ML14307B051

Final Written Examination with Answer Key (401-5 Format) (Folder 3)

ML14307B051
Person / Time
Site:	Calvert Cliffs
Issue date:	10/21/2014
From:	Lally C M Operations Branch I
To:	Calvert Cliffs
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ML14079A469	List: IR 05000317/2014301 ML14092A008 ML14304A680 ML14304A721 ML14304A727 ML14307B005 ML14307B051 ML14307B072 ML14307B104 ML15043A709 ML15146A207
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CAL VERT CLIFFS NUCLEAR POWER PLANT 2014 NRC INITIAL LICENSED OPERATOR RO WRITTEN EXAM KEY Page 1 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY Revision 1 (Incorporates Ops Reviewer Comments)

• #1 -Changed distracter "B" to make it clearly incorrect.

Changed correct answer wording and made it answer "B", and old distractor "B" is now "C" * #3 -Made editorial change ... added space in distracter "B" * #4 -Minor editorial correction to the stem of the question.

Changed "stay" to "remain" in "A" & "C" distracters

• #6 -Clarified wording in the stem of the question * #7 -Minor editorial correction to the stem of the question Changed CCW to Component Cooling * #9 -Changed stem of question and made previously correct answer "B" incorrect, changed wording on distractors "C" & "D" with "D" being correct. * #10-Changed wording on distractor "A" and re-ordered the remaining distractors so correct answer is now "B" * #12-Got rid of unnecessary wording * #13-Added clarifying information to the stem of the question * # 14 -Minor editorial correction to all four answer choices * #19-Minor editorial correction to the stem of the question * #22 -Minor editorial correction to the stem of the question * #23 -Minor editorial correction to the stem of the question * #24-Minor editorial correction to distractors "A" & "D" * #25 -Minor changes to distractor "A" to make more incorrect and minor change to "B" to make more correct. * #26 -Minor editorial correction to the stem of the question * #27 -Changed stem of question to indicate not BOC, changed all distractors to match stem. * #28 -Changed stem of question to initial start ofRCP, changed all distractors to match the stem * #29 -Minor editorial correction to the stem of the question * #31 -Changed correct answer "B" to be more straightforward

• #32-Minor editorial correction to the stem of the question * #33-Minor editorial correction to the stem of the question * #34-Changed distractor "B" & "D" to SRW TB isolations vice Main HPSI Hdrs * #36 -Minor editorial correction to the stem of the question * #38-Minor editorial correction to the stem of the question * #41 -Minor editorial correction to the stem of the question * #43 -Minor editorial correction to the stem ofthe question * #46 -Minor editorial correction to the stem of the question * #4 7 -Added seconds to all answers Page 2 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY Revision 1 (continued)

• #48 -Minor editorial correction to the stem of the question and answers "B & C" * #51 -Minor editorial correction to the stem of the question and distractor "D" * #52 -Changed "C" & "D" to say "Purge continues to operate" * #53-Changed correct answer to 1B DIG, reworded distractors "A,C & D" * #55-Changed distractor "B" to more plausible distractor

• #57-Minor editorial correction to the stem and changed wording on correct answer * #58 -Minor editorial correction to the stem of the question * #62 -Changed order of distractors

• #66 -Minor editorial correction to the stem of the question * #67 -Reworded the correct answer * #68 -Changed correct answer from 4 hrs to 3 hrs * #68 -Added ";?:" to all answers * #70-Minor editorial correction to the stem of the question * #73 -Minor editorial correction to the stem of the question Numerous changes to distractor order to equalize correct answers: 18-A, 19-B, 19-C, 19-D Revision 2 (Incorporates Collegial Review Comments)

• #1-Changed stem from switchyard troubleshooting to RPS troubleshooting

• #17-Changes stem voltage to be more within normal values * #19-Changed stem of question to be clearer, all distractors changed to be in line with new wording of stem. Correct answer changed from C to B * #23 -Minor wording in stem to clarify which switches * #25 -Changed distractor "A" (removed > 15 min) * #30-Minor change to stem to clarify * #35-Changes distractor "A" & "D" from 0.5-5psig to 0.5-8psig

• #3 7 -Changed wording from increase/decrease to raise/lower

• #38-Minor changes to stem to clarify. Fixed error in explanation

• #60-Changes to distractors "A" & "C" (removed "or Containment Purge), also minor change in format of stem * #64 -Changed distractor "B" to make it more incorrect

• #68 -Minor changes to distractor "D" to clarify temp was high * #72 -Removed using provided reference

• #73-Changed distractor "A" to clarify and make more incorrect

• 18-A, 20-B, 18-C, 19-D Page 3 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY Revision 3 (Incorporates Exam Validator Comments)

• #1 -Replaced question with a modified Bank question to match KA * #9-Minor changes to Stem to clarify * # 18 -Added "Fallowing an RFO" * #20 -Removed SDM * #32-Changed to Main Steam Leak Event and changed indications to PAMS instruments

• #35 -Changed distractors "A" & C" slightly to clarify * #39-Added" remain" to answers B & D * #43 -Bolded "Technical Specifications" * #52 -Inserted "Saturated".

Changed distractor for B&D to placing monitor in Level Cal vice putting fuel in a safe location * #60 -Bolded "Prevent Exceeding" * #61 -Capitalized "Fans" * #65 -Changed distractor "B" to > 1 minute * #69 -Changed Stem to Double Valve Isolation vice hazardous energy * #71 -Reworded distractors to clarify Revision 4 (Incorporates 2nd Round Exam Validator Comments)

• #9 -Minor change to wording in stem to clarify * #12-Added "Isolation" to LID CV's in stem * #31 -Minor editorial change to stem to clarify * #61 -Fixed grammatical error in distractor C * #62 -Minor wording change to stem to clarify what question asking * #75 -Changed 2nd part of distractor "D" to match distractor "A" Page 4 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY Revision 5 (Incorporates Fleet Reviewer & NRC Comments)

• #1-Minor editorial change to stem for ease of reading (replace "basis" with "reason")

• #2 -Minor editorial change added "significant" to stem * #3 -Minor editorial change to stem (Remove "Given" at beginning of question)

• #6 -Minor change to stem (added "for given conditions")

to ensure all incorrect answers are wrong * #8 -Added level change to all 4 answers to more closely match the Kl A, added NOP/NOT to stem * #12-Added explanation ofhigh cognitive level to answer explanation

• #14-Changed "B" and "C" distractors to equalize correct answers for exam * #20-Wrote new question based on feedback from NRC about KJ A mismatch and potential 2 right answers. * #23-Minor editorial change to stem to read easier (moved "fire effects" to end) * #24 -Minor editorial change to stem (grammatical errors fixed) * #25-Added explanation of high cognitive level to answer explanation

• #29-Changed distractors "C" & "D" to make more plausible but still wrong * #31 -Reworded distractors "A" & "C" for similar wording. So correct answer is not only one worded a certain way * #32 -Removed 1st sentence of stem, not needed * #39-Added "for 5 days" to stem to not have boron equalization occuring * #41 -Added S/G water level in stem (NRC suggestion)

• #42-Minor editorial change to stem to clarify only one answer is correct * #46 -Minor editorial change to stem "continue operation" vice "raise power" * #52-Editorial change, added "automatically" to distractors "A" & "B" * #57-Minor wording change to answer "C", "build up" vice "left in" * #63 -Minor editorial changes to stem and distractor "A" to make distractor plausible grammatically

• #67 -Added explanation ofhigh cognitive level to answer explanation

• #68 -Added explanation ofhigh cognitive level to answer explanation

• #72 -Changed question to a new question based on similarity to question on Audit Exam (different question # but very similar concept and wording) * #73-Minor editorial change to stem "directed by AOP-2A" 18-A, 20-B, 18-C, 19-D 35 Low Cog, 40 High Cog Page 5 of 157 Rev. 6

• 2014 NRC RO EXAM MASTER KEY Revision 6 (Added Learning objectives where applicable, Comments following NRC Validation incorporated)

• #1 -Changed distractor "D" per NRC request due to plausibility (Increased difficulty)

• #2 -Removed "expected" from stem, changed last bullet in stem, removed "significant" and added change of 50#. (No change to difficulty).

Changed stem to 2250 PSIA and stable. * #4 -Added the word "only" after SIAS in answer choices "A" & "B" (No change in difficulty)

• #5-Replaced distractor D with "liB RCP rotor locked. (Increased difficulty)

• #6-Added "Tube" to transfer gate valve (No change in difficulty)

• #7 -Added to distractor "A" & "C" explanations (No change in difficulty)

• #10-Added to comment about why applicant should know correct answer even though not trained on specific steps in CP-436 (No change in difficulty)

• #11 -Added "does not prevent" to explanation for correct answer (No change in difficulty)

• #12-Removed reference to ADV, reformatted to 2 X 2. (No change in difficulty)

• #13-Added plausibility statement to incorrect distractors (No change in difficulty)

• #20-Changed RPM to RFP in distractor "D" (No change in difficulty)

• #21 -Added "isolation" to valve noun name in distractor C (No change in difficulty)

• #22 -Bolded word "early" in stem. Beefed up explanations for "C" & "D" * #23 -Resampled Kl A per discussion with NRC from AA2.14 to AA2.04, removed BID from all answer choices. (No change in difficulty)

• #24 -Changed appropriate procedure to AOP-9A (No change in difficulty)

• #25 -Added LP to explanation as why an RO is expected to know ERPIP thresholds.

Changed CET to core exit thermocouple (No change in difficulty).

Changed H2 concentration to 5% * #26-Removed phrase "Using provided reference(s)" (No change in difficulty)

• #28-Added an incorrect interlock to A,C,D so that all answers had 3 parts * #29 -Bolded the word minimum, changed Operability Verification Requirement to STP-0-94A (No change in difficulty)

• #30-Changed distractor C & D to state PZR Level Control Setpoint fails low (Increased difficulty)

• #37-Changed 100 PSIA to 120 PSIA in stem (Lower difficulty).

Made changes to all answer choices to make answer choices "C" & "D" plausible (No change in difficulty)

• #3 8 -Bolded "setpoint" in stem, changed distractor A & B to include BIU htr operation (No change in difficulty), added (s) to Heater for answer choice "A" * #39-Changed stem to remove 5% steam flow change, added explanation to "D" that + 13" required to energize BIU htrs, removed period after 1st bullet. (No change in difficulty), * #42 -Change wording in stem to be more operationally valid, also bolded "setpoint" and "start" (No change in difficulty)

• #43 -Added expected RO knowledge to explanation (No change in difficulty)

Page 6 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY Revision 6 (continued)

• #47-Added statement to explanation that KIA states level ofknowledge required by RO's, added LP #(No change in difficulty)

• #53 -Resampled Kl A per discussion with NRC from Kl.08 to Kl.05, removed comment on info page (No change in difficulty)

• #59 -Changed distractor "A" to 22 BA PP (Lower Difficulty)

• #60-Changed stem from prevent exceeding flammability limit to IAW EOP-5 (Increased difficulty)

• #62 -Added "both SGFP operating" in stem, Changed "22" to "the tripped" in stem. (Lower Difficulty)

• #65-Clarified explanations for answer choices "C" & "D", and added procedure number and noun name. (No change in difficulty)

• #74-Changed "should" to "must" in question stem. (No change in difficulty)

Page 7 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY 1. 007 Reactor Trip -Stabilization

-Recovery U-1 is at 85% power and is ramping to full load when a loss ofthe SOOKV Black Bus occurs. While performing the Core and RCS Heat Removal Safety Function in EOP-0, the CRO notices T cow continuing to lower. Which ONE of the following is the con*ect operator action and the reason for the action? A. Reduce AFW flow due to excess flow from automatic initiation of steam driven pump B. Shut both MSIV's due to the MSR second stage MOV's failing to close on loss of power C. Reduce AFW flow due to excess flow from the automatic start of 13 AFW pump D. Direct ABO to manually shut S/G Feedwater Isolation valves due to loss of power Answer: B Answer Explanation:

A. Incorrect-A loss of the 500 KV Black Bus does not cause excessive AFW flow; (Steam Train). B. Correct-A loss ofthe 500 KV Black Bus will cause the loss ofMCC's 106 & 116 which will cause an excessive cooldown due to MSR 2nd stages not being isolated during the trip. C. Incorrect

-A loss of the 500 KV Black Bus does not cause an automatic start of 13 AFW pump resulting in excessive AFW Flow D. Incorrect-Loss of Black bus will cause loss of 11 4KV bus but since no mention of 1A DG problem Feed Isolation valves can be shut from the CR, also 14 4KV bus never loses power. Page 8 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

KIA Info: RO Importance:

Proposed references to be provided to applicant:

Learning Objective:

10 CFR Part 55 Content: Question source: Cognitive level: Last NRC Exam used on: Technical references:

Comments:

Question 1 (Q95013) U-1 Turbine Trip Alternate Actions 1/1 007 -Reactor Trip

• EK3 -Knowledge of the reasons for the following as they apply to a reactor trip:

• EK3.01 -Actions contained in EOP for reactor trip 4.0 None For each EOP-0 Safety Function:

Analyze and determine correct alternate action to take. 55.41 (b )(1 0) 0Bank I [SJ Modified IDNew D Memory/Fundamental Comprehension/

Analysis No record of use on a previous NRC exam EOP-0, Post-Trip Immediate Actions EOP-0, Post-Trip Immediate Actions Technical Basis None Page 9 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY 2. 008-Pressurizer Vapor Space Accident Given the following conditions:

• Unit-1 at 100% power

• Pressurizer pressure is 2250 PSIA and stable

• PZR backup and proportional heater controls are in AUTO

• 1-HS-100 (PZR pressure control) in the "Y" position

• 1-HS-100-3 (PZR Htr cutoff) in the "X+Y" position

• A steam leak develops on the sensing line for 1-PT-IOOY causing pressure change of 50 PSIA. Which ONE ofthe following describes the PROPORTIONAL HEATER response?

Assume no operator action A. Proportional heaters will operate at approximately 1/3 higher power than before the failure B. Proportional heaters will operate at approximately 1/3 lower power level than before the failure C. Proportional heaters will have maximum power level and the red lights will be illuminated D. Proportional heaters will have maximum power level and the green lights will be illuminated Answer: C Answer Explanation:

A. Incorrect-Proportional heaters will operate at approximately 1/3 higher power than before the failure. Power to the heaters will go to maximum due to a sensed lower pressure.

Normal current output is 50% of max current. B. Incorrect-Proportional heaters will operate at approximately 1/3 lower power level than before the failure. Power to the heaters will go to maximum due to a sensed lower pressure.

Normal current output is 50% of max current. C. Correct -Proportional heaters will have maximum power and the red lights will be illuminated.

Heaters will respond as ifRCS pressure were low. D. Incorrect-Proportional heaters will have maximum power and the red lights will be illuminated.

Green/red lights are function of breaker positions, supply breakers remain shut. Page 10 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

KIA Info: RO Importance:

Proposed references to be provided to applicant:

Learning Objective:

10 CFR Part 55 Content: Question source: Cognitive level: Last NRC Exam used on: Technical references:

Comments:

Question 2 (Q20438) Predict the response of Pressurizer controls to a PT -1 OOY failure 111 008-Pressurizer Vapor Space Accident

• AK2 -Knowledge of the interrelations between the Pressurizer Vapor Space Accident and the following:

• AK2.03 -Controllers and positioners 2.5 None Given a failure of any RCS pressure or temperature instrument predict the response of systems. 55.41 (b )(7) IZ] Bank jo Modified jDNew D Memory/Fundamental Comprehension/Analysis Used in the 7/2002 NRC exam 1C06-ALM, RCS Control Alarm Manual SD-064D, Reactor Coolant Instrumentation System Description None Page 11 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY ,3. 009 -Small Break LOCA U-2 was operating at 100% power in a normal lineup prior to a Loss of Coolant Accident.

Given the following:

• The reactor tripped 10 minutes ago

• RCS pressure is stable at 1270 PSIA

• PZR level is stable at 20 inches

• T COLD is stable at 531 °F

• 24 4KV bus is faulted

• No Operator actions have been taken Which ONE of the following represents indicated net charging flow (charging flow-letdown flow)? A. 44 GPM B. 58 GPM C. 88 GPM D. 132 GPM Answer: C Answer Explanation:

A. Incorrect-21 & 23 Charging Pumps are normally aligned to 21 bus and would both be running resulting a Charging flow of 88 GPM. This flow represents a single Charging Pp in operation which would be the case if23 Charging Pump were aligned to 24 Bus.) B. Incorrect-21 & 23 Charging Pumps aligned to 21 bus and will be running (this flow is for 2 Charging Pumps-min letdown but letdown is isolated by SIAS) C. Correct-21 & 23 Charging Pumps are normally aligned to 21 bus and would both be running resulting in a charging flow of 88 GPM, letdown isolated by SIAS. D. Incorrect-21 & 23 Charging Pumps aligned to 21 bus and will be running (this flow is for 3 Charging Pumps) Page 12 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

KIA Info: RO Importance:

Proposed references to be provided to applicant:

Learning Objective:

10 CFR Part 55 Content: Question source: Cognitive level: Last NRC Exam used on: Technical references:

Comments:

Question 3 {Q97133) Charging Flow during a Small Break LOCA 111 009 -Small Break LOCA

• EA2 -Ability to determine or interpret the following as they apply to a Small Break LOCA:

• EA2.13-Charging pump flow indication 3.4 None 55.4l(b)(8) 0Bank Jo Modified D Memory/Fundamental I tz;J Comprehension/

Analysis N/A-new question EOP-5, Loss of Coolant Accident SD-041, Chemical & Volume Control System AOP-71, Loss of 4KV, 480 Volt or 2081120 Volt Instrument Bus Power None Page 13 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY ,4. 011 -Large Break LOCA What is the purpose of the LOCI Sequencer and which ESF AS signals are required for actuation?

A. Ensure DIG voltage and frequency remain in band during loading; SIAS only B. Ensure ESF loads are all started to mitigate accidents:

SIAS only C. Ensure DIG voltage and frequency remain in band during loading; SIAS &UV D. Ensure ESF loads are all started to mitigate accidents; SIAS &UV Answer: C Answer Explanation:

A. Incorrect-SIAS and UV Actuations are required to trigger operation of the LOCI Sequencer B. Incorrect-Without the LOCI Sequencer all of the ESF loads would start simultaneously.

The purpose of the LOCI Sequencer is to gradually load the DG to maintain voltage and frequency in band. SIAS and UV Actuations are required to trigger operation of the LOCI Sequencer C. Correct -Without the LOCI Sequencer all of the ESF loads would start simultaneously.

The purpose of the LOCI Sequencer is to gradually load the DG to maintain voltage and frequency in band. SIAS and UV Actuations are required to trigger operation of the LOCI Sequencer D. Incorrect-Without the LOCI Sequencer all of the ESF loads would start simultaneously.

The purpose of the LOCI Sequencer is to gradually load the DG to maintain voltage and frequency in band. Page 14 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

KIA Info: RO Importance:

Proposed references to be provided to applicant:

Learning Objective:

10 CFR Part 55 Content: Question source: Cognitive level: Last NRC Exam used on: Technical references:

Comments:

Question 4 (Q97159) DIG basis for LOCA 1/1 011 -Large Break LOCA

• EK3-Knowledge of the reasons for the following responses as they apply to the Large Break LOCA:

• EK3.09-Maintaining DIG's available to provide standby power 4.2 None Define the purpose of the Shutdown and LOCI Sequencers.

55.41(b)(7) 0Bank jo Modified D Memory/Fundamental jiZI Comprehension/

Analysis Nl A -new question SD-048, Engineered Safety Features Actuation System Description None Page 15 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY is. 015/017 -RCP Malfunctions Given the following indications 10 minutes after a reactor trip with no operator actions:

• 11 S/G level -35" and 875 PSIA

• 12 S/G level-20" and 870 PSIA Which ONE of the conditions below (assume no other malfunctions) would cause these indications?

A. 12A RCP locked rotor B. Steam leak from 12 S/G C. Steam leak from 11 S/G D. liB RCP locked rotor Answer: A Answer Explanation:

A. Correct-A locked rotor on 12A RCP causes reduced flow in 12 loop resulting in less heat transfer to 12 S/G. Therefore a slightly lower S/G pressure and a slightly higher S/G level exist in 12 S/G. B. Incorrect-A steam leak from 12 S/G would cause a lower level in 12 S/G. Since 12 S/G level is higher than 11 S/G level this is incorrect.

C. Incorrect

-A steam leak from 11 S/G would cause a lower pressure in 11 S/G. Since 11 S/G pressure is higher than 12 S/G pressure this is incorrect.

D. Incorrect-A locked rotor on liB RCP causes reduced flow in 11 loop resulting in less heat transfer to 11 S/G. Therefore a slightly lower S/G pressure and a slightly higher S/G level would exist in 11 S/G. Page 16 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

KIA Info: RO Importance:

Proposed references to be provided to applicant:

Learning Objective:

10 CFR Part 55 Content: Question source: Cognitive level: Last NRC Exam used on: Technical references:

Comments:

Question 5 (Q97160) Unbalanced RCS flow thermodynamic relationship 111 015/017-RCP Malfunctions

• AK1 -Knowledge ofthe operational implications of the following concepts as they apply to Reactor Coolant Pump Malfunctions (Loss ofRC Flow):

• AK1.04-Basic steady state thermodynamic relationship between RCS loops and S/Gs resulting from unbalanced RCS flow 2.9 None 55.41(b)(5) 0Bank jo Modified _ir:8J New D Memory/Fundamental N/ A -new question None I [8J Comprehension!

Analysis Page 17 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY j6. 025-Loss ofRHR System Given the following:

• Unit-1 is in Mode 6

• Refueling Pool level is 63 Ft. in preparation for U GS removal

• A control circuit malfunction causes 1-SI-652-MOV (SDC HDR RETURN ISOL) to shut

• ALL attempts to reopen 1-SI-652-MOV are unsuccessful Which ONE of the following methods is directed, by procedure, to restore decay heat removal for these given conditions?

A. Use a SFP Cooling Pump to cool the Refueling Pool. B. Initiate Once-Thru-Cooling using a HPSI Pump or LPSI Pump. C. Open the SFP Transfer Tube Gate Valve and cool with the SFP cooling system. D. Fill the Refueling Pool using a HPSI Pump and makeup as inventory is lost to boil-off.

Answer: A Answer Explanation:

A. Correct-Per AOP-3B,Section VII, for common mode loss ofSDC with the RFP that is or can be filled. B. Incorrect-These actions are AOP-3B directions when the RFP is NOT available.

C. Incorrect-The SFP Transfer Gate is verified shut in AOP-3B. The RFP level is lower than the SFP level such that opening the transfer gate would cause a loss of inventory in the SFP. D. Incorrect-Boil-off of the Refueling Pool inventory is an option, but is not directed by AOP-3B for the given conditions.

Page 18 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

KIA Info: RO Importance:

Proposed references to be provided to applicant:

Learning Objective:

10 CFR Part 55 Content: Question source: Cognitive level: Last NRC Exam used on: Technical references:

Comments:

Question 6 (Q92092) Complete loss ofSDC occurs w/RFP level at 63Ft 1/1 025-Loss ofRHR System

• 2.1 -Conduct of Operations

• 2.1.20 -Ability to interpret and execute procedure steps . 4.6 None Using AOP-3B, successfully respond to a loss ofRCS inventory while Shutdown Cooling is in service 55.41 (b )(1 0) [SI Bank Jo Modified jDNew D Memory/Fundamental I fSI Comprehension/

Analysis No record of use on a previous NRC exam AOP-3B, Abnormal Shutdown Cooling Conditions AOP-3B, Abnormal Shutdown Cooling Conditions Technical Basis None Page 19 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY j1. 026-Loss of Component Cooling Water Unit-1 is operating at 100% power. 11 Component Cooling pump is in operation when a Component Cooling leak develops.

The following conditions exist:

• Component Cooling Head tank level is 5" and slowly lowering

• Component Cooling Standby header pressure is 75 PSIG and slowly lowering

• 1C10 alarm "CNTMT NORMAL SUMP LVL HI" is in

• 1C07B alarm "12A RCP CCW FLOW LO" is in

• 12A Thrust Bearing and CBO temperatures are rising Which ONE of the following actions is required based on existing plant conditions?

A. Shut the Component Cooling CV s to Containment, trip the reactor, and secure the RCPs. B. Secure the operating Component Cooling pump due to cavitation caused by low head tank level C. Restore Component Cooling Head tank level by opening the bypass to the makeup CV D. Trip the Reactor since no Component Cooling pumps are operating.

Answer: A Answer Explanation:

A. Correct-Based on the RCP information given, it can be deduced that a CCW leak exists on the 12A RCP. Isolating the leakage by shutting the containment isolation valves requires tripping the reactor then the RCPs once reactivity control safety function is verified.

AOP-7C Step V.C.5 states identify leak location, training is that when identified the leak will be isolated.

B. Incorrect-Based on the standby header pressure, there is adequate pump pressure.

Standby and normal CC header pressures are the same when the CCW system is in a normal alignment and does not depend on 11 or 12 CCW pump in operation.

With a pressure of75 PSIG, there is a CCW pump operating.

C. Incorrect-That a CCW System leak exists is given. The priority would be placed on isolating the leak. Once the leak is isolated, bypassing the make-up CV would only occur if the normal make-up CV was not open. No malfunction of the M/U CV was given in the stem ofthe question.

Since RCP trip criteria will be met on after a loss ofCCW flow, a manual trip would be required once the containment isolation CV s are shut. D. Incorrect

-Based on the standby header pressure, there is adequate pump pressure.

Standby and normal CCW header pressures are the same when the CCW system is in a normal alignment and does not depend on 11 or 12 CCW pump in operation.

With a pressure of75 PSIG, there is a CCW pump operating.

Based on the simulator, the CCW pumps will not show signs of cavitation when head tank level is visible. This assumes that the pump is not in a runout condition, and a standby pressure 75 PSIG does not indicate runout conditions.

Page 20 of 157 Rev.6

.-----------------------------------------------

-! 2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

KIA Info: RO Importance:

Proposed references to be provided to applicant:

Learning Objective:

10 CFR Part 55 Content: Question source: Cognitive level: Last NRC Exam used on: Technical references:

Comments:

Question 7 (Q39973) Response to a CCW leak in Mode 1 1/1 026 -Loss of Component Cooling Water

• AAl -Ability to operate and I or monitor the following as they apply to the Loss of Component Cooling Water:

• AA1.07-Flow rates to the components and systems that are serviced by the CCWS; interactions among the components.

2.9 None Given a scenario and/or plant parameters, and a copy of AOP-7C, determine the required actions. 55.41 (b )(7) IZ! Bank /D Modified /DNew D Memory/Fundamental Comprehension/

Analysis No record of use on a previous NRC exam AOP-7C, Component Cooling System OM-51, U-1 Component Cooling System None Page 21 of 157 Rev. 6 2014 NRC RO EXAM MASTER KEY 8. 027-Pressurizer Pressure Control System Malfunction While operating at NOP/NOT the Pressure transmitter associated with the selected Pressurizer Pressure Control Channel fails to 2210 PSIA. Which ONE of the following statements represents the response of Pressurizer water temperature/level and why? A. Temperature increase I level rises; Proportional heaters provide maximum current B. Temperature increase I level rises; All Pressurizer heaters energized and providing maximum current C. Temperature decrease I level lowers; Pressurizer spray valves open causing saturation temperature to lower D. Temperature decrease I level lowers; All Pressurizer heaters providing minimum current with ambient losses causing temperature to lower Answer: A Answer Explanation:

A. Correct-Sensed pressure is below setpoint therefore proportional heaters fire fully causing temperature in PZR to rise B. Incorrect

-Backup heaters come on at 2200 PSIA, since sensed pressure is above 2200 PSIA the B/U heaters stay off C. Incorrect-Sensed pressure is below setpoint therefore no additional spray (plausible if misinterpret controller response)

D. Incorrect-Sensed pressure is below setpoint therefore heaters energize not secure (plausible if misinterpret controller response)

Page 22 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

KIA Info: RO Importance:

Proposed references to be provided to applicant:

Learning Objective:

10 CFR Part 55 Content: Question source: Cognitive level: Last NRC Exam used on: Technical references:

Comments:

Question 8 (Q97192) Pressurizer response to a rise in RCS temperature 111 027 -Pressurizer Pressure Control System Malfunction

• AKl -Knowledge of the operational implications of the following concepts as they apply to Pressurizer Pressure Control Malfunctions:

• AK1.02-Expansion ofliquids as temperature increases 2.8 None Given conditions and/or parameters associated with the RCS, determine if Pressurizer Heaters and Spray are operating properly 55.41(b)(8) 0Bank jo Modified lrgj New 0 Memory/Fundamental I rgj Comprehension/

Analysis Nl A -new question None Page 23 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY 19. 029-ATWS An Automatic RPS trip signal failed to deenergize trip path relays Kl and K2. What action (if any) is required in accordance with the appropriate procedure to trip the reactor? A. None, trip paths K3 and K4 will de-energize to trip the reactor. B. A Manual trip is required by de-energizing the CEDM MG sets. C. A Manual trip is required by depressing the pushbuttons on both 1 COS and 1 C 15 D. A Manual trip is required by depressing the pushbuttons from either 1C05 or 1 C15 Answer: D Answer Explanation:

A. Incorrect

-Trip Path relays Kl and K2 maintain power to the CEAs, the reactor will not trip. B. Incorrect

-Trip Path relays K1 and K2 maintain power to the CEAs, the reactor will not trip. lAW EOP-0 pushbuttons on 1C05 or 1C15 will be depressed and the reactor will trip so de-energizing CEDM MG sets is not required.

C. Incorrect-Only the pushbuttons at 1 COS or 1 C 15 are required to be depressed to trip the reactor. D. Correct-Only the pushbuttons at 1C05 or 1Cl5 are required to be depressed to trip the reactor. Page 24 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

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Question 9 (Q20175) Effect of trip paths Kl and K2 failing to deenergize on a reactor trip signal. 1/1 029-ATWS

• EK2 -Knowledge of the interrelations between the and the following an ATWS:

• EK2.06 -Breakers, relays, and disconnects 2.9* None Recall the RPS response to a trip signal including the response of: Trip Units, Logic Matrices, Trip Paths and Trip Circuit Breakers.

55.4l(b )(7) r;gj Bank jo Modified jDNew r;gj Memory/Fundamental J 0 Comprehension/Analysis No record of use on a previous NRC exam FSAR figure 7-2, RPS Functional Diagram None Page 25 of 157 Rev.6 r----------------------------------------------1 2014 NRC RO EXAM MASTER KEY !10. 038-Steam Generator Tube Rupture When a Steam Generator tube rupture is suspected, Chemistry is called to perform qualitative samples on both SGs for activity PER CP-436 to confirm the affected S/G. Which ONE of the following actions is performed first? A. Check dose rates on the S/G blowdown piping B. Check dose rates locally on main steam line piping C. Draw a Condenser off-gas sample and analyze for activity D. Draw grab samples from both S/G's and analyze for activity Answer: B Answer Explanation:

A. Incorrect

-This action is not required by CP-436. Blowdown could be secured prior to Chemistry getting readings.

B. Correct -This action is the first specified in CP-436 and can distinguish which S/G contains activity C. Incorrect

-This is not specified in CP-436. It would supply adequate information to confirm a tube leak but could not be used to identify which Steam Generator had the leaking tube( s) D. Incorrect-This is one oflast actions in CP-436, only performed if no results gained from rad readings Page 26 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

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Question 10 (Q97134) SGTR diagnosis from local radiation readings 111 038-Steam Generator Tube Rupture

• EA2 -Ability to determine or interpret the following as they apply to a SGTR:

• EA2.11 -Local radiation readings on main steam lines 3.7* None 55.41 (b )(1 0) 0Bank jo Modified

['g) Memory/Fundamental I 0 Comprehension/

Analysis N/ A -new question EOP-6, Steam Generator Tube Rupture; CP-436, Qualitative Determination of Affected S/G in a Tube Leak Event Applicants are not specifically trained on CP-436 but are trained to expect a report on affected S/G within 15 minutes. Only correct answer would meet this requirement.

Page 27 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY Ill. 054-Loss ofMain Feedwater During a loss of all feedwater event, EOP-3 directs action when __ Steam Generator level(s) reach(es)

<-350 inches to ______ _ A. Both, ensure OTCC in service prior to S/G dryout B. One, ensure OTCC in service prior to S/G dryout C. Both, protect the tubesheet and prevent a primary to secondary leak D. One, protect the tubesheet and prevent a primary to secondary leak Answer: A Answer Explanation:

A. Correct-EOP-3 directs OTCC initiation when BOTH S/G's <-350 inches. EOP-3 Basis states that initiating OTCC prior to S/G dryout helps ensure RCS pressure does not prevent injecting water into the RCS. B. Incorrect-EOP-3 directs OTCC initiation when BOTH S/G's <-350 inches. C. Incorrect-EOP-3 directs OTCC initiation when BOTH S/G's <-350 inches. EOP-3 Basis states that initiating OTCC prior to S/G dryout helps ensure RCS pressure prevents injecting water into the RCS. D. Incorrect-EOP-3 directs OTCC initiation when BOTH S/G's <-350 inches. Page 28 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

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Question 11 (Q97135) Use ofPORVs during a LOAF 1/1 054-Loss of Main Feedwater

• AK3-Knowledge of the reasons for the following responses as they apply to the Loss of Main Feedwater:

• AK3.05-HPI/PORV cycling upon total feedwater loss 4.6 None Given availability of plant equipment, identify the actions required to cool the core 55.41 (b )(1 0) 0Bank jo Modified leg] New D Memory/Fundamental l cg] Comprehension/

Analysis N/ A -new question EOP-3, Loss of All Feedwater; EOP-3, Loss of All Feedwater Technical Basis Document None Page 29 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY !12. 055 -Station Blackout During an extended station blackout event what are the fail positions of the following valves:

• Containment Spray CV's

• Letdown Isolation CV's A. Open, Close B. Open, Open C. Close, Open D. Close, Close Answer: A Answer Explanation:

A. Correct -Fail position of Containment Spray CV's is open to ensure spray flow on loss of air, Fail position of Letdown CV's is close to minimize loss of inventory during an accident concurrent with loss of air. B. Incorrect

-LID CV s fail closed. C. Incorrect-Fail position of Containment Spray CV's is open to ensure spray flow on loss of air, Fail position of Letdown CV's is close to minimize loss of inventory during an accident concurrent with loss of air. D. Incorrect-Fail position of Containment Spray CV's is open to ensure spray flow on loss of air. Students are not expected to memorize the fail position of all air operated valves, therefore this question requires the students to process the purpose of each of these sets of valves and determine which way they should fail to keep plant safe that is reason behind the higher cognitive classification of this question.

Page 30 of 157 Rev. 6 r--------------------------------------------*--1 2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

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Question 12 (Q97136) Extended SBO II A Effects 111 055 -Station Blackout

• EA2-Ability to determine or interpret the following as they apply to a Station Blackout:

• EA2.01 -Existing valve positioning on a loss of instrument air system 3.4 None 55.41 (b )(5) 0Bank jo Modified ji:8J New 0 Memory/Fundamental ji:8J Comprehension/

Analysis N/A-new question OM-35, Unit-1 Main & Reheat Steam OM-73, Unit-1 Chemical & Volume Control System OM-74, Unit-1 Safety Injection

& Containment Spray None Page 31 of 157 Rev.6

2014 NRC RO EXAM MASTER KEY j13. 056 -Loss of Off-Site Power Given both Units operating at 100% power: One second following a complete loss of offsite power what is the status of Reactor Coolant Pumps on both units? A. U-1 RCPs are running, U-2 RCPs are off B. U-1 RCPs are off, U-2 RCPs are running C. All RCPs on both Units, are off D. All RCPs on both Units, are running Answer: D Answer Explanation:

A. Incorrect-All on due to turbine trip circuit delay on U-1 and coast down circuit on U-2. This answer is plausible if student does not remember the coast down ckt on U-2 or equivalent on U-1. B. Incorrect-All on due to turbine trip circuit delay on U-1 and coast down circuit on U-2. This answer is plausible if student remembers the coast down ckt on U-2 but not the equivalent on U-1. C. Incorrect-All on due to turbine trip circuit delay on U-1 and coast down circuit on U-2. This is plausible if studnt gets confused between U-1 & U-2. D. Correct -Unit-1: If the Turbine was paralleled to the grid, the Turbine Generator output breakers are verified open. The breakers are verified open (which directs the operator to open the breakers if they are closed) to protect against a loss of 11 125 VDC Bus, which will prevent the breakers from receiving an open signal. Following a turbine trip, the generator protection circuitry will detect reverse power, causing the Turbine Generator output breakers to trip. This trip will not occur during a turbine warm-up prior to paralleling the Generator to the grid, since the protection circuitry will not detect a reverse power condition.

If offsite power is lost, the protection circuitry will not detect a reverse power condition.

If the Turbine was paralleled to the grid, plant loads will slow the Turbine Generator until generator underfrequency opens breaker 252-11 04, which ultimately deenergizes the generator voltage regulator.

The Turbine Generator output breakers will stay closed until generator volts/hertz or another signal trips the breakers.

This has been estimated to take 12 to 1 7 seconds. If an output breaker does not open following receipt of a trip signal, a protective circuit is energized to open breakers on the 500KV Ring Bus on either side of the Turbine Generator output which is still aligned. Unit-2: -Ifthe Turbine was paralleled to the grid, the Turbine Generator output breakers are checked open following a turbine trip, limit switches in the coastdown initiation circuit close on reverse power causing the generator output breakers to trip. This trip will not occur during a turbine warmup prior to paralleling the Generator to the grid, since the protection circuitry will not detect a reverse power condition.

If an output breaker does Page 32 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY not open following receipt of this trip signal, a protective circuit is energized to prevent turbine and generator damage from resulting from the reversed power condition.

This circuit opens breakers on the 500KV Ring Bus on either side of the turbine generator output that is still aligned. It also opens the 13KV breakers powered from the bus causing a loss of power to the RCPs. If coastdown is initiated the generator output breaker will remain closed until generator voltage degrades to 80% or 20 seconds have elapsed. If coastdown is not initiated or the 20 seconds have elapsed, the operator is directed to open the breakers.

This action protects against a loss of21 125 VDC Bus, which will prevent the breakers from receiving an open signal. Topic: Tier/Group:

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Question 13 (Q97137) RCP status on LOOP 1/1 056 -Loss of Off-site Power

• 2.2 -Equipment Control

• 2.2.3 -Knowledge of the design, procedural, and operational differences between units. 3.8 None 55.4l(b )(5) 0Bank jo Modified Memory/Fundamental I 0 Comprehension!

Analysis N/ A -new question EOP-0, Post-Trip Immediate Actions Technical Basis Page 33 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY jt4. 057 -Loss ofVital AC Inst. Bus U-1 reactor tripped with fault on 11 4KV bus. During EOP-0 the CRS directs tying 1Y09 and IYIO. Which ONE of the following functions is restored and what is its purpose? A. VCT Level switch is re-energized; Allows realigning Charging Pump suction to VCT B. Pressurizer level switch is re-energized; Allows Charging Pumps to cycle off/on to control Pressurizer level C. ADV/TBV controllers are re-energized; Allows RCS temperature control from the Control Room D. PORV/Safety Valve acoustic monitor flow indications are re-energized; Allows diagnosis of PORV /Safety Valve leakage Answer: C Answer Explanation:

A. Incorrect-VCT level switch powered from 1 Yl 0 B. Incorrect-PZR level switch powered from lYlO C. Correct-ADV/TBV controllers are powered from 1 Y09 which was lost in stem of question D. Incorrect-Acoustic monitors are powered from lYlO Page 34 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

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Question 14 (Q97152) Knowledge and purpose components Loss of Vital AC 111 057-Loss ofVital AC Inst. Bus

• 2.1 -Conduct of Operations

• 2.1.28 -Knowledge of the purpose and function of major system components and controls.

4.1 None Evaluate the effect that a loss of the 125V DC/120V AC System will have on 120V AC Non-vital busses 55.41(b)(7) 0Bank jo Modified jr8J New [gl Memory/Fundamental I D Comprehension!

Analysis NIA-new question AOP-7I, Loss of 4KV, 480 Volt or 208/120 Volt Instrument Bus Power None Page 35 of 157 Rev.6 r----------------------------------------------:

2014 NRC RO EXAM MASTER KEY jts. 058-Loss of DC Power Given the following:

• Both Units are operating at 100% power when a Station Blackout occurs.

• 11, 12, 21, and 22 125V DC Bus voltages are approaching 1 05V DC Which ONE of the following Diesel Generator combinations, if any, must be restored to prevent eventual transition to EOP-8? A. None B. 1A; 2A C. 1A; OC on 14 4KV bus D. lB; OC on 24 4KV bus Answer: C Answer Explanation:

A. Incorrect -A Battery Charger must be energized on each of the four 125V DC Vital buses to maintain Bus voltages in the acceptable range. EOP-7 SFSC requires ALL 125V DC Vital Bus voltages are maintained

> 1 05V. B. Incorrect-The combination of two "A" DIGs leaves 12 and 21 125V DC Buses without Battery Chargers.

125V DC Bus voltages for these two Buses would eventually lower to < 1 05V necessitating transition to EOP-8 due to no longer meeting the SFSC C. Correct-The combination of any "A" DIG and the OC DIG on 14 or 24 4KV Bus provides power to a Battery Charger on each of the 125V DC Buses. 125V DC Bus voltages for all Buses would increase towards nominal values. Transition to EOP-8 due to no longer meeting the SFSC D. Incorrect-The combination of any "B" DIG and the OC DIG on 14 or 24 4KV Bus leaves 11 and 22 125V DC Buses without Battery Chargers.

125V DC Bus voltages for these two Buses would eventually lower to < 1 05V necessitating transition to EOP-8 due to no longer meeting the SFSC Page 36 of 157 Rev.6

2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

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Question 15 (Q97161) Loss of DC I Battery Chargers 111 2.8 058-Loss ofDC Power

• AKI -Knowledge of the operational implications of the following concepts as they apply to Loss of DC Power:

• AKl.Ol -Battery charger equipment and instrumentation None Recall the power supplies to the 120V DC Battery Chargers 55.4l(b)(8) 0Bank jDNew 0 Memory/Fundamental J C8J Comprehension/

Analysis Nl A -modified question OI-27D-2, Station Power 480 Volt System Breaker Lineup Modified from Q92882 Page 37 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY 116. 065 -Loss oflnstrurnent Air Under which ONE of the following conditions would manual operation of the Instrument Air Compressors Total Closure valves be required?

A. Removing cornpressor(s) from service (tagging).

B. Compressor operation during a loss of control power to the Unloader valve. C. Manually loading the cornpressor(s) following a total Loss oflnstrurnent Air. D. Operation of a compressor while corning up to normal speed prior to being loaded. Answer: C Answer Explanation:

A. Incorrect-Instrument Air Header pressure is unaffected by simply removing an I/ A Compressor from service. Therefore operation of the total closure valve is not required.

B. Incorrect-OI-19 states: The Total Closure Valve must be operated manually when receiver pressure is less than 30 PSIG. The condition stated does not result in a low I/ A header pressure. Therefore operation of the total closure valve is not required.

C. Correct -OI-19 states: The Total Closure Valve must be operated manually when receiver pressure is less than 30 PSIG. D. Incorrect-OI-19 states: The Total Closure Valve must be operated manually when receiver pressure is less than 30 PSIG. The condition stated does not result in a low I/ A header pressure. Therefore operation of the total closure valve is not required.

Page 38 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

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Question 16 (Q41808) Explain the S/U procedure for an Instrument Air Compressor/apply the general precautions 1/1 065 -Loss of Instrument Air

• AA1-Ability to operate and I or monitor the following as they apply to the Loss of Instrument Air:

• AA 1. 01-Remote Manual Loaders 2.7* None Identify the design features that provide a backup for Instrument Air during a partial or total loss of the system. 55.41(b)(7)

[Z] Bank Jo Modified JONew [Z] Memory/Fundamental J D Comprehension/Analysis No record of use on a previous NRC exam SD-019, Compressed Air 0 I -19, Instrument Air AOP-7D, Loss oflnstrument Air None Page 39 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY 17. 077-Generator Voltage and Electric Grid Disturbances Unit 1 Main Generator is currently carrying 200 MV ARS -LAGGING. Unit 2 Main Generator is shutdown.

A Major grid disturbance has occurred and the System Operator-Bulk Power BP) requests that Calvert Cliffs raise 500KV line voltage from 518 to 521 KV. Which ONE of the following statements is true regarding the effect of this adjustment on MV ARS being carried? A. MV ARS will be unaffected B. MV ARS will change only if MW e changes C. MV ARS will DECREASE to the LEAD direction as line voltage increases D. MVARS will INCREASE in the LAG direction as line voltage increases Answer: D Answer Explanation:

A. Incorrect-Since V ARS are currently in lag direction raising voltage will make them raise further in lag direction B. Incorrect

-V ARS change based on voltage changes, a change in actual load is not required to change V ARS C. Incorrect-Since V ARS are currently in lag direction raising voltage will make them raise further in lag direction D. Correct-Raising voltage will cause lagging V ARS to raise further in the lag direction.

Page 40 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

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Question 17 (Q24852) Turbine Generator controls effect on V ARS during grid disturbance 1/1 077 -Generator Voltage and Electric Grid Disturbances

• AAl -Ability to operate and/or monitor the following as they apply to Generator Voltage and Electric Grid Disturbances:

• AA1.02-Turbine I generator controls 3.8 None Evaluate the effect voltage adjustments have on the main generator, plant, and grid parameters 55.41(b )(5) jo Modified jDNew D Memory/Fundamental No record of use on a previous NRC exam AOP-7M, Major Grid Disturbances None Page 41 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY jts. CE/E05-Excess Steam Demand The following conditions exist on Unit-1 following a Refueling Outage Startup:

• Mode 2 with power being held at 4%

• 11 SGFP in service on Main Steam

• Pressurizer Pressure 2250 PSIA

• Performing Shell warming on the Main Turbine

• An overcooling event occurs and Tcow has reached 521 °F and continues to lower slowly

• The appropriate procedure has been implemented Which ONE of the following sets of actions are procedurally directed to prevent T coLo from continuing to lower? A. Trip the reactor, shut the MSIVs and upstream drains, and initiate auxiliary feedwater.

B. Shut the SG blowdown valves, shut the MSIV upstream drains, and shift steam seals to auxiliary steam. C. Secure main turbine shell warming, shift steam seals to auxiliary steam and initiate auxiliary feedwater.

D. Shut the MSIVs and upstream drains, reduce reactor power to maintain between 2 to 4%, and initiate auxiliary feed. Answer: B Answer Explanation:

A. Incorrect-Trip criteria in AOP-7K is RCS temp <515°, therefore tripping reactor at 521 °F is incorrect B. Correct-These actions are specified in AOP-7K, Step IV.D.4 C. Incorrect-AOP-7K directs tripping turbine vice just securing shell warming and initiating auxiliary feed at existing power level is incorrect as well (power must be lowered <1 %) D. Incorrect-Lowering power to 1% is an action called out in AOP-7K Step IV.D.4, not 2-4% Page 42 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

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Question 18 (Q28505) Determine the appropriate actions for overcooling event in Mode 2 1/1 CE/E05 -Excess Steam Demand

• EK2 -Knowledge of the interrelations between the (Excess Steam Demand)

• EK2.2-Facility's heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operations of these systems to the operation of the facility.

3.7 None Given plant conditions, indications, determine the most appropriate response scenario for overcooling in Modes 1 & 2 55.41(b)(7) rg)Bank Jo Modified IDNew 0 Memory/Fundamental J0 Comprehension!

Analysis No record of use on a previous NRC exam AOP-7K, Overcooling Event In Mode One Or Two None Page 43 of 15 7 Rev.6

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, 2014 NRC RO EXAM MASTER KEY jt9. 032-Loss of Source Range NI U-2 is preparing to parallel the turbine generator to the grid following a MOC trip recovery.

Channel B Wide Range Nuclear Instrument fails low due to a High Voltage Power Supply problem. Prior to any operator actions being performed which ONE of the statements below is correct regarding RPS trip logic? A. 1 out of 3 logic condition for SUR Trip B. 2 out of 3 logic condition for SUR Trip C. 2 out of 4 logic condition for SUR Trip D. SUR Trip bypassed on all4 channels ofRPS Answer: B Answer Explanation:

A. Incorrect-HV PS failure will fail WRNI low, this will bypass the SUR trip on Channel B. Power is <14% so other 3 channels are not bypassed.

The distractor is plausible because this answer would be correct if the PS failure caused the T /U to trip B. Correct-HV PS failure will fail WRNI low, this will bypass the SUR trip on Channel B. Power is <14% so other 3 channels are not bypassed.

Therefore 2 of 3 logic is correct. C. Incorrect-HV PS failure will fail WRNI low, this will bypass the SUR trip on Channel B. Power is <14% so other 3 channels are not bypassed.

This distractor is plausible if student did not remember that low WRNI reading also bypasses SUR trip. D. Incorrect-HV PS failure will fail WRNI low, this will bypass the SUR trip on Channel B. Power is <14% so other 3 channels are not bypassed.

This answer is plausible if student does not recognize Reactor Power is <14% (Parallel turbine is below 14%) Page 44 of 157 Rev. 6 2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

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Question 19 (Q97162) Relationship between WRNI & power supplies 112 032 -Loss of Source Range NI

• AK2 -Knowledge of the interrelations between the Loss of Source Range Nuclear Instrumentation and the following:

• AK2.01-Power supplies, including proper switch positions 2.7* None Recall automatic RPS trips and auto or manual bypass existence for each trip function 55.41 (b )(7) 0Bank Jo Modified 0 Memory/Fundamental j Comprehension/

Analysis N/A-new question SD-058, Reactor Protective System Description None Page 45 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY !20. 036 -Fuel Handling Accident U-2 is in mode 6 performing a fuel shuffle. A core-to-core fuel move is in progress.

While withdrawing a fuel assembly, Chemistry informs the Control Room that boron sample has lowered by 50 ppm unexpectedly.

Which ONE ofthe following actions is required by AOP-IA "Inadvertent Boron Dilution" with respect to the fuel assembly being moved? A. Complete current fuel move B. Leave fuel assembly in its current position C. Reinsert the fuel assembly into its original location D. Complete withdrawing fuel assembly and place it in the RFP upender Answer: D Answer Explanation:

A. Incorrect-AOP-lA, specifically directs placing the fuel assembly in the RFP upender or and approved SFP location.

Completing current move would have you reinsert assembly into core which is incorrect due to SDM concerns.

B. Incorrect-AOP-lA, specifically directs placing the fuel assembly in the RFP upender or and approved SFP location.

Leaving assembly partially inserted into core is incorrect due to SDM concerns.

C. Incorrect-AOP-IA, specifically directs placing the fuel assembly in the RFP upender or and approved SFP location.

Reinserting fuel assembly back in original location is incorrect due to SDM concerns D. Correct-AOP-IA, specifically directs placing the fuel assembly in the RFP upender or and approved SFP location.

Page 46 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

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Question 20 (Q97163) Maintaining SDM during a Fuel handling incident 1/2 036-Fuel Handling Accident

• AKl -Knowledge ofthe operational implications of the following concepts as they apply to Fuel Handling Incidents

• AK1.02-SDM 3.4 None Determine the proper location for a fuel assembly during inadvertent dilution in Modes 3,4,5,6 55.4l(b)(10) 0Bank .I 0 Modified Memory/Fundamental Nl A -new question AOP-6D, Fuel Handling Incident None I 0 Comprehension/Analysis Page 47 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY 21. 060-Accidental Gaseous RadWaste Release An uncontrolled gaseous release to the environment has been detected.

AOP-6C "Accidental Gaseous Waste Release" has been implemented.

Which valve(s) are directed to be verified shut in this condition?

A. WGDT Outlet valves B. Purification IX inlet valves C. PS-5464-CV RCS Sample Isolation valve D. Letdown Filter Inlet and Outlet valves Answer: A Answer Explanation:

All valves listed are in AOP-6C but only the WGDT valves are in the section for uncontrolled release to the environment.

Distractors B,C & Dare all in the High Airborne in the Aux Bldg section. Upper cognitive because student must determine which valves could result in release to environment.

A. Correct -See above B. Incorrect

-See above C. Incorrect

-See above D. Incorrect-See above Page 48 of 157 Rev.6

.-----------------------------------------------------------------------------------1 2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

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Question 21 (Q97164) Accidental Gaseous Release valve lineup 1/2 060 -Accidental Gaseous RadWaste Release

• AA2 -Ability to determine and interpret the following as they apply to the Accidental Gaseous Radwaste:

• AA2.06-Valve lineup for release of radioactive gases 3.6* None Given conditions, determine if the waste gas system is operating properly 55.41 (b )(1 0) 0Bank lD Modified D Memory/Fundamental I Comprehension!

Analysis N/A-new question AOP-6C, Accidental Gaseous Waste Release None Page 49 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY !22. 061-ARM System Alarms The Wide Range Noble Gas Monitor (RIC-5415) would provide early detection of radioactivity during which ONE of the following events? A. 40 GPM Leak on the VCT B. 40 GPM RCS cold leg leak C. 40 GPM Letdown HX to CC leak D. 40 GPM Spent Fuel Pool HX to SRW leak Answer: A Answer Explanation:

A. Correct-A leak on the VCT would be monitored by the WRNGM since the Auxiliary Building Ventilation System discharges thru the WRNGM B. Incorrect-An RCS Cold Leg leak would be confined to the Containment, therefore not monitored by the WRNGM. C. Incorrect-Early indication of a LDHX tube leak would be provided by the CCW system inline RMS. Eventually, the WRNGM may detect radioactivity as a result of the CC Head tank being vented to the Auxiliary Building Atmosphere.

D. Incorrect-Early indication of a SFP HX leak would be provide by the SRW system inline RMS. Eventually, the WRNGM may detect radioactivity as a result of the SRW Head tank being vented to the Auxiliary Building Atmosphere.

Page 50 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

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10 CFR Part 55 Content: Question source: Cognitive level: Last NRC Exam used on: Technical references:

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Question 22 (Q24708) The Wide Range Noble Gas Monitor (RIC-5415) would detect radioactivity during which accident?

1/2 061 -ARM System Alarms

• 2.1 -Conduct of Operations

• 2.1.27 -Knowledge of system purpose and/or function 3.9 None Recall the purpose of the WRNGM system 55.41(b )(7) jo Modified jDNew D Memory/Fundamental Comprehension/

Analysis No record of use on a previous NRC exam OI-35, Radiation Monitoring System SD-077/079, Radiation Monitoring System None Page 51 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY 123. 067 -Plant Fire On-site A severe fire has occurred in the Control Room. Which ONE of the following groups of systems/components can be electrically isolated at the Remote Shutdown Panel (112C43) to isolate the fire effects? A. Letdown, RCS Sampling, Controlled Bleed-Off B. Letdown, Controlled Bleed-Off, Atmospheric Dump Valves C. Controlled Bleed-Off, RCS Sampling, Auxiliary Feed Water D. RCS Sampling, Atmospheric Dump Valves, Auxiliary Feed Water Answer: A Answer Explanation:

A. Correct-Per AOP-9A these components are isolated from the effects of the fire through operator actions taken at the Remote Shutdown Panel B. Incorrect-Per AOP-9A these components, with the exception of ADVs, are isolated from the effects of the fire through operator actions taken at the Remote Shutdown Panel C. Incorrect-Per AOP-9A these components, with the exception of Auxiliary Feedwater and the SRW HXs, are isolated from the effects of the fire through operator actions taken at the Remote Shutdown Panel D. Incorrect-Per AOP-9A these components, with the exception of the ADVs and Auxiliary Feedwater, are isolated from the effects of the fire through operator actions taken at the Remote Shutdown Panel Page 52 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

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Question 23 (Q97140) Fire Suppression Activities 112 067 -Plant Fire On-site

• AA2 -Ability to determine and interpret the following as they apply to the Plant Fire on Site:

• AA2.04-The fire's extent ofpotential operational damage to plant equipment 3.1 None Recall the actions performed on plant systems to prevent or minimize the loss ofRCS inventory and/or depressurization 55.41 (b )(1 0) 0Bank jo Modified Memory/Fundamental I 0 Comprehension/

Analysis N/ A -new question AOP-9A-Control Room Evacuation And Safe Shutdown Due To A Severe Control Room Fire None Page 53 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY 124. 068 -Control Room Evacuation A severe fire has occurred in the Control Room concurrent with a loss of offsite power. Which buses are energized and from what source of power using AOP-9A, Control Room Evacuation and Safe Shutdown due to a Severe Control Room Fire? A. 14 4KV bus from the OC DIG; 21 4KV bus from the 2A DIG B. 11 4 KV bus from the 1A DIG; 24 4KV bus from the OC DIG C. 11 4KV bus from the OC DIG; 24 4KV bus from the OC DIG D. 14 4KV bus from the OC DIG; 21 4KV bus from the OC DIG Answer: C Answer Explanation:

A. Incorrect-AOP-9A-112 directs aligning the OC DIG to 4KV Buses 11 & 24 B. Incorrect-AOP-9A-112 directs aligning the OC DIG to 4KV Buses 11 & 24 C. Correct-AOP-9A-112 directs aligning the OC DIG to 4KV Buses 11 & 24 D. Incorrect-AOP-9A-1/2 directs aligning the OC DIG to 4KV Buses 11 & 24 Page 54 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

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Question 24 (Q97141) 4KV Busses powered during a CR Evacuation 1/2 068 -Control Room Evacuation

• AK2 -Knowledge of the interrelations between the Control Room Evacuation and the following:

• AK2.07-ED/G 3.3 None Identify why and how 480v load centers are stripped 55.41 (b )(7) 0Bank jo Modified Memory/Fundamental J D Comprehension/

Analysis N/ A -new question AOP-9A-Control Room Evacuation And Safe Shutdown Due To A Severe Control Room Fire None Page 55 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY 125. Loss of Containment Barrier per ERPIP Which ONE of the following conditions would be considered a loss/potential loss ofthe containment barrier per the Emergency Action Level Threshold for classification?

A. Core Exit Thermocouple reaches 700°F B. Containment hydrogen concentration of 5% C. S/G tube leak with a loss of condenser vacuum D. Containment pressure peaks at 47 PSIG and slowly lowers Answer: B Answer Explanation:

A. Incorrect-Per ERPIP 3.0 (EAL-Hot), Fission Product Barrier Matrix, the criteria for a Potential Loss of the Containment Barrier is > 1200°F for > 15 minutes B. Correct-Per ERPIP 3.0 (EAL-Hot), Fission Product Barrier Matrix, the criteria for a Potential Loss of the Containment Barrier is Hydrogen 4% C. Incorrect-Per ERPIP 3.0 (EAL-Hot), Fission Product Barrier Matrix, the criteria for a Potential Loss of the Containment Barrier is that the ruptured S/G is also faulted D. Incorrect-Per ERPIP 3.0 (EAL-Hot), Fission Product Barrier Matrix, the criteria for a Potential Loss of the Containment Barrier is a Containment pressure rise that is followed by a rapid unexplained depressurization RO's are not expected to memorize the EAL chart but they are expected to analyze each of the given conditions and determine which one challenges the containment as a fission product barrier, therefore higher cognitive.

At CCNPP it is an expectation that RO's have an understanding ofEAL's to aid SM in recognizing when and EAL criteria is met. Page 56 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

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Question 25 (Q97142) Loss of Containment Barrier per ERPIP 112 069 -Loss of Containment Integrity

• 2.4 -Emergency Procedures I Plan

• 2.4.41 -Knowledge ofthe emergency action level thresholds and classifications 2.9 None Lesson Plan# LOI-217-3 55.41 (b )(1 0) 0Bank jo Modified D Memory/Fundamental Comprehension/

Analysis NIA-new question ERPIP 3.0, Emergency Response Plan Implementation Procedure None Page 57 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY 126. 076-High Reactor Coolant Activity U-1 is currently escalating power to 100% following a Refueling Outage. The following conditions exist:

• The unit is currently at 50% reactor power * "RAD MON L VL HI" alarm at 1 C07 annunciated and chemistry has validated the alarm

• AOP-6A has been implemented in response to high RCS Activity.

• Chemistry reports the following values:

• Dose Equivalent I-131 = 82 uCi/gm

• Gross Activity=

115 uCi/gm

• E-Bar = 0.805 Based on these values, which ONE ofthe following sets of statements is true? A. Gross Activity is acceptable; Dose Equivalent I-131 is in the unacceptable region of the figure B. Gross Activity is acceptable; Dose Equivalent I-131 is high in the acceptable region C. Gross Activity is high; Dose Equivalent I -131 is in the unacceptable region of the figure D. Gross Activity is high; Dose Equivalent I-131 is high in the acceptable region Answer: B Answer Explanation A. Incorrect-Per TS 3.4.15 the gross activity is SAT since 100/E-bar limit is and specific activity limit is 90 f.!Cilgm I-131 B. Correct-Per TS 3.4.15 the gross activity is SAT since 100/E-bar limit is and specific activity limit is 90 ll Cil gm I-131 C. Incorrect-Per TS 3.4.15 the gross activity is SAT since 100/E-bar limit is 124 f.!Ci/gm and specific activity limit is 90 f.!Ci/gm I-131 D. Incorrect-Per TS 3.4.15 the gross activity is SAT since 100/E-bar limit is 124 J!Cilgm and specific activity limit is 90 I-131 Page 58 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

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Question 26 (Q51251) Recall parameters monitored for entry into AOP 6A 112 076 -High Reactor Coolant Activity

• AA1 -Ability to operate and I or monitor the following as they apply to the High Reactor Coolant Activity:

• AA1.04-Failed fuel-monitoring equipment 3.2 T.S. 3.4.15 and associated Bases Given plant conditions and/or values associated with RCS Activity, and using appropriate reference( s ), evaluate the condition and determine required actions IA W the Technical Specifications.

55.41 (b )(7) IZJ Bank !o Modified jDNew D Memory/Fundamental IIZJ Comprehension/

Analysis No record of use on a previous NRC exam AOP-6A, Abnormal Reactor Coolant Chemistry/

Activity T.S. 3.4.15 and associated Bases None Page 59 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY 127. CE/All -RCS Overcooling U-1 is returning to full power operation following Main Turbine Valve Testing, when the following occurred:

• Turbine Bypass Valve, 1-MS-3940-CV failed open.

• AOP-7K, Overcooling Event, has been implemented.

Which ONE of the following sets of actions/reasons is appropriate, to initially stabilize the plant, for the stated conditions?

A. Turbine load is lowered, and CEA's are withdrawn to restore Tcow to program B. The TBV is manually isolated to restore TcoLD to program C. Turbine load is lowered to restore T coLD to program D. Withdraw CEA's to restore Tcow to program Answer: C Answer Explanation:

A. Incorrect

-CEA withdrawal is not in accordance with the direction provided for stabilizing the plant. CEAs would be used to adjust reactor power, if necessary, once RCS temperature has been stabilized if power was continuing to lower. B. Incorrect-Manual isolation of the affected TBV is an action performed, in a controlled sequence, ONCE the plant has been stabilized.

C. Correct -Per AOP-7K, restoring T coLD to its program value is the appropriate actions for placing the plant in a stable condition.

The reasons stated for the action is correct. D. Incorrect

-CEA withdrawal is not in accordance with the direction provided for stabilizing the plant. CEAs would be used to adjust reactor power, if necessary, once RCS temperature has been stabilized if power was continuing to lower Page 60 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

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10 CPR Part 55 Content: Question source: Cognitive level: Last NRC Exam used on: Technical references:

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Question 27 (Q51252) Response to an Overcooling Event 1/2 CE/ All -RCS Overcooling

• AK3-Knowledge of the reasons for the following responses as they apply to the (RCS Overcooling)

• AK3.3 -Manipulation of controls required to obtain desired operating results during abnormal, and emergency situations 3.1 None Given plant conditions and/or indications determine the most appropriate response for an overcooling event in Modes 1 or 2. 55.41(b )(1 0) [8] Bank Jo Modified jDNew 0 Memory/Fundamental I [g) Comprehension/Analysis No record of use on a previous NRC exam AOP-7K, Overcooling Event In Mode One Or Two None Page 61 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY 128. 003 -Reactor Coolant Pump System Which ONE of the following describes ALL of the interlocks that must be satisfied to start a Reactor Coolant Pump? A. Oil lift pressure, RCP Oil level, sufficient CCW flow B. Oil lift pressure, RCP Oil level, Synchronizing stick inserted C. RCP Oil level, sufficient CCW flow, Synchronizing stick inserted D. Oil lift pressure, sufficient CCW flow, Synchronizing stick inserted Answer: D Answer Explanation:

A. Incorrect-RCP Oil level, while alarmed, does not provide an interlock in the RCP starting circuit. Additionally, the Synchronizing stick must be inserted.

B. Incorrect

-RCP Oil level, while alarmed, does not provide an interlock in the RCP starting circuit. Additionally, the CCW flow switches must be made up. C. Incorrect

-RCP Oil level, while alarmed, does not provide an interlock in the RCP starting circuit. Additionally, the Oil lift pump must be running for initial start of the RCP. D. Correct-The Synchronizing stick must be inserted, the CCW flow switch must be made up and the Oil lift pump discharge pressure switch must be made up for initial start of the RCP. Page 62 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

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Question 28 (Q92714) RCP Start Interlocks 2/1 003-Reactor Coolant Pump System

• K4-Knowledge ofRCPS design feature(s) and/or interlock(s) which provide for the following:

• K4.04-Adequate cooling ofRCP motor and seals 2.8 None Indicate the interlocks required to be satisfied before an RCP may be started 55.4l(b)(7) k8J Bank Jo Modified JDNew k8J Memory/Fundamental J 0 Comprehension!

Analysis No record of use on a previous NRC exam 01-lA, Reactor Coolant System and Pump Operations None Page 63 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY 129. 003-Reactor Coolant Pump System U-2 tripped due to a SGFP coupling problem, and is being held in Mode 3 awaiting repairs. Which ONE of the sets of conditions below is the minimum required to meet STP-0-94A-2, "Coolant Loop Operability Verification (Mode 3)"? A. 2 RCP's operating in each loop with both S/G levels >-40 inches B. 1 RCP operating in each loop with both S/G levels >-40 inches C. 2 RCP's operating in one loop with associated S/G level >-40 inches D. 1 RCP operating in one loop with associated S/G level >-40 inches Answer: B Answer Explanation:

A. Incorrect-STP-0-94A, Coolant Loop Operability Verification (Mode 3) requires an RCP operating in each loop with both S/G levels >-40 inches B. Correct-STP-0-94A, Coolant Loop Operability Verification (Mode 3) requires an RCP operating in each loop with both S/G levels >-40 inches C. Incorrect-STP-0-94A, Coolant Loop Operability Verification (Mode 3) requires an RCP operating in each loop with both S/G levels >-40 inches D. Incorrect-STP-0-94A, Coolant Loop Operability Verification (Mode 3) requires an RCP operating in each loop with both S/G levels >-40 inches Page 64 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

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Question 29 (Q97143) Knowledge ofRCP surveillance procedures 2/1 003-Reactor Coolant Pump System

• 2.1 -Equipment Control

• 2.1.2 -Knowledge of surveillance procedures 3.7 None Given plant conditions and/or values associated with the RCS Loops, and using appropriate reference(s), evaluate the condition and determine required actions IA W the Technical Specifications.

55.4l(b )(1 0) 0Bank jo Modified Memory/Fundamental l D Comprehension/

Analysis N/ A -new question STP-0-94A, Coolant Loop Operability Verification (Mode 3) None Page 65 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY 130. 004 -Chemical and Volume Control When collapsing the PZR bubble with RCS solid, which ONE of the following sets of (1) abnormal conditions and (2) consequences due to this abnormal condition would be true? A. (1) Backpressure Regulator setpoint fails low; (2) RCS Pressure lowers B. (1) Backpressure Regulator setpoint fails low; (2) RCS Pressure rises C. (1) Selected Pressurizer Level Controller setpoint fails low; (2) RCS Pressure lowers D. (1) Selected Pressurizer Level Controller setpoint fails low; (2) RCS Pressure rises Answer: A Answer Explanation:

A. Correct -If Backpressure regulator setpoint fails low the backpressure is higher than setpoint so backpressure control valves open to lower pressure to setpoint.

B. Incorrect

-If Backpressure regulator setpoint fails low then the backpressure is higher than setpoint so backpressure control valves open to lower pressure to setpoint.

C. Incorrect-The Letdown control valves are in the manual mode of control and are full open with the plant in this condition.

No control function occurs D. Incorrect-The Letdown control valves are in the manual mode of control and are full open with the plant in this condition.

No control function occurs Page 66 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

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Question 30 (Q97144) RCS Pressure control using CVCS 211 004-Chemical and Volume Control

• K6 -Knowledge of the effect of a loss or malfunction on the following eves components:

• K6.26 -Methods of pressure control of solid plant (PZR relief and water inventory) 3.8 None Recall how to collapse the Pressurizer Bubble. 55.4l(b)(7) 0Bank lo Modified D Memory/Fundamental Comprehension/

Analysis N/ A -new question OP-7, Shutdown Operations None Page 67 of 157 Rev. 6 2014 NRC RO EXAM MASTER KEY 131. 004-Chemical and Volume Control A Loss of Coolant Accident has occurred on U-2 which resulted in a Containment peak pressure of 5 PSIG. The CRS has directed that you lower subcooling to the lower end of the band. Which ONE of the following actions is required by EOP-5 "Loss of Coolant Accident" under these conditions?

A. Open Aux Spray Control Valve, 2-CVC-517-CV; Operate Charging Header Loop Stops 2-CVC-518

& 519-CV's as necessary B. Place the Pressurizer Spray Valve Controller, 2-HIC-100, to manual; Raise the output to open the main spray valve to desired amount C. Open the Atmospheric Steam Dump Valves by placing 2-HIC-4056, in manual; Operate the ADV's by adjusting controller output to obtain the desired result D. Place Turbine Bypass Valve Controller, 2-PIC-4056, in manual; Lower the output to obtain the desired result Answer: A Answer Explanation:

A. Correct -Per EOP-5, Auxiliary Spray will be initiated by opening CVC-517 and shutting CVC-518 & 519. B. Incorrect-Raising the output of2-HIC-100 would normally result in lowering subcooling IF the RCPs were running. RCPs would have been secured when CIS actuated at 2.8 PSIG. C. Incorrect-Raising the output of the ADV controller would lower RCS temperature resulting in an increase in RCS subcooling D. Incorrect-For the conditions stated CSAS will have actuated, closing the MSIVs. The TBVs will not be available for control ofRCS subcooling Page 68 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

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Question 31 (Q97165) Actions to control RCS subcooling 2/1 004-Chemical and Volume Control

• A4-Ability to manually operate and/or monitor in the control room:

• A4.09 -PZR spray and heater controls 3.5 None Given RCS parameters, identify the appropriate response for a Loss of Coolant Accident (LOCA) per EOP-5. 55.41 (b )(7) 0Bank jo Modified 0 Memory/Fundamental I Comprehension!

Analysis Nl A -new question EOP-5, Loss of Coolant Accident None Page 69 of 157 Rev.6

,------------------------------------------**I 2014 NRC RO EXAM MASTER KEY 132. 005 -Residual Heat Removal Which ONE ofthe following groups of instruments are designed for Post-Accident Monitoring?

1. S/G Narrow Range level, LI-1111/1121

2. PZR Low Range Pressure, PI-103/103-1

3. Containment Narrow Range Pressure, PI-5308 4. PZR Level, LI-110X/110Y A. 1, 2, 3 B. 1, 3, 4 C. 1, 2, 4 D. 2, 3,4 Answer: D Answer Explanation:

A. Incorrect-All instruments listed except S/G Narrow Range level are PAMS B. Incorrect-All instruments listed except S/G Narrow Range level are PAMS C. Incorrect-All instruments listed except S/G Narrow Range level are PAMS D. Correct-All instruments listed except S/G Narrow Range level are PAMS Page 70 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

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Question 32 (Q97145) RHR Post Accident Instrumentation 211 005 -Residual Heat Removal

• 2.4 -Emergency Procedures I Plan

• 2.4.3 -Ability to identify post-accident instrumentation 3.7 None Identify P AMS inputs and outputs. 55.41(b )(6) 0Bank Jo Modified J!Z! New [2J Memory/Fundamental N/ A -new question None I 0 Comprehension/

Analysis Page 71 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY 133. 005 -Residual Heat Removal Which ONE of the following describes (1) adequate mixing flow with SDC in service; and (2) the basis for this requirement?

A. (1) 3000 GPM SDC flow with PZR manway removed (2) Prevents dilute pockets in the RCS B. (1) 1500 GPM SDC flow with PZR manway removed (2) Maintains adequate Shutdown Margin C. (1) 3000 GPM SDC flow with S/G tubes full (2) Prevents dilute pockets in the RCS D. (1) 1500 GPM SDC flow with S/G tubes full (2) Maintains adequate Shutdown Margin Answer: C Answer Explanation:

OI-2B and OP-7 define adequate mixing as: "At least 3000 GPM flow through the core AND at least 500 GPM flow through both S/Gs." 500 GPM through both S/G's is not assured without a bubble in the PZR. The basis for these limits is to prevent dilute pockets. A. Incorrect

-With PZR manway removed there is no chance of a bubble in the PZR B. Incorrect

-1500 GPM does not meet minimum flow and maintaining SDM is not the basis for requirement C. Correct -3000 GPM meets the limit and S/G tubes full also meets requirement and basis is to prevent dilute pockets D. Incorrect-1500 GPM does not meet minimum flow and maintaining SDM is not the basis for requirement Page 72 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

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Question 33 (Q97166) Conditions permitting dilution of the RCS 211 005 -Residual Heat Removal

• KS -Knowledge of the operational implications of the following concepts as they apply the RHRS:

• K5.03-Reactivity effects ofRHR fill water 2.9* None Identify RCS dilution limitations including requirements for adequate mixing. 55.4l(b)(5) 0Bank jo Modified jr8J New IZ! Memory/Fundamental I 0 Comprehension/

Analysis N/ A -new question OI-2B, CVCS Boration, Dilution And Makeup Operations OP-7, Shutdown Operations None Page 73 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY 134. 006-Emergency Core Cooling U -1 is operating at 1 00% power when an invalid SIAS "A" actuates.

Which ONE of the following statements describe components operated and how SIAS is reset with minimal impact on plant operation?

A. Aux HPSI Header MOV's open; Match all handswitches on the SIAS verification checklist and depress the SIAS reset pushbutton on 1 C 1 0 B. All SRW Turbine Bldg Header Isolation CV's shut; Match all handswitches on the SIAS verification checklist and depress the SIAS reset pushbutton on 1 C 1 0 C. Aux HPSI Header MOV's open; Match asterisked handswitches on the SIAS verification checklist and depress the SIAS reset pushbutton on 1 C 1 0 D. All SRW Turbine Bldg Header Isolation CV's shut; Match asterisked handswitches on the SIAS verification checklist and depress the SIAS reset pushbutton on 1 Cl 0 Answer: C Answer Explanation:

A. Incorrect-The Aux HPSI Hdr MOVs open on SIAS "A", however only the handswitches annotated by an asterisk in Attachment (2) of the EOP Attachments must be matched to reset SIAS from lClO. Matching all handswitches on the SIAS verification checklist would essentially initiate a complete SIAS "A" & "B" having a significant effect on power operation (boration

& isolation ofSRW to the Turbine Building).

B. Incorrect-Only 2 SRW TB isolations shut on SIAS "A" not al14, and only the handswitches annotated by an asterisk in Attachment (2) of the EOP Attachments must be matched to reset SIAS from 1C10 C. Correct-The Aux HPSI Hdr MOVs open on SIAS "A" and only the handswitches annotated by an asterisk in Attachment (2) of the EOP Attachments must be matched to reset SIAS from 1C10 D. Incorrect-Only 2 SRW TB isolations shut on SIAS "A" not all4, and only the handswitches annotated by an asterisk in Attachment (2) of the EOP Attachments must be matched to reset SIAS from 1C10 Page 74 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

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Question 34 (Q97167) Invalid SIAS impact on plant ops 2/1 006 -Emergency Core Cooling

• A2-Ability to (a) predict the impacts of the following malfunctions or operations on the ECCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

• A2.13 -Inadvertent SIS actuation 3.9 None Evaluate plant conditions/parameters and detem1ine conditions required to reset an ESF AS signal, either remotely or locally. 55.4l(b)(5) 0Bank .l D Modified D Memory/Fundamental I Comprehension/

Analysis N/ A -new question 1 C08-ALM, ESF AS 11 Alarm Manual EOP Attachments, Attachment (2) None Page 75 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY 135. 007 -Pressurizer Reliefi'Quench Tank OP-7, Shutdown Operations, requires the Quench Tank be prepared for drawing a Pressurizer bubble. Current Quench Tank oxygen concentration is 5%. What is the maximum oxygen concentration allowed in the Quench Tank and how is it reduced, IA W OI-l B, Quench Tank Operations?

A. 4%; Cycle QT pressure between 0.5 PSIG and 10.0 PSIG 12 times by cycling QT vent and N 2 Supply one at a time. B. 4%; Open QT vent and throttle open N 2 Supply to maintain pressure between 0.5 PSIG and 1.5 PSIG for 15 minutes. C. 2%; Cycle QT pressure between 0.5 PSIG and 10.0 PSIG 12 times by cycling QT vent and Nz Supply one at a time. D. 2%; Open QT vent and throttle open N 2 Supply to maintain pressure between 0.5 PSIG and 1.5 PSIG for 15 minutes. Answer: B Answer Explanation:

A. Incorrect-OI-lB Sect. 6.11 requirements for drawing PZR bubble state that 4% is the maximum oxygen concentration therefore C & Dare incorrect.

OI-lB Sect 6.10 states that to lower concentration you purge for 15 minutes vice cycling 12 times. B. Correct-OI-lB Sect. 6.11 requirements for drawing PZR bubble state that 4% is the maximum oxygen concentration.

OI-l B Sect 6.10 states that to lower concentration you purge for 15 minutes. C. Incorrect-OI-lB Sect. 6.11 requirements for drawing PZR bubble state that 4% is the maximum oxygen concentration therefore C & Dare incorrect.

OI-lB Sect 6.10 states that to lower concentration you purge for 15 minutes vice cycling 12 times. D. Incorrect-OI-lB Sect. 6.11 requirements for drawing PZR bubble state that 4% is the maximum oxygen concentration therefore C & D are incorrect.

OI-l B Sect 6.10 states that to lower concentration you purge for 15 minutes. Page 76 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

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Question 35 (Q97168) Quench tank 0 2 limit for operability 2/1 007-Pressurizer Relief/Quench Tank

• K5 -Knowledge of the operational implications of the following concepts as the apply to PRTS:

• K5.02-Method of forming a steam bubble in the PZR 3.1 None Identify the requirements for drawing a Pressurizer bubble and venting the RCS. 55.41(b)(5) 0Bank jo Modified Memory/Fundamental N/ A -new question OI-IB, Quench Tank Operations OP-7, Shutdown Operations None I 0 Comprehension/

Analysis Page 77 of 15 7 Rev.6 2014 NRC RO EXAM MASTER KEY 136. 008 -Component Cooling Water Which ONE ofthe following conditions require controlled restoration of Component Cooling flow to the containment following a station blackout and why? A. Controlled Bleed-off temperature>

250°F; RCP seals need to be rebuilt B. Controlled Bleed-off temperature>

250°F; Thermal shock could cause an RCS leak C. Lower Seal temperature

> 280°F; Thermal shock could cause an RCS leak D. Lower Seal temperature

> 280°F; RCP seals need to be rebuilt Answer: C Answer Explanation:

A. Incorrect

-This condition does require RCP Seals be rebuilt prior to starting the affected pump, however, this condition does not prohibit the controlled restoration of CCW flow to the Containment B. Incorrect-The primary concern with exceeding a CBO temperature of250°F is that it results in the need to rebuild the seal package prior to operation of the pump. C. Correct-The primary concern with exceeding a Lower Seal temperature of280°F is that uncontrolled restoration of CCW flow could result in a thermal shock to the seals causing a seal LOCA. Controlled restoration of CCW is done to prevent thermal shock to the RCP seals and seal coolers resulting from the initiation of cooling water at elevated component temperatures D. Incorrect-The primary concern with exceeding a Lower Seal temperature of 280°F is that it could result in a seal LOCA. Page 78 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

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Question 36 (Q97146) CCW Setpoints, Interlocks, Auto Actions for EOP Entry 211 008 -Component Cooling Water

• 2.4 -Emergency Procedures I Plan

• 2.4.6 -Knowledge of EOP mitigation strategies 3.7 None Given RCP CBO and Lower Seal temperatures, identify how Component Cooling is restored to the RCPs. 55.41 (b )(7) 0Bank jo Modified ji:8J New D Memory/Fundamental I I:8J Comprehension!

Analysis N/ A -new question EOP-7, Station Blackout, Step "IV.R" None Page 79 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY 137. 008-Component Cooling Water You suspect that a Shutdown Cooling Heat Exchanger tube has ruptured while on Shutdown Cooling. RCS pressure is 120 PSIA. Which ONE of the following indications could be used to confirm a Shutdown Cooling Heat Exchanger tube leak/rupture?

A. Lowering Component Cooling head tank level; Rising PZR level; No change in Aux Building radiation levels. B. Rising Component Cooling head tank level; Lowering PZR level; Rising Aux Building radiation levels. C. Lowering Component Cooling head tank level; Rising PZR level; Rising Aux Building radiation levels. D. Rising Component Cooling head tank level; Lowering PZR level; No change in Aux Building radiation levels. Answer: B Answer Explanation:

A. Incorrect

-RCS pressure is higher than Component Cooling Header Press (SDC HX cooling medium) which would result in RCS leakage into the CCW System. Pressurizer level would lower and Aux Bldg Radiation levels would rise B. Correct -RCS pressure is higher than Component Cooling Header Press (SDC HX cooling medium) which would result in RCS leakage into the CCW System. Pressurizer level would lower and Aux Bldg Radiation levels would rise C. Incorrect

-RCS pressure is higher than Component Cooling Header Press (SDC HX cooling medium) which would result in RCS leakage into the CCW System. Pressurizer level would lower and Aux Bldg Radiation levels would rise D. Incorrect

-RCS pressure is higher than Component Cooling Header Press (SDC HX cooling medium) which would result in RCS leakage into the CCW System. Pressurizer level would lower and Aux Bldg Radiation levels would rise Page 80 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

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Question 37 (Q97183) RCS Leak on a SDCHX 2/1 008-Component Cooling Water

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

• Al.04-Surge tank level 3.1 None Using AOP-3B be able to successfully respond to a loss ofRCS inventory while Shutdown Cooling is in use. 55.41 (b )(5) 0Bank jo Modified 0 Memory/Fundamental I Comprehension!

Analysis No record of use on a previous NRC exam AOP-3B, Abnormal Shutdown Cooling Conditions, Step V.l 0 AOP-2A, Excessive Reactor Coolant Leakage, Step IX.5.1 None Page 81 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY 138. 010-Pressurizer Pressure Control Given the following:

• Unit-1 is at 100% power

• RCS Pressure Control is in AUTO

• RCS Pressure is 2250 PSIA Which ONE of the following describes the IMMEDIATE plant response if the setpoint for the selected Pressurizer Pressure Controller fails low? A. Spray valve controller goes to maximum output; Backup Heaters energize; Actual RCS pressure will rise. B. Spray valve controller goes to minimum output Backup Heaters remain deenergized Actual RCS pressure will lower. C. Spray valve controller goes to minimum output; Proportional Heaters output goes to maximum; Actual RCS pressure will rise. D. Spray valve controller goes to maximum output; Proportional Heaters output goes to minimum; Actual RCS pressure will lower. Answer: D Answer Explanation:

A. Incorrect

-Setpoint is lower than sensed press so the controller is trying to lower RCS pressure.

Spray goes to maximum, Backup heater unaffected, pressure will lower. B. Incorrect-Setpoint is lower than sensed press so the controller is trying to lower RCS pressure.

Spray goes to maximum, Backup heaters unaffected, actual RCS pressure will lower. C. Incorrect-Setpoint is lower than sensed press so the controller is trying to lower RCS pressure.

Spray goes to maximum, Proportional Heaters goes to minimum, actual RCS pressure will lower. D. Correct-Setpoint is lower than sensed press so the controller is trying to lower RCS pressure.

Spray goes to maximum, Proportional Heaters goes to minimum, actual RCS pressure will1ower.

Page 82 of 157 Rev. 6 2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

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Question 38 (Q97172) Selected Pzr Pressure Control Channel failure 2/1 010-Pressurizer Pressure Control

• K3-Knowledge of the effect that a loss or malfunction of the PZR PCS will have on the following:

• K3.01-RCS 3.8 None Given a failure, of any RCS pressure or temperature instrument, predict the response of systems to the failure. 55.41 (b )(7) 0Bank I [g) Modified IDNew D Memory/Fundamental Comprehension/

Analysis N/ A -modified question 1 C06-ALM, RCS Control Alarm Manual Modified from Q92862 Page 83 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY 139. 010-Pressurizer Pressure Control The reactor is operating at 80% power, late in core life for previous 5 days, when the following occur:

• A turbine control valve drifts shut

• Pressurizer level changes by 9 inches due to transient Which ONE of the following describes the initial response of the Pressurizer Spray valves and Pressurizer Heaters? A. Spray valves shut, Backup Heaters energize.

B. Spray valves shut, Backup Heaters remain deenergized.

C. Spray valves open, Backup Heaters energize.

D. Spray valves open, Backup Heaters remain deenergized.

Answer: D Answer Explanation:

A. Incorrect-Lowering steam demand causes Reactor Coolant System temperature to rise with a resulting increase in RCS pressure.

The Pressurizer Pressure Control System responds to deenergize Pressurizer Heaters and opens the Pressurizer Spray valves to control the rising RCS pressure B. Incorrect-Lowering steam demand causes Reactor Coolant System temperature to rise with a resulting increase in RCS pressure.

The Pressurizer Pressure Control System responds to deenergize Pressurizer Heaters and opens the Pressurizer Spray valves to control the rising RCS pressure C. Incorrect-Lowering steam demand causes Reactor Coolant System temperature to rise with a resulting increase in RCS pressure.

The Pressurizer Pressure Control System responds to deenergize Pressurizer Heaters and opens the Pressurizer Spray valves to control the rising RCS pressure D. Correct-Lowering steam demand causes Reactor Coolant System temperature to rise with a resulting increase in RCS pressure.

The Pressurizer Pressure Control System responds to deenergize Pressurizer Heaters and opens the Pressurizer Spray valves to control the rising RCS pressure, Backup heaters don't automatically energize until+ 13" on PZR lvl. Page 84 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

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Question 39 (Ql4311) Predict response of Pzr to change in steam demand 2/1 0 1 0 -Pressurizer Pressure Contro 1 A3-Ability to monitor automatic operation of the PZR PCS, including:

A3.02-PZR pressure 3.6 None Given conditions and/or parameters, associated with the RCS, determine if Pressurizer Heaters and Spray are operating properly 55.41 (b )(7) jo Modified jDNew 0 Memory/Fundamental I rgj Comprehension/

Analysis No record of use on a previous NRC exam 1C06-ALM, RCS Control Alarm Manual SD-064D, Reactor Coolant System Instrumentation System Description None Page 85 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY 140. 012 -Reactor Protection Which ONE of the following statements describes a function of the Matrix Relay Hold Pushbutton?

A. Ensures continuity from sensor to trip unit B. Applies a test voltage to relays K-1 through K-4 C. Supplies power to test the Trip Circuit Breaker Shunt and UV coils for testing. D. Prevents logic matrix relay contact from opening with normal power removed. Answer: D Answer Explanation:

A. Incorrect-The Matrix Relay Hold Pushbutton supplies power to maintain the individual matrix relays (AB, AC, AD, etc .... ) powered to prevent trip relays K-1 thru K-4 from opening (and opening their respective TCBs) during testing. B. Incorrect

-The Matrix Relay Hold Pushbutton does not directly supply power to relays K-1 thru K-4. C. Incorrect

-The Matrix Relay Hold Pushbutton does not supply power to the TCB Shunt and UV Trip coils. D. Correct-The Matrix Relay Hold Pushbutton supplies power to maintain the individual matrix relays (AB, AC, AD, etc .... ) powered to prevent trip relays K-1 thru K-4 from opening (and opening their respective TCBs) during testing. Page 86 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

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Question 40 (Q20201) Function of the Matrix Relay Hold Pushbutton 2/1 012 -Reactor Protection

• K4-Knowledge ofRPS design feature(s) and/or interlock(s) which provide for the following:

• K4.08 -Logic matrix testing 2.8* None Recall the function of the Matrix Relay Hold Pushbutton 55.41 (b )(7) [gj Bank jo Modified jDNew [gj Memory/Fundamental I D Comprehension/Analysis No record of use on a previous NRC exam SD-058, Reactor Protective System None Page 87 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY 41. 013-Engineered Safety Features Actuation U-1 was operating at 100% power when the 11 Main Feed Reg Valve goes shut concurrent with a loss of 11 120V Vital AC Bus (1Y01). 11 S/G water level is now (-)200 inches. Which ONE of the following is the most complete list of components actuated?

A. Only MS-4070-CV opened B. MS-4070-CV and MS-4071-CV opened C. MS-4070-CV opened and 13 AFW pump started D. MS-4070-CV and MS-4071-CV opened and 13 AFW pump started Answer: A Answer Explanation:

A. Correct -The loss of 11 120V Vital AC Bus results in the loss of power to the AF AS "A" Actuation Logic Cabinet and the AF AS ZD Sensor Cabinet; AF AS "A" will not be actuated.

AF AS "B" Actuation Logic Cabinet remains powered and will actuate AF AS "B" at a S/G level of (-) 170 inches on 2/4 AF AS channels for at least 20 seconds. Only components receiving an AF AS "B" signal (1-MS-4070-CV) will actuate. B. Incorrect

-The loss of 11 120V Vital AC Bus results in the loss of power to the AF AS "A" Actuation Logic Cabinet and the AF AS ZD Sensor Cabinet; AF AS "A" will not be actuated.

AF AS "B" Actuation Logic Cabinet remains powered and will actuate AF AS "B" at a S/G level of (-) 170 inches on 2/4 AF AS channels for at least 20 seconds. Only components receiving an AFAS "B" signal (1-MS-4070-CV) will actuate. 1-MS-4071-CV receives its actuation signal from the AF AS "A" logic cabinet. C. Incorrect

-The loss of 11 120V Vital AC Bus results in the loss of power to the AF AS "A" Actuation Logic Cabinet and the AFAS ZD Sensor Cabinet; AFAS "A" will not be actuated.

AF AS "B" Actuation Logic Cabinet remains powered and will actuate AF AS "B" at a S/G level of(-) 170 inches on 2/4 AFAS channels for at least 20 seconds. Only components receiving an AFAS "B" signal (1-MS-4070-CV) will actuate. 13 AFW Pp receives its actuation signal from the AF AS "A" logic cabinet. D. Incorrect-The loss of 11 120V Vital AC Bus results in the loss of power to the AFAS "A" Actuation Logic Cabinet and the AFAS ZD Sensor Cabinet; AFAS "A" will not be actuated.

AF AS "B" Actuation Logic Cabinet remains powered and will actuate AF AS "B" at a S/G level of(-) 170 inches on 2/4 AF AS channels for at least 20 seconds. Only components receiving an AFAS "B" signal (1-MS-4070-CV) will actuate. 13 AFW Pp and 1-MS-4071-CV receive their actuation signals from the AFAS "A" logic cabinet. Page 88 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

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10 CFR Part 55 Content: Question source: Cognitive level: Last NRC Exam used on: Technical references:

Comments:

Question 41 (Q97153) ESF AS loss or malfunction on sensors detectors 2/1 013-Engineered Safety Features Actuation

• K6-Knowledge of the effect of a loss or malfunction on the following will have on the ESFAS:

• K6.01 -Sensors and detectors 2.7* None Without reference, explain the effect on AF AS operation when a sensor and/or logic channel is deenergized.

55.41 (b )(7) 0Bank Jo Modified D Memory/Fundamental Comprehension/Analysis N/A-new question SD-036, Auxiliary Feedwater 1 C04-ALM, Aux Feedwater and Computer Alarm Manual None Page 89 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY 42. 022 -Containment Cooling Given an event causing Containment Pressure and RCS Pressure to degrade at which ONE of the following pressure setpoints would you expect Containment Air Coolers to automatically start? ( 1) Containment Pressure (2) RCS Pressure A. (1) 2.8 PSIG; (2) 1725 PSIA B. (1) 2.8 PSIG; (2) 1740 PSIA C. (1) 4.25 PSIG; (2) 1725 PSIA D. (1) 4.75 PSIG; (2) 1740 PSIA Answer: B Answer Explanation:

A. Incorrect-Containment Cooler Fans are started by SIAS. SIAS is actuated by a high Containment Pressure of 2.8 PSIG or a low RCS pressure of 1740 PSIA. B. Correct -Containment Cooler Fans are started by SIAS. SIAS is actuated by a high Containment Pressure of2.8 PSIG or a low RCS pressure of 1740 PSIA. C. Incorrect-Containment Cooler Fans are started by SIAS. SIAS is actuated by a high Containment Pressure of 2.8 PSIG or a low RCS pressure of 1740 PSIA. 4.25 PSIG is the setpoint for CSAS actuation which starts the Containment Spray Pumps. 1725 PSIA is the minimum RCS pressure SIAS must be actuated by per the Tech Specs. D. Incorrect-Containment Cooler Fans are started by SIAS. SIAS is actuated by a high Containment Pressure of2.8 PSIG or a low RCS pressure of1740 PSIA. 4.75 PSIG is the maximum Containment pressure SIAS must be actuated by per the Tech Specs. 1725 PSIA is the minimum RCS pressure SIAS must be actuated by per the Tech Specs. Page 90 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

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Question 42 (Q97169) Match component actuation to ESF AS signals 2/1 022-Containment Cooling

• A3 -Ability to monitor automatic operation of the CCS, including:

• A3.01 -Initiation of safeguards mode of operation 4.1 None Recall operation of ESF AS that includes setpoints 55.41 (b )(7) 0Bank !o Modified Memory/Fundamental J D Comprehension/

Analysis Nl A -new question 1C08-ALM None Page 91 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY 43. 022 -Containment Cooling What are the Technical Specification maximum Containment Pressure setpoint limits for the actuation signals that start the following components?

(1) Containment Spray Pumps start (2) Containment Spray CV's open A. (1) 2.8 PSIG; (2) 4.25 PSIG B. (1) 2.8 PSIG; (2) 4. 75 PSIG C. (1) 4.25 PSIG; (2) 4.25 PSIG D. (1) 4.75 PSIG; (2) 4.75 PSIG Answer: D Answer Explanation:

A. Incorrect-2.8 & 4.25 PSIG are the actual setpoints for SIAS (starts the Containment Spray Pumps) and CSAS (opens the Containment Spray CVs). The question asks for the Tech Spec limits for actuation signals which are both a Containment pressure of 4.75 PSIG. B. Incorrect-2.8 PSIG is the actual setpoint for SIAS (starts the Containment Spray Pumps). 4.75 PSIG is the Tech Spec limit for CSAS (opens the Containment Spray CVs). The question asks for the Tech Spec limits for actuation signals which are both a Containment pressure of 4.75 PSIG. C. Incorrect-4.25 PSIG is the actual setpoint for CSAS which opens the Containment Spray CVs. Per T.S. 3.3.4, Table 3.3.4-1, the SIAS Tech Spec limit 4.75 PSIG and the CSAS Tech Spec limit 4.75 PSIG D. Correct -The Containment Spray Pumps are started by SIAS and the Containment Spray Header CVs are opened by CSAS. Per T.S. 3.3.4, Table 3.3.4-1, the SIAS Tech Spec limit 4.75 PSIG and the CSAS Tech Spec limit is:::: 4.75 PSIG Knowledge of Tech Spec setpoints is clearly communicated as an expectation to RO's during License Class. Page 92 of 157 Rev. 6 2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

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10 CFR Part 55 Content: Question source: Cognitive level: Last NRC Exam used on: Technical references:

Comments:

Question 43 (Q97154) Containment Cooling auto operation on containment pressure 2/1 022 -Containment Cooling

• AI -Ability to monitor automatic operation of the CCS, including:

• Al.02-Containment pressure 3.6 None Evaluate plant conditions and/or parameters for RPS and ESFAS and determine compliance with Technical Specification LCO 3.3.4 55.41(b)(5) 0Bank .l cg) Modified JDNew k8J Memory/Fundamental N/ A -modified question T.S. 3.3.4, Table 3.3.4-1 Modified version of Q20788 I 0 Comprehension/

Analysis Page 93 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY 44. 026 -Containment Spray During power operation a LOCA occurred.

SIAS, CSAS and RAS have actuated.

Which ONE of the following functions are the Shutdown Cooling Heat Exchangers performing at this time? A. Providing cooling of Containment Spray. B. Providing a Shutdown Cooling flowpath.

C. Providing High Pressure Safety Injection cooling. D. Providing Low Pressure Safety Injection cooling. Answer: A Answer Explanation:

A. Correct -RAS realigns CCW flow to the SDC Heat Exchanger to cool Containment Spray Pump discharge flow. B. Incorrect-Use of the SDC Heat Exchangers as part of the Shutdown Cooling flowpath requires manual re-alignment of the LPSI Pumps. C. Incorrect

-HPSI flow does not pass through the SDC Heat Exchangers.

D. Incorrect-LPSI Pumps are secured automatically when RAS actuates.

Use ofSDC Heat Exchangers in support ofLPSI cooling requires manual re-alignment of the LPSI Pump flowpaths and starting the LPSI Pumps. Page 94 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

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1 0 CFR Part 55 Content: Question source: Cognitive level: Last NRC Exam used on: Technical references:

Comments:

Question 44 (Ql7833) SDC Heat Exchanger purpose during a LOCA 211 026 -Containment Spray

• Kl -Knowledge of the physical connections and/or cause-effect relationships between the CSS and the following systems:

• Kl.02 -Cooling water 4.1 None State the purpose and function of the Containment Spray System/Components.

55.4l(b)(8) jo Modified jDNew Memory/Fundamental j D Comprehension/

Analysis Used in NRC exam administered 12/2008 SD-052 & 061, Safety Injection and Containment Spray Systems None Page 95 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY 45. 039 -Main and Reheat Steam Given the following:

• A S/G Tube Rupture and Loss of Offsite Power occurred

• EOP-0 has been completed and the appropriate Optimal Recovery Procedure has been implemented.

Which ONE ofthe following describes the preferred method for cooldown of the RCS and when can the affected S/G be isolated?

A. TBV's; TcoLD <515oF B. ADV's; T COLD <515"F C. TBV's; T HOT <515°F D. ADV's; THOT <515°F Answer: D Answer Explanation:

A. Incorrect-Because of the Loss ofOffsite Power, the MSIVs will have been shut during the performance ofEOP-0 making the TBVs unavailable to cooldown the RCS. Additionally, EOP-6 requires the RCS be cooled down to aT HOT of <515°F. This action reduces the risk of challenging the steam generator safety valves of the affected S/G after it is isolated.

B. Incorrect-Because of the Loss ofOffsite Power, the MSIVs will have been shut during the performance ofEOP-0 necessitating the use of the ADVs to cooldown the RCS. Additionally, EOP-6 requires the RCS be cooled down to a THoT of <51YF. This action reduces the risk of challenging the steam generator safety valves of the affected S/G after it is isolated.

C. Incorrect

-Because of the Loss of Offsite Power, the MSIV s will have been shut during the performance ofEOP-0 making the TBVs unavailable to cooldown the RCS. Additionally, EOP-6 requires the RCS be cooled down to a T HOT of <515 OF. This action reduces the risk of challenging the steam generator safety valves of the affected S/G after it is isolated.

D. Correct-Because of the Loss ofOffsite Power, the MSIVs will have been shut during the performance of EOP-0 making the TBV s unavailable to cool down the RCS. Additionally, EOP-6 requires the RCS be cooled down to a THoT of <515°F. This action reduces the risk of challenging the steam generator safety valves of the affected S/G after it is isolated.

Page 96 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

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Comments:

Question 45 (Q97147) Cooldown during a SGTR 2/1 039-Main and Reheat Steam

• Kl -Knowledge of the physical connections and/or cause-effect relationships between the MRSS and the following systems:

• Kl.02-Atmospheric relief dump valves 3.3 None Recall the strategy and basis for the major actions performed in EOP-6Steam Generator Tube Rupture, and what actions are required if Safety Functions are in jeopardy of being lost. 55.41(b)(7) 0Bank jo Modified ji:8J New 0 Memory/Fundamental I C8J Comprehension/

Analysis N/A-new question EOP-6, Steam Generator Tube Rupture None Page 97 of 157 Rev. 6 2014 NRC RO EXAM MASTER KEY 46. 059-Main Feedwater Unit-2 is escalating in power, recovering from a mid-cycle forced outage. The reactor is at approximately 50% power with 21 SGFP in operation.

22 SGFP is out of service for maintenance.

Under these conditions, which ONE of the following sets of parameters on 21 SGFP would support a decision to continue operation at current power level? A. Suction flow: 15,200 GPM; Turbine speed: 5060 RPM; Suction pressure:

242 PSIG B. Suction flow: 16,200 GPM; Turbine speed: 5080 RPM; Suction pressure:

262 PSIG C. Suction flow: 17,200 GPM; Turbine speed: 5335 RPM; Suction pressure:

272 PSIG D. Suction flow: 18,200 GPM; Turbine speed: 5260 RPM; Suction pressure:

252 PSIG Answer: B Answer Explanation:

Per the Precautions stated in OI-12A-2 one pump operation is permitted if all of the following conditions are maintained:

• Suction flow rate below 18,000 GPM

• Turbine speed below 5100 RPM

• Suction pressure above 250 PSIG A. Incorrect

-Suction pressure is below the stated operating limit of 250 PSIG B. Correct-Per the Precautions stated in OI-12A-2 C. Incorrect

-SGFP speed is above the stated operating limit of 5100 RPM Plausible because speed is within the operating range for a Unit-1 SGFP D. Incorrect-SGFP speed is above the stated operating limit of5100 RPM and SGFP suction flow rate is also above the limit for single pump operation Plausible because speed is within the operating range for a Unit-1 SGFP Page 98 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

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Comments:

Question 46 (Q97170) Operating limitations on U-2 SGFPs 2/1 059-Main Feedwater

• AI -Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the MFW controls including:

• Al.03-Power level restrictions for operation ofMFW pumps and valves 2.7* None Given plant conditions and/or parameters determine if SGFP operating limitations are exceeded.

55.41 (b )(5) 0Bank jo Modified New IZJ Memory/Fundamental N/ A -new question OI-12A-l, Feedwater System OI-12A-2, Feedwater System None j D Comprehension/

Analysis Page 99 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY 47. 061 -Auxiliary/Emergency Feedwater Which ONE of the following describes the cooling/lubrication methods for (1) the Steam Driven AFW Pump bearing and (2) the AFW Pump Turbine bearings and (3) Conditions needed to auto start AFW pumps? A. (1) Air cooled rotating slinger ring; (2) Air cooled rotating slinger ring (3) -170 inches for seconds on 2 of 4 channels on either S/G B. (1) Air cooled rotating slinger ring; (2) AFW cooled rotating slinger ring supplemented by forced oil flow (3) -170 inches for seconds on 2 of 4 channels on either S/G C. (1) AFW cooled rotating slinger ring supplemented by forced oil flow; (2) Air cooled rotating slinger ring (3) -170 inches for seconds on 2 of 4 channels on both S/G's D. (1) AFW cooled rotating slinger ring supplemented by forced oil flow; (2) AFW cooled rotating slinger ring supplemented by forced oil flow (3) -170 inches for seconds on 2 of 4 channels on both S/G's Answer: B Answer Explanation:

A. Incorrect

-Per SD-036: the SDAFW Pp bearings are cooled and lubricated by an air cooled rotating slinger ring and the SDAFW Turbine Bearings are cooled by AFW flow and lubricated by a rotating slinger ring supplemented by forced oil flow from a saddle mounted pump. . -170" 2 of 4 channels on either S/G generates AF AS which starts AFW Pumps B. Correct -Per SD-036: the SDAFW Pp bearings are cooled and lubricated by an air cooled rotating slinger ring and the SDAFW Turbine Bearings are cooled by AFW flow and lubricated by a rotating slinger ring supplemented by forced oil flow from a saddle mounted pump. -1 70" 2 of 4 channels on either S/G generates AF AS which starts AFW Pumps C. Incorrect

-Per SD-036: the SDAFW Pp bearings are cooled and lubricated by an air cooled rotating slinger ring and the SDAFW Turbine Bearings are cooled by AFW flow and lubricated by a rotating slinger ring supplemented by forced oil flow from a saddle mounted pump. -170" 2 of 4 channels on either S/G generates AF AS which starts AFW Pumps D. Incorrect

-Per SD-036: the SDAFW Pp bearings are cooled and lubricated by an air cooled rotating slinger ring and the SDAFW Turbine Bearings are cooled by AFW flow and lubricated by a rotating slinger ring supplemented by forced oil flow from a saddle mounted pump. -170" 2 of 4 channels on either S/G generates AF AS which starts AFW Pumps. KIA clearly states that candidates are expected to know cooling and lubrication methods for AFW pumps Page 100 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

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Question 47 (Q97148) AFW design feature cooling and lubrication 211 061 -Auxiliary/Emergency Feedwater

• K4-Knowledge of AFW design feature(s) and/or interlock(s) which provide for the following:

• K4.13 -Initiation of cooling water and lube oil 2.7 None LOI-036A-12 55.41 (b )(7) 0Bank lD Modified Memory/Fundamental N/ A -new question SD-036, Auxiliary Feedwater None I D Comprehension!

Analysis Page 101 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY 48. 062 -AC Electrical Distribution U-1 is in a normal electrical lineup with exception of 1 Y02 has been shifted to the Inverter up bus, 1 Y11. Subsequently, a loss of the Red Bus occurs. Which ONE of the following statements describe the effect on plant equipment?

A. 11 4kV bus remains energized from P-13000-1; 1 Y11 remains powered from MCC-114; 1 Y02 remains energized.

B. 1 Y11 is de-energized due to the loss of MCC-1 04; 1B DG starts and automatically repowers 14 4kV Bus; 1 Y02 is re-energized.

C. 1 Y11 is de-energized due to the loss of MCC-114; 1A DG starts and automatically repowers 11 4kV Bus; 1 Y02 is re-energized.

D. 1 Y11 is de-energized due to the loss of P-13000-2; 1B DG must be manually started and loaded; 1 Y02 is re-energized.

Answer: D Answer Explanation:

A. Incorrect-1 Y11 (inverter backup bus) is powered from MCC 104, normally supplied by 14 4kV bus through P-13000-2 via the red bus. B. Incorrect

-May be selected if it is not recognized that the ESF AS B logic cabinet has been deenergized.

C. Incorrect

-Could be selected if there is confusion on components supplied by Red Bus/Black bus. DGs do not automatically repower 4KV Busses when the Inverter B/U Bus is in service and lost. D. Correct-1 Y11 (inverter backup bus) is powered from MCC 104, normally supplied by 14 4kV bus through P-13000-2 via the red bus. The 1B DG would not start because the ESF AS "B" logic cabinet is deenergized (results in no U/V start signal actuation).

Manually starting the 1B DG is needed to reenergize MCC-104 which reenergizes 1Yll and restores power to 1 Y02. Page 102 of 157 Rev. 6

,--------------------------------------------1 2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

KIA Info: RO Importance:

Proposed references to be provided to applicant:

Learning Objective:

10 CFR Part 55 Content: Question source: Cognitive level: Last NRC Exam used on: Technical references:

Comments:

Question 48 (Q92672) Loss of red bus when 1 Y02 is aligned to inverter backup bus 2/1 062 -AC Electrical Distribution

• K3 -Knowledge of the effect that a loss or malfunction of the ac distribution system will have on the following:

• K3.02-ED/G 4.1 None Evaluate the effect that a loss of the 125V DC /120V AC System will have on Emergency Diesel Generator operation.

55.41(b)(7)

C8J Bank jo Modified jDNew 0 Memory/Fundamental Comprehension/Analysis No record of use on a previous NRC exam None Page 103 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY 49. 063 -DC Electrical Distribution Given the following conditions:

• BOTH Units are operating at 100% power

• 1C33 Alarm Window, T-20, "125V DC Ground Detected" is illuminated Which ONE of the following describes the DC Busses potentially affected by this alarm, and the indication observed on the bus that has a ground? A. DC Busses 11, 12, 21, and 22; Red LED on Ground Detection Panel will be ILLUMINATED.

B. DC Busses 11, 12, 21, and 22; Red LED on Ground Detection Panel will be EXTINGUISHED.

C. DC Busses 11 and 21 ONLY; Red LED on Ground Detection Panel will be EXTINGUISHED.

D. DC Busses 11 and 21 ONLY; Red LED on Ground Detection Panel will be ILLUMINATED.

Answer: D Answer Explanation:

A. Incorrect

-Only 125VDC buses 11 and 21 are monitored by Annunciator window "T-20" and Red LED on Ground Detection Panel will be ILLUMINATED when a ground is present. B. Incorrect-Only 125VDC buses 11 and 21 are monitored by Annunciator window "T-20" and Red LED on Ground Detection Panel will be ILLUMINATED when a ground is present. C. Incorrect-Only 125VDC buses 11 and 21 are monitored by Annunciator window "T-20" and Red LED on Ground Detection Panel will be ILLUMINATED when a ground is present. D. Correct-Only 125VDC buses 11 and 21 are monitored by Annunciator window "T-20" and Red LED on Ground Detection Panel will be ILLUMINATED when a ground is present. Page 104 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

KIA Info: RO Importance:

Proposed references to be provided to applicant:

Learning Objective:

10 CFR Part 55 Content: Question source: Cognitive level: Last NRC Exam used on: Technical references:

Comments:

Question 49 (Q74605) Monitor and operate the 125 VDC, 120 VAC, and instrument AC Systems 2/1 063 -DC Electrical Distribution

• A3-Ability to monitor automatic operation of the DC electrical system, including:

• A3.01 -Meters, annunciators, dials, recorders, and indicating lights 2.7 None Recall basic operation of the 125V DC ground detection system. 55.41 (b )(7) [SI Bank Jo Modified JONew [SI Memory/Fundamental I 0 Comprehension/

Analysis No record of use on a previous NRC exam 1 C33 Alarm Response Manual None Page 105 of 157 Rev. 6 2014 NRC RO EXAM MASTER KEY 50. 064 -Emergency Diesel Generator Safety Related Diesel Generator auxiliaries get power from which of the following buses? Diesel Generator Bus IA 1) MCC-104 IB 2) MCC-114 2A 3) MCC-122 2B 4) MCC-123 5) MCC-204 6) MCC-214 A. 1, 2, 5, 6 B. 2, 1, 6, 5 c. 4, 1, 6, 5 D. 3, 2, 5, 6 Answer: C Answer Explanation:

A. Incorrect-1A DIG Auxiliaries are powered from MCC-123, 1B DIG Auxiliaries are powered from MCC 1BG through MCC-104 2A DIG Auxiliaries are powered from MCC 2AG through MCC-214 2B DIG Auxiliaries are powered from MCC 2BG through MCC-204 B. Incorrect-1A DIG Auxiliaries are powered from MCC-123, 1B DIG Auxiliaries are powered from MCC 1BG through MCC-104 2A DIG Auxiliaries are powered from MCC 2AG through MCC-214 2B DIG Auxiliaries are powered from MCC 2BG through MCC-204 C. Correct-1A DIG Auxiliaries are powered from MCC-123, 1B DIG Auxiliaries are powered from MCC 1BG through MCC-104 2A DIG Auxiliaries are powered from MCC 2AG through MCC-214 2B DIG Auxiliaries are powered from MCC 2BG through MCC-204 D. Incorrect-1A DIG Auxiliaries are powered from MCC-123, 1B DIG Auxiliaries are powered from MCC 1BG through MCC-104 2A DIG Auxiliaries are powered from MCC 2AG through MCC-214 2B DIG Auxiliaries are powered from MCC 2BG through MCC-204 Page 106 of 157 Rev. 6 2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

K/A Info: RO Importance:

Proposed references to be provided to applicant:

Learning Objective:

1 0 CFR Part 55 Content: Question source: Cognitive level: Last NRC Exam used on: Technical references:

Comments:

Question 50 (Q97149) Knowledge of DIG auxiliary power supplies 2/1 064 -Emergency Diesel Generator

• K2 -Knowledge of bus power supplies to the following:

• K2.02 -Fuel oil pumps 2.8* None Identify the power supplies to the Emergency Diesel Generator control circuits and auxiliary equipment.

55.4l(b)(7) 0Bank Jo Modified JC8J New C8J Memory/Fundamental J 0 Comprehension/

Analysis N/ A -new question OI-27D-2, Station Power 480 Volt System Breaker Lineup None Page 107 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY 51. 073 -Process Radiation Monitoring While discharging 11 WGDT to U-2, which ONE of the following conditions would allow the discharge to continue?

A. Critical high computer alarm 2R5415A! (U-2 Main Vent RMS) B. Alert high computer alarm O-R2191! (Waste Gas RMS) C. 12 WGDT low pressure computer alarm actuates D. Running U-2 Main Exhaust fan trips Answer: B Answer Explanation:

A. Incorrect-OI-17B Waste Gas Sect. 6.4 "Waste Gas Release "specifically requires securing release immediately if an RMS critical setpoint is exceeded.

B. Correct-OI-17B Waste Gas Sect. 6.4 "Waste Gas Release" directs entering setpoints lAW Waste Gas discharge permit. Alert Alarm is just a warning and discharge can continue.

C. Incorrect-OI-17B Waste Gas Sect. 6.4 "Waste Gas Release" directs setting low pressure alarm for WGDT's not being discharged to 5 psig below current pressure.

If 12 WGDT low pressure alarm comes in then the wrong WGDT may be discharging and release must be secured. D. Incorrect

-Waste Gas release permit directs Main Vent Exhaust fan running, if the fan tripped off there is no auto start of other fan so discharge must be secured. Page 108 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

KIA Info: RO Importance:

Proposed references to be provided to applicant:

Learning Objective:

10 CPR Part 55 Content: Question source: Cognitive level: Last NRC Exam used on: Technical references:

Comments:

Question 51 (Q97150) RMS operate/monitor from control room during effluent release 2/1 073 -Process Radiation Monitoring

• A4-Ability to manually operate and/or monitor in the control room:

• A4.01 -Effluent release 3.9 None Determine when a Waste Gas Decay Tank can be isolated and/or released.

55.41 (b )(7) 0Bank jo Modified 0 Memory/Fundamental I Comprehension/

Analysis N/A-new question OI-17B Waste Gas System Student must analyze 4 different conditions and determine if the release can continue.

Therefore this question is considered upper cognitive level Page 109 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY 52. 073 -Process Radiation Monitoring U-1 commenced a fuel shuffle 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> ago with containment purge in service when an RMS alarm occurs. Investigation shows 1-RI-5316C (Containment Area) is saturated-pegged high. Which ONE of the following describes the impact on the plant and the actions required to complete the fuel shuffle? A. Containment Purge automatically isolates; Pull fuses for RMS then reset RMS after inserting fuse B. Containment Purge automatically isolates; Place detector in Level Cal and continue to move fuel with the 3 remaining detectors C. Containment Purge continues to operate; Pull fuses for RMS then reset RMS after inserting fuse D. Containment Purge continues to operate; Place detector in Level Cal and continue to move fuel with the 3 remaining detectors Answer: C Answer Explanation:

A. Incorrect-Containment Radiation Signal (CRS) is actuated by 2 of 4 Containment area rad monitors, which would isolate Containment Purge. Only one rad monitor fails high therefore no securing of Containment Purge. OI-35 Radiation Monitoring System Sect 6.2 "Normal Operation" directs removing fuse for saturated detector and resetting detector after inserting fuse. No need to secure fuel handling based on a single pegged high area rad monitor. B. Incorrect-Containment Radiation Signal (CRS) is actuated by 2 of 4 Containment area rad monitors, which would isolate Containment Purge. Only one rad monitor fails high therefore no securing of Containment Purge. OI-35 Radiation Monitoring System Sect 6.2 "Normal Operation" directs removing fuse for saturated detector and resetting detector after inserting fuse. Moving fuel with purge on with 3 detectors is not per the procedure.

Need all 4 . C. Correct-Containment Radiation Signal (CRS) is actuated by 2 of 4 Containment area rad monitors, which would isolate Containment Purge. Only one rad monitor fails high therefore no securing of Containment Purge. OI-35 Radiation Monitoring System Sect 6.2 "Normal Operation" directs removing fuse for saturated detector and resetting detector after inserting fuse. No need to secure fuel handling based on a single pegged high area rad monitor. D. Incorrect

-Containment Radiation Signal (CRS) is actuated by 2 of 4 Containment area rad monitors, which would isolate Containment Purge. Only one rad monitor fails high therefore no securing of Containment Purge. OI-35 Radiation Monitoring System Sect 6.2 "Normal Operation" directs removing fuse for saturated detector and resetting detector after inserting fuse. Moving fuel with purge on with 3 detectors is not per the procedure.

Need all4 Page 110 of 157 Rev. 6 2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

KIA Info: RO Importance:

Proposed references to be provided to applicant:

Learning Objective:

10 CFR Part 55 Content: Question source: Cognitive level: Last NRC Exam used on: Technical references:

Comments:

Question 52 (Q97151) RMS predict impact then mitigate detector failure 211 073 -Process Radiation Monitoring

• A2-Ability to (a) predict the impacts of the following malfunctions or operations on the PRM system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

• A2.02-Detector failure 2.7 None Identify the interlocks and initiating devices associated with the Containment Purge System. 55.41(b )(5) 0Bank ID Modified lr8J New 0 Memory/Fundamental I r8J Comprehension/

Analysis N/ A -new question OI-35 Radiation Monitoring System SD-48 ESFAS System Description None Page 111 of 157 Rev. 6

,-----------------------------------------------*1 2014 NRC RO EXAM MASTER KEY 53. Components affected by loss of 12 SRW header Which ONE of the following components will be affected by the loss of 12 SRW header? A. EHC oil coolers B. 1 B Diesel Generator C. 12 Containment Air Cooler D. 12 Blowdown Heat Exchanger Answer: B Answer Explanation:

A. Incorrect-EHC cooler supplied from 11 SRW Header B. Correct-1B Diesel Generator is supplied by 12 SRW Header C. Incorrect

-12 Containment Air Cooler supplied from 11 SR W header D. Incorrect-12 Blowdown Heat Exchanger supplied from 11 SRW header Page 112 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

KIA Info: RO Importance:

Proposed references to be provided to applicant:

Learning Objective:

10 CFR Part 55 Content: Question source: Cognitive level: Last NRC Exam used on: Technical references:

Comments:

Question 53 (Q74576) Components affected by loss of 12 SR W header 2/1 076 Service Water

• K1 -Knowledge of the physical connections and/or cause-effect relationships between the SWS and the following systems:

• K1.05-DIG 3.8* None Identify the interrelationships that exist between the Service Water System and the Emergency Diesel Generators.

55.41 (b )(8) [gj Bank jo Modified jDNew [gj Memory/Fundamental I 0 Comprehension/

Analysis No record of use on a previous NRC exam AOP-7B-1 None Page 113 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY 54. 078 -Instrument Air What design features allow the 11 Salt Water Air Compressor to be operated during a Control Room fire? A. Key operated Local/Remote handswitch located at MCC-1 04 Local Control hand switch located at the Safe Shutdown Panel B. Key operated Local/Remote handswitch located at MCC-114, Local Control hand switch located at the Safe Shutdown Panel C. Key operated Local/Remote handswitch located at MCC-1 04, Local Control handswitch located at MCC-1 04 D. Key operated Local/Remote handswitch located at MCC-114, Local Control handswitch located at MCC-114 Answer: D Answer Explanation:

A. Incorrect-11 SW AC Local/Remote keyswitch and local control handswitch are both located on MCC-114 B. Incorrect

-11 SW AC Local/Remote keyswitch and local control handswitch are both located on MCC-114 C. Incorrect

-11 SW AC Local/Remote keyswitch and local control handswitch are both located on MCC-114 D. Correct-11 SWAC Local/Remote keyswitch and local control handswitch are both located on MCC-114 Page 114 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

KIA Info: RO Importance:

Proposed references to be provided to applicant:

Learning Objective:

10 CFR Part 55 Content: Question source: Cognitive level: Last NRC Exam used on: Technical references:

Comments:

Question 54 (Q20332) Operate the SWACs locally 2/1 078-Instrument Air

• K2 -Knowledge bus power supplies to the following:

• K2.02-Emergency Air compressor 3.3* None 55.41(b)(7)

[g) Bank jo Modified jDNew [g) Memory/Fundamental I 0 Comprehension/

Analysis No record of use on a previous NRC exam AOP-9A-1 None Page 115 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY 55. 103 -Containment Which ONE of the following conditions, by itself, is a loss of Containment Integrity?

A. 2-SRW-1582-CV, 21 CAC EMER DISCH, is open B. 2-PS-5464-CV, RCS SAMPLE ISOL, is open C. 2-EAD-5463-MOV, CNTMT NORM SUMP DRN, is open D. 2-SI-4144-MOV, CNTMT SUMP OUTLET ISOL, is open Answer: D Answer Explanation:

A. Incorrect-Containment Cooler SRW isolation valves get ESP AS signals. As long as the valves are operable, containment integrity is maintained.

B. Incorrect -, RCS Sample Isolation gets an ESP AS signal. As long as the valve is operable, containment integrity is maintained.

C. Incorrect

-The normal containment sump valves get an ESP AS signal which will automatically shut the valves, so containment integrity is maintained.

D. Correct-Per STP-0-55-2, Att. 2, 2-SI-4144 must be shut. Page 116 of 157 Rev. 6 2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

KIA Info: RO Importance:

Proposed references to be provided to applicant:

Learning Objective:

10 CFR Part 55 Content: Question source: Cognitive level: Last NRC Exam used on: Technical references:

Comments:

Question 55 (Q74932) Loss of Containment Integrity 2/1 103 -Containment

• K3-Knowledge of the effect that a loss or malfunction of the containment system will have on the following:

• K3 .02 -Loss of containment integrity under normal operations 3.8 None 55.41 (b )(7) (gJ Bank jo Modified IDNew (gJ Memory/Fundamental I 0 Comprehension/

Analysis Used in NRC exam administered 12/2008 STP-0-55-2, Containment Integrity Verification, Mode 1 -4 None Page 117 of 157 Rev. 6 2014 NRC RO EXAM MASTER KEY 56. 001 -Control Rod Drive Which ONE of the following set of conditions results in a CEA Withdrawal Prohibit (CWP) signal? A. 2 out of 4 pre-trips actuated from VOPT, High SUR or TM/LP condition.

B. 2 out of 4 pre-trips actuated from APD, TM/LP or VOPT condition.

C. 1 out of 4 pre-trips actuated from TM/LP, APD or High SUR condition.

D. 1 out of 4 pre-trips actuated from High SUR, VOPT or APD condition.

Answer: A Answer Explanation:

A. Correct-Per Alarm Response Manual window D-35 B. Incorrect-Logic is 2 out of 4 pre-trips actuated from VOPT, High SUR or TM/LP condition.

C. Incorrect

-Logic is 2 out of 4 pre-trips actuated from VOPT, High SUR or TM/LP condition.

D. Incorrect-Logic is 2 out of 4 pre-trips actuated from VOPT, High SUR or TM/LP condition.

Page 118 of 157 Rev. 6 2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

KIA Info: RO Importance:

Proposed references to be provided to applicant:

Learning Objective:

10 CFR Part 55 Content: Question source: Cognitive level: Last NRC Exam used on: Technical references:

Comments:

Question 56 (Q97171) Conditions causing a CWP 2/2 001 -Control Rod Drive

• K1 -Knowledge of the physical connections and/or cause-effect relationships between the CRDS and the following systems:

• K1.05 -NIS and RPS 4.5 None Identify the initiating signals and the RPS logic used to generate a Control Rod Withdrawal Prohibit (CWP) signal 55.41 (b )(7) 0Bank jo Modified lrg] New IZ! Memory/Fundamental I 0 Comprehension/

Analysis N/A -new question Alarm Response Manual 1 COS None Page 119 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY 57.002-RCS Which ONE of the following would cause a loss of conservatism in relation to the RCS flow trip from RPS? A. Grid frequency lowers to 59.9 Hz B. Grid frequency rises to 60.1 Hz C. Debris build up in primary side of S/G causes 5 tubes to plug D. Calibration of 1-PDI-101B (RCS Flow Indicator) is drifting low Answer: C Answer Explanation:

A. Incorrect

-Changing actual flow does not affect the conservatism of the setpoint B. Incorrect-Changing actual flow does not affect the conservatism of the setpoint C. Correct-This condition results in indicated flow appearing higher than actual flow, which will cause flow to be lower than RPS sensed when a trip occurs D. Incorrect-Indicated flow goes down with actual flow staying the same, low flow trip setpoint remains the same, therefore closer to low flow trip which is more conservative Page 120 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

KIA Info: RO Importance:

Proposed references to be provided to applicant:

Learning Objective:

10 CFR Part 55 Content: Question source: Cognitive level: Last NRC Exam used on: Technical references:

Comments:

Question 57 (Q97177) RCS Flow indication effect on indicated I actual reactor power 212 002-RCS

• A1 -Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the RCS controls including:

• Al.05-RCS Flow 4.0 None Recall the automatic RPS trips and their bases 55.41 (b )(5) 0Bank jo Modified 0 Memory/Fundamental I Comprehension/

Analysis N/A-new question TS 3.3.1 Basis None Page 121 of 157 Rev.6

---

-, 2014 NRC RO EXAM MASTER KEY 58. 011 -Pressurizer Level Control System U-2 tripped 2 minutes ago due to a Loss of the Red Bus. No operator actions or additional failures have occurred.

Which Charging Pump(s) if any are running? A. 21, 22, 23 Charging Pumps B. 21, 23 Charging Pumps C. 22 Charging Pump D. No Charging Pumps Answer: B Answer Explanation:

• 21 charging pump will be running because powered from black bus (not aD/G) and power lost to relay in PZR level control ckt which tells all charging pumps to start.

• 22 charging pump is not running because powered from 2B DIG. Charging pumps powered from a D/G require reset of charging spring to start (per question stem not done)

• 23 charging pump will be running because normally powered from 21 480V bus (same as 21 charging pump), power lost to relay in PZR level control ckt which tells all charging pumps to start. A. Incorrect

-22 pump is not running, see explanation above B. Correct -See explanation above C. Incorrect

-22 pump is not running, see explanation above D. Incorrect-21 & 23 pumps running Page 122 of 157 Rev.6 '

2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

KIA Info: RO Importance:

Proposed references to be provided to applicant:

Learning Objective:

10 CFR Part 55 Content: Question source: Cognitive level: Last NRC Exam used on: Technical references:

Comments:

Question 58 (Q97173) Loss of electrical effects on Charging Pumps 2/2 011 -Pressurizer Level Control System

• K6-Knowledge of the effect of a loss or malfunction on the following will have on the PZR LCS:

• K6.04-Operation ofPZR level controllers 3.1 None Given an electrical malfunction, recall the effect on the Charging Pumps. 55.41 (b )(7) 0Bank jo Modified 0 Memory/Fundamental Comprehension/

Analysis N/ A -new question AOP-7I, Loss of 4KV, 480 Volt or 208/120 Volt Instrument Bus Power None Page 123 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY 59. Knowledge of IRU power supplies An electrical bus fault occurred which caused 21 Charging Pump to lose power. Which ONE of the following pieces of equipment will also be de-energized?

A. 22 Boric Acid Pump B. 21 Battery Charger C. 23 Iodine Removal Unit D. 23 Containment Air Cooler Answer: C Answer Explanation:

A. Incorrect

-21 Charging Pump is powered from 480V Load Center 21 B; 22 Boric Acid Pump is powered from 480V MCC-204 B. Incorrect

-21 Charging Pump is powered from 480V Load Center 21 B; 21 Battery Charger is powered from 480V Load Center 24A C. Correct -21 Charging Pump and 23 Containment Filter (IRU) are powered from 480V Load Center 21 B D. Incorrect-21 Charging Pump is powered from 480V Load Center 21B; 23 Containment Air Cooler is powered from 480V Load Center 24A Page 124 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY Question 59 (Q91758) Topic: Knowledge of IRU power supplies Tier/Group:

2/2 027 -Containment Iodine Removal KIA Info:

• K2 -Knowledge ofbus power supplies to the following:

• K2.01-Fans RO Importance:

3.1

• Proposed references to be None provided to applicant:

Learning Objective:

List the power supplies to the Charging Pumps 10 CFR Part 55 Content: 55.41(b)(7)

Question source: IZ! Bank jo Modified jDNew Cognitive level: IZ! Memory/Fundamental J 0 Comprehension/

Analysis Last NRC Exam used on: No record of use on a previous NRC exam Technical references:

OI-27D-2, Station Power 480 Volt System Breaker Lineup Comments:

None Page 125 of 157 Rev. 6 2014 NRC RO EXAM MASTER KEY 60. 028 -Hydrogen Recombiner and Purge Control Following a Loss of Coolant Accident, the 1 are placed in service at a Hydrogen concentration in the Containment of (2) % IA W EOP-5, Loss of Coolant Accident A. (1) Hydrogen Recombiners (2) 0.5 B. (1) Hydrogen Recombiners and Hydrogen Purge (2) 0.5 C. (1) Hydrogen Recombiners; (2) 4.0 D. (1) Hydrogen Recombiners and Hydrogen Purge; (2) 4.0 Answer: A Answer Explanation:

EOP-5 directs use of the H 2 Recombiners ifH 2 concentrations rises to .5% and goes on to state if H 2 concentration rises to 4% consideration should be given to securing the H 2 Recombiners and placing the H 2 Purge System in operation (based on Technical Support Center recommendation).

Design of both systems is that either will maintain H 2 concentration below the flammability limit of4%. A. Correct -H 2 Recombiners are placed in service if Containment H 2 concentration reaches .5%. Design of both systems is that either will maintain H 2 concentration below the flammability limit of 4%. B. Incorrect

-H 2 Recombiners are placed in service if Containment H 2 concentration reaches .5%. Design of both systems is that either will maintain H 2 concentration below the flammability limit of 4%. C. Incorrect

-H 2 Recombiners are placed in service if Containment H2 concentration reaches .5%. Design of both systems is that either will maintain H 2 concentration below the flammability limit of 4%. D. Incorrect-H 2 Recombiners are placed in service if Containment H 2 concentration reaches .5%. Design ofboth systems is that either will maintain H 2 concentration below the flammability limit of 4%. Page 126 of 157 Rev. 6 2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

KIA Info: RO Importance:

Proposed references to be provided to applicant:

Learning Objective:

10 CFR Part 55 Content: Question source: Cognitive level: Last NRC Exam used on: Technical references:

Comments:

Question 60 (Q97174) Post-accident Containment H 2 control 2/2 028-Hydrogen Recombiner and Purge Control

• K5 -Knowledge of the operational implications of the following concepts as they apply to the HRPS:

• K5.02 -Flammable hydrogen concentration 3.4 None Given RCS parameters identify the appropriate response for a LOCA per EOP-5 55.41(b)(5) 0Bank ID Modified Memory/Fundamental I 0 Comprehension/

Analysis No record of use on a previous NRC exam EOP-5, Loss of Coolant Accident EOP-5, Loss of Coolant Accident Technical Basis document SD-060C, Hydrogen Purge System Description None Page 127 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY 61. 029-Containment Purge Given the following:

• Unit-1 is in Mode 6

• A Containment Normal Purge is in progress

• Refueling operations are in progress

• A fuel assembly is dropped within the core

• Containment Area Radiation Monitors (1-RE-5316-A through D) are in alarm at 200 mrem!hr. What automatic actions occur? A. Containment Purge Supply and Exhaust CVs SHUT; All Iodine Removal Units receive a START signal; Hydrogen Purge Outlet Containment isolation MOVs SHUT. B. Containment Purge Supply and Exhaust CVs SHUT; Containment Purge Supply and Exhaust Fans STOP; Hydrogen Purge Outlet Containment isolation MOVs SHUT. C. Containment Purge Exhaust CVs SHUT; Penetration Room Exhaust Fans receive START signal; Hydrogen Purge Outlet Containment Isolation MOVs SHUT D. Containment Purge Supply and Exhaust Fans STOP; Penetration Room Exhaust Fans receive a START signal; All Iodine Removal Units receive a START signal. Answer: B Answer Explanation:

A. Incorrect-The IRUs do not receive an automatic signal from the CRS circuitry.

B. Correct-A CRS is initiated by 2/4 coincidence from RI-5316A thru D. These six components actuate to the specified positions per LD-58A C. Incorrect

-The Purge Supply CV shuts and the Penetration Room Exhaust Fans do not receive an automatic signal from the CRS circuitry.

D. Incorrect-The Purge Supply and Exhaust CVs shut and the Penetration Room Exhaust Fans and IRUs do not receive an automatic signal from the CRS circuitry.

Page 128 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

KJA Info: RO Importance:

Proposed references to be provided to applicant:

Learning Objective:

10 CFR Part 55 Content: Question source: Cognitive level: Last NRC Exam used on: Technical references:

Comments:

Question 61 (Q24719) CRS Actuation 2/2 029 -Containment Purge

• A3 -Ability to monitor automatic operation of the Containment Purge System including:

• A3.01 -CPS isolation 3.8 None Identify the interlocks I initiating devices associated with Containment Purge 55.41 (b )(7) [gl Bank !D Modified IDNew [gl Memory/Fundamental I 0 Comprehension/

Analysis Used in NRC exam administered 12/2008 SD-048, Engineered Safety Features Actuation System Description SD-060B, Containment Purge System Description None Page 129 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY 62. 035-Steam Generator System Unit 2 is operating at 70% power, with both SGFPs in service, when a Feedwater System perturbation occurs. Which ONE of the following sets of conditions would require a Reactor Trip? A. "SGFP(S) SUCT PRESS LO" alarm annunciates B. "SGFPT TRIP" alarm annunciates and the tripped SGFP cannot be reset C. "SG L VL LO CH PRE-TRIP" alarm annunciates; feed flow is greater than steam flow D. "21 SG FW CONTR CH LVL" alarm annunciates; S/G level is approaching

(-)40 inches Answer: D Answer Explanation:

A. Incorrect-Receipt of the "SGFP(S) SUCT PRESS LO" alarm is not, by itself, sufficient criteria for tripping the unit. The alarm setpoint of 256 PSIG provides some margin to the trip setpoint which is 221 PSIG (2/2 logic) for greater than 20 seconds. B. Incorrect

-A reactor trip would not be ordered for a SGFT trip at the stated power level. One SGFPT is capable of providing required feed water flow at the stated power. C. Incorrect

-"SG L VL LO CH PRE-TRIP" alarm annunciates; feed flow is greater than steam flow D. Correct-AOP-3G, Malfunction Of Main Feedwater System, supplies the following trip criteria:

IF SG level is approaching

(-)40 inches, THEN, with the approval of the SM/CRS, trip the reactor and implement EOP-0, Post Trip Immediate Actions Page 130 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

KIA Info: RO Importance:

Proposed references to be provided to applicant:

Learning Objective:

10 CFR Part 55 Content: Question source: Cognitive level: Last NRC Exam used on: Technical references:

Comments:

Question 62 (Q97195) S/G Level related trip criteria 2/2 035-Steam Generator System

• 2.4 -Emergency Procedures I Plan

• 2.4.45 -Ability to prioritize and interpret the significance of each annunciator alarm 4.1 None Recall all automatic trips of the Steam Generator Feed Pumps. 55.41(b)(10) 0Bank ID Modified 0 Memory/Fundamental Comprehension/

Analysis No record of use on a previous NRC exam 1C03-ALM, Condensate

& Feedwater Control Alarm Manual AOP-3G, Malfunction Of Main Feedwater System None Page 131 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY 63. 041 -Steam Dump/Turbine Bypass Control Unit 1 is operating at 100% power, with a Core Bumup of 1200 MWD/MTU, when an ADV fails open. What is the effect on Unit 1 and what is/are the first action(s) taken? A. Reactor power rises causing a Reactor trip on VOPT; Implement EOP-0. B. Reactor power lowers; Raise turbine load to restore T coLD to the program band. C. Reactor power rises but reactor does NOT trip; Dispatch operators to manually isolate the ADV and insert CEAs as necessary to control reactor power. D. Reactor power rises but reactor does NOT trip; Insert CEAs as necessary to control reactor power and reduce turbine load to restore T coLD to the program band. Answer: D Answer Explanation:

A. Incorrect

-ADV will raise power approximately three percent which is below the VOPT setpoint.

Plausible because power does rise just not enough to cause trip B. Incorrect

-Raising turbine load would decrease T coLD further and cause reactor power to raise further. Plausible because BOL has +MTC at lower power C. Incorrect-Steps to dispatch operators to manually isolate the ADV follow the correct action of reducing turbine load to maintain T coLD in the program band. Plausible because everything correct except isolating ADV comes after insert CEA and adjust turbine load D. Correct-Per AOP-7K,Section IV. One ADV is good for 2.5% steam demand, so the VOPT trip setpoint should not be reached and no trip on high power should occur. Page 132 of 157 Rev. 6 2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

KIA Info: RO Importance:

Proposed references to be provided to applicant:

Learning Objective:

10 CFR Part 55 Content: Question source: Cognitive level: Last NRC Exam used on: Technical references:

Comments:

Question 63 (Q28429) ADV fails open in Mode 1 2/2 041 -Steam Dump/Turbine Bypass Control

• K3 -Knowledge of the effect that a loss or malfunction of the SDS will have on the following:

• K3.02-RCS 3.8 None Given plant conditions and several alternate actions determine the most appropriate actions, per AOP-7K, for the event in progress.

55.41(b )(7) jo Modified jDNew D Memory/Fundamental I Comprehension/

Analysis Used in NRC exam administered 12/2008 AOP-7K, Overcooling Event in Mode One or Two None Page 133 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY 64. 045 -Main Turbine Generator Why is "IMPULSE IN" selected when performing Governor Valve (GV) testing on U-2? A. To complete the circuit that allows shutting governor valves individually B. To bypass the Valve Position Limit circuit to the GV's not being tested. C. To provide for more linear and faster response to changes in load during testing D. To limit the opening of GV's during testing Answer: C Answer Explanation:

A. Incorrect

-plausible as pushbutton to perform testing B. Incorrect

-plausible because VPL is manually raised to max setting for valve testing C. Correct -Depressing the Imp In pushbutton enables impulse pressure feedback to the control circuit for better control D. Incorrect-plausible because function ofVPL (valve position limit) Higher cognitive due to fact that students will not have this memorized and must analyze system operation and procedure requirements to ascertain the correct answer Page 134 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

KIA Info: RO Importance:

Proposed references to be provided to applicant:

Learning Objective:

10 CFR Part 55 Content: Question source: Cognitive level: Last NRC Exam used on: Technical references:

Comments:

Question 64 (Q97175) Knowledge of "Impulse In" circuit 2/2 045 -Main Turbine Generator

• K4-Knowledge ofMT/G system design feature(s) and/or interlock (s) which provide for the following:

• K4.01 -Programmed controller for relationship between steam pressure at T/G inlet (impulse, first stage) and plant power level 2.7 None Recall operation of the Impulse In I Out controls on the Main Turbine Control Panel 55.41 (b )(7) 0Bank Jo Modified jCBJ New 0 Memory/Fundamental l C8J Comprehension/Analysis N/ A -new question OI-43C-2, Unit-2 Main Turbine Performance Evaluation Checks SD-93C-2 U-2 Main Turbine Controls None Page 135 of 157 Rev.6 ---; :

2014 NRC RO EXAM MASTER KEY 65. 068 -Liquid Radwaste Which ONE of the following conditions would REQUIRE a discharge of 12 RCWMT be secured per OI-17C-4, DISCHARGING REACTOR COOLANT WASTE TANKS TO THE ENVIRONMENT?

A. Composite sampler fails B. 0-RI-2201 RMS indication below background for> 1 min C. Level in 12 RCWMT reaches 1.5 feet D. Discharge flow indicator fails Answer: B Answer Explanation:

A. Incorrect

-Composite sampler failure is a problem that could result in securing the discharge but is not required.

OI-17C-4 provides instruction to inform Chemistry to obtain grab samples before and during the discharge.

B. Correct-Per OI-17C-4, IF the discharge activity indicated by the liquid waste RMS exhibits a sustained decrease below the discharge permit background values for greater than 30 seconds, THEN STOP the running RCWMT PP. A. Incorrect

-An RCWMT level of 1.5 ft does not require securing the discharge.

A note in OI-17C-4 states a level of0.9 feet will auto stop the selected RCWMT PP. Discharge of 12 RCWMT can continue, per the procedure, to the auto-stop setpoint.

B. Incorrect

-Discharge flow indicator is a problem that could result in securing the discharge but does not require securing the discharge.

OI-17C-4 provides direction for continuing the discharge in the event the flow indicator malfunctions.

Page 136 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

KIA Info: RO Importance:

Proposed references to be provided to applicant:

Learning Objective:

10 CFR Part 55 Content: Question source: Cognitive level: Last NRC Exam used on: Technical references:

Comments:

Question 65 (Q97176) Discharge termination criteria 2/2 068 -Liquid Radwaste

• A4-Ability to manually operate and/or monitor in the control room:

• A4.02-Remote radwaste release 3.2* None 55.41(b)(7) 0Bank Jo Modified Memory/Fundamental I 0 Comprehension/

Analysis N/ A -new question OI-17C-4, Discharging Reactor Coolant Waste Tanks To The Environment None Page 137 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY 66. 2.1 -Conduct of Operations Unit-1 is at 100% power when 11 SW pump trips. "U-1 4KV MOTOR OVERLOAD" alarm is annunciated.

No common mode failure is indicated and the Control Room crew has decided to align the standby pump per the applicable procedure.

Per plant administrative procedures, which ONE of the following describes the minimum acceptable direction given to the Plant Operator(s) to guide his response in this situation?

A. Direct the PPO to "Restore Saltwater per AOP-7 A, Section C". B. Direct the Outside Operator to "Align 13 Saltwater Pp to 11 SW Header per AOP-7 A-1,Section V.C.2.B, on page 11 ". C. Announce over the plant page "11 Saltwater Pp has tripped, realign 13 Saltwater Pp to 11 Saltwater Header". D. Direct the Turbine Building Operator to "align 13 Saltwater Pp power supply to 11 4KV bus" and the outside operator to "align 13 Saltwater Pp to 11 header per AOP-7 A-1" Answer: B Answer Explanation:

A. Incorrect-Provides no direction as to what components to restore. Unit, page number and specific steps of the procedure to use are not specified.

This is not in compliance with the requirements ofN0-1-200 B. Correct-Per N0-1-200:

When directing the performance of procedural steps to Operators or other plant personnel who have a copy of the procedure, the specific steps should be identified as a minimum by affected Unit, procedure number, page number, and step or section number. C. Incorrect-Per N0-1-200:

Except during emergency conditions requiring immediate PO actions, page announcements shall not be used for passing operating orders. If orders are given via the plant page, in an emergency condition, the requirements ofN0-1-200, step 5.32.A. are still applicable.

D. Incorrect-AOP-7 A-1 will not direct shifting power supplies if power is available to the swing pump. No section#, page# or step #s are provided Page 138 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

KIA Info: RO Importance:

Proposed references to be provided to applicant:

Learning Objective:

10 CFR Part 55 Content: Question source: Cognitive level: Last NRC Exam used on: Technical references:

Comments:

Question 66 (Q20575) Providing direction to personnel outside the CR Generic Knowledge

& Abilities 2.1 -Conduct of Operations

• 2.1.8 -Ability to coordinate personnel activities outside the control room. 3.4 None 55.41(b)(10) jo Modified jDNew Memory/Fundamental J 0 Comprehension/

Analysis Used in NRC exam administered 7/2002 N 0-1-200, Control of Shift Activities AOP-7 A-2, Loss Of Saltwater Cooling None Page 139 of 157 Rev. 6 2014 NRC RO EXAM MASTER KEY 67. 2.1 -Conduct of Operations Core offload is in progress.

A core to upender fuel move sequence is occurring.

During withdrawal of a fuel assembly from the core, as the hoist box position encoder readout is approaching 126.00 inches, the load increases from 1385 pounds to 2680 pounds. Which ONE of the following explains why the load readout has increased?

A. The hoist box load is transferred to the hoist, the weight gain is normal. B. The fuel assembly is bowed and contacting reactor vessel internals.

C. A grid-to-grid hang-up with an adjacent fuel assembly is occurring.

D. The camera has contacted the core support barrel. Answer: A Answer Explanation:

A. Correct-OI-25C references picking up the weight of the hoist box at about 126.5" B. Incorrect-at this height, the fuel is inside the hoist box and above other fuel assemblies.

C. Incorrect

-at this height, the fuel is inside the hoist box and above other fuel assemblies.

D. Incorrect-RFHM interlocks prevent this condition from occurring and the weight gain couldn't be attributed to camera interference.

Students are not expected to memorize specific heights associated with Refueling Machine operation but there is enough information in this question for the students to analyze each of the distractors and arrive at the correct answer, therefore higher cognitive Page 140 of 157 Rev. 6 2014 NRC RO EXAM MASTER KEY Question 67 (Q14509) Topic: RFHM operating characteristics Tier/Group:

Generic Knowledge

& Abilities KIA Info: 2.1 -Conduct of Operations

• 2.1.41 -Knowledge of the refueling process RO Importance:

2.8 Proposed

references to be None provided to applicant:

Recall the various modes of operation of the RFM Console, Learning Objective:

including operability of all electrical interlocks during manual operation of the RFM. 10 CFR Part 55 Content: 55.41(b)(10)

Question source: IZ! Bank jo Modified jDNew Cognitive level: D Memory/Fundamental jiZI Comprehension/

Analysis Last NRC Exam used on: No record of use on a previous NRC exam Technical references:

OI-25C, Refueling Machine Comments:

None Page 141 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY 68. 2.2 -Equipment Control Which ONE of the following conditions would be considered preconditioning?

A. Performing a monthly STP on weekly basis for closer monitoring of valve performance.

B. Draining the Containment Sump prior to receipt of the alarm, 3 hrs before performing a timed stroke STP. C. During performance oftime stroke STP an error occurred using the stopwatch and performed the test again immediately.

D. Opening 8" CAC emergency outlet valve to lower a high Containment temperature condition, 3 hrs before the same emergency outlet valve time stroke STP Answer: B Answer Explanation:

A. Incorrect-EN-4-108 (3.13) specifically allows increased surveillance frequency B. Correct-EN-4-108 (3.13) only allows operation if reasonably needed for plant evolution.

Draining the Containment Sump prior to receipt ofthe alarm is not normal or reasonable C. Incorrect-EN-4-108 (3.19) specifically allows retest for test equipment malfunction D. Incorrect-EN-4-1 08 (3.13) specifically allows operation if reasonably needed for plant evolution Each of these distractors would technically be preconditioning but the procedure makes allowances for only one. Students are not expected to memorize all potential preconditioning conditions but this question gives enough information that the students can analyze each situation and determine which one would be acceptable, therefore higher cognitive.

Page 142 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY "* Topic: Tier/Group:

KIA Info: RO Importance:

Proposed references to be provided to applicant:

Learning Objective:

10 CFR Part 55 Content: Question source: Cognitive level: Last NRC Exam used on: Technical references:

Comments:

Question 68 (Q97178) Pre-conditioning of SR components Generic Knowledge

& Abilities 2.2 -Equipment Control

• 2.2.12-Knowledge of the surveillance procedures 3.7 None 55.41(b )(1 0) 0Bank lD Modified D Memory/Fundamental Comprehension/

Analysis N/A-new question EN-4-108, ASME In-service Testing OfPower-Operated Valves & Manual Valves None Page 143 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY 69. 2.2 -Equipment Control When preparing a tagout, which ONE of the following sets of values require the use of two isolation points, (1) Fluid Temperature (2) System Pressure A. (1) > 200°F, (2) > 500 psig B. (1) > 200°F, (2) > 250 psig C. (1) > 120°F, (2) > 250 psig D. (1) > 120°F, (2) > 500 psig Answer: A Answer Explanation:

A. Correct--Per CNG-OP-1.01-1 007 the following are considered for the use of double isolation:;

Greater than to 200°F temperature; Greater than 500 psig; B. Incorrect--Per CNG-OP-1.01-1 007 the following are considered for the use of double isolation:

Greater than 200°F temperature; Greater than 500 psig C. Incorrect-Per CNG-OP-1.01-1007 the following are considered for the use of double isolation:

Greater than 200°F temperature; Greater than 500 psig D. Incorrect--Per CNG-OP-1.01-1 007 the following are considered for the use of double isolation:

Greater than 200°F temperature; Greater than 500 psig Page 144 of 157 Rev. 6 2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

KIA Info: RO Importance:

Proposed references to be provided to applicant:

Learning Objective:

10 CFR Part 55 Content: Question source: Cognitive level: Last NRC Exam used on: Technical references:

Comments:

Question 69 (Q97179) Knowledge of Safety Tagging procedures Generic Knowledge

& Abilities 2.2 -Equipment Control

• 2.2.13 -Knowledge of safety and tagging clearance procedures 4.1 None 55.41 (b )(1 0) 0Bank I 0 Modified IIZI New IZI Memory/Fundamental I 0 Comprehension/

Analysis N/ A -new question N0-1-112 Safety Tagging Guidance CNG-OP-1.01-1007, Clearance And Safety Tagging None Page 145 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY 70. 2.2 -Equipment Control Given the following:

• Unit 2 is in Mode 6, refueling has been completed

• SDC remains in service

• The reactor head stud bolts are being tensioned

• A penetration has a Containment closure deviation on file. It can be restored in 10 minutes

• The time to boil is 45 minutes.

• T.S 3.9.4/3.9.5 limitations are met. What is the maximum amount of time the penetration may be left open? A. 45 minutes B. 7 days C. Indefinitely (as long as mode 5 is not entered) D. Indefinitely (as long as mode 4 is not entered) Answer: D Answer Explanation:

A. Incorrect

-45 minutes is the time to boil, as long as the penetration is able to be closed within TTB then not required to be shut B. Incorrect

-7 days is requirement for STP 0-55A periodicity while shutdown, , as long as the penetration is able to be closed within TTB then not required to be shut C. Incorrect

-Indefinitely is correct but mode change we are worried about is Mode 4 because the containment is required to have integrity vice closure in Mode 1-4 D. Correct -per N 0-1-114 section 5 .l.B "If a penetration can be closed in less than the TTB and T.S. 3.9.4/3.9.5 requirements, then there are no limits on the duration that the penetration is allowed to be open. As long as unit does not enter mode of applicability for containment integrity (Modes 1-4) Page 146 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

KIA Info: RO Importance:

Proposed references to be provided to applicant:

Learning Objective:

10 CFR Part 55 Content: Question source: Cognitive level: Last NRC Exam used on: Technical references:

Comments:

Question 70 (Q44860) Limitations on closure openings Generic Knowledge

& Abilities 2.2 -Equipment Control

• 2.2.14 -Knowledge of the process for controlling equipment configuration or status. 3.9 None 55.41(b)(10) jo Modified jDNew 0 Memory/Fundamental

_I rgj Comprehension/

Analysis No record of use on a previous NRC exam None Page 147 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY 71. 2.3 -Radiation Control The following emergency situations may warrant an individual dose in excess of the established regulatory limit of 5 REM per year: (1) Life Saving (voluntary)

(2) Facility Protection Which ONE of the following represents the guidance for dose accumulated by an individual during these emergency situations?

A. (1) May exceed 25 REM; (2) Limited to 10 REM B. (1) May exceed 25 REM (2) Limited to 25 REM C. (1) Limited to 25 REM; (2) Limited to 25 REM D. (1) Limited to 10 REM, (2) Limited to 10 REM Answer: A Answer Explanation:

A. Correct -Dose limits specified in ERPIP 831 are: May exceed 25 REM for Lifesaving (voluntary) and limited to 10 REM for Facility Protection.

B. Incorrect

-Dose limits specified in ERPIP 831 are: May exceed 25 REM for Lifesaving (voluntary) and limited to 10 REM for Facility Protection.

C. Incorrect

-Dose limits specified in ERPIP 831 are: May exceed 25 REM for Lifesaving (voluntary) and limited to 10 REM for Facility Protection.

D. Incorrect

-Dose limits specified in ERPIP 831 are: May exceed 25 REM for Lifesaving (voluntary) and limited to 10 REM for Facility Protection.

Page 148 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

KIA Info: RO Importance:

Proposed references to be provided to applicant:

Learning Objective:

10 CFR Part 55 Content: Question source: Cognitive level: Last NRC Exam used on: Technical references:

Comments:

Question 71 (Q92910) Emergency dose limits Generic Knowledge

& Abilities 2.3 -Radiation Control

• 2.3.4-Knowledge of radiation exposure limits under normal or emergency conditions.

3.2 None 55.41 (b )(12) IZI Bank .I 0 Modified jDNew IZI Memory/Fundamental I 0 Comprehension/

Analysis Used in the 8/2010 NRC exam ERPIP 831, Emergency Radiation Exposure Guidance.

None Page 149 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY 72. 2.3-Radiation Control You are signed onto the Ops Low Risk RWP #2 Task #2 for an entry into the Aux Bldg, with limits of 10 mrem dose and 80 mrem!hr does rate. Which set of alarms setpoints would you expect to see when checking your EPD prior to entry? A. 8 mrem 64 mrem!hr B. 8 mrem C. lOmrem D. lOmrem Answer: B Answer Explanation:

80 mrem!hr 64mrernlhr 80mrem!hr The alarms are set on the EPD to 80% of dose limit and at the dose rate limit. Therefore correct answer is 8 mrem & 80 mrem!hr A. Incorrect-See explanation above B. Correct-See explanation above C. Incorrect-See explanation above D. Incorrect-See explanation above Page 150 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

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Proposed references to be provided to applicant:

Learning Objective:

10 CFR Part 55 Content: Question source: Cognitive level: Last NRC Exam used on: Technical references:

Comments:

Question 72 (Q74533) R WP use in emergency situations Generic Knowledge

& Abilities 2.3 -Radiation Control

• 2.3.7-Ability to comply with radiation work permit requirements during normal or abnormal conditions.

3.5 None 55.41(b)(12) 0Bank /0 Modified /[gl New [gl Memory/Fundamental I 0 Comprehension/

Analysis No record of use on a previous NRC exam RWP-2-2, Operations Activities None Page 151 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY 73. 2.3 -Radiation Control U-1 is operating normally at 100% when a S/G tube leak of approximately 50 GPM develops.

Which ONE of the following steps, directed by AOP-2A, Excessive Reactor Coolant Leakage, is performed to prevent the release of contamination?

A. Reduce turbine load to maintain T cow on program B. Ensure no leakage into Component Cooling subsystem C. Direct Rad Safety to survey Water Treatment Area D. Borate to reduce TAvE < 537°F Answer: D Answer Explanation:

A. Incorrect-Per AOP-2A: "This allows the operator to control the plant using an observable parameter while dropping load as temperature is being reduced to< 537°F. In the past the operator was trying to keep temperature near program which was constantly changing and had to be looked up. Controlling SG pressure in this band keeps the plant from swinging as power is fairly rapidly reduced." B. Incorrect

-step only performed iflocation ofleak not identified.

Information given is that a tube leak in excess of 1 charging pump therefore unit is tripped in Step VLD and this step (VLE.1 0) not performed C. Incorrect-This step is in section V, RCS leakage within capacity of 1 charging pump. Information given states 15 gpm which exceeds capacity of charging pump with LID in service. Therefore this step is not performed.

D. Correct-AOP-2A basis states: "the intent of this step is to reduce TAvE to less than 53 7°F if possible before tripping the Reactor to minimize the possibility of lifting a Steam Generator Safety Valve. The setpoint of the lowest lifting main steam safety valve corresponds to a saturation temperature of approximately 537°F. Page 152 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

KIA Info: RO Importance:

Proposed references to be provided to applicant:

Learning Objective:

10 CFR Part 55 Content: Question source: Cognitive level: Last NRC Exam used on: Technical references:

Comments:

Question 73 (Q97180) Control of contamination during a SGTR Event Generic Knowledge

& Abilities 2.3 -Radiation Control

• 2.3.14-Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities.

3.4 None Recall the reason for reducing TAvE to <537°F, prior to tripping the unit, during a Steam Generator Tube Leak. 55.41(b)(12) 0Bank jo Modified 0 Memory/Fundamental I Comprehension/

Analysis N/ A -new question EOP-6, Steam Generator Tube Rupture EOP-6, Steam Generator Tube Rupture Technical Basis Document None Page 153 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY 74. 2.4-Emergency Procedures I Plan Unit 1 is operating at 60% power when a loss of 4KV Bus 13 occurs. Which ONE of the following actions must be taken per AOP-71, Loss Of 4KV, 480 Volt Or 208/120 Volt Instrument Bus Power? A. Bypass Condensate Precoat Filters and Demineralizers and verify or start 12 or 13 Condensate Pump. B. With concurrence of the SM/CRS trip the Reactor and Turbine then implement EOP-0. C. Perform a power reduction to lower Condensate Header flow to below 8000 GPM. D. Isolate S/G Blowdown and verify 12 Heater Drain Pump is running. Answer: C Answer Explanation:

A. Incorrect-12 and 13 Condensate Pumps are powered by 4KV Bus 13. B. Incorrect-Trip criteria is provided in AOP-7I-1 and not met initially

(<70%), following the guidance to reduce power as necessary to maintain Condensate header flow less than 8000 GPM. Tripping the Reactor and Turbine would be based on exceeding the trip criteria during the course of the power reduction.

C. Correct-AOP-71-1, for the loss of 13 4KV Bus, states: IF Reactor power is less than 70% Reduce power as necessary to maintain Condensate header flow less than 8000 GPM PER OP-3, Normal Power Operation.

D. Incorrect-12 Heater Drain Pump is powered by 4KV Bus 13. Page 154 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

KIA Info: RO Importance:

Proposed references to be provided to applicant:

Learning Objective:

10 CFR Part 55 Content: Question source: Cognitive level: Last NRC Exam used on: Technical references:

Comments:

Question 74 (Q28551) Response to loss of 13 4KV Bus Generic Knowledge

& Abilities 2.4 -Emergency Procedures I Plan

• 2.4.11 -Knowledge of abnormal condition procedures . 4.0 None Recall why a loss of 4KV Bus 13(23) at> 70% power requires a Unit trip 55.41 (b )(1 0) [gl Bank /D Modified /DNew [gl Memory/Fundamental J D Comprehension/Analysis No record of use on a previous NRC exam AOP-7I, Loss of 4KV, 480 Volt or 208/120 Volt Instrument Bus Power None Page 155 of 157 Rev. 6 2014 NRC RO EXAM MASTER KEY 75. 2.4 -Emergency Procedures I Plan (1) Where does the Operations Technical Advisor report to on declaration of a Fire and (2) what is the Operations Technical Advisor's function?

A. (1) Reports to the Control Room; (2) Provides technical assistance regarding firefighting strategies B. (1) Reports to the Control Room; (2) Provides technical assistance regarding safe shutdown equipment C. (1) Reports to the scene of the Fire; (2) Provides technical assistance regarding safe shutdown equipment D. (1) Reports to the scene ofthe Fire; (2) Provides technical assistance regarding firefighting strategies Answer: C Answer Explanation:

A. Incorrect-Per SA-1-101 section 5.1.D: The Operations Technical Advisor shall be a licensed operator who responds with the Fire Brigade to provide technical assistance to the Fire Brigade Leader regarding safe shutdown equipment functions.

B. Incorrect-Per SA-1-101 section 5.1.D: The Operations Technical Advisor shall be a licensed operator who responds with the Fire Brigade to provide technical assistance to the Fire Brigade Leader regarding safe shutdown equipment functions.

C. Correct-Per SA-1-1 01, Section 4.6.A. "The Operations Technical Advisor shall be responsible to, in response to a fire, provide information on safe shutdown equipment to the Fire Brigade Leader and report the status of conditions in the area to the Control Room." Per SA-1-1 01 section 5.1.D: The Operations Technical Advisor shall be a licensed operator who responds with the Fire Brigade to provide technical assistance to the Fire Brigade Leader regarding safe shutdown equipment functions.

D. Incorrect-Per SA-1-101 section 5.1.D: The Operations Technical Advisor shall be a licensed operator who responds with the Fire Brigade to provide technical assistance to the Fire Brigade Leader regarding safe shutdown equipment functions.

Page 156 of 157 Rev.6 2014 NRC RO EXAM MASTER KEY Topic: Tier/Group:

KIA Info: RO Importance:

Proposed references to be provided to applicant:

Learning Objective:

10 CFR Part 55 Content: Question source: Cognitive level: Last NRC Exam used on: Technical references:

Comments:

Question 75 (Q50730) OT A responsibilities Generic Knowledge

& Abilities 2.4 -Emergency Procedures I Plan

• 2.4.25 -Knowledge of fire protection procedures . 3.3 None Describe the activities performed by the OT A during a Fire Brigade response to a fire lAW SA 1 01 and 1 OCFR50, Appendix "R" 55.41 (b )(1 0) !ZI Bank 10 Modified IONew !ZI Memory/Fundamental I 0 Comprehension/

Analysis No record of use on a previous NRC exam SA-1-1 01, Fire Fighting None Page 157 of 157 Rev.6 CAL VERT CLIFFS NUCLEAR POWER PLANT 2014 NRC INITIAL LICENSED OPERATOR SRO WRITTEN EXAM KEY Page 1 of 55 Rev.6 2014 NRC SRO EXAM MASTER KEY Revision 1 (Incorporates Ops Reviewer Comments)

• #77 -Minor editorial correction to the stem of the question * #79 -Minor editorial correction to the stem of the question * #82-Minor Changes to answers A & B (unaffected CEA's) * #83 -Minor editorial correction to the stem of the question * #85-Changes to stem and distractors to eliminate ambiguity

• #86-Changes to stem and distractors to eliminate ambiguity

• #87 -Minor editorial correction to the stem of the question * #90-Minor editorial correction to the stem of the question * #91-Changed the fault to LRNI (Q19 had WRNI Power Supply) * #98 -Minor editorial correction to the stem of the question * #99 -Minor editorial correction to the stem of the question * #1 00-Changed ERPIP form titles to new standard

• Numerous changes to distractor order to equalize correct answers: 6-A, 7-B, 6-C, 6-D Revision 2 (Incorporates Collegial Review Comments)

• #79 -Minor change to stem * #83 -Added "Using Provided Reference" * #87 -Reworded stem to clarify * #90-Minor changes to wording of stem and swap (1) & (2) to make clearer * #91-Minor change in stem to make clearer Revision 3 (Incorporates Exam V alidator Comments)

• #82 -Replaced question with another that matched the Kl A * #86 -Minor change to distractors A & C to make clearer Page 2 of 55 Rev.6 2014 NRC SRO EXAM MASTER KEY Revision 4 (Incorporates 2nd Round Exam Validator Comments)

• #79 -Minor style change to clarify 2 parts of question/answer

• #87 -Changed N-16 to MSLRM in stem for better operational validity * #93-Minor editorial change to stem * #100-Added using provided reference (EAL Hot Chart) Revision 5 (Incorporates Fleet Reviewer and NRC pre-review Comments)

• #76-Changed stem to cause more diagnosis based on NRC comment about LOD * #78-Minor editorial changes to distractors A, B, & C * #79 -Added explanation as to why this is as SRO question * #81 -Minor editorial change to stem, added "implemented on U-1" * #82 -Rewrote question based on NRC feedback about SRO question * #87 -Minor editorial changes to stem based on NRC feedback * #90 -Editorial change to stem (bold "all") and added "potential" * #93 -Minor editorial change to stem to clarify Kl A match Page 3 of 55 Rev.6 2014 NRC SRO EXAM MASTER KEY Revision 6 (Added Learning objectives where applicable, Integrated NRC validation week and Don Jackson comments)

• #79 -Added procedure references to answer choices * #76, 83, 89, 98, 100-Removed phrase "Using provided references" (No change in difficulty)

• #76-Added initial flow rates for auto makeup based on validator comment, approved by NRC via phone * #84 -Changed "would" to "should" in stem based on validator comment, approved by NRC via phone * #86 -Added bullet to stem about fluctuating CS PP amps (Lower difficulty)

• #88 -Added procedure references to answer choices "A" & "D" * #91 -Added the word "only" in answer choices "C" & "D" * #92 -Added IA W AOP-1 0 to the stem * #95-Changed distractors to make a 2 by 2 (Increased difficulty) and made answer choice "A" read "hours * #97 -Changed stem to say 180 degrees vice mode 5, changed distractor D from Mode 2 to Mode 1 (Increased difficulty)

Page 4 of 55 Rev.6 2014 NRC SRO EXAM MASTER KEY 176. 022-Loss ofRx-Coolant Makeup U-1 is in Mode 3 and cooling down to Mode 5 for a refueling outage. The following conditions exist:

• The RCS has been borated to refueling boron concentration.

• CVCS is aligned for auto makeup to the VCT. (15 GPM Boric Acid flow I 60 GPM Demineralized Water flow). Assuming the demineralized water flow transmitter feeding the MAKEUP WTR FLOW CONTR (FIC-21 OX) was inadvertently isolated sometime in the past 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. Which ONE of the following describes:

(1) The effect this failure has on the RCS and (2) What direction would you give to RO? A. (1) RCS boron concentration would be lower than the last sample (2) Stop cooldown, stop auto makeup to VCT, direct Chemistry to sample the RCS B. (1) RCS boron concentration would be lower than the last sample (2) Continue cooldown, immediately commence fast boration to restore SDM C. (1) RCS boron concentration would be higher than the last sample (2) Stop cooldown, stop auto makeup to VCT, direct Chemistry to sample the RCS D. (1) RCS boron concentration would be higher than the last sample (2) Continue cooldown, dilute the RCS to prevent exceeding max boron concentration.

Answer: A Answer Explanation:

A. Correct -Isolating the DI flow transmitter would cause indication of 0 gpm which would cause controller to fully open causing excessive DI flow which would dilute the RCS but 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of excessive DI flow would not be expected to cause a violation of SDM since stem states that all BA has been added up to refueling concentration therefore stopping auto MIU and sample RCS are appropriate.

With no makeup to VCT the cooldown should also be stopped. B. Incorrect-SDM not violated.

Refueling boron concentration is 2505 ppm and most restrictive SDM in mode 5 is 1517 ppm. DI M/U controller is set for 100 gpm and failing high would cause it to put out max ofDI transfer pump. To dilute from 2500 ppm to 1500 ppm is 37000 gallons of DI which would require >300gpm ofDI flow (max is <200gpm) for entire 120 mins. C. Incorrect

-RCS concentration would be lower not higher D. Incorrect-RCS concentration would be lower not higher Page 5 of 55 Rev.6 2014 NRC SRO EXAM MASTER KEY Topic: Tier/Group:

KIA Info: SRO Importance:

Proposed references to be provided to applicant:

Learning Objective:

10 CPR Part 55 Content: Question source: Cognitive level: Last NRC Exam used on: Technical references:

Comments:

Question 76 (Q97190) SRO Only RCMU controller failure 111 022 -Loss of Rx Coolant Makeup

• AA2 -Ability to determine and interpret the following as they apply to the Loss of Reactor Coolant Makeup:

• AA2.03-Failures of flow control valve or controller 3.6 NEOP-13 P.22-29 and OI-2B Figures 1-4 Given an inadvertent boron dilution, be able to identify, understand the basis, and take appropriate actions, per procedures, to mitigate the event. 55.43(b)(5) 0Bank jo Modified 0 Memory/Fundamental Comprehension/

Analysis N/A-new question AOP-lA, Inadvertent Boron Dilution None Page 6 of 55 Rev.6 2014'NRC SRO EXAM MASTER KEY 77. 040 -Stearn Line Rupture-Excessive Heat Transfer Given U-2 with the following conditions:

• RCS pressure is 1545 PSIA slowly lowering

• Containment pressure is 3.1 PSIG and slowly rising

• 21 & 22 S/G pressures are approximately 720 PSIA and slowly lowering

• CRO was delayed in shutting MSIV's, CRS has directed them to be shut Which ONE of the following describes (1) The correct Optimal Recovery Procedure to implement and (2) The first Block Step to be assigned to RO/CRO from the selected procedure?

A. (1) Implement EOP-5, Loss of Coolant Accident; (2) Monitor RCS Depressurization B. (1) Implement EOP-5, Loss of Coolant Accident; (2) Depressurize the RCS to Reduce Subcooling and Maintain Pressurizer Level C. (1) Implement EOP-4, Excess Stearn Demand Event; (2) Monitor ESF AS Actuation D. (1) Implement EOP-4, Excess Stearn Demand Event; (2) Maintain RCS Subcooling and Pressurizer Level Answer: C Answer Explanation:

A. Incorrect

-The parameters provided support diagnosis of a steam line break in the Containment, rather than a LOCA, as evidenced by a sub cooled margin of 94 op. Monitoring RCS depressurization would be the correct block step selection if the event were a LOCA. B. Incorrect

-The parameters provided support diagnosis of a steam line break in the Containment, rather than a LOCA, as evidenced by a subcooled margin of94°F. Depressurize the RCS to Reduce Subcooling and Maintain Pressurizer Level would be the correct block step selection (but not the first one) if the event were a LOCA. C. Correct -The parameters provided support diagnosis of a steam line break in the Containment warranting implementation ofEOP-4, Excess Stearn Demand Event. Monitor ESF AS Actuation is the first step assigned to one of the board operators.

D. Incorrect-The parameters provided support diagnosis of a steam line break in the Containment warranting implementation of EOP-4, Excess Stearn Demand Event. This block step provides guidance for maintaining RCS subcooling and pressurizer level and is situated after affected S/G isolation in the procedure.

Page 7 of 55 Rev.6 2014 NRC SRO EXAM MASTER KEY Topic: Tier/Group:

KIA Info: SRO Importance:

Proposed references to be provided to applicant:

Learning Objective:

10 CFR Part 55 Content: Question source: Cognitive level: Last NRC Exam used on: Technical references:

Comments:

Question 77 (Q97155) Diagnose EOP-4 entry and select 1st block step 1/1 040 -Steam Line Rupture -Excessive Heat Transfer

• 2.4 -Emergency Procedures I Plan:

• 2.4.1 -Knowledge of EOP entry conditions and immediate action steps. 4.8 None Given plant conditions and the EOP-0 diagnostic flow chart determine the correct procedure to implement.

55.43(b)(5) 0Bank ID Modified IZ! New D Memory/Fundamental IZ! Comprehension/

Analysis N/ A -new question EOP-0, POST-Trip Immediate Actions EOP-4, Excess Steam Demand Event None Page 8 of 55 Rev.6 r----------------------------------------------1 2014 NRC SRO EXAM MASTER KEY 178.055-Station Blackout A station blackout has occurred, EOP-0, Post-Trip Immediate Actions has been completed and the appropriate Optimal Recovery procedure implemented.

Which ONE of the following statements meet the procedure requirement(s) to exit this procedure?

A. The Safety Function Status Check Final Acceptance Criteria are being met permitting transition to OP-4 "Plant Shutdown from Power Operation to Hot Standby" B. Both 4KV Vital buses and associated load centers have been re-energized permitting transition to EOP-2 "Loss of Offsite Power I Loss of Forced Circulation" C. One 4KV Vital bus and associated load centers has been re-energized permitting transition to OP-4 "Plant Shutdown from Power Operation to Hot Standby" D. The Tech Support Center determines Final Safety Function Acceptance Criteria is too conservative and recommends transition to OP-4 "Plant Shutdown from Power Operation to Hot Standby" and AOP-3F "Loss Of Offsite Power While In Modes 3, 4, 5 Or 6" in parallel Answer: A Answer Explanation:

A. Correct -All of the Optimal Recovery Procedure require the Safety Function Status Check Final Acceptance Criteria are met. Additionally, EOP-2 may be an acceptable Optimal Recovery Procedure to transition to. B. Incorrect

-All of the Optimal Recovery Procedure require the Safety Function Status Check Final Acceptance Criteria are met. Information is not provided on the status of meeting the Safety Function Status Check Final Acceptance Criteria.

C. Incorrect

-All of the Optimal Recovery Procedure require the Safety Function Status Check Final Acceptance Criteria are met. Information is not provided on the status of meeting the Safety Function Status Check Final Acceptance Criteria.

D. Incorrect-All of the Optimal Recovery Procedure require the Safety Function Status Check Final Acceptance Criteria are met. Implementing the Tech Support Center's recommendations is not in accordance with the exit strategies outlined in EOP-7 and would require a procedure deviation be authorized by the Shift Manager Page 9 of 55 Rev.6 '

2014 NRC SRO EXAM MASTER KEY Topic: Tier/Group:

KIA Info: SRO Importance:

Proposed references to be provided to applicant:

Learning Objective:

10 CFR Part 55 Content: Question source: Cognitive level: Last NRC Exam used on: Technical references:

Comments:

Question 78 (Q28785) Exit EOP-7 111 055 -Station Blackout

• 2.1 -Conduct of Operations:

• 2.1.23 -Ability to perform specific system and integrated plant procedures during all modes of plant operation.

4.4 None Recall the condition when EOP-7 may be exited. 55.43(b)(5)

[gJ Bank I D Modified IDNew [gJ Memory/Fundamental D Comprehension/

Analysis No record of use on a previous NRC exam EOP-7, Station Blackout None Page 10 of 55 Rev. 6 2014 NRC SRO EXAM MASTER KEY 179. 062-Loss ofNuclear Service Water Unit-2 is at 100% power with the following indications:

• Service Water (SR W) Head Tank low level alarms

• Turbine Bldg sump alarm on 2T22

• Both SRW Head Tanks levels are approximately 40 inches and lowering at approximately 8 inches/min After the correct procedure is implemented which ONE ofthe following (1) actions is required and (2) why is this action performed?

A. (1) Trip U-2 Reactor and turbine and implement EOP-0, Post-Trip Immediate Actions (2) Reduce heat load to keep CAC's and DIG's available B. (1) Shut Turbine Bldg. SRW Isolation Valves lAW AOP-7B, Loss OfSRW (2) Isolate leak to keep CAC's and DIG's available C. (1) Open SRW Head tank level control bypass valves lAW AOP-7B, Loss OfSRW (2) To prevent SRW pumps from cavitating D. (1) Commence power reduction per OP-3, Normal Power Operation (2) Minimize plant transient if trip is required Answer: B Answer Explanation:

A. Incorrect

-Tripping the reactor is a required step but reason is to prevent damage to components cooled by SRW, not to keep CAC's and DIG's available since TB should be isolated from AB before tripping reactor. B. Correct -Indications given are rapidly lowering with indication of system rupture. Shutting TB isolation is step V.D.2.a and basis is to separate non safety related TB from safety related AB where CAC's and DIG's are supplied from. C. Incorrect-With rapidly lowering SRW Head Tank opening level control bypass is unnecessary because level will be empty by the time this action can be completed and excessive leakage flow is continuing D. Incorrect-Power reduction may be performed but intent is to reduce heat load on SRW system. SRO question because the student must determine correct AOP and the proper section of that AOP to determine correct actions. Also the student must determine the reason for performing this step. Page 11 of 55 Rev.6 2014 NRC SRO EXAM MASTER KEY Topic: Tier/Group:

KIA Info: SRO Importance:

Proposed references to be provided to applicant:

Learning Objective:

10 CFR Part 55 Content: Question source: Cognitive level: Last NRC Exam used on: Technical references:

Comments:

Question 79 (Q97194) Actions necessary for a leak on 21 SRW header 111 062-Loss ofNuclear Service Water

• AA2 -Ability to determine and interpret the following as they apply to the Loss of Nuclear Service Water:

• AA2.03 -The valve lineups necessary to restart the SWS while bypassing the portion of the system causing the abnormal condition 2.9 None Given loss of Service Water plant conditions in Modes 1 & 2 determine the most appropriate actions. 55.43(b)(5) 0Bank ID Modified D Memory/Fundamental Comprehension/

Analysis N/ A -new question AOP-7B, Loss of Service Water AOP-7B, Loss of Service Water Technical Basis None Page 12 of 55 Rev.6 2014 NRC SRO EXAM MASTER KEY 80. CE/E02 -Reactor Trip -Stabilization

-Recovery A low flow Reactor trip has just occurred due to 11 A RCP experiencing a locked rotor. The following conditions also exist during EOP-0:

• TBVs modulating RCS temperature in normal range

• 11 SG level is -80 inches and rising

• 12 SG level is -120 inches and slowly rising

• liB RCP Bentley Nevada indicates 0 mils, the OK light is out and the Danger light is lit

• RCS pressure is 2000 PSIA and rising

• PZR level is 107 inches and slowly rising Which ONE ofthe following EOPs should be implemented upon exiting EOP-0? A. EOP-1, Reactor Trip B. EOP-2, Loss ofOffsite Power/Loss of Forced Circulation C. EOP-6, Steam Generator Tube Rupture D. EOP-8, Functional Recovery Procedure Answer: A Answer Explanation:

The pump configuration of one RCP in 11 loop and two RCPs in 12 loop have created a S/G level mismatch.

The vibration information given for the liB RCP is trip criteria for the pump in 01-lA. The operator would be expected to trip the RCP during EOP-0, yielding a final pump configuration of no RCPs in 11 loop and two RCPs in 12 loop. A. Correct -The EOP that would be implemented next would be EOP-1 since the only complications were the RCPs before and after the trip and all safety functions are met. B. Incorrect-EOP-2 is implemented during a loss of all forced circulation.

Since both 12 loop RCPs are still operating, natural circulation does not exist and EOP-2 is not desired. C. Incorrect

-SG level mismatch is due to RCP configuration and not a tube leak. 'C' could be picked if the diagnostic flowchart was improperly evaluated as HR not met with feed flow and RCPs tripped. D. Incorrect-The conditions stated do not lead to meeting any of the entry criteria for implementation of the Functional Recovery Procedure.

Page 13 of 55 Rev.6

.-------------------------------------------, 2014 NRC SRO EXAM MASTER KEY Topic: Tier/Group:

KIA Info: SRO Importance:

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Learning Objective:

10 CFR Part 55 Content: Question source: Cognitive level: Last NRC Exam used on: Technical references:

Comments:

Question 80 (Q95235) EOP Transition with both IIAIIIB RCP secured Ill CE/E02 -Reactor Trip -Stabilization

-Recovery

• 2.4-Emergency Procedures I Plan:

• 2.4.31 -Knowledge of annunciator alarms, indications, or response procedures 4.1 None Given plant conditions and the EOP-0 diagnostic flow chart determine the correct procedure to implement.

55.43(b)(5)

C8J Bank I D Modified IDNew D Memory/Fundamental I C8J Comprehension/

Analysis No record of use on a previous NRC exam OI-IA Reactor Coolant System and Pump Operations; 1C06-ALM RCS Control Alarm Manual; EOP-0 Post-Trip Immediate Actions None Page 14 of 55 Rev.6

,--------------------------------------------------------------------------------------

2014 NRC SRO EXAM MASTER KEY 181. CE/E06-Loss ofFeedwater U-1 has experienced a reactor trip. Given the following conditions at the completion ofEOP-0:

• Loss of Offsite Power occurred

• 11 & 24 4 KV buses faulted

• 12 ADV cannot be opened

• 11 AFW Pump OOS for overhaul

• 12 AFW Pump tripped and cannot be reset Which ONE of the following procedures should be implemented on U-1? A. EOP-2, Loss ofOffsite Power/Loss ofForced Circulation B. EOP-3, Loss of All Feedwater C. EOP-4, Excess Steam Demand Event D. EOP-8, Functional Recovery Procedure Answer: B Answer Explanation:

A. Incorrect

-May be chosen if it is not recognized 13 & 23 AFW power supplies are deenergized due to loss of the specified buses B. Correct-Conditions presented result in a Loss of All Feedwater condition.

EOP-3, Loss of All Feedwater is written to address a Loss ofOffsite Power condition coincident with the Loss of All Feedwater.

C. Incorrect-Plausible since 12 ADV not opening may cause safety valves to open which could cause an ESDE given an equipment malfunction.

D. Incorrect

-Plausible if indications provided are misinterpreted as indicating multiple events (ESDE and LOAF) are occurring.

IF this were true, EOP-8, Functional Recovery Procedure, would be appropriate.

Page 15 of 55 Rev. 6 1 2014 NRC SRO EXAM MASTER KEY Topic: Tier/Group:

KIA Info: SRO Importance:

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10 CFR Part 55 Content: Question source: Cognitive level: Last NRC Exam used on: Technical references:

Comments:

Question 81 (Q97156) Select the appropriate EOP for stated conditions 1/1 CE/E06-Loss ofFeedwater

• EA2 -Ability to determine and interpret the following as they apply to the (Loss ofFeedwater):

• EA2.1 -Facility conditions and selection of appropriate procedures during abnormal and emergency operations 3.9 None Given plant conditions and the EOP-0 diagnostic flow chart determine the correct procedure to implement.

55.43(b)(5) 0Bank I D Modified I [g) New D Memory/Fundamental I [g) Comprehension/

Analysis No previous NRC Exam use EOP-0, Post-Trip Immediate Actions and associated Technical Basis document.

None Page 16 of 55 Rev.6 2014 NRC SRO EXAM MASTER KEY 182. 003 -Dropped Control Rod Given U-1 operating at 50% Beginning of Cycle, a malfunction occurs with the following indications:

• TcoLD 537°F

• Gross electrical output dropped by 40 MWe

• Linear Range "A" NI Power 4 7%

• Linear Range "C" NI Power 43% Which ONE of the following is (1) correct procedure to implement and (2) action required by this procedure?

A. (1) AOP-1B "CEA MALFUNCTION" (2) Lower turbine load to restore T coLo to program B. (1) AOP-7K "OVERCOOLING EVENT IN MODE 1 OR 2" (2) Lower turbine load to restore T coLD to program C. (1) AOP-1B "CEA MALFUNCTION" (2) Trip U-1 Reactor due to exceeding trip criteria D. (1) AOP-7K "OVERCOOLING EVENT IN MODE 1 OR 2" (2) Shut affected Turbine Bypass Valve upstream isolation valve Answer: A Answer Explanation:

A. Correct-Indications given in stem (T COLD 3 °F low, 4% differential in power indications) indicate a dropped CEA so AOP-1B is correct procedure to implement.

Lowering turbine load is the correct action to take (no trip criteria met yet) B. Incorrect-All of the indications except the difference in NI channel power indicate an overcooling event but all indication are consistent with a dropped CEA. C. Incorrect-Indications given in stem (TcoLD 3°F low, 4% differential in power indications) indicate a dropped CEA so AOP-1B is correct procedure to implement.

Lowering turbine load is the correct action to take (no trip criteria met yet) D. Incorrect-All of the indications except the difference in NI channel power indicate an overcooling event but all indication are consistent with a dropped CEA. If student decides an overcooling event has occurred isolating a leaking TBV would be correct. Page 17 of 55 Rev. 6 2014 NRC SRO EXAM MASTER KEY Topic: Tier/Group:

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Learning Objective:

10 CFR Part 55 Content: Question source: Cognitive level: Last NRC Exam used on: Technical references:

Comments:

Question 82 (Q40688) Response to a dropped CEA from 50% power 1/2 003 -Dropped Control Rod

• AA2 -Ability to determine and interpret the following as they apply to the Dropped Control Rod:

• AA2.03-Dropped rod, using in-core/ex-core instrumentation, in-core or loop temperature measurements.

3.8 None Given a CEA malfunction, identify the basis and determine the appropriate actions to mitigate the event lAW procedures.

55.43(b)(5) 0Bank cgj Modified 0New D Memory/Fundamental No previous NRC Exam use AOP-1B, CEA Malfunction AOP-1B Basis None I cgj Comprehension/

Analysis Page 18 of 55 Rev.6 2014 NRC SRO EXAM MASTER KEY js3. 024 -Emergency Boration 1-CVC-MOV-504 "RWT CHG PP SUCT" is tagged shut for maintenance when a fault initiates on 14 4KV bus. U-1 is manually tripped. You have been designated to review the TS and TRM implications from the events in the trip. Which ONE of the following describes the actions required, if any, to satisfy the TRM for Boration Flow Paths? A. No actions are required.

B. Log entry into and request a Functional Assessment for 15.0.3. C. Log that required boron injection flow path needs to be restored within 7 days. D. Log that required boron injection flow path needs to be restored within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Answer: D Answer Explanation:

A. Incorrect

-TRM 15 .1.2.A, requiring a flow path to be restored within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, is applicable.

B. Incorrect-With MOV-504 tagged shut, the RWT is not available as a boration flowpath.

When 14 4KV bus was lost, 12 Charging Pump, 12 BA Pump, and CVC-514 were lost due to no power. Two boration sources were available (11 BAST through MOV-509 and 12 BAST through MOV-508), but 11 and 13 Charging Pump were not powered from separate busses. Therefore TRM 15.1.2.A applied, requiring a flow path to be restored within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. C. Incorrect-With MOV-504 tagged shut, the RWT is not available as a boration flowpath.

When 14 4KV bus was lost, 12 Charging Pump, 12 BA Pump, and CVC-514 were lost due to no power. Two boration sources were available (11 BAST through MOV-509 and 12 BAST through MOV-508), but 11 and 13 Charging Pump were not powered from separate busses. Therefore TRM 15.1.2.A applied, requiring a flow path to be restored within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. D. Correct-With MOV-504 tagged shut, the RWT is not available as a boration flowpath.

When 14 4KV bus was lost, 12 Charging Pump, 12 BA Pump, and CVC-514 were lost due to no power. Two boration sources were available (11 BAST through MOV-509 and 12 BAST through MOV-508), but 11 and 13 Charging Pump were not powered from separate busses. Therefore TRM 15.1.2.A applied, requiring a flow path to be restored within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Page 19 of 55 Rev.6 2014 NRC SRO EXAM MASTER KEY Topic: Tier/Group:

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Learning Objective:

10 CFR Part 55 Content: Question source: Cognitive level: Last NRC Exam used on: Technical references:

Comments:

Question 83 (Q92671) TRM implications of loss of power 112 024 -Emergency Boration

• 2.2 -Equipment Control:

• 2.2.40-apply Technical Specifications for a system . 4.7 N 0-TRM, Technical Requirements Manual pages 1, 12-1 7 Analyze the effect of a loss of MCC-1 04 on the ability to borate. Using Tech Specs and the TRM determine if a given set of conditions satisfy the requirements for Boration Flow Paths and Borated Water Sources. 55.43(b)(2)

IZ! Bank ID Modified IDNew D Memory/Fundamental Comprehension/

Analysis No previous NRC Exam use NO-TRM, Technical Requirements Manual None Page 20 of 55 Rev.6 2014 NRC SRO EXAM MASTER KEY 184. 03 7 -Stearn Generator Tube Leak Given the following:

• A Stearn Generator Tube Rupture has occurred

• S lAS has actuated

• RCS pressure has stabilized at 1200 PSIA

• The affected S/G has been isolated

• You direct the RO to perform the next step of the Optimal Recovery procedure "DEPRESSURIZE THE RCS TO REDUCE SUBCOOLING AND MAINTAIN PRESSURIZER LEVEL" Which ONE of the following describes the optimal RCS Pressure/subcooling band you should direct the RO to maintain per the Optimal Recovery Procedure Basis and why? A. RCS pressure above RCP curve but less than 60°F subcooling; Minimize leakage from RCS to S/G and ensure forced circulation maintained B. RCS pressure above RCP curve but less than 140°F subcooling; Ensure forced circulation maintained and subcooled fluid surrounding the core C. 25 to 140°F subcooling margin; Ensures RCS remains subcooled and prevents (PTS) pressurized thermal shock D. 25-50°F subcooling margin; Minimize leakage from RCS to S/G Answer: A Answer Explanation:

A. Correct-Keeping Subcooling in this band maintains RCP's running and minimizes RCS leak rate B. Incorrect

-Because it allows high subcooling which allows higher RCS leak rates C. Incorrect

-Because with given conditions one RCP in each loop is running and 25°F subcooling is below RCP pump curve D. Incorrect-25-140°F of subcooling represents the limits specified in EOP-6, however, the basis emphasizes maintaining subcooling low in band to minimize RCS leak rate SRO only justified per 10CFR55.43(b)(5) per NRC guidance document for SRO questions.

Does the question involve one or more of the following?

Recalling what strategy or action is written into a plant procedure, including when the strategy or action is required.

Page 21 of 55 Rev.6 2014 NRC SRO EXAM MASTER KEY Topic: Tier/Group:

KIA Info: SRO Importance:

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Learning Objective:

10 CFR Part 55 Content: Question source: Cognitive level: Last NRC Exam used on: Technical references:

Comments:

Question 84 (Q97186) Maintaining SCM with a SGTR 1/2 037-Steam Generator Tube Leak

• AA2 -Ability to determine and interpret the following as they apply to the Steam Generator Tube Leak:

• AA2.16 -Pressure at which to maintain RCS during S/G cooldown.

4.3 None Recall the strategy and the basis for the major actions performed in EOP-6, Steam Generator Tube Rupture, and what actions are required if safety functions are in jeopardy of being lost. 55.43(b)(5) 0Bank Jo Modified IZ] New D Memory/Fundamental J1Z1 Comprehension/

Analysis No previous NRC Exam use EOP-6, Steam Generator Tube Rupture; EOP-6, Steam Generator Tube Rupture Technical Basis None Page 22 of 55 Rev.6 2014 NRC SRO EXAM MASTER KEY iss. CE/E09 -Functional Recovery A reactor trip and SIAS has occurred on U-1. While performing EOP-0, the RO and CRO provide the following information to the CRS:

• 2 CEAs failed to insert and boration has commenced

• Pressurizer level is 70 inches and continuing to lower with all Charging pumps running and Letdown isolated

• RCS pressure is 1500 PSIA and steady

• T coLD is 532°F

• Both S/G levels are -20 inches and rising towards 0 inches with AFW flow

• Both S/G pressures are 900 PSIA and steady

• Containment pressure is 0.8 PSIG and steady; alternate actions performed

• Containment temperature is ll9°F and steady

• S/G Blowdown and Main Vent RMS are in alarm From the information provided, which safety function report is correct, and which procedure would you implement?

A. Reactivity Control, Core/RCS Heat Removal, and Containment Environment are Complete, implement EOP-6 B. Reactivity Control, Core/RCS Heat Removal, and Containment Environment are Complete, implementEOP-8 C. Rad Levels External to Containment and RCS Pressure/Inventory Control Cannot Be Met, implement EOP-6 D. Rad Levels External to Containment and RCS Pressure/Inventory Control Cannot Be Met, implement EOP-8 Answer: D Answer Explanation:

A. Incorrect-RC is MET (boration in progress), HR is MET and CE is NOT MET (Containment Pressure).

EOP-8 would be the appropriate procedure to enter because multiple events are in progress.

B. Incorrect-RC is MET (boration in progress), HR IS MET and CE is NOT MET (Containment Pressure)

C. Incorrect

-EOP-8 would be the appropriate procedure to enter because multiple events are in progress.

Page 23 of 55 Rev. 6

2014 NRC SRO EXAM MASTER KEY D. Correct-Safety Function status reports are correct for stated conditions.

EOP-8 would be the appropriate procedure to enter because multiple events are in progress.

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10 CFR Part 55 Content: Question source: Cognitive level: Last NRC Exam used on: Technical references:

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Question 85 (Q97157) EOP-8 Parameters and logic 112 CE/E09 Functional Recovery

• 2.4 -Emergency Procedures I Plan:

• 2.4.21 -Knowledge of the parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc. 4.6 None Given plant conditions, correctly report Safety Functions in EOP-0. 55.43(b )(5) 0Bank ID Modified 1[2] New D Memory/Fundamental I [2] Comprehension/

Analysis No previous NRC Exam use EOP-0, Post-Trip Immediate Actions; EOP-8, Functional Recovery Procedure None Page 24 of 55 Rev.6 I i 2014 NRC SRO EXAM MASTER KEY 186. 026 -Containment Spray System U-2 has the following indications:.

• A Design Basis Loss of Coolant Accident occurred 45 minutes ago

• 21 & 23 HPSI Pump amps are fluctuating

• 21 & 22 CS Pump amps are fluctuating

• HPSI flow has been throttled equally among the four headers to the minimum allowed Which ONE of the following describes:

(1) What actions are to be taken IA W applicable procedure for conditions above? (2) Impact of these actions? A. (1) Stop both HPSI Pumps and implement EOP-8 (2) Potential to exceed maximum containment pressure limit B. (1) Stop both CS Pumps and implement EOP-8 (2) Potential to exceed maximum containment pressure limit C. (1) Stop both HPSI Pumps and implement EOP-8 (2) Potential to raise NPSH to CS pumps D. (1) Stop both CS Pumps and implement EOP-8 (2) Potential to raise NPSH to HPSI pumps Answer: D Answer Explanation:

A. Incorrect

-The four Containment Air Coolers supply a 100% redundancy to the Containment Spray System in supplying 100% of the capacity required in cooling the Containment.

EOP-5, Loss of Coolant Accident, directs securing both CS Pumps. B. Incorrect

-The four Containment Air Coolers supply a 100% redundancy to the Containment Spray System in supplying 100% of the capacity required in cooling the Containment.

Given that HPSI has been throttled to the minimum allowed flow, securing the CS Pumps is the next step in the procedure followed by checking the HPSis for acceptable performance and then implementation ofEOP-8. C. Incorrect

-EOP-5, Loss of Coolant Accident, directs securing both CS pumps then checking the performance of the HPSI Pumps. Securing both HPSI pumps is directed later in procedure if securing CS PP's did not cause SAT HPSI operation.

D. Correct-EOP-5, Loss of Coolant Accident, directs securing both CS pumps then checking the performance of the HPSI Pumps. Page 25 of 55 Rev.6 2014 NRC SRO EXAM MASTER KEY Topic: Tier/Group:

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10 CFR Part 55 Content: Question source: Cognitive level: Last NRC Exam used on: Technical references:

Comments:

Question 86 (Q97191) Diagnosing/responding 2/1 026-Containment Spray System (CSS)

• A2-Ability to (a) predict the impacts of the following malfunctions or operations on the CSS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

• A2.07 -Loss of containment spray pump suction when in recirculation mode, possibly caused by clogged sump screen, pump inlet high temperature exceeded cavitation, voiding), or sump level below cutoff (interlock) limit 3.9 None Given RCS parameters, identify appropriate Safety Injection System response.

55.43(b)(5) 0Bank I D Modified D Memory/Fundamental No previous NRC Exam use EOP-5 EOP-5 Basis Document None Comprehension/

Analysis Page 26 of 55 Rev.6 2014 NRC SRO EXAM MASTER KEY 187. 039 -Main Steam and Reheat U-2 was operating steady state at 100% when 21 Main Steam Line Radiation Monitor comes into alarm and is validated.

AOP-2A "Excessive Reactor Coolant Leakage" has been implemented.

Which combination below of(1) plant condition and (2) action from appropriate section of AOP-2A is correct? A. (1) Pressurizer level rising slowly with LID at minimum (2) Borate to reduce TAvE to< 537°F B. (1) Pressurizer level lowering until 2nd Charging Pump starts (2) Borate to reduce T AVE to < 53 7°F C. (1) Pressurizer level rising slowly with LID at minimum; (2) Commence an expeditious power reduction; maintain S/G pressure 800-825 PSIA D. (1) Pressurizer level lowering until 2nd Charging Pump starts (2) Commence an expeditious power reduction; maintain S/G pressure 800-825 PSIA Answer: B Answer Explanation:

A. Incorrect-PZR level rising with LID at min indicates

< capacity of 1 charging pump.Section V of AOP-2A does not direct borating to 537°F TAVE* B. Correct-PZR level lowering with LID at minimum indicates

> capacity of 1 charging pump. Sect VI of AOP-2A directs borating to 53 7°F T AVE* C. Incorrect

-PZR level rising with LID at min indicates

< capacity of 1 charging pump.Section V of AOP-2A does not direct maintaining S/G press 800-825 PSIA. D. Incorrect

-PZR level lowering with LID at minimum indicates

> capacity of 1 charging pump. Sect VI of AOP-2A directs borating to 537°F TAvE not an expeditious power reduction.

Page 27 of 55 Rev.6 2014 NRC SRO EXAM MASTER KEY Topic: Tier/Group:

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10 CFR Part 55 Content: Question source: Cognitive level: Last NRC Exam used on: Technical references:

Comments:

Question 87 (Q97182) Steam Generator Tube Leak response 2/1 039-Main Steam and Reheat

• A2-Ability to (a) predict the impacts of the following malfunctions or operations on the MRSS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

• A2.03 -Indications and alarms for Main Steam and area radiation monitors (during SGTR) 3.7 None Determine the plant parameter criteria evaluated to identify the following RCS leaks: * >the capacity of One Charging Pump * <the capacity of One Charging Pump 55.43(b )(5) 0Bank jo Modified 0 Memory/Fundamental I Comprehension/

Analysis No previous NRC Exam use AOP-2A, Excessive Reactor Coolant Leakage None Page 28 of 55 Rev.6 2014 NRC SRO EXAM MASTER KEY 88. 061 -Auxiliary/Emergency Feedwater Unit 1 is at 75% power when two of three (2 of3) Condensate Pumps trip. The TBO reports the Condensate Header has ruptured, spraying water near the Condensate Booster Pumps. Which ONE of the following strategies is employed to combat this condition?

A. Perform a Rapid Power Reduction to remove U-1 from service, secure Main Feed and Condensate systems and initiate AFW flow at <1% reactor power lAW AOP-3G, Malfunction Of Main Feedwater System. B. Secure remaining Condensate Pump, trip reactor prior to RPS Low S/G Level auto trip, implement EOP-0, Post-Trip Immediate Actions. C. Trip the Reactor, verify Reactivity Control, secure the Main Feed and Condensate systems, start an AFW pump, and continue with EOP-0, Post-Trip Immediate Actions. D. Commence a Rapid Power Reduction to lower Condensate Header flow to< 8000 GPM and direct TBO to isolate leak, if possible, lAW AOP-3G, Malfunction Of Main Feedwater System. Answer: C Answer Explanation:

A. Incorrect-AOP-3G directs tripping reactor, performing reactivity control safety function, securing the Main feed and condensate systems, starting an AFW pump and continuing with EOP-0. B. Incorrect-AOP-3G directs tripping reactor first and performing reactivity control safety function then securing main feed and condensate systems, starting an AFW pump and continuing with EOP-0. C. Correct -IA W AOP-3G-1, Sect. VIII, these actions are all required in this order. D. Incorrect-AOP-3G directs tripping reactor first and performing reactivity control safety function then securing main feed and condensate systems, starting an AFW pump and continuing with EOP-0. Page 29 of 55 Rev.6 2014 NRC SRO EXAM MASTER KEY Topic: Tier/Group:

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10 CFR Part 55 Content: Question source: Cognitive level: Last NRC Exam used on: Technical references:

Comments:

Question 88 (Q28769) AOP-3G Actions 2/1 061 -Auxiliary/Emergency Feedwater

• 2.4 -Emergency Procedures I Plan

• 2.4.6 -Knowledge of EOP mitigation strategies 4.7 None 55.43(b)(5)

IZ! Bank D Modified jDNew IZ! Memory/Fundamental No previous NRC Exam use EOP-3, Loss of All Feedwater None I D Comprehension/

Analysis Page 30 of 55 Rev.6 2014 NRC SRO EXAM MASTER KEY 189. 064-Emergency Diesel Generator Given the following conditions following a Loss of Offsite Power (LOOP):

• lA, lB, & OC DIG's cannot be aligned due to 11 and 14 4KV buses faulted

• Repairs to 11 4KV Bus are projected to take 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />

• Repairs to 14 4KV Bus are projected to take 30 minutes

• Unit-1 RCS temperature is currently at 350°F

• A plant heatup to hot standby was in progress on Unit-1 prior to LOOP

• 4KV Bus 21 has been repowered from the 2A DG

• 4KV Bus 24 can be repowered from the OC DG as soon as the disconnects are shifted Which ONE of the following EAL classifications is appropriate for the stated conditions?

A. General Emergency B. Site Area Emergency C. Alert D. Unusual Event Answer: B Answer Explanation:

A. Incorrect-With 14 4KV Bus expected to return in 0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> conditions are met for a SAE not a General Emergency.

A General Emergency declaration would be appropriate ifUnit-1 were to be without power to a vital4KV Bus for greater than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. B. Correct-With 14 4KV Bus expected to return in 0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> conditions are met for declaration of a Site Area Emergency for Unit-1. C. Incorrect

-Unit-2 is one power supply loss away from a Station Blackout which meets Alert declaration criteria.

However, there is no power to either Unit-1 4KV bus therefore declaration of a Site Area Emergency is appropriate, D. Incorrect-Unit-2 is meeting the conditions for Alert declaration.

If the OC DG were pre-aligned to 24 4KV Bus, Unit-2 would be meeting the criteria for an Unusual Event declaration.

However, Unit-1 conditions require declaring a SAE due to higher EAL category.

Page 31 of 55 Rev.6 2014 NRC SRO EXAM MASTER KEY Topic: Tier/Group:

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10 CFR Part 55 Content: Question source: Cognitive level: Last NRC Exam used on: Technical references:

Comments:

Question 89 (Q15827) Evaluate 4KV Bus status for EAL Declaration purposes 2/1 064 -Emergency Diesel Generator

• 2.4-Emergency Procedures I Plan

• 2.4.41 -Knowledge of the EAL thresholds and classifications 4.6 EAL Hot and Cold Charts Given any CCNPP event, use the ERPIP and Attachments to respond to the event and take appropriate actions. 55.43(b)(5)

[8J Bank jo Modified jDNew D Memory/Fundamental

[8J Comprehension/

Analysis No previous NRC Exam use ERPIP-3.0, Immediate Actions None Page 32 of 55 Rev. 6

,----------------------------------------------------------------------------------, 2014 NRC SRO EXAM MASTER KEY 190. 078 -Instrument Air U-1 is in Mode 5 with SDC in-service when a malfunction occurs on the in-service instrument air dryer. Which ONE ofthe following describes (1) which procedure should be implemented to address all plant effects ofthis malfunction and (2) potential impacts to Shutdown Cooling from this malfunction?

A. AOP-3B, Loss of SDC; SI-306-CV fails open, SI-657-CV fails closed B. AOP-7D, Loss oflnstrument Air; SI-306-CV fails open, SI-657-CV fails closed C. AOP-3B, Loss of SDC; SI-306-CV fails closed, SI-657-CV fails open D. AOP-7D, Loss oflnstrument Air; SI-306-CV fails closed, SI-657-CV fails open Answer: B Answer Explanation:

A. Incorrect-Loss of Instrument Air causes SI-306-CV to fail open and SI-657-CV fails shut. AOP-3B has steps for most malfunctions that could affect SDC (loss of power, leaks, etc.) but not a loss of instrument air. Therefore AOP-7D is required to correct an air dryer malfunction.

B. Correct-Loss oflnstrument Air causes SI-306-CV to fail open and SI-657-CV fails shut. AOP-3B has steps for most malfunctions that could affect SDC (loss of power, leaks, etc.) but not a loss of instrument air. Therefore AOP-7D is required to correct an air dryer malfunction.

C. Incorrect-Loss oflnstrument Air causes SI-306-CV to fail open and SI-657-CV fails shut. AOP-3B has steps for most malfunctions that could affect SDC (loss of power, leaks, etc.) but not a loss of instrument air. Therefore AOP-7D is required to correct an air dryer malfunction.

D. Incorrect-Loss oflnstrument Air causes SI-306-CV to fail open and SI-657-CV fails shut. AOP-7D is required to correct an air dryer malfunction.

Page 33 of 55 Rev.6 2014 NRC SRO EXAM MASTER KEY Topic: Tier/Group:

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10 CFR Part 55 Content: Question source: Cognitive level: Last NRC Exam used on: Technical references:

Comments:

Question 90 (Q97189) Determine Air Dryer malfunction effects and correct AOP 211 078 -Instrument Air

• A2 -Ability to (a) predict the impacts of the following malfunctions or operations on the lAS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

• A2.01 -Air dryer and filter malfunctions 2.9 None Given AOP-7D, determine operator actions for a loss of Instrument Air in Modes 3, 4, 5, 6 or defueled 55.43(b)(5) 0Bank D Modified 1:8:1 New 0 Memory/Fundamental ji:8J Comprehension/

Analysis No previous NRC Exam use AOP-3B, Abnormal Shutdown Cooling Conditions AOP-7D, Loss of Instrument Air None Page 34 of 55 Rev.6 2014 NRC SRO EXAM MASTER KEY 191. 015-Nuclear Instrumentation U-1 is operating at 100% power when Channel B LRNI Power Summer failure occurs. Associated RPS Trip Units have been bypassed within required time. Which ONE of the following describes (1) which RPS T/U' s are affected and (2) how long until Channel B must be restored or additional actions must be taken. A. (1) 1-High Power, 2-High SUR, 7-TM/LP, '8-APD and 10-Loss of Load (2) 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> B. (1) 1-High Power, 2-High SUR, 7-TM/LP, 8-APD and 10-Loss of Load (2) 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> C. (1) 1-High Power, 7-TM/LP and 10-Loss of Load only (2) 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> D. (1) 1-High Power, 7-TM/LP and 10-Loss of Load only (2) 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> Answer: B Answer Explanation:

A. Incorrect

-TS 3.3 .1 allows 48 hrs to restore B. Correct-T/U's 1,2,7,8,10 affected, and TS 3.3.1 allows 48 hrs to restore C. Incorrect-T/U 2 & 8 are affected and must also be bypassed and TS 3.3.1 allows 48 hrs to restore D. Incorrect-T/U 2 & 8 are affected and must also be bypassed Page 35 of 55 Rev.6 2014 NRC SRO EXAM MASTER KEY Topic: Tier/Group:

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10 CFR Part 55 Content: Question source: Cognitive level: Last NRC Exam used on: Technical references:

Comments:

Question 91 (Q97184) Evaluate alarms and indications 2/2 015 -Nuclear Instrumentation

• 2.1 -Conduct of Operations

• 2.1. 7 -Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation 4.7 None Given CNG-NL-1.01-1004 and/or Tech Specs, determine whether LCO for RPS operation is met and identify required actions. 55.43(b)(2) 0Bank ID Modified D Memory/Fundamental Comprehension/

Analysis No previous NRC Exam use N0-1-200, Control of Shift Activities Tech Spec 3.3.1 None Page 36 of 55 Rev. 6 2014 NRC SRO EXAM MASTER KEY 192. 056 -Condensate System Unit-1 is operating at 100% power when multiple condenser tube failures occur in 12A Waterbox:

• A power reduction was begun.

• At 90% power 12A Waterbox is secured.

• Plant Chemistry determines S/G chemistry is in Action Level 3 for cation conductivity

• It is now 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> since the tube failures, power is at 40%. S/G chemistry is in Action level 2 and improving

• Action Level 1 is expected in 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> at the current cleanup rate (1) Predict the impact if no actions are taken for condenser tube failure and; (2) What action should be directed IA W AOP-1 0, Abnormal Secondary Chemistry Conditions?

A. (1) Potential breach of 1 fission product barrier; (2) Place the unit in Mode 2 as quickly as safe operation permits. B. (1) Potential breach of 1 fission product barrier; (2) Reduce power to less than 30% within the next 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. C. (1) Potential breach of 2 fission product barriers; (2) Terminate the power reduction when Action Level 1 is reached. D. (1) Potential breach of2 fission product barriers; (2) Reduce power to less than 30% then when Action Level 1 is reached, return to 100% power. Answer: A Answer Explanation:

A. Correct-Condenser tube leak can cause a S/G tube leak if no action taken. AOP-1 0 states: IF Plant Chemistry determines SG Chemistry is in Action Level 3, as a result of a Condenser tube leak, THEN commence an orderly plant shutdown to be less than 5% power as quickly as safe operation permits PER OP-3 and OP-4. A note preceding this step in AOP-10 states: If SG chemistry levels are reduced below the Action Level 3 value, before or during the power reduction, power level is still required to be reduced below 5%. B. Incorrect

-Condenser tube leak can cause a S/G tube leak if no action taken. This action is correct if Action Level2 was declared at the time ofthe tube failure. AOP-10 states: Within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of initiating Action Level2, reduce power to less than 30% PER OP-3, NORMAL POWER OPERATION.

Page 37 of 55 Rev.6 2014 NRC SRO EXAM MASTER KEY C. Incorrect-Condenser tube leak can cause a S/G tube leak if no action taken, this only affects 1 fission product barrier. 2 barriers is plausible if student thinks ADV operation bypasses containment barrier or abnormal chemistry could enter RCS affecting cladding barrier. This action would be appropriate if the Unit had peaked out in an Action Level 2 condition and Action Level 1 values have been attained.

Entry into Action Level 3 requires reducing power to less than 5%. D. Incorrect-Condenser tube leak can cause a S/G tube leak if no action taken, this only affects 1 fission product barrier. 2 barriers is plausible if student thinks ADV operation bypasses containment barrier or abnormal chemistry could enter RCS affecting cladding barrier. This action would be appropriate if the Unit had peaked out in an Action Level2 condition.

Entry into Action Level 3 requires reducing power to less than 5%. Page 38 of 55 Rev.6 2014 NRC SRO EXAM MASTER KEY Topic: Tier/Group:

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10 CFR Part 55 Content: Question source: Cognitive level: Last NRC Exam used on: Technical references:

Comments:

Question 92 (Q41948) Condenser Tube Leak 2/2 056 -Condensate System

• A2-Ability to (a) predict the impacts of the following malfunctions or operations on the Condensate system; and (b) based on those predictions, use procedures to correct, control or mitigate the consequences of those malfunctions or operations:

• A2.05-Condenser Tube Leakage 2.5* None Recall required actions for chemistry levels 55.43(b)(5)

ID Modified IDNew 0 Memory/Fundamental Comprehension/Analysis No previous NRC Exam use AOP-1 0 Abnormal Secondary Chemistry None Page 39 of 55 Rev.6 2014 NRC SRO EXAM MASTER KEY 193. 072 -Area Radiation Monitoring U-1 containment pressure is -0.4 psig and preparing for a negative pressure vent. Containment High Range Radiation Monitor, 1-RI-5317 A, reading is erratic due to a power supply problem. Which ONE of the following describes (1) the impact I mitigating strategy of removing this rad monitor from service, and (2) Tech Spec implications of this failure (if applicable)?

A. (1) Negative pressure vent prevented (2) No TS implications B. (1) Negative pressure vent prevented (2) 30 days to return to service per TS C. (1) Negative pressure vent allowed after verifying RI-5317B in service (2) No TS implications D. (1) Negative pressure vent allowed after verifying RI-5317B in service (2) 30 days to return to service per TS Answer: D Answer Explanation:

A. Incorrect-Negative pressure vent Initial Conditions require only one Containment High Range Rad Monitor, butTS 3.3.10 requires two Containment High Range Rad Monitor with 30 days to return B. Incorrect-Negative pressure vent Initial Conditions require only one Containment High Range Rad Monitor, butTS 3.3.10 requires two Containment High Range Rad Monitor with 30 days to return C. Incorrect-Negative pressure vent Initial Conditions require only one Containment High Range Rad Monitor, butTS 3.3.10 requires two Containment High Range Rad Monitor with 30 days to return D. Correct-Negative pressure vent Initial Conditions require only one Containment High Range Rad Monitor, butTS 3.3.10 requires two Containment High Range Rad Monitor with 30 days to return Page 40 of 55 Rev. 6 2014 NRC SRO EXAM MASTER KEY Topic: Tier/Group:

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Comments:

Question 93 (Q97187) Containment Hi-Range RMS OOS 2/2 072 -Area Radiation Monitoring

• A2-Ability to (a) predict the impacts of the following malfunctions or operations on the ARM; and (b) based on those predictions or mitigate the consequences of those malfunctions or operations:

• A2.01 -Erratic or failed power supply 2.9 None Describe the response to and the condition that will indicate a system failure of the Containment Hi-Range Radiation Monitor. 55.43(b)(5) 0Bank jo Modified IZ! New IZ! Memory/Fundamental No previous NRC Exam use OI-41B Hydrogen Purge Tech Spec 3.3.10 None I D Comprehension!

Analysis Page 41 of 55 Rev.6 2014 NRC SRO EXAM MASTER KEY 194. Conduct of Operations Unit 1 is in Mode 6 with movement of irradiated fuel in progress.

A safety tagger requests the Fuel Handling Supervisor (FHS) to independently verify a safety tagged component located on the 69' containment elevation.

Which ONE of the following describes the restrictions, if any, placed on the FHS performing the independent verification (IV)? A. The FHS may not perform the IV under any circumstances.

B. The FHS may perform the IV at any time during the movement of irradiated fuel. C. The FHS may perform the IV only during suspension of movement of irradiated fuel. D. The FHS may perform the IV only ifthe Refueling Machine Operator is a licensed reactor operator Answer: C Answer Explanation:

The FHS may leave the refueling machine if a licensed operator is the RFM operator.

The FHS may tour the 69' area of containment.

The FHS supervises refueling operations with no concurrent duties. A. Incorrect

-The designated FHS may perform the independent verification IF fuel handling activities have been suspended.

B. Incorrect-N0-1-200, Step 5.1.B.a. states that during fuel loading, transferring or handling an SRO shall be designated the FHS and further states: The FHS shall directly supervise from the Containment 69 foot elevation and have no other concurrent duties. Therefore, the IV cannot be performed by the FHS unless fuel handling activities are suspended.

C. Correct-N0-1-200, Step 5.1.B.a. states that during fuel loading, transferring or handling an SRO shall be designated the FHS and further states: The FHS shall directly supervise from the Containment 69 foot elevation and have no other concurrent duties. Therefore, the IV cannot be performed by the FHS unless fuel handling activities are suspended.

D. Incorrect-Having a licensed operator on the RFM allows the FHS to leave the RFM as long as he remains on the 69' elevation of the Containment and focused solely on the fuel handling activities.

IV cannot be performed by the FHS unless fuel handling activities are suspended.

Page 42 of 55 Rev.6 201'4 NRC SRO EXAM MASTER KEY Topic: Tier/Group:

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10 CFR Part 55 Content: Question source: Cognitive level: Last NRC Exam used on: Technical references:

Comments:

Question 94 (Q44770) Determine the limits of travel for the FHS Generic Knowledge

& Abilities 2.1 -Conduct of Operations

• 2.1.40 -Knowledge of refueling administrative requirements.

3.9 None Determine limits of travel for the FHS when the RFM is being operated.

55.43(b)(7)

/D Modified /DNew Memory/Fundamental I D Comprehension/

Analysis No previous NRC Exam use N0-1-200, Control of Shift Activities (Step 5.1.B.1.)

CNG-OP-1.-1-1000, Conduct of Operations (Step 4.8) None Page 43 of 55 Rev.6 -1 I I 2014 NRC SRO EXAM MASTER KEY 195. 2.1 -Conduct of Operations Given the following:

• BOTH Units are in Mode 1.

• The shift is manned to the Technical Specification minimum composition.

• The shift has 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> remaining.

• An RO has become ill and must leave the site for emergency medical treatment.

Which ONE of the following describes the requirements regarding the shift composition and the MINIMUM required action in this situation?

A. The RO may leave the site immediately after turnover of responsibilities to the CRS. A replacement must arrive within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. B. The RO may leave the site immediately after turnover of responsibilities to the CRS. A replacement must arrive within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. C. The RO may leave the site immediately after turnover of responsibilities to the CRO. A replacement must arrive within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. D. The RO may leave the site immediately after turnover of responsibilities to the CRO. A replacement must arrive within 1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. Answer: C Answer Explanation:

Tech Spec 5.2.2.c specifies: "Shift crew composition may be less than the minimum requirement of 10 CFR 50.54(m)(2)(i), 5.2.2.a, and 5.2.2.g for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements." A. Incorrect

-The RO must be relieved by the CRO not the CRS, CRS is a plausible distractor since he is in CR B. Incorrect-1 hr is incorrect, many of the Tech Specs and Tech Spec Surveillance requirements have completion times/Frequencies of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> making this a plausible distractor.

C. Correct -Relieved by CRO and 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> is the correct answer IA W TS. D. Incorrect-1 hr is incorrect, many of the Tech Specs and Tech Spec Surveillance requirements have completion times/Frequencies of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> making this a plausible distractor..

Page 44 of 55 Rev. 6 2014 NRC SRO EXAM MASTER KEY Topic: Tier/Group:

KIA Info: SRO Importance:

Proposed references to be provided to applicant:

Learning Objective:

10 CFR Part 55 Content: Question source: Cognitive level: Last NRC Exam used on: Technical references:

Comments:

Question 95 (Q54961) Administrative requirements for minimum shift staffing Generic Knowledge

& Abilities 2.1 -Conduct of Operations

• 2.1.5 -Ability to use procedures related to shift staffing, such as minimum crew complement, overtime limitations, etc. 3.8 None 55.43(b )(2) [gl Bank D Modified /DNew [gl Memory/Fundamental I D Comprehension/

Analysis No previous NRC Exam use Technical Specification 5.2.2.c for Unit Staff None Page 45 of 55 Rev.6 2014 NRC SRO EXAM MASTER KEY 196. 2.2 -Equipment Control Given the following:

• Unit-2 is operating at 100% power

• Following completion of shift turnover, at 1830 on a Saturday evening, the 2B DG is declared out of service due to problems with the fuel oil transfer pump.

• You have directed the 2B DG be removed from service. Which ONE of the following offsite organizations must you, as the CRS or Shift Manager, notify per OI-21B, 2B DIESEL GENERATOR?

A. The responsible Work Week Manager B. The Nuclear Regulatory Commission (NRC) C. Southern Maryland Electric Co-operative (SMECO) D. System Operator-Transmission Systems Operation (SO-TSO) Answer: D Answer Explanation:

A. The Work Week Manager must be notified to assist the OWC in completing an updated risk assessment, however, this notification is not required by OI-21B B. The NRC is not required to be notified ofDG unavailability per OI-21B. The Resident Inspector would be notified by voice-mail or page per CNG-OP-1.01-2001, Communications and Briefings C. SMECO is not required to be notified of DG unavailability.

D. OI-21B states: The CRS or SM will contact the SO-TSO when removing a D/G from service during the weekend. Page 46 of 55 Rev.6 2014 NRC SRO EXAM MASTER KEY Topic: Tier/Group:

K/A Info: SRO Importance:

Proposed references to be provided to applicant:

Learning Objective:

10 CFR Part 55 Content: Question source: Cognitive level: Last NRC Exam used on: Technical references:

Comments:

Question 96 (Q74876) Reporting requirement for DG OOS Generic Knowledge

& Abilities 2.2 -Equipment Control

• 2.2.17-Knowledge of the process for managing maintenance activities during power operations, such as risk assessments, work prioritization, and coordination with the transmission system operator.

3.8 None 55.43(b)(l)

D Modified 0New Memory/Fundamental I D Comprehension/

Analysis No previous NRC Exam use OI-21B, 2B DIESEL GENERATOR None Page 47 of 55 Rev.6 2014 NRC SRO EXAM MASTER KEY 197. 2.2-Equipment Control Given the following conditions:

• Unit-2 RCS temperature is 180°F

• 21 AFW Pump is out of service for repairs that will take 2 to 3 days to complete Which ONE of the following describes the HIGHEST OPERATIONAL MODE that may be entered with 21 AFW Pump out of service? A. Mode 5 B. Mode4 C. Mode 3 D. Mode 1 Answer: B Answer Explanation:

A. Incorrect

-Turbine Driven AFW Pumps are required operable in Modes 1, 2 & 3. Plausible if student believes no mode changes allowed due to misunderstanding of LCO 3.0.4 B. Correct-LCO 3.7.3.a is the applicable TS and contains a note stating LCO 3.0.4.b is not applicable C. Incorrect

-Tech Specs provide an allowance for proceeding to Mode 1/2/3 prior to performing Turbine Driven AFW Pump Surveillance Testing D. Incorrect-Tech Specs provide an allowance for proceeding to Mode 1/2/3 prior to performing Turbine Driven AFW Pump Surveillance Testing Page 48 of 55 Rev.6 2014 NRC SRO EXAM MASTER KEY Topic: Tier/Group:

KIA Info: SRO Importance:

Proposed references to be provided to applicant:

Learning Objective:

10 CFR Part 55 Content: Question source: Cognitive level: Last NRC Exam used on: Technical references:

Comments:

Question 97 (Q51274) Evaluate a Mode change with a TDAFW Pp OOS Generic Knowledge

& Abilities 2.2 -Equipment Control

• 2.2.35 -Ability to determine Technical Specification Mode of Operation 4.5 None ' Given the Tech Specs and a plant conditions, stste the applicable LCO 55.43(b)(2)

IZI Bank jo Modified jDNew 0 Memory/Fundamental No previous NRC Exam use Technical Specifications Technical Specification Bases None jiZI Comprehension/

Analysis Page 49 of 55 Rev.6 2014 NRC SRO EXAM MASTER KEY 198. 2.4-Emergency Procedures I Plan During AFW testing it was discovered that there is blockage in the U-2 AFW suction line upstream of the level indicator in 27Ft. Purge Supply Fan Room. An Auxiliary Operator has been sent for "for cause" fitness-for-duty testing for an unrelated event. Condition Reports have been written. Which ONE of the following is the soonest report required?

A. Four (4) hour Emergency Notification System (ENS) Report B. Eight (8) hour Emergency Notification System (ENS) Report C. Twenty-four (24) hour Emergency Notification System (ENS) Report D. Sixty (60) day Licensee Event Report (LER) Answer: B Answer Explanation:

A. Incorrect-A 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> ENS report, per 10 CFR 50.72(b)(2)(i), is not required since T.S. 3.7.3 Action F states "LCO 3.0.3 and all other LCO Required Actions requiting MODE changes are suspended until one AFW train is restored to OPERABLE status". B. Correct-An 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> report ENS is required, under 10 CFR 50.72(b)(3)(v), due to the blockage affecting the ability to removal residual heat. C. Incorrect -A 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> ENS report may be determined necessary per 10 CFR 26.719(b)(2) when results of fitness for duty screening are obtained.

However, this is not the soonest report required.

D. Incorrect-A 60 day LER may be determined necessary per either 10 CFR 50.73(a)(2)(i)(B) or 50.73(a)(2)(vii)(B).

However, this is not the soonest report required.

Page 50 of 55 Rev.6 2014 NRC SRO EXAM MASTER KEY Topic: Tier/Group:

KJA Info: SRO Importance:

Proposed references to be provided to applicant:

Learning Objective:

10 CFR Part 55 Content: Question source: Cognitive level: Last NRC Exam used on: Technical references:

Comments:

Question 98 (Q50852) AFW System Impairment reporting requirements Generic Knowledge

& Abilities 2.4-Emergency Procedures I Plan

• 2.4.30 -Knowledge of events related to system operation/status that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator.

3.8 CNG-NL-1.01-1004, Regulatory Reporting Att. 2 Tech Spec 3.7.3, Auxiliary Feedwater (AFW) System Tech Spec Basis 3.7.3 AFW System Given CNG-NL-1.01-1 004, correctly analyze plant conditions to determine if the conditions are reportable and type of report required.

55.43(b)(l)

IZI Bank jo Modified jDNew 0 Memory/Fundamental J1Z1 Comprehension/

Analysis No previous NRC Exam use CN G-NL-1. 01-1 004, Regulatory Reporting Tech Spec 3.7.3, Auxiliary Feedwater (AFW) System None Page 51 of 55 Rev.6 2014 NRC SRO EXAM MASTER KEY 199. 2.3 -Radiation Control U-1 is in Mode 5 with the following conditions:

• RCS pressure 150 PSIA

• PZR level slowly lowering

• Component Cooling RMS in alarm

• Component Cooling Head Tank level slowly rising Which ONE of the following choices identifies the correct AOP to enter and the EARLIER action to take in that procedure?

A. AOP-2A Excessive Reactor Coolant Leakage; Initiate HPSI flow with 11 & 13 HPSI pumps B. AOP-2A Excessive Reactor Coolant Leakage; Verify all available charging pumps operating C. AOP-3B Abnormal Shutdown Cooling Conditions; Start Charging pumps as necessary to maintain RCS level D. AOP-3B Abnormal Shutdown Cooling Conditions; Shut Component Cooling containment isolation valves (1-CC-3832

& 1-CC-3833)

Answer: C Answer Explanation A. Incorrect-AOP-2A is incorrect procedure but there is a section for RCS leakage (shutdown) but only for Mode 3, not Mode 5 and initiating HPSI flow is in this section B. Incorrect

-AOP-2A is incorrect procedure but there is a section for RCS leakage shutdown but only for Mode 3, not Mode 5 and verifying all charging pumps running is in this section C. Correct-AOP-3B is the correct procedure and starting charging pumps is step V.E.l D. Incorrect-AOP-3B is the correct procedure but isolating containment is step V.F.l 0.1 Page 52 of 55 Rev.6 2014 NRC SRO EXAM MASTER KEY Topic: Tier/Group:

KIA Info: SRO Importance:

Proposed references to be provided to applicant:

Learning Objective:

10 CFR Part 55 Content: Question source: Cognitive level: Last NRC Exam used on: Technical references:

Comments:

Question 99 (Q97193) CCRMS Generic Knowledge

& Abilities 2.3 -Radiation Control

• 2.3.15-Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc. 3.1 None Using AOP-3B, respond to a loss ofRCS Inventory while Shutdown Cooling is in use. 55.43(b)(4) 0Bank 0 Modified 0 Memory/Fundamental Comprehension/

Analysis Nl A new question AOP-2A Excessive Reactor Coolant Leakage AOP-3B Abnormal Shutdown Cooling Conditions None Page 53 of 55 Rev.6 2014 NRC SRO EXAM MASTER KEY 100. Use of the "Follow-up Communications" form Given the following conditions:

• 0900 -A Seismic Event, resulting in a Loss of Offsite Power occurs

• 0912 -The Shift Manager declares an Alert, due to the seismic event

• 0925-An aftershock occurs resulting in 14 4KV Bus deenergizing

• Investigation, by the Electricians, indicates 14 4KV Bus is grounded.

Electricians believe it will take several hours to clear the ground so the Bus can be re-energized.

Which ONE of the following is the appropriate form to complete for notifying offsite agencies of events occurring since the Alert declaration and why? A. EP-Form-ALL23RO, "Initial Notification Form", since this is a significant plant condition change resulting in Alert criteria being met for a different EAL. B. EP-Form-ALL23RO, "Initial Notification Form", since conditions warrant an upgrade to an EAL at a higher classification.

C. EP-Form-ALL22RO, "Follow-up Communications Form", since this is a significant plant condition change resulting in Alert criteria being met for a different EAL D. EP-Form-ALL22RO, "Follow-up Communications Form", since conditions warrant an upgrade to an EAL at a higher classification.

Answer: C Answer Explanation:

A. Incorrect-Although this is a significant plant change, EAL SA1.1 is NOT an upgrade in classification from HA1.1. Initial Notification is NOT used to inform offsite agencies.

B. Incorrect-EAL SAl. I is NOT an upgrade in classification from HALl and Initial Notification is NOT used to inform offsite agencies.

C. Correct-EAL SAl. I is NOT an upgrade in classification from HALl. Follow Up Communication is used to inform offsite agencies of this change in plant status. D. Incorrect-This is a significant plant change, EAL SA1.1 is NOT an upgrade in classification from HAl.l. If an upgrade in EAL occurred, Initial Notification is used to inform offsite agencies.

Page 54 of 55 Rev.6 2014 NRC SRO EXAM MASTER KEY Topic: Tier/Group:

KIA Info: SRO Importance:

Proposed references to be provided to applicant:

Learning Objective:

10 CFR Part 55 Content: Question source: Cognitive level: Last NRC Exam used on: Technical references:

Comments:

Question 100 (Q97013) Use of the "Follow-up Communications" form Generic Knowledge

& Abilities 2.4 -Emergency Procedures I Plan

• 2.4.40 -Knowledge of SRO responsibilities in emergency plan implementation 4.5 EAL Hot Chart 55.43(b)(5)

IZ! Bank 0 Modified 0New D Memory/Fundamental No previous NRC Exam use ERPIP 3.0 EAL Mode 1-4 Chart None IZ! Comprehension/

Analysis ' Page 55 of 55 Rev.6