the offset between the pressure setpoint and equalizing header that provides the signal to open the governor valves would choose this option.

C. This option is incorrect because reactor pressure rises to greater than the new equalizing header pressure. This choice is a plausible misconception of reactor pressure and equalizing header differential pressure lowering (and equalizing) as reactor power is raised to rated (head losses become negligible at such a high steam flow). The candidate who confuses the head loss between the reactor and the equalizing header as power is raised would choose this option.

Technical Reference(s): SOP 2.2.77.1 Rev.35, Digital Electro-Hydraulic Control System SOP 2.2.77 Rev.111, Turbine Generator Proposed references to be provided to applicants during examination: None Learning Objective: COR002-09-02 Rev.17

5. Explain the interrelation between the following parameter sets, and describe how their interrelationship affects operation of the DEH Control system.
b. Turbine inlet pressure vs. reactor pressure Question Source: Bank #

Modified Bank #

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 (5)

Comments:

LOD 3 119

120 121 122 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO Tier # 2 Group # 2 K/A # 259001 Reactor Feedwater K6.04 Importance Rating 2.8 259001 Reactor Feedwater - Knowledge of the effect that a loss or malfunction of the following will have on the reactor feedwater system: (CFR: 41.7 / 45.7) - K6.04 Extraction steam Question: 62 The plant is operating at 95% power on the 100% rod line when the Non-Return Isolation Check Valve for Feedwater Heater 1-A-5 goes full closed due to a short in its motor operator.

(1) How does feed water temperature respond?

(2) Where does operation stabilize on the power-to-flow map?

(1) (2)

Feedwater temperature 100%Rod Line A. Lowers and returns to near the previous temperature. Below B. Stabilizes at a lower temperature ONLY. Below C. Lowers and returns to near the previous temperature. Above D. Stabilizes at a lower temperature ONLY. Above Answer:

D. Stabilizes at a lower temperature ONLY. Above Explanation:

The Extraction Steam system conducts steam from Main Turbine connections to two parallel Feedwater heater strings to improve the overall efficiency of the reactor by preheating the incoming feedwater and reducing the reactor heat load. The NON-RETURN ISOLATION CHECK VALVES (NRVs) are installed in the extraction steam supply lines to the Feedwater Heaters. NRV-1 is designed with an integral motor operator and when closed will essentially function as a stop check valve. When NRV-1 goes closed, feedwater Heater 1-A-5 heat transfer rate significantly lowers. This causes the FW temperature to lower.

ESAR Vol. V.,Section XIV, Part 5.2.1; Loss of Feedwater Heating A decrease in feedwater temperature due to loss of feedwater heating results in a core power increase. This power rise (at the same reactor recirculation flow) raises the operating point above the 100% rod line. So at a constant flow, power is higher and therefore the operating point is higher on the power to flow map.

Distractors:

123

A. This answer is incorrect because with the loss of extraction steam, feedwater is not being preheated so feedwater temperature lowers and remains lower. The operating point on the power-to-flow map when compared to the rod line is above the 100% rod line due to the now higher reactor power. A candidate who believes that the malfunctioning NRV causes more steam to be admitted to the heater (as is the case with closure of a heater steam dump valve) or who does not know the plant response on the power -to-flow map when feedwater temperature lowers would choose this answer.

B. This answer is incorrect because the operating point on the power-to-flow map is higher not lower after conditions stabilize. The candidate who understands the effect of the NRV malfunction but not the final impact on the reactor may choose this answer.

C. This answer is incorrect because with the loss of extraction steam, feedwater is not being preheated so feedwater temperature lowers and remains lower. The candidate who does not understand the extraction steam system and evaluates the effect of the NRV as that of a dump valve would choose this answer.

Technical Reference(s): 2.4EX-STM, Rev. 18 2.1.10 Station Power Changes, Attachment 1 Power to Flow Map, Rev. 107 Proposed references to be provided to applicants during examination: NONE Learning Objective: OPS Extraction Steam and Heater Drains COR001-04-01 LO-06 Given a specific extraction steam and heater drains malfunction, determine the effect on any of the following:

d. Reactor Feedwater Question Source: Bank #

Modified Bank #

New X Question History: Last NRC Exam n\a Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 (7)

Comments:

LOD 3 124

NRV that closes 125

126 USAR Chapter XIV 127

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO Tier # 2 Group # 2 K/A # 239001 A1.08 Importance Rating 3.8 239001 Main and Reheat Steam System: Ability to predict and/or monitor changes in parameters associated with operating the MAIN AND REHEAT STEAM SYSTEM controls including: A1.08 Reactor pressure (CFR: 41.5 / 45.5)

Question: 63 Reactor power is reduced to 65% for performance of Surveillance Procedure 6.MS.201 (Main Steam Isolation Valve Operability Test).

The MSIV Test Pushbutton is depressed and sticks in the depressed position.

How long does it take the MSIV to fully close from the time the pushbutton is first depressed and what is the final reactor pressure when conditions stabilize?

Time to Close Final Reactor Pressure A. < 5 seconds > Initial Pressure B. < 5 seconds < Initial Pressure C. > 20 seconds > Initial Pressure D. > 20 seconds < Initial Pressure.

Answer:

C. > 20 seconds > Initial Pressure Explanation: Requires operating a Main Steam Isolation Valve and predicting the reactor pressure change.

MSIV Spring Only Closure Test vents air off the MSIV operator to allow closure by spring pressure only which takes greater than 20 seconds to reach full closure. With the pushbutton in the depressed position and when conditions stabilize with the MSIV closed, reactor pressure will be higher due to the increased head loss of 3 Main Steam Lines vs. 4 open Main Steam Lines.

Distracters:

A. This answer is incorrect due to MSIV slow closure being greater than 5 seconds. This choice is plausible if confused with TS maximum MSIV closure time of 5 seconds. The candidate who confuses MSIV slow closure time and correctly identifies reactor pressure response would choose this answer.

128

B. This answer is incorrect due to MSIV slow closure being greater than 5 seconds and reactor pressure rising greater than the original pre-closure pressure. This choice is plausible if confused with TS maximum MSIV closure time of 5 and if confused with equalizing header pressure which would be maintained at the original pressure by the DEH system. The candidate who confuses MSIV slow closure time and confuses equalizing header pressure or DEH response would choose this answer.

D. This answer is incorrect due to reactor pressure rising greater than the original pre-closure pressure. This choice is plausible if confused with equalizing header pressure which would be maintained at the original pressure by the DEH system. The candidate who correctly identifies MSIV slow closure time and confuses equalizing header pressure or DEH response would choose this answer.

Technical Reference(s):

Procedure 6.MS.201 (Main Steam Isolation Valve Operability Test), Rev. 15 Procedure 2.2.56 (Main Steam System), Rev. 49 Proposed references to be provided to applicants during examination: None Learning Objective: COR0021402001070F - Predict the consequences of the following items on the MAIN STEAM SYSTEM: Closure of one or more MSIV's at power Question Source: Bank #

Modified Bank #

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 (5)

Comments:

LOD 3 129

130 131 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO Tier # 2 Group # 2 K/A # 272000 A2.15 Importance Rating 2.5 272000 Radiation Monitoring - Ability to (a) predict the impacts of the following on the radiation monitoring system; and (b) based on those predictions, use procedures to correct, control or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5) A2.15 Maintenance operations Question: 64 Maintenance is being performed on RMP-RM-150B (OFFGAS RAD MONITOR B) while operating at rated power with the following annunciator in alarm:

OFFGAS RAD PANEL/WINDOW:

MON DOWNSCALE OR INOP 9-4-1/D-4 (1) What is the impact on the Off-Gas system if radiation levels rise causing RMP-RM-150A (OFF GAS RAD MONITOR A) to reach the Hi Hi Trip Setpoint?

(2) What action is required following the Off-Gas System isolation IAW Procedure 2.4OG (Off-Gas Abnormal)?

A. (1) Off-Gas isolates IMMEDIATELY.

(2) SCRAM and enter Procedure 2.1.5 (Reactor Scram).

B. (1) Off-Gas isolates IMMEDIATELY.

(2) Lower reactor power per Procedure 2.1.10 (Station Power Changes).

C. (1) Off-Gas isolates after a 15 minute time delay.

(2) SCRAM and enter Procedure 2.1.5 (Reactor Scram).

D. (1) Off-Gas isolates after a 15 minute time delay.

(2) Lower reactor power per Procedure 2.1.10 (Station Power Changes).

Answer:

C. (1) Off-Gas isolates after a 15 minute time delay.

(2) SCRAM and enter Procedure 2.1.5 (Reactor Scram).

Explanation:

Requires knowledge of maintenance activity impact on the OG system and Scram actions of 2.4OG. The off-gas stream is monitored by two radiation monitors: RMP-RM-150A and RMP-RM-150B. Both monitors must be in a tripped condition to start a 15 minute timer which will 132

isolate off-gas after the timer times out. With one channel inoperable for maintenance (mode switch not in operate) half of the trip logic is actuated and (9-4-1/D-4) sounds. When RM-150A goes upscale, ( 9-4-1/C-4) sounds. At this point a 15 minute timer begins counting down. After 15 minutes, an off-gas isolation signal will occur. OG-254 and AOG-902 close after 15 minutes which isolates the off-gas system. IAW 2.4OG - if the OG system isolates due to Hi Radiation, Scram and enter Procedure 2.1.5.

Distracters:

A. This answer is incorrect due Off-gas does NOT isolate immediately under these conditions.

This choice is plausible due to not recalling OG having a 15 min time delay - if stem were change to reflect RB HVAC RM - immediately isolate would be correct. The candidate that confuses OG isolation time delay and correctly identifies a scram is required following isolation due to valid radiation levels would choose this answer.

B. This answer is incorrect due Off-gas does NOT isolate immediately under these conditions and reducing power not required under the given conditions. This choice is plausible due to not recalling OG having a 15 min time delay - if stem were change to reflect RB HVAC RM

- immediately isolate would be correct and power reduction being required to maintain main condenser vacuum and to reduce the OG radiation levels prior to OG isolation. The candidate that confuses OG isolation timeframe and does not recognize AOP Scram action would choose this answer.

D. This answer is incorrect reducing power not required under the given conditions. This choice is plausible due to power reduction being required to maintain main condenser vacuum and to reduce the OG radiation levels prior to OG isolation. The candidate that correctly identifies OG isolation time delay and does not recognize AOP Scram action would choose this answer.

Technical Reference(s):

Procedure 2.4OG (Off-Gas Abnormal), Rev. 22 Proposed references to be provided to applicants during examination: None Learning Objective: COR001-18-02

7. Given a specific Radiation Monitoring system malfunction, determine the effect on any of the following:
b. Station Area Radiation monitoring
11. Predict the consequences of the following items on the Radiation Monitoring system.
g. Maintenance/Surveillance operations Question Source: Bank #

Modified Bank #

New X Question History: Last NRC Exam 133

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 (5)

Comments:

LOD 3 OFF GAS RAD MON RMP-RM-150A(B), Upon a downscale signal DOWNSCALE OR INOP, 1 mR/hr or monitor from both monitors an Off 9-4-1/D-4 inop. (Mode switch Gas system isolation will not in operate or high occur after a 15-min. time voltage level low) delay.

Upon an inoperative signal from both monitors, an Off Gas system isolation will occur after a 15-min. time delay.

ALARM PROCEDURE 2.3_9-4-1 PANEL 9 ANNUNCIATOR 9-4-1 D-4 OFFGAS RAD MON DOWNSCALE OR INOP (Page 134)

OFFGAS RAD PANEL/WINDOW:

MON DOWNSCALE 9-4-1/D-4 OR INOP AUTOMATIC ACTIONS Off-gas timer initiates upon a simultaneous trip signal (High-High, Inoperable, or Downscale) from both off-gas monitors.

134

OFFGAS TIMER PANEL/WINDOW:

INITIATED 9-4-1/C-4 SETPOINT CIC 9-4-1/C-4 (1758) OFFGAS TIMER INITIATED on any RMP-RM-150A and RMP-RM-150B simultaneous combination of an A and B trip due to:

1. Channel inoperable
2. Downscale at 1.05 mR/hr
3. Hi-Hi trip at 1.58E3 for Channel A or Channel B (Tech Spec 1 Ci/sec)

AUTOMATIC ACTIONS After 14 minutes (ODAM 15 minutes) of continuous alarm condition, following valves close:

  • OG-AO-254, OFF/GAS SYSTEM ISOLATION.

OPERATOR OBSERVATION AND ACTION

  • Reduce power, as necessary, per Procedure 2.1.10 to clear alarm.
  • Enter Procedures 2.4OG and 5.2FUEL.

135

136 137 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO Tier # 2 Group # 2 K/A # 288000 A3.01 Importance Rating 3.8 288000 Plant Ventilation - Ability to monitor automatic operations of the plant ventilation systems including: (CFR: 41.7 / 45.7) - A3.01 Isolation/initiation signals Question: 65 What automatically starts the Essential Control Building Ventilation system?

A. Battery room exhaust fan trip.

B. Low Critical Switchgear room temperature.

C. High Critical Switchgear room temperature.

D. Non-essential Control Building supply fan trip.

Answer:

C. High Critical Switchgear room temperature.

Explanation:

The Essential Control Building Ventilation System is located in the Critical AC Switchgear Rooms and is comprised of two 100% capacity redundant supply and exhaust fans. These fans are powered from Critical Division 1 and Division 2 power supplies. When EF-SWGR-1F or EF-SWGR-1G, EXHAUST FAN, is running, the isolation dampers to the RHR SWBP Room and from the Non-Essential Control Building HVAC System close. Either Essential Control Building Ventilation Subsystem can remove any potential hydrogen buildup in the Battery Rooms.

The Essential Control Building Ventilation system functions to ensure the 4160V Critical Switchgear Rooms do not overheat. Essential Control Building HV controls Battery Room temperatures between 50°F and 120°F. On a critical switchgear room high temperature (110°F for room 1F and 115°F for room 1G) the respective system automatically starts.

Distracters:

A. This answer is incorrect because there is no interlock between the battery room exhaust fan operation and the essential control building ventilation system. One design criteria of the essential control building ventilation system is to remove any potential hydrogen buildup in the Battery Rooms should a Battery Room exhaust fan not be in operation. Because the essential control building ventilation system services the Battery Rooms, this answer is plausible.

B. This answer is incorrect because a low temperature of 50°F causes the essential control building ventilation system to trip not start. This answer is plausible because the essential control building ventilation system does have an interlock with the critical switchgear room temperature. If the interlock is not correctly determined, this answer may be chosen by a candidate.

138

D. This answer is incorrect because the trip of the non-essential fan does not have an interlock with the essential control building ventilation system. However, this answer is plausible because the initiation of the essential control building ventilation system will cause the non-essential control building fans to trip. The candidate who knows one of the system causes the other to trip would select this answer. The normal convention of HVAC fans is for one to be in standby and automatically start if the running fan trips. Because this area is served by both systems, the interlock may be blurred.

Technical Reference(s): SOP 2.2.38, HVAC CONTROL BUILDING, Rev. 40 Proposed references to be provided to applicants during examination: NONE Learning Objective: OPS Heating, Ventilation and Air Conditioning COR001-08-01 R28

11. Describe the HVAC design features and interlocks that provide for the following:
c. Automatic starting and stopping of fans Question Source: Bank #

Modified Bank #

New X Question History: Last NRC Exam n\a Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 (7)

Comments:

LOD 3 139

140 141 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO Tier # 3 Group #

K/A # 2.1.8 Importance Rating 3.4 2.1.8 Ability to coordinate personnel activities outside the control room. (CFR: 41.10 /

45.5 / 45.12 / 45.13)

Question: 66 With the plant at power, a manual valve is closed in order to maintain Primary Containment OPERABLE IAW TS LCO 3.6.1.3 (PCIVs).

The valve must be opened for maintenance purposes. The NLO opens the valve and remains at the valve.

Which of the following identifies the MINIMUM additional actions required IAW Procedure 2.0.1 (Plant Operations Policy)?

A. Establish intermittent communication with the Control Room.

B. Instruct the NLO to close the valve in event of an accident condition ONLY.

C. Establish continuous communication with the Control Room and instruct the NLO to close the valve in event of an accident condition ONLY.

D. Establish continuous communication with the Control Room, instruct the NLO to close the valve in event of an accident condition and document the instructions in the Control Room log.

Answer:

D. Establish continuous communication with the Control Room, instruct the NLO to close the valve in event of an accident condition and document the instructions in the Control Room log.

Explanation:

This question requires the Reactor Operator to coordinate personnel activities outside the control room (Plant Operator local valve operations under TS Administrative Controls).

Procedure 2.0.1 (Plant Operations Policy) provides the following guidance:

Isolation valves closed to satisfy LCO 3.6.1.3 may be re-opened on an intermittent basis following administrative controls:

  • A person shall be stationed at valve controls while valve is open.
  • If valve is being controlled outside of Control Room, person at valve controls shall be in continuous communication with Control Room.
  • Person at valve controls shall be instructed to close valve in event of an accident condition. These instructions shall be documented (the Control Room log satisfies this requirement).

142

Distracters:

A. This answer is incorrect; specifically continuous communication and direction for valve operation during accident conditions is not included. There is an allowance for opening the containment isolation valve on an intermittent basis under administrative controls. This choice is plausible if the candidate does not remember all of the specific requirements for opening a PCIV under administrative controls.

B. This answer is incorrect due to not providing ALL the procedural requirements; specifically direction for valve operation during emergency conditions is not included in the selection.

This choice is plausible if the candidate does not remember all of the requirements for opening a PCIV under administrative controls.

C. This answer is incorrect due to not providing ALL the procedural requirements; specifically the requirement to document the instructions in the Control Room log. This choice is plausible if all the requirements for opening a PCIV under administrative controls are not specifically known. The candidate does not recall the requirement to document valve closure directions in the CR log would select this answer.

Technical Reference(s):

Procedure 2.0.1 (Plant Operations Policy), Rev. 62 Proposed references to be provided to applicants during examination: None Learning Objective:

SKL0080102 Ops Watchstanding Principles for Licensed Operators 0010400 Briefly describe the administrative controls for primary or secondary containment manual valve opening or associated cap removal when primary or secondary containment is required.

Question Source: Bank #

Modified Bank #

New X Question History: Last NRC Exam n\a Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 (10)

Comments:

LOD 2 143

144 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO Tier # 3 Group #

K/A # 2.1.18 Importance Rating 3.6 2.1.18 Ability to make accurate, clear, and concise logs, records, status boards, and reports.

(CFR: 41.10 / 45.12 / 45.13)

Question: 67 The Reactor has just been declared CRITICAL during plant startup. In addition to:

  • Date/Time
  • Sequence
  • Reactor Pressure What additional criticality information is required to be logged in the Control Room Operators Log IAW Procedure 2.1.1 (Startup Procedure)?

A. SRM Counts AND IRM Overlap.

B. SRM Counts AND Moderator Temperature.

C. Reactor Period AND IRM Overlap.

D. Reactor Period AND Moderator Temperature.

Answer:

D. Reactor Period AND Moderator Temperature.

Explanation:

This answer requires knowledge/ability of required accurate criticality log entries. Procedure 2.1.1 requires the following narrative log criticality data - Log control rod number, control rod position, moderator temperature, reactor pressure, sequence, reactor period, and time on Procedure 10.13, Attachment 1, and in Control Room Operator's log.

Distracters:

A. This answer is incorrect due to SRM Counts and IRM overlap not required to be logged in the narrative log for criticality data. This choice is plausible due to SRM & IRM chart recorders are required to be annotated and IRM overlap required to be document in Procedure 2.1.1. The candidate who confuses annotating charts and documenting IRM overlap would choose this answer.

B. This answer is incorrect due to SRM Counts not required to be logged in the narrative log for criticality data. This choice is plausible due to SRM & IRM chart recorders are required to 145

be annotated IAW Procedure 2.1.1. The candidate who confuses annotating charts and correctly identifies Moderator Temperature would choose this answer.

C. This answer is incorrect because IRM overlap is not required to be logged in the narrative log for criticality data. This choice is plausible due to IRM overlap is required to be documented in Procedure 2.1.1 but not at the time of criticality. The candidate who correctly identifies Period and confuses documenting IRM overlap would choose this answer.

Technical Reference(s): GOP 2.1.1, Startup Procedure, Rev. 178 Proposed references to be provided to applicants during examination: NONE Learning Objective:

INT0320104 CNS Administrative Procedures General Operating Procedures (Startup and Shutdown) Procedures 00A0800 Describe the required actions to be completed upon achieving criticality as described in Procedure 2.1.1, Startup Procedure.

Question Source: Bank #

Modified Bank #2377 New Question History: Last NRC Exam n\a Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 (10)

Comments:

LOD 3 146

147 148 149 150 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO Tier # 3 Group #

K/A # 2.1.31 Importance Rating 4.6 2.1.31 Ability to locate control room switches, controls, and indications, and to determine that they correctly reflect the desired plant lineup. (CFR: 41.10 / 45.12)

Question: 68 Which of the following identifies the indication and location that confirms the reactor is in Hot Shutdown when the Mode Switch is in the Shutdown position during a reactor cooldown?

A. Feedwater Temperature indicates 200°F on Panel A.

B. RR Suction Temperature indicates 240°F on Panel 9-4.

C. RHR HX Inlet Temperature indicates 200°F on Panel 9-3.

D. Vessel Head Adjacent to Flange temperature indicates 240°F on Panel 9-21.

Answer:

B. RR Suction Temperature indicates 240°F on Panel 9-4.

Explanation:

Requires knowledge of indication location and indication which supports the plant being in Mode 3, IAW TS Definitions - Hot Shutdown (Mode 3) is determined by Mode Switch position (Shutdown) and Average Reactor Coolant Temperature (>212°F) with all reactor vessel head closure bolts fully tensioned. The RR suction temperature is an accurate measure of average reactor coolant temperature and one of the temperatures monitored on reactor cooldown.

200°F is a plausible distractor due to confusing the FPC Time to 200°F (was previously Time to Boil) curve.

Distracters:

A. This answer is incorrect because FW Temperature indicating 200°F is the temperature of the water entering the reactor and not average reactor coolant. This choice is plausible due to Panel A being the location for FW Temperature (correct location) and confusing FW Temperature as a valid indication of Average Reactor Coolant Temperature and temperature required for HOT shutdown confused with COLD shutdown of < 212°F. The candidate who correctly identifies FW Temperature indication is located on Panel A and confuses Cold Shutdown with Hot Shutdown temperature requirements would choose this answer.

C. This answer is incorrect because RHR Hx inlet Temperature indicating 200°F is indication of cold shutdown conditions. This choice is plausible due to Hx inlet Temperature being located on Panel 9-3 (correct temperature) and RHR Hx inlet temperature being reactor coolant temperature if indicating 200°F. The candidate who correctly identifies RHR Hx inlet 151

temperature but confuses Cold Shutdown with Hot Shutdown temperature requirements would choose this answer.

D. This answer is incorrect because RPV metal temperatures are not specified on T.S. Table 1.1-1 when determining Reactor Modes. This answer is plausible because of the heat transport mechanism and temperature gradients involved across the RPV shell wall. The candidate who incorrectly believes that RPV metal temperatures instead of average reactor coolant temperature would provide indication of Reactor Mode conditions would choose this answer.

Technical Reference(s): TS Table 1.1-1 Proposed references to be provided to applicants during examination: None Learning Objective: INT00705010010400 From memory, given a set of plant conditions, determine the plant MODE.

Question Source: Bank #

Modified Bank #

New X Question History: Last NRC Exam n\a Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 (10)

Comments:

LOD 3 152

153 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO Tier # 3 Group #

K/A # 2.2.13 Importance Rating 4.1 2.2.13 Knowledge of tagging and clearance procedures. (CFR: 41.10 / 45.13)

Question: 69 Which of the following identifies the FIRST two tagging order steps for placing a system pump under clearance, and the reason for this order IAW Procedure 0.9 (Tagout)?

OPEN the pump breaker and then close the A. suction valve to minimize draining time.

B. discharge valve to minimize draining time.

C. suction valve to protect low pressure components.

D. discharge valve to protect low pressure components.

Answer:

D. discharge valve to protect low pressure components.

Explanation:

Requires knowledge of tagging procedure 0.9. Procedure 0.9 provides guidance for pump tagging to remove the power source first. If isolating the pump, the discharge valve is closed before the suction valve to prevent possible over-pressurization of low pressure components on the suction side.

Distracters:

A. This answer is incorrect due to the suction valve being closed prior closing the discharge valve and minimizing draining time. This choice is plausible due to confusing the reason for closing the discharge valve prior to the suction and minimizing draining time is desired but not the reason. The candidate who confuses valve closure order and reason for the order would choose this answer.

B. This answer is incorrect because minimizing draining time is not the reason for the order of operations. This choice is plausible due to confusing the reason for closing the discharge valve prior to the suction and minimizing draining time is desired but not the reason. The candidate who correctly identifies the valve closure order and confuses the reason for the order would choose this answer.

C. This answer is incorrect due to the suction valve being closed prior closing the discharge valve. This choice is plausible due to confusing the reason for closing the discharge valve prior to the suction. The candidate who confuses valve closure order and correctly identifies the reason for the order would choose this answer.

154

  • Technical Reference(s): Procedure 0.9, Tagout, Rev. 85 Proposed references to be provided to applicants during examination: None Learning Objective: SKL00803020010600 Describe the proper sequence for hanging and picking up tags with regards to Tagging Orders.

Question Source: Bank #

Modified Bank #

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 (10)

Comments:

LOD 3 155

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO Tier # 3 Group #

K/A # 2.2.41 Importance Rating 3.5 2.2.41 Ability to obtain and interpret station electrical and mechanical drawings.

(CFR: 41.10 / 45.12 / 45.13)

Question: 70 An Operator is preparing to write a Clearance Order on a Solenoid Valve.

(1) Where can the Operator obtain Controlled Copies of the electrical and mechanical drawings needed to prepare the Clearance?

(2) When interpreting the electrical drawing, what is the status of Contact 3-4 when the switch is in the CLOSE position?

(1) (2)

A. Control Room Open B. Operations Support Center Open C. Control Room Closed D. Operations Support Center Closed Answer:

A. Control Room Open Explanation:

GENERATING A TAGOUT SECTION Controlled Distribution drawings are located in the areas identified below:

156

a. Aperture cards located in Technical Support Center (TSC), Information Resource Center (IRC), AND Central Alarm Station (CAS).
b. Full size drawings with copies located in the TSC, EOF, Control Room, Simulator Control Room, I&C Shop, E-Shop, Work Control Center (WCC), and Planning.

The embedded switch development depicts a switch with two positions: Close (1) and Open (2). An x in the column is used to determine the status of the contacts (1-2, 3-4) associated with the switch.

In the given switch development matrix, an x is in column (2), which corresponds to the Open position. The contacts 3-4 will be in the condition as drawn (i.e. closed). Should the switch be in the Close position, column (1), it does not have an x in it; therefore contacts 3-4 would be open.

Distractors:

B. This option is incorrect because the Operations Support Center is not on the Controlled Drawing distribution list. Contact 3-4 is open when the switch is in the Close position. A candidate may believe that because the operations support center uses many drawing that this would be where controlled copies exist would choose this option. It should also be noted that this may be confused with the Operational Support Center which is in the same area as the TSC and does maintain controlled documents for emergency purposes.

C. This option is incorrect because contact 3-4 is open when the switch is in the Close position. A candidate may believe that the position of the x in the switch development means open and would choose this option. This answer is plausible because the drawing location is correct.

D. This option is incorrect because the Operations Support Center is not on the Controlled Drawing distribution list and contact 3-4 is open when the switch is in the Close position.

A candidate may believe that the position of the x in the switch development means open and that the OSC is on the distribution list and choose this option. This answer is plausible because the switch contact 3-4 do exist. It should also be noted that this may be confused with the Operational Support Center which is in the same area as the TSC and does maintain controlled documents for emergency purposes.

Technical Reference(s): 3.DRAWING, Drawing Control, Rev. 4 0.9, Tagout, Rev. 85 OTH015-09-08, Electrical Print Reading for Clearance Order and Circuit Evaluation Applications, Rev. 0 Proposed references to be provided to applicants during examination: none Learning Objective: OTH015-09-08/Electrical Print Reading for Clearance Order and Circuit Evaluation Applications

5. Describe the basic types of electrical drawings.
a. Determine the status of prints and actions required when generating a clearance order.
b. Determine where electrical prints are available and how to retrieve them.

Question Source: Bank #

Modified Bank #

New X Question History: Last NRC Exam 157

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 (10)

Comments:

LOD 2 158

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO Tier # 3 Group #

K/A # 2.3.4 Importance Rating 3.2 2.3.4 Knowledge of radiation exposure limits under normal or emergency conditions.

(CFR: 41.12 / 43.4 / 45.10)

Question: 71 Which of the following is the FIRST accumulated dose value ABOVE WHICH a tour member is required to have an NRC Form 5 or equivalent issued IAW 9.ALARA.1 (Personnel Dosimetry and Occupational Radiation Exposure Program)?

A. 100 mrem B. 500 mrem C. 1000 mrem D. 5000 mrem Answer:

A. 100 mrem Explanation:

9.ALARA.1 Personnel Dosimetry and Occupational Radiation Exposure Program step 5.5 specifies the dose value which requires an NRC Form 5 or equivalent to be issued. The accumulated dose value specified is >100 mrem. 9.ALARA.13 provides the instructions for completing the NRC Form 5.

Distractors:

B. This option is incorrect because the selection is above the >100 mrem accumulated value.

This choice is plausible because it is the accumulated dose allowed for a pregnant female per 10CFR20.1208. If the candidate confuses these two values this option would be chosen due to 500 mrem being a familiar dose value.

C. This option is incorrect because the selection is above the >100 mrem accumulated value.

This choice is plausible because it is the CNS administrative dose limit for the year. If the candidate confuses these two values this option would be chosen due to 1000 mrem being a familiar dose value.

D. This option is incorrect because the selection is above the >100 mrem accumulated value.

This choice is plausible because it is the annual allowed 10CFR20.1201 TEDE dose. If the candidate confuses these two values this option would be chosen due to its potential familiarity.

159

Technical Reference(s): 9.ALARA.1 (Personnel Dosimetry and Occupational Radiation Exposure Program), Rev. 43 Proposed references to be provided to applicants during examination: NONE Learning Objective: INT032-01-15R06 OPS CNS Administrative Procedures Radiation Protection and Chemistry Procedures.

D. Procedure 9.ALARA.1, Personnel Dosimetry and Occupational Radiation Exposure Program

g. Monitoring Tour Groups Question Source: Bank #

Modified Bank #

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 (12)

Comments:

LOD: 4 160

161 162 163 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO Tier # 3 Group #

K/A # 2.3.5 Importance Rating 2.9 2.3.5 Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.

Question: 72 What is/are the MINIMUM personnel monitoring requirement(s) for exiting a contaminated area in the Reactor Building and dressing in street clothing IAW 9.EN-RP-100 (Radiation Worker Expectations)?

A. Perform a hand and foot frisk with a frisker ONLY.

B. Perform a whole body contamination monitor scan ONLY.

C. Perform a whole body frisk with a frisker then a whole body contamination monitor scan.

D. Perform a hand and foot frisk with a frisker then a whole body contamination monitor scan.

Answer:

D. Perform a hand and foot frisk with a frisker then a whole body contamination monitor scan.

Explanation:

When exiting a contaminated area within the Reactor Building, personnel are required as a minimum to perform a hand and foot frisk (with a frisker) as soon as practical upon exiting the CA (RB airlock has friskers to support as soon as practical) and then proceed to a PCM for whole body contamination monitoring. Clothing can then be changed within the RCA. If Exiting the RCA personnel are required monitor themselves for contamination with a whole body contamination monitor and a gamma portal monitor.

The candidate should recognize that contamination control requires a minimum of a hand and foot frisk to reduce the potential for spread of contamination. The requirement to exit the RCA via a whole body monitor is always a requirement.

A. This answer is incorrect because it does not include the whole body monitor. The candidate could choose this distractor if he/she does not equate the whole body monitor with contamination control. This answer is plausible because performing hand and foot frisk is required.

B. This answer is incorrect because it does not include hand and foot frisk. The candidate could choose this distractor if he/she does not equate hand and foot frisk with contamination control. This answer is plausible because using the whole body monitor is required.

164

C. This answer is incorrect because it does not properly show the allowance for a hand and foot frisk. The candidate could choose this distractor if he/she did not remember the minimum requirement of the procedure. This answer is plausible because using the whole body monitor is required.

Technical Reference(s): 9.EN-RP-100, Radiation Worker Expectations, Rev. 4 Proposed references to be provided to applicants during examination: None Learning Objective:

Question Source: Bank #

Modified Bank #

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 (11)

Comments:

LOD 2 165

166 167 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO Tier # 3 Group #

K/A # 2.3.13 Importance Rating 3.4 2.3.13 Knowledge of radiological safety procedures pertaining to licensed operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.

(CFR: 41.12 / 43.4 / 45.9 / 45.10)

Question: 73 An Operator has to enter a room to close a valve to stop a small water leak.

  • The affected work area general radiation level is 1600 mrem/hour.

What type of entry permit is required?

Is continuous RP coverage required during the entry?

Entry Permit Continuous RP Coverage A. SWP NOT Required B. SWP Required C. RWP Required D. RWP NOT Required Answer:

B. SWP Required Explanation:

Requires knowledge of LHRA entry requirements. IAW 9.EN-RP-101, entry into LHRAs require a SWP and continuous HP coverage. A Locked High Radiation Area is an area accessible to individuals in which radiation levels from sources external to the body could result in an individual receiving a deep dose equivalent > 1 rem (10 mSv) in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> at 30 cm (~ 12") from the radiation source or from any surface that the radiation penetrates. A Special Work Permit Area (SWP) is an area where a SWP has been issued to control access to, and work within, which involves any one or combination of the conditions: a Very High Radiation Area, Locked High Radiation Area, a High Radiation Area, High Contamination Area, Discrete Radioactive Particle Area, or an Airborne Radiation Area.

Distracters:

A. This answer is incorrect because continuous RP coverage is required. This answer is plausible because an SWP permit is required for entry and under some conditions 168

Operations personnel can take action to protect the health and safety of the public without continuous RP coverage. The isolation of a small leak is not an instance where health and safety of the public is an issue. The candidate who believes continuous RP coverage is NOT required would select this answer.

C. This answer is incorrect because an SWP is required for entry. This answer is plausible because there are instances where Operations personnel can take actions to protect the health and safety of the public without signing on a permit or signing on the SWP. However, the action to isolate a small leak is not an instance where the health and safety of the public is an issue. The candidate who understands continuous RP coverage is required but doesnt realize an SWP is required would select this answer.

D. This answer is incorrect because an SWP and Continuous RP coverage are required to enter a LHRA. This answer is plausible because there are instances where Operations personnel can take actions to protect the health and safety of the public without signing on a permit or signing on the SWP and continuous HP coverage not being required. The candidate who believes the action can be taken and doesnt realize the requirements of a locked high radiation area entry would select this answer.

Technical Reference(s): Procedure 9.EN-RP-101 (Access Control For Radiologically Controlled Areas), Rev 15 Proposed references to be provided to applicants during examination: NONE Learning Objective: INT032-01-100 OPS CNS Administrative Procedures Radiation Protection H. 9-EN-RP-101, Access Control for Radiologically Controlled Areas

1. Precautions and limitations
2. RCA access and egress Question Source: Bank #

Modified Bank #

New X Question History: N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 (12)

Comments:

LOD 2 169

170 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO Tier # 3 Group #

K/A # 2.4.45 Importance Rating 4.1 2.4.45 Ability to prioritize and interpret the significance of each annunciator or alarm.

(CFR: 41.10 / 43.5 / 45.3 / 45.12)

Question: 74 The Plant is at power when a transient occurs.

  • Multiple Black and Yellow outlined annunciators are in alarm.

Which colored alarm take precedence and why IAW Procedure 2.3.1 (General Alarm Procedure)?

A. BLACK; an EOP entry is required.

B. YELLOW; a plant shutdown condition may be present.

C. BLACK; BOTH an EOP entry is required AND a Plant Shutdown condition may be present.

D. YELLOW; BOTH an EOP entry is required AND a Plant Shutdown condition may be present.

Answer:

B. YELLOW; a Plant Shutdown condition may be present.

Explanation:

The window box assembly is a matrix of divided lamp windows with engraved legend plates and multi-colored window bezels. Each window has been given a "PRIORITY" signifying the importance of the alarm:

Priority I - RED; alarms that alert of EOP entry conditions or conditions requiring or causing an automatic or manual plant shutdown, or significant system setpoints.

Priority II - YELLOW; alarm conditions which may require or rapidly cause a plant shutdown or radiation release.

Priority III - BLACK; alarms that indicate off normal plant conditions that affect plant or component operability but should not lead to plant shutdown or radiation release.

Distracters:

A. This option is incorrect because black outlined alarms do not represent EOP entry conditions and do not have a higher priority than yellow outlines alarms. The candidate may 171

choose this if he/she does not understand the color of EOP entry tiles. This option is plausible because black outlines alarms do indicate off normal conditions.

C. This option is incorrect because yellow outlined alarms do not indicate an EOP entry condition is present. The candidate may choose this if he/she does not understand the meaning of color-coded alarms. This option is plausible because yellow outlined alarms do indicate off normal conditions and do have a higher priority than black outline alarms.

D. This option is incorrect because the black outlined alarms do not indicate a plant shutdown condition may be present. The candidate may choose this if he/she does not understand the meaning of color-coded alarms. This option is plausible because black outlined alarms do indicate off normal conditions.

Technical Reference(s): Procedure 2.3.1 (General Alarm Procedure), Rev. 63 Proposed references to be provided to applicants during examination: NONE Learning Objective: COR002-35-02, Plant Annunciator System LO-02 State the purpose of the following components related to the Plant Annunciator System.:

k. Alarm Window Boxes Question Source: Bank #

Modified Bank #

New X Question History:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 (10)

Comments:

LOD 3 172

173 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO Tier # 3 Group #

K/A # 2.4.49 Importance Rating 4.6 2.4.49 Ability to perform without reference to procedures those actions that require immediate operation of system components and controls.

(CFR: 41.10 / 43.2 / 45.6)

Question: 75 There are operational circumstances when operators must perform immediate operator actions without reference to procedures.

Which statement represents one of those circumstances?

A. When an EOP directs performing the action.

B. When a scram is directed by Technical Specifications.

C. When an alarm procedure directs performing the action.

D. When an abnormal procedure directs performing the action.

Answer:

D. When an abnormal procedure directs performing the action.

Explanation:

Abnormal and Emergency (Non-EOP) procedures contain Immediate Operator Actions which the control room operator has committed to memory. Should a condition exist that requires Immediate Operator Actions, Procedure 2.0.1.2 directs performing the action without use of procedures.

Distractors:

A: This answer is incorrect because EOPs do not contain immediate operator actions. Actions are taken per the EOPs without the control room operator having the procedure in hand, but the actions are directed by the CRS. This answer is plausible because actions are directed to be taken without use of the procedure but they are not immediately performed from memory. The candidate who recalls immediately performing actions directed from EOPs would select this answer.

B: This answer is incorrect because no actions are taken immediately from Technical Specifications. This answer is plausible because Technical Specifications have completion times to be taken immediately, but per TS immediate means to pursue without delay and in a controlled manner.

C: This answer is incorrect because alarm procedures do not contain immediate operator actions. This answer is plausible because actions can be performed immediately after 174

entering the procedure. The candidate who recalls scram actions contained in alarm procedures may believe the action can be taken prior to entering the alarm procedure would select this answer.

Technical Reference(s): Procedure 2.0.1.2 (Operations Procedure Policy), Rev. 44 Proposed references to be provided to applicants during examination: none Learning Objective: INT032-01-03 (OPS CNS Administrative Procedure Conduct of Operations and General Alarm Procedures (Formal Classroom/Pre-OJT Training)

G. Procedure 2.0.1.2, Operations Procedure Policy

1. Discuss the following as described in Procedure 2.0.1.2, Operations Procedure Policy.
f. Attachment 1, Immediate Operator Actions Question Source: Bank #

Modified Bank #

New X Question History:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 (10)

Comments:

LOD 2 175

176 177 178 179 180 RO References NOTE 3 Torus overpressure is sum of torus pressure and hydrostatic head above suction strainer Torus pressure (psig) ______________

Hydrostatic head (psig)

PC water level (ft.) ____________

Strainer level (ft.) -4 0.43 x ____________ = +

Torus overpressure (psig) ______________

Figure 1 - TIME TO BOILING - WATER LEVEL AT HIGH LEVEL TRIP Figure 2 - TIME TO BOILING - WATER LEVEL AT FLANGE Figure 3 - TIME TO BOILING - WATER TO LEVEL FLOODED TO 1001

U.S. Nuclear Regulatory Commission Site-Specific SRO Written Examination Applicant Information Name:

Date: Facility/Unit: Cooper Nuclear Station Region: I II III IV Reactor Type: W CE BW GE Start Time: Finish Time:

Instructions Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. To pass the examination you must achieve a final grade of at least 80.00 percent overall, with 70.00 percent or better on the SRO-only items if given in conjunction with the RO exam; SRO-only exams given alone require a final grade of 80.00 percent to pass. You have 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to complete the combined examination, and 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> if you are only taking the SRO portion.

Applicant Certification All work done on this examination is my own. I have neither given nor received aid.

Applicants Signature Results RO/SRO-Only/Total Examination Values / / Points Applicants Scores / / Points Applicants Grade / / Percent 1

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level SRO Tier # 1 Group # 1 K/A # 295006G2.1.6 Importance Rating 4.8 295006 SCRAM 2.1.6: Ability to manage control room crew during plant transients.

Question: 76 Which of the following completes the statement below identifying the CRS direction to the Reactor Operators following an Automatic Scram from rated power and the procedure which provides guidance for this direction?

Assign the RO Procedure 2.1.5 (Reactor Scram) Attachments ____(1)____ IAW Procedure

____(2)____.

Attachment 1 Mitigating Task Scram Actions Attachment 2 Reactor Power Control Attachment 3 Reactor Water Level Control Attachment 4 Reactor Pressure Control Attachment 5 Balance of Plant Actions A. (1) 1 & 2 ONLY AND assign the BOP Attachments 3, 4, and 5.

(2) 2.0.3 (Conduct of Operations)

B. (1) 1 & 2 ONLY AND assign the BOP Attachments 3, 4, and 5.

(2) 2.0.1.3 (Time Critical Operator Action Control and Maintenance)

C. (1) 1, 2 & 3 ONLY AND assign the BOP Attachments 4 and 5.

(2) 2.0.1.3 (Time Critical Operator Action Control and Maintenance)

D. (1) 1, 2 & 3 ONLY AND assign the BOP Attachments 4 and 5.

(2) 2.0.3 (Conduct of Operations)

Answer:

D. (1) 1, 2 & 3 ONLY AND assign the BOP Attachments 4 and 5.

(2) 2.0.3 (Conduct of Operations).

2

Explanation:

Requires knowledge and coordination of procedure 2.1.5 (Reactor Scram) Attachments and selecting the procedure which provides for division of operator responsibilities during a transient. IAW procedure 2.0.3, The CRO-RO is normally responsible for reactivity control, safe reactor shutdown, and mitigating task scram actions and post-scram reactor level control (Attachments 1, 2, and 3 of Procedure 2.1.5). The CRO-BOP is normally responsible for post-scram Pressure Control and BOP System operation (Attachments 4 and 5 of Procedure 2.1.5).

Assign the RO Procedure 2.1.5 (Reactor Scram) Attachments 1, 2, & 3 AND the BOP Attachments 4& 5 IAW Procedure 2.0.3 (Conduct of Operations).

Distracters:

A. This answer is incorrect due to RPV Level control being assigned to the RO. This answer is plausible if ATWS conditions were present, assigning level control to the BOP would be correct. (One of four validators chose this answer due to a common misconception). The candidate who confuses reactor conditions following a scram and correctly identifies the procedure providing guidance would select this answer.

B. This answer is incorrect due to RPV Level control being assigned to the RO and the procedure not providing division of operator responsibilities. This answer is plausible if ATWS conditions were present, assigning level control to the BOP would be correct and scram mitigating actions being a plausible misconception with time critical operator actions.

The candidate who confuses reactor conditions following a scram and confuses time critical operator actions with division of operator responsibilities would select this answer.

C. This answer is incorrect due to the procedure not providing division of operator responsibilities. (One of four validators chose this answer due to a common misconception). This answer is plausible due to scram mitigating actions being a plausible misconception with time critical operator actions. The candidate who correctly identifies Attachments 1, 2, & 3 and confuses time critical operator actions with division of operator responsibilities would select this answer.

Technical Reference(s):

Procedure 2.0.3 (Conduct of Operations), Rev. 87 Procedure 2.0.1.3 (Time Critical Operator Action Control and Maintenance), Rev. 02 Proposed references to be provided to applicants during examination: NONE Learning Objective:

INT032010400D0400 Describe the general sequence of events performed in the Reactor Scram section of procedure 2.1.5, Reactor Scram.

Question Source:

Bank #

Modified Bank #

New X Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 LOD 3 3

SRO Only - 10CFR55.43 b (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Requires coordination of procedure attachments and selection of procedure which provides this guidance.

4

5 6

Examination Outline Cross-

Reference:

Level SRO Tier # 1 Group # 1 K/A # 295018AA2.03 Importance Rating 3.5 295018 Partial or Complete Loss of Component Cooling Water AA2 Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER:

AA2.03: Cause for partial or complete loss Question: 77 The plant is operating at 100% power when the following conditions occur:

  • SW Pressure on both Divisions has risen but is still in the green band.
  • REC system pressure is steady and is in the green band.
  • REC Surge Tank Level High alarms.
  • RR MG Set Oil Temperatures are rising.
  • Drywell temperature and pressure are rising.
  • RWCU F/D Inlet Temp High alarms.

Which one of the following identifies the cause of these conditions and the required action to correct the problem?

The REC Heat Exchanger A. SW Outlet valve has closed, shift REC heat exchangers IAW 5.2REC (Loss of REC).

B. REC Outlet valve has closed, shift REC heat exchangers IAW 5.2REC (Loss of REC).

C. SW Outlet valve has closed, shift REC heat exchangers IAW 5.2SW (Service Water Casualties).

D. REC Outlet valve has closed, shift REC heat exchangers IAW 5.2SW (Service Water Casualties).

Answer:

A. SW Outlet valve has closed, shift REC heat exchangers IAW 5.2REC (Loss of REC).

Explanation:

7

A loss of SW flow to the REC Heat Exchanger recovery is covered in both 5.2REC and 5.2SW (assumes loss of SW pumps or piping). If the pressure of the service water system lowers to <

38 psig the system will isolate non-critical loads. The subsequent steps will have the Operators place the other loops REC Heat Exchanger in service. For the given condition, SW Pressure rising indicates that there was some restriction in the flow path. 5.2SW shift is to use a good SW loop vs. 5.2REC due to REC cooling issues. The actions for shifting REC HXs in 5.2SW allow for bypassing Group 6 signal to SW-MO-650 (REC HX A SERVICE OUTLET), opening REC-19 & 21 (REC HX INLETs), and transferring REC-TIC-451B to MANUAL. These steps differ from the steps in 5.2REC due to shifting heat exchangers for different reasons and would not be appropriate (see highlighted differences provided).

Distracters:

B. This answer is incorrect because the REC outlet valve closing will not cause SW pressure to rise and 5.2SW not being the procedure utilized to shift heat exchangers under the provided conditions. This answer is plausible because closure of this valve would provide the other indications and if the stem were changed to reflect REC pressure rising would be correct and 5.2SW provides guidance to shift (use a good SW loop vs. 5.2REC due to REC cooling issues). The candidate who confuses indications provided and which procedure provides the correct guidance would select this answer.

C. This answer is incorrect because 5.2SW is not the procedure utilized to shift heat exchangers under the provided conditions. This answer is plausible because 5.2SW provides guidance to shift (use a good SW loop vs. 5.2REC due to REC cooling issues).

The candidate who correctly identifies the cause and confuses which procedure provides the correct guidance would select this answer.

D. This answer is incorrect because the REC outlet valve closing will not cause SW pressure to rise. This answer is plausible because closure of this valve would provide the other indications and if the stem were changed to reflect REC pressure rising would be correct along. The candidate who confuses indications provided and correctly identifies the procedure providing the correct guidance would select this answer.

Technical Reference(s):

Emergency Procedure 5.2REC, Loss of REC, Rev. 16 Emergency Procedure 5.2SW, Service Water Casualties, Rev. 24.

Proposed references to be provided to applicants during examination: None Learning Objective:

INT0320126L0L0100 Given plant condition(s), determine from memory the appropriate Abnormal/Emergency Procedure(s) to be utilized to mitigate the event(s).

Question Source:

Bank #

Modified Bank # X (See attached)

New Question History Last NRC Exam: 2011 Question 78 Question Cognitive Level: Memory or Fundamental Knowledge X 8

Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 5 LOD 3 SRO Only - 10CFR55.43 b (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Requires assessment of plant conditions and selection of procedure to mitigate the conditions.

QUESTION: S 3 78 The plant is operating at 100% power when the following conditions occur:

  • SW Pressure on both Divisions have risen but are still in the green band
  • REC system pressure is steady and is in the green band
  • REC Surge Tank Level High alarms
  • RR MG Set Oil Temperatures are rising
  • Drywell temperature and pressure are rising
  • RWCU F/D Inlet Temp High alarms What is the cause and which procedure should be entered to correct the problem?
a. REC Heat Exchanger SW Outlet valve failed closed; enter 5.2REC to correct the problem.
b. Service water non-critical loop isolated; enter 5.2SW to correct the problem.
c. REC system has developed a leak; enter 5.2REC to correct the problem.
d. Service water pump tripped; enter 5.2SW to correct the problem.

ANSWER: S3 78

a. REC Heat Exchanger SW Outlet valve failed closed; enter 5.2REC to correct the problem.

9

10 11 12 13 14 15 Examination Outline Cross-

Reference:

Level SRO Tier # 1 Group # 1 K/A # 295019G2.4.4 Importance Rating 4.7 295019 Partial or Complete Loss of Instrument Air 2.4.4 Ability to recognize abnormal indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures.

Question: 78 An air leak occurs with the plant operating at rated power causing Instrument Air (IA) pressure to lower.

At what pressure is 5.2AIR, Attachment 2 (IA Pressure Loss) FIRST required to be implemented?

A. 90 psig B. 85 psig C. 75 psig D. 60 psig Answer:

C. 75 psig Explanation:

The Control Room Operator must recognize the Emergency Procedure entry due to abnormal system parameter indications. The Control Room Supervisor must evaluate plant conditions and determine when to implement attachments of the procedure. 5.2AIR entry is required if SA or IA pressure is below green band and does not recover back into green band. Per 5.2AIR Subsequent Operator Actions procedure section, when Instrument Air header pressure lowers below 77 psig, Attachment 2 entry is required and the attachment provides instructions that are performed when system pressure is considered to be too low to support continued operation.

Distracters:

A. This option is incorrect because Attachment 2 does not require entry at this high a pressure.

This option is plausible because this is the pressure that starts the second standby air compressor. At this pressure all 3 air compressors are operating loaded. The candidate who believes this pressure is the pressure requiring Attachment 2 entry would select this option.

B. This option is incorrect because 85 psig does not require Attachment 2 performance. This option is plausible because 5.2AIR subsequent operator actions do require actions to be taken if this pressure is reached and 85 psig is the setpoint for SA low pressure annunciator.

The candidate who recalls the 85 psig pressure but does not recognize that Attachment 2 entry is not required would select this option.

16

D. This option is incorrect because 60 psig is significantly below the FIRST pressure requiring Attachment 2 performance. This pressure is plausible because this is the pressure which triggers low air pressure alarms in the main control room. The candidate who recalls the 60 psig pressure being related to low air pressure alarm and recognizes 60 psig is abnormally low would select this option.

Technical Reference(s):

Procedure 5.2Air (Loss of Instrument Air), Rev. 19.

Proposed references to be provided to applicants during examination: NONE Learning Objective:

COR0011702001070A Given a specific Plant Air system malfunction, determine the effect on any of the following: Plant Operation Question Source:

Bank #

Modified Bank #

New X Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 2 55.45 6 Difficulty: 3 SRO Only - 10CFR55.43 b (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

This meets SRO ONLY due to requiring knowledge of when to implement attachments associated with emergency procedures.

17

18 19 20 21 22 Examination Outline Cross-

Reference:

Level SRO Tier # 1 Group # 1 K/A # 295023A2.05 Importance Rating 4.6 295023 Refueling Accidents AA2 Ability to determine and/or interpret the following as they apply to REFUELING ACCIDENTS:

A2.05 Entry conditions of emergency plan Question: 79 The plant is in MODE 5 with refueling in progress when an irradiated fuel bundle is dropped over the core.

The Refueling Floor ARM (RA-1) is 5.5 x 104 mR/hr and rising.

The ERP Kaman is indicating 1.80E +06 µCi/sec and rising.

(1) What is the current MINIMUM required Emergency Classification?

(2) What ERP Kaman reading requires escalation to the next higher classification?

A. (1) Unusual Event (2) 3.70E+06 Ci/sec B. (1) Unusual Event (2) 3.70E+07 Ci/sec C (1) Alert (2) 3.70E+06 Ci/sec D. (1) Alert (2) 3.70E+07 Ci/sec Answer:

D. (1) Alert (2) 3.70E+07 Ci/sec Explanation:

This question was on the 2011 CNS NRC Written exam as Question #77 and included the appropriate EAL handout. An Alert declaration is required if irradiated fuel is damaged and refuel floor radiation levels exceed 50 R/hr. If refuel floor ARM is rising combined with reactor cavity or spent fuel pool level lowering, the threshold for an Unusual Event declaration is met.

The ERP Kaman indication is also a NOUE. The Site Area Emergency for release rate from the ERP is 3.50E+07 Ci/sec. With the information given, an Alert is required to be declared. Note 2 for the EAL Category A, states not to wait for the 15 minute time to elapse before declaring if the start time is unknown. No times were given in the question so this requirement is met.

23

Distracters:

A. This answer is incorrect because an Alert is required. A UE would be correct if radiation level was provided with a lowering SFP level. All other SAE values in Table A-1 are established at 3.50E+06 Ci/sec. Since 3.70E+06 is greater than this value, it is possible to consider the threshold being exceeded. The candidate who chooses an incorrect release point would select this answer. This answer is plausible because the UE threshold is met.

B. This answer is incorrect because an Alert is required. A UE would be correct if radiation level was provided with a lowering SFP level. 3.70E+07 is correct for this release point to require a higher EAL classification. The candidate who misses the damaged fuel EAL would select this answer. This answer is plausible because the UE threshold is met.

C. This answer is incorrect because the listed radiation release is too low for escalation to the next higher EAL. All other SAE values in Table A-1 are established at 3.70E+06 Ci/sec.

Since 3.7E+06 is greater than this value, it is possible to consider the threshold being exceeded. If the candidate misreads the table, then this answer would be chosen. This answer is plausible because the release rates are contained in Table A-1.

Technical Reference(s):

Procedure EPIP 5.7.1 Attachment 4, (Emergency Action Level Matrix), Rev. 11 Procedure EPIP 5.7.1 (Emergency Classification), Rev. 50.

Proposed references to be provided to applicants during examination: EAL Matrix Category A Learning Objective:

GEN0030401C0C050E Concerning event classification: Given a copy of EPIP 5.7.1 and hypothetical abnormal plant symptoms, indications, or events, determine any and all EALs which have been exceeded and specify the appropriate emergency classification.

Question Source:

Bank #

Modified Bank # 19335 New Question History: 2011 NRC exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Difficulty: 3 SRO Only - 10CFR55.43 b (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

The SRO is responsible for EAL classification declarations.

24

REFUEL AREA PANEL/WINDOW:

HIGH RAD 9-3-1/A-10 SETPOINT CIC 9-3-1/A-10

1. (1448) RX BLDG FUEL POOL (HR) AREA 1. RMA-RA-1 RAD HIGH at 500 mR/Hr
2. (1449) RX BLDG FUEL POOL (LR) AREA 2. RMA-RA-2 RAD HIGH at 10 mR/Hr 25

26 27 28 Examination Outline Cross-

Reference:

Level SRO Tier # 1 Group # 1 K/A # 295028G2.4.11 Importance Rating 4.2 295028 High Drywell Temperature 2.4.11 Knowledge of abnormal condition procedures.

Question: 80 While operating at 80% power, the following indications are observed:

  • Drywell Temperature is 148°F and rising 1°F/5 min.
  • Drywell Pressure is 0.35 psig and rising slowly (0.05 psig/min).

What action is required to mitigate these conditions?

Ensure all available DW FCU control switches are in A. RUN IAW 2.4PC (Primary Containment Control).

B. RUN IAW H-1/A-2 (Drywell Zone 1 High Temp).

C. OVERRIDE IAW H-1/A-2 (Drywell Zone 1 High Temp).

D. OVERRIDE IAW 2.4PC (Primary Containment Control).

Answer:

A. RUN IAW 2.4PC (Primary Containment Control).

Explanation: Requires knowledge of abnormal procedure 2.4PC supplemental actions because DW temperature and pressure are rising. 2.4PC requires verification of drywell FCU control switches in RUN vs placing switches in OVERRIDE. The EOPs are the only procedures that allow placing the switches in OVERRIDE. The Drywell Zone 1 alarm provides guidance to check the FCUs are operating and ensuring the cooling to the FCU cooling coils is correctly aligned. The CRS must know that with the temperature and pressure rising if the FCUs are performing their function so it is appropriate to ensure their control switches are in RUN.

29

Distracters:

B. This answer is incorrect because the alarm procedure does not provide direction to ensure all available FCU control switches are in RUN. This answer is plausible because the alarm procedure does contain a step to check they are operating. The candidate who believes the procedure guidance is adequate for the given conditions would select this answer.

C. This answer is incorrect because the procedure is wrong and the action is incorrect. This answer is plausible because the FCU control switches are taken to run but only after drywell temper or pressure rise to EOP 3A levels. The candidate who recalls placing the FCU switches in OVERRIDE is a mitigating strategy but does not recognize that the EOPs are not entered would select this answer.

D. This answer is incorrect because the procedure is wrong and the action is incorrect. This answer is plausible because the FCU control switches are taken to run but only after drywell temper or pressure rise to EOP 3A levels. The candidate who recalls placing the FCU switches in OVERRIDE is a mitigating strategy but does not recognize that the EOPs are not entered would select this answer.

Technical Reference(s):

Procedure 2.4PC (Primary Containment Control), Rev. 17.

ARP 2.3_H-1, Rev. 10.

Proposed references to be provided to applicants during examination: None Learning Objective:

INT0320128K0K0100 Given plant condition(s), and the applicable Abnormal/Emergency Procedure, determine the correct subsequent actions required to mitigate the event(s).

Question Source:

Bank #

Modified Bank #

New X Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 5 55.45 13 Difficulty: 3 SRO Only - 10CFR55.43 b (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

This meets SRO ONLY due to requiring knowledge supplemental AOP actions and when to implement the actions associated with abnormal procedures.

30

31 32 33 Examination Outline Cross-

Reference:

Level SRO Tier # 1 Group # 1 K/A # 295005EA2.04 Importance Rating 3.3 295005 Main Turbine Generator Trip EA2 Ability to determine and/or interpret the following as they apply to MAIN TURBINE GENERATOR TRIP:

EA2.08 Electrical distribution status Question: 81 The plant is operating at rated power with the Startup Transformer out of service for maintenance.

The main generator trips on load reject causing ALL 4160 VAC Buses to de-energize.

(1) What power source automatically reenergizes the Critical Busses?

(2) What action is required IAW 5.3EMPWR (Emergency Power During MODES 1, 2, or 3)?

A. (1) Diesel Generators (2) Direct DCC to perform the CNS-Black Plant Procedure.

B. (1) Diesel Generators (2) Coordinate with DCC to backfeed through the Normal Transformer.

C. (1) Emergency Transformer (2) Direct DCC to perform the CNS-Black Plant Procedure.

D. (1) Emergency Transformer (2) Coordinate with DCC to backfeed through the Normal Transformer.

Answer:

C. (1) Emergency Transformer (2) Direct DCC to perform the CNS-Black Plant Procedure.

Explanation:

With the Startup Station Service Transformer out of service, the Normal Station Transformer is providing power to Critical Buses 1F and 1G through 4160V Buses 1A and 1B. When the main generator trips, the Normal Station Service Transformer becomes de-energized. Buses 4160 1A and 1B become de-energized which for one second de-energizes 1F and 1G. The Emergency Station Service Transformer repowers 1F and 1G directly. The Diesel Generators receive a start signal because of the short-lived (1 second) de-energization of 4160V buses1F and 1G. With all 4160V buses are de-energized for a short period of time a Station Blackout condition exists. However, due to the short lived duration, the proper procedure to enter is 5.3EMPWR. A common misconception is that only procedure 5.3SBO has guidance for directing DCC to enter the CNS Black Plant procedure. Procedure 5.3EMPWR Attachment 3 has been recently revised (within the last two years) to direct DCC to enter the CNS Black Plant procedure.

Distracters:

A. This answer is incorrect because the Critical Buses are energized from the Emergency Service Station Transformer. This answer is plausible if the order in which emergency 34

power supplies energize the Critical Buses is confused or if the stem were changed to reflect a Loss of Offsite Power (LOOP - the emergency transformer is unavailable). The candidate who confuses the order in which emergency power supplies energize the Critical Buses and correctly identifies 5.3EMPWR directs the DCC to enter the Black Plant procedure would select this option.

B. This answer is incorrect because the Critical Buses are energized from the Emergency Service Station Transformer and backfeed is not directed in 5.3 EMPWR. This answer is plausible if the order in which emergency power supplies energize the Critical Buses is confused or if the stem were changed to reflect a Loss of Offsite Power (LOOP - the emergency transformer is unavailable) AND due to the Normal transformer being available for backfeed (Off Site power remains available). The candidate who confuses the order in which emergency power supplies energize the Critical Buses and does not know backfeed through the Normal transformer is only directed in 5.3SBO would select this answer.

D. This answer is incorrect because backfeed is not directed in 5.3 EMPWR. This answer is plausible to the Normal transformer being available for backfeed (Off Site power remains available). The candidate who correctly identifies the order in which emergency power supplies energize the Critical Buses and does not know backfeed through the Normal transformer is only directed in 5.3SBO would select this answer.

Technical Reference(s):

Procedure 5.3EMPWR (Emergency Power During Modes 1, 2, or 3), Rev. 48.

Procedure 5.3SBO (Station Blackout) Rev. 33 Proposed references to be provided to applicants during examination: NONE Learning Objective: INT0320126Q0Q0100, Given plant condition(s), and the applicable Abnormal/Emergency Procedure, determine the correct subsequent actions required to mitigate the event(s).

Question Source:

Bank #

Modified Bank #

New X Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 5 55.45 13 Difficulty: 3 SRO Only - 10 CFR 55.43(b)(5) - Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Requires assessing plant conditions to determine if supplemental actions contained in Attachment 3 (Electrical Systems Guideline) of 5.3EMPWR (Emergency Power During Modes 1, 2, OR 3) are required.

35

36 37 38 Examination Outline Cross-

Reference:

Level SRO Tier # 1 Group # 1 K/A # 600000G2.2.44 Importance Rating 4.4 600000 Plant Fire On Site 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.

Question: 82 While the Fire Brigade is combatting a Temporary Trailer fire (south of the Training Building) with fire hoses, the following indications are observed on the Main Control Room Fire Protection Panel:

G R G R G R ELECTRIC FIRE DIESEL FIRE ELECTRIC FIRE PUMP C PUMP D PUMP E AUTO AUTO AUTO PULL TO PULL TO LOCK LOCK The operator then places the control switch for Fire Pump E to START and the pump starts.

(1) What is the known MINIMUM pressure reached in the fire protection header prior to the operator taking the above action?

(2) What is the operability status of the Fire Suppression Water System?

A. (1) 68 psig (2) Operable B. (1) 68 psig (2) Inoperable C. (1) 141 psig (2) Operable D. (1) 141 psig (2) Inoperable 39

Answer:

B. (1) 68 psig (2) Inoperable Explanation:

Requires candidate to determine Fire Protection header pressure which automatically starts ALL Fire Pumps and recognize failure of Fire Pump E to auto start (interpret control room indications). Based on the information given to the candidate, it is known the header pressure lowered at least to 68 psig which is the automatic start setpoint of Fire Pump C. 141 psig is the automatic start setpoint for Fire Pump D plus a 10 second time delay. With Fire Pump D running this choice is plausible. Second part requires knowledge of TRM bases (requires Fire Pump E to support operability) and TSR 3.11.2.13 which requires each pump to sequentially start based upon lowering system pressure. Fire Pump E failed to auto start on low system pressure and is therefore Inoperable even if manually started (operator action impact) from the Control Room.

Distracters:

A. This answer is incorrect due to the Fire Suppression system being inoperable. This answer is plausible if the SR for this pump to auto start on low system pressure is not known or Fire Pumps D & E are required to support system operability. This is a plausible misconception because there are 2 pumps required and the C pump does not count. The candidate who knows fire pump auto start setpoints and does not know pump auto start on low system pressure is required for operability or which pumps are required for operability would choose this answer.

C. This answer is incorrect due to 141psig being above the auto start setpoint of Fire pump C and the Fire Suppression system being inoperable. This answer is plausible if the Fire pump auto start setpoint are confused (If the stem were changed to indicate Fire Pump C not running, 141 psig would be correct) and the SR for this pump to auto start on low system pressure is not known or Fire Pumps D & E are required to support system operability. This is a plausible misconception because there are 2 pumps required and the C pump does not count. The candidate who confuses fire pump auto start setpoints and does not know pump auto start on low system pressure is required for operability or which pumps are required for operability would choose this answer.

D. This answer is incorrect due to 141psig being above the auto start setpoint of Fire pump C.

This answer is plausible if the Fire pump auto start setpoint are confused (If the stem were changed to indicate Fire Pump C not running, 141 psig would be correct). The candidate who confuses fire pump auto start setpoints and knows pump auto start on low system pressure is required for operability and which pumps are required for operability would choose this answer.

Technical Reference(s):

Technical Requirements Manual (TRM), Rev. 8/27/2014.

Procedure 2.3_FP-4 (Fire Protection - Annunciator 4), Rev. 11.

Procedure 2.2.30 (Fire Protection System), Rev. 62.

Surveillance Procedure 6.FP.102 (Annual Testing of Fire Pumps), Rev. 33 Procedure 0.23 (CNS Fire Protection Plan), Rev. 71 Proposed references to be provided to applicants during examination: NONE Learning Objective:

40

COR00105020010200 Given condition(s) and/or parameters associated with the Fire Protection system, determine if related Technical Requirements Manual Limiting Condition for Operation are met.

Question Source:

Bank #

Modified Bank #

New X Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41.4 55.43.2 55.45.12 Difficulty: 3 SRO Only - 10CFR55.43 b (2) Facility operating limitations in the TS and their bases.

Requires knowledge of Fire Suppression Water System surveillance requirements, bases and CNS Fire Protection Plan.

41

42 43 44 45 Examination Outline Cross-

Reference:

Level SRO Tier # 1 Group # 2 K/A # 295007A2.01 Importance Rating 4.1 295007 High Reactor Pressure AA2 Ability to determine and/or interpret the following as they apply to HIGH REACTOR PRESSURE:

A2.01 Reactor pressure Question: 83 The plant is operating in Mode 1, End-of-Cycle with pressure being maintained at 1015 psig.

A pressure adjustment is made at 0815 on September 1st.

The RO observes STABLE pressure on the following 3 indicators at the specified time:

Time 0817 RFC-PI-90A, RX PRESS is indicating 1015 psig.

Time 0818 RFC-PI-90C, RX PRESS is indicating 1020 psig.

Time 0819 RFC-PI-90B, RX PRESS is indicating 1025 psig.

What is the LATEST time the reactor is required to be in MODE 3 IAW TS 3.4.10, Reactor Steam Dome Pressure if these conditions remain unchanged?

A. 2033 B. 2034 C. 2118 D. 2119 Answer:

B. 2034 Explanation:

RFC-PI-90A and RFC-PI-90C are within specification limits. RFC-PI-90B is noted to be exceeding TS LCO 3.4.10 requirements at 0819. At this time, the CRS is notified and the CRS enters LCO 3.4.10 Condition A. Technical Specification LCO 3.4.10 requires the reactor steam dome pressure be 1020 psig when in Modes 1 and 2. Condition A requires the pressure to be restored within limits within 15 minutes. Condition B requires the unit to be in Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> if Condition A cannot be met. 0819 + 15 minutes (Condition A) + 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> = 0834 +12 hours = 2034.

Distracters:

A. This answer is incorrect because 2033 does not incorporate the 15 min Condition A completion time from the time 1020 psig was exceeded. This answer is plausible if 15 minutes is applied to the time 1020 psig is reached. The candidate that applies the 15 min completion time of Condition A from when pressure reaches 1020 psig would select this answer.

C. This answer is incorrect because 2118 does not incorporate the 15 min Condition A completion time from the time 1020 psig was exceeded. This answer is plausible due to a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> completion time being applicable to other TSs and from memory can be easily 46

confused for Condition A completion time and LCO entry is confused with pressure at 1020 psig vs. >1020. The candidate that applies a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> completion time of Condition A and confuses the actual LCO entry pressure would select this answer.

D. This answer is incorrect because 2119 does not incorporate the 15 min Condition A completion time from the correct time 1020 psig was exceeded. This answer is plausible due to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> completion time being applicable to other TSs and from memory can be easily confused for Condition A completion time. The candidate that applies a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> completion time to Condition A and correctly identifies the actual LCO entry pressure would select this answer.

Technical Reference(s):

Technical Specifications LCO 3.4.10 Proposed references to be provided to applicants during examination: LCO 3.4.10 with LCO Pressure removed and Condition A Completion Time blanked.

Learning Objective:

INT007-05-05, OPS Tech Specs 3.4 Reactor Coolant System (RCS)

3. Given a set of plant conditions that constitutes noncompliance with a section 3.4 LCO, determine the ACTIONS that are required.
10. From memory, in MODES 1 and 2, state the actions required in one hour if the reactor steam dome pressure LCO is not met (LCO 3.4.10).

Question Source:

Bank #

Modified Bank #

New X Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41.10 55.43.5 55.45.13 Difficulty: 3 SRO Only - 10CFR55.43 b (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Application of Required Actions (Section 3) IAW with rules of application requirements (Section 1).

47

48 49 50 51 52 Examination Outline Cross-

Reference:

Level SRO Tier # 1 Group # 2 K/A # 295020G2.2.14 Importance Rating 4.3 295020 Inadvertent Containment Isolation 2.2.14 Knowledge of the process for controlling equipment configuration or status.

Question: 84 A loss of Division 1 RPSPP1A power occurs while operating at power. Troubleshooting the problem is about to commence.

What is the response of the RWCU PCIV(s)?

What procedure controls the configuration of the RWCU PCIV(s) during troubleshooting?

A. (1) ONLY RWCU-MO-15 (INBOARD ISOLATION VALVE) closes.

(2) 0.31(Equipment Status Control) is the controlling procedure.

B. (1) ONLY RWCU-MO-15 (INBOARD ISOLATION VALVE) closes.

(2) 0-CNS-WM-102 (Work Implementation and Closeout) is the controlling procedure.

C. (1) BOTH RWCU-MO-15 and RWCU-MO-18 (OUTBOARD ISOLATION VALVE) close.

(2) 0.31(Equipment Status Control) is the controlling procedure.

D. (1) BOTH RWCU-MO-15 and RWCU-MO-18 (OUTBOARD ISOLATION VALVE) close.

(2) 0-CNS-WM-102 (Work Implementation and Closeout) is the controlling procedure.

Answer:

C. (1) BOTH RWCU-MO-15 and RWCU-MO-18 (OUTBOARD ISOLATION VALVE) close.

(2) 0.31(Equipment Status Control) is the controlling procedure.

Explanation:

The loss of Div 1 RPS causes both RWCU PCIVs (Inboard and Outboard) to close due to loss power to the Non-regenerative heat exchanger outlet temperature instrument. The loss of Div 2 RPS power only closes the OUTBD isolation valve for RWCU. Second part requires knowledge

& selection of procedures which provide guidance for configuration control during troubleshooting activities following a plant transient. Procedure 0.31 provides direction for configuration control of Motor operated isolation valves.

Distracters:

A. This option is incorrect because both RWCU-MO-15 and 18 PCIVs close. The procedure controlling configuration is correct. This answer is plausible because generally an RPS bus de-energizing causes one division valve only to close. The candidate who does not recall the power supply to the Non-regenerative heat exchanger outlet temperature switch and that it solely causes both PCIVs to close may select this option.

B. This option is incorrect because both RWCU-MO-15 and 18 PCIVs close. The procedure controlling configuration is not correct. This answer is plausible because generally an RPS bus de-energizing causes one division valve only to close. The candidate who does not recall the power supply to the Non-regenerative heat exchanger outlet temperature switch 53

and that it solely causes both PCIVs to close may select this option. The candidate who does not recall the procedure that controls configuration in this instance may select this answer.

D. This answer is incorrect because the controlling procedure is 0.31 not 0-CNS-WM-102 The correct valve response is listed. This answer is plausible because the correct valve response is listed. The candidate who does not correctly recall the proper configuration control procedure but does recall the correct valve response may select this answer.

Technical Reference(s):

Procedure 2.2.22 (Vital Instrument Power System), Rev. 71 Procedure 0.31(Equipment Status Control), Rev. 71.

Procedure 0-CNS-WM-102 (Work Implementation and Closeout), Rev. 01.

Proposed references to be provided to applicants during examination: None Learning Objective:

COR0022102001080C Given a specific RPS malfunction, determine the effect on any of the following: c. PCIS INT032010100H010D Discuss the following as described in Administrative Procedure 0.31, Equipment Status Control: System component checklist requirements Question Source:

Bank #

Modified Bank #

New X Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41.10 55.43.3 55.45.13 Difficulty: 3 SRO Only - 10CFR55.43 b (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Requires assessment of plant conditions and selection of procedure which provides guidance for configuration control during troubleshooting.

54

55 56 57 58 59 60 Examination Outline Cross-

Reference:

Level SRO Tier # 1 Group # 2 K/A # 295034A2.01 Importance Rating 4.2 295034 Secondary Containment Ventilation High Radiation EA2 Ability to determine and/or interpret the following as they apply to SECONDARY CONTAINMENT VENTILATION HIGH RADIATION:

EA2.01 Ventilation radiation levels Question: 85 The plant is in Mode 1 and annunciator 9-4-1/E-4, RX BLDG VENT HI-HI RAD alarms due to a valid signal.

(1) What is the LOWEST radiation level which causes this alarm to actuate?

(2) What is the TS Bases for the allowable value of this instrument setpoint?

A. (1) 5 mR/hr (2) Detect a steam leak in Secondary Containment.

B. (1) 5 mR/hr (2) Detect gross fuel cladding failure.

C. (1) 10 mR/hr (2) Detect a steam leak in Secondary Containment.

D. (1) 10 mR/hr (2) Detect gross fuel cladding failure.

Answer:

D. (1) 10 mR/hr (2) Detect gross fuel cladding failure.

Explanation:

Requires candidate to first determine the annunciator setpoint (radiation level) for Rx Bldg Vent Hi-Hi Rad (10 vs. 5 mR/hr which is the Hi Rad). The second part requires knowledge of TS 3.3.6.2 (Secondary Containment Isolation Instrumentation) Bases for the setpoint. The setpoint is based upon detecting gross fuel cladding failure.

Distracters:

A. This option is incorrect as the setpoint for the Reactor Building Vent HI Hi rad is 10 mr/hr NOT 5 mr/hr. The basis for the setpoint is to detect gross clad fuel failure and NOT a steam leak in Secondary Containment. This choice is plausible if the Hi Rad is confused with the Hi HI Rad setpoint (change stem to reflect Hi Rad annunciator answer becomes correct) and steam leak in Secondary Containment would provide elevated radiation levels but is not the bases. A candidate may select this answer if they confuse the Hi setpoint of 5 mr/hr with the Hi Hi and believe the bases for the setpoint is a steam leak in Secondary Containment.

B. This option is incorrect as the setpoint for the Reactor Building Vent HI Hi rad is 10 mr/hr NOT 5 mr/hr. This choice is plausible if the Hi Rad is confused with the Hi HI Rad 61

setpoint (change stem to reflect Hi Rad annunciator). A candidate may select this answer if they confuse the Hi setpoint of 5 mr/hr with the Hi Hi and know the bases for the setpoint gross fuel cladding failure.

C. This option is incorrect as the setpoint is to detect gross clad fuel failure and NOT a steam leak in Secondary Containment. This choice is plausible due to a steam leak in Secondary Containment would provide elevated radiation levels but is not the bases. A candidate may select this answer if they correctly identify the Hi Hi setpoint and believe the bases for the setpoint is a steam leak in Secondary Containment.

Technical Reference(s):

Procedure 2.3_9-4-1 (PANEL 9 ANNUNCIATOR 9-4-1), Rev. 46.

Technical Specification LCO 3.3.6.2, Secondary Containment Isolation Instrumentation Proposed references to be provided to applicants during examination: NONE Learning Objective:

COR00118020010200 Given condition(s) and/or parameters associated with the Radiation Monitoring System, determine if related Technical Specification, Technical Requirements Manual, and Off Site Dose Assessment Manual Limiting conditions for Operation are met.

Question Source:

Bank #

Modified Bank #

New X Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41.10 55.43.3 55.45.13 Difficulty: 3 SRO Only - 10CFR55.43 b (2) Facility operating limitations in the TS and their bases.

Requires knowledge of TS Bases for the alarm setpoint.

62

RX BLDG VENT PANEL/WINDOW:

HIGH RAD 9-4-1/E-5 SETPOINT CIC 9-4-1/E 5 mR/hr

1. (1777) RX BLDG VENT MONITOR DIV I 1. RMP-RR-455 CH-1 or CH-3 HIGH RAD, Monitor A or C
2. (1778) RX BLDG VENT MONITOR DIV II 2. RMP-RR-455 CH-2 or CH-4 HIGH RAD, Monitor B or D PROBABLE CAUSES Refueling activities.

REFERENCES Technical Specification LCO 3.3.6.2, Secondary Containment Isolation Instrumentation.

Off-Site Dose Assessment Manual DLCO 3.2.1, Gaseous Effluents Concentration.

Off-Site Dose Assessment Manual DLCO 3.2.2, Noble Gases Dose.

Off-Site Dose Assessment Manual DLCO 3.2.3, Iodine and Particulates.

Emergency Procedure 5.1RAD, Building Radiation Trouble.

RX BLDG VENT PANEL/WINDOW:

HI-HI RAD 9-4-1/E-4 SETPOINT CIC 9-4-1/E

1. (1763) RX BLDG VENT MONITOR A 1. RMP-RM-452A HI-HI RAD at 10 mR/hr
2. (1764) RX BLDG VENT MONITOR B 2. RMP-RM-452B HI-HI RAD at 10 mR/hr
3. (1779) RX BLDG VENT MONITOR C 3. RMP-RM-452C HI-HI RAD at 10 mR/hr
4. (1780) RX BLDG VENT MONITOR D 4. RMP-RM-452D HI-HI RAD at 10 mR/hr PROBABLE CAUSES Refueling floor accident.

Line external to primary containment breaks.

REFERENCES General Operating Procedure 2.1.22, Recovering from a Group Isolation.

Emergency Procedure 5.1RAD, Building Radiation Trouble.

63

64 Examination Outline Cross-

Reference:

Level SRO Tier # 2 Group # 1 K/A # 203000A2.04 Importance Rating 3.6 203000 Residual Heat Removal /Low Pressure Coolant Injection: Injection Mode A2 Ability to (a) predict the impacts of the following on the RHR/LPCI: INJECTION MODE (PLANT SPECIFIC); and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

A2.04 A.C. failures Question: 86 The following conditions exist during a large break LOCA from rated power:

  • No Off-Site is power available.
  • DG1 is unavailable.
  • RHR Pump D is unavailable.
  • Reactor Building is inaccessible.
  • RPV Pressure is 35 psig and steady.

(1) Which RHR Loop is available for LPCI injection?

(2) What action is procedurally required if additional injection is required to assure adequate core cooling?

A. (1) A (2) Enter 5.3EMPWR (Emergency Power During MODES 1, 2, OR 3) and use the Supplemental Diesel Generator (SDG) to re-energize 4160V 1F Bus.

B. (1) A (2) Enter 5.3ALT-STRATEGY (Alternate Core Cooling Mitigating Strategies) and inject using Fire Protection to RHR.

C. (1) B (2) Enter 5.3ALT-STRATEGY (Alternate Core Cooling Mitigating Strategies) and inject using Fire Protection to RHR.

D. (1) B (2) Enter 5.3EMPWR (Emergency Power During MODES 1, 2, OR 3) and use the Supplemental Diesel Generator (SDG) to re-energize 4160V 1F Bus.

ANSWER:

A. (1) A (2) Enter 5.3EMPWR (Emergency Power During MODES 1, 2, OR 3) and use the Supplemental Diesel Generator (SDG) to re-energize 4160V 1F Bus.

Explanation of answer:

65

RHR Loop A consists of RHR pumps A & C. RHR Loop B consists of RHR pumps B & D. Loss of Div 1 4160V Bus 1F de-energizes RHR pumps A & B and AC valves in Loop A. The only pump with power is RHR Pump C (Loop A) with RHR-MO-27A (Outboard Injection Valve) de-energized in the OPEN position provides LPCI flow to the RPV with RHR-MO25A (Inboard Injection Valve) being DC powered. No RHR pumps are available in RHR Loop B (RHR pump D unavailable and B with no power). The connection of Fire Protection to SW Emergency Core Flooding ties into RHR Loop A piping. This is accomplished by tying in Fire Protection to the suction side of the RHRSW Booster pump via a 3 inch fire hose. The Diesel Fire pump discharge pressure is approximately 160 psig. With the friction loss going through the fire hose and the pressure drop across the RHRSW Booster pump the Fire Protection pressure is too low to inject with the RHR system running. At full injection flow, the RHR discharge pressure is approximately 175 psig. In order to connect Fire Protection to RHR Loop B requires opening the normally closed MO-20 valve which has no power and access to the Reactor Building is not allowed. Loss of all off-site power with only DG2 supplying Div 2 4160V Bus 1G with a LOCA requires entry into EOP-1A (RPV Control) and 5.3EMPWR. 5.3EMPWR provides guidance for RPV and Containment, Balance of Plant, and Electrical Systems guidance via Attachments 1, 2,

& 3. Requires SRO to determine the most effective and timely means of gaining additional RPV injection under conditions challenging Adequate Core Cooling. Utilizing the SDG will provide 2 additional RHR pumps for injection. EOP-1A provides guidance to utilize procedure 5.3ALT-STRATEGY to align required/desired systems, but due to extreme system/plant conditions, normal operation is not possible. RCIC is the preferred injection system, but with RPV at 35 psig RCIC will not provide sufficient makeup (injecting fire protection is the next preferred but requires Reactor Building access). NOTE: The RHR System operating pressure listed here was obtained from Surveillance Procedure 6.1RHR.101 data and the Diesel Fire Protection system operating pressure was obtained from Surveillance Procedure 6.FP.101 data.

Distracters:

B. This answer is incorrect because aligning Fire Protection to RHR will not add water due to RHR system pressure being much higher than Fire Protection header pressure while RHR Pump C is operating. This choice is plausible because the alignment of FP can be performed. The candidate who correctly identifies the available RHR loop and confuses the differences of system driving head for injection would choose this answer.

C. This answer is incorrect because B Loop RHR is unavailable and the reactor building is inaccessible to manually open crosstie to Loop B (MO-20). This choice is plausible due to the diversity of power supplies to the RHR pumps & valves and 5.3ALT-STRATEGY provides guidance for Fire Protection connection to RHR but this will not work due to RHR system pressure and Fire Protection pressure differences. The candidate who incorrectly identifies the available RHR loop and confuses the Service Water availability for injection would choose this answer.

D. This answer is incorrect because B Loop RHR is unavailable. This choice is plausible due to the diversity of power supplies to the RHR pumps & valves. The candidate who incorrectly identifies the available RHR loop and correctly identifies the need to energize Bus 1F for the SDG would choose this answer.

Technical Reference(s):

Procedure 5.8 Attachment 1(RPV Control EOP-1A), Rev. 17.

Procedure 5.3ALT-STRATEGY Alternate Core Cooling Mitigating Strategies, Rev. 42.

Procedure 5.3EMPWR (Emergency Power During MODES 1, 2, OR 3), Rev. 43.

66

Proposed references to be provided to applicants during examination: NONE Learning Objective:

COR0022302001080A Predict the consequences a malfunction of the following will have on the RHR system: A.C. electrical power (including RPS)

Question Source:

Bank #

Modified Bank # 4029 (Attached)

New Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41.5 55.45.6 Difficulty: 3 SRO Only - 10CFR55.43 b (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Requires assessment of conditions and selection of procedure with which to proceed.

67

68 69 70 71 72 73 74 75 Examination Outline Cross-

Reference:

Level SRO Tier # 2 Group # 1 K/A # S212000G2.1.20 Importance Rating 4.6 212000 Reactor Protection System 2.1.20 Ability to interpret and execute procedure steps.

Question: 87 The plant is in Mode 5 with all OPERABLE Control Rods (Core Cell contains fuel) fully inserted.

Procedure 4.5 (Reactor Protection/Alternate Rod Insertion Systems), Section 7 (Bypass Reactor Mode Switch - Shutdown Position Scram and transfer MODE switch), directs the following Step:

7.3.2.2 Inform Shift Manager that Reactor Mode Switch - Shutdown Position Scram (LCO 3.3.1.1, Function 10) is inoperable due to being bypassed.

a. Enter applicable Conditions and Required Actions for following LCOs as required:
1. LCO 3.3.1.1.
2. LCO 3.10.3.
3. LCO 3.10.4.

What is the MINIMUM Required Action(s) for this step IAW TS LCO 3.3.1.1 (RPS Instrumentation) if an OPERABLE Control Rod is fully withdrawn with the DIV 1 Reactor Mode Switch Scram bypass jumper installed?

Place the bypassed channel in trip...

A. within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> ONLY.

B. OR Insert the Control Rod within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

C. OR Insert the Control Rod within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

D. OR Insert the Control Rod within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Answer:

B. OR Insert the Control Rod within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

Explanation:

Requires the step to be interpreted and executed to determine the TS Required actions.

Requires knowledge of TS Bases and what is being bypassed by the jumpers (ONLY Mode Switch SHUTDOWN position is in RPS Channels A3 & B3). LCO is applicable in MODEs 1, 2 and 5 (With any control rod withdrawn from a core cell containing one or more fuel assemblies).

Requires application of multiple TS ACTIONS to determine correct action with knowledge of TS 3.3.1.1 Bases (Loss of trip capability). Condition C requires restoration of RPS trip capability within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. If not completed within 1 hr, enter Condition H (as directed by TABLE 3.3.1.1-1) which requires initiating action to insert the Control Rod immediately. Options are to exit the 76

MODE of applicability (all Control Rods inserted), place bypassed channel in trip or remove jumper to restore trip capability. With the bypassed channel in trip, trip capability is restored.

Distracters:

A. This answer is incorrect due to placing the bypassed channel in trip within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> not being the only additional TS Action. This choice is plausible if loss of function in one channel and RPS trip (full scram) capability is not recognized and not understanding the impact of inserting the withdrawn control rod (no longer in the MODE of applicability). The candidate who does not recognize a loss of function & trip capability would choose this answer.

C. This answer is incorrect due to inserting the Control Rod within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> not being the only additional TS Action. This choice is plausible if loss of function in one channel and RPS trip (full scram) capability is not recognized and not understanding the impact of inserting the withdrawn control rod (no longer in the MODE of applicability). The candidate who does not recognize a loss of RPS trip capability would choose this answer. Placing the bypassed channel in trip OR Inserting the Control Rod within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is plausible if both jumpers were installed with failure to recognize loss of trip (full scram) capability.

D. This answer is incorrect due to inserting the Control Rod within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> not meeting the additional TS Action. This choice is plausible if loss of function in one channel and RPS trip (full scram) capability is not recognized and not understanding the impact of inserting the withdrawn control rod (no longer in the MODE of applicability). The candidate who does not recognize a loss of function & RPS trip capability but does understand exiting the Mode of applicability (insert Control Rod) would choose this answer.

Technical Reference(s):

Procedure 4.5 (Reactor Protection/Alternate Rod Insertion Systems), Rev. 31.

LCO 3.3.1.1 RPS Instrumentation LCO 3.10.4 Single Control Rod Withdrawal--Cold Shutdown Proposed references to be provided to applicants during examination: LCO 3.3.1.1 &

Table with Mode Switch info only Learning Objective:

COR00221020010200 Given conditions and/or parameters associated with the RPS, determine if related Technical Specification and Technical Requirements Manual Limiting Condition for Operations are met.

COR0022102001050D Briefly describe the following concepts as they apply to RPS: Mode switch position Question Source:

Bank #

Modified Bank #

New X Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41.10 55.43.5 55.45.12 Difficulty: 3 77

SRO Only - 10CFR55.43 b (2) Facility operating limitations in the TS and their bases.

Requires application of TS action statements with knowledge to TS Bases. The SRO is required to determine "loss of function" for a given equipment condition.

78

79 80 81 82 83 84 Examination Outline Cross-

Reference:

Level SRO Tier # 2 Group # 1 K/A # 215003A2.05 Importance Rating 3.5 215003 Intermediate Range Monitor System A2 Ability to (a) predict the impacts of the following on the INTERMEDIATE RANGE MONITOR (IRM) SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

A2.05 Faulty or erratic operation of detectors/system Question: 88 A reactor startup is in progress with the following conditions present:

  • Reactor power is below the point of adding heat.

The following annunciator alarms and clears multiple times within 30 seconds and repeats for 3 minutes due to IRM A spiking.

IRM RPS CH A PANEL/WINDOW:

UPSCALE TRIP OR INOP 9-5-1/D-7 (1) What is the result of IRM A spiking?

(2) What is the MOST Restrictive required Technical Specification action?

A. (1) Rod block ONLY (2) Place Channel or Associated Trip system in Trip within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

B. (1) Rod block and 1/2 scram (2) Place Channel or Associated Trip system in Trip within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

C. (1) Rod block ONLY (2) Place Channel in one trip system OR one trip system in trip within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

D. (1) Rod block and 1/2 scram (2) Place Channel in one trip system OR one trip system in trip within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Answer:

B. (1) Rod block and 1/2 scram (2) Place Channel or Associated Trip system in Trip within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Explanation:

Procedure 4.1.2 provides Abnormal IRM Readings on Attachment 1 to determine IRM operability due to Spiking. Since Rod Block & Scram setpoints (102.5 & 117.5/125 of scale) are reached (>59/125 scale) IRM A is inoperable. The Trip System is made up of 4 Channels (IRMs A, C, E, and G are in Division 1 and B, D, F, and H are in Division 2). TS 3.3.1.1 Condition A is entered for IRMs A and G being inoperable. Table 3.3.1.1-1 requires 3 operable 1

IRMs per trip system. With IRM A &G Inoperable, Table 3.3.1-1 Required Channels per Trip System requirements are not met. LCO 3.3.1.1, Condition A applies and must be met. If 2 IRMs were INOP in two different trip systems, then Condition B would apply.

Distracters:

A. This answer is incorrect because an RPS trip would also occur. The required action is correct. This answer is plausible because an IRM does cause a Rod Block and the required action is correct. The candidate who does not realize the RPS trip occurs would select this answer.

C. This answer is incorrect because and RPS trip would also occur and listed required action is incorrect. If the candidate did not realize the IRMs were in the same trip system may select this answer based upon the required action. The 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to place the channel in one trip system or one trip system in trip is the action taken if the IRMs were in different trip systems (e.g. Div 1 and Div 2). This answer is plausible because the listed required action is taken when its conditions are met.

D. This answer is incorrect because the listed required action is not correct. The system response is correct. If the candidate did not realize the IRMs were in the same trip system may select this answer. The 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to place the channel in one trip system or one trip system in trip is the action taken if the IRMs were in different trip systems (e.g. Div 1 and Div 2). This answer is plausible because the system response is correct and the listed required action is an action one would take if the conditions were met.

Technical Reference(s):

Procedure 4.1.2 (Intermediate Range Monitoring System), Rev. 21.

Procedure 4.5 (Reactor Protection/Alternate Rod Insertion Systems), Rev. 31.

TS LCO 3.3.1.1, Reactor Protection System Instrumentation Proposed references to be provided to applicants during examination: LCO 3.3.1.1 Learning Objective:

COR00212020010200 Given conditions and/or parameters associated with the IRM's, determine if related Technical Specification and Technical Requirements Manual Limiting Condition for Operation are met.

COR0021202001090A Given plant conditions, determine if the following IRM actions should occur: Rod Block.

COR0021202001090B Given plant conditions, determine if the following IRM actions should occur: Reactor Scram.

INT007-05-05, OPS CNS Tech Specs 3.3, Instrumentation

3. Given a set of plant conditions that constitutes non-compliance with a Section 3.3 LCO, determine the ACTIONS that are required.

Question Source:

Bank #

Modified Bank #

New X Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41.50 55.43.2 Difficulty: 3 2

SRO Only - Facility operating limitations in the technical specifications and their bases.

(10 CFR 55.43(b)(2)).

The SRO is responsible for TS LCO required action determination.

IRM PANEL/WINDOW:

UPSCALE 9-5-1/E-7 SETPOINT CIC 9-5-1/E-7 (2354) 102.5/125 of scale (TRM MNI-NAM-41A through MNI-NAM-41H 108/125 of full scale)

IRM RPS CH A PANEL/WINDOW:

UPSCALE TRIP OR INOP 9-5-1/D-7 SETPOINT CIC 9-5-1/D-7 (2353) Upscale trip at 117.5/125 of scale NMI-NAM-41A, NMI-NAM-41C, (Tech Spec 121/125 of scale) or inop NMI-NAM-41E, or NMI-NAM-41G due to:

1. IRM module unplugged
2. High voltage low
3. MODE switch not in operate
4. Loss of negative supply voltage 3

4 5

6 7

8 9

10 Examination Outline Cross-

Reference:

Level SRO Tier # 2 Group # 1 K/A # 259002G2.2.40 Importance Rating 4.5 259002 Reactor Water Level Control System 2.2.40 Ability to apply Technical Specifications for a system Question: 89 The Plant is operating at rated power on March 10th.

At 1300 the RO observes the following:

  • RFC-LI-94A indicates 36 inches and is slowly rising.
  • RFC-LI-94B indicates 32 inches and is slowly lowering.
  • RFC-LI-94C indicates 37 inches and is slowly rising.

If these conditions continue, when is the LATEST reactor power is required to be less than 25%

RTP IAW TS 3.3.2.2, Feedwater and Main Turbine High Water Level Trip??

A. March 10 at 1500 B. March 10 at 1900 C. March 17 at 1300 D. March 17 at 1700 Answer:

D. March 17 at 1700 Explanation:

Requires knowledge of TS instrument surveillance requirements and application of required actions associated with the RVLC System. The RVLC system utilizes NR level and FW flow instruments to properly control RPV water level. If one level instrument differs from the average mean by 8 inches or more, the RVLCS will remove the anomalous input and control utilizing the remaining valid instruments. The RVLC halts the vessel rise after it removes LI-94B from processing. SR 3.3.2.2.1 requires a channel check which is documented in 6.LOG.601 requiring these Narrow Range level indicators to indicate within 2 inches of each other. Since the B NR is greater than 2 inches from A & C, this channel is considered inoperable (failed channel check). TS 3.3.2.2, Condition A is entered at the time the failure is identified and the required time to place the channel in a tripped condition is 7 days (March 17 at 1300). Once the completion time is expired, reactor power must be lowered below 25% RTP IAW Condition C within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. 7 days + 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> = March 17 at 1700.

11

Distracters:

A. This answer is incorrect because only 1 NR instrument being inoperable. This answer is plausible due to 2 NR level indicators being higher than normal setpoint (35) and reflects a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> completion time. The candidate who incorrectly determines 2 level NR Level instruments are inoperable and confuses completion time to restore trip capability vs.

reducing power to <25% would select this answer.

B. This answer is incorrect because only 1 NR instrument being inoperable. This answer is plausible due to 2 NR level indicators being higher than normal setpoint (35) and reflects a 2 + 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> (6 hrs) completion time. The candidate who incorrectly determines 2 level NR Level instruments are inoperable but correctly identifies the additional 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to reducing power to <25% would select this answer.

C. This answer is incorrect due not applying the 4 additional hours allowed to reduce power below 25%. This answer is plausible due to 1 NR level indicator being lower than normal setpoint (35) with higher FW flow and reflects the 7 day completion time (if stem were changed to ask when required to be in trip - would be correct). The candidate who correctly determines 1 level NR Level instrument is inoperable but confuses placing the channel in trip vs. reducing power to <25% would select this answer.

Technical Reference(s):

Technical Specifications LCO 3.3.2.2, Feedwater and Main Turbine High Water Level Trip Instrumentation.

Procedure 4.4.1 (Reactor Vessel Level Control System), Rev. 7 Proposed references to be provided to applicants during examination: TS 3.3.2.2 Learning Objective:

INT007-05-04, OPS CNS Tech Specs 3.3, Instrumentation

3. Given a set of plant conditions that constitutes non-compliance with a Section 3.3 LCO, determine the ACTIONS that are required.

Question Source:

Bank #

Modified Bank #

New X Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41.10 55.43.2 55.43.5 55.45.13 Difficulty: 3 SRO Only - 10CFR55.43 b (2) Facility operating limitations in the technical specifications and their bases.

Requires assessing instrument / plant response and application of TS required actions.

12

13 14 Examination Outline Cross-

Reference:

Level SRO Tier # 2 Group # 1 K/A # 262001A2.06 Importance Rating 2.9 262001 A.C. Electrical Distribution A2 Ability to (a) predict the impacts of the following on the A.C. ELECTRICAL DISTRIBUTION; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

A2.06 Deenergizing a Plant Bus Question: 90 The plant is in Mode 3.

An event requires IMMEDIATELY de-energizing 4160V Bus 1A.

(1) What is the impact of IMMEDIATELY de-energizing 4160V Bus 1A?

(2) How many Off Site Power Circuits are Inoperable IMMEDIATELY following power transfer?

A. (1) A momentary loss of Bus 1F occurs.

(2) One B. (1) A momentary loss of Bus 1F occurs.

(2) Two C. (1) A momentary loss of Bus 1G occurs.

(2) One D. (1) A momentary loss of Bus 1G occurs.

(2) Two Answer:

A. (1) A momentary loss of Bus 1F occurs.

(2) One Explanation:

Requires predicting the impact of de-energizing Bus 1A and knowledge of TS 3.8.1 Bases for Off Site power source OPERABILITY which would then be utilized to control the consequences of this Bus loss IAW TSs. Conditions provided that the plant is in Mode 3. De-energizing Bus 1A without manually transferring Bus 1F prior to de-energizing Bus 1A results in 1 second time delay for auto transfer to the Emergency Station Service Transformer (ESST) (Impact = short duration de-energization). De-energizing Bus 1A results in loss of 1 offsite circuit which is the Startup Station Service Transformer (SSST) because it no longer has fast transfer capability to both 4160V Critical Buses (1F and 1G). The ESST remains operable because it is powering the 4160V Critical Bus 1F (fast transfer complete) and it maintains its fast transfer capability to the other division Critical Bus 1G. When a 4160V critical bus is manually transferred to the ESST IAW Procedure 2.2.18, the opposite divisions ESST supply breaker is placed in pull-to-lock requiring declaring both Off Site power circuits inoperable. The difference between manual transfer and automatic transfer is a common misconception for operability of the Off Site power circuits. Procedurally de-energizing Bus 1A is only done with the plant in Mode 4 or 5, but due 15

to the emergent nature of this event, the bus would be de-energized by opening supply breakers and then followed up by performing applicable steps of procedure 2.2.18.

Distracters:

B. This answer is incorrect because ONLY the SSST is inoperable. This choice is plausible due to the common misconception of OPERABLE Off Site circuits and if de-energized IAW Procedure 2.2.18 requires declaring TWO offsite circuits INOPERABLE. The candidate that correctly recognizes momentary loss of power to Bus 1F and knows both Off Site circuits are declared inoperable due to breaker alignment during manual transfer per procedure would select this option.

C. This answer is incorrect due to Bus 1G not being impacted by de-energizing Bus 1A. This choice is plausible if electric plant alignment is not known or confused. Candidate that confuses momentary loss of power to Bus 1F and correctly identifies the number of INOPERABLE Off Site circuits would select this option.

D. This answer is incorrect due to Bus 1G not being impacted by de-energizing Bus 1A and ONLY the SSST is inoperable. This choice is plausible if electric plant alignment is not known or confused and the common misconception of OPERABLE Off Site circuits and if de-energized IAW Procedure 2.2.18 requires declaring TWO offsite circuits INOPERABLE.

Candidate that confuses momentary loss of power to Bus 1F and does not recognize the correct number of OPERABLE offsite circuits or knows both Off Site circuits are declared inoperable due to breaker alignment during manual transfer per procedure would select this option.

Technical Reference(s):

Procedure 2.2.18 (4160V Auxiliary Power Distribution System), Revision 178.

TS LCO 3.8.1 Proposed references to be provided to applicants during examination: NONE Learning Objective:

INT00705090010300 Given a set of plant conditions that constitutes non-compliance with a Section 3.8 LCO, determine the ACTIONS that are required.

Question Source:

Bank #

Modified Bank #

New X Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41.5 55.45.6 Difficulty: 3 SRO Only - 10 CFR 55.43(b)(2) - Facility operating limitations in the TS and their bases.

Requires knowledge of TS Bases for determination of Off Site circuit INOPERABILITY.

16

17 18 19 20 21 22 Examination Outline Cross-

Reference:

Level SRO Tier # 2 Group # 2 K/A # 201006G2.1.23 Importance Rating 4.4 201006 Rod Worth Minimizer System 2.1.23 Ability to perform specific system and integrated plant procedures during all modes of plant operation.

Question: 91 6.RWM.301 (Rod Worth Minimizer Functional Test For Startup) is being performed.

Which of the following completes the statements below regarding how the Select Error function of the RWM is tested IAW 6.RWM.301 AND the significance of maintaining the RWM operable during startup IAW TS Bases?

The operator is required to verify the SELECT ERROR indicator turns red while selecting

____(1)____ from each RWM group (except Group 1).

The RWM enforces compliance with BPWS which ensures that the initial conditions of the analysis for a ____(2)____ is NOT violated.

A. (1) a single rod (2) control rod drop accident B. (1) a single rod (2) single control rod withdrawal error C. (1) ALL individual rods (2) control rod drop accident D. (1) ALL individual rods (2) single control rod withdrawal error Answer:

A. (1) a single rod (2) control rod drop accident Explanation: Requires knowledge of RWM Functional Test and TS 3.3.2.1 Bases. 6.RWM.301 (Rod Worth Minimizer Functional Test For Startup) requires the operator to verify the SELECT ERROR functions (indicator on PMIS turns red) when selecting a single rod from each RWM group (except Group 1). The RWM enforces BPWS which ensures that the initial conditions of the CRDA analysis are not violated.

Distracters:

B. This answer is incorrect due to a single control rod withdrawal error (RWE) not being the TS bases for the RWM. This choice is plausible due to a CRDA being easily confused with a RWE which is the bases for the Rod Block Monitor. The candidate that correctly identifies selecting only 1 rod in the remaining groups and confuses RBM vs. RWM bases would select this choice.

23

C. This answer is incorrect due to selecting all rods in the remaining groups not being required to support SELECT ERROR function verification. This choice is plausible due having many groups requiring single rod selection being easily confused with all the rods within each group. The candidate that confuses only 1 rod in the remaining groups vs. all rods and correctly identifies the RWM bases would select this choice.

D. This answer is incorrect due to selecting all rods in the remaining groups not being required to support SELECT ERROR function verification and a single control rod withdrawal error (RWE) not being the TS bases for the RWM. This choice is plausible due having many groups requiring single rod selection being easily confused with all the rods within each group and a CRDA being easily confused with a RWE which is the bases for the Rod Block Monitor. The candidate that confuses only 1 rod in the remaining groups vs. all rods and confuses RBM vs. RWM bases would select this choice.

Technical Reference(s):

Procedure RWM.301 (Rod Worth Minimizer Functional Test For Startup), Rev. 10.

TS Bases Proposed references to be provided to applicants during examination: None Learning Objective:

INT007-05-04:

2. Discuss the applicable Safety Analysis in the Bases associated with each Section 3.3 Specification.

Question Source:

Bank #

Modified Bank #

New X Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41.10 55.43.5 55.45.2 55.45.6 Difficulty: 3 SRO Only - 10CFR55.43 b (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Requires knowledge of TS Surveillance Requirements and Bases. It is the SRO who determines how much of a system is required to be tested. In this case, the SRO must know that testing only the first control rod of each RWM group must be tested.

24

25 26 27 Examination Outline Cross-

Reference:

Level SRO Tier # 2 Group # 2 K/A # 202002A2.06 Importance Rating 3.3 202002 Recirculation flow Control System A2 Ability to (a) predict the impacts of the following on the RECIRCULATION FLOW CONTROL SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

A2.06 Low reactor water level: Plant-Specific Question: 92 The plant is operating in MODE 1 at 97% power when RFP A trips.

(1) How is the Recirculation Flow Control System affected if RPV level lowers to 25 inches on NR Level instruments?

(2) If RRMG B speed does not change and CANNOT be reduced, what attachment provides direction to trip RRMG B with RPV level stable at 20 inches on NR Level instruments IAW 2.4RR (Reactor Recirculation Abnormal)?

A. (1) Both Recirculation Pumps run back to 22% speed.

(2) Attachment 1 {Trip of Reactor Recirculation Pump(s)}.

B. (1) Both Recirculation Pumps run back to 22% speed.

(2) Attachment 4 (Reactor Recirculation Flow Control Failure/RRMG Scoop Tube Lockout).

C. (1) Both Recirculation Pumps run back towards 45% speed.

(2) Attachment 1 {Trip of Reactor Recirculation Pump(s)}.

D. (1) Both Recirculation Pumps run back towards 45% speed.

(2) Attachment 4 (Reactor Recirculation Flow Control Failure/RRMG Scoop Tube Lockout).

Answer:

D. (1) Both Recirculation Pumps run back towards 45% speed.

(2) Attachment 4 (Reactor Recirculation Flow Control Failure/RRMG Scoop Tube Lockout).

Explanation:

lf both Reactor Recirculation (RR) pumps are running and not locked out, RR pumps run back towards 45% speed if the following condition is met:

Total steam flow > 9 Mlbm/hr with at least 1 RFP tripped/flow < 1 Mlbm/hr and selected reactor water level < 27.5 inches. Both RR pumps runback towards 45% and stops when the condition causing the runback is no longer true or no other 45% runback conditions exist. With power at 97%, FW flow is around 9.27 Mlbm/hr. As the RR pumps start running back FW flow will lower below 9 Mlbm/hr and runback will stop. If the stem were changed to reflect Discharge valve closure or FW Flow <20% - 22% runback would be correct.

On a trip of a RFP entry into Abnormal Procedure 2.4MC-RF, Condensate and Feedwater Abnormal is required. Additionally the inability to change speed of RRMG B requires entry into 2.4RR (Reactor Recirculation Abnormal). With RPV level not recovering, Attachment 4 (Reactor Recirculation Flow Control Failure/RRMG Scoop Tube Lockout) is required to be performed which directs tripping the RRMG B if a faster power reduction is needed.

28

Distractors:

A. This answer is incorrect because both RR pumps run back towards 45% speed and Attachment 1 not providing guidance to trip RRMG B. Once the condition that caused the runback (lowering FW flow) is clear the RR pumps stop running back. This answer is plausible because the RR pumps would runback to 22% if the Discharge valve closes or FW Flow lowers below 20% and Attachment 1 titled Trip of RR Pump(s). Attachment 1 would be required to be entered following the trip of RRMG B. The candidate that confuses RR Runback logic and is not familiar with 2.4RR attachments and thinks tripping RRMG B IAW Attachment 1 would select this answer.

B. This answer is incorrect because both RR pumps run back towards 45% speed. Once the condition that caused the runback (lowering FW flow) is clear the RR pumps stop running back. This answer is plausible because the RR pumps would runback to 22% if the Discharge valve closes or FW Flow lowers below 20%. The candidate that confuses RR Runback logic and is familiar with 2.4RR attachments would select this answer.

C. This answer is incorrect because Attachment 1 does not provide guidance to trip RRMG B.

This answer is plausible because Attachment 1 titled Trip of RR Pump(s). Attachment 1 would be required to be entered following the trip of RRMG B. The candidate that correctly identifies RR Runback logic and is not familiar with 2.4RR attachments and thinks tripping RRMG B IAW Attachment 1 would select this answer.

Technical Reference(s):

Procedure 2.2.68, Reactor Recirculation System Operations, Rev 77.

Procedure 2.4MC-RF, Condensate and Feedwater Abnormal, Rev 14 Procedure 2.4RR, Reactor Recirculation Abnormal, Rev 40.

Proposed references to be provided to applicants during examination: None Learning Objective:

4. Describe the interrelationships between the Reactor Recirculation system or the Recirculation Flow Control system and the following:
j. Reactor water level/pressure
10. Describe the Reactor Recirculation system and/or Recirculation Flow Control system design features and/or interlocks that provide for the following:
l. Recirculation pump runback
13. Given plant conditions, determine if any of the following should occur:
c. Recirculation pump runback to the 45% speed limiter Question Source:

Bank #

Modified Bank #

New X Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 41.5 45.6 Difficulty: 3 SRO Only - 10CFR55.43 b (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

29

Assessing plant conditions and selecting a procedure with which to proceed. The SRO assesses plant conditions and determines whether RPV level will or will not recover and make priority decisions about responding to a plant transient.

30

31 32 33 34 Examination Outline Cross-

Reference:

Level SRO Tier # 2 Group # 2 K/A # 271000G2.4.20 Importance Rating 4.3 271000 Offgas System 2.4.20 Knowledge of operational implications of EOP warnings, cautions, and notes.

(CFR: 41.10 / 43.5 / 45.13)

Question: 93 The following conditions exist following an earthquake:

  • SRVs cannot to be opened for pressure control.
  • No means of RPV injection are available.
  • RPV water level is -150 inches corrected fuel zone and steady.
  • One MSL remains open.
  • RPV pressure is 950 psig.
  • Torus water level is 9.5 feet and lowering 0.2 feet/minute.

(1) What is the required CRS response?

(2) What CAUTION is of concern when carrying out this response?

A. (1) Transition to EOP 2A and direct Emergency Depressurization.

(2) When placing SJAEs in service, caution of sending personnel through potentially high radiation areas must be addressed.

B. (1) Transition to EOP 2A and direct Emergency Depressurization.

(2) When placing SJAEs in service, caution of forcing the shutter slide on the breaker could cause damage to the breaker preventing closure.

C. (1) Transition to EOP 2A and direct Steam Cooling.

(2) When placing SJAEs in service, caution of sending personnel through potentially high radiation areas must be addressed.

D. (1) Transition to EOP 2A and direct Steam Cooling.

(2) When placing SJAEs in service, caution of forcing the shutter slide on the breaker could cause damage to the breaker preventing closure.

Answer:

A. (1) Transition to EOP 2A and direct Emergency Depressurization.

(2) When placing SJAEs in service, caution of sending personnel through potentially high radiation areas must be addressed.

35

Explanation:

The CRS must direct Emergency Depressurization due to EOP 3A direction on torus water level being below 9.6' of water. When the downcomers become uncovered, the drywell and torus pressure suppression function is lost because any possible future steam leakage into the drywell will not be directed below the torus water level and condensed. The result could be primary containment failure due to overpressure. Emergency depressurization is the primary action to take with the given conditions. To accomplish the ED, the steam jet air ejectors are going to be used. The CRS is knowledgeable of all the plant conditions and can prioritize which system(s) is/are to be used and the priority placed on the order to use the systems. EOPs allow using one or all of the listed systems and the CRS directs placing as many in service that is required to perform the task of lowering RPV pressure less than 50 psig below tours space pressure. The CRS must be knowledgeable about the CAUTIONS applicable to which system to use and prioritize based upon procedure guidance and plant conditions. At this point ED is complete and the RPV will not again pressurize.

Distracters:

B, This answer is incorrect because the CAUTION of concern is not correct. There are no breaker trip actions to take when placing the SJAEs in service. This caution is based upon placing the AOG third stage SJAEs in service which is contained in the same procedure.

The guidance is contained in the section of EOP procedure 5.8.2 which is knowledge the SRO is required to know but is not general RO knowledge based upon the placement being well inside the procedure guidance. This answer is plausible because the action to transfer to EOP 2A and emergency depressurize is required. The second part is plausible because it is a caution that is in the procedure for placing a like system in service.

C. This answer is incorrect because the requirement to steam cool is not met. RPV level is steady and as long as level can be maintained above -183 inches with no means of injection available. The second part of the answer is correct because there is a potential of sending personnel through a high radiation field when performing the task. It is the SROs responsibility to determine if personnel safety risk merits performing the task. The SRO has determined it safe as inferred in the question stem. This answer is plausible because the correct caution is stated D. This answer is incorrect because both parts of the answer are incorrect. The correct action is to emergency depressurize and the correct caution to regard is sending the personnel through the potential high radiation field. This answer is plausible because the actions given are correct if given different circumstances. The same procedure is used for using the AOG third stage SJAEs in service and this system can be used as an emergency depressurization system. The SRO must know detailed procedure guidance to realize there are no breaker trips with the slide shutter when using the SJAEs.

36

Technical Reference(s):

PSTG/SATG AMP-TBD00 rev 8 APDX B EOP 5.8.2, RPV Depressurization Systems (Table 2), Rev. 40 Proposed references to be provided to applicants during examination: NONE Learning Objective:

6. Identify any EOP support procedure addressed in Flowchart 2A and apply any associated special operating instructions or cautions.
7. Given plant conditions and EOP flowchart 2A, EMERGENCY RPV DEPRESSURIZATION / STEAM COOLING, determine required actions.

Question Source:

Bank #

Modified Bank #

New X Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41.10 55.43.5 55.45.13 LOD 3 SRO Only - 10CFR55.43 b (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Knowledge of diagnostic steps and decision points in the emergency operating procedures (EOP) that involve transitions to event specific sub-procedures or emergency contingency procedures.

37

38 39 40 41 42 43 44 Examination Outline Cross-

Reference:

Level SRO Tier # 3 Group # 1 K/A # 2.1.4 Importance Rating 3.8 2.1.4 Knowledge of individual licensed operator responsibilities related to shift staffing, such as medical requirements, no-solo operation, maintenance of active license status, 10CFR55, etc.

Question: 94 Which of the following completes the statements below regarding SRO License requirements per CNS procedures?

A MINIMUM of ____(1)____ 12-hour shifts under instruction is required to allow taking the watch after a four month absence from watchstanding.

A Special Prescription Respirator Glasses Verification is required IAW ____(2)____?

A. (1) four (2) NTP8.1 (Administration of Licensed Operator Medical Examination Program)

B. (1) four (2) 2.0.7 (Licensed Operator Active/Reactivation/Medical Status Maintenance Program)

C. (1) five (2) NTP8.1 (Administration of Licensed Operator Medical Examination Program)

D. (1) five (2) 2.0.7 (Licensed Operator Active/Reactivation/Medical Status Maintenance Program)

Answer:

B. (1) four (2) 2.0.7 (Licensed Operator Active/Reactivation/Medical Status Maintenance Program)

Explanation:

Requires knowledge that four 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> shift watches are required to support license reactivation.

In order to maintain a license active, the operator is required to perform five twelve hour shifts (seven 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> shifts) per calendar quarter. With the license away from watchstanding duties for four months the license is no longer active. The license must be reactivated which requires four 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> shift watches under instruction. It is plausible to choose five since this is required to support license maintenance (five 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> shifts). The Special Prescription Respirator Glasses Verification is contained on the SRO On-Shift Time & Reactivation Attachments in procedure 2.0.7. If a change in medical condition were to occur, Procedure 2.0.7 directs licensee to complete Attachment 2 of Procedure NTP8.1. It is plausible to choose procedure NTP8.1 due 45

being titled Medical Status Maintenance Program and addressing prescription medication changes..

Distracters:

A. This answer is incorrect because NTP8.1 does not contain the Special Prescription Respirator Glasses Verification. This choice is plausible because this procedure is required to support change in licensed operator medical status. The candidate who correctly identifies the number of watches required to reactivate and confuses Respirator Glasses Verification with a change in license medical status would select this answer.

C. This answer is incorrect because the number of shifts stated is incorrect and NTP8.1 does not contain the Special Prescription Respirator Glasses Verification guidance. This choice is plausible due 5 shifts being required for initial license activation and quarterly license maintenance and this procedure being required to support change in licensed operator medical status. The candidate who confuses the number of watches required to activate/maintain vs. reactivate and confuses Respirator Glasses Verification with a change in license medical status would select this answer.

D. This answer is incorrect because the number of shifts required is incorrect. This choice is plausible because 5 shifts is required for initial license activation. The candidate who confuses the number of watches required to activate/maintain vs. reactivate and correctly identifies the procedure requiring Special Prescription Respirator Glasses Verification would select this answer.

Technical Reference(s):

Procedure 2.0.7 (Licensed Operator Active/Reactivation/Medical Status Maintenance Program) Rev. 09 Procedure NTP8.1 (Administration of Licensed Operator Medical Examination Program), Rev. 17 Proposed references to be provided to applicants during examination: None Learning Objective:

INT00705130010100 Given a set of plant conditions, recognize non-compliance with a Chapter 5.0 Requirement.

Question Source:

Bank #

Modified Bank #

New X Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41.10 55.43.2 Difficulty: 4 SRO Only - 10CFR55.43 b (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Requires knowledge of NRC license maintenance requirements and selection of procedure requiring Special Prescription Respirator Glasses Verification.

46

47 48 Examination Outline Cross-

Reference:

Level SRO Tier # 3 Group # 1 K/A # 2.1.41 Importance Rating 3.7 2.1.41 Knowledge of the refueling process.

Question: 95 Which activity REQUIRES Refuel Floor Supervisor permission during refueling operations in Mode 5?

A. Allowing under vessel access.

B. Allowing access to the refuel floor.

C. Re-commencing fuel handling operations.

D. Using greater than 50 gallons of demineralized water on the refuel floor.

Answer:

C. Re-commencing fuel handling operations.

Explanation:

Requires knowledge of Refuel Floor SRO responsibilities during refueling operations. Refuel Floor Supervisor permission is required to re-commence fuel handling operations IAW Attachment 4 (Reset Checklist) which shall be used each time the normal fuel handling process is stopped/interrupted. This includes, but is not limited to, Shift Turnover, Fuel Mover/Spotter mid-shift role change, or following a distraction which interrupts the normal fuel handling process flow. Putting the applicable procedure in the stem would eliminate under vessel access plausibility due to title being Fuel Handling - Refueling Platform.

Distracters:

A. This answer is incorrect because Refuel Floor Supervisor permission is not required to allow access to the under vessel area. This answer is plausible because under vessel area gets posted to prohibit access without Shift Manager's permission. The candidate who confuses access permission authority would select this choice.

B. This answer is incorrect because Refuel Floor Supervisor permission is not required to allow access to the refuel floor. This choice is plausible due to the Refuel floor SRO permission is required to access the fuel handling area - the fuel handling area is located within the refueling floor. The candidate who confuses refuel floor with fuel handling area would choose this answer.

D. This answer is incorrect because Refuel Floor Supervisor permission is not required to use greater than 50 gallons of demineralized water on the refuel floor. This choice is plausible due to the Refuel floor SRO is required to brief available refueling floor personnel on limiting demineralized water usage and requirement to notify Control Room if using > 50 gallons demineralized water each shift. The candidate who confuses briefing vs. giving permission would choose this answer.

Technical Reference(s):

Procedure 2.1.20.3, RPV Refueling Preparation (Wet Lift of Dryer and Separator), Rev. 45 49

Procedure 10.25 (Refueling - Core Unload, Reload, and Shuffle), Rev. 59 Procedure 2.2.31 (Fuel Handling - Refueling Platform), Rev. 47 Proposed references to be provided to applicants during examination: None Learning Objective:

INT0231002001160A Identify the administrative duties and responsibilities of the each of the following: Refueling Floor Supervisor Question Source:

Bank #

Modified Bank #

New X Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41.2 55.41.10 55.43.6 55.45.13 Difficulty: 3 10 CFR 55.43(b)(7) - Fuel handling facilities and procedures.

Requires knowledge of Refuel floor SRO responsibilities.

50

51 52 53 54 Examination Outline Cross-

Reference:

Level SRO Tier # 3 Group # 2 K/A # 2.2.11 Importance Rating 3.3 2.2.11 Knowledge of the process for controlling temporary design changes.

Question: 96 The plant is operating at power in Mode 1.

What is the MAXIMUM time a Temporary Alteration In Support of Maintenance (TASM) can be installed on plant equipment WITHOUT performing a 10CFR50.59 Review IAW Procedure 3.4.4 (Temporary Configuration Change)?

A. 30 days B. 60 days C. 90 days D. 120 days ANSWER:

C. 90 days Explanation:

A temporary alteration is necessary to support maintenance if it makes the maintenance activity easier, or the maintenance activity has been planned to allow prompt restoration. TASMs have regulatory considerations specific to duration under 10CFR50.59. Engineering Procedure 3.4.4, Temporary Configuration Change, Attachment 7, Step 4.1.1 and 4.1.2 describe the requirements for a 10CFR50.59 review prior to installation if it is expected to be in place > 90 days, or if after installation, it is going to be installed > 90 days. The procedure states that if a TASM that was installed and originally not expected to exceed 90 day that a 50.59 review should be performed. Procedure 0-EN-HU-106, Procedure and Work Instruction Use and Adherence, Step 3.12 defines should and states: Should - Denotes strong recommendation and indicates an action that is expected to be performed as described unless there is a compelling reason not to do so.

Distracters:

A. This answer is incorrect because the time listed is not the maximum time a TASM can be installed without a 10CFR50.59 Review being performed. The maximum time is 90 days as allowed by federal regulations. This answer is plausible because some non-emergency events require NRC notification reports and the candidate may recall the 30 days without tying it to the TASM requirements.

B. This answer is incorrect because the time listed is not the maximum time a TASM can be installed without a 10CFR50.59 Review being performed. The maximum time is 90 days as allowed by federal regulations. This answer is plausible because configuration change affected documents must be processed within 60 days of CED installation and the candidate may remember that time frame.

55

D. This answer is incorrect because the time listed exceeds the maximum time a TASM can be installed without a 10CFR50.59 Review being performed. The maximum time is 90 days as allowed by federal regulations. This answer is plausible because there is a time limit of 120 days in Tech Specs for time shutdown which requires scram timing on startup prior to 40%

power. The candidate may recall the 120 days but be unsure of its source.

Technical Reference(s):

Procedure 3.4.4 (Temporary Configuration Change), Rev. 15.

Proposed references to be provided to applicants during examination: NONE Learning Objective:

INT0320109A0A010A Procedure 3.4.4, Temporary Configuration Change - Discuss the following as described in Engineering Procedure 3.4, Configuration Change Control: Temporary Configuration Changes (TCCs)

Question Source:

Bank #

Modified Bank #

New X Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41.10 55.43.3 55.45.13 Difficulty: 3 SRO Only - 10CFR55.43 b (3) Facility licensee procedure required to obtain authority for design and operating changes in the facility.

Requires knowledge of Administrative processes for temporary modifications/configuration changes.

SRO Task: 200001W0303 Approve Installation of a Plant Temporary Configuration Change (TCC) Order. The SRO is responsible for knowledge of the TCC process.

56

57 Examination Outline Cross-

Reference:

Level SRO Tier # 3 Group # 2 K/A # 2.2.19 Importance Rating 3.4 2.2.19 Knowledge of maintenance work order requirements.

Question: 97 Which of the following identifies an item required to be addressed by a SRO performing a Work Order Standard Plant Impact Statement IAW 0-CNS-WM-102 (Work Implementation and Closeout)?

A. Verification of required parts availability.

B. Verifying the Work Instructions are accurate.

C. Identification of single valve isolation requirements.

D. Identifying required walkdowns by Shop/Responsible Work Center.

Answer:

C. Identification of single valve isolation requirements.

Explanation: Requires knowledge of SRO/FIN SRO/WCCA/WCC Supervisor impact review of maintenance work orders. The Standard Plant Impact Statement addresses the following items:

1. Does the work activity affect SSCs identified in TSs?
2. Will the maintenance activity require the equipment to be declared inoperable/unavailable?
3. Is a power reduction required?
4. Any other special plant conditions required to perform this work activity?
5. Is there a potential to affect the Operability of systems, structures, or components required to be operable?
6. Does the maintenance activity create the potential for inadvertent actuations (RPS trip, turbine trip, ECCS actuation, SDC isolation, system discharge valve closure)?
7. Does the maintenance activity have an actual or potential reactivity impact? (CRD, recirc flow, feed flow, feed temperature, etc.)?
8. Does the work activity increase the potential for loss of off-site power?
9. Will single valve isolation be required?
10. Are contingency plans or compensatory actions necessary?

Distracters:

A. This answer is incorrect due to verification of required parts availability not being part of the impact review by operations. This answer is plausible due to SROs normally performing this item. Parts availability is the most common reason for removing work from the schedule and is highly scrutinized by Operations to maintain schedule stability and minimize unnecessary burden on the operators and site workers. The candidate who recognizes the impact of parts availability on planning & scheduling during the operations impact review would select this answer.

B. This answer is incorrect due to verifying the accuracy of the Work Instructions not being part of the impact review by operations. This answer is plausible due to SROs often reviewing 58

the instructions for impact on Operations but is not required to be verified by the operations organization. The candidate who incorrectly believes reviewing Work Instructions is verifying their accuracy as part of the operations impact review would select this answer.

D. This answer is incorrect due to determining if a Shop/Responsible Work Center walkdown is required not being part of the impact review by operations. This answer is plausible due operators being required to walkdown clearance orders in support of work orders but is not required to be determined by the operations organization. The candidate who incorrectly believes determining clearance order walkdowns are part of the operations impact review would select this answer.

Technical Reference(s):

Procedure CNS-WM-102, Work Implementation and Closeout, Rev.01.

Proposed references to be provided to applicants during examination: None Learning Objective: SKL0110101001290A, 0.40, Work Control Program, Discuss the following as described in Administrative Procedure 0.40, Work Control Program: 1)

Precautions and limitations 2) Minor maintenance 3) Emergency MWR processing.

Question Source:

Bank #

Modified Bank #

New X Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41.10 55.43.5 55.45.13 Difficulty: 3 SRO Only - 10CFR55.43 b (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Knowledge of administrative procedures that specify implementation of plant normal procedures.

59

60 61 62 63 Examination Outline Cross-

Reference:

Level SRO Tier # 3 Group # 3 K/A # 2.3.6 Importance Rating 3.8 2.3.6 Ability to approve release permits.

Question: 98 Waste Sample Tank A is required to be discharged.

What is the MINIMUM recirculation time which allows approval of the release by the Shift Manager IAW Procedure 8.8.11 (Liquid Radioactive Waste Discharge Authorization)?

A. 1/2 hour prior to being sampled.

B. 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> prior to being sampled.

C. 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> prior to being sampled.

D. 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior to being sampled.

Answer:

B. 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> prior to being sampled.

Explanation:

The following items are verified by the Shift Manager prior to approving a liquid RW discharge:

  • Tank recirculation time > 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> prior to being sampled
  • Proper dilution flow (159,000 gpm)
  • Tank volume sampled matches volume to be released Distracters:

A. This answer is incorrect due to tank recirculation not meeting the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> procedural requirement. This answer is plausible due to 30 minutes is the normal tank recirculation time prior to sampling to support transfer vs. discharge. The candidate who confuses tank sample recirc time for transfer would select this answer.

C. This answer is incorrect due to tank recirculation being greater than the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> procedural minimum requirement. This answer is plausible due to 3 hrs satisfying the required recirulation time but is not the minimum. The candidate who does not recognize the minimum tank sample recirc time for discharge would select this answer.

D. This answer is incorrect due to tank recirculation being greater than the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> procedural minimum requirement. This answer is plausible due to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> satisfying the required recirulation time but is not the minimum and can be confused with Phase Separator decanting time limitations based upon turbidity results being > 10 FTU requiring stopping 64

decanting for at least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> before initiating another decant. The candidate who does not recognize the minimum tank sample recirc time for discharge would select this answer.

Technical Reference(s):

Procedure 8.8.11 Liquid Radioactive Waste Discharge Authorization, Rev. 32.

Procedure 2.5.1.6 (RW Low Conductivity Liquid Waste Sample Tank Fluid Transfer), Rev. 43 Procedure 2.5.1.7 (RWCU Phase Separator Tank Transfer), Rev. 27 Proposed references to be provided to applicants during examination: NONE Learning Objective:

INT0320115B0B0200 State who, by title, is required to grant permission to perform a liquid discharge with the liquid radwaste monitor inoperable.

Question Source:

Bank #

Modified Bank #

New X Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41.13 55.43.4 55.45.10 Difficulty: 3 SRO Only - 10CFR55.43 b (4) Radiation hazard that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.

Requires knowledge of the process for liquid release approval.

65

66 67 Examination Outline Cross-

Reference:

Level SRO Tier # 3 Group # 4 K/A # 2.4.29 Importance Rating 4.4 2.4.29 Knowledge of the Emergency Plan.

Question: 99 Given the following:

  • At 1200 the threshold for a NOUE is exceeded.
  • At 1210 the Emergency Director declared classification of a NOUE.
  • At 1215 the threshold for an ALERT is exceeded.
  • At 1220 the Emergency Director declared classification of an ALERT.

What is the LATEST time that State/Local agency notifications of the NOUE classification is required to be performed IAW EPIP 5.7.6 (Notification)?

A. 1215 B. 1225 C. 1230 D. 1235 Answer:

B. 1225 Explanation:

Requires knowledge of E-Plan notification requirements. Initial notification to State/local agencies of E-plan classification is required to be performed within 15 minutes of emergency declaration. There is a common misapplication of start times when staggering (subsequent)

EALs are met. The SRO must keep separate times running for each EAL as it is entered and not reset an earlier time due to subsequent EALs. This is the responsibility of the Emergency Director (SRO) until relieved by another Emergency Director.

Distracters:

A. This answer is incorrect due to not being the latest time requiring notification. This answer is plausible due to being 15 minutes from exceeding an EAL threshold. The candidate who cannot differentiate between the time to notify State/Local from exceeding EAL threshold vs. Declaration would choose this answer.

C. This answer is incorrect due exceeding the latest time requiring notification. This answer is plausible due to being 15 hour1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> from exceeding an Alert EAL threshold. The candidate who cannot differentiate between the time to notify State/Local from Alert EAL threshold vs.

NOUE Declaration would choose this answer.

D. This answer is incorrect due exceeding the latest time requiring notification. This answer is plausible due to being 15 minutes from the ALERT declaration. The candidate who cannot 68

differentiate between the time to notify State/Local from Alert EAL Declaration vs. NOUE Declaration would choose this answer.

Technical Reference(s):

EPIP 5.7.2 (Emergency Director EPIP), Rev. 32 Emergency Procedure 5.7.6 (Notification), Rev 24.

Proposed references to be provided to applicants during examination: NONE Learning Objective:

GEN0030401B0B030B Emergency Notifications and Communications Systems: State the time requirements for initial and/or follow-up notifications to offsite agencies.

Question Source:

Bank #

Modified Bank # 6082 New Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41.10 55.43.5 55.45.13 Difficulty: 2 SRO Only - 10CFR55.43 b (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

The SRO is responsible for task 344022O0303, Direct Emergency Response as Emergency Director (Emergency Plan) 69

70 71 72 Question 100 and associated references redacted due to SUNSI considerations

SRO References INFORMATION ONLY INFORMATION ONLY

INFORMATION ONLY INFORMATION ONLY

INFORMATION ONLY INFORMATION ONLY

INFORMATION ONLY INFORMATION ONLY

INFORMATION ONLY INFORMATION ONLY

AG1.1 1 2 3 4 5 DEF AS1.1 1 2 3 4 5 DEF AA1.1 1 2 3 4 5 DEF AU1.1 1 2 3 4 5 DEF Any valid gaseous monitor reading > Table A-1 column GE Any valid gaseous monitor reading > Table A-1 column SAE Any valid gaseous monitor reading > Table A-1 column Any valid gaseous monitor reading > Table A-1 column UE for 15 min. (Note 1) for 15 min. (Note 1) Alert for 15 min. (Note 2) for 60 min. (Note 2)

AG1.2 1 2 3 4 5 DEF AS1.2 1 2 3 4 5 DEF AA1.2 1 2 3 4 5 DEF AU1.2 1 2 3 4 5 DEF Dose assessment using actual meteorology indicates doses Dose assessment using actual meteorology indicates doses Any valid liquid effluent monitor reading > Table A-1 column Any valid liquid effluent monitor reading > Table A-1 column 1 > 1 Rem TEDE or > 5 Rem thyroid CDE at or beyond the site boundary

> 0.1 Rem TEDE or > 0.5 Rem thyroid CDE at or beyond the site boundary Alert for 15 min. (Note 2) UE for 60 min. (Note 2)

Offsite Rad Conditions AG1.3 1 2 3 4 5 DEF AS1.3 1 2 3 4 5 DEF AA1.3 1 2 3 4 5 DEF AU1.3 1 2 3 4 5 DEF Field survey results indicate closed window dose rates Field survey indicates closed window dose rate > 0.1 Rem/hr Confirmed sample analyses for gaseous or liquid releases Confirmed sample analyses for gaseous or liquid releases

> 1 Rem/hr expected to continue for 60 min. at or beyond that is expected to continue for 60 min. at or beyond the site indicate concentrations or release rates > 200 x ODAM limits indicate concentrations or release rates > 2 x ODAM the site boundary (Note 1) boundary (Note 1) for 15 min. (Note 2) limits for 60 min. (Note 2)

OR OR Analyses of field survey samples indicate thyroid CDE Field survey sample analysis indicates thyroid CDE > 0.5 Rem

> 5 Rem for 1 hr of inhalation at or beyond the site boundary for 1 hr of inhalation at or beyond the site boundary

m. AA2.1 1 2 3 4 5 DEF AU2.1 1 2 3 4 5 DEF Table A-1 Effluent Monitor Classification Thresholds e Damage to irradiated fuel OR loss of water level (uncovering Unplanned water level drop in the reactor cavity or spent fuel GE SAE ALERT UE irradiated fuel outside the RPV) that causes EITHER of the pool as indicated by any of the following:

Monitor nt for 15 min. for 15 min. for 15 min. for 60 min. following:

  • LI-86 (calibrated to 1001' elev.)

2 Valid RMA-RA-1 Fuel Pool Area Rad reading > 50 R/hr

  • Spent fuel pool low level alarm OR
  • Visual observation ERP 3.50E+08 µCi/sec 3.50E+07 µCi/sec 2.80E+06 µCi/sec 2.24E+05 µCi/sec AND Valid RMP-RM-452 A-D Rx Bldg Vent Exhaust Plenum Valid area radiation monitor reading rise on RMA-RA-1 or Onsite Rad Hi-Hi alarm Conditions RMA-RA-2 Rx Bldg Vent 3.50E+07 µCi/sec 3.50E+06 µCi/sec 5.45E+05 µCi/sec 8.48E+04 µCi/sec GASEOUS Spent Fuel Pool Turb Bldg Vent AA2.2 1 2 3 4 5 DEF AU2.2 1 2 3 4 5 DEF 3.50E+07 µCi/sec 3.50E+06 µCi/sec 5.62E+05 µCi/sec 9.02E+04 µCi/sec Events A water level drop in the reactor refueling cavity or spent fuel Unplanned valid area radiation monitor reading or survey RW / ARW Bldg Vent 3.50E+07 µCi/sec 3.50E+06 µCi/sec 5.64E+05 µCi/sec 9.08E+04 µCi/sec pool that will result in irradiated fuel becoming uncovered results rise by a factor of 1,000 over normal levels*
  • Normal levels can be considered as the highest reading in the past 24 The lesser of *: The lesser of *: hours excluding the current peak value 200 x calculated 2 x calculated LIQUID Rad Waste Effluent ----- ----- alarm values alarm values AA3.1 1 2 3 4 5 DEF 3 OR monitor upscale OR monitor upscale Dose rates > 15 mRem/hr in EITHER of the following areas requiring continuous occupancy to maintain plant safety MCR/CAS Service Water Effluent ----- ----- 4.80E-04 µCi/cc 4.80E-06 µCi/cc functions:

Rad Main Control Room (RM-RA-20)

OR

  • with effluent discharge not isolated CAS

N for possible escalation above the Unusual Event due inoperable, or out of service, before the event d as it will have no adverse impact on the ability of the ond that already allowed by Technical Specifications at likely longer than the specified time interval. If off-ed the backfeed, its power to the safety-related buses

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