ML13156A032
| ML13156A032 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 06/05/2013 |
| From: | NRC/RGN-II |
| To: | Florida Power & Light Co |
| References | |
| Download: ML13156A032 (72) | |
Text
L<MF Examination Outline Cross-reference:
1 Group#
1 KIA#
015 2.4.8 Importance Rating 4.5 Emergency Procedures I Plan: Knowledge of how abnormal operating procedures are used in conjunction with EOPs.
Proposed Question:
SRO Question # 76 Plant conditions:
Unit 3 is operating at 100% power.
o The crew has entered 3-ONOP-041.1, Reactor Coolant Pump Off Normal, due to problems with 3B RCP seal package.
Total #1 seal flow is approximately 8.4 GPM.
Seal DP is lowering slowly.
Seal Inlet and Outlet temperatures are rising slowly.
Which ONE of the following describes the procedural action required for this condition?
A.
Initiate a plant shutdown to have the RCP secured within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
B.
Monitor 3B RCP using Enclosure 1, Number 1 seal leakoff, and refer to NOP 041.01 B, 3B Reactor Coolant Pump Operations, prior to determining if reactor trip is required.
C.
Trip the reactor, perform immediate actions, trip 3B RCP, then isolate seal return. No additional actions of 3-ONOP-041.1 will be performed while in 3-EOP-E-0.
D.
Trip the reactor, perform immediate actions, trip 3B RCP, then continue with 3-ONOP-041.1 in parallel wfth 3-EOP-E-0 until directed to transition.
Proposed Answer:
C Explanation (Optional):
A.
Incorrect. The conditions presented are a failure of #1 seal. This option describes the ONOP actions taken for failure of #2 or #3 seal B.
Incorrect. This option describes actions taken for seal degradation that does not yet require a reactor trip lAW ONOP-041.1
C.
CORRECT D.
Incorrect. Correct with the exception of the performance of entire ONOP-041.1.
In this case, only the required steps are performed as listed in Option C 3-ONOP-041.1 (Foldout page Rev6 11/8/11)
Technical Reference(s):
0-ADM-211 (p 17; rev 2 (Attach if not previously provided) 9/20/11)
Proposed References to be provided to applicants during examination:
N Learning Objective:
(As available)
Question Source:
Bank #
WTSI 95748 Modified Bank #
(Note changes or attach parent)
New Question History:
Last NRC Exam:
2011 Ginna Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 55.43 5
Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Comments:
KA is matched because it is related to an RCP malfunction and examines how ONOP-041.1 would be performed in conjunction with EOPs for an RCP seal failure.
SRO criteria is met because use of AOPs in conjunction with EOPs is at the US/SM discretion, and this item does not ask an overall strategy or entry to EOP/AOP, but usage of both as defined within the question
Examination Outline Cross-reference:
I Group#
1 KIA#
022 2.4.20 Importance Rating 4.3 Emergency Procedures! Plan: Knowledge of operational implications of EOP warnings, cautions, and notes.
Proposed Question:
SRO Question # 77 Plant conditions:
Unit 3 is in Mode 3.
The following alarms are received:
A 9/3, PZR CONTROL HI/LO LEVEL G 1/2, CHARGING PUMP HI SPEED Charging flow indication is lowering on Fl-3-122A.
Which ONE of the following describes the event in progress, and the procedural actions required to mitigate the event?
A.
Air intrusion into the CVCS; Enter 3-ONOP-047.1, Loss of Charging flow in Modes lthrough 4 to vent the Charging System.
B.
Air intrusion into the CVCS; Enter 3-ONOP-046.4, Malfunction of Boron Concentration Control System, and vent the Charging System.
C.
Leak in the CVCS regenerative heat exchanger; Loss of Charging flow in Modes lthrough 4 D.
Leak in the CVCS regenerative heat exchanger; Enter 3-ONOP-046.4, Malfunction of Boron Concentration Control System, and vent the Charging System.
Proposed Answer:
A Explanation (Optional):
A.
Correct. See reference B.
Incorrect. Plausible because the 1st part is correct and also because Charging hasnt been lost like with a charging pump trip. ONOP-046.4 can also be entered from ONOP 047.1
C.
Incorrect. If the leak was in the RHX, charging flow would be rising, not lowering. 2nd part is correct.
D.
Incorrect. See explanation for B and C 3-ONOP-047.1 (p.3; Rev I Technical Reference(s):
BD-ONOP-047.1 (p4; Rev (Attach if not previously provided) 11/4/08)
Proposed References to be provided to applicants during examination:
None Learning Objective:
6902234, Objectives 1, 2, 7 (As available)
Question Source:
Bank #
Modified Bank #
(Note changes or attach parent)
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 55.43 5
Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Comments:
KA is matched because the item evaluates the implication of a note/caution in the ONOP related to Loss of RCS Makeup for air intrusion.
Question is SRO level because it meets the criteria of 1 OCFR55.43(b) item 5 for assessment of conditions and selection of procedures, it does not test overall AOP strategy, but tests procedure selection based on plant conditions specific to a malfunction of a system
Examination Outline Cross-reference:
1 Group#
1 KIA#
027 2.4.11 Importance Rating 4.2 Emergency Procedures I Plan: Knowledge of abnormal condition procedures.
Proposed Question:
SRO Question # 78 Plant conditions:
Unit 4 is operating at 100% power.
The following is observed:
Annunciator A 9/4, PZR LO LEVEL/HEATER OFF/LTDN SECURED, is received.
Letdown flow is 0 GPM.
All Pressurizer Heaters are OFF.
Pressurizer Level indicates 53.3% on all LI-4-459A, LI-4-460, and LI-4-461.
Pressurizer pressure is 2230 psig and lowering slowly Subsequently:
The crew has entered 4-ONOP-041.6, Pressurizer Level Control Malfunction.
I&C reports that the 4-459CX relay has failed.
Which ONE of the following describes the correct action for the restoration of Pressurizer Heaters, AND identifies the technical specification implication?
A.
Use the button on relay 4-459CX to operate pressurizer heaters; Plant shutdown is required in accordance with technical specifications.
B.
Remove channel 4-459 from the pressurizer level control circuit; Plant shutdown is required in accordance with technical specifications.
C.
Use the button on relay 4-459CX to operate pressurizer heaters; Plant shutdown is NOT required in accordance with technical specifications.
D.
Remove channel 4-459 from the pressurizer level control circuit; Plant shutdown is NOT required in accordance with technical specifications.
Proposed Answer:
A
Explanation (Optional):
A.
Correct. 1st part correct, 2nd part correct. According to 4-ONOP-041.6 (p5; Rev 12/4/08), a NOTE prior to Step 1 of the Subsequent Actions states that If Pressurizer Level Malfunction is a result of a failure of the 4-459CX or 4-460CX relays (as indicated by a loss of letdown flow with a loss of Pressurizer Heaters with no concurrent failure of Level Transmitters 4-459A, 4-460, 4-461), use 4ONOP003.6 Attachment 4, for 4-460CX failure, OR 4ONOP003.9 Attachment 4, for 4-459CX failure as guidance for establishing Letdown flow and Pressurizer Heaters. Consequently, 4-ONOP-003.9 should be used. A second Note states that If the button on relays 4-459CX or 4-460CX are used to restore Letdown flow and Pressurizer Heaters, comply with Tech Spec Action Statement 3.4.3 Action b. According to 4-ONOP-003.9, Attachment 4, the operator will need to press the button on Relay LC 459CX (Step 2), and direction is provided in Step 6 to Comply with the 6-hour Action b of Technical Specification 3.4.3, Pressurizer. According to Technical Specification, the ACTION for LCO 3.4.3.b requires the operator to be in at least HOT STANDBY with the Reactor Trip System breakers open within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
B.
Incorrect. 1st part wrong, 2nd part correct. This is incorrect because the wrong procedure has been selected to recover. This is plausible because according to 4-ONOP-041.6 Step 5.5 through 5.7, if the operator were to remain in the current ONOP, direction would be provided to restore both Letdown and Pressurizer Heaters.
However, because of the failure this direction will not be successful. The operator may not recognize this, and continue in the current ONOP.
C.
Incorrect. 1 st part wrong, 2nd part wrong. See A and B.
D.
Incorrect. 1 st part correct, 2nd part wrong. This is incorrect because the wrong Technical Specification ACTION has been chosen. See description for option A on TS action.
It is not blatantly obvious that shutdown will be required because failure of this relay makes the pressurizer inoperable (PZR Heaters) 4-ONOP-041.6 (p5; Rev 12/4/08) 4-ONOP-003.9, Attachment 4, Technical Reference(s):
(p17; Rev 8/6/07)
(Attach if not previously provided)
Technical Specification 3/4 4-3.b, pg 3/4 4-9 Proposed References to be provided to applicants during examination:
None 6910254 Objectives 4, 6 and 8 Learning Objective:
(As available)
Question Source:
Bank #
Modified Bank #
(Note changes or attach parent)
New X
Question History:
Last NRC Exam:
NA Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 55.43 5
Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Comments:
The KA is matched because the operator must demonstrate knowledge of abnormal condition procedures associated with a Pressurizer Pressure Control Malfunction (Pzr Heaters have tripped).
The question is at the Comprehension/Analysis (2DR) cognitive level because the operator must recall bits of information (ONOP Strategies, TS ACTIONS), and then relate this information to the current conditions and required procedural actions to answer the question correctly.
The question is SRO-ONLY because it cannot be answered solely by knowing system knowledge, immediate operator actions, plant parameters that require direct entry into EOPs, or knowing the purpose, overall sequence of events, or overall mitigating strategy of a procedure; AND requires the operator to assess plant conditions, recall the location of the proper strategy and then select a procedure or section of a procedure to mitigate, recover, or with which to proceed; including the demonstration of knowledge of when to implement attachments, including how to coordinate these items with procedure steps. Additionally, the question is SRO-ONLY because it cannot be answered solely by knowing 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> TSITRM action statements, the LCO/TRM information listed above the line, or by knowing the TS Safety Limits; AND requires the operator to apply the Required Actions in accordance with the rules of application.
NOTE: This question could also be connected to 10CFR43(b)(2).
Examination Outline Cross-reference:
1 Group#
1 KIA#
038 EA2.03 Importance Rating 4.6 Ability to determine or interpret the following as they apply to a SGTR: Which SIG is ruptured Proposed Question:
SRO Question # 79 Plant conditions:
Unit 3 at 100% power.
3A RCP trips, causing the reactor to trip.
While in 3-EOP-ES-0.1, Reactor Trip Response, the reports the following parameters:
3A SG level 50% narrow range and stable. AFW flow has been throttled to approximately 50 GPM.
3B SG level off-scale low narrow range. AFW flow is approximately 300 gpm.
3C SG level 65% narrow range and rising. AFW flow is isolated.
RCS pressure is 1860 psig and lowering.
PZR level is off-scale low.
SG pressures are all approximately 1000 psig and stable.
Which ONE of the following identifies the action required?
A.
Manually start SI Pumps as needed and immediately transition to 3-EOP-E-3, Steam Generator Tube Rupture, to isolate 3C Steam Generator.
B.
Initiate Safety Injection and return to 3-EOP-E-0, Reactor Trip or Safety Injection, then transition to E-3, Steam Generator Tube Rupture to isolate 3A Steam Generator.
C.
Initiate Safety Injection and return to 3-EOP-E-0, Reactor Trip or Safety Injection, then transition to E-3, Steam Generator Tube Rupture to isolate 3C Steam Generator.
D.
Manually start SI Pumps as needed and immediately transition to 3-EOP-E-3, Steam Generator Tube Rupture, to isolate 3A Steam Generator.
Proposed Answer:
C Explanation (Optional):
A.
Incorrect. This is incorrect because SI is not actuated if the crew is in ES-0.1, and SI actuation requires transition to E-0 before going to another EOP.(E-3). 3C SG is correct B.
Incorrect.
. First part is correct but 2nd part incorrect but plausible. Because of RCP trip, post trip SG level is higher in band with lower AFW flow due to reduced heat input to SG from the RCP, making it similar to a ruptured SG C.
Correct. PZR pressure and SG level require initiation of SI. The appropriate action once SI is initiated is to return to E-0 and then use E-3 to isolate 3C SG D.
Incorrect. Same reason as option A, and SG 3A is incorrect SG to isolate. See Option B
3-EOP-ES-0.1 (Foldout; Rev 5C Technical Reference(s):
3-EOP-E-0 (p.16; Rev 5 (Attach if not previously provided) 8/7/12)
Proposed References to be provided to applicants during examination:
None 6902321 Objective 7 Learning Objective:
6902323 Objective 10 (As available)
Question Source:
Bank #
Modified Bank #
(Note changes or attach parent)
New X
Question History:
Last NRC Exam:
NA Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 55.43 5
Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Comments:
The KA is matched because the operator must demonstrate the ability to determine or interpret which S/G is ruptured during a plant event in which the EOP network is being implemented. This is accomplished by requiring the SRO to decide whether or not enough
information is present to state that the 3C SG is ruptured given conflicting information on SG 3A level.
The question is at the Comprehension/Analysis (3PEO) cognitive level because the operator must recall bits of information (Foldout Page Items, minimum indications for SGTR), and then apply this information to a set of plant conditions to answer the question correctly.
The question is SRO-ONLY because it cannot be answered solely by knowing system knowledge, immediate operator actions, plant parameters that require direct entry into EOPs, or knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure; AND requires the operator to assess plant conditions and then select a procedure or section of a procedure to mitigate, recover, or with which to proceed. NOTE: This question does require the operator to make a selection based on plant parameters that require an indirect entry (i.e. from a Foldout Page of another EOP).
Examination Outline Cross-reference:
1 Group#
1 KIA#
058 AA2.01 Importance Rating 4.1 Ability to determine and interpret the following as they apply to the Loss of DC Power: That a loss of dc power has occurred; verification that substitute power sources have come on line Proposed Question:
SRO Question # 80 Plant conditions:
Unit 3 is operating at 100% power.
Unit 4 has tripped from 100% power.
The following indications are observed in the control room The left half of all alarming annunciators on Unit 4 are DARK Annunciator G4/1 ANNUNCIATOR POWER FAILURE is LIT MOV-4-1403 STM TO AFW PUMPS position indicating lights are DARK B AFW Trip and Throttle Valve position indicating lights are DARK The crew has completed 4-EOP-E-0, Reactor Trip or Safety Injection, and has now stabilized the plant using 4-EOP-ES-0.1, Reactor Trip Response.
Which ONE of the following (1) identifies the applicable procedure for this event and (2) identifies the MINIMUM Tech Spec required ACTION?
REFERENCE PROVIDED A.
(1) 4-ONOP-003.5, Loss of DC Busses 4D23 and 4D23A (4A) to verify that CVCS Blender valves, FCV-4-1 1 3B and FCV-4-1 1 4A are closed (2) Place Unit 4 in Cold Shutdown within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
B.
(1) 4-ONOP-003.4, Loss of DC Bus 4D01 and 4DO1A (4B) to verify Inverter 4B is being supplied by CVT 4Y02A (2) Place Unit 4 in Cold Shutdown within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
C.
(1) 4-ONOP-003.5, Loss of DC Busses 4D23 and 4D23A (4A) to verify that CVCS Blender valves, FCV-4-113B and FCV-4-114A are closed (2) Place Unit 4 in Cold Shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
D.
(1) 4-ONOP-003.4, Loss of DC Bus 4D01 and 4DOIA (4B) to verify Inverter 4B is being supplied by CVT 4Y02A (2) Place Unit 4 in Cold Shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
Proposed Answer:
D Explanation (Optional):
A.
Incorrect. This is incorrect because the procedure is not associated with the DC Bus that has failed, and because it does not address the required procedure implementation at Unit 3. This is plausible because the operator may misdiagnose the event.
If so, according to 4-ONOP-003.5 Step 5.4, WHEN required by the Shift Manager THEN commence shutdown of Unit 4 to Mode 5, Cold Shutdown, using 4-GOP-305, HOT STANDBY TO COLD SHUTDOWN, within the time specified in Technical Specification 3.0.3.
B.
Incorrect. 1st part is correct. This is incorrect because it does not address the required shutdown time of unit 4 correctly.
C.
Incorrect. This is incorrect because the procedure is not associated with the DC Bus that has failed. This is plausible because the operator may misdiagnose the event.
If so, according to 4-ONOP-003.5 Step 5.4, WHEN required by the Shift Manager THEN commence shutdown of Unit 4 to Mode 5, Cold Shutdown, using 4-GOP-305, HOT STANDBY TO COLD SHUTDOWN, within the time specified in Technical Specification 3.0.3. TS 3.0.3 requires 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> D.
Correct. According to 4-ONOP-003.4, for a loss of DBOI/DBO1A the reactor/turbine will trip and the MCB Annunciators will be de-energized. According to 4-ONOP-003.5, for a loss of DB23/DB23A the reactor/turbine will trip and the MSIVs will go closed. Since the Annunciators will not be lost on a loss of DC Bus DB23/DB23A, the cause of this event is a loss of DC Bus DBO1/DBOIA, and the operator must perform 4-ONOP-003.4, Loss of DC Bus 4D01 and 4DOIA (4B), concurrent with 4-EOP-ES-0.1. According to 4-ONOP-003.4 Step 5.3, the operator will be directed to verify that the Inverter (4B or BS),
that was supplying 120V Vital Instrument Bus 4P08 before the loss of DC Bus 4D01 AND 4DO1A (4B), has transferred to CVT 4Y02A. Step 5.6 requires that Unit 4 commence shutdown of Unit 4 to Mode 5, Cold Shutdown using 4-GOP-305, HOT STANDBY TO COLD SHUTDOWN OR 4-EOP-ES-0.2, NATURAL CIRCULATION COOLDOWN, within the time specified in Technical Specification, 3.0.3.
4-ONOP-003.4 (p4-5; Rev 1 9/12/12)
Technical Specification LCO Technical Reference(s):
3.8.3.1 (p3/4 8-18-20; (Attach if not previously provided)
Amendment 138 and 133)
Proposed References to be provided to applicants during examination:
TS 3.8.3.1 6902253 Objective 4 Learning Objective:
(As available)
Question Source:
Bank #
Modified Bank #
(Note changes or attach parent)
New X
Question History:
Last NRC Exam:
NA Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 55.43 5
Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Comments:
The KA is matched because the operator must demonstrate the ability to determine and interpret that a loss of DBOI!DBO1A has occurred and that verification that substitute power sources have come on line has taken place. This is accomplished by requiring that the operator select the correct procedure initially for the appropriate DC Bus which provides direction for verifying that the associated 12OVAC Instrument Bus has transferred to its CVT.
This is the only situation for this event that meets the KA for checking alternate sources of power have come on-line.
The question is at the Comprehension/Analysis (3PEO) cognitive level because the operator must recall bits of information (What Bus has failed, TS requirements on Bus), and then apply this information to the overall integrated plant response and Technical Specifications, in order to select appropriate recovery procedures and answer the question correctly.
The question is SRO-ONLY because it cannot be answered solely by knowing system knowledge, immediate operator actions, plant parameters that require direct entry into EOPs, or knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure; AND requires the operator to assess plant conditions and then select a procedure or section of a procedure to mitigate, recover, or with which to proceed. The question is SRO ONLY because it cannot be answered solely by knowing 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> TS/TRM action statements, the LCO information listed above the line, or by knowing the TS Safety Limits; AND requires the operator to apply the Required Actions in accordance with the rules of application.
NOTE: This question could be connected to 10CFR55.43(b)(2).
Examination Outline Cross-reference:
I Group#
1 KIA#
062 AA2.03 Importance Rating 2.9 Ability to determine and interpret the following as they apply to the Loss of Nuclear Service Water: The valve lineups necessary to restart the SWS while bypassing the portion of the system causing the abnormal condition Proposed Question:
SRO Question # 81 Plant conditions:
Unit 4 is operating at 25% power.
TPCW/ICW Isolation Valves, POV-4-4882 and 4883 have failed CLOSED.
The crew has entered 4-ONOP-019, intake Cooling Malfunction.
POV-4-4882 has been opened manually using the local handwheel.
POV-4-4883 will NOT open.
Which ONE of the following identifies the status of TPCW cooling, and identifies the technical specification implications of this event, if any?
A.
TPCW Cooling has been restored; Declare the 4A ICW header inoperable ONLY, enter the action statement for LCO 3.7.3, Intake Cooling Water System, and restore the inoperable header within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
B.
TPCW Cooling has been restored; Declare the 4A ICW header AND 4A train of Safety Injection Automatic Actuation Logic inoperable and enter action for LCD 3.3.2, Engineered Safety Feature Actuation System, and restore the inoperable train within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
C.
TPCW Cooling has NOT been restored; Declare the 4A ICW header AND 4A train of Safety Injection Automatic Actuation Logic inoperable and enter action for LCD 3.3.2, Engineered Safety Feature Actuation System, and restore the inoperable train within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
D.
TPCW Cooling has NOT been restored; Declare the 4A ICW header inoperable, enter the action statement for LCD 3.7.3, Intake Cooling Water System, and restore the inoperable header within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
Proposed Answer:
A
Explanation (Optional):
A.
Correct. According to 4-ONOP-019 (p9; Rev 119!O1C) a Caution proceeds Step 7 which states If POV-4-4882 or POV-4-4883 must be locally opened using the handwheel, then the Technical Specification 72-hour action statement for an inoperable ICW header is required to be entered. The operator will be directed at Step 7 to check both valves open.
If they are not, as is the case, the RNO will need to be performed.
The RNO directs the operator to locally open one ICW to TPCW Heat Exchanger valve using the handwheel, and then states IF a valve was opened using the handwheel, THEN enter the Technical Specification 72-hour action statement for an inoperable ICW header.
B.
Incorrect. This is incorrect because although TPCW has been restored but the second half of the statement is incorrect because although the statement is true that TPCW is not TS, ICW is, and the action statement for lOW inoperable must be entered.
C.
Incorrect. This is incorrect because opening ONE valve of the pair is required to restore TPCW cooling. Therefore, cooling is restored. Also, TS action is taken due to ICW operability. This may not be obvious because TPCW is affected, although cooling is from lOW D.
Incorrect. This is incorrect because TPCW has been restored as discussed in C above. The applicant would consider TS 3.0.3 if they believed cooling was not restored if it took 2 valves to restore operability of ICW.
It is logical to assume if cooling is not restored then ICW is inoperable and apply TS 3.0.3 4-ONOP-019 (p9 Rev 119101C)
Technical Reference(s):
(Attach if not previously provided)
Proposed References to be provided to applicants during examination:
None 6902277 Objective 4 and 6 Learning Objective:
(As available)
Question Source:
Bank #
Modified Bank #
(Note changes or attach parent)
New X
Question History:
Last NRC Exam:
NA Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 55.43 5
Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Comments:
The KA is matched because the operator must demonstrate the ability to determine and interpret the valve lineups necessary (one valve or two) to restart the SWS while bypassing the portion of the system causing the abnormal condition (i.e. use of handwheel) as they apply to the Loss of Nuclear Service Water.
The question is at the Comprehension cognitive level because the operator must recall bits of information (how many valves opened per ONOP, Action required when valve opened using HW) to answer the question correctly.
The question is SRO-ONLY because it cannot be answered solely by knowing system knowledge, immediate operator actions, plant parameters that require direct entry into EOPs, or knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure; AND requires the operator to assess plant conditions and then select a procedure or section of a procedure, or recall a stratecw to mitigate, recover, or with which to proceed.
Additionally, The question is SRO-ONLY because it cannot be answered solely by knowing 1
hour TS/TRM action statements, the LCO!TRM information listed above the line, or by knowing the TS Safety Limits; AND requires the operator to apply the Required Actions and Surveillance Requirements in accordance with the rules of application.
NOTE: This Question could also be connected to IOCFR55.43(b)(2).
Examination Outline Cross-reference:
1 Group#
2 KIA#
003 2.1.19 Importance Rating 3.8 Conduct of Operations: Ability to use plant computers to evaluate system or component status.
Proposed Question:
SRO Question # 82 Plant conditions:
Unit 3 is at 97% power and stable.
A dropped rod has occurred.
The crew has entered 3-ONOP-028.3, Dropped RCC.
The current QPTR has been calculated to be 1.04 using the Excore NIS Current readings.
Which ONE of the following identifies (1) where QPTR is monitored in the control room, and (2) the technical specification required action and basis for the action required?
A.
DCS; Reduce power to 85% to reinstate the margin for uncertainty for FQ(Dh).
B.
DCS; Reduce power to 88% to reinstate the margin for uncertainty for FQ(Z).
C.
QSPDS; Reduce power to 88% to reinstate the margin for uncertainity for FQ(Dh).
D.
QSPDS; Reduce power to 85% to reinstate the margin for uncertainity for FQ(Z).
Proposed Answer:
B Explanation (Optional):
A.
Incorrect. jst part correct, 2nd part wrong. 85% is plausible because it represents a 12% reduction from current power level. Basis is incorrect as it identifies the wrong hot channel factor B.
Correct. 1 st part correct, 2nd part correct. see references C.
Incorrect. 1 st part correct, 2nd part wrong See A and D.
D.
Incorrect. 1st part wrong, 2nd part correct. This is incorrect because power level must be reduced to 88% not 85%. This is plausible because the operator may subtract the
penalty of LCD 3.1.2.4 ACTION a.2.a-b from the current power level rather than the Rated Thermal Power. Additionally, the correct TS basis is identified 3-ONOP-028.3 (p7; Rev 3/28/12) 3-OSP-059.10 (p9-13; Rev 0B)
Technical Reference(s):
Technical Specification LCD (Attach if not previously provided) 3.2.4 (p3!4 2-13; Amendment 137 and 132)
Proposed References to be provided to applicants during examination:
None 6910207 Objective 6 Learning Objective:
6902280 Objective 4 and 6 (As available)
Question Source:
Bank #
Modified Bank #
(Note changes or attach parent)
New X
Question History:
Last NRC Exam:
NA Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 55.43 2
Facility operating limitations in the technical specifications and their bases.
Comments:
The KA is matched because the operator must demonstrate the ability to use plant computers to evaluate system or component status during a dropped rod event. This is accomplished by requiring the operator to identify which computer is used for monitoring QPTR, a significant power distribution concern during the dropped rod mitigation process.
The question is at the Comprehension cognitive level because the operator must recall bits of information (where is QPTR read (memory), but apply the TS Action Statement (application) to answer the question correctly.
The question is SRO-ONLY because it cannot be answered solely by knowing 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> TSITRM action statements, the LCO information listed above the line, or by knowing the TS Safety Limits; AND requires the operator to identify methods that are acceptable to be used to comply with the Required Actions and Surveillance Requirements; and to apply the Technical Specification Action requirements within the rules of usage.
Examination Outline Cross-reference:
1 Group#
2 K/A #
028 2.4.30 Importance Rating 4.1 Emergency Procedures I Plan; Knowledge of events related to system operation / status that must be reported to internal organizations or external agencies, such as the state, the NRC, or the transmission system operator.
Proposed Question:
SRO Question # 83 Plant conditions:
Unit 3 is operating at 100% power.
Pzr Level instrument LT-460 is failed HIGH and has been removed from service.
Which ONE of the following events will subsequently result in the need for the Shift Manager to notify the NRC Operations Center, AND identifies the time frame within which this notification must be made?
A.
A turbine runback occurs; 1 Hour B.
A turbine runback occurs; 4 Hours C.
Pressurizer Level Channel LT-461 fails LOW; 4 Hours D.
Pressurizer Level Channel LT-461 fails LOW; 1 Hour Proposed Answer:
C Explanation (Optional):
A.
Incorrect. 1st part wrong, 2nd part wrong. See B and D.
B.
Incorrect. 1 st part wrong, 2nd part correct. This is incorrect because according to 0-ADM-1 15 Enclosure 1 (p1 0; Rev 10/26/11), Turbine runbacks are not part of RPS and therefore not reportable. This is plausible because the operator may not be aware of this interpretation, and incorrectly believe that the Turbine Runback does constitute an
actuation of the RPS.
If so, according to Enclosure 1, An event that results in actuation of the Reactor Protection System (RPS) when the reactor is critical requires a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> notification.
C.
Correct. 1 st part correct, 2nd part correct. According to Technical Specification LCO 3.3.1, Table 3.3-1, Functional Unit 9 (p3!4 3-2; Amendment 140-1 35), a minimum of two Pressurizer Level Channels must be OPERABLE in Mode 1. Since this is not the case, and there is no ACTION addressing this issue, TS 3.0.3, which will require a plant shutdown, will apply. According to 0-ADM-1 15 (p9; Rev 10/26/11), IF an event meets criteria listed in Enclosures 1, 2, or 3 THEN perform Attachment 1. According to 0-ADM-1 15 (p25; Rev 10/26/11), Attachment 1, Step 1.a.2, directs the SM to Notify the NRCOC using the ENS telephone. According to 0-ADM-115 (plO; Rev 10/26/11), a plant event is reportable under 10 CFR 50.72(b)(2)(i) if an initiation of any nuclear plant shutdown required by Technical Specifications occurs. Initiation of any Nuclear Plant Shutdown is defined as the performance of any action to start reducing reactor power to achieve an operational condition or mode that requires the reactor to be subcritical as a result of a Tech Spec requirement. This includes any means of power reduction, such as control rod insertion, boration, or turbine load reduction; and it cites as an example:
(1) Exceeding an LCO Action Statement, (2) Entry into Technical Specification 3.0.3, or (3) a Safety Limit violation. Consequently, this event will require NRC Notification. states that this notification must be made within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
D.
Incorrect. 1 st part correct, 2nd part wrong. This is incorrect because according to 0-ADM-115 Enclosure 1 (plO; Rev 10/26/11) this notification requires a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> reporting time. This is plausible because at least one event (Deviation from Tech Specs allowed by 10 CFR 50.54(x) and 10 CFR 72.32(d)) that does not rise to the level of an emergency classification, requires a notification to be made to the NRC within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />; and the operator may incorrectly believe that this does, as well.
Technical Specification LCO 3.3.1, Table 3.3-1, Functional Unit 9 (p3!4 3-2; Amendment 140-135)
Technical Reference(s):
0-ADM-1 15 (p9-10 & 25; Rev (Attach if not previously provided) 10/26/11)
SD-063 (p50; Rev 9/10/11)
Proposed References to be provided to applicants during examination:
None 6902006 Objective 4 Learning Objective:
(As available)
Question Source:
Bank #
Modified Bank #
(Note changes or attach parent)
New X
Question History:
Last NRC Exam:
NA Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 55.43 1
Conditions and limitations in the facility license Comments:
The KA is matched because the operator must demonstrate knowledge of events related to system operation/status that must be reported to internal organizations or external agencies, such as the state, the NRC, or the transmission system operator. This is accomplished by providing the operator a set of initial conditions, and the requiring the operator to identify a subsequent event that will result in the need to call the NRCOC, and the time frame within which the NRCOC must be notified.
The question is at the Comprehension/Analysis (2DR) cognitive level because the operator must recall bits of information (reportable thresholds, time periods), and then evaluate two different plant events, one of which requires entry into TS 3.0.3, to determine if a reportable threshold has been exceeded, in order to answer the question correctly.
The question is SRO-ONLY because it requires the operator to possess knowledge of reportability requirements of regulations imposed by I OCFR5O.72, Reportability Requirements, which is a condition and requirement of the Facility License.
Examination Outline Cross-reference:
1 Group#
2 KIA#
037 AA2.07 Importance Rating 3.6 Ability to determine and interpret the following as they apply to the Steam Generator Tube Leak: Flowpath for dilution of ejector exhaust air Proposed Question:
SRO Question # 84 Plant conditions:
Unit 3 is operating at 100% power.
The crew has entered 3-ONOP-071.2, Steam Generator Tube Leakage.
The SJAE SPING reading is 2E-4 pCi/cc.
The R-15 reading is 1250 cpm.
The Turbine Operator reports that the Air Ejector In-Leakage is 5.6 SCFM.
Chemistry reports that the tube leakage is in the 3C Steam Generator.
The crew is evaluating actions of Attachment 1 of 3-ONOP-071.2, Guidelines For Continued Plant Operation With Primary-To-Secondary Leakage.
Which ONE of the following identifies the correct assessment of the leak rate based on these radiation monitors, AND any required action?
REFERENCES PROVIDED A.
Neither monitor has estimated a leakrate in excess of the Technical Specifications; Continue to monitor in accordance with Attachment 1 of 3-ONOP-071.2.
B.
One of the monitors has estimated a leakrate in excess of the Technical Specifications; Continue to monitor in accordance with Attachment 1 of 3-ONOP-071.2, until the limit can be confirmed by Chemistry.
C.
Both monitors have estimated a leakrate in excess of the Technical Specifications; Take action in accordance with Attachment 1 of 3-ONOP-071.2 to place the plant in HOT STANDBY within a maximum time of 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.
D.
Both monitors have estimated a leakrate in excess of the Technical Specifications; Take action in accordance with Attachment 1 of 3-ONOP-071.2 to place the plant in HOT STANDBY within a maximum time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
Proposed Answer:
C
Explanation (Optional):
A.
Incorrect. This is incorrect because both monitors are reading in excess of the TS Limits. This is plausible because the operator may miss the need to correct the approximation for SJAE in-leakage.
If so, the action pertaining to Attachment I is appropriate.
B.
Incorrect. This is incorrect because both monitors are reading in excess of the TS Limits. This is plausible because the operator may miss the need to correct the approximation for SJAE in-leakage on one instrument, or make a math error and conclude incorrectly.
If so, the action pertaining to Attachment 1 is appropriate.
Lending additional plausibility to this answer is that according to 3-ONOP-071.2 (p31; Rev 11/2/07), Action Levels 2 and above should be confirmed with 2 independent radiation monitors trending in the same direction and the same order of magnitude.
If the operator incorrectly believed that only one monitor was in excess of the TS Limit, then again, the Attachment I action would be appropriate.
C.
Correct. See reference D.
Incorrect. See explanation for options A and B. First part is correct, action is wrong 3-ONOP-071.2 (p7-13 and 31; Rev 7/26/12)
LCO 3.4.6.2 (p3/4 4-19; Amendment 233 and 228)
Technical Reference(s):
Plant Curve Book Section 5, (Attach if not previously provided)
Figure 14 Plant Curve Book Section 5, Figure 15 Plant Curve Book Proposed References to be provided to applicants during examination:
Section 5, Figures 14 & 15 6910236 Objectives 4 6 and 7 Learning Objective:
(As available)
Question Source:
Bank #
Modified Bank #
(Note changes or attach parent)
New X
Question History:
Last NRC Exam:
NA
Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 55.43 2
Facility operating limitations in the technical specifications and their bases.
Comments:
The KA is matched because the operator must demonstrate the ability to determine and interpret the flowpath for dilution of ejector exhaust air as they apply to the Steam Generator Tube Leak. This is accomplished by providing an air in-leakage value that will result in a lower reading of the SGTL monitors due to dilution, and requiring the operator to adjust the reading accordingly.
The question is at the Comprehension/Analysis (3PEO) cognitive level because the operator must recall bits of information (TS Limits and Actions), use provided graphs to provide new information, and then apply this information to a set of plant conditions to answer the question correctly.
The question is SRO-ONLY because it cannot be answered solely by knowing 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> TSITRM action statements, the LCO/TRM information listed above the line or by knowing the TS Safety Limits; AND requires the operator to apply the Required Actions in accordance with the rules of application.
Examination Outline Cross-reference:
1 Group#
2 KIA#
059 AA2.01 Importance Rating 3.5 Ability to determine and interpret the following as they apply to the Accidental Liquid Radwaste Release: The failure-indication light arrangement for a radioactive-liquid monitor Proposed Question:
SRO Question # 85 Plant conditions:
Both Units are operating at 100% power.
A liquid radioactive waste release is in progress.
The following Control Room indications are received:
H1/6, PRMS CHANNEL FAILURE annunciator is received R-18 WARNING ALARM LIGHT OFF.
R-18 Fail indicator is ON and its display is failed low.
Which ONE of the following describes the action required by 3-ONOP-067, Radioactive Effluent Release, AND what actions must be met to restart the release without performing maintenance on the monitor?
A.
Release must be manually stopped; At least two independent samples are analyzed in accordance with the ODCM surveillance requirements, and at least two technically qualified members of the facility staff independently verify the release rate calculations and discharge valve line-up prior to initiating the release.
B.
Release must be manually stopped; At least two independent samples are analyzed in accordance with the 00CM surveillance requirements prior to the release, and grab samples are taken during the release at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and these samples are analyzed for radioactivity within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
C.
Release Es automatically stopped; At least two independent samples are analyzed in accordance with the ODCM surveillance requirements prior to the release, and grab samples are taken during the release at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and these samples are analyzed for radioactivity within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
D.
Release is automatically stopped; At least two independent samples are analyzed in accordance with the 00CM
surveillance requirements, and at least two technically qualified members of the facility staff independently verify the release rate calculations and discharge valve line-up prior to initiating the release.
Proposed Answer:
A Explanation (Optional):
A.
Correct. 1 st part correct, 2nd part correct. According to 3-ONOP-067 (p7, 10 and 13; Rev 9/27/07), the operator will check that a High Alarm exists on R-18 at Step 1. Since it does not, the operator will perform the RNO and to Step 4. At Step 4, the operator will check to see if the Warning Light is on lit on R-1 8. Since it is not, the operator will perform the RNO and move to Step 8. At Step 8, the operator will check for a PRMS Channel Failure. Since a Failure light is LIT, the operator will again perform the RNO.
The RNO directs the operator to stop the release if R-1 8 has failed low (Which it has).
It then notifies the SM to refer to Technical Specifications and take all the required actions. According to the ODCM (p2-7; Rev 12/13/11), ACTION 2.1.1 states that with the number of channels FUNCTIONAL less than required by the Minimum Channels FUNCTIONAL requirement, effluent releases via this pathway may continue provided that prior to initiating a release: (1) At least two independent samples are analyzed in accordance with the surveillance requirement of Control 2.2.1, and (2) At least two technically qualified members of the facility staff independently verify the release rate calculations and discharge valve line-up; otherwise, suspend release of radioactive effluents via this pathway.
B.
Incorrect. 1st part correct, 2nd part wrong. This is incorrect because it contains wrong actions to re-initiate the release. This is plausible because it contains part of the correct ODCM ACTION. However, the ACTION for ODCM Control 3.1, ACTION 3.1.3 is substituted for the second ACTION listed as part of ODCM Control 2.1.1. According to the ODCM (p3-14; Rev 12/13/11), effluent releases via this pathway may continue provided grab samples are taken at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and these samples are analyzed for radioactivity within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. This ACTION would be associated with a failure of the Unit 3 or4 Noble Gas activity monitor; and the operator may incorrectly believe it is the ACTION required under the stated circumstances.
C.
Incorrect. 1 st part wrong, 2nd part wrong. See B and D.
D.
Incorrect. 1st part wrong, 2nd part correct. This is incorrect because the valve will not automatically close on a failed channel. This is plausible because according to 3-ONOP-067 (p9; Rev 5/27/10), the operator is provided direction to verify that RCV-01 8 is closed when a High Alarm has occurred. The operator may incorrectly believe that when the channel failed it de-energized altogether, and allowed its automatic action to fail safe.
If so, this would be the correct action.
3-ONOP-067 (p7, 10 and 13; Technical Reference(s):
Rev 9/27/07)
(Attach if not previously provided)
ODCM (p2-7 and 3-14; Rev
12/13/11 Proposed References to be provided to applicants during examination:
None 6902168 Objectives 4 and 8.e Learning Objective:
6902242 Objectives 4 and 6 (As available)
Question Source:
Bank #
Modified Bank #
(Note changes or attach parent)
New X
Question History:
Last NRC Exam:
NA Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 55.43 2
Facility operating limitations in the technical specifications and their bases.
Comments:
The KA is matched because the operator must demonstrate the ability to determine and interpret the failure-indication light arrangement for a radioactive-liquid monitor as they apply to the Accidental Liquid Radwaste Release.
The question is at the Comprehension/Analysis (2DR) cognitive level because the operator must recall bits of information (, failed detector indications, detector failures effect on RCV 018), and then relate this information to itself to identify a correct procedural response, in order to answer the question correctly.
The question is SRO-ONLY because it cannot be answered solely by knowing 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> TS/TRM action statements, the LCO/ODCM information listed above the line or by knowing the TS Safety Limits; AND requires the operator to apply the Required Actions in accordance with the rules of application.
Examination Outline Cross-reference:
2 Group#
1 KIA#
005 A2.01 importance Rating 2.9 Ability to (a) predict the impacts of the following malfunctions or operations on the RHRS, and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Failure modes for pressure, flow, pump motor amps, motor temperature, and tank level instrumentation Proposed Question:
SRO Question # 86 Plant conditions:
Unit 3 is in Mode 5.
The A Train of RHR is in service.
RCS Temperature is 190°F.
RCS pressure is 120 psig.
The B Train of RHR is removed from service for motor replacement of the 3B RHR Pump.
The secondary sides of the Steam Generators are at 50% NR.
Subsequently, the following annunciators are received:
H6/2, RHR HX HI/LO FLOW H6/3, RHR PP NB Motor Overload Attempts to restart 3A RHR Pump have failed.
RCS temperature has risen approximately 6°F in 5 minutes.
Which ONE of the following identifies the action required in accordance with 3-ONOP-050, Loss of RHR?
A.
Start one RCP; feed the associated SG using AFW pumps B.
Isolate containment; Start one RCP; feed the associated SG using Standby Feedwater Pu m PS C.
Open SG atmospheric dump valves; feed SGs using AFW pumps, and align SG Blowdown if necessary D.
Isolate containment; open SG atmospheric dump valves; feed SGs using Standby Feedwater pumps, and align SG Blowdown if necessary
Proposed Answer:
D Explanation (Optional):
A.
Incorrect. Starting an RCP will not be the success path with RCS pressure at 120 psig.
Plausible because both SGs are at 50% NR, which satisfies secondary heat sink criteria for starting an RCP B
Incorrect.
Starting an RCP will not be performed for same reason as option A.
Additional plausibility is brought by the operation of feedwater and the fact that containment will be isolated C.
Incorrect. Two of the actions are correct but the crew will not align AFW, they will align Standby Feedwater.
D.
Correct. Containment must be isolated, the unit is on the verge of changing modes, and RCS temperature is approaching 200 degrees F. the crew will attempt to stabilize temperature by use of atmospheric dump valves, Standby feedwater, and if necessary, SC blowdown.
3-ARP-097.CR.H 6/2, 6/3 Technical Reference(s):
3-ONOP-050 p6-14 (Attach if not previously provided)
Proposed References to be provided to applicants during examination:
None 3-ARP-097.CR.H 6/2, 6/3 Learning Objective:
3-ONOP-050 p6-14 (As available)
Question Source:
Bank #
Modified Bank #
(Note changes or attach parent)
New X
Question History:
Last NRC Exam:
NA Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 55.43 5
Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Comments:
The KA is matched because the operator must demonstrate the ability to (a) predict the impacts of an RHR Pump motor failure on the RHRS by determining which of the ONOP success paths is applicable for recovery The question is at the Comprehension/Analysis (2DR) cognitive level because the operator must recall bits of information (Equipment availability, procedure requirements), and then apply this information to a set of plant conditions to answer the question correctly.
H The question is SRO-ONLY because it cannot be answered solely by knowing system knowledge, immediate operator actions, plant parameters that require direct entry into EOPs, or knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure; AND requires the operator to assess plant conditions and then select a procedure or section of a procedure to mitigate, recover, or with which to proceed.
Examination Outline Cross-reference:
2 Group#
1 KJA#
008 2.1.23 Importance Rating 4.4 Conduct of Operations: Ability to perform specific system and integrated plant procedures during all modes of plant operation.
Proposed Question:
SRO Question # 87 Plant conditions:
Unit3isinModel.
3C CCW Heat Exchanger is out of service for mechanical cleaning.
The Outside SNPO requests to backwash 3A ICW/CCW basket strainer.
In accordance with 3-NOP-019, Intake Cooling Water System, which ONE of the following identifies requirements that must be met prior to granting permission to the SNPO?
A.
Restore 3C CCW heat exchanger to operable status, AND; Notify the STA to determine the minimum required ICW/CCW flowrate.
B.
Perform CCW heat exchanger performance monitoring in accordance 3-OSP-019.4, AND; Notify the STA to determine the minimum required ICW/CCW flowrate.
C.
Complete On-Line Risk Monitor risk assessment in accordance with 0-ADM-225, On Line Risk Management, AND; Notify the STA to determine the minimum required ICW/CCW flowrate.
D.
Complete On-Line Risk Monitor risk assessment in accordance with 0-ADM-225, On Line Risk Management, AND; Restore 3C CCW heat exchanger to operable status.
Proposed Answer:
C Explanation (Optional):
A.
Incorrect.
3C CCW heat exchanger is not required to be restored to operable status to allow a strainer backwash. Plausible because it is logical that restoring a heat exchanger would be required. Second half is correct.
B.
Incorrect. Heat exchanger performance monitoring would be performed when there a suspected fouling of CCW heat exchangers, but would not be used under this particular
circumstance. Second half is correct, same as in option A C.
Correct. See reference D.
Incorrect. Both halves of statement are incorrect.. See discussions of choices A and B 3-NOP-019 (p33-37, 122-124, Technical Reference(s):
127; Rev hA)
(Attach if not previously provided)
Proposed References to be provided to applicants during examination:
None 6902140 Objective 9.a, 10.b and Learning Objective:
11.c (As available)
Question Source:
Bank #
Modified Bank #
(Note changes or attach parent)
New X
Question History:
Last NRC Exam:
NA Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 55.43 2
Facility operating limitations in the technical specifications and their bases.
Comments:
The KA is matched because the operator must demonstrate the ability to perform specific system and integrated plant procedures during all modes of plant operation.(3-NOP-01 9)
The question is at the Comprehension/Analysis (2DR) cognitive level because the operator must recall bits of information (LCO requirements, operability issues), and then apply this information to a set of plant conditions which require correct interpretation of plant status and requirements to answer the question correctly
The question is SRO.ONLY because it cannot be answered solely by knowing 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> TS action statements, the LCO information listed above the line, or by knowing the TS Safety Limits; AND requires the operator to apply the operability Requirements in accordance with the rules of application.
Examination Outline Cross-reference:
2 Group#
1 KIA#
012 A2.01 Importance Rating 3.6 Ability to (a) predict the impacts of the following malfunctions or operations on the RPS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Faulty bistable operation Proposed Question:
SRO Question # 88 Plant conditions:
Unit 3 is operating at 100% power.
Annunciator C2/1, SG A NARROW RANGE LEVEL HI, alarms.
3A Steam Generator Narrow Range Level is normal.
I&C reports that Bistable BS-3-476-1 has failed and its associated Bistable light on VPB is LIT, S/G A HI LEVEL LC476-1.
The crew has entered 3-ONOP-049.1, Deviation or Failure of Safety Related or Reactor Protection Channels.
Which ONE of the following describes the action required, and the technical specification basis for the action?
Within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />...
A.
Place ONLY the failed Bistable in the tripped position; Maintains coincidence logic and redundancy to ensure safety analysis assumptions are met B.
Place the failed Bistable and all other Bistables associated with the SG level channel in the tripped position; Maintains coincidence logic and redundancy to ensure safety analysis assumptions are met C.
Place ONLY the failed Bistable in the tripped position; Ensures overall system functional capability is maintained comparable to design standards D.
Place the failed Bistable and all other Bistables associated with the SG level channel in the tripped position; Ensures overall system functional capability is maintained comparable to design standards Proposed Answer:
B
Explanation (Optional):
A.
Incorrect. This is incorrect because all of the Bistables for the failed channel must be tripped within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. This is plausible because it is clear that the failure is isolated to a failed Bistable ONLY, and the operator may incorrectly believe that only this Bistable must be placed in the tripped conditions. Second half is correct.
B.
Correct. According to 3-ARP-097.CR.C (p10; Rev 1), IF alarm is due to instrument failure, THEN REFER TO 3-ONOP-049.1, Deviation or Failure of Safety Related or Reactor Protection Channels. According to 3-ONOP-049.1 (p6; 6/27/11) the operator will be directed to verify instrument loop failure by comparison to adjacent loops and known plant parameters and conditions. In step 5.6 the operator will be directed to Refer to Technical Specifications 3/4.3, Instrumentation AND verify the minimum channels operable. According to Technical Specification LCD 3.3.1 (p3!4 3-3; Amendment 249) Functional Units 11 and 12, a total of three channels exists per SG, and a minimum of two are required per Technical Specifications and to trip, when in Modes 1 and 2. ACTION 6 is applicable otherwise. According to Technical Specifications (p3/4 3-6; Amendment 179 and 173), ACTION 6 states that with the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed until performance of the next required ANALOG CHANNEL OPERATIONAL TEST provided the inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. According to Technical Specification LCD 3.3.2 (p3!4 3-18; Amendment 249), a similar Specification is associated with Table 3.3-2 Functional Unit 5.c; with similar ACTION. According to 3-ONOP-049.1 (p6; 6/27/11), a Caution is provided just prior to Step 5.7 which states The failed channel bistable(s) is required to be placed in the tripped mode within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> of the failure determination, except if other channel bistable(s) are in the tripped or test position and would result in an undesired Engineered Safety Features actuation or Reactor Trip actuation. The overall effect of a failure of this type is a reduction of instrumentation redundancy and, therefore, a possible reduction in plant protection. Furthermore, according to 3-ONOP-049.1 (p7; 6/27/11) IF plant conditions are such that all required bistables associated with the failed channel may be tripped without an undesired RPS or ESF actuation, THEN perform the following: (1) Place all bistable switches for the affected loop in test position using Attachment 4, and (2) verify bistables tripped by observing corresponding status light (VPB) lit.
In other words, the procedure does not permit the tripping of only one Bistable under the stated conditions.
C.
Incorrect. This is incorrect because all of the Bistables for the failed channel must be tripped within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and AMSAC does not need to be bypassed and reset. This is plausible because it is clear that the failure is isolated to a failed Bistable ONLY, and the operator may incorrectly believe that only this Bistable must be placed in the tripped conditions. Second half incorrect because the basis is for other reactor protection system functions.
D.
Incorrect. First half is correct but second half describes an incorrect basis.
3-ARP-097.CR.C (plO; Rev 2)
Technical Specification LCD Technical Reference(s):
3.3.1 (p 3
4 3-3 and 3-6);
(Attach if not previously provided)
Technical Specification LCD
3.3.2 (p3/4 3-18)
(Amendment 249) 3-ONOP-049.1 (p6-7; 8/3/1 2, 6/27/11)
Proposed References to be provided to applicants during examination:
None 6910243 Objective4and6 Learning Objective:
(As available)
Question Source:
Bank #
Modified Bank #
(Note changes or attach parent)
New X
Question History:
Last NRC Exam:
NA Question Cognitive Level:
Memory or Fundamental Knowledge X
Comprehension or Analysis 10 CFR Part 55 Content:
55.41 55.43 2
Facility operating limitations in the technical specifications and their bases.
Comments:
The KA is matched because the operator must demonstrate the Ability to (a) predict the impacts of faulty bistable operation on the RPS (I.e. removes entire instrument channel from service); and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations (i.e. tripped within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> by Technical Specifications).
The question is at the Memory (1 P) cognitive level because the operator must recall bits of information to answer the question correctly.
L The question is SRO-ONLY because it cannot be answered solely by knowing 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> TS/TRM action statements, the LCO information listed above the line, or by knowing the TS Safety Limits; AND requires the operator to apply the Required Actions in accordance with the rules of application.
Examination Outline Cross-reference:
2 Group#
I KIA#
026 2.4.45 Importance Rating 4.3 Emergency Procedures / Plan: Ability to prioritize and interpret the significance of each annunciator or alarm.
Proposed Question:
SRO Question # 89 Plant conditions:
Unit 3 has tripped from 100% power due to a LOCA.
Containment Spray has automatically actuated.
The crew is implementing 3-EOP-ES-1.2, Post LOCA Cooldown and Depressurization.
Unit 3 RWST level is 152,000 gallons and lowering Containment pressure is 24 psig and slowly lowering.
Annunciator H7/5, CSP A/B COOLING WATER LO FLOW alarm is lit:
Which ONE of the following identifies the action required?
A.
Transition to 3-EOP-ES-1.3, Transfer to Cold Leg Recirculation; Immediately place both Containment Spray Pumps in Pull-To-Lock to prevent damage.
B.
Transition to 3-EOP-ES-1.3, Transfer to Cold Leg Recirculation; Refer to 3-NOP-30, CCW System, to locally raise CCW flow to the affected Containment Spray pump as time allows.
C.
Remain in 3-EOP-ES-1.2; Refer to 3-NOP-30, CCW System, to locally raise CCW flow to the affected Containment Spray pump as time allows.
D.
Remain in 3-EOP-ES-1.2; Immediately place both Containment Spray Pumps in Pull-To-Lock to prevent damage.
Proposed Answer:
B Explanation (Optional):
A.
Incorrect. Correct transition is identified based upon RWST level, but tripping containment spray pumps while containment pressure is still high is incorrect, although logical based on loss of cooling water flow
B.
Correct. See reference. SM/US may perform annunciator response at their discretion, and with Containment Spray pumps required under the current plant conditions, they may respond as required as time permits C.
Incorrect. Wrong procedure entry based upon plant conditions, but second half is correct.
D.
Incorrect. 1 st half and 2 half incorrect, but consistent with other 3 distractors and plausible because they are already in the procedure described and because loss of cooling water to a pump typically requires a trip of that pump SD-025 (p22; Rev 2/23/12) 3-ARP-097.CR.H (p34 and 44; Rev 5) 3-EOP-ES-1.2 (p 6 and Foldout Technical Reference(s):
Rev 6/15/12)
(Attach if not previously provided) 3-EOP-ES-i.3 (p7; Rev 6/1 5/1 2)
Proposed References to be provided to applicants during examination:
None 6902125 Objectives 5, 9.a, and Learning Objective:
i0.a (As available)
Question Source:
Bank #
Modified Bank #
(Note changes or attach parent)
New X
Question History:
Last NRC Exam:
NA Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 55.43 5
Assessment of facility conditions and selection of appropriate procedures during normal,
abnormal, and emergency situations.
Comments:
The KA is matched because the operator must demonstrate the Ability to prioritize and interpret the significance of annunciators associated with the operation of the Containment Spray System.
The question is at the Comprehension/Analysis (3PEO) cognitive level because the operator must recall bits of information (ES-i.3 transition criteria, ARP responses for CSP5), and then apply this information to a set of plant conditions to answer the question correctly.
The question is SRO-ONLY because it cannot be answered solely by knowing system knowledge, immediate operator actions, plant parameters that require direct entry into EOPs, or knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure; AND requires the operator to assess plant conditions and then select a procedure or section of a procedure to mitigate, recover, or with which to proceed.
Examination Outline Cross-reference:
2 Group#
I KIA#
062 A2.04 Importance Rating 3.4 Ability to (a) predict the impacts of the following malfunctions or operations on the ac distribution system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Effect on plant of de-energizing a bus Proposed Question:
SRO Question # 90 Plant conditions:
Unit 4 is operating at 100% power.
Power Range Channel N-4-43 has failed.
All the required action in accordance with 4-ONOP-059.8, Power Range Nuclear Instrumentation Malfunction, has been completed.
Subsequently, the Supply Breaker to 4P08 trips open.
Which ONE of the following describes the correct implementation of procedures and technical specifications, if applicable?
A.
Perform 4-ONOP-003.8, Loss of 120V Vital Instrument Panel 4P08, to stabilize the plant; Technical specifications allow ONE hour to restore power prior or be in MODE 3 within the next SIX hours.
B.
Perform 4-ONOP-003.8, Loss of 120V Vital Instrument Panel 4P08, to stabilize the plant; Technical Specifications allow EIGHT hours to restore power or be in MODE 3 within the next SIX hours.
C.
Perform 4-EOP-E-0, Reactor Trip or Safety Injection; Use 4-ONOP-003.8, Loss of 120V Vital Instrument Panel 4P08, concurrently with 4-EOP-ES-0.1 Reactor Trip Response, to restore power to 4P08.
D.
Perform 4-EOP-E-0, Reactor Trip or Safety Injection; Use 4-ONOP-003.8, Loss of 120V Vital Instrument Panel 4P08, ONLY after 4-EOP-ES-0.1 Reactor Trip Response, has been used to stabilize the plant.
Proposed Answer:
A Explanation (Optional):
A.
Correct. According to Lesson Plan 6900104 (p37; Rev 12/8/07), Excore NIS Power Supplies are as follows: Channel I is powered from vital AC bus 4P06. Channel II is powered from vital AC bus 4P07. Channel III is powered from vital AC bus 4P08, and Channel IV from vital AC bus 4P09. Consequently, when 4P08 is de-energized a second Power Range channel is NOT de-energized, and no automatic reactor trip will occur. According to 4-ONOP-003.8 (p3; Rev 8/10/10) Loss of the 120V Vital Instrument Panel 4P08 results in a loss of automatic feedwater control, and a loss of power to all channel III instrumentation. Enclosure I of this procedure contains a list of instrumentation lost in the Control Room due to the loss of Vital Instrument Panel 4P08.
It will not result in a reactor trip directly, but may if aggressive operator action is not taken to control Steam Generator water levels. Consequently, the correct procedure entry is into 4-ONOP-003.8. According to 4-ONOP-003.8 (p5-7; Rev 8/1 0/1 0) Step I the operator will be directed to check if a reactor trip has occurred. Since there is nothing that will require a reactor trip, the operator will perform the RNO and move to Step 2. The operator will move to Step 3 when it is determined that the plant is operating in Mode 1. Step 3 will control Pressurizer Level and Step 4 will direct the operator to control the 4C SG water level using manual control. Step 5 will direct the operator to maintain Tavg, reactor power, pressurizer pressure and level and SG water level stable. At Step 6 the operator will be directed to check power restored to 4P08.
Since no action has been taken to do this, the bus will be de-energized, and the operator will perform the Step 6 RNO. The RNO will direct the operator to Continue efforts to restore power to 4P08, and direct the operator that IF power can NOT be restored to 4P08 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, THEN perform the actions required by Technical Specifications as directed by the NPS. According to Technical Specification LCO 3.8.3.1 (p3/4 8-18; Amendment 138 and 133) 120 Volt AC Vital Panel 4P08 and 4P23 energized from its associated inverter connected to D.C. Bus 4B, while in modes 1-4.
With one A.C. vital panel either not energized from its associated inverter, or with the inverter not connected to its associated D.C. bus: (1) Reenergize the A.C. vital panel within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or be in at least HOT STANDBY within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />; and (2) reenergize the A.C. vital panel from an inverter connected to its associated D.C. bus. According to BD-ONOP-003.8 (p7; Rev 1/8/02) If power cannot be restored within the prescribed time frame (i.e. 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> per procedure), a unit shutdown is commenced to ensure compliance with Tech Specs. In other words, the shutdown is started early to ensure compliance with Technical Specification in a situation where several of the plant controllers are in MANUAL.
B.
Incorrect. This is incorrect because by procedure the shutdown must be started if 4P08 is not re-energized within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. This is plausible because the Technical Specification action 3.8.3.1.a allows 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. The student may incorrectly determine that a loss of a vital train in the most restrictive. This Tech Spec does apply to this situation but is not the most restrictive action statement..
C.
Incorrect. This is incorrect because this failure will not directly result in a reactor trip.
This is plausible because one Power Ranger instrument is OOS, and the operator may not realize that this Vital Bus powers the OOS Power Range.
If not, the operator may incorrectly conclude that a reactor trip will occur.
If so, this selection offers one potential
method of moving through the EOP Network and implementing an ONOP at the same time.
D.
Incorrect. This is incorrect because this failure will not directly result in a reactor trip.
This is plausible because one Power Ranger instrument is OOS, and the operator may not realize that this Vital Bus powers the OOS Power Range.
If not, the operator may incorrectly conclude that a reactor trip will occur.
If so, this selection offers an alternative method from choice C of moving through the EOP Network and implementing an ONOP at the same time.
Lesson Plan 6900104 (p37; Rev 12/8/07) 4-ONOP-003.8 (p3, 5-7; Rev 8/10/10)
Technical Specification LCO Technical Reference(s):
3.8.3.1 (p 3
4 8-18; Amendment (Attach if not previously provided) 138 and 133)
BD-ONOP-003.8 (p7; Rev 1/8/02)
Proposed References to be provided to applicants during examination:
None 6900104 Objective 5 Learning Objective:
6902260 Objectives 4 and 6 (As available)
Question Source:
Bank #
WTSI 53409 Modified Bank #
(Note changes or attach parent)
New Question History:
Last NRC Exam:
2007 Robinson Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 55.43 5
Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Comments:
The KA is matched because the operator must demonstrate ability to (a) predict the effect on the plant of de-energizing a bus (4P08) on the ac distribution system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions.
The question is at the Comprehension/Analysis (3PEO) cognitive level because the operator must recall bits of information (N43 power supply, NIS powered by 4P08), and then apply this information to a set of plant conditions and a procedure to answer the question correctly.
The question is SRO-ONLY because it cannot be answered solely by knowing system knowledge, immediate operator actions, plant parameters that require direct entry into EOPs, or knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure; AND requires the operator to assess plant conditions and then select a procedure or section of a procedure to mitigate, recover, or with which to proceed.
Examination Outline Cross-reference:
2 Group#
2 K!A#
034 A2.01 Importance Rating 4.4 Ability to (a) predict the impacts of the following malfunctions or operations on the Fuel Handling System ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Dropped fuel element Proposed Question:
SRO Question # 91 Plant conditions:
Unit4isinMode6.
Core alterations are in progress.
Subsequently:
A spent fuel assembly has been dropped and is lying horizontally on the top of the core basket.
The Control Room has been informed and 4-ONOP-033.3, Accidents Involving New or Spent Fuel, is being implemented.
Which ONE of the following describes the Containment evacuation that is required, AND identifies the procedure(s) that will be implemented by the Unit Supervisor?
A.
All non-essential personnel ONLY must evacuate Containment, and access control will be a responsibility of the Refueling SRO; The Refueling SRO will implement 4-ONOP-033.3, Accidents Involving New or Spent Fuel, ONLY.
B.
All personnel must evacuate Containment, and access control will shift to the RP Supervisor; The Refueling SRO will implement 4-ONOP-033.3, Accidents Involving New or Spent Fuel, ONLY.
C.
All personnel must evacuate Containment, and access control will shift to the RP Supervisor; The Refueling SRO will implement 4-ONOP-033.3, Accidents Involving New or Spent Fuel, and 4-ONOP-067, Radioactive Effluent Release.
D.
All non-essential personnel ONLY must evacuate Containment, and access control will be a responsibility of the Refueling SRO; The Refueling SRO will implement 4-ONOP-033.3, Accidents Involving New or Spent
Fuel, and 4-ONOP-067, Radioactive Effluent Release.
Proposed Answer:
B Explanation (Optional):
A.
Incorrect. 1 st part wrong, 2nd part correct. This is incorrect because the procedure speaks nothing of evacuating only non-essential personnel. Rather, the procedure requires aM personnel to evacuate, and then actions to be taken from outside the Containment. This is plausible because there are no additional indications of gas bubbles emanating from the dropped assembly, and the operator may incorrectly conclude that the event does not rise to the occasion of requiring all personnel to evacuate.
B.
Correct. 1 st part correct, 2 nd part correct. According to 4-ONOP-033.3 (p4.; Rev 8/29/1 2), this procedure will be entered upon notification that a fuel assembly has been dropped or damaged. According to 4-ONOP-033.3 (p5; Rev 8/29/1 2) Step 5.1.2, all personnel will be evacuated once this procedure is implemented. According to 4-ONOP-033.3 (p6; Rev 8/29/1 2) Step 5.2.1.7, the operator must Verify that the Containment has been evacuated, AND the Personnel Hatch, Equipment Hatch and Escape Hatch are closed. The Containment SRO appears the most likely candidate to do this for two reasons. First of all, according to 4-NOP-040.02 (p6; Rev 3) 2.1.10.C, the Containment SRO is responsible to Maintain command and control of the Refueling evolution with oversight of activities in Containment. Secondly, according to 4-NOP-040.02 (p1 2; Rev
- 3) 4.2.8, 4-ONOP-033.3 is one of nine procedures that are required to be located in the fuel handling areas of Containment during Refueling. Consequently, the Containment SRO will have this procedure available for reference. Ultimately, the RP Supervisor will control entry into the Containment. According to 4-ONOP-033.3 (p6; Rev 8/29/1 2) Step 5.2.1.11, once evacuated, personnel are directed to obtain permission from the Radiation Protection Shift Supervisor, THEN enter the containment. The Containment SRO will implement 4-ONOP-033.3 only. Although, 4-ONOP-033.3 (p3; Rev 8/29/1 2)
Step 5.2.3 requires that the crew implement 4-ONOP-067 concurrently with 4-ONOP-033.3, this procedure will be implemented by Control Room personnel, and not the Containment SRO. According to 4-NOP-040.02 (p12; Rev 3) 4.2.8, 4-ONOP-067 is NOT one of nine procedures that are required to be located in the fuel handling areas of Containment during Refueling.
C.
Incorrect. 1 st part correct, 2 nd part wrong. This is incorrect because the Containment SRO does not need to implement 4-ONOP-067 because this is done from the Control Room. This is plausible because according to 4-ONOP-033.3 (p3; Rev 8/29/1 2), Step 5.2.3 requires that the crew implement 4-ONOP-067 concurrently with 4-ONOP-033.3; and the operator may incorrectly believe that this is a responsibility of the Containment SRO.
D.
Incorrect. 1 st part wrong, 2nd part wrong. See A and C.
4-ONOP.-033.3 (p4-6; Rev 8/29/12)
Technical Reference(s):
4-NOP-040.02 (p6 and 12; Rev (Attach if not previously provided) 3)
Proposed References to be provided to applicants during examination:
None 6902283 Objectives 4 6 and 8 Learning Objective:
(As available)
Question Source:
Bank #
Modified Bank #
(Note changes or attach parent)
New X
Question History:
Last NRC Exam:
NA Question Cognitive Level:
Memory or Fundamental Knowledge X
Comprehension or Analysis 10 CFR Part 55 Content:
55.41 55.43 7
Fuel handling facilities and procedures.
Comments:
The KA is matched because the operator must demonstrate the ability to (a) predict the impacts of the dropped fuel element on the Fuel Handling System (i.e. system abandoned or control maintained); and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences (Which procedures implemented by Containment SRO).
The question is at the Memory (1 P) cognitive level because the operator must recall bits of information (type of evacuation, procedures to be implemented) to answer the question correctly.
ft H
The question is SRO-ONLY because it deals directly with Containment SRO (i.e. Refuel floor SRO) responsibilities during Refueling, which are not shared by the RO. While it is noted that an RO would be expected to know the type of evacuation that is required, the SRO-ONLY would be expected to know who controls access once evacuated, and what procedures must be implemented by the Containment SRO.
Examination Outline Cross-reference:
2 Group#
2 KIA#
071 A2.02 Importance Rating 3.6 Ability to (a) predict the impacts of the following malfunctions or operations on the Waste Gas Disposal System ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Use of waste gas release monitors, radiation, gas flow rate, and totalizer Proposed Question:
SRO Question # 92 Plant conditions:
Both units are operating at 100% power.
It is necessary to release Waste Gas Decay Tank F.
Plant Vent Gas Monitor, R-14, is OPERABLE.
Plant Vent SPING, RAD-6304 is inoperable.
Which ONE of the following identifies the MINIMUM action required to perform the release?
REFERENCE PROVIDED A.
The release may be performed using 0-NOP-061.014F, Waste Gas Disposal System Controlled Release of Decay Tank F, using existing operable equipment.
B.
Perform maintenance on Plant Vent SPING to restore Channel 5, Low Range Noble Gas Monitor, to service, ONLY.
C.
Perform maintenance on Plant Vent SPING to restore Channel 10, Plant Ventilation Flow Monitor, to service, ONLY.
D.
Perform maintenance on Plant Vent SPING to restore Channel 5, Low Range Noble Gas Monitor, and Channel 10, Plant Ventilation Flow Monitor, to service.
Proposed Answer:
A Explanation (Optional):
A.
Correct. Under the current conditions ACTION is needed with the flow monitor OOS but the procedure covers that by insuring a portable flow monitoring device is installed prior to release.
B.
Incorrect. This is incorrect because Channel 5 of the SPING4 is not required by the ODCM to perform this release.
In fact, according to 0-NOP-061.14F (p4; Rev 0)
Precautions 2.1.6 & 7, Plant Vent SPING-4, Channel 5 shall NOT be used to satisfy the Minimum Channel OPERABILITY requirements of ODCM, Table 3.1-1; and Plant Vent SPING-4, Channel 5 shall NOT be used in lieu of the sampling requirements of ODCM, Table 3.1-1. This is plausible because according to Technical Specification 3.3.3 (p 3
4 3-36; Amendment 149 and 144), the SPING4 monitor may be used to satisfy the instrument requirement for the Unit 4 Spent Fuel Storage Pool Areas. Consequently, the operator may incorrectly believe that the SPING4 is required to avoid reliance upon an ACTION statement under the current conditions.
C.
Incorrect. Plausible because according to the ODCM (p3-12; Rev 12/13/11), Table 3.1-1, there are two instruments needed to be OPERABLE to avoid reliance upon an ACTION statement to conduct the release. The first is that the Noble Gas Activity Monitor - Providing Alarm and Automatic Termination of Release (Plant Vent Monitor) must be OPERABLE. According to Lesson Plan 6902150 (p44; Rev 9/3/08), RCV-14 will trip closed on high radiation as sensed at RE-14. Consequently, the first required instrument is OPERABLE. However, the second instrument required to be OPERABLE is the Effluent System Flow Rate Measuring Device. According to 0-NOP-061.14F (p21; Rev 0) this instrument is Channel 10 of the SPING4 Monitor. Since it is NOT OPERABLE, reliance upon an ACTION statement will be needed. Consequently, the MINIMUM ACTION necessary to avoid reliance upon the ACTION statement is to perform maintenance on SPING4 to restore Channel 10, Plant Ventilation Flow Monitor, to service or provide a flow monitoring device in its place.
D.
Incorrect. This is incorrect because Channel 5 of the SPING4 is not required by the ODCM to perform this release. This is plausible because according to Technical Specification 3.3.3 (p3/4 3-36; Amendment 149 and 144), the SPING4 monitor may be used to satisfy the instrument requirement for the Unit 4 Spent Fuel Storage Pool Areas.
Consequently, the operator may incorrectly believe that the SPING4 is required to avoid reliance upon an ACTION statement under the current conditions.
ODCM (p3-12; Rev 12/13/11),
Table 3.1-1 Lesson Plan 6902150 (p44; Rev 9/3/08) 0-NOP-061.14F(p4and2l; Technical Reference(s):
Rev 0)
(Attach if not previously provided)
Technical Specification 3.3.3 (p3/4 3-36; Amendment 149 and 144)
Proposed References to be provided to applicants during examination:
ODCM Table 3.1-1
6918150 Objective 7.b, 8.c, 9.e, Learning Objective:
and 12 (As available)
Question Source:
Bank #
Modified Bank #
(Note changes or attach parent)
New X
Question History:
Last NRC Exam:
NA Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 55.43 2
Facility operating limitations in the technical specifications and their bases.
Comments:
The KA is matched because the operator must demonstrate the ability to (a) predict the impacts of the use of waste gas release monitors; radiation, gas flow rate, and totalizer on the Waste Gas Disposal System (need to enter Action Statement without maintenance);
and (b) based on those predictions, use procedures (i.e. take actions per ODCM) to correct, control, or mitigate the consequences of those malfunctions or operations.
The question is at the Comprehension/Analysis (2DR) cognitive level because the operator must recall bits of information (ODCM requirements for instrumentation needed for release),
and then relate this information to itself, by applying the ODCM requirements, to answer the question correctly.
The question is SRO-ONLY because it cannot be answered solely by knowing 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> TS/TRM action statements, the LCD information listed above the line, or by knowing the TS Safety Limits.
It requires the operator to identify required instrumentation needed to perform a radioactive release, which is found in a Table in the ODCM.
Examination Outline Cross-reference:
2 Group#
2 K/A #
079 2.4.35 Importance Rating 4.0 Emergency Procedures I Plan: Knowledge of local auxiliary operator tasks during emergency and the resultant operational effects.
Proposed Question:
SRO Question # 93 Plant conditions:
Unit3isinMODEl.
A loss of instrument air is in progress.
All Instrument Air Compressors are operating.
The TO reports that the instrument air particulate filters are clogged.
Which ONE of the following identifies the actions that are required in accordance with 0-ONOP-013, Loss of Instrument Air?
The SM/US must A.
notify System Operations to enable the NERC load shedding contingency software; then direct the TO to open IAS-3-012 and IAS-4-012, Instrument Air cross connect valves.
B.
notify System Operations to enable the NERC load shedding contingency software; then direct the TO to align Service Air to Instrument Air via 40-215, Instrument Air Isolation Valve from Service Air.
C.
apply the requirements of 10CFR5O.54(x) and (y); then direct the TO to open IAS 012 and IAS-4-012, Instrument Air cross connect valves.
D.
apply the requirements of 1 OCFR5O.54(x) and (y); then direct the TO to align Service Air to Instrument Air via 40-215, Instrument Air Isolation Valve from Service Air.
Proposed Answer:
A Explanation (Optional):
A.
Correct. The US must notify NERC due to the fact that the lineup to be performed will bypass automatic safety features and put both units at risk. The TO will open 2 manual
cross connect valves that are upstream of the Instrument Air filters to ensure Instrument Air supply when filters are clogged. This will bypass the ability of the Instrument Air system to isolate on low air pressure.
B.
Incorrect. First part is correct, 2nd part is incorrect because it identifies the wrong valve to supply air to the instrument air system. Using this valve would not help the situation, as it is downstream of the clogged filters C.
Incorrect. IOCFR5O.54(x) is incorrect but plausible because the action taken will bypass a safety feature of the instrument air system. (isolation on low IA pressure.)
However, instrument air is not a TS system. 2nd half of this option is correct.
D.
Incorrect. See explanations above Technical Reference(s):
0-ONOP-013 rev. 4, pg 12 (Attach if not previously provided)
Proposed References to be provided to applicants during examination:
None LP6910286obj4 Learning Objective:
(As available)
Question Source:
Bank #
Modified Bank #
(Note changes or attach parent)
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 55.43 5
Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Comments:
KA is matched because it evaluates knowledge of AO tasks during a loss of Instrument Air and the resultant operational effects (Where to cross-connect Instrument Air when filters are clogged).
Question is written at comprehension level because the SRO must understand system alignments that will mitigate the event, as well as notifications to off-site agencies when the unit is in a condition that elevates risk of a dual unit trip Question is SRO level because the actions contained related to this KA will be performed by the SRO, and actions to notify NERO are SRO responsibility. Placing the unit in a condition susceptible to dual unit trip is a design basis decision also made only by the SRO
Examination Outline Cross-reference:
3 Group#
1 K!A#
Gi 2.1.37 Importance Rating 4.6 Conduct of Operations: Knowledge of procedures, guidelines, or limitations associated with reactivity management.
Proposed Question:
SRO Question # 94 Which ONE of the following identifies the requirement for reactivity management at low power operations in accordance with O-ADM-200, Conduct of Operations?
A.
Operations Manager must approve extended low power operation; approval must be recorded in the reactor operator narrative log.
B.
Operations Manager must approve extended low power operation; approval must be documented in an AR.
C.
Plant General Manager must approve extended low power operation; approval must be recorded in the reactor operator narrative log.
D.
Plant General Manager must approve extended low power operation; approval must be documented in an AR.
Proposed Answer:
A Explanation (Optional):
A.
Correct. Per O-ADM-200 Rev 13, p28 B.
Incorrect. First part right, second part wrong C.
Incorrect. First part wrong, second part right D.
Incorrect. Both parts wrong OADM-200 rev 13 p.28 Technical Reference(s):
(Attach if not previously provided)
Proposed References to be provided to applicants during examination:
None Learning Objective:
(As available)
Question Source:
Bank #
Modified Bank #
(Note changes or attach parent)
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 55.43 5
Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Comments:
KA is matched because it directly evaluates knowledge of reactivity management policy during low power operations.
Question is written at memory cognitive level because the applicant must know who is responsible for the decision to operate at low power for extended periods, which is a reactivity management issue SRO criteria is met because the application of this policy is controlled by the SRO
Examination Outline Cross-reference:
3 Group#
2 KIA#
G2 2.2.11 Importance Rating 3.3 Equipment Control: Knowledge of the process for controlling temporary design changes.
Proposed Question:
SRO Question # 95 Which ONE of the following is the responsibility of the Unit Supervisor with STA Responsibility in regard to Temporary Modifications in accordance with O-ADM-503, Temporary Modification?
A.
Process extensions to Temporary Modification expiration dates B.
Determine whether a 10CFR5O.59 applicability screening is required prior to installation of the Temporary Modification C.
Maintain the Temporary Modification Log Index (Form F213) in the Temporary Modification File D.
Update plant drawings to reflect redlines resulting from Temporary Modifications Proposed Answer:
C Explanation (Optional):
A.
Incorrect. The US does not process extensions, but this is plausible because the US would identify Temporary Changes that may require extensions. Qualified Engineer would actually process them B.
Incorrect. This task would be performed by a Qualified Reviewer or Engineering. The US would take part in a screening by assisting in answering questions provided by the screening C.
Correct. See reference.
D.
Incorrect. The US would not update plant drawings, this task is performed by Records Management, but if the Temporary Mod were to be performed off-hours, the US would ensure that the affected drawings and red-lines were available in the control room Technical Reference(s):
O-ADM-503 (plO
- 14; Rev 3)
(Attach if not previously provided)
Proposed References to be provided to applicants during examination:
None 6902029 Objective 2 Learning Objective:
(As available)
Question Source:
Bank #
Modified Bank #
(Note changes or attach parent)
New X
Question History:
Last NRC Exam:
NA Question Cognitive Level:
Memory or Fundamental Knowledge X
Comprehension or Analysis 10 CFR Part 55 Content:
55.41 55.43 3
Facility licensee procedures required to obtain authority for design and operating changes in the facility.
Comments:
The KA is matched because the operator must demonstrate Knowledge of the process for controlling temporary design changes, specifically responsibilities in the temporary modification program.
The question is at the Memory (1 P) cognitive level because the operator must recall information about responsibilities and requirements for performing temporary plant modifications The question is SRO-ONLY because it pertains to facility licensee procedures required to obtain authority for design and operating changes in the facility, specifically, administrative processes for temporary modifications.
Examination Outline Cross-reference:
3 Group#
2 KIA#
G2 2.2.35 Importance Rating 4.5 Equipment Control: Ability to determine Technical Specification Mode of Operation.
Proposed Question:
SRO Question # 96 Plant conditions:
Unit 3 is operating at 25% power.
Unit 4 is in a plant heatup.
Unit 4 RCS Tavg is 342° F.
Subsequently, the 4A High Head Safety Injection Pump is declared inoperable due to an inoperable flow path.
Which ONE of the following identifies the Unit 4 Operational Mode, AND the requirements needed to enter the next highest Operational Mode?
REFERENCE PROVIDED A.
Mode 3; Mode 2 can be only entered when the 4A HHSI Pump is restored to OPERABLE status within 30 days.
B.
Mode 3; Mode 2 can be entered as long as the HHSI flowpath is restored to OPERABLE prior to Tave exceeding 380°F.
C.
Mode 4; Mode 3 can be only entered when the 4A HHSI Pump is restored to OPERABLE status within 30 days..
D.
Mode 4; Mode 3 can be entered as long as the HHSI flowpath is restored to OPERABLE prior to Tave exceeding 380°F.
Proposed Answer:
D
Explanation (Optional):
A.
Incorrect. This is incorrect because the plant is in Mode 4. This is plausible because the difference between Mode 3 and 4 is entirely related to the RCS Temperature threshold, which the operator may not know.
If the operator incorrectly believes that the plant is in Mode 3, it is reasonable to assume that the 4A HHSI Pump must be returned to OPERABLE status before entering Mode 2, because LCO 3.0.4 requires that that the LCD be met without reliance upon the ACTION statements when the statement includes shutdown requirements. The operator may not be aware of the stated LCD 3.0.4 exception.
B.
Incorrect. This is incorrect because the plant is in Mode 4. This is plausible because the difference between Mode 3 and 4 is entirely related to the RCS Temperature threshold, which the operator may not know.
If the operator incorrectly believes that the plant is in Mode 3, it is reasonable to assume that the 4A HHSI Pump does not need to be returned to OPERABLE status before entering Mode 2, because an exception to LCD 3.0.4 is stated in the LCD.
C.
Incorrect. This is incorrect because an entry can be made into Mode 3 with the 4A HHSI Pump inoperable. This is plausible because the operator may incorrectly believe that the 4A HHSI Pump must be returned to OPERABLE status before entering Mode 3, because LCD 3.0.4 requires that that the LCO be met without reliance upon the ACTION statements when the statement includes shutdown requirements. The operator may not be aware of the stated LCO 3.0.4 exception.
D.
Correct. According to Technical Specifications Table 1.2 (p1-8; Amendment 137 and 132) Mode 4, Hot Shutdown, is defined as Keff <0.99, 0% rated Thermal Power, and 350°F > Tavg > 200°F. Consequently, under the current conditions the plant is in Mode
- 4. According to Technical Specifications LCO 3.0.4 (p3/4 0-2; Amendment 235 and 230) entry into an OPERATIONAL MODE or other specified condition shall not be made when the conditions for the Limiting Conditions for Operation are not met and the associated ACTION requires a shutdown if they are not met within a specified time interval. Entry into an OPERATIONAL MODE or specified condition may be made in accordance with ACTION requirements when conformance to them permits continued operation of the facility for an unlimited period of time. Exceptions to these requirements are stated in the individual specifications. According to Technical Specifications LCO 3.5.2 (p3!4 5-3; Amendment 212 and 206) with one of the four required Safety Injection pumps inoperable and the opposite unit in MODE 1, 2, or 3, restore the pump to OPERABLE status within 30 days or be in at least HOT STANDBY within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. A note is provided indicating that the provisions of Specifications 3.0.4 and 4.0.4 are not applicable. In other words, an exception exists. Consequently, the heatup can continue, and Mode 3 can be entered, as long as ACTION Statement 3.5.2.c is being complied with.
Technical Specifications Table 1.2 (p1-B; Amendment 137 and Technical Reference(s):
132)
(Attach if not previously provided)
Technical Specifications LCO 3.0.4 (p 3
4 0-2; Amendment
235 and 230)
Technical Specifications LCO 3.5.2 (p3/4 5-3; Amendment 212 and 206)
Proposed References to be provided to applicants during examination:
with 6910517 Objectives 5, 10 and 14 Learning Objective:
6902121 Objective 13 (As available)
Question Source:
Bank #
Modified Bank #
(Note changes or attach parent)
New X
Question History:
Last NRC Exam:
X Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 55.43 2
Facility operating limitations in the technical specifications and their bases.
Comments:
The KA is matched because the operator must demonstrate the Ability to determine Technical Specification Mode of Operation. This is accomplished by providing the operator with a set of conditions and requiring that the Mode of Operation be identified The question is at the Comprehension/Analysis (2DR) cognitive level because the operator must recall bits of information (Definition of Mode 4, LCD 3.0.4 requirements, 3.5.2 ACTION requirements, Does LCO 3.5.2 except LCD 3.0.4), and then apply this information to a set of plant conditions to answer the question correctly.
The question is SRO-ONLY because it cannot be answered solely by knowing 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> TS action statements, the LCD information listed above the line, or by knowing the TS Safety Limits; AND requires the operator to apply the Required Actions in accordance with the rules of
application and apply the generic LCO requirements.
Examination Outline Cross-reference:
3 Group#
3 KIA#
G3 2.3.4 Importance Rating 3.7 Radiation Control: Knowledge of radiation exposure limits under normal or emergency conditions.
Proposed Question:
SRO Question # 97 You are the Emergency Coordinator (EC) during a LOCA outside containment.
A worker is critically injured and unconscious in the Auxiliary Building.
The Shift RP Tech estimates that each of the two proposed rescue team members will receive 30 REM while rescuing the injured person.
Which ONE of the following describes the authorization required in accordance with EPIP 0-20111, Re-Entry?
A.
You may authorize jy volunteers to rescue the injured person. They must understand that they may receive in excess of 25 Rem.
B.
You may assign personnel to rescue the injured person, but jjy for dose up to 25 Rem C.
You must receive the Radiation Protection and Chemistry Managers permission to authorize exceeding the 4 REM dose limit for the volunteer rescuers.
D.
You may not authorize the entry with this expected dose. Plant Managers approval is required to allow volunteers to use emergency dose limits.
Proposed Answer:
A Explanation (Optional):
A.
Correct. Authorization is by Emergency Coordinator.
B.
INCORRECT. Emergency Exposure is voluntary. EC cannot assign C.
INCORRECT. RP & Chemistry Manager controls dose up to 4 REM. Above 4 REM the EC is the approval authority
D.
INCORRECT. EC can authorize this exposure 0-EPIP-201 1 1 rev 1 3/23/11 Technical Reference(s):
pp. 8-9 14-16 (Attach if not previously provided)
Proposed References to be provided to applicants during examination:
N Learning Objective:
(As available)
Question Source:
Bank #
WTSI 99867 Modified Bank #
(Note changes or attach parent)
New Question History:
Last NRC Exam:
2006 Ginna Question Cognitive Level:
Memory or Fundamental Knowledge X
Comprehension or Analysis 10 CFR Part 55 Content:
55.41 55.43 5
Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
KA is matched because the applicant must display knowledge of emergency exposure limits Question is memory level because the item requires memorization of the rules for allowing high radiation exposure to volunteers Question is SRO level because it evaluates a decision made only by SROs in the function of Emergency Coordinator during a plant emergency Comments:
Examination Outline Cross-reference:
3 Group#
3 KJA#
G3 2.3.5 Importance Rating 2.9 Radiation Control: Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.
Proposed Question:
SRO Question # 98 Plant conditions:
Unit 3 is in the process of Reactor Vessel Head disassembly for Refueling activities.
Area Radiation Monitor (ARMS) RD-1402B, 58 Elevation, has a broken dose rate indicator on the local monitor.
In accordance with 3-OP-038.1, Preparation for Refueling Activities, which ONE of the following identifies the action necessary to allow the Refueling Supervisor to resume core alterations?
A.
RP survey of the 58 elevation every 15 minutes.
B.
Dedicated operator to constantly monitor Control Room (ARMS) RD-1402B.
C.
Local indication for (ARMS) R-2 replaced with portable monitor with alarm.
D.
Control Room functions for (ARMS) R-2 verified operable every 15 minutes.
Proposed Answer:
C Explanation (Optional):
A.
Incorrect. This is incorrect because the action is NOT required in accordance with of O-OP-038.1. This is plausible because it appears to be equivalent to having R-2 OPERABLE; and the operator may incorrectly believe that this is an acceptable alternative to using R-2.
B.
Incorrect. This is incorrect because the action is NOT required in accordance with of O-OP-038.1. This is plausible because according to O-ADM-200 (p38; Rev 12) the use of a dedicated operator, and the rules for their usage, is proceduralized; and the operator may incorrectly believe that this is an acceptable alternative.
C.
Correct. According to 3-OP-038.1, (p14; Rev 4/15/09), Step 5.2.2.4, the Refueling Supervisor must commence performance of Attachment 3 prior to the start of and during control rod unlatching and removal of upper internals. According to 0-OP-038.1, (p35; Rev 8/31/1 0), Area Radiation Monitor R-2 is required to be OPERABLE (both remote and local) in order to move control rods or the upper internals.
If the instrument is inoperable, Refueling Operations must be stopped immediately.
If area monitor R-2 is NOT operable, a portable monitor, with alarm may be used.
D.
Incorrect. This is incorrect because the action is NOT required in accordance with of 3-OP-038.1. This is plausible because if this were implemented the Control Room would be able to stop core alterations upon observing high radiation; and the operator may incorrectly believe that this is an acceptable alternalive.
3-OP-038.1, (p14; Rev 4/15/09) 3-OP-038.1, Attachment 3 (p35; Technical Reference(s):
Rev 8/31/1 0)
(Attach if not previously provided) 0-ADM-200 (p38; Rev 12)
Proposed References to be provided to applicants during examination:
N 6902953 Objectives 1-3 Learning Objective:
(As available)
Question Source:
Bank #
WTSI 99797 Modified Bank #
(Note changes or attach parent)
New Question History:
Last NRC Exam:
2009 Turkey Point Question Cognitive Level:
Memory or Fundamental Knowledge X
Comprehension or Analysis 10 CFR Part 55 Content:
55.41 55.43 7
Fuel handling facilities and procedures.
Comments:
The KA is matched because the operator must demonstrate the ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.
The question is at the Memory (1 P) cognitive level because the operator must recall bits of information (potential alternatives to local meter inoperability) to answer the question correctly.
The question is SRO-ONLY because it deals directly with Containment SRO (i.e. Refuel floor SRO) responsibilities during Refueling, which are not shared by the RO, specifically Refuel SRO floor responsibilities and prerequisites for vessel disassembly and reassembly.
NOTE: This question was used on the 2009 Turkey Point SRO NRC Exam. (Not last 2)
Examination Outline Cross-reference:
3 Group#
4 KJA #
G4 2.4.34 Importance Rating 4.1 Emergency Procedures I Plan: Knowledge of RO tasks performed outside the main control room during an emergency and the resultant operational effects.
Proposed Question:
SRO Question # 99 Plant conditions:
25 minutes ago the Control Room was evacuated due to a fire.
The crew has implemented O-ONOP-105, Control Room Evacuation.
The SM has not determined whether plant control has been established.
The RO is performing Attachment 14 and preparing to initiate RCS Boration.
Which ONE of the following correctly completes the statement below?
The Boration to Cold Shutdown for Unit 3 is performed by (1)
. The Shift Manager must declare a(n)
(2)
, as a minimum, for this event.
REFERENCE PROVIDED A.
(1) borating for 136 minutes from Boric Acid Storage Tanks with 3B Charging pump in local speed control at 12 psig.
(2) Alert B.
(1) locally aligning charging pump suction to the RWST until sample determines that (2) Alert C.
(1) locally aligning charging pump suction to the RWST until sample determines that (2) Site Area Emergency D.
(1) borating for 136 minutes from Boric Acid Storage Tanks with 3B Charging pump in local speed control at 12 psig.
(2) Site Area Emergency Proposed Answer:
D Explanation (Optional):
A.
Incorrect. 1 st part wrong, 2 nd part wrong. See B and C.
B.
Incorrect. 1 st part correct, 2nd part wrong. This is incorrect because the conditions require an SAE to be declared as a minimum. This is plausible because according to EPIP-201 01, Attachment 1 Hot Conditions Table (p28; Rev 5) HA5 the initiation of a Control Room Evacuation requires an ALERT. Also actions for boration are plausible because it would borate directly to RCS in this lineup C.
Incorrect. 1 5t part wrong, 2nd part correct. This is incorrect because the emergency boration actions are different than what is presented in this option. According to 0-ONOP-1 15 (p65; Rev 5/4/04), the emergency boration of Unit 3 takes places by using BASTs for 136 minutes D.
Correct. 1 st part correct, 2nd part correct. According to 0-ONOP-105 (p72; Rev 5 11/9/1 2), the emergency boration of Unit 3 takes places in Step 25 of Attachment 14.
NOTE prior to Step 19 of Attachment 13 states that failure to establish control of shutdown system from outside the Control Room within 15 minutes may place the plant in a Site Area Emergency condition. The Shift Manager should consult 0-EPIP-201 01, DUTIES OF THE EMERGENCY COORDINATOR. SAE must be declared if Control Room Evacuation Has Been Initiated and Plant Control Cannot be Established. This is applicable during all modes of operation, and the specific thresholds are (1) Control Room evacuation has been initiated, and (2) Control of the plant cannot be established within 15 minutes. The basis for this classification to capture those events where control of the plant cannot be reestablished in a timely manner. In this case, expeditious transfer of control of safety systems has not occurred (although fission product barrier damage may not yet be indicated). The intent of the Threshold is to establish control of important plant equipment and knowledge of important plant parameters in a timely manner. Primary emphasis should be placed on those components and instruments that supply protection for and information about safety functions. These safety functions are reactivity control, RCS inventory, and secondary heat removal. The determination of whether or not control is established at the remote shutdown panel is based on Emergency Coordinator judgment. The Emergency Coordinator is expected to make a reasonable, informed judgment within the time for transfer that the licensee has control of the plant from the remote shutdown panel. Additionally, actions for emergency boration are correct 0-ONOP-105 (p50 and 65; Rev 5, 11/9/12) 0-BD-ONOP-105 (p6; Rev 5 Technical Reference(s):
EPIP-201 01, Attachment 1 Hot (Attach if not previously provided)
Conditions Table (p27-28; Rev 5)
Proposed References to be provided to applicants during examination:
EPIP 20101 Att. 1 6902252 Objectives 3 4, 5 and 6 Learning Objective:
(As available)
Question Source:
Bank #
Modified Bank #
(Note changes or attach parent)
New X
Question History:
Last NRC Exam:
NA Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 55.43 6
Procedures and limitations involved in initial core loading, alterations in core configuration, control rod programming, and determination of various internal and external effects on core reactivity.
Comments:
The KA is matched because the operator must demonstrate Knowledge of RO tasks performed outside the main control room during an emergency and the resultant operational effects.
Specifically, the operator must know that because 15 minutes has elapsed since the control room evacuation, and the SM has not verified control, a Site Area Emergency Must be declared.
The question is at the Comprehension/Analysis (2DR) cognitive level because the operator must recall bits of information, and then relate this information to itself by applying this to to answer the question correctly.
The question is SRO-ONLY because it deals directly with the determination of various external effects of core reactivity, specifically, evaluating core conditions and emergency classifications based on core conditions.
Examination Outline Cross-reference:
3 Group#
4 KIA#
G4 2.4.37 Importance Rating 4.1 Emergency Procedures I Plan: Knowledge of the lines of authority during implementation of the emergency plan.
Proposed Question:
SRO Question # 100 Given the following:
A General Emergency has been declared at Unit 3.
The EOF and TSC are not yet operational.
Which ONE of the following identifies (1) g Emergency Coordinator responsibilities that can be delegated to another individual, AND (2) the individual who has the responsibility to declare PARs once the TSC ONLY is declared operational?
A.
(1) The decision to issue Potassium Iodide (K) and the decision to notify federal, state and local agencies; (2)The Shift Manager.
B.
(1) The decision to issue Potassium Iodide (KI) and the decision to notify federal, state and local agencies; (2) The Emergency Coordinator in the TSC.
C.
(1) The decision to evacuate site personnel and the decision to waive emergency response training requirements; (2) The Emergency Coordinator in the TSC.
D.
(1) The decision to evacuate site personnel and the decision to waive emergency response training requirements; (2) The Shift Manager.
Proposed Answer:
C Explanation (Optional):
A.
Incorrect. 1 st part wrong, 2nd part correct. This is incorrect because the decision to notify federal, state and local agencies cannot be waived.
It is plausible because it is one of several responsibilities of the Emergency Coordinator, and the operator may
incorrectly believe that this can be delegated. Additionally, according to 0-EPIP-20101 (p8; Rev 3/23/11), the Emergency Coordinator shall authorize the issuance of Potassium Iodide (KI) to emergency workers upon the recommendation from the TSC Radiation Protection Supervisor based on a thyroid CDE of greater than or equal to 5 rem actual or estimated. This activity is NOT listed among the responsibilities that cannot be delegated.
B.
Incorrect. 1 st part wrong, 2nd part wrong. See A and C.
C.
Correct.
D.
Incorrect. 1 st part correct, 2 nd part incorrect. Once the SM has transferred responsibility, he is no longer responsible for EC functions 0-EPIP-20101 (p8, 11 and 13; Technical Reference(s):
Rev 10/22/1 2)
(Attach if not previously provided)
Proposed References to be provided to applicants during examination:
None Learning Objective:
(As available)
Question Source:
Bank #
Modified Bank #
(Note changes or attach parent)
New X
Question History:
Last NRC Exam:
NA Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 55.43 1
Conditions and limitations in the facility license Comments:
The KA is matched because the operator must demonstrate Knowledge of the lines of authority during implementation of the emergency plan. This is accomplished by requiring the operator to identify responsibilities of the emergency coordinator that can be delegated to another individual when implementing the emergency plan, and who is responsible for issuing the
PARs once the EOF is operational.
The question is at the Comprehension/Analysis (2DR) cognitive level because the operator must recall bits of information (Responsibilities of EC and RM, Responsibilities that cannot be delegated), and then relate this information to itself by requiring that responsibilities that can be waived be identified, and subsequent responsibilities of the RM be identified to answer the question correctly.
The question is SRO-ONLY because it requires the operator to possess knowledge of emergency plan requirements of regulations imposed by 1 OCFR5O.47, Emergency Plans.