ML12159A391

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Draft Written Exam (Folder 2)
ML12159A391
Person / Time
Site: Oyster Creek
Issue date: 06/01/2012
From: D'Antonio J
Operations Branch I
To:
Exelon Generation Co, Exelon Nuclear
Jackson D
Shared Package
ML120230007 List:
References
TAC U01848
Download: ML12159A391 (324)
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EXAMINATION ANSWER KEY oc RO NRC Exam 1 10: 11-1 NRO 01 Points: 1.00 The plant is at rated power. An event then occurred and the following annunciator came into alarm:

  • FCS/RFCS - DUAL LINK FAILURE Which ONE of the following describes the effect on the Digital Feedwater Control System (DFCS) AND RPV Water Level?

DFCS (i).

RPV Water Level (2).

A. (1) transfers to the Moore Stations (2) drops until operator action is taken B. (1) does NOT transfer to the Moore Stations (2) drops until operator action is taken C. (1) transfers to the Moore Stations (2) remains constant since DFCS functions to maintain last known setpoint D. (1) does NOT transfer to the Moore Stations (2) remains constant since DFCS functions to maintain last known setpoint Answer: D IAnswer Explanation QID: 11-1 NRO 01 Question # I 1 I Developer / Date: JJR /5-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO SRO 259002 Reactor Water Level Control K1.03 - Knowledge of the physical connections and/or cause- effect relationships between 3.8 3.9 REACTOR WATER LEVEL CONTROL SYSTEM and the following: Reactor water level OCSOPSILT Page: 1 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam Level RO Tier 2 Group 1 General RAP-J2c, J1c References o is Correct. A Dual Link failure alarm informs the control room operators that the Moore Stations are disabled and the OCCs continue to function normally based on the last settings obtained from the Moore stations. As a result, the FRV's will continue to maintain RPV water level at the same setpoint so there is no impact on level.

A is Incorrect but plausible. Moore stations are disabled upon a Explanation dual link failure. This would be the expected condition for a dual computer failure. Level will remain the same.

B is Incorrect but plausible since it is the correct OCFS response, but incorrect RPV water level response.

o is Incorrect but plausible since it is the incorrect OFCS response, but correct RPV water level response.

Lesson Plan 2621.828.0.0018, Feedwater Control System FWC-10449, State the function and interpretation of system Learning alarms, alone and in combination, as applicable in accordance with the RAPS.

Question Source (New, Modified, Bank) I Bank If Bank Qr MQdifi~d VISION System/Question 10: 505981 Question Source: ILT Bank Previous 2 Exams: No

EXAMINATION ANSWER KEY oc RO NRC Exam Memory or X Comprehension Fundamental 3:PEO or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: Predict an Event or Outcome 10CRF55 55.41b I 7 55.43b I

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

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes Point Value: 1 System 10 No.: 295002 PRA:

I No Safety [81 Initial License Level Function:

2 o LORT

EXAMINATION ANSWER KEY oc RO NRC Exam 2 10: 11-1 NRO 02 Points: 1.00 The plant is in cold shutdown and is cooling down with Shutdown Cooling. The following conditions currently exist:

  • RBCCW Pump 1-1 is in service
  • A, B, C, and E Reactor Recirculation Loops are idle; Reactor Recirculation Pump D is operating Which of the following will result in the GREATEST impact (after 3 minutes) on reactor coolant cooldown rate, with NO Operator action?

A. A loss of Unit Substation 1A2 due to overload.

B. SDC Pump B sensed suction temperature rises to 360 OF.

C. An 86/S1A lockout occurs due to the trip of the differential relay 87SA.

D. A loss of Drywell cooling which results in a Drywell pressure of 2.6 psig.

Answer: A IAnswer Explanation QID: 11-1 NRO 02 Question # J 2 I Developer 1 Date: JJR 15-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO SRO 205000 Shutdown Cooling K1.05 - Knowledge of the physical connections andlor cause- effect relationships between 3.1 3.1 SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) and the followin I: Component cooling water s~stems Level RO I Tier I 2 I Group I 1 OCSOPS ILT Page: 4 of 241 02 April 2012

EXAMINATION ANSWER KEY DC RD NRC Exam General ABN-45 References Answer A is Correct. A loss of Unit Substation1A2 results in the loss of SOC Pump A and RBCCW Pump 1-1. SOC Pump B remains in service but there is no RBCCW flow since there is no auto start signal for the RBCCW Pump 1-2.

Answer B is Incorrect but plausible. When SOC Loop B senses 360°F, then SOC Pump B ONLY trips. SOC Pump A and RBCCW remain in service.

Answer C is Incorrect but plausible. A lockout on Startup transformer SA results in the loss of 4160 VAC Bus 1A and 480 Explanation VAC Bus 1A2 (which looses RBCCW Pump 1-1 and SOC Pump A). But, EG01 will fast start and load onto 4160 VAC Bus 1C and 480 VAC Bus 1A2 will re-energize and pickup RBCCW Pump 1-1 after 166 seconds. Since SOC Pump B is still running and RBCCW restored, SOC cooling remains, although diminished.

Answer 0 is Incorrect but plausible. A LOCA signal (90" RPV water level AND Hi OW pressure of 2.9 psig) combined with LOOP will auto trip RBCCW and will not allow the auto re-start.

There are no indications of a LOOP. Also, 90" RPV water level OR 2.9 psig OW pressure will isolate SOC. The values provided do not incur a SOC isolation.

2621.828.0.0045 Shutdown Cooling System 205-10446 Identify and explain system operating controls/indications under all conditions.

Question Source (New, Modified, Bank) Bank If BanK or MQdified VISION System/Question 10: 811697 Question Source: ILT Bank Previous 2 Exams: No Memory or X Cognitive Comprehension Fundamental 3:PEO Level or Analysis Knowledge OCSOPS ILT Page: 5 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam NUREG 1021 Appendix B: Predict an Event or Qutcome 10CRF55 55.41b I 7 I 55.43b I

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

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes I Point Value: 1 System ID No.: 205000 PRA:

I No Safety [81 Initial License Level 4

Function: o LORT OCSOPSILT Page: 6 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 3 10: 11-1 NRO 03 Points: 1.00 The plant was at rated power when the following annunciators alarmed:

  • S1A SUDDN PRESS
  • LKOUT RELAY 86/S1A TRIP Which of the following states ALL power supplies that can provide power to the listed Bus, if the main generator tripped AND with NO Operator action?

Bus 1C Bus 10 A. EDG 1 EDG2 CT Bank 6 CT B. EDG 1 EDG2 ONLY Bank 6 ONLY C. EDG 1 EDG2 ONLY Bank 6 CT D. EDG 1 Bank 6 CT CT ONLY Answer: B IAnswer Explanation QID: 11-1 NRO 03 Question # I 3 I Developer / Date: J..IR /5-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO I SRO

EXAMINATION ANSWER KEY oc RO NRC Exam 262001 AC Electrical Distribution K2.01 - Knowledge of electrical power supplies 3.3 3.6 to the followin : Off-site sources of nn\JIIII'"r Level Tier 2 Group 1 General RAP-S1b, S4b ABN-37 BR 3000 References B is Correct. The listed alarms describe a loss of offsite power (transformer SiA or Bank 5) to Bus iA which feeds Bus iC. The remaining power supplies to Bus 1C are EDG 1 only. Power supplies available to power Bus 1D, with NO operator action are startup transformer (or Bank 6), and EDG 2. The Combustion Turbines are always available to supply Bus 1B (and feed Bus 1D) but manual Operator actions are required (ABN-37). Thus, Explanation EDG 1 to Bus iC and EDG 2 and Bank 6 to Bus 1D are ready and available for power, with no operator actions.

All distractors are Incorrect but plausible if the applicant does not recall the correct power supplies during emergency conditions with no operator actions.

Lesson Plan 2621.828.0.0016, Electrical Distribution ACD-i0444, Describe the interlock Signals and setpoints for the Learning affected system components and expected system response includi loss or failed com nents.

Question Source (New, Modified, Bank) I Bank If Bank Qr MQdifiBd VISION System/Question ID: 609063 Question Source: ILT Bank Previous 2 Exams: No OCS OPS ILT Page: 8 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam Memory or X Comprehension Fundamental 2:DR or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: Describing or recognizing Relationships 10CRF55 55.41b I 7 55.43b I

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

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes Point Value: 1 System 10 No.: 262001 PRA: I No Safety ~ Initial License Level 6

Function: D LORT OCSOPSILT Page: 9 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 4 10: 11-1 NRO 04 Points: 1.00 A reactor startup is in progress. Reactor power is 3% when the following annunciator goes into alarm:

  • 24 VDC PP - B PWR LOST Which one of the following conditions will result from this event?

A. Inability to insert/withdraw IRM detectors.

B. Loss of IRM/APRM recorders on Panel4F.

C. A half-scram due to a downscale failure of IRM detectors 15-18.

D. A half-scram due to an inoperative failure of IRM detectors 15-18.

Answer: D Answer Explanation QID: 11-1 NRO 04 Question # I 4 I Developer I Date: JJR I 5-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO SRO 2150031RM K2.01 - Knowledge of electrical power supplies 2.5 2.7 to the following: IRM channels/detectors Level I RO I Tier 2 I Group J 1 General RAP-9XF8d 401.2 ABN-58 References oes OPS ILT Page: 10 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam o is Correct. A loss of 24 VDC Panel B results in an INOP condition on IRMs 15-18 (due to a loss of high voltage) when the Mode Switch is in STARTUP or REFUEL. The indication for IRM instruments 15-18 on Panel 4F also fail downscale. Based on the conditions given (reactor power is 3%), the Mode Switch would be in STARTUP.

Explanation A is Incorrect but plausible since the IRM detector drive motors are powered from 120 VAC PaneIIP-4.

B is Incorrect but plausible since the IRM/APRM recorders on 4F are powered from 120 VAC Panel CIP-3.

C is Incorrect but plausible since IRMs 15-18 fail downscale; the loss of power causes an INOP trip and half-scram when the Mode Switch is in STARTUP or REFUEL.

Lesson Plan 2621.828.0.0029, Nuclear Instrumentation NIS-10444, Describe the interlock signals and setpoints for the Learning affected system components and expected system response ective/ includi loss or failed nents.

Question Source (New, Modified, Bank) Bank If Bank Qr MQdified VISION System/Question 10: 510728 Question Source: ILT Bank Previous 2 Exams: No Memory or X Comprehension Fundamental 3:PEO or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: Predict an Event or Outcome 10CRF55 55.41b I 7 55.43b I

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

OCSOPS ILT Page: 11 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes I Point Value: 1 System 10 No.: 215003 PRA:

I No Safety ~ Initial License Level 7

Function: D LORT OCSOPSILT Page: 12of241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 5 10: 11-1 NRO 05 Points: 1.00 The plant is at rated power when the following annunciator alarmed:

  • RPS MG SET 1 TRIP Which of the following states the impact on indicated power?

A. APRMs 1-4 indicate 100%; APRMs 5-8 indicate 0%

B. APRMs 1-4 indicate 0%; APRMs 5-8 indicate 100%

C. APRMs 1, 3, 5, 7 indicate 0%; APRMs 2, 4, 6, 8 indicate 100%

D. APRMs 1, 3, 5, 7 indicate 100%; APRMs 2, 4, 6, 8 indicate 0%

Answer: B Answer Explanation QID: 11-1 NRO 05 Question # I 5 I Developer / Date: JJR /5-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO SRO 212000 RPS K3.03 - Knowledge of the effect that a loss or malfunction of the REACTOR PROTECTION 3.5 3.6 SYSTEM will have on following: Average power range monitoring system: Plant-Specific level I RO I Tier 2 J Group I 1 General RAP-9XF3a References OCSOPSILT Page: 13 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam B is Correct. The annunciator provided shows a loss of RPS MG Set 1. This failure results in the loss of APRMs 1-4, which will Explanation indicate 0% power. APRMs 5-8 will continue to indicate normal.

All other distractors are Incorrect but plausible if the candidate does know how the in the APRMs.

Lesson Plan 2621.828.0.0029, Nuclear Instrumentation NIS-10444, Describe the interlock signals and setpoints for the Learning affected system components and expected system response o r loss or failed """'rnn"n,~n'~~

Question Source (New, Modified, Bank) Bank If Banis Qr Modified VISION System/Question 10: 663329 Question Source: ILT 08-1 NRC Exam Previous 2 Exams: No Memory or X Comprehension Fundamental 1:1 or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: interlocks, setpoints, or system (singular) response 10CRF55 55.41b I 7 55.43b I

Design, components, and functions of control and safety systems, Content including instrumentation, Signals, interlocks, failure modes, and automatic and manual features.

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes Point Value: 1 System 10 No.: 212000 PRA:

I No Safety 7

181 Initial License Level Function: D LORT OCSOPS ILT Page: 14 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 6 ID: 11*1 NRO 06 Points: 1.00 The plant was at rated power when an Electric ATWS occurred. SP-16, Bypassing MSIV Lo-Lo level Isolation Interlocks And The RBCCW Interlocks, has been executed by the crew.

With the above conditions, if a complete loss of Instrument Air were to occur, which of the following MSIVs, if any, would lose their CURRENT pneumatic supply?

A. Inboard ONLY B. Outboard ONLY C. BOTH the inboard AND outboard D. NEITHER the inboard NOR outboard Answer: B Answer Explanation QID: 11*1 NRO 06 Question # I 6 I Developer I Date: JJR I 5-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO SRO 300000 Instrument Air K3.01 - Knowledge of the effect that a loss or malfunction of the (INSTRUMENT AIR 2.7 2.9 SYSTEM) will have on the following:

Containment air system Level I RO I Tier 2 I Group I 1 General EMG-SP16 ABN*35 References

EXAMINATION ANSWER KEY oc RO NRC Exam B is Correct. The normal pneumatic supply to the Containment (Orywell) air system at power is Nitrogen. The normal air supply to the outboard MSIVs is Instrument Air and inboard MSIVs is Nitrogen. If an event occurred where Nitrogen supply to the Orywell was lost, the Nitrogen supply would isolate automatically and Instrument Air would supply Orywell (Containment) Air system loads. If a non-ATWS complete loss of Instrument Air were to occur, however, both the inboard and outboard MSIVs would close due to V-6-395, MSIV Isolation Signal Bypass Valve, Explanation closing, securing Nitrogen to the Oryweilioads too. Since SP-16 was performed, MSIV EOP interlocks have been bypassed, V-6 395 was placed in Bypass, and during a complete loss of Instrument Air, the MSIVs would still have Nitrogen Supply to them, and will remain open.

All distractors are Incorrect but plausible if the applicant does not recall the effects of completing SP-16 on the Orywell Air System or recall the interrelationship and interlocks between Instrument and Air.

Lesson Plan 2621.845.0.0053, RPV Control - with A TWS Learning EWA-2257, Given the EOP, describe in detail each Objective/ step/statement, including the technical basis, and how to verify or each Question Source (New, Modified, Bank) New If Bank or Mgdified VISION System/Question 10:

Question Source: N/A Previous 2 Exams:

Memory or X Comprehension Fundamental 3:SPK or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: Solve a Problem using Knowledge and its meaning OCSOPS ILT Page: 16 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 10CRF55 55.41b I 7 I 55.43b I

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

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes I Point Value: 1 System 10 No.: 300000 PRA: I No Safety [81 Initial License Level Function:

8 o LORT OCSOPS IlT Page: 17 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 7 10: 11*1 NRO 07 Points: 1.00 The plant is shutdown.

Which of the following shows the correct auto-start condition for the associated plant system pump? (Note: LOOP is Loss of Offsite Power; LOCA is Loss of Coolant Accident)

A. ESW: LOOP B. TBCCW: Low system pressure C. Service Water: Combined LOOP AND LOCA D. RBCCW: Combined Low system pressure AND LOOP Answer: B Answer Explanation QID: 11-1 NRO 07 Question # I 7 I Developer / Date: JJR /5-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO SRO 400000 Component Cooling Water K4.01 - Knowledge of CCWS design feature(s) 3.4 3.9 and or interlocks which provide for the followin J: Automatic start of standby pump Level RO I Tier 2 I Group I 1 General RAP-Q5f, Q1f References OCSOPS ILT Page: 18 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam B is Correct. TBCCW will auto start on low system pressure (79 pSig w/10 second time delay). TBCCW pumps trip on a LOOP and are not restored when EDGs start and load.

A is Incorrect but plausible. The applicant may not recall the ESW pumps have no auto start feature associated with the Explanation LOOP.

C is Incorrect but plausible since Service water will auto start from a LOOP, but not from a combined LOOP + LOCA.

D is Incorrect but plausible. RBCCW has no auto start from rs but does auto start from a LOOP.

Lesson Plan 2621.828.0.0048, TBCCW TBC-10443: Given the system logic/electrical drawings describe Learning the system component starts or trips [breaker logic] and Objective/ expected system response including power loss or failed com Question Source (New, Modified, Bank) Bank If Bank or MQdified VISION System/Question ID: 718183 Question Source: ILT Bank Previous 2 Exams: No Memory or X Comprehension Fundamental 2:DR or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: Describing or recognizing Relationships 10CRF55 55.41b I 7 55.43b I

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

OCSOPSILT Page: 19 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1N2 minutes I Point Value: 1 System 10 No.: 400000 PRA:

I No Safety [81 Initial License Level 8

Function: D LORT OGS OPS ILT Page: 20 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 8 10: 11*1 NRO 08 Points: 1.00 The plant was at rated power when an event occurred. The following conditions exist:

  • 4160 VAC Bus 1C is de-energized
  • Attempts to Fast Start EDG-1 from the Control Room have failed An operator has manually started EDG-1 from its local cubicle lAW procedure 341, Emergency Diesel Generator Operation, section 8 'Manual Control for Deadline Pickup From The Diesel Generator Switchgear'.

A steady state loading condition has been attained lAW the procedure.

Which of the following is the correct response to EDG-1 FREQUENCY and KILOWATTS if the EDG-1 GOVERNOR CONTROL switch at the local panel is momentarily placed in the LOWER position for 1 second?

FREQUENCY KILOWATTS A. lower remain constant B. lower lower C. remain constant lower D. remain constant remain constant Answer: A IAnswer Explanation QID: 11*1 NRO 08 Question # I 8 I Developer I Date: JJR /5-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO I SRO OCSOPS ILT Page: 21 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 264000 EDGs K4.06 - Knowledge of EMERGENCY GENERATORS (DIESEUJET) design feature(s) 2.6 2.7 and/or interlocks which provide for the followi : Governor control Level RO Tier 2 Group 1 General GFES: Motors and 341 section 8 References Generators A is Correct. When the EDG is not in parallel with another generator/grid/electric power source, the GOVERNOR CONTROL switch controls EDG Frequency (Speed). The VOLTAGE CONTROL switch controls EDG Kilowatts (KW) (real load) in this situation. Placing the GOVERNOR CONTROL switch in LOWER will cause EDG Speed to lower but will have no affect on EDG Kilowatts.

Answer B is Incorrect. EDG KW will remain constant. Distractor Explanation is plausible if the candidate believes that load will lower.

Answer C is Incorrect. EDG Speed will lower. Distractor is plausible if the candidate believes that speed will lower.

Answer D is Incorrect. EDG Speed will lower. Distractor is plausible if the candidate believes that a 1 second switch manipulation will not have a significant affect on frequency or KW.

Lesson Plan 2621.828.0.0013, Emergency Diesel Generators EDG-10446, Identify and explain system operating controls /

Learning indications under all plant operating conditions.

Question Source (New, Modified, Bank)

If Bank Qr MQdified I Bank VISION System/Question ID: 811748 Question Source: ILT Bank Previous 2 Exams: No OCSOPS lLT Page: 22 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam Memory or X Comprehension Fundamental 1:1 or Analysis Knowledge Cognitive Level NLiREG 1021 Appendix B: Interlocks, setpoints, or system (singular) response 10CRF55 55.41b I 7 55.43b I

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

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes I Point Value: 1 System ID No.: 264000 PRA: I No Safety ~ Initial License Level 6

Function: D LORT OCSOPS ILT Page: 23 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 9 10: 11-1 NRO 09 Points: 1.00 The plant is starting up with the REACTOR MODE SELECTOR switch in STARTUP. All SRMs indicate between 50 - 100 CPS.

While withdrawing the 8th control rod, the following annunciator alarmed:

  • SRM DNSCl The Operator reports the following indications:
  • SRM All IN light extinguished
  • IRM All IN light is in alarm Which ONE of the following states the plant response to the conditions listed above AND what directly caused the alarm and indications above?

Plant Response Cause of the Event A. Rodblock SRM downscale B. Rodblock SRM not fully inserted C. Alarm ONLY SRM not fully inserted D. Rodblock IRM downscale Answer: B Answer Explanation QID: 11-1 NRO 09 Question # I 9 I Developer / Date: JJR /5-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO I SRO

EXAMINATION ANSWER KEY oc RO NRC Exam 215004 Source Range Monitor K5.01 - Knowledge of the operational implications of the following concepts as they 2.6 2.6 apply to SOURCE RANGE MONITOR (SRM)

SYSTEM : Detector 0 on Level RO 2 Group 1 General RAP-G5d RAP-H7a References B is Correct. The question stem shows a plant startup when the SRM downscale annunciator alarms. The impact of this alarm is an alarm only, if the associated SRM is fully inserted. If the associated SRM is not fully inserted, then a rodblock is applied.

With all SRMs fully inserted, the SRM ALL IN light will be energized. With this light extinguished, then at least 1 SRM is not full in. The operator cannot tell which SRM is not fully inserted, only that 1 or more are not fully inserted. A non-fully inserted SRM would produce a reduction in counts, and thus, the SRM downscale annunciator is expected when counts go down enough. The rod block RAP (H7a) also states that a rod block is inserted when any SRM is < 500 cps and not fully inserted with the Mode Switch in STARTUP.

Explanation A is Incorrect but plausible. Even though a Rodblock does exist, the cause of the event was a SRM not fully inserted, not an SRM downscale condition C is Incorrect but plausible. SRM not fully inserted does not energize any audible alarm but does produce a visual alarm on the rod block panel. But the given conditions also produce a rod block.

D is Incorrect but plausible if the applicant does not recognize that an IRM downscale is an expected condition on a startup. Any impact from a downscale IRM is bypassed while in Range 1, which is where the IRMs should currently be.

oes OPS ILT Page: 25 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam Lesson Plan 2621.828.0.0029, Nuclear Instrumentation NIS-10444, Describe the interlock signals and setpoints for the Learning affected system components and expected system response Objectivel including power loss or failed components.

Question Source (New, Modified, Bank) Modified If Bank or Modified VISION System/Question 10: 667517 Question Source: ILT 08-1 Audit Exam Previous 2 Exams: No Memory or X Comprehension Fundamental 3:SPK or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: .solve a f.roblem using Knowledge and its meaning 10CRF55 55.41b I 6 55.43b I

Content Design, components, and functions of reactivity control mechanisms and instrumentation.

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes I Point Value: 1 System 10 No.: 215004 PRA:

I No Safety [81 Initial License Level 7

Function: D LORT OCSOPSILT Page: 26 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 10 Points: 1.00 Given the following plant conditions:

  • Both sets of ADS timers have initiated due to a LOCA.
  • Drywell pressure drops to 2.5 psig due to containment failure.
  • ADS Timer "A" bypass switch has been taken to BYPASS.
  • ADS Timer "B" bypass switch cannot be repositioned from AUTO due to switch failure.
  • NO other operator actions are taken.

Based on these plant conditions, the Automatic Depressurization System (ADS) will A. NOT initiate because it is bypassed.

B. initiate and ALL 5 EMRVs will open.

C. initiate but "AU and "D" EMRVs will open ONLY.

D. NOT initiate because drywell pressure is 2.5 psig.

Answer: B Answer Explanation QID: 11-1 NRO 10 Question # I 10 J Developer I Date: JJR I 5-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO SRO 218000 ADS K5.01 - Knowledge of the operational implications of the following concepts as they 3.8 3.8 apply to AUTOMATIC DEPRESSURIZATION SYSTEM: ADS logic 0 )eration Level I RO Tier 2 I Group I 1 General GE729Ei82 RAP-Big References OCSOPS ILT Page: 27 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam B is Correct. ADS is actuated on simultaneous occurrence of high drywell pressure (> 2.9 psig), 10-10-10 reactor water level (64.6" TAF), and core spray system operation as verified by a differential pressure across the core spray booster pump (DP >

about 30 psid) after the ADS timers have timed out for 105 seconds. Upon initiation all five EMRVs open in a staggered fashion within 5 seconds.

A is Incorrect but plausible. Both bypass switches must be taken Explanation to bypass in order to prevent ADS initiation following the 105 second time delay.

C is Incorrect but plausible since the "A" timer is bypassed. It is true that ADS will initiate, however, after the "A" and "0" EMRVs initially open the other valves will open following a short time delay.

o is Incorrect but plausible. Hi drywell pressure is a seal-in contact and must be reset to prevent the initiation from Lesson Plan 2621.828.0.0005, Automatic Depressurization System ADS-368, Describe the EMRV initiation logic for both Learning overpressure operation and operation in the ADS mode. Include Objective/ the following: 1. Initiation signals and setpoints 2. Timers and

3. Control switches 4. Panel indications Question Source (New, Modified, Bank) Bank If Bank Q[ MQdified VISION System/Question 10: 505517 Question Source: ILT Bank Previous 2 Exams: No Memory or X Com prehension Fundamental 3:PEO or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: Predict an Event or Outcome

EXAMINATION ANSWER KEY oc RO NRC Exam 55.41b I 5 I 55.43b I

Facility operating characteristics during steady state and transient 10CRF55 conditions, including coolant chemistry, causes and effects of Content temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes J Point Value: 1 System 10 No.: 218000 PRA:

I No Safety 3

181 Initial License Level Function: D LORT OCSOPS ILT Page: 29 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 11 10: 11*1 NRO 11 Points: 1.00 Which one of the following statements describes what happens on a loss of Vital MCC 1B2?

=

(ATS Automatic Transfer SWitch)

(CIP = Continuous Instrument Panel)

A. ATS IT-3 will transfer CIP-3 to the alternate power source and then back to the rotary inverter after the DC motor starts.

B. ATS IT-3 will transfer CIP-3 to the alternate power source but will NOT transfer back to the rotary inverter after the DC motor starts.

C. The rotary inverter AC motor will continue to run ensuring power to CIP-3 is not interrupted; ATS IT-3 will NOT transfer to the alternate power source.

D. The rotary inverter DC motor will automatically start ensuring power to CIP-3 is not interrupted; ATS IT-3 will NOT transfer to the alternate power source.

Answer: D IAnswer Explanation QID: 11-1 NRO 11 Question # I 11 I Developer 1 Date: JJR 15-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO SRO 262002 UPS (AC/DC)

K6.01 - Knowledge of the effect that a loss or malfunction of the following will have on the 2.7 2.9 UNINTERRUPTABLE POWER SUPPLY (A.C.lO.C.) : A.C. electrical power Level I RO I Tier I 2 I Group I 1 OCSOPS ILT Page: 30 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam General ABN-51 References D is Correct. The rotary inverter DC motor will start, maintaining an uninterrupted generator output (due to the flywheel); ATS IT-3 will not transfer. CIP-3 is a UPS at Oyster Creek.

A is Incorrect but plausible. ATS IT-3 will not transfer since the Explanation rotary output power remains essentially constant.

B is Incorrect but plausible. ATS IT-3 will not transfer since the rotary output power remains essentially constant.

C is Incorrect but plausible. A loss of VMCC-1 B2 results in loss of to the inverter AC motor.

Lesson Plan 2621.828.0.0056, Vital AC Distribution Learning VAC-10441, Given the system logic/electrical drawings, Objective/ describe the system trip signals, setpoints and expected includi loss or failed comn""n.:.."'~.,.

Question Source (New, Modified, Bank) Modified If Bank or Mgdified VISION System/Question ID: 505413 Question Source: ILT Bank Previous 2 Exams: No Memory or X Comprehension Fundamental 1:1 or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: lnterlocks, setpoints, or system (singular) response 10CRF55 55.41b I 7 55.43b I

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

OCSOPS ILT Page: 31 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes I Point Value: 1 System 10 No.: 262002 PRA:

I No Safety ~ Initial License Level Function:

6 o LORT

EXAMINATION ANSWER KEY oc RO NRC Exam 12 Points: 1.00 The plant was at rated power when a STATION BLACKOUT occurred.

Under the conditions:

1. Can the Isolation Condenser System be manually initiated from the Control Room?
2. Can makeup water be provided to the Isolation Condenser shells (includes both Control Room and local actions)?

1 A. Yes No B. Yes Yes C. No No D. No Yes Answer: B Answer Explanation QID: 11-1 NRO 12 Question # I 12 I Developer / Date: JJR /5-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO SRO 207000 Isolation (Emergency) Condenser K6.07 - Knowledge of the effect that a loss or malfunction of the following will have on the 3.0 3.2 ISOLATION (EMERGENCY) CONDENSER:

A.C. ~ower: BWR-2,3 Level I RO I Tier 2 I Group I 1 General ABN-37 307 References OCS OPS ILT Page: 33 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam B is Correct. The plant was at power when a station blackout occurred. There is no AC power in the station. In the normal configuration, the steam admission valves to each IC are open, one condensate return valve is open, and the second condensate return valve is closed. The closed valve is DC powered and can be manipulated with a loss of AC power.

Filling of the shells usually requires AC power to a water pump.

With AC gone, these AC powered pumps are lost. But the shells can also be filled by the Fire Protection water system, which under the given conditions, will be pressurized by diesel driven fire pumps. The makeup valves are air operated, with air Explanation accumulators, and fail closed on loss of air. Even if the accumulators discharged, they can be manually manipulated in the plant locally.

Therefore, the isolation condensers can be initiated in the control room and the shells can be filled from fire protection with the total loss of AC power, All distractors are Incorrect but plausible if the applicant does not recall about the use of fire protection water supplied by the fire diesels or power supplies to the system valves and the normal standby lineup.

Lesson Plan 2621.828.0.0023, Isolation Condenser System ICS-2338, Given plant conditions, evaluate the impact on the Learning Isolation Condenser System and the plant.

Ob ective/

Question Source (New, Modified, Bank) I Bank If Bank or Modified VISION System/Question ID: 663375 Question Source: ILT Bank Previous 2 Exams: No OCSOPS ILT Page: 34 of 241 02 April 2012

EXAMINATION ANSWER KEY DC RD NRC Exam Memory or X Comprehension Fundamental 1:F or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: Eacts 10CRF55 55.41b I 7 55.43b I

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

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes I Point Value: 1 System 10 No.: 207000 PRA:

I No Safety 181 Initial License Level Function:

4 o LORT OCSOPSILT Page: 35 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 13 10: 11-1 NRO 13 Points: 1.00 The plant is at 100% power with the following conditions:

  • Battery "A" is on the Battery Charger M-G "A"
  • Battery "B" is on the A-B Static Charger The breaker from USS 1B2 to MCC 1B21 then opens.

Based on the above conditions, with no operator action, how is battery voltage affected?

Voltage indication for ...

A. "A" battery will begin to lower ONLY.

B. liB" battery will begin to lower ONLY.

C. "A" & liB" battery will begin to lower.

D. "A" & "B" battery will NOT change.

Answer: B Answer Explanation QID: 11-1 NRO 13 Question # I 13 I Developer 1 Date: JJR 15-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO SRO 263000 DC Electrical Distribution A 1.01

  • Ability to predict and/or monitor changes in parameters associated with 2.5 2.8 operating the D.C. ELECTRICAL DIS'rRIBUTION controls including: Battery chargin~ Idischarging rate Level RO I Tier I 1 I Group I 1 OCSOPS ILT Page: 36 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam General RAP-U3e References B is Correct. lAW RAP U-3-e, AlB BAT CHG DRV MOT TRIP, loss of MCC 1B21 with B battery on static charger will place DC-B bus directly on the battery and its voltage indication will begin to lower.

Explanation A & C are Incorrect but plausible if the applicant doesn't recall A Battery is still on its MG set (VMCC 1B2) and its bus voltage will not lower.

o is Incorrect but plausible if the applicant doesn't recall B battery voltage will begin to lower.

Lesson Plan 2621.828.0.0012, DC Distribution DCD-10445, Given a set of system indications or data, evaluate Learning and interpret them to determine limits, trends and system Objective/ status.

Question Source (New, Modified, Bank) Bank If Bank Qr MQdified VISION System/Question 10: 510794 Question Source: ILT Bank Previous 2 Exams: No Memory or Comprehension X

Fundamental 2:RI or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: Recognizing interaction between systems (plural), including consequences and implications (AC and DC system interaction)

OCSOPS ILT Page: 37 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 55.41b I 5 I 55.43b I

Facility operating characteristics during steady state and transient 10CRF55 conditions, including coolant chemistry, causes and effects of Content temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operati.,g characteristics.

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes I Point Value: 1 System 10 No.: 263000 PRA:

I No Safety 6

181 Initial License Level Function: o LORT OCSOPSILT Page: 38 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 14 10: 11-1 NR014 Points: 1.00 The plant was at rated power when an event occurred. The BOP observed the following indications on Panel 3R:

D FOR CLARITY, INDICATING LIGHTS HIGHLIGHTED ARE LIT Based on these indications, which of the following Annunciator Panel G indications correctly corresponds to the Panel 3R indications above?

OCSOPSILT Page: 39 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam A.

FUEL STDBY REACTOR POOL LIQ ------N-EU-T-RO-N-CNTRL RPS MONITORS oes OPS ILT Page: 40 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam B.

STDBY REACTOR FUEL LIQ POOL CNTRL RPS NEUTRON MONITORS CHANNELl

~.VJL\IIS _-WI'JUItUu:1 C~II OPE" 8RM fRill

....HI HI t~l.SiEt; GRill IRtII Hl/W Hl/INQP ONSCL SRM DNS9L OCSOPS ILT Page: 41 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam C.

FUEL STDBY REACTOR POOL LIQ ------NE-U-T-RO-N-CNTRL RPS MONITORS OCSOPS ILT Page: 42 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam D.

FUEL STDBY REACTOR POOL LIQ - - - - - - - . - - - -

CNTRL RPS NEUTRON MONITORS Answer: C Answer Explanation QID: 11-1 NRO 14 Question # I 14 I Developer / Date: JJR /5-14-2012 Knowledge and Ability Reference Information K&A I Importance Rating OCSOPSILT Page: 43 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 215005 APRM / LPRM A 1.05 - Ability to predict and/or monitor changes in parameters associated with 3.3 3.2 operating the AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM controls includin : Li and alarms Level RO 2 Group 1 General RAP-G1c, G6f, G1f References C is Correct. The indications provided show an upscale trip of LPRM 44-25C. At rated power, this will result in this LPRMs corresponding APRM to indicate HI-HI. For this failure, annunciators G1c, G1d, G1f, G3f, and G6f will all be in the alarm condition on annunciator panel G.

A is Incorrect but plausible. The applicant may recognize an LPRM has failed upscale but not recognize this results in its Explanation corresponding APRM failing upscale. At low power levels, this is true, but not at rated power.

B is Incorrect but plausible. The applicant may not recognize annunciator G6f is not in alarm, which it would be based on the Panel 3R indications.

o is Incorrect but plausible if the applicant confuses which RPS was affected.

Lesson Plan 2621.828.0.0029, Nuclear Instrumentation NIS-10445, Given a set of system indications or data, evaluate Learning and interpret them to determine limits, trends and system ective/ status.

Question Source (New, Modified, Bank)

If Bank or MQdified I New VISION System/Question 10:

Question Source: N/A Previous 2 Exams:

OCSOPS ILT Page: 44 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam Memory or X Comprehension Fundamental 3:SPK or Analysis Knowledge Cognitive Level NliREG 1021 Appendix B: .solve a Problem using Knowledge and its meaning 55.41b I 5 55.43b I

Facility operating characteristics during steady state and transient 10CRF55 conditions, including coolant chemistry, causes and effects of Content temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes I Point Value: 1 System 10 No.: 215005 PRA: I No Safety ~ Initial License Level 7

Function: D LORT OCSOPS ILT Page: 45 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 15 ID: 11-1NR015 Points: 1.00 The plant is at 20% power during an ascension to rated power. An event then occurs resulting in the crew executing Emergency Depressurization (ED). Plant conditions include the following:

  • All Control Rod indications on Panel 4F indicate a green backlight
  • All EMRV Control switches on Panel 1F/2F are in MAN
  • Reactor Pressure indicates 5 psig
  • RPV Water Level indicates 165 inches
  • Torus Pressure indicates 1.5 psig What is the correct status of all EMRV acoustic indications on Panel 1F/2F AND required action (lAW the ED procedure) associated with the EMRVs, if any?

All EMRVs A~oU&ti~s Indicate In The ... Required Action A. VALVE OPEN REGION Place All EMRVs in AUTO B. VALVE CLOSED REGION Leave All EMRVs in MAN C. VALVE OPEN REGION Leave All EMRVs in MAN D. VALVE CLOSED REGION Place All EMRVs in AUTO Answer: B IAnswer Explanation QID: 11-1 NRO 15 Question # I 15 J Developer / Date: JJR /5-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO I SRO OCSOPSILT Page: 46 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 239002 SRVs A2.05 - Ability to (a) predict the impacts of the following on the RELIEF/SAFETY VALVES; and (b) based on those predictions, use 3.2 3.4 procedures to correct, control, or mitigate the consequences of those abnormal conditions or Low reactor Level RO 2 Group 1 General ED-no A TWS EOP EOP User's Guide References B is Correct. The question stem provides a condition where all EMRVs have been manually opened for ED. When RPV pressure lowers to where there is < 50 psid between the RPV and Torus, the EMRVs will close. The ED procedure has the operator leave the EMRVs in MAN until the ED procedure has been exited.

A is Incorrect. This distractor is plausible if the applicant does not recall that EMRVs solenoid indication will indicate closed when there is < 50 psid between RPV pressure and Torus pressure. In addition, the ED procedure has the crew leave all Explanation EMRVs in MAN.

C is Incorrect. This distractor is plausible if the applicant does not recall that EMRVs solenoid indication will indicate closed when there is < 50 psid between RPV pressure and Torus pressure.

o is Incorrect. This distractor is plausible if the applicant does not recall that the ED procedure has the crew leave all EMRVs in MAN.

Lesson Plan 2621.845.0.0054, Emergency Depressurization Learning EED-9572, Given a copy of the ED EOP, describe the technical basis for each or conditional statement of the Question Source (New, Modified, Bank) Bank oes OPS ILT Page: 47of241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam If Bank or Modified VISION System/Question 10: 811782 Question Source: ILT 10-1 NRC Exam Previous 2 Exams: Yes Memory or X Comprehension Fundamental 3:PEO or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: Predict an Event or Outcome 10CRF55 55.41b I 10 55.43b I

Content Administrative, normal, abnormal, and emergency operating procedures for the facility.

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes I Point Value: 1 System 10 No.: 239002 PRA:

I No Safety 3

IZI Initial License Level Function: D LORT OCSOPSILT Page: 48 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 16 ID: 11-1 NRO 16 Points: 1.00 The plant was at rated power when an event occurred requiring entry into ABN-58, Instrument Power Failures.

Which ONE of the following correctly describes a plant impact AND the action required to correct it?

Plant Impact Required Action Perform actions in ABN-58 for a loss of...

A. EXACTLY 1/2 of the MSIV LEDs inside VACP-1.

Panel 11 F are OFF B. ALL CRD HYDRAULIC SYSTEM analog VACP-1.

meters on Panel 4F are DOWNSCALE C. EXACTLY 1/2 of the MSIV LEDs inside Panel 1 'I F are OFF D. ALL CRD HYDRAULIC SYSTEM analog CIP-3.

meters on Panel 4F are DOWNSCALE Answer: C Answer Explanation QID: 11-1 NRO 16 Question # I 16 I Developer / Date: JJR /5-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO I SRO OCSOPSILT Page: 49 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 223002 PCIS/Nuclear Steam Supply Shutoff A2.06 - Ability to (a) predict the impacts of the following on the PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT -OFF; and (b) based on those 3.0 3.2 predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Containment instrumentation failures Level I RO I Tier 2 I Group I 1 General RAP-J1a, J8b ABN-58 References OCSOPS ILT Page: 50 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam C is Correct. The MSIVs are controlled by solenoids (to allow air/nitrogen operation). There is an AC solenoid for each MSIV (powered from CIP-3) and a DC solenoid for each MSIV (from OC o or DC-F). To close the MSIVs, both the DC solenoid and the AC solenoid must de-energize for each MSIV. Inside Panel 11 F, there are 4 sets of 2 LEOs: one set per MSIV. One LED is powered from the AC solenoid power supply, and the second LED per MSIV, is powered from the DC solenoid power supply. All LEOs are lit when all power supplies are normal. When CIP-3 is lost, the AC solenoid to each MSIV is de-energized and the respective LED for each MSIV goes out. Therefore, with CIP-3 lost, 1/2 of the LEOs are out, and the other 1/2 of the LEOs are energized (from DC power). The correct action is for the applicant to recognize that a loss of CIP-3 would result in this paticular containment instrument failure and the correct action to correct it is so perform actions required by ABN-58 for a loss of CIP-3.

Explanation A is Incorrect but plausible if the applicant does not recall the correct power supply which would result in a loss of AC MSIV LEOs. VACP-1 is a vital AC power source and also has its own actions within ABN-58.

B is Incorrect but plausible. The loss of CIP-3 will result in a loss of CRO RETURN FLOW INO on Panel 4F and render the Reactor Manual Controls System inoperable. The applicant may assume that CRO Hydraulic System analog meter indications were also affected and confuse these with a loss ofVACP-1, also being a vital power supply to many indicators in the Control Room.

o is Incorrect but plausible. The loss of CIP-3 will result in a loss of CRD RETURN FLOW INO on Panel 4F and render the Reactor Manual Controls System inoperable. The applicant may assume that CRD Hydraulic System analog meter indications were also anlectea. which are not.

Lesson Plan 2621.828.0.0026, Main Steam System Learning MSS-10446, Identify and explain system operating controls /

indications under all conditions.

Question Source (New, Modified, Bank) New OCS OPS ILT Page: 51 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam If Bank Q[ MQdified VISION System/Question 10:

Question Source: N/A Previous 2 Exams:

Memory or X Comprehension Fundamental 2:DR or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: Describing or recognizing Relationships 10CRF55 55.41b I 7 55.43b I

Design, components, and functions of control and safety systems, Content including instrumentation, Signals, interlocks, failure modes, and automatic and manual features.

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes I Point Value: 1 System 10 No.: 223002 PRA:

I No Safety ~ Initial License Level Function:

5 o LORT OCSOPS ILT Page: 52 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 17 ID: 11-1 NRO 17 Points: 1.00 The plant is at rated power. An event with a radioactive source has resulted in the following conditions:

  • REACTOR BUILDING VENT MANIFOLD #1 indicates 14 mRlhr
  • REACTOR BUILDING VENT MANIFOLD #2 indicates 12 mRlhr
  • Annunciator RX BLDG - VENT HI is in alarm Which of the following states the correct Control Room indications from this event after all automatic action(s) have occurred?

A. RX BLDG DIFFERENTIAL PRESS indicates a slightly positive LlP B. STANDBY GAS OUTLET TEMP shows a higher than normal temperature C. SGTS CROSSTIE valve V-28-48 indicates red light ON and green light OFF D. REACTOR BUILDING VENT MANIFOLD #1 and #2 indicate a valid rising dose rate Answer: B Answer Explanation QID: 11-1 NRO 17 Question # I 17 I Developer / Date: JJR /5-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO SRO 261000 SGTS A3.04 - Ability to monitor automatic operations 3.0 3.1 of the STANDBY GAS TREATMENT SYSTEM includin I: System tem perature Level RO Tier 2 I Group I 1 General RAP-10F1f BR 2011 sh. 2 GU 3E*822*21-1000 References OCSOPS ILT Page: 53 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam B is Correct. Since the SGTS takes a suction on RB atmosphere, which includes steam, plus heaters in the STGS, and the energy from decay of radioactive particles in the stream, the SGTS discharge air temperature will be greater than normal.

A is Incorrect but plausible. The indications provided in the stem will result in the isolation of the normal RB HVAC System and the auto start of the SGTS. SGTS is designed to maintain a negative Explanation RB .6P.

C is Incorrect but plausible. When SGTS auto starts, the SGTS CROSSTIE valve V *28-48 closes (green light ON).

o is Incorrect but plausible. Since there is no air flow past the RB vent manifold radiation monitors when the normal RB HVAC isolates, the reading cannot be considered as valid reading of the RB ~tn'lI\c:,nnArA Lesson Plan 2621.828.0.0042, Secondary Containment and SGTS SGT-10445, Given a set of system indications or data, evaluate Learning and interpret them to determine limits, trends and system ective/ status.

Question Source (New, Modified, Bank) Bank If Bank Qr MQdified VISION System/Question 10: 608590 Question Source: ILT Bank Previous 2 Exams: No Memory or Comprehension X

Fundamental 3:PEO or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: Predict an Event or .outcome 10CRF55 55.41b I 7 55.43b I

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

OCSOPS ILT Page: 54 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes I Point Value: 1 System 10 No.: 261000 PRA:

I No Safety 9

IZI Initial License Level Function: D LORT OCSOPS ILT Page: 55 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 18 10: 11-1 NRO 18 Points: 1.00 The plant was at rated power when an event occurred. Present plant conditions are as follows:

  • Drywell pressure is 3.6 psig and rising
  • RPV water level is 120" and rising
  • FEED PUMPS DISCHARGE PRESSURE indicates 800 psig The Operator notes the following Core Spray System indications:
  • MAIN PUMP AMPS NZ01A indicates 50 AC AMPERES
  • MAIN PUMP AMPS NZ01D indicates 0 AC AMPERES
  • SYS 1 FLOW indicates approximately 100 GPM
  • SYS 2 PUMP DISCH PRESS BOOSTERS indicates approximately 330 psig Which of the following is correct regarding the observed Core Spray indications?

A. Core Spray Pump NZ01D has tripped.

B. Core Spray Pump NZ01A is running on minimum flow.

C. Core Spray System 2 is NOT indicating the expected discharge head.

D. Core Spray System 1 CANNOT provide core cooling when the RPV depressurizes.

Answer: B Answer Explanation QID: 11-1 NRO 18 Question # I 18 I Developer I Date: JJR /5-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO SRO 209001 LPCS A3.02 - Ability to monitor automatic operations 3.8 3.7 of the LOW PRESSURE CORE SPRAY SYSTEM including: Pump start Level I RO I Tier I 2 I Group I 1 OCSOPSILT Page: 56 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam General 341 RAP-B1e, B2e UFSAR 6.3.1.3.3 References B is Correct. The question stem describes the plant at power when an event resulted in a low RPV water condition and a high drywell pressure condition. Under the given conditions, core spray 1 (main pump A and booster pump a) and core spray 2 (main pump B and booster pump B) will start. With feedwater discharge pressure at 800 pSig, then RPV pressure is close to this value. With core spray running at an RPV pressure> 305 psig, the core spray parallel isolation valves are closed and core spray is running on minimum flow back to the torus. This flow is approximately 100 gpm. Therefore, core spray A has started and is running on minimum flow.

Explanation A is Incorrect but plausible. As stated, core spray A and B start on their signals. Core spray C and 0 will still be in standby (off),

unless a preferred core spray system fails. Since there is no indication of this in the question stem, then core spray 0 will be off and no amps is the expected condition - not tripped.

C is Incorrect but plausible. With core spray system B running on minimum flow, the discharge pressure is approximately as listed in answer C.

o is Incorrect but plausible since the provided indications are the expected indications, and core spray A will provide core cooling, as desi when RPV < 305 Lesson Plan 2621.828.0.0010, Core Spray System Learning CSS-10444, Describe the interlock Signals and setpoints for the Objective/ affected system components and expected system response includ loss or failed ,.nllTll"l,nnAn,rc:

Question Source (New, Modified, Bank)

If Bank or MQdifhl~d I Bank VISION System/Question 10: 609285 Question Source: ILT Bank Previous 2 Exams: No QCSQPS ILT Page: 57 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam Memory or X Comprehension Fundamental 3:PEO or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: .fredict an Event or .outcome 10CRF55 55.41b I 7 55.43b I

Design, components, and functions of control and safety systems, Content including instrumentation, Signals, interlocks, failure modes, and automatic and manual features.

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes I Point Value: 1 System 10 No.: 209001 PRA:

I No Safety 2&4 181 Initial License Level Function: o LORT OCSOPS ILT Page: 58 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 19 10: 11-1 NRO 19 Points: 1.00 The plant is at rated power. The following Panel4F indications are observed:

STANDBY UQUID CONTROL PIIIIIII ~II "1I1II~1I1111~111I1.11 ~ 111111111 o 1 )( UDlGAUONS 234 5 o

11111'3001'1'1GOO'1'1'1 1111'11200

$00

'1'1'1150011

""S""C'"

An event then occurred which resulted in an Electric ATWS. Actions required by RPV Control- with ATWS are being implemented by the crew.

lAW the EOP User's Guide, which of the following Panel 4F indications is the FIRST to indicate the reactor will remain shutdown under ALL conditions, regardless of control rod position or RPV water temperature?

EXAMINATION ANSWER KEY oc RO NRC Exam A.

STANDBY UQUID CONTROL o

'IIIIIII~111I~lIIqllllllllllllllllll.1I1111II11 2 3 4 5

)( f!DlGAlIDNS OCSOPSILT Page: 60 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam B.

STANDBY LIQUID CONTROL 111'11111111111111111' 111111111  !

o ~O 600 SOO

,.a,u_el'14 1200 1$00 OCSOPS ILT Page: 61 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam C.

STANDBY UQUID CONTROL OCS OPS ILT Page: 62 of 241 02 April 2012

EXAMINATION ANSWER KEY DC RD NRC Exam D.

STANDBY UQUID CONTROL 11111111111111111111111111111'1 I o 300 600 900 ftCSWfl'Clt1i 1200 1500 Answer: C Answer Explanation QID: 11-1 NRO 19 Question # I 19 I Developer / Date: JJR /5-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO I SRO OCSOPS ILT Page: 63 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 211000 SLC A4.01 - Ability to manually operate and/or 4.1 4.1 monitor in the control room: Tank level Level I RO I Tier 2 I Group I 1 General RPVC - with ATWS EOP User's Guide References EOP

EXAMINATION ANSWER KEY oc RO NRC Exam C is Correct. The EOP Users Guide provides the following: The Cold Shutdown Boron Weight (CSBW) is defined to be the least weight of soluble boron which, if injected into the RPVand mixed uniformly, will maintain the reactor shutdown under all conditions regardless of control rod position or RPV water temperature.

In the RPV Control* with ATWS and EMERGENCY DEPRESSURIZATION - with ATWS procedures, this amount is expressed as "Liquid Poison tank level at or below 150 gallons."

150 gallons in the Liquid Poison tank is equivalent to the elevation of the Liquid Poison pump suction line. When the Liquid Poison tank reaches this level, all available boron in the tank has been injected. With adherence to Technical Specifications limits, the Liquid Poison Tank will contain an amount of boron that is greater than the CSBW. For conservatism, all the liquid in the tank above the suction line will be injected. Therefore, injecting SLC until the tank volume indicates < 150 gallons ensures the CSBW is met and the reactor will remain shutdown under ALL conditions regardless of control rod positions. Choice C SLC tank indicates 150gal. The Explanation applicant's ability to manually monitor SLC tank level is required to answer this question.

A is Incorrect but plausible. SLC tank indicates 1450 gal. The Hot Shutdown Boron Weight (HSBW) is achieved when 650

=

gallons of boron has been injected (2100 gal

  • 650 gal 1450 gal).

This corresponds to the amount of boron that will maintain the reactor shutdown under all hot standby conditions. The question specifically asks under all conditions.

B is Incorrect but plausible. SLC tank indicates 650 gal. The applicant may confuse the requirement and assume the HSBW has been injected when SLC tank level indicates 650 gal, not when 650 gal has been injected.

D is Incorrect. SLC tank indicates 0 gal. It is true the CSBW has been injected at this point however the question asks which is the FIRST of the four SLC tank indications the applicant can call the reactor shutdown under all conditions. The applicant may not recall the CSBW is considered injected when SLC tank level is at 150 OCSOPS ILT Page: 65 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam Lesson Plan 2621.845.0.0053, RPV Control - with A TWS Learning EWA-2257, Given the EOP, describe in detail each Objectivel step/statement, including the technical basis, and how to verify or perform each step.

Question Source (New, Modified, Bank) New If Bank gr Modifi~d VISION System/Question 10:

Question Source: N/A Previous 2 Exams:

Memory or X Comprehension Fundamental 3:SPK or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: Solve a Problem using Knowledge and its meaning 10CRF55 55.41b I 10 55.43b I

Content Administrative, normal, abnormal, and emergency operating procedures for the facility.

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes Point Value: 1 System 10 No.: 211000 PRA: I No Safety ~ Initial License Level 1

Function: D LORT OCSOPSILT Page: 66 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 20 10: 11-1 NRO 20 Points: 1.00 The plant is starting up after an outage and the reactor has just been declared CRITICAL lAW Procedure 201, Plant Startup.

Which of the following SRM recorder charts shows that the reactor is critical with the LONGEST period? (See below)

A B c D OCSOPS ILT Page: 67 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam A. A B. B C. C D. D Answer: A IAnswer Explanation QID: 11-1 NRO 20 Question # I 20 I Developer 1 Date: JJR 15-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO SRO 215004 Source Range Monitor A4.02 - Ability to manually operate andlor 3.0 3.1 monitor in the control room: SRM recorder Level I RO I Tier 2 I Group I 1 General 201 References OCSOPS ILT Page: 68 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam A is Correct and B is Incorrect but plausible. The reactor has just been declared critical during a startup. When declared critical, the reactor state is actually slightly supercritical with counts rising at a constant rate. Procedure 201 defines the reactor is critical when neutron flux is increasing with a stable positive period, without additional control rod movement. 4 SRM charts are provided. SRM counts goes from left to right increasing. The top of each trace represents the current time and time goes from top to bottom increasing. Trace A shows a constant increase in Explanation counts, with a smaller slope than that in trace B, which is also critical.

C & 0 are Incorrect but plausible. Trace C shows constant counts at the top of the trace and trace 0 shows lowering counts at the top of the trace. The applicant must recognize the reactor state when declared critical and must also know that counts is on the horizontal axis and time is on the vertical axis.

Lesson Plan 2621.828.0.0029, Nuclear Instrumentation NIS-10445, Given a set of system indications or data, evaluate Learning and interpret them to determine limits, trends and system ective/ status.

Question Source (New, Modified, Bank) Bank If Bank or Modified VISION System/Question 10: 663735 Question Source: ILT 08*1 NRC Exam Previous 2 Exams: No Memory or X Comprehension Fundamental 3:SPK or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: .solve a Problem using Knowledge and its meaning OCSOPSILT Page: 69 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 10CRF55 55.41b I 1 I 55.43b I

Fundamentals of reactor theory, including fission process, neutron Content multiplication, source effects, control rod effects, criticality indications, reactivity coefficients, and poison effects.

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes I Point Value: 1 System ID No.: 215004 PRA:

I No Safety ~ Initial license Level 7

Function: LORT OCSOPS ILT Page: 70 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 21 10: 11-1 NRO 21 Points: 1.00 The plant is at rated power. Electrical maintenance was hanging a tag on the breaker for the electrical bus that powers Shutdown Cooling (SOC) Outlet Isolation Valve, V-17 54.

An event then occurred and the electrical bus tripped offline. Which of the following electrical busses is V-17-54 powered from AND was an LCO for Tech Spec 3.7 affected?

V-17-54 Power Supply TS 3.7 LCO affected?

A. MCC 1A12 Yes B. MCC 1AB2 Yes C. MCC 1A12 No O. MCC 1AB2 No Answer: B Answer Explanation QID: 11-1 NRO 21 Question # J 21 I Developer 1 Date: JJR 15-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO SRO 205000 Shutdown Cooling 2.2.22 - Equipment Control: Knowledge of 4.0 4.7 limiting conditions for operations and safety limits.

Level I RO I Tier 2 I Group I 1 General 305 References

EXAMINATION ANSWER KEY oc RO NRC Exam B is Correct. The electrical bus that powers SOC Outlet Isolation Valve, V-17-54, is Vital MCC 1AB2. TS 3.7 states that a loss of this electrical bus requires that plant is placed in the COLD SHUTDOWN condition within 30 hrs. A RO is required to know Explanation that this TS is affected but not the actual LCO.

All distractors are Incorrect but plausible if the applicant fails to recall the correct power supply or that TS 3.7 is affected. MCC 1A12 is not affected TS 3.7.

Lesson Plan 2621.828.0.0045, Shutdown Cooling System Learning SDC*10453, Explain or describe how this system is interrelated o ective/ with other Question Source (New, Modified, Bank) New If Banis Qr M,uUfied VISION System/Question 10:

Question Source: N/A Previous 2 Exams:

Memory or X Comprehension Fundamental 1:F or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: facts 55.41b I 5 55.43b I

Facility operating characteristics during steady state and transient 10CRF55 conditions, including coolant chemistry, causes and effects of Content temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes Point Value: 1 System 10 No.: I 205000 I PRA: I No oes OPS ILT Page: 72 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam Safety 4

181 Initial License Level Function: o LORT OCSOPS ILT Page: 73 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 22 10: 11-1 NRO 22 Points: 1.00 Which of the following design features of the EMRVs and associated piping prevents siphoning of water into the discharge piping?

EMRV discharge piping _ _ _ _ _ _ _ _ _ __

A. Y-quencher B. slotted openings C. vacuum breakers D. in-line check valves Answer: C Answer Explanation QID: 11-1 NRO 22 Question # I 22 I Developer / Date: JJR /5-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO SRO 239002 SRVs 2.1.28 - Conduct of Operations: Knowledge of 4.1 4.1 the purpose and function of major system components and controls.

Level I RO I Tier 2 I Group I 1 General BR2002 sh.1 References OCSOPS ILT Page: 74of241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam C is Correct. The purpose/function of the EMRV discharge piping vacuum breakers is that it is designed to break vacuum in the piping upon closure of the EMRV so that the vacuum established by steam condensation does not draw water into the discharge piping.

Explanation All distractors are Incorrect but plausible if the applicant is not familiar with major components of the SRV/EMRV system. Y quenchers represents the shape of the piping and does not prevent siphoning. There are no in-line check valves in the

............... and there are no slotted s in the I1I~jl"nJ:a nts.

Question Source (New, Modified, Bank) Bank If Bank Qr MQdified VISION System/Question 10: 609059 Question Source: ILT Bank Previous 2 Exams: No Memory or X Comprehension Fundamental 1:8 or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: Bases or purpose 10CRF55 55.41b I 7 55.43b I

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

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes I Point Value: 1 System 10 No.: I 239002 I PRA: I No OCSOPSILT Page: 75 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam Safety 3

IZI Initial License Level Function: D LORT OCSOPS ILT Page: 76 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 23 10: 11-1 NRO 23 Points: 1.00 One of the prerequisites for operating 125 VDC Distribution Systems A & B, is to ensure Battery Room ventilation is operating lAW procedure 331, Office Building Heating, Ventilation and Air Conditioning System operating procedure.

Which of the following correctly describes the basis of this requirement?

A. Battery Rooms must be maintained at a negative pressure.

B. Battery Room ventilation is required to achieve proper float voltage.

C. Battery Room ventilation is required to prevent hydrogen accumulation.

D. Battery Room ventilation raises the capacity of the battery by keeping its temperature in the procedural range.

Answer: C Answer Explanation QID: 11-1 NRO 23 Question # I 23 I Developer / Date: JJR /5-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO SRO 263000 DC Electrical Distribution K5.01 - Knowledge of the operational implications of the following concepts as they 2.6 2.9 apply to D.C. ELECTRICAL DISTRIBUTION :

Hydrogen generation during battery charging.

Level I RO I Tier 2 I Group I 1 General 331 340.1 References OCSOPS ILT Page: 77 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam C is Correct. Procedure 331 and 340.1 requires battery room ventillation be in service prior when the batteries are in service to prevent hydrogen accumulation which is an explosive hazard.

Unless there is an emergency or surveillance, the batteries always operate on a float charge (KIA match). Only in an emergency or surveillance are the batteries discharging and supplying power to DC loads.

Explanation A is Incorrect but plausible. The ventilation basis is not to maintain negative pressure.

B is Incorrect but plausible. The ventilation does not affect the float voltage, however the applicant may not recognize this is not the correct basis.

o is Incorrect but plausible since batteries are more efficient at lower tem however this is not the correct basis.

Lesson Plan 2621.828.0.0012, DC Distribution DCD-10447, Given normal operating procedures and documents Learning for the system, describe or interpret the procedural steps.

ectivel Question Source (New, Modified, Bank) Bank If Bank or Modified VISION System/Question 10: 510793 Question Source: ILT Bank Previous 2 Exams: No Memory or X Comprehension Fundamental 1:P or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: Procedure steps and cautions OCSOPS ILT Page: 78 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 55.41b I 5 I 55.43b I Facility operating characteristics during steady state and transient 10CRF55 conditions, including coolant chemistry, causes and effects of Content temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes I Point Value: 1 System ID No.: 263000 PRA: I No Safety ~ Initial License Level 6

Function: D LORT OCSOPS ILT Page: 79 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 24 ID: 11-1 NRO 24 Points: 1.00 The plant was at rated power when a LOCA occurred. The crew is executing the RPV Control - no ATWS EOP. The following plant conditions currently exist:

  • Drywell Pressure indicates 1.8 psig and slowly lowering
  • Reactor water level lowered to 50 in and is now 100 inches and slowly rising The US then directs the BOP to RESET THE ADS TIMERS (see EOP figure below).

RESETTHEADS TMERS lAW the RPV Control - no ATWS EOP, which of the following actions are required to reset the ADS timers?

1. Place all EMRV AUTO DEPRESS VALVE control switches to OFF then back to AUTO.
2. Place both ADS TIMER AUTO BYPASS keylock switches to BYPASS then back to AUTO.
3. Place both HI DRYWELL PRESSURE SWITCH keylock switches to RESET then back to AUTO.

A. 2 ONLY B. 3 ONLY C. 1 .and. 2 D. 2.and. 3 OCS OPS ILT Page: 80 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam Answer: A Answer Explanation QID: 11-1 NRO 24 Question # I 24 I Developer / Date: JJR /5-14-2012 Importance Rating K&A 218000 ADS A4.05 - Ability to manually operate and/or 4.2 4.2 monitor in the control room: ADS timer reset Level RO Tier 2 Group 1 General EOP User's Guide References A is Correct. In order to reset ADS timers following initiation, both ADS timer keylock switches must be placed in the BYPASS position momentarily, then placed back in AUTO.

All distractors are Incorrect but plausible if the applicant Explanation confuses the steps to RESET ADS following timer initiation (in RPV Control-no ATWS) and the steps to reset ADS logic when all initiating signals are clear (lAW procedure 308). Step 3 in the question stem is required to reset ADS logic lAW 308. Step 1 is plausible if the applicant does not recall how to reset ADS timers This action is uired if an EMRV is 0 n or I Lesson Plan 2621.845.0.0052, RPV Control - no A TWS Learning ENA-2257, Given the EOP, describe in detail each Objective/ step/statement, including the technical basis, and how to verify or each ste Question Source (New, Modified, Bank) New OCS OPS ILT Page: 81 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam If Bank Qr MQdified VISION System/Question 10:

Question Source: N/A Previous 2 Exams:

Memory or X Comprehension Fundamental 1:P or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: Procedure steps and cautions 10CRF55 55.41b I 10 55.43b I

Content Administrative, normal, abnormal, and emergency operating procedures for the facility.

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes I Point Value: 1 System 10 No.: 218000 PRA: I No Safety 3

181 Initial License Level Function: o LORT OCSOPS ILT Page: 82 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 25 10: 11-1 NRO 25 Points: 1.00 The Standby Liquid Control System explosive (squib) valves are powered from which of the following sources?

A. 24 I 48 VDC Distribution B. Safety Related 125 VDC Distribution C. Vital Motor Control Center (VMCC) Distribution D. The respective pump Motor Control Center (MCC)

Answer: D IAnswer Explanation QID: 11-1 NRO 25 Question # I 25 I Developer / Date: JJR /5-14-2012 Importance Rating K&A 211000SLC K2.02 - Knowledge of electrical power supplies 3.1 3.2 to the foil  : Ex osive valves Level Tier 2 Group 1 General BR 3004 sh. 1 References o is Correct. The MCC of the SLCS pump selected for injection provides the 480VAC power for both squib valves for that SLCS train.

Explanation All distractors are Incorrect but plausible since they are logical sources of power if the applicant does not know the correct OCSOPS ILT Page: 83 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam Lesson Plan 2621.828.0.0046, Standby Liquid Control SLC-10436, Using plant procedures and electrical drawings, Learning determine electrical power supply for system equipment and Objective/ any associated/applicable logic, including power loss effects.

Question Source (New, Modified, Bank) Modified If Bank gr Mgdified VISION System/Question 10: N/A Question Source: Peach Bottom ILT Bank Previous 2 Exams: No Memory or X Comprehension Fundamental 1:F or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: facts 10CRF55 55.41b I 7 55.43b I

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

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes I Point Value: 1 System 10 No.: 211000 PRA: I No Safety ~ Initial License Level 1

Function: o LORT OCSOPS ILT Page: 84 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 26 10: 11-1 NRO 26 Points: 1.00 The plant was at rated power when LPRM 28-33A (input to APRM 1) failed resulting in APRM 1 indicating 87.5%. All other LPRMs and APRMs indicate normally.

Which of the following other indications are correct? Assume that the APRM Gain is 1.000.

A. APRM 1 DNSCL OR INOP light will be ON (on Panel4F) AND a rod block is present.

B.LPRM 28-33A local analog meter on Panel 4F full core display will have a red back-light.

C.LPRM 28-33A amber light on Panel 4F full core display will be OFF AND a rod block is present.

D.LPRM 28-33A local meter on Panel4F full core display indicates downscale.

Answer: 0 IAnswer Explanation QID: 11-1 NRO 26 Question # I 26 I Developer J Date: JJR J 5-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO SRO 215005 APRM J LPRM K3.05 - Knowledge of the effect that a loss or malfunction of the AVERAGE POWER RANGE 3.8 3.8 MONITOR/LOCAL POWER RANGE MONITOR SYSTEM will have on following: Reactor power indication Level I RO I Tier 2 I Group I 1 General RAP-G4f, G7f References

EXAMINATION ANSWER KEY oc RO NRC Exam o is Correct. APRM 1 will read 100% at full power and will read 87.5 when 1 LPRM goes downscale (700/8) [given the gain is 1.000]. Therefore, the LPRM has failed downscale. When the LPRM fail downscale, several things happen: the associated APRM produces a rodblock; the amber light (on full core display) goes ON; the APRM reading goes down; and, the local reactor power meter of the full core display (individual LPRM readings) will go downscale. Therefore, there will be a rod block and the local meter will read downscale.

Explanation A is Incorrect but plausible. The APRM will neither be INOP nor downscale.

B is Incorrect but plausible. There is no red backlighting for the associated LPRM, the red backlighting on control rod indications.

C is Incorrect but plausible. Since the local reactor power indication for the LPRM failed downscale, its amber light will be ON not OFF.

Lesson Plan 2621.828.0.0029, Nuclear Instrumentation NIS-10444, Describe the interlock signals and setpoints for the Learning affected system components and expected system response includin loss or failed comn",nOIMll'ec Question Source (New, Modified, Bank) Bank If Baok Qr MQdifi~d VISION System/Question 10: 609055 Question Source: ILT Bank Previous 2 Exams: No Memory or X Comprehension Fundamental 3:SPK or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: .solve a Problem using Knowledge and its meaning 10CRF55 Content 55.41b 7 55.43b I

OCSOPS ILT Page: 86 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes I Point Value: 1 System ID No.: 215005 PRA:

I No Safety ~ Initial License Level 7

Function: LORT OCSOPSILT Page: 87 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 27 10: 11-1 NRO 27 Points: 1.00 The plant was at rated power when a LOCA occurred.

Which of the following states the sequence of automatic Recirculation Pump trips and automatic Isolation Condenser (IC) initiations as RPV water level steadily drops from 95" to 82"?

Occurs First Occurs Second Occurs Third A. ALL Recirculation IC condensate No other actions Pumps Trip return valves open and vent valves close B. IC condensate A, B, E ONLY C, D Recirculation return valves open Recirculation Pumps Trip and vent valves Pumps Trip close C. IC condensate ALL Recirculation No other actions return valves open Pumps Trip and vent valves close D. A, B, E ONLY IC condensate C, D Recirculation Recircu lation return valves open Pumps Trip Pumps Trip and vent valves close Answer: A Answer Explanation QID: 11*1 NRO 27 Question # I 27 I Developer / Date: JJR /5-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO I SRO OCSOPSILT Page: 88 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 216000 Nuclear Boiler Inst.

K1.15 - Knowledge of the physical connections and/or cause- effect relationships between 3.9 4.1 NUCLEAR BOILER INSTRUMENTATION and the following: Isolation condenser: Plant

.,...nDrIIT'C RO Tier 2 Group 2 General RAP-C1a, C2a BR 3029 sh. 2 References A is Correct. The isolation condensers auto initiate (after 1.5 seconds) on Nuclear Boiler Instrumentation signals from either a 10-10 RPV water level (90") or RPV high pressure (1051 psig).

Recirculation pumps also trip from the same parameters. On 10-10 water level, all recirculation pumps trip immediately. On high pressure, recirculation pumps A, B & E trip immediately, and pumps C & D trip after sustained high pressure of 10.5 seconds.

Explanation Therefore, as RPV water level lowers through the 10-10 setpoint, all recirculation pumps trip and the ICs initiate after a time delay of 1.5 seconds.

All distractors are Incorrect but plausible since they occur but in they're in the incorrect sequence or the applicant may confuse the RPV and Lo-Lo water level 3C,uu.t:n Lesson Plan 2621.828.0.0023, Isolation Condensers ICS-2030, Describe the Isolation Condenser design features Learning and/or interlocks which provide for the following: automatic m initiation and isolation Question Source (New, Modified, Bank) Bank If Bank Qr MQ{Ufied VISION System/Question ID: 666839 Question Source: ILT Bank Previous 2 Exams: No Memory or X Cognitive Comprehension Fundamental 2:DR Level or Analysis Knowledge OCSOPS ILT Page: 89 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam NliREG 1021 Appendix B: .Describing or recognizing Relationships 10CRF55 55.41b I 7 I 55.43b I

Design, components, and functions of control and safety systems, Content including instrumentation, Signals, interlocks, failure modes, and automatic and manual features.

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes I Point Value: 1 System 10 No.: 216000 PRA: I No Safety 7

IZI Initial License Level Function: o LORT

EXAMINATION ANSWER KEY oc RO NRC Exam 28 10: 11*1 NRO 28 Points: 1.00 The plant was at rated power when a reactor scram occurred.

Which ONE of the following correctly describes the electrical power supplies to the Condensate Pumps five minutes after the reactor scram?

CQndimsat~ E!umR CQnd~nsat~ PumR CQnd~Dsate PumR A B .c.

A. Transformer 1A via Transformer 1A via Transformer 1B via Bus 1A Bus 1A Bus 1B B. Transformer 1A via Transformer 1B via Transformer 1B via Bus 1A Bus 1B Bus'IB

c. Transformer S1A via Transformer S1A via Transformer S1 B via Bus 1A Bus 1A Bus 1B D. Transformer S1A via Transformer S1 B via Transformer S 1B via Bus 1A Bus 1B Bus 1B Answer: 0 Answer Explanation QID: 11-1 NRO 28 Question # I 28 I Developer / Date: JJR /5-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO SRO 256000 Reactor Condensate K2.01 - Knowledge of electrical power supplies 2.7 2.8 to the following: System pumps Level I RO I Tier 2 I Group I 2 General BR 3001A BR 3001B References OCSOPSILT Page: 91 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam o is Correct. The power supply to Condensate Pump A is 4160VAC Bus 1A and the power supply to Condensate Pumps B

& C is 4160VAC Bus 1B. When shutdown (post-scram). these buses are powered from the Startup Transformers S1A and S1B respectively (offsite power). The startup transformer breakers automatically close when the Main Turbine trips. The Main Explanation Turbine will trip immediately following the scram.

All distractors are Incorrect but plausible if the applicant confuses either the transformer labels, when the startup transformers pick up power to the 4160VAC buses, or they do not recall which bus Condensate Pump B is powered from since one bus two Lesson Plan 2621.828.0.0017, Feed and Condensate System CNS-10453, Explain or describe how this system is interrelated Learning with other plant systems.

Ohl"",."""...,

Question Source (New, Modified, Bank) New If Bank Qr MQdifi~d VISION System/Question 10:

Question Source: N/A Previous 2 Exams:

Memory or X Comprehension Fundamental 1: I or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: interlocks, setpoints, or system (singular) response 10CRF55 55.41b I 7 55.43b I

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

Justification for LORT questions with N/A KIA values < 3.0 OCS OPS ILT Page: 92 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam Time to Complete: 1*2 minutes I Point Value: 1 System 10 No.: 256000 PRA: I No Safety t8I Initial License Level Function:

2 o LORT OCSOPSILT Page: 93 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 29 10: 11-1 NRO 29 Points: 1.00 A plant shutdown is in progress due to the loss of 125 VDC Bus C, when a catastrophic loss of instrument air occurred.

TWO MINUTES LATER, the URO observes the following:

  • INSTR AIR SUPPLY indicates 0 psig
  • RPV pressure currently indicates 1100 psig and steady
  • 230 KV Breakers GD1 and GC1 indicate GREEN lights ON In reference to RPV Pressure control, which of the following are correct for the above conditions? (assume NO operator action had been taken)
1. ALL Turbine Bypass Valves are OPEN
2. ALL EMRVs are OPEN
3. ONLY 3 EMRVs are OPEN
4. BOTH Isolation Condensers are in service
5. SOME SRVs are OPEN A. 2 ONLY B. 3 and 4 ONLY C. 3 and 5 ONLY D. 1,2, and 4 ONLY Answer: A Answer Explanation QID: 11-1 NRO 29 Question # I 29 I Developer / Date: JJR /5-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO SRO 239001 Main and Reheat Steam K3.16 - Knowledge of the effect that a loss or malfunction of the MAIN AND REHEAT STEAM 3.6 3.6 SYSTEM will have on following: Relief/safety valves

EXAMINATION ANSWER KEY oc RO NRC Exam Level I RO I Tier 2 I Group I 2 General RAP-B4g GE 729Ei82 EB 0-3033 References RAP-Cia OCSOPS ILT Page: 95 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam This question examines how the EMRVs will respond to a loss of the Main Steam System (MSIV Closure).

The plant is at power with the loss of 125 VDC Bus C. With this loss, 125 VDC Busses DC-F and DC-2 are also lost.

All EMRVs receive 2 DC power supplies (DC-D and DCF), such that the loss of 1 DC power supply will not prevent any EMRV from functioning. Therefore, all EMRVs can operate normally.

Each Isolation Condenser has a normally closed condensate return valve, which are DC powered: IC-A valve receives power from DC-1 and IC-B receives power from DC-2 which is powered from DC-C. With the loss of DC power, then IC-B condensate return valve cannot open, but IC-A is fully operable.

With the loss of all instrument air, the outboard MSIVs will auto close, which prevents operation of the Turbine Bypass Valves.

The EMRVs open under an ADS Signal or from RPV high pressure (some open at 1065 psig and the others open at 1085 psig). Since RPV pressure is 1100 psig and steady, and since all EMRVs are Explanation fully functional, then all EMRVs are currently open and indicate in the VALVE OPEN REGION.

The RPV safety valves open under RPV high pressure (initial opening at 1212 psig). Since RPV pressure is only 1100 psig, then no SRVs have opened.

Isolation Condensers initiate on 10-10 RPV water level and from RPV high pressure (sustained 1051 psig) and are therefore in service, except that IC-B condensate return valve is closed with no electrical power. Thus, only IC-A is in service.

Choice 1 is false since the TBV are closed.

Choice 2 is true since all EMRVs have opened due to RPV high pressure.

Choice 3 is false since all EMRVs have opened due to RPV high pressure.

Choice 4 is false since IC-B is not in-service (but is true for ONLY IC-A).

Choice 5 is false since the SRV lift setpoint has not been reached.

Therefore, Answer A (Choice 2 ONLY) is Correct.

OCSOPS ILT Page: 96 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam Lesson Plan 2621.828.0.023 Isolation Condensers ICS-2338, Given plant conditions, evaluate the impact on the Learning Isolation Condenser System and the plant.

0"""',""1',,,""-1 Question Source (New, Modified, Bank) Bank If Bank or MQdified VISION System/Question 10: 811719 Question Source: ILT Bank Previous 2 Exams: No Memory or X Comprehension Fundamental 3:SPK or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: Solve a Problem using Knowledge and its meaning 10CRF55 55.41b I 7 55.43b I

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

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes I Point Value: 1 System 10 No.: 239001 PRA:

I No Safety ~ Initial License Level 4

Function: o LORT OCSOPS ILT Page: 97 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 30 10: 11-1 NRO 30 Points: 1.00 The plant is at rated power.

Which of the following annunciators/indications by themselves, indicate that an automatic protective action has occurred or will occur to mitigate an offsite radiological release?

A. BOTH Offgas Radiation Monitors indicate upscale.

B. Service Water Discharge Radiation Monitor indicates upscale.

C. Spent Fuel Pool Area radiation monitor C5 indicates upscale.

D. BOTH Stack RAGEMS noble gas effluent monitors indicate upscale.

Answer: A IAnswer Explanation QID: 11-1 NRO 30 Question # I 30 I Developer 1 Date: JJR 15-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO SRO 272000 Radiation Monitoring K4.01 - Knowledge of RADIATION MONITORING System design feature(s) andlor 2.7 2.8 interlocks which provide for the following:

Redundancy Level I RO I Tier 2 I Group I 2 General RAP-10F1c References

EXAMINATION ANSWER KEY oc RO NRC Exam A is Correct. RAP-10F1c (answer a) states the automatic actions for this alarm (OFFGAS HI-HI): Closure of V-7-31 [AOG Bypass Valve], V-7-29 [48" hold-up drain valve] and OG-AOV-001A (001 B)

[Recombiner inlet valves] to isolate the off gas system at the stack and trip the mechanical vacuum pump (if running) after a 15 +0 -1 (14 to 15) minute time delay with coincident upscale trip of both channels, or an upscale trip in 1 channel and downscale trip in the other channel (redundancy required to actuate the automatic interlock). The mechanical vacuum pumps are not in service given the plant conditions. With both offgas rad monitors upscale, the offgas system will be isolated from the stack after 15 minutes. The expected annunciator prior to this Hi-Hi alarm Explanation (Offgas Hi 10F2c) has no protective functions.

B is Incorrect but plausible. Service Water Discharge Radiation Monitor has no automatic actions.

C is Incorrect but plausible. Fuel pool area radiation monitors B9 and C9 will isolate RB HVAC and initiate SGTwhen either rad monitor goes high. Area rad monitor C5, in the same vicinity, only produces a control room alarm (RAP-F1 k).

D is Incorrect but plausible. There are no protective functions from u stack RAGEMS.

Lesson Plan 2621.828.0.033A, Plant Radiation Monitoring System RAD-10449, State the function and interpretation of system Learning alarms, alone and in combination, as applicable in accordance O... i ........i",... with the RAPS.

Question Source (New, Modified, Bank)

If Bank Q[ MQdified I Bank VISION System/Question ID: 510682 Question Source: ILT Bank Previous 2 Exams: No OCSOPSILT Page: 99 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam Memory or X Comprehension Fundamental 2:RI or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: Recognizing Interaction between systems (plural), including consequences and implications 10CRF55 55.41b I 7 55.43b I

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

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes I Point Value: 1 System ID No.: 272000 PRA:

I No Safety [gI Initial License Level 9

Function: o LORT OCSOPSILT Page: 100 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 31 10: 11*1 NRO 31 Points: 1.00 A reactor shutdown is in progress with reactor power at 8%. The currently latched group of 8 rods has an insert limit of 28 and a withdraw limit of 32. This group's rods are at the following positions:

  • 2 rods are at 26
  • 4 rods are at 32
  • 1 rod is at 34
  • 1 rod is at 28, and selected Which of the following statements is correct concerning the RWM status?

A. 3 INSERT ERRORS exist AND an INSERT BLOCK is applied to ALL control rods.

B. The RWM must be bypassed since it would not have allowed this configuration.

C. 2 INSERT ERRORS exist AND a WITHDRAW BLOCK is applied to ALL control rods.

D. 1 WITHDRAW ERROR exists AND a WITHDRAW BLOCK is applied to SOME control rods.

Answer: C IAnswer Explanation QID: 11-1 NRO 31 Question # I 31 I Developer J Date: JJR /5-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO SRO 201006 RWM K5.12 - Knowledge of ROD WORTH MINIMIZER SYSTEM (RWM) (PLANT SPECIFIC) design 3.5 3.5 feature(s) and/or interlocks which provide for the following: Withdraw block Level I RO I Tier I 2 I Group I 2 OCSOPS ILT Page: 101 of241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam General 409 VM-RW-1312 References C is Correct. There are two insert errors (the two rods at 26) and there is a withdraw block due to the rod at 34, which is applied to all control rods.

A is Incorrect but plausible. There are only two insert errors; the two rods at 26. There is an insert block, however, since the rod that is withdrawn past its withdraw limit is not selected.

Explanation B is Incorrect but plausible if the applicant does not recall RWM operations since it would have allowed the given configuration.

o is Incorrect but plausible. There is a withdraw error, but the withdraw block would be applied to all control rods since the withdraw error rod is not selected.

Lesson Plan 2621.828.0.0041, Rod Worth Minimizer RWM-10444, Describe the interlock signals and setpoints for Learning the affected system components and expected system ective/ res includi loss or failed nents.

Question Source (New, Modified, Bank) Bank If Bank or MQgified VISION System/Question 10: 506426 Question Source: ILT Bank Previous 2 Exams: No Memory or X Comprehension Fundamental 3:PEO or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: Predict an .Event or .outcome oes OPS ILT Page: 102 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 10CRF55 55.41b I 7 J 55.43b I

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

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes I Point Value: 1 System 10 No.: 201006 PRA: I No Safety 7

181 Initial License Level Function: D LORT OCSOPS ILT Page: 103 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 32 Points: 1~OO The plant is at rated power. An event then occurs resulting in a loss of USS 1B2.

Which of the following will lose power from this event?

A. Reactor Feed Pump 'c' Aux Oil Pump, P-2-9C.

B. Control Room Master Fire Alarm Panels A and B.

C. The 'c' Reactor Recirc Pump Discharge Valve, V-37-32.

D. Panel ER-42 (Screen Wash Control Panel) normal power.

Answer: B

!Answer Explanation QID: 11-1 NRO 32 Question # I 32 J Developer I Date: JJR /5-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO SRO 286000 Fire Protection K6.01 - Knowledge of the effect that a loss or malfunction of the following will have on the 3.1 3.1 FIRE PROTECTION SYSTEM: A. C. electrical distribution: Plant-Specific Level I RO 1 Tier 2 I Group J 2 General ABN-48 References oes OPS ILT Page: 104 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam B is Correct. USS 1B2 powers the Control Room Master Fire Alarm Panels A and B. This will cause the Control Room to lose its ability to detect fires from those panels in the Main Control Room.

Explanation All distractors are incorrect but plausible if the applicant does not recall the correct power supply to the CR Fire Alarm Panels.

None of the other choices are from USS-1 B2.

Lesson Plan 2621.828.0.0019, Fire Protection Learning FPS-10449, State the function and interpretation of system Objective/ alarms, alone and in combination, as applicable in accordance with the RAPS.

Question Source (New, Modified, Bank) New If Bank or Modifigd VISION System/Question 10:

Question Source: N/A Previous 2 Exams:

Memory or X Comprehension Fundamental 1:F or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: facts 10CRF55 55.41b I 7 55.43b I

Design, components, and functions of control and safety systems, Content including instrumentation, Signals, interlocks, failure modes, and automatic and manual features.

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes I Point Value: 1 System 10 No.: I 286000 I PRA: I No OCSOPS ILT Page: 105 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam Safety [81 Initial License Level 8

Function: o LORT OCSOPS ILT Page: 106 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 33 10: 11-1 NRO 33 Points: 1.00 The plant is at rated power. Plant conditions include the following:

Five (5) minutes after the SGTS initiation, with no operator action, which of the following is the correct fan/valve configuration if the lead system developed/maintained a low flow signal?

SGTS 2 Orific~

SGTS 1 Fan SGTS 2 Fan Valve V-28-28 A. ON ON OPEN B. ON OFF CLOSED C. OFF ON CLOSED D. ON OFF OPEN Answer: A Answer Explanation QID: 11-1 NRO 33 Question # I 33 I Developer 1 Date: J..IR 15-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO SRO 290001 Secondary CTMT A 1.01 - Ability to predict andlor monitor changes in parameters associated with 3.1 3.1 operating the SECONDARY CONTAINMENT controls including: System lineups Level I RO I Tier I 2 I Group I 2 OCSOPSILT Page: 107 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam General 330 RAP-L5b References A is Correct. On an automatic system initiation, both SGTS fans start. If the lead fan develops adequate flow within the first 2-3 minutes, the lag fan will shutdown and the associated inlet/outlet valves close. If the lead fan does not develop adequate flow, the lag fan continues and the lead fan continues to run, but with the lead system inlet/outlet valves closed. The system orifice valves are normally closed (with the systems in standby) and stays closed when the lead system starts with proper flow. If the lead running system sees low flow, then besides what's already been said, the lead system orifice valve also opens (and inlet/outlet Explanation valves close and the redundant system assumes the SGTS function). Therefore, 5 minutes after an auto initiation, system 2 fan (which was selected as lead) will be running with the loop inlet/outlet valves closed and loop orifice valve open. System 1 fan is also running performing the SGTS function.

All distractors are Incorrect but plausible if the applicant does not recognize the correct fan/valve lineup for the stated conditions.

Lesson Plan 2621.828.0.0042, Secondary Containment & SGTS SGT-10445, Given a set of system indications or data, evaluate Learning and interpret them to determine limits, trends and system status.

Question Source (New, Modified, Bank)

If Bilnk or Modifi~d I Bank VISION System/Question 10: 510845 Question Source: ILT Bank Previous 2 Exams: No OCSOPS ILT Page: 108 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam Memory or X Comprehension Fundamental 3:PEO or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: Predict an Event or .outcome 10CRF55 55.41b I 7 55.43b I

Design, components, and functions of control and safety systems, Content including instrumentation, Signals, interlocks, failure modes, and automatic and manual features.

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes I Point Value: 1 System ID No.: 290001 PRA: I No Safety ~ Initial License Level 5

Function: D LORT OCSOPS ILT Page: 109 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 34 10: 11-1 NRO 34 Points: 1.00 The plant is shutdown for a refuel outage. Plant conditions include the following:

  • The 'B' Fuel Pool Cooling (FPC) pump is in service
  • TBCCW Pumps 1 and 2 are in service An event then occurred resulting in a loss of MCC 1A21.

Based on the conditions above, which of the following correctly states the plant impact AND the required action for this event?

Plant Impact Action Refer to AN O/OR Loss of... Perform Actions Required by...

A. the 'B' SOC Loop ABN-3, Loss of Shutdown Cooling B. the 'B' FPC Pump ABN-16, Loss of Fuel Pool Cooling C. TBCCW Pumps 1 & 2 ABN-20, Loss ofTBCCW O. power to the Refueling Bridge ABN-45, Loss of USS 1A2 Answer: 0 Answer Explanation QID: 11-1 NRO 34 Question # I 34 I Developer 1 Date: JJR 15-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO I SRO

EXAMINATION ANSWER KEY oc RO NRC Exam 234000 Fuel Handling Equipment A2.03 - Ability to (a) predict the impacts of the following on the FUEL HANDLING EQUIPMENT

and (b) based on those predictions, use 2.8 3.1 procedures to correct, control, or mitigate the consequences of those abnormal conditions or
Loss of electrical oo"",er Tier 2 Group 2 General 205.0 ABN-45 References o is Correct. The power supply to the Refuel Bridge is MCC 1A21. The correct action is to restore power to the refuel bridge.

Of the choices listed the operators will refer to ABN-48, Loss of USS 1A2. In ABN-48, the operator will also determine the extent of condition by referring to the load lists in the back of the ABN.

Explanation All loads off of MCC 1A21, 1A21A, and 1A21B are listed.

All distractors are Incorrect but plausible if the applicant does not recall the loads that were lost when MCC 1A21 de-energized.

Each ABN is the correct ABN for the loss stated in each distractor.

Lesson Plan 2621.812.0.0003, Refueling RFL-00291, Describe the Refueling Platform major components Learning location, function and power supply.

Question Source (New, Modified, Bank)

If Bank Qr MQdified I New VISION System/Question 10:

Question Source: N/A Previous 2 Exams:

OCSOPS ILT Page: 111 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam Memory or X Comprehension Fundamental 3:PEO or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: Predict an Event or Outcome 10CRF55 55.41b I 10 55.43b I

Content Administrative, normal, abnormal, and emergency operating procedures for the facility.

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes I Point Value: 1 System 10 No.: 234000 PRA:

I No Safety 8

181 Initial License Level Function: o LORT OCSOPS ILT Page: 112 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 35 10: 11-1 NRO 35 Points: 1.00 Which of the following could be indicative of a Reactor Manual Control System control rod movement timer malfunction?

A. The red WITHDRAW light ON for 3 seconds during a control rod ROD OUT NOTCH.

B. The green INSERT light ON for 3.5 seconds during a control rod single notch ROD IN.

C. The amber SETTLE light ON for 5 seconds following a control rod single notch ROD IN evolution D. The green INSERT light ON for 1 second during a control rod ROD OUT NOTCH.

Answer: A Answer Explanation QID: 11-1 NRO 35 Question # I 35 I Developer / Date: JJR /5-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO SRO 201002 RMCS A3.02 - Ability to monitor automatic operations 2.8 2.7 of the REACTOR MANUAL CONTROL SYSTEM includin I: Rod movement sequence lights Level RO I Tier 2 I Group I 2 General ABN-6 302.2 References

EXAMINATION ANSWER KEY oc RO NRC Exam A is Correct. lAW 302.2, the red WITHDRAWAL light is illuminated approximately 2 seconds following switch movement and remains on for approximately 1.5 seconds. Since it is on for 3 seconds, this could indicate a timer malfunction and actions of Explanation ABN-6 should be taken.

All distractors are Incorrect but plausible. All conditions described in the distractors are expected indications for rod movement.

Lesson Plan 2621.828.0.0036, Reactor Manual Control System RMC-10446, Identify and explain system operating controls I Learning indications under all plant operating conditions.

Question Source (New, Modified, Bank) Bank If Bank or Modifi~d VISION System/Question 10: 510850 Question Source: ILT Bank Previous 2 Exams: No Memory or X Comprehension Fundamental 1 :F or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: facts 10CRF55 55.41b I 7 55.43b I

Design, components, and functions of control and safety systems, Content including instrumentation, Signals, interlocks, failure modes, and automatic and manual features.

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes Point Value: 1 System 10 No.: I 201002 I PRA: I No OCSOPSILT Page: 114of241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam Safety 1

181 Initial License Level Function: D LORT OCSOPS ILT Page: 11Sof241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 36 10: 11-1 NRO 36 Points: 1.00 The plant was at rated power when an Off-Gas Deflagration occurred.

lAW ABN-25, Off-Gas Deflagration, which of the following combinations of alarms are required to be cleared before the Off-Gas system can be reset?

1. OFF GAS ISOL ACT I
2. OFF GAS ISOL ACT 1\
3. OFF GAS PRESS HI
4. OFF GAS TEMP HI A. 1 and 2 ONLY B. 3 and 4 ONLY C. 1,2, and 4 ONLY D. 1,2,3, and 4 Answer: B IAnswer Explanation QID: 11-1 NRO 36 Question # I 36 I Developer 1 Date: JJR 15-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO SRO 271000 Off-gas A4.01 - Ability to manually operate andlor 2.8 2.8 monitor in the control room: Reset system isolations Level I RO I Tier 2 I Group I 2 General ABN-25 References OCSOPSILT Page: 116 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam B is Correct. lAW ABN-25, the OFF GAS PRESS HI and OFF GAS TEMP HI alarms are required to be clear before continuing to reset the Off-Gas isolation logic.

Explanation All distractors are Incorrect but plausible since they all initiate following an Off-gas deflagration. The applicant may not recall what alarms clear or what alarms are required to be clear for this event.

Lesson Plan 2621.845.0.0053, RPV Control - with A TWS EWA-2257, Given the EOP, describe in detail each Learning step/statement, including the technical basis, and how to verify ective/ or each Question Source (New, Modified, Bank) New If Bank Qr Modifigd VISION System/Question 10:

Question Source: N/A Previous 2 Exams:

Memory or X Comprehension Fundamental 1:P or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: frocedure steps and cautions 10CRF55 55.41b I 10 55.43b I Content Administrative, normal, abnormal, and emergency operating procedures for the facility.

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes I Point Value: 1 System 10 No.: I 271000 I PRA: I No OCSOPSILT Page: 117 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam Safety 9

IZI Initial License Level Function: D LORT OCSOPSILT Page: 118 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 37 10: 11*1 NRO 37 Points: 1.00 The plant was at rated power. The Operator had just placed TIP 1 and 2 at the core top location, when the following annunciators alarmed:

  • DW PRESS HI-HI RV46 AlB
  • DW PRESS HI-HI RV46 C/D 12 minutes later, the Operator reports the following observations:

- IN SHIELD white light is energized

- DETECTOR POSITION displays 02

- IN SHIELD white light is de-energized

- DETECTOR POSITION displays 255

  • The TIP red light (Panel 11 F) is energized
  • NO TIPs can be moved lAW 405.2, Operation of the TIP System, which of the following states the required action for the stated conditions?

A. Manually retract TIP 1 locally B. Fire the shear valve for TIP 1 C. Manually retract TIP 2 locally D. Fire the shear valve for TIP 2 Answer: D Answer Explanation QID: 11*1 NRO 37 Question # I 37 I Developer I Date: JJR 15*14*2012 Knowledge and Ability Reference Information Importance Rating K&A RO I SRO OCSOPS ILT Page: 1190f241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 215001 Traversing In-core Probe 2.1.23 - Conduct of Operations: Ability to perform specific system and integrated plant 4.3 4.4 procedures during all modes of plant RO Tier 2 Group 2 General 405.2 RAP-C1f, C2f References D is Correct. The plant is at power with TIPs 1 & 2 at the core top. The provided annunciators show that a LOCA signal has been generated (3 psig Drywell pressure or RPV water level at or below 86"). These signals also isolate the Primary Containment and RPV, including the TIPs. On an isolation, the TIPs automatically retract and the ball valves close. Conditions show that with the Panel 11 F TIP red light on, then at least one TIP ball valve is open. It also shows that the in shield light for TIP 2 is de energized, which means that the TIP 2 has not retracted to the in shield position and the ball valve will be open. The ball valve normally auto closes when the TIP is retracted into the shield.

Explanation The TIP 2 detector position (lowest is in shield and counts up as the detector moves out of the shield) shows that it is not in shield. lAW the 405.2, with a ball valve open and cannot be closed, then it directs that the shear valve be fired for the applicable TIP.

A & C are Incorrect but plausible since the TIPs can be manually cranked locally, however 405.2 directs actuating the shear valve for this condition. Choice A also specifies the wrong TIP.

B is Incorrect but plausible if the applicant is confused about which TIP retracted.. It is not the correct TIP.

Lesson Plan 2621.828.0.0029, Nuclear Instrumentation NIS-10445, Given the EOP, describe in detail each Learning step/statement, including the technical basis, and how to verify rm each Question Source (New, Modified, Bank) Modified

EXAMINATION ANSWER KEY oc RO NRC Exam If Bank or Modified VISION System/Question 10: 718311 Question Source: ILT Bank Previous 2 Exams: No Memory or Comprehension X

Fundamental 3:SPK or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: Solve a Problem using Knowledge and its meaning 10CRF55 55.41b I 10 55.43b

.~

Content Administrative, normal, abnormal, and emergency operating procedures for the facility.

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes Point Value: 1 System 10 No.: 215001 PRA: I No Safety ~ Initial License Level 7

Function: D LORT OCSOPSILT Page: 121 of241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 38 10: 11-1 NR038 Points: 1.00 Given the following plant conditions:

  • The plant is at 100% power
  • The Electric Pressure Regulator (EPR) pressure transmitter fails, the EPR relay position strokes to the 0 % position.
  • No operator actions have yet occurred Based on these plant conditions, which one of the following parameters will initially lower?

A. Reactor Power B. Generator Output C. Reactor Pressure Vessel Pressure D. Mechanical Pressure Regulator Relay Position Answer: B Answer Explanation QID: 11-1 NRO 38 Question # J 38 I Developer I Date: JJR /5-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO SRO 245000 Main Turbine Gen. I Aux.

K1.08 - Knowledge of the physical connections and/or cause- effect relationships between 3.4 3.5 MAIN TURBINE GENERATOR AND AUXILIARY SYSTEMS and the following: Reactor/turbine pressure control system: Plant-Specific Level I RO I Tier 2 I Group I 2 General Turbine Tech GE 223R309 References Manual Tab 10 OCSOPS ILT Page: 122 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam B is Correct. If a loss of power occurs to the EPR, the actual relay position and setpoint indicators go downscale; The MPR takes control and regulates at a higher RPV pressure of 1024 psig. This causes the TCVs to close resulting in a loss of generator load initially because less steam is going to the turbine thus dropping generator load.

A is Incorrect but plausible if the applicant doesn't recall how the Explanation Turbine Control System responds to a loss of the EPR. Reactor power will increase.

C is Incorrect but plausible jf the applicant doesn't recall how the Turbine Control System responds to a loss of the EPR. Initial RPV pressure will rise.

D is Incorrect but plausible if the applicant doesn't recall how the Turbine Control System responds to a loss of the EPR. The MPR will rise as the EPR re lowers.

Lesson Plan 2621.828.0.0051, Turbine Control System EWA-2257, Given the EOP, describe in detail each Learning step/statement, including the technical basis, and how to verify ective/ or each ste Question Source (New, Modified, Bank) Bank If Bank or Modifi~d VISION System/Question 10: 607955 Question Source: ILT Bank Previous 2 Exams: No Memory or X Comprehension Fundamental 3:PEO or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: Predict an Event or Outcome OCSOPSILT Page: 123 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 10CRF55 55.41b I 7 I 55.43b I

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

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes I Point Value: 1 System 10 No.: 245000 PRA:

I No Safety IZI Initial License Level Function:

4 o LORT OCSOPS ILT Page: 124 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 39 Points: 1.00 The plant is shutdown and cooling down with the Shutdown Cooling System (SOC).

The following plant conditions currently exist:

  • All Recirculation Pumps are in service
  • SOC Pump A is tagged out of service due to an oil leak
  • RPV water level indicates 160"
  • RECIRC PUMP SUCTION TEMPS indicate 197 OF The following annunciators then alarmed:
  • 1B2 MN BRKR TRIP
  • 1B2 MN BRKR OL TRIP The Operator reports that RECIRC PUMP SUCTION TEMPS are rising. Which of the following states the required action to provide adequate core cooling for the given conditions?

A. Raise RPV water level up to at least 170".

B. Bypass the SOC isolation and restart the SOC System.

C. Establish alternate RPV cooldown with the RWCU System.

O. Initiate the Isolation Condensers lAW 307, Isolation Condenser System.

Answer: C IAnswer Explanation QID: 11-1 NRO 39 Question # I 39 I Developer I Date: JJR /5-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO I SRO

EXAMINATION ANSWER KEY oc RO NRC Exam 295021 Loss of Shutdown Cooling I 4 AK1.03 - Knowledge of the operational implications of the following concepts as they 3.9 3.9 apply to LOSS OF SHUTDOWN COOLING:

uate core cooli Level RO Tier 1 Group 1 General ABN-3 References C is Correct. The plant is shutdown and cooling down with SOC pumps Band C. The provided alarms show a loss of 480 volt USS 1B2, which is providing power to SOC pumps Band C. SOC A is tagged out due to an oil leak and is not available. Therefore, these conditions present a total loss of SOC. lAW ABN-3, Loss of Shutdown Cooling, is SOC is lost, then restore cooling lAW attachment ABN-3-3. Of the methods listed while in cold shutdown, aligning alternate cooling with RWCU is allowed and is available.

A is Incorrect but plausible. ABN-3 states that SOC is isolated, Explanation then raise RPV water level >185" to establish circulation flow through the steam separators.

B is Incorrect but plausible. The indications show a bus over load and loss of power to the SOC pumps. Even if a SOC isolation were to occur, bypassing the isolation would still not result in forced flow from SOC.

o is Incorrect but plausible. Initiating the ICs is listed as an alternate cooldown method lAW ABN-3, but there must be steam in the RPV. With coolant temperature currently at 197 OF, there is no steam to flow th h and be condensed the ICs.

Lesson Plan 2621.828.0.0045, Shutdown Cooling System Learning SOC-10445, Given a set of system indications or data, evaluate Objective/ and interpret them to determine limits, trends and system status.

Question Source (New, Modified, Bank) Bank OCSOPSILT Page: 126 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam If Bank Qr MQdified VISION System/Question 10: 616816 Question Source: ILT Bank Previous 2 Exams: No Memory or X Comprehension Fundamental 3:SPK or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: Solve a Problem using Knowledge and its meaning 10CRF55 55.41b I 8 55.43b I

Content Components, capacity, and functions of emergency systems.

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes Point Value: 1 System 10 No.: 295021 PRA:

I No Safety ~ Initial License Level 4

Function: o LORT OCSOPS ILT Page: 127 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 40 10: 11-1 NRO 40 Points: 1.00 The plant was at rated power when a LOCA then occurred. Plant conditions include the following:

  • CONT SPRAY FLOWS SYSTEM 1 indicates 4000 GPM
  • CONT SPRAY FLOWS SYSTEM 2 indicates 4100 GPM
  • Core Spray SYS 2 FLOW indicates 4200 GPM utilizing Core Spray Pumps NZ01 B/NZ03B An event then occurred resulting in Torus Water Level lowering. Torus Water Level has now stabilized. Current plant conditions include the following:
  • Torus water level is 102 inches
  • Torus pressure is 4.4 psig
  • Torus water temperature is 180 OF Which of the following Core Spray System(s), if any, have exceeded their NPSH limits?

A. None B. Core Spray System 1 C. Core Spray System 2 D. BOTH Core Spray Systems Answer: B Answer Explanation QID: 11-1 NRO 40 Question # I 40 I Developer / Date: JJR /5-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO I SRO oes OPS ILT Page: 128 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO 1\1 RC Exam 295030 Low Suppression Pool Water Levell 5 EK1.02 - Knowledge of the operational implications of the following concepts as they 3.5 3.8 apply to LOW SUPPRESSION POOL WATER LEVEL: Pum NPSH Level Tier 1 Group 1 General EMG-SP4 EOP User's Guide References B is Correct. The question stem describes an event where both Core Spray and Containment Spray Systems are in service to mitigate the events of a LOCA. An event occurs resulting in Torus level lowering (then stabilizing). Due to this event, and Explanation lAW EMG-SP4, Core Spray System 1 now does not meet it's NPSH requirements.

All distractors are Incorrect but plausible if the applicant does not co determine if NPSH has been violated lAW EMG-SP4.

Lesson Plan 2621.845.0.0056, Primary Containment Control Learning PCC-10445, Given a set of system indications or data, evaluate Objectivel and interpret them to determine limits, trends and system status.

Question Source (New, Modified, Bank) Modified If Bank or Modified VISION SystemlQuestion 10: 718225 Question Source: ILT Bank Previous 2 Exams: No Memory or X Comprehension Fundamental 3:SPR or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: Solve a Problem using References 10CRF55 Content 55.41b I 10 55.43b I

OCSOPS ILT Page: 129 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam Administrative, normal, abnormal, and emergency operating procedures for the facility.

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes I Point Value: 1 System 10 No.: 295030 PRA:

I No Safety ~ Initial License Level 5

Function: o LORT OCSOPS ILT Page: 130 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 41 10: 11-1 NR041 Points: 1.00 While at the controls during a fuel shuffle, you are notified that an irradiated fuel bundle was dropped while being moved over the core.

Which ONE of the following would be an accurate radiation monitoring response from this event, if the design basis release were to occur?

Panel 2R radiation monitor. ..

A. C5, SPENT FUEL POOL AREA, will indicate elevated radiation levels, and when tripped high, will isolate the DW vent/purge valves (after a time delay).

B. C10, FUEL POOL HI RANGE, will indicate elevated radiation levels, and when tripped high, will initiate the Standby Gas Treatment System (after a time delay).

C. C9, FUEL POOL LOW RANGE, will indicate elevated radiation levels, and when tripped high, will initiate the Standby Gas Treatment System (after a time delay).

D. 89, REACTOR OPEN FLR EQUIP HATCH, will indicate elevated radiation levels, and when tripped high, will isolate the DW vent/purge valves (after a time delay).

Answer: C IAnswer Explanation QID: 11-1 NRO 41 Question # I 41 I Developer / Date: JJR /5-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO SRO 295023 Refueling Acc Cooling Mode I 8 AK1.01 - Knowledge of the operational implications of the following concepts as they 3.6 4.1 apply to REFUELING ACCIDENTS: Radiation exposure hazards OCSOPS ILT Page: 131 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam Level RO Tier 1 Group 1 General RAP-10F3m References C is Correct. 'rhe question stem describes an event where an irradiated fuel bundle was dropped over the core. When radiation monitor C9 reaches 50mRlhr, it will trip RB ventilation, and start the Standby Gas Treatment System (SGTS) following a 2-min time delay.

A is Incorrect but plausible. It is true that C5 will indicate elevated radiation levels, however it will not trip the DW vent/purge valves. _I  ?

Explanation <...

B is Incorrect but plausible. It is true that C10 will jPdicate elevated radiation levels, however it will trip RB ventilation, and start the Standby Gas Treatment System (SGTS) following a 2 min time delay.

D is Incorrect but plausible. When B9 reaches 50mRlhr, it will trip RB ventilation, and start the Standby Gas Treatment System (SGTS) following a 2-min time delay, not isolate the DW valves.

Lesson Plan 2621.828.0.0033A, Plant Radiation Monitoring System 273-10449, State the function and interpretation of system Learning alarms, alone and in combination, as applicable in accordance o ive/ with the RAPS.

Question Source (New, Modified, Bank)

If Bank gr Modified I Bank VISION System/Question ID: 510828 Question Source: ILT 05-1 NRC Exam Previous 2 Exams: No oesops ILT Page: 132 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam Memory or X Comprehension Fundamental 3:SPK or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: S.olve a Problem using Knowledge and its meaning 10CRF55 55.41b I 11 55.43b I

Content Purpose and operation of radiation monitoring systems, including alarms and survey equipment.

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1*2 minutes I Point Value: 1 System 10 No.: 295023 PRA:

I No Safety 1:81 Initial License Level 8

Function: D LORT OCSOPSILT Page: 133 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 42 10: 11-1 NRO 42 Points: 1.00 The plant is at rated power. The Fire Protection System lineup is as follows:

  • FIRE SYSTEM POND PUMP 1 is in AUTO
  • FIRE SYSTEM POND PUMP 2 is in MANUAL
  • FIRE SYSTEM DIESEL PUMP 1 is in AUTO
  • FIRE SYSTEM DIESEL PUMP 2 is in AUTO The following annunciators then alarmed:
  • XFMRlTURB AREA FIRE
  • LFAP 2 FLOW ALARM If the fire system header pressure dropped to 60 psig, which of the following states the status of the Diesel Pumps and Pond Pumps?

Fire Pond Pumps Fire Diesel Pumps A. NO pumps operating BOTH pumps operating B. BOTH pumps operating BOTH pumps operating C. NO pumps operating Diesel pump 1 operating ONLY D. BOTH pumps operating Diesel pump 2 operating ONLY Answer: A IAnswer Explanation QID: 11-1 NRO 42 Question # I 42 I Developer 1 Date: JJR 15-14-2012 Knowled1le and Ability Reference Information Importance Rating K&A RO I SRO OCSOPS ILT Page: 134 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 600000 Plant Fire On~site / 8 AK2.03 - Knowledge of the interrelations 2.5 2.6 between PLANT FIRE ON SITE and the followin : Motors Level RO Tier 1 Group 1 General 333 RAP-N2a, N2b RAP-MFAP A(7-c)

References A is Correct. The question stem describes an event where a fire erupted in the area of the Main Transformer. In addition, fire header pressure dropped to 60 pSig. Diesel Fire pump 2 will auto start on fire header low pressure at 85 +/-10 psig. Diesel Fire pump 2 will auto start on fire header pressure at 75 +/-10 psig.

At 60 psig, both Diesel Fire pumps should have started and will be operating. Either Diesel Fire pump starting will trip the operating Fire Pond pump and also prevent the Pond pump in AUTO from starting therefore neither Pond pump will be operating at 60 psig.

Explanation B is Incorrect but plausible. Neither Pond pump will be operating.

C is Incorrect but plausible. Diesel pump 2 will be operating, not just Diesel pump 1.

D is Incorrect but plausible. Neither Pond pump will be operating. Diesel pump 1 will be operating, not just Diesel pump 2.

Lesson Plan 2621.828.0.030, Nuclear Steam Supply System NIS-1029, Given a drawing of the NSSS, trace the flowpaths and Learning locate the major components associated with the system, and its 0 within the ""U'~'I"""'~

Question Source (New, Modified, Bank)

If Bank Qr MQdifi~d I Modified VISION System/Question ID: 811698 Question Source: ILT Bank Previous 2 Exams: No

EXAMINATION ANSWER KEY oc RO NRC Exam Memory or X Comprehension Fundamental 1:1 or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: Interlocks, setpoints, or system (singular) response 10CRF55 55.41b I 7 55.43b I

Design, components, and functions of control and safety systems, Content including instrumentation, Signals, interlocks, failure modes, and automatic and manual features.

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes Point Value: 1 System ID No.: 600000 PRA: I No Safety 8

181 Initial License Level Function: o LORT OCSOPS ILT Page: 136 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 43 10: 11-1 NRO 43 Points: 1.00 Given the following:

  • The reactor is operating at 100% power when a turbine trip occurs
  • Reactor pressure spikes to 1061 psig for 3 seconds and then lowers to 1015 psig Which one of the following describes reactor recirculation pump status following this event?

A. All recirc pumps are running B. No recirc pumps are running C. ONLY recirc pumps C and 0 are running D. ONLY recirc pumps A, Band E are running Answer: C IAnswer Explanation QID: 11-1 NRO 43 Question # I 43 I Developer 1 Date: JJR 15-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO SRO 295025 High Reactor Pressure 13 EK2.04 - Knowledge of the interrelations 3.9 4.1 between HIGH REACTOR PRESSURE and the followin : ARI/RPTIATWS: Plant-Specific Level RO I Tier 1 I Group I 1 General RAP-E1a 420 References OCSOPS ILT Page: 137 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam C is Correct. Reactor recirc pumps A, Band E trip at 1051 psig (for 1.5 seconds) (ATWS high pressure trip); recirc pumps C and D trip if reactor pressure exceeds 1051 psig for 10.5 seconds.

For the given conditions, reactor pressure would not have exceeded 1051 psig for 10.5 seconds. Therefore, A, Band E Explanation pumps are tripped; C and D pumps are running.

All distractors are Incorrect but plausible if the applicant does not recall the Recirc Pump ATWS logic or recognize this logic initiated.

Lesson Plan 2621.828.0.0038, Reactor Recirculation System RRS-10441, Given the system logic/electrical drawings, describe the system trip signals, setpoints and expected

"""~<T""lrn r ... ~n'nse includin loss or failed ents.

Question Source (New, Modified, Bank) Bank If Bank Qr MQdified VISION System/Question ID: 506355 Question Source: ILT Bank Previous 2 Exams: No Memory or X Comprehension Fundamental 3:PEO or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: E.redict an Event or Qutcome 10CRF55 55.41b I 7 55.43b I

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

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes I Point Value: 1 System ID No.: I 295025 I PRA:

I No OCSOPS ILT Page: 13801241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam Safety [81 Initial License Level 3

Function: o LORT OCSOPSILT Page: 139 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 44 10: 11-1 NRO 44 Points: 1.00 The reactor was at rated power, when the following annunciators alarmed:

  • REACTOR PRESS - RX PRESS HI-HI I
  • REACTOR PRESS - RX PRESS HI-HIli Which of the following states (1) where the Feedwater Control System will control RPV water level in AUTO (PRIOR to any Operator actions), and (2) the procedurally required manual operator actions to control RPV water level?

(1) The Feedwater Control System will (2) Action control RPV water level at the...

Trip two feedwater pumps when RPV water level ...

A. pre-scram level setpoint begins to rise B. post-scram level setdown level setpoint begins to rise C. post-scram level setdown level setpoint reaches 142" D. pre-scram level setpoint reaches 142" Answer: B Answer Explanation QID: 11-1 NRO 44 Question # I 44 I Developer / Date: JJR /5-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO SRO 295006 SCRAM /1 AK2.02 - Knowledge of the interrelations 3.8 3.8 between SCRAM and the following: Reactor water level control system Level I RO I Tier I 1 I Group I 1 OCSOPS ILT Page: 140 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam General RAP-H1f ABN-1 References B is Correct. RAP-H1f and -H2f {RX Press Hi-Hi} will initiate an automatic reactor scram. The operator is required to verify actuation of the ~ scram level setdown and to perform followup actions of ABN-1. (SP-2 of RPV Control- No ATWS also says the same correct answer.) Following a scram and lowering RPV water level, feedwater level control will attempt to control RPV water level at the reactor level setdown setpoint (142")

(when feedwater level control is left in AUTO). ABN-1, Reactor Scram, requires that when RPV water level begins to rise, to trip two feedwater pumps. Then to place the main feed regulating Explanation valves in manual and close, them.

All distractors are Incorrect but plausible. The applicant may not recall if they are required to verify whether the Rx feedwater control system is at the pre or post scram setpoint. In addition, the post scram level setpoint is 142" (the feed control system automatically changes the Master Feedwater Controller setpoint from its initial value to 142"). If the applicant is not familiar with this value they may assume it is the point where they are required to close the Main Feed Reg Valves (which would be Lesson Plan 2621.828.0.0018, Feedwater Control System FCS-10444, Describe the interlock signals and setpoints for the Learning affected system components and expected system response ectivel includi loss or failed com nents.

Question Source (New, Modified, Bank) I Modified If Bank Qr MQdifi~d VISION System/Question 10: 667521 Question Source: ILT Bank Previous 2 Exams: No OCSOPS ILT Page: 141 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam Memory or X Comprehension Fundamental 3:SPK or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: .solve a Problem using Knowledge and its meaning 10CRF55 55.41b I 7 55.43b I

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

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1*2 minutes I Point Value: 1 System 10 No.: 295006 PRA: I No Safety 1

181 Initial License Level Function: D LORT oes OPS ILT Page: 142 of 241 02 April 2012

EXAMINATION ANSWER KEY DC RD NRC Exam 45 10: 11-1 NRO 45 Points: 1.00 The plant was at rated power when the crew entered ABN-12, Generator Excitation Equipment Malfunction, due to erratic operation of the Voltage Regulator.

(1) lAW ABN-12. which ONE of the following conditions would require the crew to scram the reactor? AND; (2) What is the reason for scramming at this time?

A. (1) Voltage control CANNOT be adjusted below 24.5 KV.

(2) This is above the design operating limit of the voltage regulator.

B. (1) Voltage control CANNOT be adjusted higher than 23.5 KV.

(2) This is below the design operating limit of the voltage regulator.

C. (1) Manual Voltage control failed to correct the voltage instability.

(2) To prevent permanent damage to the voltage regulator.

D. (1) Automatic Voltage control cannot be stabilized within 15 minutes.

(2) To prevent permanent damage to the voltage regulator.

Answer: C IAnswer Explanation QID: 11-1 NRO 45 Question # I 45 I Developer 1 Date: JJR 15-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO SRO 700000 Generator Voltage and Electric Grid Disturbances 1 6 AK3.01 - Knowledge of the reasons for the following responses as they apply to 4.6 4.6 GENERATOR VOL TAGE AND ELECTRIC GRID DISTURBANCES: Reactor and turbine trip criteria Level I RO I Tier I 1 I Group I 1 OCSOPS ILT Page: 143 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam General ABN-12 GEK-5522 References C is Correct. lAW ABN-12, during periods of erratic voltage regulation, manually scram the reactor if manual voltage control fails to correct the instability. This is to prevent possible permanent damage to the main generator voltage regulator.

Explanation A & B are Incorrect but plausible if the applicant does not recall the normal operating band for voltage regulation (23.3 - 24.7 KV).

D is Incorrect but plausible if the applicant does not recall when it is required to scram lAW ABN-12. The reason is correct but the uirement to scram is incorrect.

Lesson Plan 2621.828.0.0025, Main Generator Learning GEN-10450, Describe and interpret procedure sections and steps Objectivel for plant emergency or off-normal conditions that involve this system including personnel allocation and equipment operation lAW applicable ABN, EOP & EOP support procedures and EP res.

Question Source (New, Modified, Bank) New If Bank or MQdified VISION System/Question ID:

Question Source: N/A Previous 2 Exams:

Memory or X Comprehension Fundamental 1:P or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: Procedure steps and cautions OCSOPSILT Page: 144 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 10CRF55 55.41b I 10 I 55.43b I

Content Administrative, normal, abnormal, and emergency operating procedures for the facility.

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes I Point Value: 1 System 10 No.: 700000 PRA:

I No Safety 6

IZI Initial License Level Function: D LORT OCSOPS ILT Page: 145 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 46 10: 11-1 NRO 46 Points: 1.00 The plant was at rated power when an event occurred resulting in an ATWS. Note the EOP step below from RPV Control - with ATWS:

NO YES CONFIRM RECIRCULATION FLOW IS RUN BACK TO MINIMLt.1 lAW the EOP User's Guide, what is the basis for confirming recirculation flow is runback to minimum if the main generator is on-line?

A. To protect the recirculation pumps from carryunder.

B. To ensure the main turbine doesn't trip on high RPV water level.

C. To reduce recirculation pump power consumption during an emergency condition.

D. To prevent a main turbine runback by proactively reducing recirculation flow and reactor power.

Answer: B Answer Explanation QID: 11-1 NRO 46 Question # J 46 I Developer / Date: J..IR /5-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO I SRO OCSOPS ILT Page: 146 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 295037 SCRAM Conditions Present and Reactor Power Above APRM Downscale or Unknown /1 EK3.01 - Knowledge of the reasons for the following responses as they apply to SCRAM 4.1 4.2 CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN:

Recirculation pump trip/runback: Plant-RO Tier 1 Group 1 General RPVC - with ATWS EOP User's Guide References EOP B is Correct. lAW the EOP Users Guide, if the Main Turbine is on line, the Recirculation pump speeds are reduced prior to tripping them to prevent a large RPV level swell, or moisture separator drain tank level increase which could trip the Main Turbine. The Main Turbine on-line provides the priority heat sink during ATWS Explanation conditions.

All distractors are Incorrect but plausible reasons for reducing recirculation flow if the student does not recall the bases for this action.

Lesson Plan 2621.845.0.0052, RPV Control - no A TWS EWA-3055, Given a copy of RPV Control, describe in detail each Learning step or conditional statement, including technical basis, and how to each Question Source (New, Modified, Bank)

If Bank Qr MQdifi~d I Bank VISION System/Question 10: 811734 Question Source: ILT Bank Previous 2 Exams: No OCSOPS ILT Page: 147 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam Memory or X Comprehension Fundamental 1:B or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: Bases or purpose 10CRF55 55.41b I 10 55.43b I

Content Administrative, normal, abnormal, and emergency operating procedures for the facility.

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes Point Value: 1 System 10 No.: 295037 PRA: I No Safety ~ Initial License Level 1

Function: D LORT OCSOPSILT Page: 148 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 47 10: 11-1 NRO 47 Points: 1.00 The plant is at 25% power when an event required entry into ABN-10, Turbine Generator Trip.

IN ORDER, which IMMEDIATE OPERATOR ACTIONS are required by ABN-10, and what is the reason for that specific order?

Confirm the Main (1) is tripped, then confirm the Main (2) is tripped. This order will prevent (3) ill m ill A. Generator Turbine overspeeding the Turbine/Generator B. Turbine Generator overspeeding the Turbine/Generator C. Generator Turbine motoring the Main Generator D. Turbine Generator motoring the Main Generator Answer: D Answer Explanation QID: 11-1 NRO 47 Question # l 47 I Developer / Date: JJR /5-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO SRO 295005 Main Turbine Generator Trip I 3 AK3.04 - Knowledge of the reasons for the following responses as they apply to MAIN 3.2 3.2 TURBINE GENERATOR TRIP: Main generator trip Level I RO I Tier I 1 I Group I 1 OCSOPS ILT Page: 149 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam General GEK*5522 ABN*10 References D is Correct. lAW GEK*5522 section GEK-46517, Sequential Tripping and Prevention of Motoring, the reason the Main Generator is tripped immediately following a Main Turbine trip is if the Main Generator didn't immediately (or relatively soon after) the Main Turbine Trip, it will result in motoring the Main Generator and cause rapid heating of the L.P. turbine exhaust hoods and L.P. turbine last stage buckets. The immediate Explanation actions of ABN*10 (below 30% power) are to Confirm the Main Turbine is tripped, then Confirm the Main Generator is tripped.

All distractors are Incorrect but plausible if the applicant does not recall the order of Immediate Actions of ABN-1 0 or the reason. Overspeeding is plausible if the applicant believes that the Main Generator motoring would result in maintaining or the s of the Main Turbine/Generator.

Lesson Plan 2621.828.0.0050, Turbine and Turbine Auxiliaries Learning MTA-10444, Describe the interlock signals and setpoints for the Objective/ affected system components and expected system response includ loss or failed com"'I'\lnAI"I'Q Question Source (New, Modified, Bank) New If Bank Q[ MQdified VISION System/Question ID:

Question Source: N/A Previous 2 Exams:

Memory or X Comprehension Fundamental 1:B or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: Bases or purpose 10CRF55 55.41b 5 55.43b Content OCSOPSILT Page: 150 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes I Point Value: 1 System 10 No.: 295005 PRA:

I No Safety 3

IZI Initial License Level Function: D LORT OCSOPS ILT Page: 151 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 48 10: 11-1 NRO 48 Points: 1.00 The plant was at rated power when an event resulted in the crew executing ABN-30, Control Room Evacuation. Plant conditions include the following:

  • The REACTOR MODE SELECTOR switch is in SHUTDOWN
  • Annunciator SCRAM CONTACTOR OPEN is in alarm lAW ABN-30, which of the following actions are required BEFORE evacuating the Control Room?

A. Initiate the "A" Isolation Condenser AND start both EDG's.

B. Trip all Reactor Recirculation Pumps AND start both EDG's.

C. CONFIRM all control rods are inserted to or beyond position 04.

D. Initiate the "B" Isolation Condenser AND trip all but one Feed pump.

Answer: C

~wer Explanation QID: 11-1 NRO 48 Question # I 48 I Developer / Date: JJR /5-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO SRO 295016 Control Room Abandonment / 7 AA1.03 - Ability to operate and/or monitor the 3.0 3.1 following as they apply to CONTROL ROOM ABANDONMENT: RPIS Level l RO I Tier 1 I Group I 1 General ABN-30 References

EXAMINATION ANSWER KEY oc RO NRC Exam C is Correct. lAW ABN-30, before or immediately after evacuating the Control Room, Scram the reactor and confirm all control rods are inserted to or beyond position 04.

A is Incorrect but plausible. The 'B' Isolation Condenser (IC) is placed in service (the applicant may not recall the correct IC) and the EDGs start on a LOCA or LOOP signal. The applicant may assume with the CR Evacuation an EDG LOOP or LOCA start signal has been generated.

C is Incorrect but plausible since tripping the Reactor Explanation Recirculation Pumps is a correct action. The EDGs start on a LOCA or LOOP signal. The applicant may assume with the CR Evacuation an EDG LOOP or LOCA start signal has been generated.

o is Incorrect but plausible since plaCing the 'B' IC in service is a correct action. Tripping all the Feed Pumps is the correct action, however on a normal scram, one Feed Pump is left operating.

The applicant may confuse actions of ABN-30 with ABN-1, Reactor Scram. The question specifically asks for ABN-30 actions.

Lesson Plan 2621.828.0.0064, Alternate Shutdown Facility ASF-10450, Describe and interpret procedure sections and Learning steps for plant emergency or off-normal conditions that involve Objectivel this system including personnel allocation and equipment operation lAW applicable ABN, EOP & EOP support procedures, and EP ures.

Question Source (New, Modified, Bank)

I New If BaD k o[ Modified VISION System/Question 10:

Question Source: N/A Previous 2 Exams:

Des OPS ILT Page: 153 of 241 02 AprH 2012

EXAMINATION ANSWER KEY oc RO NRC Exam Memory or X Comprehension Fundamental 1:P or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: Procedure steps and cautions 10CRF55 55.41b I 10 55.43b I

Content Administrative, normal, abnormal, and emergency operating procedures for the facility.

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes Point Value: 1 System 10 No.: 295016 PRA:

I No Safety ~ Initial License Level 7

Function: D LORT OCSOPS ILT Page: 154 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 49 10: 11-1 NRO 49 Points: 1.00 The plant was at rated power when a LOCA occurred. The following conditions currently exist:

  • Drywell pressure is 13 pSig and lowering The following annunciators then alarmed:
  • S1A BRKR TRIP
  • BUS 1A UN Which of the following states the response of the Containment Spray Pumps 51A and 51C?

Containment Spray Containment Spray Pump 51A Pump 51C A. Trips AND can be re-started Trips AND can be re-started immediately after AC power is immediately after AC power is restored restored B. Trips AND will automatically Remains running restart after a time delay after AC power is restored C. Remains running Trips AND will automatically restart after AC power is restored D. Trips AND can be re-started Remains running after a time delay after AC power is restored Answer: D IAnswer Explanation QID: 11-1 NRO 49 Question # I 49 I Developer / Date: JJR /5-14-2012 OCSOPS ILT Page: 155 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam Importance Rating K&A 295024 High Orywell Pressure / 5 EA1.17 - Ability to operate and/or monitor the following as they apply to HIGH ORYWELL 3.9 3.9 PRESSURE: Containment spray: Plant-RO Tier 1 Group 1 General 237E901 sh 1 RAP-S1f BR 3000 References 116B8328 sh11a o is Correct. The question stem states there is a LOCA with Orywell with Containment Spray in service (which indicates OW press and temp are high). The question shows that containment spray pump 51A (powered from USS Bus 1A2, which is powered from 4160 VAC Bus 1C) is spraying the drywell, and that containment spray pump 51C (powered from USS Bus 1B2, which is powered from 4160 VAC Bus 10) is cooling the torus.

The alarm given describes a loss of the startup transformer (SA) to Bus 1A and onto Bus 1C (which powers bus 1A2). When this occurs, containment spray pump 51A will trip, and EOG1 will start and load onto bus 1C, which will automatically re-energize bus 1A2. But, there is a 200 second time delay after the EOG has Explanation loaded onto the bus to allow for sequenced loading. There is no auto start of the pumps, even if they were previously running when the startup power was lost. Therefore, containment spray pump 51A will trip, and can be manually re-started after a time delay after the bus power is restored. The loss of the startup transformer SA does not impact the running containment spray pump 51C, since it is still powered from the second startup transformer, SB. Therefore, it will remain running.

All distractors are Incorrect but plausible if the applicant does not recall the power supplies to the Containment Spray Pumps and does not recall which Containment Spray Pumps are affected a loss of to 4160 VAC Bus 1A & 1C.

OCSOPS ILT Page: 156 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam Lesson Plan 2621.828.0.0009, Containment Spray/ESW EWA-2257, Given the EOP, describe in detail each Learning step/statement, including the technical basis, and how to verify Objective/ or perform each step.

Question Source (New, Modified, Bank) Bank If Bank or MOdifigd VISION System/Question 10: 609265 Question Source: ILT 07-1 SRO NRC Exam Previous 2 Exams: No Memory or X Comprehension Fundamental 3:SPK or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: .solve a Problem using Knowledge and its meaning 10CRF55 55.41b I 7 55.43b I

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

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes Point Value: 1 System 10 No.: 295024 PRA:

I No Safety [gl Initial License Level 5

Function: D LORT OCSOPS III Page: 157 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 50 10: 11-1 NRO 50 Points: 1.00 The plant was at rated power making preparations to shutdown due to a loss of USS 1A2. An event then occurred resulting in a small leak in the Drywell. The following plant parameters were observed:

  • Drywell Pressure is 1.7 psig 0

The US has ordered the BOP to perform Support Procedure 27 (SP-27), Maximizing Drywell Cooling.

Which of the following Drywell Recirc Fans will be running following the completion of SP-27?

1. DW RECIRC FAN 1-1
2. DW RECIRC FAN 1-2
3. DW RECIRC FAN 1-3
4. DW RECIRC FAN 1-4
5. DW RECIRC FAN 1-5 A. 1 and 2 ONLY B. 4 and 5 ONLY C. 1,2, and 3 ONLY D. 1,2,3,4, and 5 Answer: B Answer Explanation QID: 11-1 NRO 50 Question # I 50 I Developer / Date: JJR /5-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO I SRO OCSOPS ILT Page: 158 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 295028 High Drywell Temperature / 5 EA1.03 - Ability to operate and/or monitor the 3.9 3.9 following as they apply to HIGH DRYWELL TEMPERATURE: I coo Ii Level Group 1 General EMG-SP27 PCC EOP References B is Correct. The normal system lineup for Drywell Recirc fans are the 1-1, 1-2, 1-4, & 1-5 fan in operation with the 1-3 fan off.

The question states there is a loss of USS-1A2. The power supply to Drywell recirc fans 1-1, 1-2, & 1-3 is MCC-1A23 via USS 1A2, therefore Drywell Recirc fans 1-1, 1-2, & 1-3 do not have power available to operate. Only Drywell Recirc fans 1-4 & 1-5 are available for Drywell cooling.

Explanation A, C, & D are Incorrect but plausible if the applicant does not recall the actions required by SP-27 or recall the power supply to the drywell cooling pumps. Drywell Recirc fans 1-1,1-2, & 1-3 are powered from MCC-1A23 via USS-1A2. Since USS-1A2 has been lost, those fans are not operable. Answers A, C, & D include one or more of Drywell Recirc fans 1-1, 1-2, or 1-3 as a choice, therefore which is incorrect. The only fans that have power available are I Recirc fans 1-4 & 1-5.

Lesson Plan 2621.845.0.0056, Primary Containment Control PCC-10445, Given a set of system indications or data, evaluate Learning and interpret them to determine limits, trends and system status.

Question Source (New, Modified, Bank) Bank If Bank Qr Modified VISION System/Question ID: 811733 Question Source: ILT Bank Previous 2 Exams: No Memory or X Cognitive Comprehension Fundamental 2:RI Level or Analysis Knowledge OCSOPS ILT Page: 159 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam NUREG 1021 Appendix B: Recognizing Interaction between systems (plural), including consequences and implications 10CRF55 55.41b I 7 I 55.43b I

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

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes I Point Value: 1 System 10 No.: 295028 PRA: I No Safety [&I Initial License Level 5

Function: o LORT OCS OPS ILT Page: 160 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 51 ID: 11-1 NRO 51 Points: 1.00 The plant was at 57% power in 5 loop operation. An event then occurred resulting in a trip of the *C* Reactor Reactor Recirculation Pump. All Immediate Operator Actions of ABN-2, Recirculation System Failures, have been completed by the crew.

The following conditions exist:

  • Reactor power is 45% and steady
  • Reactor recirculation flow is 6.5 X 104 GPM and steady lAW 202.1, Power Operation, which of the following actions are required?

A. Manually scram the reactor.

B. Raise reactor power to 60% with control rods.

C. Lower reactor power to 30% with control rods.

D. Raise reactor recirculation flow to 7.0 x 104 GPM.

Answer: D IAnswer Explanation QID: 11-1 NRO 51 Question # I 51 I Developer I Date: JJR /5-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO SRO 295001 Partial or Complete loss of Forced Core Flow Circulation /1 & 4 AA2.01 - Ability to determine and/or interpret 3.5 3.8 the following as they apply to PARTIAL OR COMPLETE lOSS OF FORCED CORE FLOW CIRCULATION: Power/flow map Level I RO I Tier I 1 I Group I 1 OCS OPS ILT Page: 161 of241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam General 202.1 References o is Correct. The question describes an event in which power and recirculation flow place the plant in the Exclusion Zone on the Power Operations Curve (Power/flow map). lAW procedure 202.1, Power Operation, the operator is to exit the zone using rods or flow. The recirculation flow in answer 0 places the plant outside of the zone.

A is Incorrect but plausible. 202.1 states if the Exclusion Zone is entered, EXIT it immediately using rods or flow. The applicant may interpret a reactor scram as an acceptable method of using rods, however this is incorrect and not the intent of this Explanation procedural direction.

B is Incorrect. RaiSing reactor power to 60% would move the plant out of the zone, but it would also pass the scram setpoint.

It is plausible the applicant may not interpret the Power Operations curve correctly and not recognize this.

C is Incorrect. Lowering reactor power to 30% would not place the plant outside of the zone. It is plausible the applicant may not interpret the Power Operations curve correctly and not rct"'"nize this.

Lesson Plan 2621.828.0.0038, Reactor Recirculation System EWA-22S7, Given the EOP, describe in detail each Learning step/statement, including the technical basis, and how to verify or each Question Source (New, Modified, Bank) Modified If Bank or MQdified VISION System/Question 10: 609320 Question Source: ILT 07-1 NRC Exam Previous 2 Exams: No Memory or X Cognitive Comprehension Fundamental 3:SPR Level or Analysis Knowledge OCSOPSILT Page: 162 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam NUREG 1021 Appendix B: ,Solve a Problem using References 10CRF55 55.41b I 10 I 55.43b I

Content Administrative, normal, abnormal, and emergency operating procedures for the facility.

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes I Point Value: 1 System 10 No.: 295001 PRA: I No Safety ~ Initial License Level 1&4 Function: o LORT OCSOPS ILT Page: 163 of241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 52 10: 11-1 NRO 52 Points: 1.00 The plant was at rated power when the 'following annunciator alarmed:

  • 1B2 MN BRKR OL TRIP If DC-B voltage was 133 volts just prior to the event, and is lowering at a constant 2 volts/minute, which of the following is correct? (SEE BELOW)

(Xl is currently OPERABLE and will be INOPERABLE in (Y) minutes.

ATTACHMENT ABN-4B-3 B BATTERY MINIMUM VOLTAGE FOR EQUIPMENT OPERABILITY REQUIRED BATTERY BATTERY LOAD VOLTAGE

  • 'A' IC V-14-31 Battery Charger available
  • 'A' IC V-14-34 Battery Charger available
  • C/U Iso. Valve V-16-2 Battery Charger available

EXAMINATION ANSWER KEY oc RO NRC Exam Answer: C IAnswer Explanation QID: 11-1 NRO 52 Question # I 52 I Developer 1 Date: JJR 15-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO SRO 295004 Partial or Total Loss of DC Pwr 1 6 AA2.03 - Ability to determine and/or interpret the following as they apply to PARTIAL OR 2.8 2.9 COMPLETE LOSS OF D.C. POWER: Battery voltage Level I RO I Tier 1 I Group I 1 General ABN-48 3028 sh. 1 References

EXAMINATION ANSWER KEY oc RO NRC Exam C is Correct. The alarm in the question stem shows a loss of USS 1B2. This results in the loss of all battery chargers to DC*A and DC*B. In 12 minutes, DC*B voltage will lower to 109 volts

=

(122- [2x12] 109), which is less than the minimum voltage for operability of 111 for the CRD pump.

A is Incorrect but plausible does not interpret the discharge rate for the B Battery correctly. The table provided shows that A IC V*

14-34 is inoperable when the charger is inoperable. Thus, the valve is inoperable at the time of the initial breaker annunciator.

Explanation B is Incorrect but plausible does not interpret the discharge rate for the B Battery correctly. In 9 minutes, DC-B voltage will lower to 115 volts (122-[2x9] :: 115), which is greater than the minimum of 113.3 volts for the pump. Thus the pump is still operable.

o is Incorrect but plausible does not interpret the discharge rate for the B Battery correctly. In 14 minutes, DC-B voltage will lower

=

to 105 volts (122*[2x14]) 105), which is greater than the minimum of 101 volts for the relays. Thus the relays are still nn,Dr!:lllhle.

Lesson Plan 2621.828.0.0012, DC Distribution Learning DCD*10445, Given a set of system indications or data, evaluate Objective/ and interpret them to determine limits, trends and system status.

Question Source (New, Modified, Bank) Bank If Bank Qr MQdified VISION System/Question 10: 666832 Question Source: ILT Bank Previous 2 Exams: No Memory or X Comprehension Fundamental 3:SPR or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: Solve a Problem using References OCSOPS ILT Page: 166 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 10CRF55 55.41b I 10 I 55.43b I

Content Administrative, normal, abnormal, and emergency operating procedures for the facility.

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes I Point Value: 1 System 10 No.: 295004 PRA: I No Safety t8j Initial License Level 6

Function: o LORT OCSOPSILT Page: 167 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 53 10: 11*1 NRO 53 Points: 1.00 The plant is operating at 100% power when a small unisolable leak develops on the common piping of the VARIABLE leg of the "B" Yarway level indicator.

(Assume this leak does not significantly affect reference leg temperature)

Based on the above information, which of the following will occur?

A. Reactor scram ONLY B. Reactor scram AND LO-LO initiations/isolations

c. LO-LO AND LO-LO-LO initiations/isolations ONLY D. Reactor scram plus LO-LO AND LO-LO-LO initiations/isolations Answer: B IAnswer Explanation QID: 11-1 NRO 53 Question # J 53 I Developer / Date: JJR /5-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO SRO 295031 Reactor Low Water Level/2 EA2.01 - Ability to determine and/or interpret 4.6 4.6 the following as they apply to REACTOR LOW WATER LEVEL: Reactor water level Level I RO I Tier 1 I Group I 1 General [[procedure::RAP.H1e|RAP.H1e], H3e, References H5e OCSOPSILT Page: 168 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam B is Correct. A leak on the B Yarway variable leg will indicate a lowering reactor water level. Yarway level instrument generates a reactor scram and a Lo-Lo isolation and initiation signal. Lo-Lo Lo signal is generated from a Gemac instrument. The applicant must correctly determine and interpret what indicated RPV water Explanation level will do and what protective functions will actuate from the event.

All distractors are Incorrect but plausible if the applicant does not correctly identify what protective functions will be actuated indicated RPV water level In\lllArllnn Lesson Plan 2621.828.0.0055, Rx Vessel Instrumentation Learning RVI-10445, Given a set of system indications or data, evaluate Objective/ and interpret them to determine limits, trends and system status.

Question Source (New, Modified, Bank) Bank If Bank or MQdifiid VISION System/Question 10: 608203 Question Source: ILT Bank Previous 2 Exams: No Memory or X Comprehension Fundamental 2:RI or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: Recognizing Interaction between systems (plural), including consequences and implications 10CRF55 55.41b I 7 55.43b I

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

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes I Point Value: 1 OCSOPS ILT Page: 169 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam System 10 No.: 295031 PRA: I No Safety 2

181 Initial License Level Function: o LORT OCSOPSILT Page: 170 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 54 10: 11-1 NRO 54 Points: 1.00 The plant was at power when an EMRV opened unexpectedly, and could NOT be closed. Primary Containment Control EOP was entered due to high Torus Water Temperature.

In the Torus Water Temperature leg, you are directed to enter RPV Control- No ATWS EOP, as shown in the EOP steps below:

YES ENTER EMG-3200,01A. RPV ~

CONTROl- NO AT\NS. AT ***

AND PERFORM IT CONCURRENTLY WITH Tl-IIS PROCEDURE Which of the following lists the basis for entering RPV Control - No A TWS?

Entering RPV Control- No ATWS is required to ensure that. ..

A. the total integrated heat available to be discharged to the torus through the open EMRV is minimized.

B. the reactor will be able to be shut down prior to reaching the requirement for boron injection.

C. the torus load limit will NOT be exceeded prior to the need to emergency depressurize.

D. the hydrodynamic loads on the EMRV discharge line components are minimized.

OCSOPSILT Page: 171 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam Answer: A IAnswer Explanation QID: 11-1 NRO 54 Question # I 54 I Developer I Date: JJR 15-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO SRO 295026 Suppression Pool High Water Temp. 15 2.4.18 - Emergency Procedures I Plan: 3.3 4.0 Knowledge of the specific bases for EOPs.

Level I RO I Tier 1 I Group I 1 General EOP User's Guide PCC EOP References OCSOPS ILT Page: 172 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam A is Correct. The EOP User's Guide provides the following: If any EMRV cannot be closed, shut down of the Reactor is ensured through entry to the RPV CONTROL - NO ATWS procedure. If the scram is successful, the total integrated heat available to be discharged to the Torus through the open EMRV(s) is minimized.

This action is appropriate even if the Technical Specification limit requiring a scram on high Torus water temperature (110" F) has not yet been reached.

B is Incorrect but plausible. This is the justification for the next requirement in Primary Containment Control to enter RPV Control - No ATWS: prior to reaching BIIT, then enter RPV Control - No ATWS. Entry into EMG-3200.01A to scram is Explanation required prior to reaching BIIT to ensure the reactor is shutdown.

From the EOP User's Guide: Scramming the Reactor before Torus temperature reaches the Boron Injection Initiation Temperature (BIIT) gives the benefit of knowing whether the Reactor will be able to be shut down prior to reaching the requirement for boron injection.

C is Incorrect but plausible since it is related to the basis for entering RPV Control- No ATWS due to high torus water level.

The applicant may not recall the correct basis.

o is Incorrect but plausible since it is also related to the basis for entering RPV Control - No ATWS due to high torus water level.

The a icant not recall the correct basis.

Lesson Plan 2621.845.0.0056, Primary Containment Control EWA-2257, Given the EOP, describe in detail each Learning step/statement, including the technical basis, and how to verify or each Question Source (New, Modified, Bank) I Bank If Bank o[ Modified VISION System/Question 10: 510679 Question Source: ILT Bank Previous 2 Exams: No OCSOPS ILT Page: 173 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam Memory or X Comprehension Fundamental 1:B or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: Bases or purpose 10CRF55 55.41b I 10 55.43b I

Content Administrative J normal, abnormal, and emergency operating procedures for the facility.

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes Point Value: 1 System 10 No.: 295026 PRA:

I No Safety 5

181 Initial License Level Function: o LORT oes OPS ILT Page: 174 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 55 Points: 1.00 The plant is at rated power when the following annunciator came into alarm (and is confirmed valid):

A. the feedwater control valves lockup due to a loss of air signal.

B. Service Air valve V-6S-2 is NOT isolated or is bypassed.

C. two or more control rods begin to drift into the core.

D. the RWCU system isolation valves close.

Answer: C IAnswer Explanation QID: 11-1 NRO 55 Question # I 55 J Developer / Date: JJR /5-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO SRO 295019 Partial or Total Loss of Inst. Air /8 2.4.50 - Emergency Procedures I Plan: Ability to verify system alarm setpoints and operate 4.2 4.0 controls identified in the alarm response manual.

Level I RO I Tier 1 I Group I 1 General RAP-H1a References OCSOPS ILT Page: 175 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam C is Correct. lAW RAP*H1a, a manual scram must be inserted if control air pressure lowers to 55 psig or if i:!: 2 control rods begin to drift into the core.

Explanation All distractors are Incorrect but plausible since they will occur due on lowering instrument air pressure, however the only choice listed that requires a scram per the alarm response rocedure is choice C.

Lesson Plan 2621.828.0.0043, Service, Instrument, and Breathing Air CAS-10449, State the function and interpretation of system Learning alarms, alone and in combination, as applicable in accordance ective/ with the RAPS.

Question Source (New, Modified, Bank) Modified If BallK or M~ulifiid VISION System/Question 10: 506586 Question Source: ILT Bank Previous 2 Exams: No Memory or X Comprehension Fundamental 1:P or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: .f.rocedure steps and cautions 10CRF55 55.41b I 10 55.43b J

Content Administrative, normal, abnormal, and emergency operating procedures for the facility.

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1*2 minutes I Point Value: 1 System 10 No.: I 295019 I PRA:

J No OCSOPS ILT Page: 176 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam Safety 181 Initial License Level Function:

8 o LORT OCSOPSILT Page: 177 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 56 10: 11-1 NRO 56 Points: 1.00 Given the following plant conditions and sequence of events occur:

  • A plant startup is in progress
  • RPV temperature is 200 OF
  • RBCCW temperatures are high in band and rising Based on these conditions, what action(s) is(are) required to improve RBCCW cooling lAW ABN-19, RBCCW Failure Response?

A. Trip all recirculation pumps ONLY.

B. Scram the reactor and trip all recirculation pumps.

C. Place TBCCW heat exchangers on Circulating water.

D. Place RBCCW heat exchangers on Circulating water.

Answer: C

~nswer Expla,!a_t_lo_n~~~ ~~___________________-'

QID: 11-1 NRO 56 Question # I 56 I Developer / Date: JJR /5-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO SRO 295018 Partial or Total Loss of CCW 18 2.1.23 - Conduct of Operations: Ability to perform specific system and integrated plant 4.3 4.4 procedures during all modes of plant operation.

Level I RO I Tier 1 I Group I 1 General ABN-19 References OCSOPSILT Page: 178 of 241 02 April 2012

EXAMINATION ANSWER KEY DC RO NRC Exam C is Correct. lAW ABN-19, rising RBCCW temperature is considered a partial loss of RBCCW cooling and requires entry into the ABN. The only choice provided which the ABN requires for this condition is to transfer TBCCW heat exchanger cooling to the Circulating Water System.

Explanation A & B are Incorrect but plausible since they are actions required by the ABN, however only if RPV temperature is >212 OF. The question states RPV temperature is 200 of.

o is Incorrect but plausible if the applicant does not recall the actions of ABN-19 or believes that RBCCW heat exchangers can be ned to the Circulati Water which cannot.

Lesson Plan 2621.828.0.0035, Reactor Building Closed Cooling Water RBC-10450, Describe and interpret procedure sections and Learning steps for plant emergency or off-normal conditions that involve Objective/ this system including personnel allocation and equipment operation in accordance with applicable ABN, EOP and EOP su and EP Procedures.

Question Source (New, Modified, Bank) Bank If Bank or Modified VISION System/Question 10: 607825 Question Source: ILT Bank Previous 2 Exams: No Memory or X Comprehension Fundamental 1:P or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: Procedure steps and cautions 10CRF55 55.41b I 10 55.43b I

Content Administrative, normal, abnormal, and emergency operating procedures for the facility.

OCSOPS ILT Page: 179 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes I Point Value: 1 System 10 No.: 295018 PRA: I No Safety ~ Initial License Level 8

Function: D LORT OCSOPS ILT Page: 180 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 57 10: 11*1 NRO 57 Points: 1.00 The plant was at rated power when a complete Loss of Offsite Power (LOOP) occurred.

The event was captured on the Plant Process Computer (SEE BELOW).

Which of the following correctly states the trend of Reactor Power, RPV Pressure, and RPV Water Level the FIRST two minutes following the LOOP? (Assume NO operator action)

OCSOPS1LT Page: 181 of241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam Reactor Power RPV Pressure RPV Water Level A. Lower ONLY Lower ONLY Lower ONLY B. Rise, then Lower Rise, then Lower Lower, then Rise C. Lower ONLY Rise, then Lower Lower, then Rise D. Rise, then Lower Rise, then Lower Lower ONLY Answer: C IAnswer Explanation QID: 11-1 NRO 57 Question # I 57 I Developer / Date: JJR / 5-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO SRO 295003 Partial or Complete Loss of AC / 6 AA2.02 - Ability to determine and/or interpret the following as they apply to PARTIAL OR 4.2 4.3 COMPLETE LOSS OF A.C. POWER: Reactor power, pressure, and level Level I RO I Tier 1 I Group I 1 General ABN-36 ABN-10 References OCSOPS ILT Page: 182 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam C is Correct. Immediately occurring on a Loss of Offsite Power (LOOP) from rated power, RPS de-energizes, the MSIVs close, the Turbine Trips, and all Reactor Feed Pumps trip. When RPS de energizes, all Control Rods will immediately insert causing reactor power to lower, and trend lower. When the Turbine trips (on load rejection) and MSIVs close, RPV Pressure will rise greater than the Isolation Condenser ATWS setpoint (1051 psig) and EMRV setpoints (1065#) for a few seconds. After that, Isolation Condensers immediately go in service and RPV Pressure will lower at a steady rate (until the RPV is depressurized without operator action). Immediately after the LOOP, all Reactor Feed Pumps trip. This, combined with all Control Rods inserting and the rapid drop in reactor power, will Explanation result in RPV water level 'shrink'. As the transient stabilizes (within about 30 seconds) RPV water level will start to rise. In addition, 60 seconds after the LOOP, both CRD pumps will start resulting in an additional rise in RPV water level.

All distractors are Incorrect but plausible if the applicant does not recall plant critical parameters immediately following a LOOP. It is plausible that they would think reactor power would rise with and RPV pressure rise, however the negative reactivity from all control rods inserting is much greater than the positive reactivity added by voids collapsing. It is also plausible they might not recall that RPV water level will start rising following a transient that results in RPV water level 'shrink'.

Lesson Plan 2621.828.0.0016, Electrical Distribution System ACD-10450, Describe and interpret procedure sections and Learning steps for plant emergency or off-normal conditions that involve Objectivel this system including personnel allocation and equipment operation lAW applicable ABN, SDRP, EOP & EOP support res and EP Procedures.

Question Source (New, Modified, Bank)

If Bank Qr Modified I New VISION System/Question ID:

Question Source: N/A Previous 2 Exams:

OCS OPS ILT Page: 183 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam Memory or X Comprehension Fundamental 3:PEO or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: Predict an Event or Qutcome 55.41b I 5 55.43b I

Facility operating characteristics during steady state and transient 10CRF55 conditions, including coolant chemistry, causes and effects of Content temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes Point Value: 1 System 10 No.: 295003 PRA:

I No Safety ~ Initial License Level 6

Function: D LORT OCSOPS ILT Page: 184 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 58 10: 11*1 NRO 58 Points: 1.00 Which of the following have a Control Room annunciator to indicate a potential Liquid Off-site Radioactivity Release is in progress?

A. Radwaste Overboard AND Sump 1-5 Collection Pit B. Emergency Service Water AND Condensate Transfer C. Service Water AND Reactor Building Closed Cooling Water D. Radwaste Service Water AND Turbine Building Closed Cooling Water Answer: C IAnswer Explanation QID: 11-1 NRO 58 Question # I 58 I Developer / Date: JJR /5-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO SRO 295038 High Off*site Release Rate / 9 EK2.06 - Knowledge of the interrelations between HIGH OFF-SITE RELEASE RATE and 3.4 3.7 the following: Process liquid radiation monitoring system Level I RO I Tier 1 I Group I 1 General RAP-10F3g RAP-10F3f ABN-27 References OCSOPS ILT Page: 185 of241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam C is Correct. The Service Water System and RBCCW System are monitored by the Process Radiation Monitoring System. High activity in either system will result in a Control Room annunciator and entry into ABN-27, Inadvertent Overboard Radioactive Liquid Release or Cross-Contamination.

Explanation A is Incorrect but plausible. The 1-5 Sump Collection Pit is monitored, but the Radwaste Overboard monitoring system is retired and no longer functions.

B & 0 Incorrect but plausible if the applicant believes these systems are monitored for radioactivity in the Control Room, which are not.

Lesson Plan 2621.828.0.0033A, Plant Radiation Monitoring System Learning 273-10449, State the function and interpretation of system Objectivel alarms, alone and in combination, as applicable in accordance with the RAPS.

Question Source (New, Modified, Bank) New If Bank or Modified VISION System/Question 10:

Question Source: N/A Previous 2 Exams:

Memory or X Comprehension Fundamental 1:F or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: Eacts 10CRF55 55.41b I 11 55.43b I

Content Purpose and operation of radiation monitoring systems, including alarms and survey equipment.

Justification for LORT questions with N/A KIA values < 3.0

EXAMINATION ANSWER KEY oc RO NRC Exam Time to Complete: 1-2 minutes I Point Value: 1 System 10 No.: 295038 PRA:

I No Safety 9

181 Initial License Level Function: o LORT OCSOPSILT Page: 187 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 59 10: 11-1 NRO 59 Points: 1.00 The plant is at rated power when an ATWS occurred.

Which of the following describes two ways to add negative reactivity to the core under these conditions?

A. Lowering RPV Pressure AND raising RPV water level B. Lowering RPV pressure AND lowering reactor water level C. Initiating Standby Liquid Control AND raising RPV water level D. Initiating Standby Liquid Control AND lowering RPV water level Answer: 0 IAnswer Explanation QID: 11-1 NR059 Question # I 59 I Developer I Date: JJR I 5-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO SRO 295015 Incomplete SCRAM 11 AK1.03 - Knowledge of the operational implications of the following concepts as they 3.8 3.9 apply to INCOMPLETE SCRAM: Reactivity effects Level I RO I Tier 1 I Group I 2 General RPVC - with ATWS EOP User's Guide References EOP OCSOPS ILT Page: 188 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam o is Correct. Two methods used to lower reactor power (add negative reactivity) are injecting SLC (adds boron which is a poison) and lowering RPV water level (voids core).

A is Incorrect but plausible if the applicant does not recall the actions to reduce reactor power in the RPV Control

  • with A TWS EOP. Lowering RPV Pressure reduces RPV Temperature adding positive reactivity. Raising RPV water level collapses voids adding positive reactivity.

Explanation C is Incorrect but plausible if the applicant does not recall the actions to reduce reactor power in the RPV Control* with ATWS EOP. It is true lowering RPV water level will add negative reactivity (voids core) however lowering RPV Pressure reduces RPV Temperature adding positive reactivity.

o is Incorrect but plausible if the applicant does not recall the actions to reduce reactor power in the RPV Control* with A TWS EOP. It is true initiating SLC adds negative reactivity however raiSing RPV water level collapses voids which adds positive Lesson Plan 2621.845.0.0053, RPV Control - with ATWS EWA-22S7, Given the EOP, describe in detail each Learning step/statement, including the technical basis, and how to verify or each Question Source (New, Modified, Bank) Modified If flank Qr MQdified VISION System/Question 10: 332471 Question Source: Dresden ILT Exam Bank Previous 2 Exams: No Memory or X Comprehension Fundamental 1:F or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: Facts OCSOPSILT Page: 189 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 10CRF55 55.41b I 1 I 55.43b I

Fundamentals of reactor theory, including fission process, neutron Content multiplication, source effects, control rod effects, criticality indications, reactivity coefficients, and poison effects.

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes I Point Value: 1 System 10 No.: 295015 PRA:

I No Safety 1

181 Initial License Level Function: [] LORT OCSOPS ILT Page: 190 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 60 10: 11-1 NRO 60 Points: 1.00 The plant was at rated power. Plant conditions include the following:

  • A operator is inserting a TIP into the core on #2 TIP machine
  • An event then resulted in a spurious LOCA signal In addition to the TIP purge valve closing, which one of the following statements correctly describes the response of the TIP system?

First the (1). Then the (2) ill (2}

A. TIP drive withdraws the detector shear valve fires to the in-shield position B. TIP drive withdraws the detector ball valve closes to the in-shield position C. ball valve closes TIP drive withdraws the detector to the in-shield position D. shear valve fires TIP drive withdraws the detector to the in-shield position Answer: B

! Answer Explanation QID: 11-1 NRO 60 Question # I 60 I Developer / Date: JJR /5-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO SRO 295020 Inadvertent Cont. Isolation / 5 & 7 AK2.08 - Knowledge of the interrelations between INADVERTENT CONTAINMENT 2.5 2.6 ISOLATION and the following: Traversing in-core probes: Plant-Specific

EXAMINATION ANSWER KEY oc RO NRC Exam Level RO Tier 1 Group 2 General 405.2 References B is Correct. The stem provides a condition where a spurious LOCA signal was received. On a Primary Containment isolation signal (LOCA signal), any TIP detectors that are not in-shield will automatically retract. Once in-shield, the ball valve(s) will Explanation automatically close.

All distractors are Incorrect but plausible if the applicant does not recall what happens to the TIPs on a containment isolation nal.

Lesson Plan 2621.828.0.0029, Nuclear Instrumentation NIS-10444, Describe the interlock signals and setpoints for the Learning affected system components and expected system response includi loss or failed I'nlrnn.nn,,,nlrc:

Question Source (New, Modified, Bank) Modified If Bank or Modifi§d VISION System/Question ID: 608248 Question Source: ILT Bank Previous 2 Exams: No Memory or X Comprehension Fundamental 1:1 or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: Interlocks, setpoints, or system (singular) response 10CRF55 55.41b I 7 55.43b l

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

OCSOPS ILT Page: 192 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes I Point Value: 1 System 10 No.: 295020 PRA: I No Safety 5&7 181 Initial License Level Function: D LORT OCSOPSILT Page: 193 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 61 10: 11-1 NR061 Points: 1.00 The following is a partial summary of steps contained in the temperature leg of the Secondary Containment Control EOP:

IF A PRIMARY SYSTEM IS DISCHARGING INTO SECON DARY CON TAINMENT:

1. BEFOREA PARAMETER (RADIATION.

TEMPERATliRE. OR LE\iEL) REACHES AMAXSAFEVAUJE ENTER R PV C ONm OL* NO ATWS 2.IF THE SAMI: PARAMETER EXCEEDS AMAXSAFEVAUJE IN 2AREAS.

EMERGENCY DEPRESS URIZE lAW the EOP Users Guide, which of the following states the bases for Emergency Depressurization above?

1. It places the RPV in the lowest energy state.
2. It reduces the driving head on primary systems discharging into the Secondary Containment.
3. It allows RPV injection from low pressure systems to makeup for the primary system leak.
4. It minimizes the amount of energy available to be deposited into the Primary Containment.

A. 1 ONLY B. 1 and 2 ONLY C. 2 and 3 ONLY D. 1,3, and 4 ONLY Answer: B Answer Explanation QIO: 11-1 NRO 61 Question # I 61 I Developer / Date: JJR /5-14-2012 Knowledge and Ability Reference Information OCSOPS ILT Page: 194 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam Importance Rating K&A 295032 High Secondary Containment Area Temperature / 5 EK3.01 - Knowledge of the reasons for the following responses as they apply to HIGH 3.5 3.8 SECONDARY CONTAINMENT AREA TEMPERATURE: Emergency/normal Tier 1 Group 2 General SCC EOP EOP User's Guide References B is Correct and A is Incorrect. lAW the EOP Users Guide, the temperature increases is so wide spread that is poses a direct threat to secondary containment integrity, equipment located in the secondary containment or continued safe operation. ED will place the plant in its lowest energy state and will reduce the driving head and flow from primary systems that are discharging into the secondary containment.

Explanation C is Incorrect but plausible. It is true that lowering RPV pressure will make alternate, low pressure systems available for RPV injection, but it is not the bases for the ED.

o is Incorrect but plausible. ED is performed by opening the EMRVs which releases the energy from the RPV into the Torus.

ED does not reduce the amount of energy to be released to the containment.

Lesson Plan 2621.845.0.0057, Secondary Containment Control Learning SCC-3082, Using the Secondary Containment Control EOP, Objective/ evaluate the technical basis for each step and apply this evaluation to determine the correct course of action under An'lIArIAru''U conditions.

Question Source (New, Modified, Bank) Bank

EXAMINATION ANSWER KEY oc RO NRC Exam If Bank Q[ MQdifi~d VISION System/Question 10: 663558 Question Source: ILT 08-1 NRC Exam Previous 2 Exams: No Memory or X Comprehension Fundamental 1:8 or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: Bases or purpose 10CRF55 55.41b I 10 55.43b I

Content Administrative, normal, abnormal, and emergency operating procedures for the facility.

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes I Point Value: 1 System 10 No.: 295032 PRA: I No Safety [81 Initial License Level 5

Function: D LORT OCSOPSILT Page: 196 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 62 10: 11 ..1 NRO 62 Points: 1.00 The plant is shutdown for a refuel outage when an event occurred requiring entry into the Secondary Containment Control EOP. SP-50, Reactor Building Ventilation Restart, has just been completed.

A short time later the following annuciator came into alarm and was confirmed valid:

  • RX BLDG - VENT HI Which of the following correctly states the impact of this alarm? (Assume NO operator action)

A. Reactor Building temperatures will rise due to the reduction in Reactor Building forced air flow.

B. Reactor Building LlP will become less negative due to the reduction in Reactor Building forced air flow.

C. Air from the Reactor Building will be directed through filters prior to discharge, to minimize the off-site radioactivity release.

O. Air from the Reactor Building will NOT be directed through SGTS filters prior to discharge and thus the off-site radioactivity release is riSing.

Answer: 0 IAnswer Explanation QID: 11-1 NRO 62 Question # I 62 I Developer 1 Date: JJR 15..14..2012 Knowled-.ge and Ability Reference Information Importance Rating K&A RO SRO 295033 High Secondary Containment Area Radiation Levels 1 9 EA1.03 .. Ability to operate and/or monitor the 3.8 3.8 following as they apply to HIGH SECONDARY CONTAINMENT AREA RADIATION LEVELS:

Secondary containment ventilation OCSOPS ILT Page: 197 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam Level RO Tier 1 Group 2 General SCC EOP EOP User's Guide SP-50 References o is Correct. The conditions in the stem indicate the RB ventillation monitors are indicating HI-HI (VENT HI alarm). If the RB vent system is secured and certain conditions are met, Secondary Containment Control EOP overrides directs re-start of the normal RB vent system by performing SP-50. In both the EOP override and in the SP-50, it requires verification that the RB vent rad monitors are not tripped << 9 mr/hr). In the SP-50, it requires the removal/insertion of EOP bypass plugs. When these are removed/installed, the auto start feature of SGT and isolation of normal RB vent is bypassed. Therefore a valid high-high signal Explanation on the RB vent rad monitors has no effect of either the normal RB vent system or SGTS. Therefore, there is no change in forced air flow in the RB (which would occur if SGTS did auto start and normal vent isolated), no air will be discharged through the SGTS filters, and normal RB vent system remains in service (no filtering to minimize offsite release).

All distractors are Incorrect but plausible if the applicant does not recall the actions of SP-50 or believes that RB ventilation will secure and the SGTS will start.

Lesson Plan 2621.845.0.0057, Secondary Containment Control SCC-3082, Given the EOP, describe in detail each Learning step/statement, including the technical basis, and how to verify or each Question Source (New, Modified, Bank) Modified If Bank Qr Modified VISION System/Question 10: 609041 Question Source: ILT Bank Previous 2 Exams: No Memory or X Cognitive Comprehension Fundamental 3:PEO Level or Analysis Knowledge OCSOPS ILT Page: 198 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam NUREG 1021 Appendix B: Predict an Event or Qutcome 10CRF55 55.41b I 7 I 55.43b I

Design, components, and functions of control and safety systems, Content including instrumentation, Signals, interlocks, failure modes, and automatic and manual features.

Justification for lORT questions with N/A KIA values < 3.0 Time to Complete: 1*2 minutes I Point Value: 1 System 10 No.: 295033 PRA:

I No Safety 9

181 Initial license level Function: o lORT oes OPSILT Page: 199 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 63 10: 11*1 NRO 63 Points: 1.00 A plant startup is in progress. Plant conditions include the following:

  • All APRMs indicate 12% power
  • All IRMs are on Range 10
  • The mode switch is in STARTUP 4
  • Recirculation flow is 11 x 10 gpm
  • A spike in reactor pressure to 1043 psig
  • A spike in reactor power to 40%

What is the correct status of the reactor (assume NO operator action)?

A. At power B. Scrammed due to High RPV Pressure C. Scrammed due to High IRM neutron 'Hux D. Scrammed due to High APRM neutron flux Answer: C Answer Explanation QID: 11-1 NRO 63 Question # I 63 I Developer I Date: JJR /5-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO SRO 295006 SCRAM 11 AA2.05 - Ability to determine and/or interpret the following as they apply to SCRAM:

3.5 3.6 Whether a reactor SCRAM has occurred Level I RO I Tier I 1 I Group I 2 OCSOPSILT Page: 200 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam General RAP-G1e References C is Correct. Based on conditions in the question stem (Mode Switch in STARTUP at 12% power). the reactor is operating on IRM Range 10. The scram setpoint for IRM Range 10 is 38% (on 0-40% scale) and 118% (on 0-125% scale). APRM indication rising to 40% would exceed both these setpoints (regardless of which one the IRMs are set to) and a reactor scram would occur on High IRM flux.

Explanation A is Incorrect but plausible if the applicant does not recall the plant would scram with the Mode Switch in STARTUP under these conditions.

B is Incorrect but plausible if the applicant does not recall the High RPV Pressure Scram setpoint (1045 psig).

o is Incorrect but plausible if the applicant believes a scram would occur but does not recognize the scram was a result of Hi IRM neutron fI not h APRM neutron flux.

Lesson Plan 2621.828.0.0037, Reactor Protection System Learning RPS-10453, Explain or describe how this system is interrelated with other ms.

Question Source (New, Modified. Bank) Bank If Bank Qr Modified VISION System/Question 10: 510863 Question Source: ILT 05-1 NRC Exam Previous 2 Exams: No Memory or X Comprehension Fundamental 1:1 or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: Interlocks, setpoints, or system (singular) response OCSOPSILT Page: 201 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 10CRF55 55.41b I 10 I 55.43b I

Content Administrative, normal, abnormal, and emergency operating procedures for the facility.

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes I Point Value: 1 System 10 No.: 295006 PRA:

I No Safety ~ Initial License Level 1

Function: o LORT OCSOPS ILT Page: 202 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 64 10: 11-1 NR064 Points: 1.00 The plant is at rated power with all systems normally aligned. You are performing a walkdown of the control room prior to relieving the watch. Electrical Maintenance had completed maintenance activities on Panel2R on the previous shift. With NO Control Room annunciators in alarm, you note the following:

REACTOR BUILDING I

,r-"'\ ,"

,~~ . lU'

"-.,J MRlHI!

INOtCAtQR& ~rOR&

TRlPUHlT tRIP"'"T Which of the following Tech Spec LCOs are impacted, jf any?

A. 3.5, Containment ONLY B. NO Tech Spec LCOs are impacted C. 3.1, Protective Instrumentation ONLY D. 3.1, Protective Instrumentation AND 3.5, Containment OCS OPS IlT Page: 203 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam Answer: D Answer Explanation QID: 11-1 NRO 64 Question # I 64 I Developer / Date: J ..IR /5-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO SRO 295034 Secondary Containment Ventilation High Radiation /9 2.2.36 - Equipment Control: Ability to analyze 3.1 4.2 the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations.

Level I RO I Tier 1 I Group I 2 General TS 3.1 RAP-10F1f GE 706E841 References TS3.5 OCSOPS ILT Page: 204 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam o is Correct. The question shows that both Reactor Building Ventilation radiation monitors are in the alarm condition following electrical maintenance activities on Panel 2R. The radiation monitors in alarm should result in a VENT HI annunicator, RBHVAC isolation, and the auto start of the SGTS.

SGTS & RBHVAC Annunciators L-1-b, L-4-b, L-4-c, and L-8-c are expected to be in alarm when RBHVAC isolates and the SGTS auto starts. TS 3.5.B.5 states that two separate SGTS shall be operable. The system that was selected as the priority system did not auto start and cannot be considered operable. TS 3.5 has an LCO impacted. With no other annunciators in alarm and no other indications of a condition in which a high radiation condition would exist in the Reactor Building ventilation system, it can be assumed that both Rx Bldg ventilation rad monitors are inoperable. TS 3.1 also is impacted.

Explanation Answer A is Incorrect but plausible. TS 3.1 states that at least 1 RBHVAC monitor must be Operable. Since both RBHVAC monitors are affected, TS 3.1 has an LCO impacted.

Answer C is Incorrect but plausible. The question stem shows that there is a Reactor Building Ventilation radiation monitor in the alarm condition following electrical maintenance activities on Panel 2R. With no additional annunciators in the Control Room in alarm (as stated in the question stem), the SGTS did not auto start as required. SGTS & RBHVAC Annunciators L-1-b, L-4-b, L-4-c, and L-8-c are expected to be in alarm when RBHVAC isolates and the SGTS auto starts. TS 3.5.B.5 states that two separate SGTS shall be operable. The system that was selected as the priority system did not auto start and cannot be considered operable. TS 3.5 has an LCO impacted.

B is Incorrect but plausible. See explanation for distractors A &

C. Both TS 3.1 and 3.5 are imu.:;.. IL........

Lesson Plan 2621.828.0.033A, Plant Radiation Monitoring System Learning 272-10453, Explain 01' describe how this system is interrelated ectivel with other Question Source (New, Modified, Bank) Modified OCSOPSILT Page: 205 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam If Bank Q[ MQdified VISION System/Question 10: 811735 Question Source: ILT Bank Previous 2 Exams: No Memory or X Comprehension Fundamental 3:SPK or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: ,Solve a Problem using Knowledge and its meaning 10CRF55 55.41b I 10 55.43b I

Content Administrative, normal, abnormal, and emergency operating procedures for the facility.

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes I Point Value: 1 System 10 No.: 295034 PRA:

I No Safety 9

181 Initial License Level Function: D LORT OCSOPSILT Page: 206 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 65 ID: 11-1 NR065 Points: 1.00 The plant is at rated power when an event resulted in a fuel clad failure.

lAW ABN-26. High Main Steam I Off-Gas I Stack Effluent Activity. which of the following conditions would procedurally require the crew to insert a manual scram?

A. BOTH OFF GAS HI-HI annunciators have been in alarm for 17 minutes.

B. The STACK EFFLUENT HI-HI annunciator has been in alarm for 20 minutes.

C. ONE Main Steam Line Radiation Monitor is indicating 900 mRlhr with Off-gas activity steady.

D. TWO Main Steam Line Radiation Monitors are indicating 600 mRlhr with Off-gas activity rising.

Answer: A Answer Explanation QID: 11-1 NRO 65 Question # I 65 I Developer I Date: JJR I 5-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO SRO 295017 High Otf-site Release Rate /9 AA1.10 - Ability to operate and/or monitor the 3.6 3.7 following as they apply to HIGH OFF-SITE RELEASE RATE: RPS Level I RO I Tier 1 I Group I 2 General ABN-26 References oes OPS ILT Page: 207 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam A is Correct. lAW ABN-26, if both Off-Gas Hi-Hi alarms have not cleared within 15 minutes, the crew must operate RPS and insert a manual scram lAW ABN-1, Reactor Scram.

B is Incorrect but plausible since it is an annuciator that will alarm as a result of a fuel failure and it has been in alarm> 15 Explanation min, however the requirement to scram is from the Off-Gas Hi-Hi alarms.

C & 0 are Incorrect but plausible. The requirement is to scram when two or more Main Steam Radiation Monitors are> 800 mRlhrwith Lesson Plan 2621.828.0.0033A, Plant Radiation Monitoring System 273-0838, Given auto isolation setpoints, list or identify Learning cause(s), system response, and affected Process RAO Monitors C;:VCtT~lm com Question Source (New, Modified, Bank) New If Bank or Modified VISION System/Question 10:

Question Source: N/A Previous 2 Exams:

Memory or X Comprehension Fundamental 3:SPR or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: Solve a Problem using References 10CRF55 55.41b I 10 55.43b I

Content Administrative, normal, abnormal, and emergency operating procedures for the facility.

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes I Point Value: 1

EXAMINATION ANSWER KEY oc RO NRC Exam System 10 No.: 295017 PRA: I No Safety 181 Initial License Level Function:

9 o LORT OCSOPS ILT Page: 209 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 66 10: 11-1 NR066 Points: 1.00 The plant was at rated power when an event occurred which allowed the use of Transient Alarm Response.

lAW OP-OC-1 01-111-1 001, Strategies For Successful Transient Mitigation, which of the following states the expectation for alarm announcement by this response AND when Transient Alarm Response is exited?

Transient Alarm Response Transient Alarm Alarm Announcement Response Exited A. ONLY those alarms associated with When announced by the EOP entry conditions should be Unit Supervisor announced B. ONLY those alarms associated with When all EOPs have EOP entry conditions should be been exited announced C. ONLY critical alarms should be When announced by announced the Unit Supervisor D. ONLY critical alarms should be When all EOPs have announced been exited Answer: C Answer Explanation QID: 11-1 NRO 66 Question # I 66 I Developer / Date: JJR /5-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO SRO Conduct of Operations 2.1.1 - Knowledge of conduct of operations 3.8 4.2 requirements.

Level I RO I Tier I 3 I Category I COO OCSOPS ILT Page: 210 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam General OP-OC-101-111 References 1001 C is Correct. lAW procedure OP-OC-101-111-1001, Strategies for Successful Transient Mitigation, when transient alarm response is allowed, only critical alarms and results should be announced to the US (Unit Supervisor). The US shall appraise the transient and as conditions permit, exit transient alarm response by Explanation announcing to the crew that transient alarm response is being exited.

All distractors are Incorrect but plausible in that some are related to normal alarm response or are misinterpretations of the transient alarm uideline in the ure.

Lesson Plan 2621.830.0.0017, Conduct of Operations - Admin 2.1.1 - Knowledge of conduct of operations requirements.

Learning Question Source (New, Modified, Bank) Bank If Bank or Modified VISION System/Question 10: 609318 Question Source: ILT 07-1 NRC Exam Previous 2 Exams: No Memory or X Comprehension Fundamental 1:P or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: Procedure steps and cautions 10CRF55 55.41b I 10 55.43b J

Content Administrative, normal, abnormal, and emergency operating procedures for the facility.

OCSOPS ILT Page: 211 of241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes I Point Value: 1 System 10 No.: N/A PRA:

I No Safety N/A IZJ Initial License Level Function: o LORT

EXAMINATION ANSWER KEY oc RO NRC Exam 67 ID: 11-1 NRO 67 Points: 1.00 WHICH ONE of the following activities is considered a CORE ALTERATION?

A. Removal of control rod 30-31 from the core.

B. Installation of a new LPRM string in core location 24-25.

C. Withdrawal of control rod blade 30-35 for CRD exercises.

D. Removal of an irradiated LPRM string from core location 48-41.

Answer: A Answer Explanation QID: 11-1 NRO 67 Question # I 67 I Developer / Date: JJR /5-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO SRO Conduct of Operations 2.1.36 - Knowledge of procedures and 3.0 4.1 limitations involved in core alterations.

Level I RO I Tier 3 I Category I COO General TS 1.21 References OCSOPSILT Page: 213 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam A is Correct. CORE ALTERATION shall be the movement of any fuel, sources, or reactivity control components within the reactor vessel with the vessel head removed and fuel in the vessel. The following exceptions are not considered to be CORE ALTERATIONS:

a) Movement of source range monitors, local power range monitors, intermediate range monitors, traversing incore probes, or special moveable detectors (including undervessel replacement); and b) Control rod movement, provided there are no fuel assemblies Explanation in the associated core cell.

Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.

All distractors are Incorrect but plausible does not recall the Tech Spec definition for Core Alterations and therefore does not recognize the procedural requirements and limitations associated with Core Alterations.

Lesson Plan 2621.830.0.0017, Conduct of Operations 2.1.36 - Knowledge of procedures and limitations involved in Learning core alterations.

Question Source (New, Modified, Bank) Bank If Bank Qr MQdifi~d VISION System/Question 10: 690118 Question Source: Limerick ILT Bank Previous 2 Exams: No Memory or X 1:[, Comprehension Fundamental or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: Definitions 10CRF55 Content 55.41b I 10 5S.43b I

EXAMINATION ANSWER KEY oc RO NRC Exam Administrative, normal, abnormal, and emergency operating procedures for the facility.

Justification for LORT questions with NJA KIA values < 3.0 Time to Complete: 1-2 minutes I Point Value: 1 System ID No.: NJA PRA: I No Safety NJA 181 Initial License Level Function: o LORT OCSOPS ILT Page: 215 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 68 ID: 11-1 NRO 68 Points: 1.00 Which of the following would require the use of a grounding device for a clearance lAW Procedure OP-MA-109-101, Clearance and Tagging? The work will require replacing the motor in each case.

A. SOC Pump B. ESW Pump C. Core Spray Booster Pump O. Containment Spray Pump Answer: B IAnswer Explanation QID: 11-1 NRO 68 Question # I 68 I Developer / Date: JJR /5-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO SRO Equipment Control 2.2.13 - Knowledge of tagging and clearance 4.1 4.3 procedures.

Level I RO I Tier 3 I Category I EQC General OP-MA-109-101 References OCSOPS ILT Page: 216 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam B is Correct. lAW the reference, proper safety grounding shall be applied prior to working on high voltage equipment when contact with exposed conductors is planned or possible. The reference also defines high voltage as an energy source 600 volts or above.

In the work activities listed in the question stem, all will require removal of the motor and the potential for exposed conductors Explanation exists. Of the equipment listed, only the ESW Pump is powered from a bus greater than 600 VAC (Bus 1C or 10).

All distractors are Incorrect but plausible since they are large pumps, however are all powered from 480VAC and do not require device.

Lesson Plan 2621.830.0.0018, Equipment Control Learning 2.2.13 - Knowledge of tagging and clearance procedures.

Question Source (New, Modified, Bank) Bank If Bank Qr MQdifiid VISION System/Question 10: 609321 Question Source: ILT Bank Previous 2 Exams: No Memory or X Comprehension Fundamental 3:SPK or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: Solve a Problem using Knowledge and its meaning 10CRF55 55.41b I 10 55.43b I

Content Administrative, normal, abnormal, and emergency operating procedures for the facility.

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1*2 minutes I Point Value: 1 OCSOPSILT Page: 217 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam System 10 No.: N/A PRA: I No Safety [gI Initial License Level N/A Function: D LORT

EXAMINATION ANSWER KEY oc RO NRC Exam 69 10: 11-1 NRO 69 Points: 1.00 The plant was at 80% power and shutting down, with the following abnormal switch configuration:

  • AUTO DEPRESS VALVE NR108A switch is in the OFF position
  • The NORMAL/DISABLE switch for EMRV NR108B is in the DISABLE position Which of the following states those EMRVs which can function in the Pressure Relief Mode to control RPV pressure and/or in the ADS Mode during a LOCA?

Pressure Relief Mode. ADS Function A. EMRVs B, C, D and E ONLY All EMRVs B. EMRVs A, C, D and E ONLY All EMRVs C. EMRVs C, D and E ONLY EMRVs A, C, D and E ONLY D. EMRVs C, D and E ONLY EMRVs B, C, D and E ONLY Answer: C IAnswer Explanation QID: 11-1 NRO 69 Question # I 69 J Developer I Date: JJR /5-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO SRO Equipment Control 2.2.37 - Ability to determine operability and / or 3.6 4.6 availability of safety related equipment.

Level I RO I Tier 3 I Category I EQC General 729E182 References OCSOPS ILT Page: 219 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam C is Correct. The plant is at power with an abnormal switch configuration. For an EMRV to open, its solenoid must energize.

With the front control panel switch in OFF, the affected EMRV will not function in the pressure relief mode, but will function in the ADS mode. With the interior panel switch in DISABLE, the solenoid will not energize at all: the affected EMRV will not function in the pressure relief mode or the ADS mode. Therefore, Explanation all EMRVs will function in the ADS mode, except that EMRV which is in DISABLE (NR018B). All EMRVs, except NR108A and NR108B, will function in the pressure relief mode.

All other distractors are plausible but incorrect if the student does not recall how EMRVs will function with their control switch in OFF or DISABLE.

Lesson Plan 2621.828.0.0005, Automatic Depressurization System ADS-0368, Describe the EMRV initiation logic for both Learning overpressure operation and operation in the ADS mode. Include Objective/ the following: 1. Initiation signals and setpoints 2. Timers and ATr.t'\IIr'lT~ 3. Control switches 4. Panel indications Question Source (New, Modified, Bank) Bank If Bank Qr MQdified VISION System/Question ID: 811675 Question Source: ILT Bank Previous 2 Exams: No Memory or X Comprehension Fundamental 3:SPK or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: .solve a Problem using Knowledge and its meaning 10CRF55 55.41b I 7 55.43b I

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

OCSOPSILT Page: 220 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes I Point Value: 1 System 10 No.: N/A PRA:

I No Safety N/A 181 Initial License Level Function: D LORT OCSOPS ILT Page: 221 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 70 10: 11-1 NRO 70 Points: 1.00 The plant was at rated power when the following annunciator alarmed:

  • RADIATION MON ITORS AREA - AREA MON HI The Operator reports the following:
  • ARM R010-A2, CONTROL ROOM ACCESS AREA, shows the HIGH light lit
  • ARM R010-A3, MAIN CONTROL ROOM, shows a rising trend lAW the associated RAP, which of the following is correct regarding the Control Room Ventilation System?

The Control Room Ventilation System shall be ...

A. secured.

B. placed in PURGE mode.

C. placed in FULL RECIRC mode.

D. placed in PART RECIRC mode.

Answer: D Answer Explanation QIO: 11-1 NRO 70 Question # I 70 I Developer 1 Date: JJR 15-14*2012 Knowledge and Ability Reference Information Importance Rating K&A RO SRO Radiation Control 2.3.15* Knowledge of radiation monitoring systems, such as fixed radiation monitors and 2.9 3.1 alarms, portable survey instruments, personnel monitoring equipment, etc.

Level I RO I Tier I 3 I Category I RPT OCSOPS ILT Page: 222 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam General RAP-10F1k 331.1 References o is Correct. lAW RAP-10F1k, if ARM R010-A1, -A2, or A3 is in alarm, then place the CR HVAC in PART RECRIC. Procedure 331.1 also says to place in PART RECIRC for a radiological release with offsite power available.

Explanation All distractors are Incorrect but plausible if the applicant is not familiar with the actions required for RAP-10F1k or does not recognize that plant conditions require a change in Control Room ventilation.

Lesson Plan 2621.828.0.0054, Turbine Building and Misc Ventilation Systems TMV-02324: Explain the basis, with use of procedure, for the Learning four different modes of control room ventilation damper Objective/ alignment and the effects of the damper alignment modes on control room habitabil Question Source (New, Modified, Bank) Bank If Bank or M~utifjid VISION System/Question 10: 718246 Question Source: ILT Bank Previous 2 Exams: No Memory or X Comprehension Fundamental 1:P or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: Procedure steps and cautions 10CRF55 55.41b I 12 55.43b I

Content Radiological safety principles and procedures.

OCSOPSILT Page: 223 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes I Point Value: 1 System 10 No.: N/A PRA: I No Safety [gJ Initial License Level N/A Function: D LORT OCS OPS ILT Page: 224 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 71 10: 11-1 NRO 71 Points: 1.00 Which of the following sets of indications would require entry into an EOP?

A.

OCSOPS ILT Page: 225 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam B. REACTOR LEVEL NARROW RANGE YARWAY GEMAC

,.......190

......... 110' ,;..... 110

, f'OO ,

fleo

~150 -150

~

" 1<IQ Il 1'10 l! --130 S T -130 A

120 F ~1:W T

A F --110 ~110

(;0 c.

OCSOPSILT Page: 226 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam D. CLEANUP SY$ FI)ELPOOL FILTEf;' AReA ISOLATION CONtI AREA

$HlITOOWN (Lei HXAREA P\JMPAAEA

0. .* *...

Answer: D IAnswer Explanation QID: 11-1 NRO 71 Question # I 71 J Developer I Date: JJR 15-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO SRO Radiation Control 2.3.5 - Ability to use radiation monitoring systems, such as fixed radiation monitors and 2.9 2.9 alarms, portable survey instruments, personnel monitoring equipment, etc.

Level I RO J Tier 3 I Category I RPT General SCC EOP References OCSOPS ILT Page: 227 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam o is Correct. This picture shows that CLEANUP SYS PUMP AREA is in the Alarm condition and also above the EOP entry condition of 15 mRlhr which is an entry into. This question examines the KIA by having the student interpret fixed radiation monitor indication and also determine that an EOP entry into Secondary Containment Control has been met.

A is Incorrect but plausible if the applicant does not read the indications provided correctly. This picture shows Reactor Pressure at 1035#. The entry condition into RPV Control - no ATWS is 1045#.

Explanation B is Incorrect but plausible if the applicant does not read the indications provided correctly. This picture shows all RPV water levels at 140". The EOP entry condition into RPV Control - no ATWS is 138".

C is Incorrect but plausible if the applicant does not read the indications provided correctly. This picture shows Torus water level at 153" and Torus Temperature at 85F. The EOP entry into Primary Containment Control is a Torus water temperature >95F or Torus water level >154".

Lesson Plan 2621.830.0.0015, Radiation Control - Admin 2.3.5, Ability to use radiation monitoring systems, such as fixed Learning radiation monitors and alarms, portable survey instruments, o monitorin etc.

Question Source (New, Modified, Bank) Bank If Bank o[ Modified VISION System/Question 10: 811747 Question Source: ILT Bank Previous 2 Exams: No Memory or X Comprehension Fundamental 3:SPK or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: Solve a Problem using Knowledge and its meaning OCSOPS ILT Page: 228 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 10CRF55 55.41b I 11 I 55.43b I

Content Purpose and operation of radiation monitoring systems, including alarms and survey equipment.

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes I Point Value: 1 System 10 No.: N/A PRA:

I No Safety ~ Initial License Level N/A Function: o LORT oes OPS ILT Page: 229 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 72 10: 11-1 NRO 72 Points: 1.00 The plant was shutdown and was cooling down with the Shutdown Cooling System.

Present plant conditions are as follows:

  • RPV water level indicates 160" and steady
  • E RECIRC PUMP SUCTION TEMP indicates 300 OF and lowering slowly
  • Shutdown Cooling Loops A and B are in service The Operator then reported that Drywell pressure was rising slowly and that UNIDENTIFIED DRYWELL LEAKAGE rose and steadied out at 8 GPM.

10 minutes later, and with UNIDENTIFIED DRYWELL LEAKAGE UNCHANGED, the following annunciators alarmed:

  • DW PRESS HI-HI I
  • DW PRESS HI-HIli Which of the following actions will have the GREATEST affect on PREVENTING steam line flooding?

A. Close all LFRVs B. Trip all Condensate Pumps C. Place Core Spray Parallel Isolation Valves in CLOSE D. Override Core Spray Signals and place Core Spray Pumps in STOP Answer: D IAnswer Explanation QID: 11-1 NRO 72 Question # I 72 I Developer / Date: JJR /5-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO I SRO OCSOPS ILT Page: 230 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam Emergency Procedures I Plan 2.4.9

  • Knowledge of low power I shutdown implications in accident (e.g., loss of coolant 3.8 4.2 accident or loss of residual heat removal) ation ies.

RO Tier 3 Category EOP General 201 EMG-SP10 References D is Correct. The plant is shutdown and cooling down with SDC.

At a coolant temperature of 300 OF, this equates to an RPV pressure of 53 psig. At this reactor power level, feedwater flow is minimal (one condensate pump running through a LFRV).

A leak then occurs in the drywell and steadies out at 8 gpm, when the drywell pressure scram and isolation setpoint is reached (as provided by the given annunciators). This same parameter results in the start of core spray, and since RPV pressure is < 305 pSig, the core spray parallel isolation valves will open immediately. At this pressure, each loop of core spray will inject several thousand GPM to the RPV.

To stop core spray with an initiation signal present, the signals Explanation must first be overridden, then the core spray pumps can be stopped. Since core spray is adding water to the RPV at the largest rate, as compared to condensate, then securing core spray has the largest impact on controlling RPV water level.

A & B are Incorrect but plausible since closing the only in-service LFRV and stopping the only running condensate pump are very similar in their impacts, but their flow is much less than core spray.

C is Incorrect but plausible. Because there is a core spray initiation signal present, placing the core spray parallel isolation valve switches to close will not result in the valves going and closed.

OCSOPSILT Page: 231 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam Lesson Plan 2621.828.0.0010, Core Spray System CSS-10446, Identify and explain system operating controls and Learning indications under all plant operating conditions.

Objectivel Question Source (New, Modified, Bank) Bank If B~nk gr Mgdifjed VISION System/Question 10: 663328 Question Source: ILT 08-1 NRC Exam Previous 2 Exams: No Memory or X Comprehension Fundamental 3:SPK or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: .solve a f.roblem using Knowledge and its meaning 10CRF55 55.41b I 10 55.43b J

Content Administrative, normal, abnormal, and emergency operating procedures for the facility.

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes Point Value: 1 System 10 No.: N/A PRA:

I No Safety [8J Initial License Level N/A Function: o LORT OCSOPS ILT Page: 232 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 73 10: 11-1 NRO 73 Points: 1.00 The plant was at rated power when a spontaneous fire erupted in the Reactor Building.

Which of the following is the LOWEST Emergency Classification Level that, if declared, requires activation of BOTH the Technical Support Center (TSC) and Emergency Operations Facility (EOF) from this event?

A. Unusual Event B. Alert C. Site Area Emergency D. General Emergency Answer: B IAnswer Explanation QID: 11-1 NR073 Question # I 73 I Developer / Date: JJR /5-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO SRO Emergency Procedures / Plan 2.4.42 - Knowledge of emergency response 2.6 3.8 facilities.

Level I RO I Tier 3 I Category I EOP General EP-AA-1010 References OCSOPS ILT Page: 233 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam B is Correct. The lowest Emergency Plan classification that will result in requiring activation of both the TSC and EOF from a spontaneous fire is an Alert. This is specified as spontaneous so the applicant does not think it was due to a security issue (the Explanation TSC and EOF are activated at the Unusual Event level for security threats).

All distractors are Incorrect but plausible if the applicant cannot recall when both the TSC and EOF are ired to be activated.

Lesson Plan 2621.830.0.0016, Emergency Procedures / Plan - Admin 2.4.42 - Knowledge of emergency response facilities.

Learning ective/

Question Source (New, Modified, Bank) New If Bank or MQdifi~d VISION System/Question 10:

Question Source: N/A Previous 2 Exams:

Memory or X Comprehension Fundamental 1:IP or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: frocedure steps and cautions 10CRF55 55.41b I 10 55.43b I

Content Administrative, normal, abnormal, and emergency operating procedures for the facility.

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes Point Value: 1 System 10 No.: I N/A I PRA: I No OCSOPS ILT Page: 234 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam Safety t8I Initial License Level Function:

N/A o LORT OCSOPS ILT Page: 235 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 74 10: 11-1 NRO 74 Points: 1.00 The plant is at rated power. An event then occurred and plant conditions include the following:

  • Drywell Pressure indicates 1.4 psig and rising slowly The BOP is executing RAP-C3f, OW PRESS HI/LO. lAW station procedures, which of the following is the correct way to communicate Drywell pressure to the Crew?

The BOP will raise one hand announce the following ...

A. "Attention for a brief, Drywell Pressure is 1.4 psig and rising slowly, end of brief."

B. "Attention for an update, Drywell Pressure is 1.4 psig and rising slowly, end of update."

C. "Attention for a brief, Drywell Pressure is 1.4 psig and increasing slowly, end of brief."

D. "Attention for an update, Drywell Pressure is 1.4 psig and increasing slowly, end of update."

Answer: B Answer Explanation QID: 11-1 NRO 74 Question # I 74 I Developer / Date: J ..IR /5-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO SRO Conduct of Operations 2.1.38 - Knowledge of the station's 3.7 3.8 requirements for verbal communications when implementing procedures.

Level I RO I Tier I 3 I Category I COO oes OPS IlT Page: 236 of 241 02 April 2012

EXAMINATION ANSWER KEY DC RD NRC Exam General OP..QC*101*111 HU-AA-101 References 1001 B is Correct. There are two types of formally communicate plant status/parameters to the crew. One is a crew 'update'; the second is a crew 'brief. An update is used to communicate a specific parameter or parameters to the crew, without any comments from crew members and a brief is used to communicate parameters/plant evolutions where crew feedback is solicited. lAW OP*OC-101*1001 and HU-AA-101, when reporting a plant parameter, the correct communication must be to report the VALUE (with units), direction of TREND (up, down, Explanation rising, steady, etc.), and RATE (slow, fast, slowly, etc.) oftrend.

It is not acceptable to use trends such as increase or decrease.

The BOP will call for an update by raising one hand and saying "Attention for an update, (report parameter, value, and trend),

end of update."

All distractors are Incorrect but plausible if the applicant does not recall the stations communications requirements of OP-OC 101*1001 and HU*AA-101.

Lesson Plan 2621.830.0.0017, Conduct of Operations* Admin Learning 2.1.38 - Knowledge of the station's requirements for verbal communications when im ures.

Question Source (New, Modified, Bank) New If Bank Qr Mo~Ufi~d VISION System/Question ID:

Question Source: N/A Previous 2 Exams:

Memory or X Comprehension Fundamental 1:F' or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: Procedure steps and cautions OCSOPS ILT Page: 237 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam 10CRF55 55.41b I 10 I 55.43b I

Content Administrative, normal, abnormal, and emergency operating procedures for the facility.

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes I Point Value: 1 System 10 No.: N/A PRA:

I No Safety [81 Initial License Level N/A Function: D LORT

EXAMINATION ANSWER KEY oc RO NRC Exam 75 10: 11-1 NRO 75 Points: 1.00 Which of the following activities could result in INTERNAL radioactive contamination to workers in the Reactor Building?

A. Initiating the Isolation Condensers for an RPV cooldown.

B. Venting the Torus through the hardened vent post-LOCA.

C. Opening the Individual Scram Test Switches during an ATWS.

D. De-energizing the Scram Solenoids during an ATWS with the Scram buttons.

Answer: C IAnswer Explanation QID: 11-1 NRO 75 Question # I 75 I Developer I Date: JJR /5-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO SRO Radiation Control 2.3.14 - Knowledge of radiation or containment hazards that may arise during normal, 3.4 3.8 abnormal, or emergency conditions or activities.

Level I RO I Tier 3 I Category I EQC General EMG-SP21 EOP User's Guide References OCSOPS ILT Page: 239 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam C is Correct. lAW the reference, when opening the scram test switches, potentially radioactive steam may be released and RB airborne concentration levels may increase. Increased airborne radioactivity could lead to internal radioactive contamination.

Explanation All distractors are Incorrect but plausible if the applicant does not recall plant evolutions which can result in airborne radioactivity. The distractors may lead to increased external dose in the RB, or elevated internal contamination outside the RB.

Lesson Plan 2621.845.0.0053, RPV Control - with ATWS EWA-03056, Given a copy of RPV Control, describe in detail Learning each Caution or Note, including the technical basis and how to

" ........., conformance at time.

Question Source (New, Modified, Bank) Bank If Bank Qr MQdified VISION System/Question 10: 718247 Question Source: ILT Bank Previous 2 Exams: No Memory or X Comprehension Fundamental 1:1 or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: Interlocks, setpoints, or system (singular) response 10CRF55 55.41b I 12 55.43b I

Content Radiological safety principles and procedures.

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes Point Value: 1 System 10 No.: I N/A J PRA:

I No OCSOPS ILT Page: 240 of 241 02 April 2012

EXAMINATION ANSWER KEY oc RO NRC Exam Safety r8l Initial License Level Function:

N/A o LORT OCSOPSILT Page: 241 of 241 02 April 2012

EXAMINATION ANSWER KEY OC SRO NRC Exam 1 10: 11-1 NSRO 01 Points: 1.00 The plant is shutdown for a refuel outage, with fuel moves in progress. An event occurred as shown by the following time-line.

At 1020; The following annunciators have alarmed:

  • AREAMON HI
  • CRIT MON C5 HI
  • NORTH WALL C10 HI
  • NORTH WALL C9 HI VENT TRIP
  • OPER FLOOR B9 HI VENT TRIP At 1021; An Operator observes the following:
  • All refuel floor radiation monitors have the HIGH light lit and that Radiation Monitor B9, REACTOR OPEN FLR EQUIP HATCH, indicates upscale.
  • Main Stack RAGEMS indicates 4.0 E+OO cps At 1023; The refuel floor SRO notifies the control room that the loaded hoist became separated from the bridge and has fallen into the core area.

What is the correct emergency plan classification, if any, for these conditions?

A. None B. Unusual Event C. Alert D. Site Area Emergency Answer: C Answer Explanation QID: 11-1 NSRO 01 Question #1 is I Developer / Date: JJR /5-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO I SRO OCSOPS ILT Page: 1 of 83 02 April 2012

EXAMINATION ANSWER KEY OC SRO NRC Exam 295023 Refueling Acc Cooling Mode / 8 AA2.05 - Ability to determine and/or interpret the following as they apply to REFUELING 3.2 4.6 ACCIDENTS: Entry conditions of emergency SRO Tier 1 Group 1 General EP-AA-1010 rev. 3 References C is Correct. For the given conditions, the applicant must declare an Alert, RA2, on the Cold Matrix, due to a valid reading of >1000 mRlhr or upscale reading on one or more of the radiation monitors.

Explanation All distractors are Incorrect but plausible if the applicant does not recognize the correct EAL classification. Site Area Emergency is plausible if the applicant does not correctly interpret the indications provided. An SAE would be correct if Main Stack RAGEMS were at 4.0 E+OO IJCi/cc HRM, not 4.0 E+OO hich are the units for the Low W'lS,'U'16 I\AI'\,n.~nrl Lesson Plan 2621.812.0.0003, Refueling RFL-326, State actions required in the event of unexplained Learning criticality or high SR.M counts.

o ective/

Question Source (New, Modified, Bank)

If Bank or MQdified I Modified VISION System/Question 10: 608369 Question Source: ILT Bank Previous 2 Exams: No OCS OPS ILT Page: 20f83 02 April 2012

EXAMINATION ANSWER KEY OC SRO NRC Exam Memory or X Comprehension Fundamental 3:SPR or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: Solve a Problem using References 10CRF55 55.41b I 55.43b I 5 Content Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes I Point Value: 1 System 10 No.: 295023 PRA:

I No Safety ~ Initial License Level 8

Function: o LORT OCSOPSILT Page: 30f83 02 April 2012

EXAMINATION ANSWER KEY OC SRO NRC Exam 2 10: 11-1 NSRO 02 Points: 1.00 The plant was at rated power when an event occurred. Present plant conditions include the following:

  • RPV water level indicates 35" and lowering
  • RPV pressure indicates 465 psig and lowering slowly
  • Drywell pressure indicates 18 psig and is being controlled with Drywell Sprays
  • Both Startup Transformers have failed to energize their respective busses
  • One alternate subsystem is lined-up to Core Spray 1, and a second alternate subsystem is lined up to Core Spray 2 Which of the following shall the SRO direct NEXT?

A. When RPV water level lowers to 0", direct entry into the Steam Cooling EOP.

B. When RPV water level lowers to -20", direct entry into the SAMG for Primary Containment Flooding.

C. Rapidly depressurize the RPV using the turbine bypass valves to allow Fire Water to inject into the RPV.

D. When RPV water level lowers to 0", direct entry into the Emergency Depressurization - No ATWS EOP.

Answer: D Answer Explanation QID: 11-1 NSRO 02 Question # I 2S I Developer 1 Date: JJR 15-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO SRO 295031 Reactor Low Water Level 12 EA2.04 - Ability to determine andlor interpret 4.6 4.8 the following as they apply to REACTOR LOW WATER LEVEL: Adequate core cooling OCSOPS ILT Page: 4 of 83 02 April 2012

EXAMINATION ANSWER KEY OC SRO NRC Exam Level SRO Tier 1 Group 1 General RPVC - no A TWS References EOP o is Correct. The plant was at rated power when an event occurred. The present plant conditions show that RPV water level is 35" and lowering, offsite power has been lost, and no core spray pumps are available for RPV injection. Firewater to Core Spray, a RPV alternate subsystem injection source, is lined-up.

With firewater lined-up to core spray, the guiding support procedure directs opening the core spray parallel isolation valves to allow flow, when RPV pressure drops below 310 psig.

Since RPV pressure is above this pressure, and RPV water level is currently lowering, it will continue to lower until the isolation valves are opened.

lAW the RPV Control - no ATWS EOP, under these circumstances (with no core spray and no condensate, and any injection source lined-up and running), when RPV water leyel lowers to 0", then emergency depressurization is required to assure adequate core Explanation COOling.

A is Incorrect but plausible. Entry into the Steam Cooling EOP would be required when RPV water level reaches 0" if no RPV injection sources are lined-up and running.

B is Incorrect but plausible. Answer C could be a correct direction, but a decision prior to lowering to -20" must be made first. Thus, this would not be the NEXT direction as asked for in the question.

C is Incorrect but plausible. If core spray or condensate were available for RPV injection, then directing an RPV pressure reduction to allow these low pressure systems to inject would be correct. But with no condensate or core spray, the direction is incorrect.

OCS OPS ILT Page: 50f83 02 April 2012

EXAMINATION ANSWER KEY OC SRO NRC Exam Lesson Plan 2621.845.0.0052, RPV Control - no ATWS ENA-10445, Given a set of system indications or data, evaluate Learning and interpret them to determine limits, trends and system Objective/ status.

Question Source (New, Modified, Bank) New If Bank Q[ MQdified VISION System/Question 10:

Question Source: N/A Previous 2 Exams:

Memory or X Comprehension Fundamental 3:SPK or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: .solve a Problem using Knowledge and its meaning 10CRF55 55.41b I 55.43b I 5 Content Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes I Point Value: 1 System 10 No.: 295006 PRA: I No Safety 1

IZI Initial License Level Function: D LORT OCSOPS ILT Page: 6of83 02 April 2012

EXAMINATION ANSWER KEY OC SRO NRC Exam 3 10: 11*1 NSRO 03 Points: 1.00 The plant was shutting down for an outage, with the following conditions:

  • RPV coolant temperature is 320 OF and lowering
  • Shutdown Cooling pumps A and B are in service with a combined total SOC system flow of 4000 GPM
  • RBCCW flow through each of the in-service SOC heat exchangers is 1000 GPM
  • RPV water level is being maintained at 155" TAF The following annunciator then came into alarm:
  • SHUT ON CLG - PUMP B TRIP The new plant conditions are as follows:
  • Investigation shows that the SOC pump B tripped on over-current Which of the following lists the required action to raise RPV cooling?

A. Initiate an alternate RPV cooldown lAW ABN-3, Loss of Shutdown Cooling.

B. Raise RPV water level to above 170" TAF lAW procedure 305, Shutdown Cooling System Operation.

C. Maximize SOC loop A flow to no more than 3400 GPM lAW procedure 305, Shutdown Cooling System Operation.

O. Maximize RBCCW cooling water flow through the SOC loop A heat exchanger to no more than 2000 GPM lAW procedure 309.2, Reactor Building Closed Cooling Water System.

Answer: C Answer Explanation QID: 11-1 NSRO 03 Question # I 3S I Developer 1 Date: J ..tR 15*14-2012 Knowledge and Ability Reference Information OCSOPS ILT Page: 70f83 02 April 2012

EXAMINATION ANSWER KEY OC SRO NRC Exam Importance Rating K&A 295021 Loss of Shutdown Cooling I 4 AA2.02 - Ability to determine and/or interpret the following as they apply to LOSS OF 3.4 3.4 SHUTDOWN COOLING: RHRlshutdown coolin flow Level SRO Tier 1 Group 1 General 305 References Initial conditions indicate a partial loss of shutdown cooling flow.

The RAP for the tripped SOC pump directs another pump be started if possible (which is not possible).

C is Correct. Procedure 305 explains how to place SOC in service. The SOC pump discharge valves are throttled to establish the desired cooldown rate. The same procedure sets a limit on SOC flow of 3400 GPM through a heat exchanger. Since current SOC flow is only 2000 GPM, then SOC flow can be increased up to 3400 GPM.

Explanation A is Incorrect but plausible. Initiating alternate cooling through the cleanup system is only performed when all SOC flow is lost or cannot be established, lAW ABN-3.

B is Incorrect but plausible. lAW procedure 305, with reactor recirc pumps running, RPV water level should be maintained within the normal band. Raising water level is only required during a partial SOC flow loss when no reactor recirc pumps are running.

o is Incorrect. In both procedures 309.2 and 305, it stipulates that RBCCW cannot exceed 1500 GPM through a SOC heat OCSOPS ILT Page: 80183 02 April 2012

EXAMINATION ANSWER KEY OC SRO NRC Exam Lesson Plan 2621.828.0.0045, Shutdown Cooling System SOC-10450, Describe and interpret procedure sections and Learning steps for plant emergency or off-normal conditions that involve Objective/ this system including personnel allocation and equipment operation lAW applicable ABN, EOP & EOP support procedures and EP procedures.

Question Source (New, Modified, Bank) Bank If Bank or Modified VISION System/Question 10: 510950 Question Source: ILT 05-1 SRO NRC Exam Previous 2 Exams: No Memory or X Comprehension Fundamental 3:SPK or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: .solve a f!.roblem using Knowledge and its meaning 10CRF55 55.41b I 55.43b I 5 Content Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes Point Value: 1 System 10 No.: 295021 PRA: I No Safety t8j Initial License Level 4

Function: D LORT OCSOPS ILT Page: 90f83 02 April 2012

EXAMINATION ANSWER KEY OC SRO NRC Exam 4 10: 11*1 NSRO 04 Points: 1.00 IF Torus water temperature exceeds 95"F while at power, what action is required per Tech Specs AND what is the Tech Spec basis for the 95°F limit? (Assume NO EMRV testing is in progress)

TS Action TS Bases A. Be in COLD SHUTDOWN Ensure that the maximum peak within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Torus temperatu re does not exceed 110°F if an ED was performed B. Be in COLD SHUTDOWN Ensure that the maximum peak within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Torus temperature does not exceed 160°F if an ED was performed C. Be in COLD SHUTDOWN Ensure that the maximum peak within 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> Torus temperature does not exceed 160°F if an ED was performed D. Be in COLD SHUTDOWN Ensure that the maximum peak within 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> Torus temperature does not exceed 11 oaF if an ED was performed Answer: B Answer Explanation QID: 11-1 NSRO 04 Question # I 4S I Developer I Date: JJR 15-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO SRO 295026 Suppression Pool High Water Temp.! 5 2.2.25 - Equipment Control: Knowledge of 3.2 4.2 bases in technical specifications for limiting conditions for operations and safety limits.

OCSOPS ILT Page: 10 of 83 02 April 2012

EXAMINATION ANSWER KEY OC SRO NRC Exam Level SRO Tier 1 Group 1 General TS 3.5.A.1 TS 3.5 Bases References Answer B is Correct. lAW TS 3.5.A.1, the Maximum Torus water temperature is 95F at power. TS 3.5.A.1.d states that if this limit is exceeded, be in the COLD SHUTDOWN condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The basis for this action is to avoid excessive Torus loading following a depressurization using EMRVs. This is accomplished by ensuring Torus temperature does not exceed 160F following any period of EMRV operation.

TS 3.5 Bases state the following in regards to maximum Torus temperature: Experimental data indicate that excessive steam condensing loads can be avoided if the peak temperature of the suppression pool is maintained below 160F during any period of relief valve operation with sonic conditions at the discharge exit.

Explanation Specifications have been placed on the envelope of reactor operating conditions so that the reactor can be depressurized in a timely manner to avoid the regime of potentially high suppression chamber loadings.

All distractors are Incorrect but plausible if the student does not remember the exact value of the TS limit or TS bases. The normal shutdown LCO action statement to be in Cold Shutdown if one is not given is 30 hrs. The Torus temp tech specs gives a value of 24 hrs. The candidate may not recall this making it plausible. The value of 11 OF is plausible if the student confuses this with the maximum temperature allowed where a reactor red.

Lesson Plan 2621.845.0.0056, Primary Containment PCC-422, Given tech specs, identify and explain each LCO for Learning the Primary Containment System under described conditions.

o nU::I,,.TI'VA I Question Source (New, Modified, Bank) Modified OCS OPS ILT Page: 11 of 83 02 April 2012

EXAMINATION ANSWER KEY OC SRO NRC Exam If Bank Q[ MQdified VISION System/Question 10: 811770 Question Source: ILT Bank Previous 2 Exams: No Memory or X Comprehension Fundamental 1:B or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: Bases or Purpose 10CRF55 55.41b I 55.43b I 2 Content Facility operating limitations in the technical specifications and their bases.

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes I Point Value: 1 System 10 No.: 295026 PRA:

I No Safety 5

I8l Initial License Level Function: D LORT OCSOPS ILT Page: 12 of 83 02 April 2012

EXAMINATION ANSWER KEY OC SRO NRC Exam 5 10: 11-1 N5RO 05 Points: 1.00 The plant is in cold shutdown and is cooling down with the Shutdown Cooling System (SOC). The following conditions currently exist:

  • RECIRC PUMP SUCTION TEMPS indicates 209 OF
  • RPV water level is 175" and steady An event then occurs as shown in the timeline below:

0800 Annunciator RBCCW - SURGE TANK LVL HI/LO alarms 0804 The EO reports the RBCCW Surge Tank indicates 1" and lowering and the Tank makeup valve is full open 0806 The Radwaste Operator reports RB Floor Drain Sump 1-7 high level is in alarm 0808 Maintenance reports that they are unable to repair the leak 0809 The SM observes Drywell pressure at 1.7 psig and steady and Drywell temperature at 140 OF and steady 0810 The SM starts the 1-hour clock to monitor entry into EAL MA5( 1)

Which of the following shall the SRO direct NEXT?

A. Trip all Recirculation Pumps lAW ABN-19, RBCCW Failure Response.

B. Operate all available Drywell Coolers, lAW SP-27, Maximizing Drywell Cooling.

C. Isolate the Reactor Water Cleanup System lAW the Secondary Containment Control EOP.

D. Initiate Isolation Condensers by placing the Condensate Return DC valves to OPEN, lAW 307, Isolation Condenser System.

Answer: A Answer Explanation QID: 11-1 N5RO 05 Question # I 55 I Developer / Date: JJR /5-14-2012 Knowledge and Ability Reference Information K&A I Importance Rating OCSOPS ILT Page: 13 of 83 02 April 2012

EXAMINATION ANSWER KEY OC SRO NRC Exam 295018 Partial or Total loss of CCW /8 2.1.23 - Conduct of Operations: Ability to perform specific system and integrated plant 4.3 4.4 procedures during all modes of plant level SRO Tier 1 Group 1 General EP-AA-1010 References A is Correct. The plant is < 212 OF and cooling down with SOC with all 3 SOC pumps in service. Then, indications are provided which show an unisolable leak in RBCCW (lowering surge tank level and high level in the floor drain tank, and not corrected quickly). Some time later, the SM starts the 1-hour clock for entry into emergency EAl MA5(1). This EAl entry shows that RPV temperature is> 212 OF for at least 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. Therefore, RPV temperature has reached 212 OF and is rising due to the lack of RBCCW cooling to SOC. With the RBCCW leak and RPV water temperature> 212 OF, the RBCCW ABN directs that all recirculation pumps be tripped.

Answer B is Incorrect but plausible. Operation of the Orywell coolers lAW SP-27 is directed from the Primary Containment Control EOP. Conditions show parameters less than the entry Explanation conditions (OW temperature & pressure) for the EOP. Thus, the SP cannot be used.

Answer C is Incorrect but plausible. With a loss of RBCCW, it is suggested that RWCU be removed from service. The RB floor drain sump 1-7 is an entry into the Secondary Containment Control EOP. In the Secondary Containment Control EOP, it directs isolation of leaking systems, which in this case, is RBCCW - not RWCU. Thus isolation/removal of RWCU is directed from the loss of RBCCW ABN and not the EOP.

Answer 0 is Incorrect but plausible. Since the RPV has lost its cooling medium and is heating up, Isolation Condensers can be used now that RPV temperature is > 212 OF. But with RPV water level> 160" initiation the normal is not allowed.

EXAMINATION ANSWER KEY OC SRO NRC Exam Lesson Plan 2621.828.0.0035, Reactor Building Closed Cooling Water System Learning RBC-0061, Using the procedure, identify and explain normal Objective/ and emergency operations of the RBCCW System.

Question Source (New, Modified, Bank) Bank If Bank or Modified VISION System/Question 10: 6636'17 Question Source: ILT 08-1 SRO NRC Exam Previous 2 Exams: No Memory or X Comprehension Fundamental 2:RI or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: Recognizing Interaction between systems (plural), including consequences and implications 10CRF55 55.41b I 55.43b I 5 Content Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes I Point Value: 1 System 10 No.: 295018 PRA: I No Safety 8

I:8J Initial License Level Function: LORT OCSOPS ILT Page: 15 of 83 02 April 2012

EXAMINATION ANSWER KEY OC SRO NRC Exam 6 10: 11*1 N5RO 06 Points: 1.00 The plant was at rated power when an event occurred. Plant parameters include the following:

  • ESW Pump 1-2 has lost breaker indications
  • The RWCU System has automatically isolated
  • A control rod block exists from SDV high level Complete the following statement which describes the required action.

The reactor shall be placed in the COLD SHUTDOWN CONDITION within ill hours due to the loss of 125 VDC Distribution Center .(2)..

ill A. 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> DC-B B. 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> DC-C C. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> DC-B D. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> DC-C Answer: B Answer Explanation QID: 11-1 N5RO 06 Question # I 65 I Developer / Date: JJR /5-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO SRO 295004 Partial or Total Loss of DC Pwr I 6 2.2.38 - Equipment Control: Knowledge of 3.6 4.5 conditions and limitations in the facility license.

Level I SRO I Tier I 1 I Group I 1

EXAMINATION ANSWER KEY OC SRO NRC Exam General TS 3.7 ABN-55 References B is Correct. The indications provided are caused by a loss of 125 VDC distribution center DC-C (see ABN-55, DC Bus C And Panel/MCC Failures). TS 3.7.A.A.1.e requires DC distribution centers DC-B and DC-C to make the reactor critical. TS 3.7.B provides the following: The reactor shall be PLACED IN the COLD SHUTDOWN CONDITION if the availability of power falls below that required by Specification A above, except that... (3 conditions which do not apply). OP-OC-100, Oyster Creek Conduct of Operations, states that for the for the statement requiring the plant be placed in the COLD SHUTDOWN CONDITION, with no time interval provided, that the plant be Explanation placed in cold shutdown in 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. Only answer B provides the correct DC Bus loss and the correct TS requirement.

lAW section 2.C.(2) of the Facility License, Technical Specifications are part of the Facility License at Oyster Creek.

References are not provided for this question raising the difficulty level of the question.

All distractors are Incorrect but plausible if the applicant does not recall the correct LCO or recognize which DC power supply was lost.

Lesson Plan 2621.828.0.0012, DC Distribution DCD*10451 Given Technical Specifications, identify and explain Learning associated actions for each section of the Technical Objective/ Specifications relating to this system, including personnel allocation and Question Source (New, Modified, Bank) I Bank If Bank gr Modified VISION System/Question 10: 608160 Question Source: ILT Bank Previous 2 Exams: No oes OPS ILT Page: 17 of 83 02 April 2012

EXAMINATION ANSWER KEY OC SRO NRC Exam Memory or X Comprehension Fundamental 2:RI or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: Recognizing interaction between systems (plural), including consequences and implications 10CRF55 55.41b I 55.43b I 1 Content Conditions and limitations in the facility license.

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes Point Value: 1 System 10 No.: 295004 PRA: I No Safety [&/ Initial License Level 6

Function: [] LORT OCSOPSILT Page: 18 of 83 02 April 2012

EXAMINATION ANSWER KEY OC SRO NRC Exam 7 10: 11..1 NSRO 07 Points: 1.00 The plant is at rated power when a 'fire started inside a back panel (confirmed fire at time 1100). The Shift Manager declares that a Control Room Evacuation was required.

The following actions were performed prior to evacuation:

  • The reactor was scrammed (all rods in)
  • The Recirculation Pumps were tripped
  • The Feedwater Pumps were tripped The following conditions have occurred:
  • The Control Room evacuation was initiated at 1105 and completed at 1110
  • Control of the Local Shutdown Panels and the Remote Shutdown Panel was delayed until 1135
  • The Shift Manager declared at this time (1135) that the Operators have control of the plant
  • The Fire Brigade reports that the Control Room fire is extinguished at 1140 and several panels are heavily damaged What is the HIGHEST emergency plan classification required for this event?

A. HU6 B. HA4 C. HA6 D. HS4 Answer: 0 Answer Explanation QID: 11-1 NSRO 07 Question # I 7S I Developer I Date: JJR 15*14*2012 Knowledge and Ability Reference Information Importance Rating K&A RO SRO 600000 Plant Fire On-site I 8 3.1 4.4 2.4.29 - Knowledge of the emergency plan.

EXAMINATION ANSWER KEY OC SRO NRC Exam Level SRO Tier 1 Group 1 General EP-AA-1010 ABN-30 References o is Correct. Plant conditions provided show a fire in the Control Room requiring an evacuation and implementation of the Emergency Plan. lAW EP-AA-1010, a site area emergency is required if a control room evacuation is initiated and control of the plant cannot be established in < 15 minutes per ABN-30, Explanation Control Room Evacuation (HS4). The highest classification due to the fire is an alert (HA6).

All distractors are Incorrect but plausible if the applicant does not interpret the correct EAL based on plant conditions provided.

Lesson Plan 2621.828.0.0019, Fire Protection FPS-10450, Describe and interpret procedure sections and Learning steps for plant emergency or off-normal conditions that involve Objective/ this system including personnel allocation and equipment operation lAW applicable ABN, EOP & EOP support procedures and EP ures.

Question Source (New, Modified, Bank) Bank If Bank or Modified VISION System/Question 10: 609239 Question Source: ILT Bank Previous 2 Exams: No Memory or X Comprehension Fundamental 3:SPR or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: .solve a Problem using References OCSOPSILT Page: 20 of 83 02 April 2012

EXAMINATION ANSWER KEY OC SRO NRC Exam 10CRF55 55.41b I I 55.43b I 5 Content Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Justification for LORT questions with N/A KiA values < 3.0 Time to Complete: 1-2 minutes I Point Value: 1 System 10 No.: 600000 PRA:

I No Safety 8

t8I Initial License Level Function: o LORT OCSOPS ILT Page: 21 of 83 02 April 2012

EXAMINATION ANSWER KEY OC SRO NRC Exam 8 10: 11-1 NSRO 08 Points: 1.00 The plant was at rated power when a TOTAL loss of Feed and Condensate occurred.

Plant conditions include the following:

  • BOTH Isolation Condensers are unavailable
  • The B CRD Pump is unavailable
  • The A CRD Pump is in operation lAW SP-3, CRD System Operation
  • Drywell Pressure is 2.2 psig
  • Reactor Pressure is 340 psig
  • Torus Temperature is 108 OF
  • Plant Cooldown is in progress utilizing SP-14, ALTERNATE PRESSURE CONTROL SYSTEMS CLEAN-UP IN LETDOWN MODE
  • Letdown f10wrate is 100 GPM
  • RPV water level is 95" and lowering Which ONE of the following shall the SRO direct NEXT?

A. Bypass ADS Timers lAW RPV Control - no ATWS prior to reaching 91" RPV water level.

B. Place the Shutdown Cooling System in service lAW 305, Shutdown Cooling System Operation.

C. Initiate boron injection lAW SP-22, Initiating The Liquid Poison System, prior to Torus Temperature reaching 110°F D. Secure from the plant Cooldown lAW SP-14, Alternate Pressure Control Systems, Clean-Up In Letdown Mode, prior to reaching 91" RPV water level.

Answer: D Answer Explanation QID: 11-1 NSRO 08 Question # I 8S I Developer / Date: JJR /5-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO I SRO OCSOPS ILT Page: 22 of 83 02 April 2012

EXAMINATION ANSWER KEY OC SRO NRC Exam 295009 Low Reactor Water Levell 2 AA2.03 - Ability to determine andlor interpret the following as they apply to LOW REACTOR 2.9 2.9 WATER LEVEL: Reactor water cleanup blowdown rate Level I SRO I Tier 1 I Group I 2 General RPVC - no A TWS EOP User's Guide EMG-SP14 References EOP D is Correct. The question stem describes a condition where a plant cooldown has commenced via the RWCU (Reactor Water Cleanup) letdown system (letdown is the terminology used at Oyster Creek for RWCU blowdown). The RWCU system is required to be in operation to utilize the letdown function. At 91" (instrument setpoint) the RWCU system will automatically isolate on RPV Lo-Lo setpoint. The correct decision the SRO shall direct next would be to secure letdown.

A is Incorrect but plausible. If RPV water level continued to lower the SRO would direct actions required by RPV Level Restoration.

The first action required is to Bypass the ADS Timers, however this is only if level cannot be maintained >61", not 91". With the

'A' CRD pump still available, the next action would be for the SRO to secure RWCU letdown and re-evaluate whether the one CRD pump is enough to make up to RPV inventory losses before Explanation entering level restoration.

B is Incorrect but plausible if the applicant does not recognize the plant does not meet conditions for satisfying the Shutdown Cooling interlocks. RPV temperature is required to be below 350F, which corresponds to an RPV pressure of approximately 120 psig.

C is Incorrect but plausible if the applicant doesn't recall the correct procedure for initiating Liquid Poison for RPV inventory.

The applicant could direct initiating Liquid Poison lAW SP-7, but SP-22 is used for injecting Liquid Poison for maintaining Torus Temperature below the BIIT curve. In addition, to utilize SP-14, boron injection is required to be secured so securing RWCU letdown would be performed before injecting Liquid Poison.

OCSOPS ILT Page: 23 of 83 02 April 2012

EXAMINATION ANSWER KEY DC SRD NRC Exam Lesson Plan 2621.845.0.0052, RPV Control - with ATWS ENA-2257, Given the EOP, describe in detail each Learning step/statement, including the technical basis, and how to verify or each Question Source (New, Modified, Bank) New If Bank or Modifi~d VISION System/Question 10:

Question Source: N/A Previous 2 Exams:

Memory or X Comprehension Fundamental 3:SPK or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: Solve a Problem using Knowledge and its meaning 10CRF55 55.41b I 55.43b I 5 Content Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes I Point Value: 1 System 10 No.: 295009 PRA: I No Safety ~l Initial License Level 2

Function: o LORT OCSOPSILT Page: 24 of 83 02 April 2012

EXAMINATION ANSWER KEY OC SRO NRC Exam 9 10: 11 ..1 NSRO 09 Points: 1.00 The plant is at rated power when indications and local reports confirm an earthquake has just occurred.

5 minutes later, the following events occur:

  • The Operator reports that TORUS LEVEL WIDE RANGE indicates 139" and lowering
  • The EO calls the Control Room and reports 1:looding in the NW RB corner room from a Core Spray Pump line break, and that water level in the room is above the MAX SAFE value At the direction of the US, the Operator closes the Core Spray Pumps AlC suction valves and Torus water level stabilizes at 135".

Which ONE of the following is the SRO REQUIRED to direct NEXT?

A. Manually scram the reactor, lAW ABN-38. Station Seismic Event.

B. Initiate a manual reactor shutdown. lAW the Primary Containment Control EOP.

C. Emergency Depressurize the RPV, lAW the Secondary Containment Control EOP.

D. Anticipate Emergency Depressurization, lAW the Primary Containment Control EOP.

Answer: A Answer Explanation QID: 11-1 N8RO 09 Question # I 98 I Developer / Date; J..IR /5-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO I 8RO OCS OPS ILT Page: 25 of 83 02 April 2012

EXAMINATION ANSWER KEY OC SRO NRC Exam 295036 Secondary Containment High SumplArea Water Levell 5 2.4.47 - Emergency Procedures I Plan: Ability 4.2 4.2 to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material.

Level I SRO I Tier 1 I Group I 2 General ABN-38 PCC EOP References OCSOPS ILT Page: 26 of 83 02 April 2012

EXAMINATION ANSWER KEY OC SRO NRC Exam A is Correct. The plant is at power when an earthquake occurs resulting in a core spray line break downstream of the core spray pump suction valve. A lowering torus water level and flooding in the NW RB Corner Room becomes evident. The applicant must utilize the appropriate ABN and EOP (reference materials used by the crew during the event) to determine the correct course of action. Two EOPs are entered - Primary Containment Control EOP, and Secondary Containment Control EOP. In the torus water level leg of the PCC EOP, a scram is required if torus water level cannot be maintained above 110". The leak is stopped at 135" (by closing the core spray suction valve). In the SCC EOP, it is necessary to recognize if a primary system is discharging into the secondary containment. In this case, it is not. The SCC EOP then directs a reactor shutdown if water level goes above the max safe value in 2 or more areas. In the question, only 1 area has flooding.

With flooding stopped at 135" and in only one area, the EOPs do not require either a scram or reactor shutdown, however the flooding (Secondary Containment High Water Level) directly resulted in Core Spray System 1 Pumps inoperable. ABN*38, Station Seismic Event, requires a reactor scram if: 1) the seismic Explanation event caused a spurious actuation; 2) directly resulted in the inoperability of any safety system or a system required to complete a safe shutdown; or, 3) can potentially affect the public safety. The break in the core spray suction line is a direct result in the loss of a safety system from the earthquake and it is inoperable due to flooding at the MAX SAFE line in the NW Corner Room. Also, the torus leak itself was a result of the earthquake and this loss of primary containment integrity could potentially affect public health. Because of this, a manual reactor scram is required lAW the ABN, although not required by the EOPs.

All distractors are Incorrect but plausible. A scram is required if a primary system is discharging into the reactor Building and flooding in 2 areas exceeds the max safe value, lAW the Secondary Containment Control EOP. If flooding not from a primary system exceeds the max safe value in 2 areas, then a reactor shutdown is required in the same EOP. An emergency Depressurization is required in the Primary Containment EOP if torus water level lowers to 110". The drywell vent header downcomers begin uncovered at this point and the primary containment suppression function is lost. There is no procedurally required action to perform a rapid power reduction.

OCSOPSILT Page: 27 of 83 02 April 2012

EXAMINATION ANSWER KEY oc SRO NRC Exam Lesson Plan 2621.845.0.0056, Primary Containment Control PCC-10445, Given a set of system indications or data, evaluate Learning and interpret them to determine limits, trends and system ective/ status.

Question Source (New, Modified, Bank) Bank If Bank or MQdified VISION System/Question 10: 663639 Question Source: ILT 08-1 NRC Exam Previous 2 Exams: No Memory or X Comprehension Fundamental 3:SPK or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: Solve a Problem using Knowledge and its meaning 10CRF55 55.41b I 55.43b I 5 Content Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes I Point Value: 1 System 10 No.: 295036 PRA:

I No Safety 5

18l Initial License Level Function: o LORT OCSOPSILT Page: 28 of 83 02 April 2012

EXAMINATION ANSWER KEY OC SRO NRC Exam 10 10: 11-1 NSRO 10 Points: 1.00 A plant startup was in-progress. The following conditions currently exist:

  • AIIIRM Range switches are on Range 10
  • The REACTOR MODE SELECTOR switch is in STARTUP An event occurred which resulted in reduced recirculation pump flow, and NO operator actions have occurred. Total core flow is 30.2 x 106Ib/hr.

Which of the following states the correct Technical Specification requirements due to this plant condition, AND the basis for this requirement?

The plant shall be placed in ...

A. the SHUTDOWN CONDITION within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to prevent fuel cladding failure during a LOCA.

B. COLD SHUTDOWN within 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> to prevent exceeding 1% plastic strain on the cladding during a transient.

C. COLD SHUTDOWN within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to prevent fuel cladding temperature from exceeding 1500 OF during a LOCA.

D. COLD SHUTDOWN within 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> to ensure that the Onset of Transition Boiling (OTB) does not occur during a transient while operating Answer: D Answer Explanation QID: 11-1 NSRO 10 Question # I 10S I Developer / Date: JJR I 5-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO SRO 295014 Inadvertent Reactivity Addition /1 AA2.04 - Ability to determine and/or interpret the following as they apply to INADVERTENT 4.1 4.4 REACTIVITY ADDITION: Violation of fuel thermal limits

EXAMINATION ANSWER KEY OC SRO NRC Exam Level SRO Tier 1 Group 2 General TS 3.3.H References o is Correct. The question stem provides a condition where an event resulted in an unexpected reduction in Recic System below that which is required by TS 3.3.H. This reduction in system flow is an inadvertent addition of negative reactivity and violates MCPR limits (MCPR limits are set to ensure OTB does not occur during transients while operating). TS 3.3.H says that a minimum flow of 39.65 x 106 Ib/hr is required while in Range 10 of the IRMs and the Reactor Mode switch in STARTUP. This is done to ensure Explanation transient MCPR limits for operation are not violated. Because this TS does not provide any actions if exceeded, then TS 3.0.A applies, which requires the plant be placed in Cold Shutdown within 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. Answer 0 lists the correct TS requirement and the correct basis.

All distractors are Incorrect but plausible if the applicant does not recall the correct TS action or basis.

oes ops ILT Page: 30 of 83 02 April 2012

EXAMINATION ANSWER KEY OC SRO NRC Exam Lesson Plan 2621.850.0.0003, Overview/Highlights of Technical Specifications Learning TSX-10451, Referencing plant Technical Specifications (* from Objectivel memory for Initial Candidates) and given a set of plant conditions, determine, as applicable, the:

  • Definitions*
  • Safety Limits and Bases*
  • Limiting Safety System Settings and Bases*
  • Limiting Conditions for Operation and Applicability
  • LCO Action Requirements (SRO ONLy)
  • Surveillance Requirements (SRO ONLY)
  • Design Features, Containment, Auxiliary Equipment, Administrative Controls, and Appendix B Environmental Technical Specifications (SRO ONLY)
  • Bases for Surveillance Requirements, Design Features, Containment, Auxiliary Equipment, Administrative Controls, and Appendix B Environmental Technical Specifications. (SRO ONLY)*

Question Source (New, Modified, Bank) Modified If Bank Qr MQdifi~d VISION System/Question ID: 609455 Question Source: ILT 07-1 SRO NRC Exam Previous 2 Exams: No Memory or X Comprehension Fundamental 1:8 or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: Bases or purpose 10CRF55 55.41b I 55.43b I 2 Content Facility operating limitations in the technical specifications and their bases.

Justification for LORT questions with N/A KIA values < 3.0 OCSOPS ILT Page: 31 of 83 02 April 2012

EXAMINATION ANSWER KEY OC SRO NRC Exam Time to Complete: 1*2 minutes I Point Value: 1 System 10 No.: 295014 PRA:

I No Safety ~ Initial License Level Function:

1 o LORT OCSOPSILT Page: 32 of 83 02 April 2012

EXAMINATION ANSWER KEY OC SRO NRC Exam 11 10: 11*1 NSRO 11 Points: 1.00 The plant was at rated power when the following annunciator alarmed:

  • 460V STATION POWER - 1A2 MN BRKR TRIP Which of the following states the impact on the Core Spray System (Consider Active Components ONLY) and the MOST LIMITING Technical Specification (T.S.) action statement for the current plant conditions?

Cgre Sgril~ Core SRra~

S~~tem 1 S~stem 2 Most Limiting T.S.

InoRerable InoRerable Action Statem&nt Comggnents CgmRgnents A. One Booster Pump One Booster Pump The reactor may AND AND remain in operation One Main Pump One Main Pump not to exceed 15 days B. One Booster Pump One Booster Pump The reactor shall be ONLY ONLY placed in the COLD SHUTDOWN CONDITION C. One Booster Pump One Booster Pump The reactor shall be AND AND placed in the COLD One Main Pump One Main Pump SHUTDOWN CONDITION D. One Booster Pump One Booster Pump The reactor may ONLY ONLY remain in operation not to exceed 15 days Answer: B Answer Explanation QID: 11-1 NSRO 11 Question # I 11S I Developer I Date: JJR I 5-14*2012 OCSOPS ILT Page: 33 of 83 02 April 2012

EXAMINATION ANSWER KEY OC SRO NRC Exam Knowledge and Ability Reference Information Importance Rating K&A RO SRO 262001 AC Electrical Distribution A2.06

  • Ability to (a) predict the impacts of the following on the A.C. ELEC'rRICAL DISTRIBUTION; and (b) based on those 2.7 2.9 predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Deenergizing a plant bus Level I SRO I Tier 2 I Group I 1 General ABN-48 References Answer B is Correct. The annunciator in the stem describes a loss of power to USS 1A2 (which powers a core spray booster pump in each Core Spray System. System 1 includes the AlC booster pumps and System 2 includes the BID booster pumps.

When USS 1A2 is lost, 1 booster in each core spray system is lost. Also, USS 1A2 must be declared inoperable since it is unable to perform its design function. TS 3.7.B requires a cold shutdown condition action with the inoperability of Bus USS 1A2.

None of the (Parallel Isolation Valves (PIVs) are directly affected by the loss of DC control power to USS 1A2, and are all still functioning. One booster pump is included in each answer choice since it is LOD-1 that one CS booster pump in each loop Explanation is powered from USS 1A2.

All distractors are Incorrect but plausible. With USS 1A2 deenergized, the plant must be placed in the cold shutdown condition lAW TS 3.7. Due to Core Spray still being able to operate at designed flowrate, even with a loss of one booster pump in each system, the student may believe the plant is in a 15 day LCO from the TS 3.4 requirements with one or two non redundant CS components in each loop inoperable. In addition, Core Spray Main Pumps are powered from 4160 VAC power and are not affected. With no reference provided the student must recall both TS 3.4 and 3.7 TS section requirements from memory, as well as which Core Spray components are affected.

EXAMINATION ANSWER KEY OC SRO NRC Exam Lesson Plan 2621.828.0.0016, Electrical Distribution Learning ACD*10445, Given a set of system indications or data, evaluate Objective/ and interpret them to determine limits, trends and system status.

Question Source (New, Modified, Bank) Modified If Bank or Modified VISION System/Question ID: 811754 Question Source: ILT Bank Previous 2 Exams: No Memory or X Comprehension Fundamental 3:SPK or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: Solve a Problem using Knowledge and its meaning 10CRF55 55.41b I 55.43b I 2 Content Facility operating limitations in the technical specifications and their bases.

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1*2 minutes I Point Value: 1 System ID No.: 262001 PRA:

I No Safety [8J Initial License Level 6

Function: o LORT OCSOPS ILl Page: 35 of 83 02 April 2012

EXAMINATION ANSWER KEY OC SRO NRC Exam 12 10: 11*1 NSRO 12 Points: 1.00 A plant startup is in progress with the following conditions:

  • RPV pressure is 750 psig and rising slowly
  • RPV water level is in the normal band
  • Feedwater Pump C is in service An event then occurred. Plant conditions now include the following:
  • RV46 A indicates upscale
  • RPV water level swelled to 162" and continues to slowly rise Based on the above conditions, which of the following RPV pressure control strategies shall the SRO direct?

A. Use EMRVs lAW RPV Control- no ATWS B. Adjust the MPR setpoint lAW 201, Plant Startup C. Use the Isolation Condensers lAW RPV Control - no ATWS D. Use the Bypass Valve Opening Jack lAW 201, Plant Startup Answer: B Answer Explanation QIO: 11-1 NSRO 12 Question # I 125 I Developer 1 Date: JJR 15*14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO I 5RO OCSOPS ILT Page: 36 of 83 02 April 2012

EXAMINATION ANSWER KEY OC SRO NRC Exam 212000 RPS A2.09 - Ability to (a) predict the impacts of the following on the REACTOR PROTECTION SYSTEM; and (b) based on those predictions, 4.1 4.3 use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: High containme ure Level 2 Group 1 General 201 237E566 Sh. 3 References B is Correct. Under the conditions in the stem, with RPV pressure less that 825 pSig (TS value), a single RPS Bus loss will result in a full reactor scram and closure of the MSIVs. A failure of Drywell pressure instrument RV46 A by itself will result in a 1/2 scram on RPS Channel 1. The correct answer is to continue pressure control lAW 201, Plant startup on the MPR.

A & D are Incorrect but plausible if the applicant believes a full scram occurred due to RV46 A inserting a 1/2 scram on a RPS Channel. They might believe a full scram had occurred since this Explanation is true on a loss of power to one RPS bus at an RPV pressure <

825 psig. The High Drywell Pressure inserted a 1/2 scram, but RPS Bus 'A' still has power and a full scram did not occur. Post scram, the applicant may direct using RPV pressure control using the EMRVs and/or BPV Opening Jack.

C is Incorrect. This distractor is plausible if the applicant does not recognize that the Isolation Condensers are not available due to RPV level> 160". In addition, the RPV Control - no ATWS nrt"l"IlI,t'ture has not been entered.

Lesson Plan t2621.828.0.0037, Reactor Protection System RPS-10445, Given a set of system indications or date, evaluate Learning and interpret them to determine limits, trends, and system o ective/ status.

Question Source (New, Modified, Bank) Modified OCSOPSILT Page: 37 of 83 02 April 2012

EXAMINATION ANSWER KEY OC SRO NRC Exam If Baok gr Mgdified VISION System/Question 10: 811856 Question Source: ILT Bank Previous 2 Exams: No Memory or X Comprehension Fundamental 3:SPK or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: Solve a Problem using Knowledge and its meaning 10CRF55 55.41b I 55.43b I 5 Content Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes I Point Value: 1 System 10 No.: 212000 PRA:

I No Safety ~ Initial License Level 7

Function: o LORT oes OPS ILT Page: 38 of 83 02 April 2012

EXAMINATION ANSWER KEY OC SRO NRC Exam 13 ID: 11*1 NSRO 13 Points: 1.00 The plant is at rated power. An operator reports the feeder breaker to 125V MCC DC-2 tripped on overload.

Which ONE of the following is a correct statement for this condition?

A. The reactor may remain in operation for a period not to exceed 7 days, providing the 'A' Isolation Condenser is operable.

B. The reactor may remain in operation for a period not to exceed 7 days, providing the '8' Isolation Condenser is operable.

C. The reactor may remain in operation for a period not to exceed 7 days, providing BOTH Isolation Condensers are operable.

D. The reactor shall be placed in the COLD SHUTDOWN condition within 30 hrs regardless of Isolation Condenser operability.

Answer: A Answer Explanation QID: 11-1 NSRO 13 Question # I 13S I Developer / Date: J ..IR /5-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO SRO 207000 Isolation (Emergency) Condenser 2.2.40 - Equipment Control: Ability to apply 3.4 4.7 Technical Specifications for a system.

Level I SRO I Tier 2 I Group I 1 General TS 3.7.B.2 TS 3.8 References OCSOPS ILT Page: 39 of 83 02 April 2012

EXAMINATION ANSWER KEY OC SRO NRC Exam A is Correct. With a loss of DC-2, the 'B' Isolation Condenser is considered INOPERABLE. Even though DC-2 is required for plant operation and would normally require a 30hr LCO, TS 3.7.B.2 provides an exception that if the requirements of TS 3.8 are met (the 'A' Isolation Condenser being operable), plant operation may continue for a period not to exceed 7 days. Only Explanation an SRO is required to know LCO action statements, and therefore a RO would not be able to eliminate all distractors for this question.

All distractors are Incorrect but plausible if the applicant doesn't recall the correct LCO action statement for the conditions Lesson Plan 2621.828.0.0003, Isolation Condensers ICS-08653, Referencing plant Technical Specifications (* from Learning memory for Initial Candidates) and given a set of plant Objective/ conditions, determine, as applicable, the:

  • Definitions*
  • Safety Limits and Bases*
  • Limiting Safety System Settings and Bases*
  • Limiting Conditions for Operation and Applicability
  • LCO Action Requirements (SRO ONLY)
  • Surveillance R.equirements (SRO ONLY)
  • Design Features, Containment, Auxiliary Equipment, Administrative Controls, and Appendix B Environmental Technical Specifications (SRO ONLY)
  • Bases for Surveillance Requirements, Design Features, Containment, Auxiliary Equipment, Administrative Controls, and Appendix B Environmental Technical S ns. ON Question Source (New, Modified, Bank) New If Bank Qr MQdified VISION System/Question ID:

Question Source: N/A Previous 2 Exams:

Memory or X Cognitive Com prehension Fundamental 3:SPK Level or Analysis Knowledge OCSOPS ILT Page: 40 of 83 02 April 2012

EXAMINATION ANSWER KEY OC SRO NRC Exam NUREG 1021 Appendix B: Solve a .f.roblem using Knowledge and its meaning 10CRF55 55.41b I I 55.43b I 2 Content Facility operating limitations in the technical specifications and their bases.

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes I Point Value: 1 System 10 No.: 207000 PRA:

I No Safety 6

181 Initial License Level Function: o LORT OCS OPS ILT Page: 41 of 83 02 April 2012

EXAMINATION ANSWER KEY OC SRO NRC Exam 14 ID: 11*1 N5RO 14 Points: 1.00 The plant was at rated power. An event then occurred resulting in a major unisolable TBCCW leak.

The crew entered ABN-20, TBCCW FAILURE RESPONSE, and have completed all Immediate Operator Actions. Present plant conditions include the following:

  • RPV pressure indicates 950 psig and steady
  • RPV water level indicates 159" and slowly rising
  • Primary Containment parameters are normal The operator observes on Panel 12XR that ALL Feed and Condensate Pump bearing temperatures indicate 220 OF and rising.

Which of the following shall the SRO direct NEXT?

A. Bypass ADS Timers lAW the RPV Control- with ATWS EOP.

B. Trip ALL Feed Pumps ONLY lAW ABN-1, Reactor Scram.

C. Trip ALL operating Feed and Condensate Pumps lAW ABN-20, TBCCW Failure Response.

D. Secure all but ONE operating Feed and Condensate Pump lAW ABN-20, TBCCW Failure Response.

Answer: C Answer Explanation QID: 11-1 N5RO 14 Question # I 145 I Developer / Date: JJR /5-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO SRO 400000 Component Cooling Water 2.4.11 - Knowledge of abnormal condition 4.0 4.2 procedures.

EXAMINATION ANSWER KEY OC SRO NRC Exam Level SRO Tier 2 Group 1 General ABN-20 References C is Correct. The question stem describes an event where an unisolable TBCCW leak has occurred. Immediate Operator Actions require a reactor scram and a trip of all operating recirculation pumps. Subsequent operator actions require tripping all Feed & Condensate Pumps if bearing temperatures, as indicated on Panel 12XR, are ~195 OF. With bearing temperatures indicating 220 OF, all Feed & Condensate Pumps are required to be tripped.

A is Incorrect but plausible if the applicant believes an A TWS is Explanation in progress. Bypassing the ADS timers would be the next required action of the BOP.

B is Incorrect but plausible if the applicant believes the Feed Pumps should be tripped lAW ABN-1. ABN-1 requires tripping the Feed Pumps at 170". With RPV water level at 159" and slowly riSing, the high bearing temperature takes priority and all Feed and Condensate Pumps must be tripped lAW ABN-20.

o is Incorrect but plausible if the applicant does not recall the emental actions of ABN-20.

Lesson Plan 2621.828.0.0017, Feed and Condensate System CFW-10445, Given a set of system indications or data, evaluate Learning and interpret them to determine limits, trends and system Objective/ status.

Question Source (New, Modified, Bank)

I Modified If Bank or Modifhut VISION System/Question 10: 718367 Question Source: ILT Bank Previous 2 Exams: No OCSOPS ILT Page: 43 of 83 02 April 2012

EXAMINATION ANSWER KEY DC SRD NRC Exam Memory or X Comprehension Fundamental 3:SPK or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: Solve a Problem using Knowledge and its meaning 10CRF55 55.41b I 55.43b I 5 Content Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes Point Value: 1 System 10 No.: 400000 PRA: I No Safety 8

IZI Initial License Level Function: o LORT

EXAMINATION ANSWER KEY OC SRO NRC Exam 15 ID: 11-1 N8RO 15 Points: 1.00 The plant is at 65% power and stable. Plant conditions include the following:

Which of the following actions is REQUIRED lAW Technical Specifications?

A. Bypass APRM 4 B. CONFIRM a half-scram on RPS 1 C. CONFIRM a half-scram on RPS 2 D. Commence a normal plant shutdown Answer: B Answer Explanation QID: 11-1 N8RO 15 Question # I 158 I Developer / Date: JJR /5-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO 8RO 215005 APRM / LPRM 2.2.22 - Equipment Control: Knowledge of 4.0 4.7 Limiting Conditions for operations and safety limits.

Level I 8RO I Tier 2 I Group I 1 General TS 3.1.8.1 TS Table 3.1.1 References OCSOPS ILT Page: 45 of 83 02 April 2012

EXAMINATION ANSWER KEY OC SRO NRC Exam Answer B is Correct. APRM 4 is inoperable with four failed inputs (T.S. 3.1.B.1). Since there are now less than the required minimum APRM channels (3) for RPS-1, a half-scram must be inserted on RPS-1.

Answer A is Incorrect. It is true the crew may bypass APRM 4 however the required action per tech specs is to place/confirm a half-scram on RPS channel 1.

Answer C is Incorrect. RPS-1 has two inoperable APRMs (1 and

4) and would be tripped. RPS-2 has one inoperable APRM (6) and Explanation would not be tripped since it meets the operability requirements of Table 3.1.1.

Answer 0 is Incorrect. This action is not required and if taken, would not eliminate the requirement to trip RPS-1.

All distractors are plausible if the student does not recall action requirements for inoperable APRMs. The only action that is REQUIRED is that RPS 1 is placed in the TRIP condition, making all distractors incorrect as these are not required actions per Technical or Lesson Plan 2621.828.0.0029, Nuclear Instrumentation NIS-10445, Given a set of system indications or data, evaluate Learning and interpret them to determine limits, trends and system status.

Question Source (New, Modified, Bank) Bank If Bank Qr MQdified VISION System/Question 10: 718262 Question Source: ILT Bank Previous 2 Exams: No Memory or X Comprehension Fundamental 3:SPK or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: S.olve a Problem using Knowledge and its meaning oes OPS ILT Page: 46 of 83 02 April 2012

EXAMINATION ANSWER KEY OC SRO NRC Exam 10CRF55 55.41b I I 55.43b I 2 Content Facility operating limitations in the technical specifications and their bases.

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes I Point Value: 1 System 10 No.: 215005 PRA:

I No Safety [gI Initial License Level Function:

7 o LORT OCSOPS ILT Page: 47 of 83 02 April 2012

EXAMINATION ANSWER KEY OC SRO NRC Exam 16 10: 11*1 NSRO 16 Points: 1.00 The plant was at rated power with USS '1A2 out of service. A combined LOOP and LOCA then occurred. Plant conditions include the following:

  • RPV pressure is 780 psig and lowering
  • RPV water level is 102" and lowering 0
  • Torus temperature is 112 F and rising
  • DW pressure is 15 psig and rising 0
  • Drywell temperature is 282 F and rising
  • EDG 1 BREAKER indicates RED light ON
  • EDG 2 BREAKER indicates GREEN light ON Which of the following is the SRO required to direct NEXT?

A. Emergency Depressurize the RPV, lAW the Primary Containment Control EOP.

B. Initiate one loop of Containment Spray in the TORUS CLG mode, lAW the Primary Containment Control EOP.

C. Initiate one loop of Containment Spray in the DW SPRAY mode, lAW the Primary Containment Control EOP.

D. Lower reactor pressure to allow low pressure systems to inject into the RPV, lAW the RPV Control- no ATWS EOP.

Answer: A Answer Explanation QID: 11-1 NSRO 16 Question # I 165 I Developer / Date: JJR /5-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO I SRO

EXAMINATION ANSWER KEY OC SRO NRC Exam 223001 Primary CTMT and Aux.

A2.10 - Ability to (a) predict the impacts of the following on the PRIMARY CONTAINMENT SYSTEM AND AUXILIARIES; and (b) based on 3.6 3.8 those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: High rv\A'PIl ternnIPr::n SRO Tier 2 Group 2 General PCC EOP EOP User's Guide References A is Correct. The question stem describes a combined LOOP and LOCA event. The stem shows that EDG-2 is not powering Bus 1D, therefore System 2 Containment Spray Pumps are not available. In addition, with USS 1A2 not in service, Containment Spray System 1 Pumps are also unavailable. With Drywell Temperature >281 of and unable to be lowered, the PCC EOP directs Emergency Depressurization due to High Drywell Temperature to protect the Primary Containment.

B & C are Incorrect but plausible if the applicant doesn't recognize that BOTH Containment Spray systems are Explanation unavailable. Since Torus Temperature is >95 of, initiating Torus Cooling is required, but is not possible. Initiating Drywell Sprays is also required, however is also not possible with both Containment Spray systems unavailable.

D is Incorrect but plausible if the applicant prioritizes ensuring adequate core cooling due to the LOCA. Lowering RPV pressure to allow low pressure systems to inject would be required once level restoration has been entered. This will occur when RPV water level lowers to < 61". At 102" RPV level, the only REQUIRED EOP action that is possible is to Emergency De urize on Hi II TemnPlr::lTI OCSOPS ILT Page: 49 of 83 02 April 2012

EXAMINATION ANSWER KEY OC SRO NRC Exam Lesson Plan 2621.845.0.0056, RPV Control - no ATWS PCC-3000, Using Procedure EMG-3200.02, evaluate the Learning technical bases for each step in the procedure and apply this Objectivel evaluation to determine correct courses of action under emergency conditions.

Question Source (New, Modified, Bank) New If Bank Qr MQdified VISION System/Question 10:

Question Source: N/A Previous 2 Exams:

Memory or X Comprehension Fundamental 3:SPK or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: .solve a Problem using Knowledge and its meaning 10CRF55 55.41b I 55.43b I 5 Content Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes I Point Value: 1 System 10 No.: 223001 PRA:

I No Safety [8l Initial License Level 5

Function: D LORT

EXAMINATION ANSWER KEY OC SRO NRC Exam 17 10: 11*1 NSRO 17 Points: 1.00 The plant was starting up after a refuel outage. The Reactor Operator was withdrawing control rods, when the control rod position indication went dark for the control rod being withdrawn.

Control rod position indication was NOT regained when the URO inserted the control rod one notch. The Operator then attempted to fully insert the control rod in preparation for isolating the control rod. Neutron monitoring showed no change in counts as the control rod was inserted.

With the control rod valved out of service, lAW procedure 302.1 Control Rod Drive System, and the control rod position not known, which of the following Technical SpeCifications actions is required?

A. Immediately initiate action to fully insert all insertable control rods.

B. The reactor must be placed in the SHUTDOWN CONDITION within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

C. The SHUTDOWN MARGIN must be verified within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, including the effects of the unknown-positioned control rod.

D. Verify there are no more than 8 inoperable control rods valved out of service, prior to continuing with control rod withdrawals.

Answer: C Answer Explanation QID: 11-1 NSRO 17 Question # I 17S I Developer / Date: J ..tR /5-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO SRO 214000 RPIS 2.2.40 - Equipment Control: Ability to apply 3.4 4.7 Technical Specifications for a system.

Level I SRO I Tier I 2 I Group I 2

EXAMINATION ANSWER KEY OC SRO NRC Exam General TS 3.2.A References C is Correct. Shutdown margin is determined with the strongest reactivity control rod assumed fully withdrawn and all other control rods fully inserted. But since the control rod in the question has no position indication, and they are unable to verify that the control rod is fully inserted, its position is unknown.

Because of this, shutdown margin must be verified with this control not fully inserted. This is required lAW TS 3.2.A.2.

A is Incorrect but plausible. If SDM cannot be met while in Explanation REFUEL mode, then TS 3.2.A.5 requires that all control rods be fully inserted. The stem states a plant startup is in progress.

B is Incorrect but plausible. Only if the SDM cannot be verified within the time allowed (6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />), the plant must be placed in the shutdown condition within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> lAW TS 3.2.A.3.

o is Incorrect. TS 3.2.B.4 allows only 6 inoperable, valved out of service control rods. In any event, the startup cannot continue even if this verification was made.

Lesson Plan 2621.850.0.0090, Overview/Highlights of Technical Specifications Learning TSX-10451 , Referencing plant Technical Specifications (* from Objective/ memory for Initial Candidates) and given a set of plant conditions, determine, as applicable, the:

  • Definitions*
  • Safety Limits and Bases*
  • Limiting Safety System Settings and Bases*
  • Limiting Conditions for Operation and Applicability
  • LCO Action Requirements (SRO ONLY)
  • Surveillance Requirements (SRO ONLY)
  • DeSign Features, Containment, Auxiliary Equipment, Administrative Controls, and Appendix B Environmental Technical Specifications (SRO ONLy)
  • Bases for Surveillance ReqUirements, Design Features, Containment, Auxiliary Equipment, Administrative Controls, and Appendix B Environmental Technical ROON OCSOPSILT Page: 52 of 83 02 April 2012

EXAMINATION ANSWER KEY OC SRO NRC Exam Question Source (New, Modified, Bank) Bank If Bank or Modified VISION System/Question 10: 510956 Question Source: ILT Bank Previous 2 Exams: No Memory or X Comprehension Fundamental 3:SPK or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: Solve a Problem using Knowledge and its meaning 55.41b I 55.43b I 2&6 Facility operating limitations in the technical specifications and 10CRF55 their bases; Procedures and limitations involved in initial core Content loading, alterations in core configuration, control rod programming, and determination of various internal and external effects on core reactivity.

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes Point Value: 1 System 10 No.: 214000 PRA: I No Safety ~ Initial License Level 7

Function: [] LORT OCSOPS ILT Page: 53 of 83 02 April 2012

EXAMINATION ANSWER KEY OC SRO NRC Exam 18 10: 11-1 NSRO 18 Points: 1.00 The plant was at rated power when an event then occurred. Ten minutes later, the following plant conditions are observed:

  • ALL recirculation pumps are tripped
  • RPV water level is 112" and lowering
  • EMRV NR108B indicates RED light ON
  • ABN-40, Stuck Open EMRV, is being executed
  • Drywell pressure is 8 psig and rising slowly
  • Drywell temperature is 205 OF and rising slowly
  • Torus water temperature is 96 OF and rising slowly 180 170 160 10RUS 150 R

TEMPERATURE A (2.1~~

(OF) 140 130

" IiIr..

~

120 110

"'" "- c (11r 1O)

D 100

  • I
  • 02468101214161820 RE.AC1OR POWER (%)

FIG.L BIIT BORON INJECT ION INrrlATION TI""PERATURE OCS OPS ILT Page: 54 of 83 02 April 2012

EXAMINATION ANSWER KEY oc SRO NRC Exam CONTAINMENT SPRAY INITIATION LIMIT 600

- B 550

- I (7,550) c 500 450

- I 400

- I r

ORYWELL TEMPERATURE 350 f'F) 300

- I 250

- I 200

- I

- I 150 100

- JA o 1.8 5 10 15 20 25 30 35 40 DRYWELL PRESSURE (PSIG)

Which of the following actions is required AT THIS TIME?

A. Initiate the Liquid Poison System lAW Support Procedure 22, Initiating the Liquid Poison System, in the RPV Control- with ATWS EOP.

B. Initiate Torus Cooling lAW Support Procedure 25, Initiation of the Containment Spray System in the Torus Cooling Mode, in the Primary Containment Control EOP.

C. Initiate Drywell Sprays lAW Support Procedure 29, Initiation of the Containment Spray System for Drywell Sprays, from the Pressure Leg in the Primary Containment Control EOP.

D. Restore RPV water level 138" - 175" lAW Support Procedure 19, Feedwater/Condensate and CRD System Operation, in the RPV Control

- With ATWS EOP.

Answer: B I

Answer Explanation QID: 11-1 NSRO 18 OCSOPSILT Page: 55 of 83 02 April 2012

EXAMINATION ANSWER KEY DC SRD NRC Exam Question # 18S I Developer / Date: JJR / 5-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO SRO 219000 RHRlLPCI: Torus/Pool Cooling Mode A2.13 - Ability to (a) predict the impacts of the following on the RHRlLPCI:

TORUS/SUPPRESSION POOL COOLING MODE

and (b) based on those predictions, use 3.5 3.7 procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations
High suppression pool temperature Level I SRO I Tier 2 I Group I 2 General RPVC - with A TWS EOP User's Guide References EOP OCSOPSILT Paige: 56 of 83 02 April 2012

EXAMINATION ANSWER KEY OC SRO NRC Exam B is Correct. The plant was at rated power when an event resulted in an electrical ATWS. Indications are then provided 10 minutes after the scram: power is> 2%,1 EMRV is stuck open resulting in an elevated torus water temperature of 97F.

Containment Spray is available and Containment Spray is required to be placed in the TORUS CLG mode due to High Suppression Pool Temperature lAW the PCC EOP.

A is Incorrect but plausible since it is true that Liquid Poison would have to be initiated to maintain Torus Temperature below the BIIT curve, however with temperature at 97F and slowly rising, the only REQUIRED action at this time is to initiate Torus Cooling. ABN-40 is still being executed and there is a chance if NR108B closes the BIIT will never be reached.

Explanation C is Incorrect but plausible. Orywell sprays can be initiated in the Primary Containment Control EOP in 2 legs: OW temperature and OW pressure. In the temperature leg, conditions must first be allowed by the Containment Spray Initiation Limit (CSIL) curve.

To spray from the pressure leg, then OWlTorus pressure must first exceed 12 psig (this is a WAIT requirement). Therefore OW sprays can be initiated from the temperature leg (as allowed by CSIL) but not the pressure leg.

o is Incorrect but plausible. lAW the RPV Control* with ATWS EOP, RPV water level is NOT restored/maintained at 138-175, but lowered intentionally to control reactor power (Terminate and Prevent RPV ection Lesson Plan 2621.845.0.0053, RPV Control* with ATWS EWA.10445, Given a set of system indications or data, evaluate Learning and interpret them to determine limits, trends and system status.

Question Source (New, Modified, Bank)

If Bank or Modified I New VISION System/Question 10:

Question Source: N/A Previous 2 Exams:

EXAMINATION ANSWER KEY OC SRO NRC Exam Memory or X Comprehension Fundamental 3:SPK or Analysis Knowledge Cognitive Level NliREG 1021 Appendix B: Solve a Problem using Knowledge and its meaning 10CRF55 55.41b I 55.43b I 5 Content Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes I Point Value: 1 System 10 No.: 219000 PRA:

I No Safety [8J Initial License Level Function:

5 o LORT OCSOPS ILT Page: 58 of 83 02 April 2012

EXAMINATION ANSWER KEY OC SRO NRC Exam 19 10: 11-1 NSRO 19 Points: 1.00 The plant is shutdown for a refuel outage. Control rod 30-35 is to be replaced. Which of the following lists, in the correct order, the steps to prepare the cell to remove the control rod from the core?

A. 1. Remove fuel bundles A and B

2. Insert double blade guide
3. Remove fuel bundles C and D
4. Uncouple control rod
5. Withdraw control rod to position 48 B. 1. Remove fuel bundles A and C
2. Insert double blade guide
3. Remove fuel bundles Band D
4. Withdraw control rod to position 48
5. Uncouple control rod C. 1. Remove fuel bundles A and B
2. Remove fuel bundles C and D
3. Insert double blade guide
4. Uncouple control rod
5. Withdraw control rod to position 48 D. 1. Remove fuel bundles A and insert single blade guide
2. Remove fuel bundles B and insert single blade guide
3. Remove fuel bundles C and insert single blade guide
4. Remove fuel bundles D and insert single blade guide
5. Withdraw control rod to position 48
6. Uncouple control rod ocs OPS ILT Page: 59 of 83 02 April 2012

EXAMINATION ANSWER KEY OC SRO NRC Exam Answer: B Answer Explanation QID: 11-1 NSRO 19 Question # I 19S I Developer I Date: JJR I 5-14-2012 Importance Rating K&A Conduct of Operations 2.1.42 - Knowledge of ne w and spent fuel 2.5 3.4 movement rocedures.

Level SRO Tier 3 Category COO General 205.0 205.5 References B is Correct. Procedures 205.0 (Reactor Refueling) and 205.5 (Rod Withdrawaillnsertion During Refueling) provide the general guidance to remove a control rod from the core: 1. remove 2 opposite bundles; 2. insert blade guide; 3. remove last 2 bundles; Explanation 4 withdraw rod to 48; 5. Uncouple.

All distractors are Incorrect but plausible if the applicant is not familiar with the control rod removal process during refuel activities.

Lesson Plan 2621.812.0.0003, Reactor Refueling RFL-7442, Describe, in general, refueling and fuel handling Learning procedures to include precautions and limitations per Procedure 205 series.

Question Source (New, Modified, Bank) Bank

EXAMINATION ANSWER KEY OC SRO NRC Exam If Bank qr Modified VISION System/Question 10: 811860 Question Source: ILT 10-1 SRO NRC Exam Previous 2 Exams: Yes Memory or X Comprehension Fundamental 1:P or Analysis Knowledge Cognitive Level NLIREG 1021 Appendix B: Procedure steps and cautions 10CRF55 55.41b I 55.43b I 7 Content Fuel handling facilities and procedures.

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes Point Value: 1 System 10 No.: N/A PRA: I No Safety ~ Initial License Level N/A Function: o LORT OCSOPS ILT Page: 61 of 83 02 April 2012

EXAMINATION ANSWER KEY OC SRO NRC Exam 20 10: 11-1 NSRO 20 Points: 1.00 The plant is in a refuel outage. Due to the loss of SRM 24, fuel was being shuffled in core quadrants 1,2, and 3.

While reviewing work packages for the following day, you note a maintenance activity requiring a tagout to de-energize 24 VDC Power Panel A.

If the maintenance activity were allowed to occur as scheduled, which of the following states the impact on refueling, if any?

A. There will be NO impact on the core alterations.

B. Core alterations will be restricted to core quadrant 3 ONLY.

C. ALL core alterations must cease due to the loss of the required number of operable SRMs.

D. ALL core alterations must cease due to the loss of Secondary Containment Integrity and the auto start of SGTS.

Answer: C Answer Explanation QID: 11-1 NSRO 20 Question # I 20S I Developer / Date: JJR /5-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO SRO Equipment Control 2.2.36 - Ability to analyze the effect of maintenance activities, such as degraded 3.1 4.2 power sources, on the status of limiting conditions for operations.

Level I SRO I Tier 3 I Category I EQC General 706E812 TS3.9 References sh.3,5,6 oes ops ILl Page: 62 of 83 02 April 2012

EXAMINATION ANSWER KEY OC SRO NRC Exam C is Correct. The plant is in a refuel outage with SRM 24 inoperable. Core alterations are occurring in the other core quadrants with operable SRMs. If 24 VOC Power Panel A is de energized, this will render SRMs 21 and 22 inoperable. TS 3.9.0 provides the following: During CORE ALTERATIONS at least two (2) source range monitor (SRM) channels shall be OPERABLE and inserted to the normal operating level. One of the OPERABLE SRM channel detectors shall be located in the core quadrant where CORE ALTERATIONS are being performed, and another shall be located in an adjacent quadrant. TS 3.9.G provides the following: With any of the above requirements not met, cease CORE ALTERATIONS or control rod removal as appropriate, and initiate action to satisfy the above requirements.

Since only 1 SRM remains operable in quadrant 3, the requirement for 2 operable SRMs will not be met and core alterations must cease.

Explanation A is Incorrect but plausible since the refuel activities are impacted. The applicant may not recall the loss of 24 VOC Power Panel A would affect SRMs 21 and 22 and/or they may not recall the Tech Spec requirements for operable SRMs while performing core alterations.

B is Incorrect but plausible. Since SRM 23, in core quadrant 3 is still operable, the candidate may think that fuel moves are still allowed in that single quadrant. But as shown, 2 SRMs are required.

o is Incorrect but plausible. The loss of 24 VOC Power Panel will isolate RB normal Vent and initiate the Standby Gas treatment System (SGTS). This will not cause SGTS or Secondary Containment to be ino ble.

OCSOPSILT Page: 63 of 83 02 April 2012

EXAMINATION ANSWER KEY OC SRO NRC Exam Lesson Plan 2621.828.0.0029, Nuclear Instrumentation NIS-10451, Referencing plant Technical Specifications (* from Learning memory for Initial Candidates) and given a set of plant Objective/ conditions, determine, as applicable, the:

  • Definitions*
  • Safety Limits and Bases*
  • Limiting Safety System Settings and Bases*
  • Limiting Conditions for Operation and Applicability
  • LCO Action Requirements (SRO ONLY)
  • Surveillance Requirements (SRO ONLY)
  • Design Features, Containment, Auxiliary Equipment, Administrative Controls, and Appendix B Environmental Technical Specifications (SRO ONLY)
  • Bases for Surveillance Requirements, Design Features, Containment, Auxiliary EqUipment, Administrative Controls, and Appendix B Environmental Technical Specifications. (SRO ONLY)*

Question Source (New, Modified, Bank) Bank If Bank Q[ MQdified VISION System/Question ID: 718376 Question Source: ILT 09-1 SRO NRC Exam Previous 2 Exams: Yes Memory or X Comprehension Fundamental 3:PEO or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: Predict an Event or Qutcome 55.41b I 55.43b I 2,6,7 Facility operating limitations in the technical specifications and 10CRF55 their bases; Procedures and limitations involved in initial core Content loading, alterations in core configuration, control rod programming, and determination of various internal and external effects on core reactivity; Fuel handling facilities and procedures.

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes Point Value: 1

EXAMINATION ANSWER KEY oc SRO NRC Exam System 10 No.: N/A PRA: I No Safety ~ Initial License Level Function:

N/A o LORT OCSOPS ILT Page: 65 of 83 02 April 2012

EXAMINATION ANSWER KEY OC SRO NRC Exam 21 10: 11-1 NSRO 21 Points: 1.00 The plant was at rated power when an event occurred.

20 minutes later, the following plant conditions exist:

  • Offgas Radiation Monitors have risen and continue to rise
  • Several Turbine Building AND Reactor Building Area Radiation Monitors are in alarm (but on-scale)
  • Turbine Building ap is positive
  • The Shift Manager has declared an Alert due to Radiological Effluent Which of the following actions is required?

A. Close the MSIVs lAW the Radioactivity Release Control EOP.

B. Close the MSIVs lAW ABN-26, High Main Steam/Offgas/Stack Effluent Activity.

C. Emergency Depressurize the RPV lAW the Radioactivity Release Control EOP.

D. Emergency Depressurize the RPV lAW the Secondary Containment Control EOP.

Answer: A Answer Explanation QID: 11-1 NSRO 21 Question # I 21S I Developer / Date: JJR /5-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO SRO Radiation Control 3.8 4.3 2.3.11 - Ability to control radiation releases.

Level I SRO I Tier I 3 I Category I RPT OCSOPSILT Page: 66 of 83 02 April 2012

EXAMINATION ANSWER KEY OC SRO NRC Exam General RREOP EOP User's Guide References A is Correct. The question states that an event had occurred.

The conditions show that MSL and offgas radiation has increased, TB ARMs are in alarm and that TB AP is positive.

These indicate a primary steam leak in the TB. The stem also shows that an alert emergency condition has been declared due to radiological effluents. This is an entry condition into the Radioactivity release Control EOP. The first step is to isolate primary systems discharging outside the primary and secondary containments. Closing the MSIVs would stop the leak into the TB.

Explanation B is Incorrect but plausible since ABN-26 requires closing the MSIVs when MSL radiation is > 800 mr/hr and the stem shows only 500 and rising slowly.

C is Incorrect but plausible since the Radioactivity Release Control EOP does require ED, but only after a GE is declared.

o is Incorrect but plausible since ED is also required in the Secondary Containment Control EOP, but the MAX SAFE must first be exceeded leak in the in 2 areas first.

Lesson Plan 2621.830.0.0015, Radiation Control - Admin Learning 2.3.11, Ability to control radiation releases Question Source (New, Modified, Bank)

If Bank or Modified I Bank VISION System/Question 10: 811863 Question Source: ILT 10-1 SRO NRC Exam Previous 2 Exams: Yes OCSOPS ILT Page: 67 of 83 02 April 2012

EXAMINATION ANSWER KEY OC SRO NRC Exam Memory or X Comprehension Fundamental 3:SPR or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: Solve a Problem using References 10CRF55 55.41b I 55.43b I 4 Radiation hazards that may arise during normal and abnormal Content situations, including maintenance activities and various contamination conditions.

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes I Point Value: 1 System 10 No.: N/A PRA:

I No Safety ~ Initial License Level N/A Function: [] LORT oes OPS ILT Page: 68 of 83 02 April 2012

EXAMINATION ANSWER KEY OC SRO NRC Exam 22 10: 11*1 NSRO 22 Points: 1.00 The plant was at rated power when a LOCA and A TWS occurred. Plant conditions include the following:

  • Reactor power is 15% and steady
  • RPV water level indicates -16" and lowering
  • Emergency Depressurization has been performed
  • SP-17, Terminate and Prevent Injection, has been completed
  • RPV Pressure has just lowered below the Minimum Steam Cooling Pressure (MSCP) lAW the RPV Control - with A TWS EOP, which of the following systems shall the SRO direct FIRST to restore RPV water level AND, lAW the EOP User's Guide, which is the correct basis for this action?

A. Feed and Condensate lAW SP-19, Feedwater/Condensate and CRD System Operation, since it injects outside the core shroud.

B. Fire Water via the Core Spray System lAW SP-20, Low Pressure Injection During an A TWS, due to its ability to be throttled and controlled.

C. Core Spray System lAW SP-20, Low Pressure Injection During an ATWS, due to its ability to restore RPV water level faster than other injection systems.

D. Condensate Transfer via the Core Spray System lAW SP-20, Low Pressure Injection During an ATWS, due to its ability to throttle and is at a higher water purity than Fire Water.

Answer: A Answer Explanation QIO: 11-1 NSRO 22 Question # I 22S I Developer / Date: JJR /5-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO SRO Emergency Procedures I Plan 2.4.22 - Knowledge of the bases for prioritizing safety functions during abnormal/emergency 3.6 4.4 operations.

OCSOPS ILT Page: 69 of 83 02 April 2012

EXAMINATION ANSWER KEY OC SRO NRC Exam Level SRO Tier 3 Category EOP General RPVC - with A TWS EOP User's Guide References EOP A is Correct. The question stem describes a condition where there is both a LOCA and ATWS. When ED is performed during an ATWS, pressure is allowed to lower below the MSCP, then makeup to the RPV commences via a series of preferred Safety Systems. Feed/Condensate and CRD are the FIRST priority since they inject outside the Core Shroud, allowing the cold water injected to warm and mix with borated water before entering the core. The first makeup source the SRO shall direct is Feed and Condensate.

Explanation B & 0 are Incorrect but plausible. Fire Water and Condensate Transfer via Core Spray are one of the next sources of water in line for makeup due to their ability to be throttled. Feed and Condensate has a higher priority though due to it injecting outside the core shroud where Fire Water and Condensate Transfer would inject cold water directly on top of the core.

C is Incorrect but plausible. The Core Spray system is the last source of makeup during an A TWS due to its injection of large quantities of cold unborated water injecting directly on the core.

Lesson Plan 2621.845.0.0053, RPV Control - with A TWS EWA-2257, Given the EOP, describe in detail each Learning step/statement, including the technical basis, and how to verify n ....-T""'r... each Question Source (New, Modified, Bank)

If Bank gr M~ulified I New VISION System/Question 10:

Question Source: N/A Previous 2 Exams:

OCSOPS ILT Page: 70 of 83 02 April 2012

EXAMINATION ANSWER KEY OC SRO NRC Exam Memory or X Comprehension Fundamental 3:SPK or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: Solve a Problem using Knowledge and its meaning 10CRF55 55.41b I 55.43b I 5 Content Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes I Point Value: 1 System ID No.: N/A PRA:

I No Safety N/A r8J Initial License Level Function: D LORT

EXAMINATION ANSWER KEY OC SRO NRC Exam 23 10: 11-1 N5RO 23 Points: 1.00 The plant is at rated power with CRD Pump B tagged out of service to replace the pump oil.

The work order requires running surveillance test 617.4.001, CRD Pump Operability Test as a Post Maintenance Test (PMT) following work completion.

lAW MA-AA-716-012, Post Maintenance Testing, which of the following states an additional requirement for the PMT of this pump, if any?

A. NO other actions outside of the surveillance are required.

B. Motor current should also be monitored and documented.

C. Bearing temperatures should also be monitored and documented.

D. A VT-2 leakage inspection should be performed and documented.

Answer: C Answer Explanation QID: 11-1 N5RO 23 Question # I 235 I Developer / Date: JJR /5-14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO 5RO Equipment Control 2.2.19 - Knowledge of maintenance work order 2.3 3.4 requirements.

Level I 5RO I Tier 3 I Category I EQC General MA-AA-716-012 617.4.001 References

EXAMINATION ANSWER KEY OC SRO NRC Exam C is Correct. lAW MA-AA-716-012, Post Maintenance Testing, Attachment 1, Generic Post Maintenance Test Matrix, there are 3 types of tests for pump lubricant changeout: bearing temperature, external leakage, and lubrication level checks. The surveillance test does not test or verify any of these Explanation recommended actions. Of those actions listed, only answer C specifies one of the listed actions.

All distractors are Incorrect but plausible if the applicant is not familiar with the requirements for PMT.

Lesson Plan 2621.828.0.0011, Control Rod Drive System CRD-0021, Identify and interpret the test and surveillance procedures for the CRD System, including personnel and nrn,p"T allocation.

Question Source (New, Modified, Bank) Bank If Bank Qr Modified VISION System/Question ID: 663659 Question Source: ILT 08-1 SRO NRC Exam Previous 2 Exams: No Memory or X Com prehension Fundamental 2:DR or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: Describing or recognizing Relationships 10CRF55 55.41b I 55.43b I 5 Content Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes I Point Value: 1 OCSOPS ILT Page: 73 of 83 02 April 2012

EXAMINATION ANSWER KEY OC SRO NRC Exam System 10 No.: N/A PRA: I No Safety [8j Initial License Level N/A Function: o LORT

EXAMINATION ANSWER KEY OC SRO NRC Exam 24 10: 11-1 NSRO 24 Points: 1.00 The plant was at rated power when an event occurred. Present plant conditions are as follows:

  • RECIRC PUMP SUCTION TEMPS indicates 440 OF and lowering
  • RPV water level indicates 134" and rising slowly
  • Drywell pressure indicates 11 psig
  • Drywell temperature indicates 225 OF
  • Torus water temperature indicates 153 OF and rising slowly
  • Torus water level indicates 160" and stable 180 170 160 lORUS 150 R lEMPERATURE A (2.15~

(OF) 140 130 120 "

110

'"' "- C D

(111.110) 100

  • . * . I . .

02468101214161820 REAClOR POWER (%)

FIG. L lilT 10RON IN.tECTION INITIATION TEMPERATURE

EXAMINATION ANSWER KEY oc SRO NRC Exam TORUS HIGH LEVEL 230 220 210 TORUS TEMPERATURE 190 (I:J TORUS 180 LEVEL 170

-144 160 -154 164

-174 150 +---+---+---+---~--~--~~~~~-4--~~~ --180

-188 140 ~~+-~+-~+~~~~~~t~~~~~~~~~~

0 100 200 300 400 500 600 700 800 900 1000 1100 RPV PRESSURE (PSIG)

TORUS LOAD LI MIT 200 19

°L

~

i:

(530,188}

TORUS WATER 18 oj  !

1 i

I (700,182) j LEVEL (IN.)

17o t *ii I

(900, 174.5) *

~

i  ! (1015,168)

B 10o 1 1 ...I - ~

~

!  ! I 1

1 j i 1 I! (1125,151) 15o ~

i  : /'/

A/

1'/

0 i .1

~

o 100 200 300 400 500 600 700 800 900 1000 1100 1200 RPV PRESSURE (PSIG)

OCSOPSILT 02 April 2012

EXAMINATION ANSWER KEY OC SRO NRC Exam Temperldure CF, w

8 U1 CJ o ~

i 50 ~ i 100 ~ '\

I i

"\

I 150 i 200

\ i 250

\

j 300 I 350 \ I I

400 \ I 450 I~ i rn I!

"C

\

I i

c I!

500 II C

II 550 \ I CD CL 11

IS'

!I.

i

\ i *3 sa 600 650 \ I I

i 1\ I 700 750 \ I 800 \ i I

850 \

900 \

950 \

1000 \ i I

I 1050 1100 ~---" - \ I OCSOPS ILT Page: 77 of 83 02 April 2012

EXAMINATION ANSWER KEY OC SRO NRC Exam Which ONE of the following actions shall the SRO direct? (See EOP Figures above)

A. Initiate Liquid Poison lAW Support Procedure 22, Initiating the Liquid Poison System B. Emergency Depressurize the RPV lAW the Emergency Depressurization

- With ATWS EOP C. Before RECIRC PUMP SUCTION TEMPS indicate about 425 OF, stop injection into the RPV from the Core Spray System, lAW the RPV Control

-NoATWS EOP D. Terminate and prevent RPV injection from sources external to the Primary Containment not required for adequate core cooling lAW the Primary Containment Control EOP Answer: C Answer Explanation QID: 11-1 N5RO 24 Question # I 245 I Developer I Date: JJR 15*14-2012 Knowledge and Ability Reference Information Importance Rating K&A RO 5RO Conduct of Operations 2.1.25 - Ability to interpret reference materials, 3.9 4.2 such as graphs, curves, tables, etc.

Level 5RO I Tier 3 I Category I COO General RPV Control* no References ATW5 EOP

EXAMINATION ANSWER KEY OC SRO NRC Exam C is Correct. The plant was at power when an event occurred.

The conditions provided require entry into the RPV Control- No ATWS and Primary Containment Control EOP. The recirculation pump suction temperature (which is trending down) shows that RPV pressure is about 375 psig and lowering. A note in the RPV Control - No A TWS EOP states that if Core Spray is running (which it is), then to secure Core Spray before RPV pressure drops to 310 psig (which equates to about 425 OF), as long as core cooling is assured. RPV water level is given as 134" and slowly rising; thus, adequate core cooling is assured. With this, and RPV pressure lowering, Core Spray should be prevented from injecting.

A is Incorrect but plausible. The indications show that several control rods did not insert to position 00 but inserted to position

04. Under these conditions, the reactor can still be declared shutdown under all conditions and thus the RPV Control - No Explanation ATWS is the correct EOP. The conditions show that the Boron Injection Initiation Temperature (BIIT) curve is violated, and SLC should be injected if the RPV Control- With ATWS EOP was the correct EOP. Since there is no ATWS EOP entry, there is no direction to inject SLC due to violation of the BIIT curve.

B is Incorrect but plausible. Emergency Depressurization would be required if the Heat Capacity Temperature Limit (HCTL) were violated. But the conditions show that the point is on the good side of the HCTL curve and no ED is required.

D is Incorrect but plausible. Terminating RPV injection from those sources external to the Primary Containment not required for adequate core cooling is appropriate when the Torus Load Limit (TLL) curve is violated. The provided conditions show that TLL is not violated and thus termination of external sources is not required.

Lesson Plan 2621.845.0.0052, RPV Control - no A TWS ENA-3055, Given a copy of RPV Control - No A TWS, describe in Learning detail each step or conditional statement, including technical Objectivel basis, and how to perform each step as required.

OCSOPS ILT Page: 79 of 83 02 April 2012

EXAMINATION ANSWER KEY OC SRO NRC Exam Question Source (New, Modified, Bank) Bank If Bank or Modified VISION System/Question ID: 663915 Question Source: ILT Bank Previous 2 Exams: No Memory or X Comprehension Fundamental 3:SPK or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: Solve a E.roblem using Knowledge and its meaning 10CRF55 55.41b I 55.43b I 1 Content Conditions and limitations in the facility license.

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes I Point Value: 1 System ID No.: N/A PRA:

I No Safety [8J Initial License Level N/A Function: D LORT OCSOPSILT Page: 80 of 83 02 April 2012

EXAMINATION ANSWER KEY OC SRO NRC Exam 25 10: 11 ..1 NSRO 25 Points: 1.00 Given the following:

  • A Site Area Emergency has been declared at Oyster Creek.
  • An Emergency Exposure of greater than 5 Rem TEDE is required to terminate a radioactive release.

According to EP-AA-113, Personnel Protective Actions, who must authorize the Emergency Exposure?

1. The Shift Manager in the Control Room
2. The Station Emergency Director in the TSC
3. The Corporate Emergency Director in the EOF A. 1 ONLY B. 2 ONLY C. BOTH 1 and 2 D. BOTH 2 and 3 Answer: B
  • Answer Explanation QID: 11-1 NSRO 25 Question # I 25S I Developer 1 Date: J..IR 15*14*2012 Knowledge and Ability Reference Information Importance Rating K&A RO SRO Radiation Control 2.3.4 - Knowledge of radiation exposure limits 3.2 3.7 under normal or emer~ ency conditions.

Level I SRO Tier I 3 I Category I RPT OCSOPSILT Page: 81 of 83 02 April 2012

EXAMINATION ANSWER KEY oc SRO NRC Exam General EP-AA-113 EP-AA-1007 References B is Correct. Per EP-AA-1007 (among others), emergency exposure controls are non-delegable responsibilities that remain with the Station Emergency Director in the TSC. Since the TSC is activated, the Shift Manager (Shift Emergency Director) has transferred this responsibility to the Station Emergency Director.

Per EP-AA-113, the Station Emergency Director (TSC) authorizes emergency exposures greater than 5 Rem TEDE.

Explanation All distractors are Incorrect but plausible. The Shift Manager does authorize emergency exposure limits, however only until Command and Control is transferred to the TSC. The applicant may not recall authorizing emergency exposure is a non delegable responsibility. Since the EOF is activated, they may assume emergency exposure control authorization has now been transferred to the EOF .

Question Source (New, Modified, Bank) Bank If Bank Qr Modified VISION System/Question ID: None Question Source: Peach Bottom 2009 NRC Exam Previous 2 Exams: No Memory or X Comprehension Fundamental 1:P or Analysis Knowledge Cognitive Level NUREG 1021 Appendix B: Procedure steps and cautions 10CRF55 Content 55.41b I 5S.43b I 4 OCSOPS ILT Page: 82 of 83 02 April 2012

EXAMINATION ANSWER KEY OC SRO NRC Exam Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.

Justification for LORT questions with N/A KIA values < 3.0 Time to Complete: 1-2 minutes I Point Value: 1 System 10 No.: N/A PRA:

I No Safety N/A I8l Initial License Level Function: o LORT OCS OPS ILT Page: 83 of 83 02 April 2012

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